ML20137K902

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Proposed Tech Specs Revising Definitions for ECCS Response Time,Isolation Sys Response Time, & Reactor Protection Response Time
ML20137K902
Person / Time
Site: Brunswick  Duke Energy icon.png
Issue date: 03/24/1997
From:
CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20137K882 List:
References
NUDOCS 9704070120
Download: ML20137K902 (35)
v • d • e


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ENCLOSURE 5 t

' BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 .!

NRC DOCKET NOS. 50-325 AND 50-324 OPERATING LICENSE NOS. DPR-71 AND DPR-62 f REQUEST FOR EMERGENCY / EXIGENT LICENSE AMENDMENTS  :

INSTRUMENTATION RESPONSE TIME TESTING t l

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i MARKED-UP TECHNICAL SPECIFICATION PAGES - UNIT 1 l i

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'3 9704070120 970324 PDR ADOCK 05000324 P PDR ,

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DEFINITIONS EMERCENCY CORE COOLING SYSTEM (ECCS) RESPONSE SME The EMERCENCY CORE C00LINC SYSTEM (ECCS) RESPt,NSE TIME shall be that time interval f rom when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

I Adct .Ibse:rt A FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

CASEOUS RADWASTE TREATMENT SYSTEM I

A CASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous ef fluents by collecting primary coolant system of fgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKACE

. IDENTIFIED LEAKACE shall be:

a. Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or
b. Leakage into the containment atmosphere from' sources that are both specifically located and known either not to interfere with the l operation of the leakage detection ' ystems s or not be PRESSURE I BOUNDARY LEAKAGE.

l ISOLATION SYSTEM RESPONSE TIME The ISOLATION SYSTEM RESPONSE TIMr. sna t t ce that time interval t rem when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence loading delays where applicable. fl-]d lCnsed B LIMITINC CONTROL ROD PATTERN A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the core  !

being on a limiting value for APLHCR or MCPR.

BRUNSWICK - UNIT 1 1-3 Amendment No. 1/* 7

l insert A In lieu of the methodology defined above, the EMERGENCY CORE COOLING RESPONSE TIME may also be determined using an alternate methodology that has been l I

reviewed and approved by the NRC.

Insert 11 I

In lieu of the methodology defined above, the ISOLATION SYSTEM IESPONSE TIME . l may also be determined using an attemate methodology that has been reviewed and i l

approved by the NRC.

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a DEFINITIONS PROCESS CONTROL

The PROCESS CONTROL PROGRAM (PCP) shall contain the current formula, sampling,  ;

analyses, tests and determinations to be made to ensure that the processing and pack. aging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to l l

assure compliance with 10 CFR Part 20,10 CFR Part 71, and Federal and State regulations and other. requirements governing the disposal of the radioactive

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waste.

PURCE - PURCING .

PURCE or PURCING is the controlled process of discharging. air or gas from a i confinement to maintain temperature, pressura, humidity, concentration or other operating condition, in such a manner that replacement air or ' gas is i I

required to purify the containment.

RATED THERMAL POWER

' RATED THERMAL POWER shall be total reactor core heat transfer rate to the reactor coolant of 2436 MVt.

REACTbRPROTECTIONSYSTEMRESPONSETIME .

REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the serai pilot valve solenoids. /)dd f~n. sere d-REFERENCE LEVEL ZERO The. REFERENCE LEVEL ZERO point is arbitrarily set at 367 inches above the vessel 2,ero point. This REFERENCE LEVEL ZERO is approximately mid point on

  • the top fuel guide and is the single reference for all specifications of vessel water level.

REPORTABLE EVENT A REPORTABLE EVENT shall be any of those' conditions specified'in Section 50.73 to 10 CFR Part 50.

ROD DENSITY ROD DENSITY shall be~ the number of control red " notches inserted as a fraction of the total number of notches. All rods fully inserted is equivalen1 to 100Z ROD DENSITY.

BRUNSVICK - UNIT 1 1-6 Amendment No. 131 f n

Insert C In lieu of the methodology defined above, the REACTOR PROTECTION SYSTEM RESPONSE TIME may also be determined using an alternate methodology that has been reviewed and approved by the NRC.  ;

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i e 3}4.3 INSTRUMENTATION t

BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:

a. Preserve the integrity of the fuel cla'dding.
b. Preserve the integrity of the reactor coolant system.
c. Minimize the energy which must be adsorbed following a loss-of-coolant accident, and prevent inadvertent criticality.

This specification )rovides the limiting conditions for operation necessary to preserve tie ability of the system to perform its intended 1 function even during periods when instrument channels may be out of service because of maintenance. When necessary, one channel may be made inoperable for brief intervals to conduct the required surveillance tests.

Specified surveillance intervals and allowed out-of-service times were established based on the reliability analyses documented in GE reports NEDC-30851P-A. " Technical Specification Improvement Analyses for BWR Reactor Protection System." March 1988 and MDE-81-0485. Rev. 1. " Technical Specification Im)rovement Analysis for the Reactor Protection System for Brunswick Steam Electric Plant. Units 1 and 2." August 1994, as modified by BWROG-92102. Letter from C. L. Tully (BWROG) to B. K. Grimes (NRC). "BWR Owners' Group (BWROG) Topical Reports on Technical Specification Improvement Analysis for BWR Reactor Protection Systems - Use for Relay and Solid State Plants (NEDC-30844 and NEDC-30851P)." November 4. 1992.

The reactor protection system is made up of two independent trip systems.

There are usually four channels to monitor each parameter, with two in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram. The system meets the intent of IEEE-279 for nuclear power plant protection systems.

The measurement of response time at the specified frequencies provides assurance that the ]rotective, isolation, and emergency core cooling functions associated with eac1 channel are completed within the time limit assumed in the accident analysis. No credit was taken for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential. overlapping

' or total channel test measurements provided such tests demonstrate the total channel response time as defined. Sensor response time verification may-be demonstrated by either 1) inplace, onsite, or offsite test measurements, or 2) utilizing replacement sensors with certified response times.

1 The bases for the trip settings of the reactor protection system are discussed in the bases for Specification 2.2.

A Add % cd Q BRUNSWICK - UNIT 1- B 3/4 3-1 Amendment No. 175 1

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t insert D On December 29,1993, the NRC staffissued Generic Letter 93-08, " Relocation of Technical Specification Tables ofInstrument Response Time Limits." Relocation to the Updated Final Safety Analysis Report of the instrument response time table associated  ;

with Technical Specification 3/4.3.1 was approved by Amendment No.171 to the Facility Operating License. Updated Final Safety Analysis Report Table 7.2.1-3 provides the Reactor Protection System instrumentation response times. The NRC staff has reviewed and accepted the alternate methodologies of BWR Owners' Group Licensing Topical Report NEDO-32291-A, " System Analyses For the Elimination of Selected Response ,

Time Testing Requirements." As a result, the alternate methodologies provided in NEDO-32291-A have been incorporated into the instrument response table which has  ;

been relocated to the Updated Final Safety Analysis Report and are an acceptable means of demonstrating instrument response times.

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INSTRUMENTATION -

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3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION' I

This specification ensures the effectiveness of the instrumentation used to mitigate the consaquences of accidents by prescribing the' trip settings for  !

isolation of the reactor systems. When necessary, one channel may be 1 inoperable for brief intervals to conduct required surveillance. Some of the  ;

trip settings have tolerances explicitly stated where both the high and low  !

values are 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 the safety, are established at a level away .i from t1e normal operating range to prevent inadvertent actuation of the systems involved. j ;

- Specified surveillance intervals and allowed out-of-service times were  !

established based on the reliability analyses documented in GE reports l NEDC-30851P-A. Supplement 2. " Technical Specification Improvement Analysis for i

-BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation." March  :

1989 and NEDC-31677P-A. " Technical Specification Improvement Analysis for BWR ~

isolation Actuation Instrumentation." July 1990, as modified by OG90-579-32A.

Letter to'Hillard L. Wohl (NRC) from W. P. Sullivan and J. F. Klapproth (GE).

" Implementation Enhancements to Technical Specification Changes Given in Isolation Actuation Instrumentation Analysis." June 25. 1990 and supplemented by GE letter report GENE-A31-00001-02. " Assessment of Brunswick Nuclear Plant  :

Isolation Actuation Instrumentation Against NEDC-31677P-A Bounding Analyses."  !

[idd gugust1994. j 5 p T 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to I initiate actions to mitigate the consequences of accidents that are beyond the  !

operator's ability to control. This specification provides the tria point

- settings that will ensure effectiveness of the systems to provide t1e design protection. Although the instruments are listed by system. in some cases the I same instrument is used to send the start signal to several systems at the same time. The out-of-service times for the instruments are consistent with i the requirements of the specifications in Section 3/4.5.

Specified surveillance intervals and allowed out-of-service times were established based on the reliability analyses documented in GE reports NEDC-30936P-A. Parts 1 and 2. "BWR Owners' Group Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation)." December 1988 and RE-011. Rev.1. " Technical Specification

, Improvement Analysis for the Emergency Core Cooling System Actuation Instrumentation for Brunswick Steam Electric Plant. Units 1 & 2." August-1994, as modified by 0G90-319-320. letter from W. P. Sullivan and J. F. Klapproth (GE) to Millard L. Wohl (NRC). " Clarification of Technical Specification Changes Given in ECCS Actuation Instrumentation Analysis." March 22. 1990.

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BRUNSWICK - UNIT 1 B 3/4 3-2 Amendment No. 175 1

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Insert E On December 29,1993, the NRC staffissued Generic Letter 93-08, " Relocation of f

Tecimical Specification Tables ofInstmment Response Time Limits." Relocation to the Updated Final Safety Analysis Report of the instrument response time table associated with Technical Specification 3/4.3.2 was approved by Amendment No.171 to the Facility Operating License. Updated Final : ifety Analysis Report Table 7.3.1-3A provides the Isolation System instrumentation ;ponse times. The NRC staff has reviewed and accepted the alternate methodologies of BWR Owners' Group Licensing Topical Report NEDO-32291-A, " System Analyses For the Elimination of Selected Response Time Testing Requirements." As a result, the alternate methodologies provided in NEDO-32291-A have been incorporated into the instrument response table which has been relocated to the Updated Final Safety Analysis Report and are an acceptable means of demonstrating instrument response times. .

Insert F j On December 29,1993, the NRC staffissued Generic Letter 93-08, " Relocation of Technical Specification Tables ofInstrument Response Time Limits." Relocation to the Updated Final Safety Analysis Report of the instrument response time table associated with Technical Specification 3/4.3.3 was approved by Amendment No.171 to the Facility Operating License. Updated Final Safety Analysis Report Table 7.3.3-5,provides the Emergency Core cooling System instrumentation response times. The NRC staff has reviewed and accepted the alternate methodologies of BWR Owners' Group Licensing Topical Report NEDO-32291-A, " System Analyses For the Elimination of Selected Response Time Testing Requirements." As a result, the alternatemethodologies provided in NEDO-32291-A have been incorporated into the instrument response table which has been relocated to the Updated Final Safety Analysis Report and are an acceptable means of demonstrating instrument response times.

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I i i ENCLOSURE 6

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BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 (

NRC DOCKET NOS. 50-325 AND 50-324

,- OPERATING LICENSE NOS. DPR-71 AND DPR-62  ;

i REQUEST FOR EMERGENCY / EXIGENT LICENSE AMENDMENTS i INSTRUMENTATION RESPONSE TIME TESTING

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i DEFINITIONS EMERCENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME The EMERCENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety , function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel g[enerator n # .Y W jtarting and sequence loading delays where applicable.

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END-61-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval to recirculation pump breaker trip from initial movement of the associated: P

a. Turbine stop valves, and
b. Turbine control valves.

I FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

. CASEOUS RADWASTE TREATMENT SYSTEM .

AGkSEOUSRADWASTETREATMENTSYSTEMisanysystemdesignedandinstalledto reduce radioactive gaseous effluents by collecting primary coolant system off gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKACE IDENTIFIED LEAKACE shall be:

a. Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or
b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the '

operation of the leakage detection systems or not be PRESSURE BOUNDARY

- LEAKACE.

-n BRUNSWICK - UNIT 2 1-3 Amendment No.168

Insert A In lieu of the methodology defined above, the EMERGENCY CORE COOLING RESPONSE TIME may also be determined using an alternate methodology that has been reviewed and approved by the NRC.

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. i DEFINITIONS - ,

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ISOLATION SYSTEM RESPONSE TIME The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence loading delays where i applicable. McM, h erd 8 LIMITINC CONTROL ROD PATTERN A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the core '.

being on a limiting value for APLHCR or MCPR. (

l LOGIC SYSTEM FUNCTIONAL TEST A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all relays and contacts of a logic circuit, from sensor output to activated device, to ensure that components are OPERABLE.

MAXIMUM AVERACE PLANAR LINEAR HEAT CENERATION RATE RATIO The MAXIMUM AVERACE PLANAR LINEAR HEAT GENERATION RATE RATIO (MAPRAT) for a bundle shall be the largest value in the bundle of the ratio of the APLHGR at a specific height in the bundle divided by the exposure dependent APLHCR limit for that specific height.

MEMBER (S) OF THE PUBLIC MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the t utility, its contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This ,

category does not include persons who use portions of the site for recreational, I occupational or other purposes not associated with the plant.

MINIMUM CRITICAL POWER RATIO The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core.

ODYN OPTION A

, ODYN OPTION A shall be analyses which refer to minimum critical power ratio limits which are determined using a transient analysis plus an analysis uncertainty penalty. .

ODYN OPTION B ODYN OPTION B shall be analyses which refer to minimum critical power ratio limits determined using a transient analysis which includes a requirement for 20% scram insertion times to reduce the analysis uncertainty penalty.

BRUNSWICK - UNIT 2 1-4 Amendment No. 161 i

Insert B In lieu of the methodology defined above, the ISOLATION SYSTEM RESPONSE TIME may also be determined using an attemate methodology that has been reviewed and approved by the NRC.

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, DEFINITIONS

.l PRIMARY CONTAINMENT INTEGRITY (Continued)

b. All equipment hatches are closed and sealed.
c. Each containment air lock is OPERABLE pursuant to Specification 3.6.1.3.
d. The containment leakage rates are within the limits of Specification 3.6.1.2.
e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

PROCESS CONTROL PROGRAM (PCP)

The PROCESS CONTROL PROGRAM (PCP) shall contain the current for:ula, sa=pling, analyses, tests and determinations to be made to ensure that the processing and packaging of solid radioactive vastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Part 20, 10 CFR Part 71, and Federal and State regulations and other requirements governing the disposal of the radioactive waste.

. PURGE - PURGING

.J PURGE OR PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or o ther operating condition, in such a manner that replacement air or gas is required to purify the containment.

RATED THERMAL POWER i I

2ATED THERMAL POWER shall be a total reactor care heat transfer rate to the reactor coolant of 2436 MWt. 1 REACTOR PROTECTION SYSTEM RESPONSE TIME REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the ti=e interval f rom when I the =cnitored para =eter exceeds its trip setpoint at the channel sensor until i

.de-energization of the scram pilot valve solenoids. $dd l~rtsedd_

1 REFERENCE LEVEL ZERO I

, The REFERENCE LEVEL ZERO point is arbitrarily set at 367 inches above the vessel zero point. This REFERENCE LEVEL ZERO is approx 1=ately sid point on the top fuel guide and is the single reference for all specifications of vessel water level.

l BRUNSWICK - UNIT 2 1-6 Amendment No. 88

Insert C In lieu of the methodology defined above, the REACTOR PROTECTION SYSTEM RESPONSE TIME may also be determined using an alternate methodology that has been ,

reviewed and approved by the NRC.

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3/4.3 TNSTRUMENTATTON BASES

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3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:

a. Preserve the integrity of the fuel cladding.
b. Preserve the integrity of the reactor coolant system.
c. Minimize the energy which must be adsorbed folicavi,ig a loss-of-coolant accident, and prevent inadvertent criticality.

This specification arovides the limiting conditions for operation necessary to preserve t1e ability of the system to perform its intended function even during periods when instrument channels may be out of service because of maintenance. When necessary, one channel may be made inoperable for brief intervals to conduct the required surveillance tests.

Specified surveillance intervals and allowed out-of-service times were established based on the reliability analyses documented in GE reports NEDC-30851P-A. " Technical Specification Improvement Analyses for BWR Reactor ,

Protection System." March 1988 and MDE-81-0485. Rev. 1. " Technical Specification Improvement Analysis for the Reactor Protection System for Brunswick Steam Electric Plant. Units 1 and 2." August 1994, as modified by BWROG-92102. Letter from C. L. Tully (BWROG) to B. K. Grimes (NRC). "BWR Owners' Group (BWROG) Topical Reports on Technical Specification Improvement Analysis for BWR Reactor Protection Systems - Use for Relay and Solid State Plants (NEDC-30844 and NEDC-30851P)." November 4. 1992.

The reactor protection system is made up of two independent trip systems.

There are usually four channels to monitor each parameter with two in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram. The system meets the intent of IEEE-279 for nuclear power plant protection systems.

The measurement of response time at the specified frequencies provides assurance that the 3rotective, isolation and emergency core cooling functions associated with eac1 channel are completed within the time limit assumed in the accident analysis. No credit was taken for those channels with response )

l times indicated as not applicable.

Resporse time may be demonstrated by any series of sequential, overlapping l or total channel test measurements, provided such tests demonstrate the total  !

channel response time as defined. Sensor response time verification may be demonstrated by either 1) inplace, onsite, or offsite test measurements,-or 2) {

i utilizing replacement sensors with certified response times.

The bases for the trip settings of the reactor protection system are discussed in the bases for Specification 2.2.

L-Odd Lusert D BRUNSWICK - UNIT 2 8 3/4 3-1 Amendment No. 206

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Insert D On December 29,1993, the NRC staffissued Generic Letter 93-08, " Relocation of Tecimical Specification Tables ofInstrument Response Time Limits." Relocation to the Updated Final Safety Analysis Report of the instntment response time table associated with Technical Specification 3/4.3.1 was approved by Amendment No. 202 to the Facility Operating License. Updated Final Safety Analysis Report Table 7.2.1-3 provides the Reactor Protection System instrumentation response times. The NRC staff has reviewed and accepted the alternate methodologies of BWR Owners' Group Licensing Topical Report NEDO-32291-A, " System Analyses For the Elimination of Selected Response Time Testing Requirements." As a result, the alternate methodologies provided in NEDO-32291-A have been incorporated into the instrument response table which has been relocated to the Updated Final Safety Analysis Report and are an acceptable means of demonstrating instrument response times.

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INSTRUMENTATTON BASES 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the trip settings for isolation of the reactor systems. When necessary, one channel ma inoperable for brief intervals to conduct required surveillance. y Some be of the trip settings have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety. The setpoints of other instrumentation where only the high or low end of the setting from thehas a direct normal o bearing on the safety, inadvertent actuation of theare established at a le systems involved.perating range to prevent Specified surveillance intervals and allowed out-of-service times were established based on reliability analyses documented in GE reports NEDC-30851P-A Supplement 2. " Technical S)ecification Improvement Analysis for BWR Isolation Instrumentation Commcq to R)S and ECCS Instrumentation." March 1989 and NEDC-31677P-A. " Technical Sp'ecification Improvement Analysis for BWR Isolation Actuation Instrumentation. July 1990. as modified by OG90-579-32A.

Letter to Millard L. Wohl (NRC) from W. P. Sullivan and J. F. Klapproth (GE).

" Implementation Enhancements to Technical Specification Changes Given in Isolation Actuation Instrumentation Analysis." June 25, 1990 and supplemented by GE letter report GENE-A31-00001-02. " Assessment of Brunswick Nuclear Plant Isolation Actuation Instrumentation Against NEDC-31677P-A Bounding Analyses."

jg[ August 1994.

[** N 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the operator's ability to control. This specification provides the tri) point settings that will ensure effectiveness of the systems to provid? tie design  :

protection. Although the instruments are listed by system, in some cases the l same instrument is used to send the start signal to several systems at the I same time. The out-of-service times for the instruments are consistent with the requirements of the specifications in Section 3/4.5.

Specified surveillance intervals and allowed out-of-service times were established based on the reliability analyses documented in GE reports NEDC-30936P-A. Parts 1 and 2. "BWR Owners' Group Technical Specification .

Improvement Methodology (With Demonstration for BWR ECCS Actuation l Instrumentation)." December 1988 and RE-011. Rev.1. " Technical Specification l Improvement Analysis for the Emergency Core Cooling System Actuation Instrumentation for Brunswick Steam Electric Plant. Units 1 & 2." August 1994, as modified by 0G90-319-32D. letter from W. P. Sullivan and J. F. Klapproth (GE) to Millard L. Wohl (NRC). " Clarification of Technical Specification Changes Given in ECCS Actuation Instrumentation Analysis." March 22, 199,0.

&c{[hser{f BRUNSWICK - UNIT 2 B 3/4 3-2 Amendment No. 206 I

Insert E On December 29,1993, the NRC staffissued Generic Letter 93-08, " Relocation of )

Technical Specification Tables ofInstrument Response Time Limits." Relocation to the l I

Updated Final Safety Analysis Report of the instrument response time table associated with Technical Specificaticn 3/4.3.2 was approved by Amendment No. 202 to the Facility j Operating License. Updated Final Safety Analysis Report Table 7.3.1-3 A provides the  :

Isolation System instrumentation response times. The NRC staff has reviewed and  !

3 accepted the alternate methodologies of BWR Owners' Group Licensing Topical Report NEDO-32291-A, " System Analyses For the Elimination of Selected Response Time Testing Requirements." As a result, the alternate methodologies provided in NEDO-32291-A have been incorporated into the instrument response table which has been relocated to the Updated Final Safety Analysis Report and are an acceptable means of demonstrating instrument response times.

Insert F On December 29,1993, the NRC staffissued Generic Letter 93-08, " Relocation of Technical Specification Tables ofInstrument Response Time Limits." Relocation to the Updated Final Safety Analysis Report of the instrument response time table associated with Technical Specification 3/4.3.3 was approved by Amendment No. 202 to the Facility Operating License. Updated Final Safety Analysis Report Table 7.3.3-5 provides the Emergency Core cooling System instrumentation response times. The NRC staff has reviewed and accepted the attemate methodologies of BWR Owners' Group Licensing Topical Report NEDO-32291-A, " System Analyses For the Elimination of Selected Response Time Testing Requirements." As a result, the alternate methodologies provided in NEDO-32291-A have been incorporated into the instrument response table which has been relocated to the Updated Final Safety Analysis Report and are an acceptable means of demonstrating instrument response times.

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ENCLOSURE 7 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 -

! NRC DOCKET NOS. 50-325 AND 50-324 OPERATING LICENSE NOS. DPR-71 AND DPR-62 l l REQUEST FOR EMERGENCY / EXIGENT LICENSE AMENDMENTS l INSTRUMENTATION RESPONSE TIME TESTING I i

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> TYPED TECHNICAL SPECIFICATION PAGES - UNIT 1 t.

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. DEFINITIONS EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME

-The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump  ;

discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable. In lieu of the methodology defined above, the EMERGENCY CORE COOLING RESPONSE TIME may also be determined using an alternate methodology that has been reviewed and approved by the NRC.

FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1. j GASEOUS RADWASTE TREATMENT SYSTEM ,

A GASE0US RADWASTE TREATMENT SYSTEM is any system designed and installed to -

. reduce radioactive gaseous effluents by collecting primary coolant system off-gases from the primary system and providing for delay or holdup for the l purpose of reducing the total radioactivity prior to release to the environment.

JDENTIFIED LEAKAGE i IDENTIFIED LEAKAGE shall be:  :

a. Leakage into collection systems, such as pump seal or valve  !

packing leaks. that is captured and conducted to a sump or collecting tank, or  ;

b. Leakage into the containment atmos 3here from sources that are both specifically located and known eitler not to interfere with the operation of the leakage detection systems or not be PRESSURE BOUNDARY LEAKAGE. =

ISOLATION SYSTEM RESPONSF TIME The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and secuence loading delays where applicable. In lieu of the methodology definec above, the ISOLATION SYSTEM RESPONSE TIME may also be determined using an alternate methodology that has been reviewed and approved by the NRC.

LIMITING CONTROL ROD PATTERN A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the core '

being on a limiting value for APLHGR or MCPR.

i BRUNSWICK - UNIT 1 1-3 Amendment No.

l

..ta.. .4

~

iDEFINIT10NS .

]/

'!~

PROCESS CONTROL G0 GRAM (PCP) i

'The'PROCESSCONTROLPROGRAM(PCP)'shallicontainthecurrent. formula,~ sampling.. 1 analyses tests and determinations to be.made to, ensure that the processing  :

and packaging of solid radioactive wastes based on demonstrated 3rocessing of 1 actual or simulated wet solid wastes will be accomplished in suc1 a way as^to .;

assure compliance with 10 CFR Part 20.-10 CFR.Part 71, and Federal and State regulations and other requirements governing!the disposal of the radioactive waste. j PURGE -' PURGING .

PURGE OR PURGING is the controlled process of discharging air or gas!from a  !

-confinement to maintain temperatt.re, pressure,. humidity. concentration or-other operating condition. in such a manner that replacement air ~or gas is -i required to purify the containment. -

RATED THERMAL POWER  !

RATED THERMAL POWER shall be total reactor core heat transfer rate to the reactor coolant of 2436 MWt. '

REACTOR PROTECTION SYSTEM RESPONSE TIME  !

REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until  !

de-energization of the scram pilot valve solenoids. In lieu of the  !

methodology defined above the REACTOR PROTECTION SYSTEM RESPONSE TIME may i also be determined using an alternate methodology that has been reviewed and

. approved by the NRC.

REFERENCE LEVEL ZERO f

i The REFERENCE LEVEL ZERO aoint is arbitrarily set at 367 inches above the  !

.~ vessel zero point. This REFERENCE LEVEL ZERO is approximately-mid-point on ,

the top fuel guide and is the single reference for all specifications of i vessel water level.

REPORTABLE EVENT  !

A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73  !'

to 10 CFR Part 50.

R00 DENSITY-ROD. DENSITY shall be the number of control rod notches inserted as a fraction

.of the total number of notches. All rods fully inserted 1.s equivalent to 100% ROD-DENSITY, ,

. )

BRUNSWICK a UNIT 1 1-6 Amendment No.

.- '3/4.3 INSTRUMENTATION [

BASES  ;

4 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:  !

a. Preserve the integrity of the fuel cladding. l t
b. Preserve the integrity of the reactor coolant system. -
c. Minimize the energy which must be adsorbed following a loss-of- '

coolant accident. and prevent inadvertent criticality.  ;

This specification 3rovides the limiting conditions for operation i necessary to preserve t1e ability of the system to perform its intended function even during periods when instrument channels may be out of service because of maintenance. When necessary one channel may be made inoperable for brief intervals to conduct the required surveillance tests.  ;

Specified survoillance intervals and allowed out-of-service times were established based on the reliability analyses documented in GE reports '

NEDC-30851P-A, " Technical Specification Improvement Analyses for BWR Reactor Protection System." March 1988 and MDE-81-0485. Rev. 1. " Technical  :

Specification Improvement Analysis for the Reactor Protection System for  ;

Brunswick Steam Electric Plant. Units 1 and 2." August 1994 as modified by i BWROG-92102. Letter from C. L. Tully (BWROG) to B. K. Grimes (NRC), "BWR Owners' Group (BWROG) Topical Reports on Technical Specification Improvement Analysis for BWR Reactor Protection Systems - Use for Relay and Solid State Plants (NEDC-30844 and NEDC-30851P)," November 4. 1992.

4 The reactor protection system is made up of two independent trip systems.

There are usually four channels to monitor each parameter, with two in each l trip system. The outputs of the channels in a trip system are combined in a  !

, logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram. The system meets the intent of i IEEE-279 for nuclear power plant protection systems.

The measurement of response time at the specified frequencies provides assurance that the 3rotective, isolation, and emergency core cooling functions associated with eac1 channel are completed within the time limit assumed in the accident analysis. Nc 'redit was taken for those channels with response times indicated as not ar cable.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements, provided such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) inplace, onsite, or offsite test measurements. or 2) utilizing replacement sensors with certified response times.

' BRUNSWICK - UNIT 1 8 3/4 3-1 Amendment No.

3/4.3 INSTRUMENTATION BASES '

3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION On December 29, 1993. the NRC staff issued Generic Letter 93-08.

" Relocation of Technical Specification Tables of Instrument Response Time ,

Limits." Relocation to the Updated Final Safety Analysis Report of the instrument response time table associated with Technical Specification 3/4.3.1 was approved by Amendment No. 171 to the Facility Operating License. Updated Final Safety Analysis Report Table 7.2.1-3 provides the Reactor Protection System instrumentation res)onse times. The NRC staff has reviewed and ,

accepted the alternate metlodologies of BWR Owners' Group Licensing Topical Report NED0-32291-A. " System Analyses For the Elimination of Selected Response ,

Time Testing Requirements." As a result, the alternate methodologies )rovided in NED0-32291-A have been incorporated into the instrument response ta)le 1 which has been relocated to the Updated Final Safety Analysis Report and are an acceptable means of demonstrating instrument response times. ,

Tha bases for the trip settings of the reactor protection system are disc: 1 in the bases for Specification 2.2.

l l

I i

i BRUNSWICK - UNIT 1 8 3/4 3-la Amendment No. I

i

" INSTRUMENTATION i BASES j 3/4 3.2 ISOLATION ACTUATION INSTRUMENTATION  ;

This specification ensures the effectiveness of the instrumentation used l to mitigate the consequences of accidents by prescribing the trip settings for  :

isolation of the reactor systems. When necessary, one channel may be j inoperable for brief intervals to conduct- required surveillance. Some of the i trip settings have tolerances explicitly stated where both the high and low  !

values are critical and may have a substantial effect on safety. The i

setpoints of other instrumentation where only the high or low end of the l setting has a direct bearing on the safety. are established at a level away i from the normal operating range to prevent inadvertent actuation of the i systems involved. l t

Specified surveillance intervals and allowed out-of-service times were r established based on the reliability analyses documented in GE reports  !

I NEDC-30851P-A. Supplement 2. " Technical S)ecification Improvement Analysis for BWR' Isolation Instrumentation Common to R)S and ECCS Instrumentation." March i 1989 and NEDC-31677P-A. '" Technical Specification Improvement Analysis for BWR f Isolation Actuation Instrumentation," July 1990, as modified by OG90-579-32A.

Letter to Millard L. Wohl (NRC) from W. P. Sullivan and J. F. Klapproth (GE).

" Implementation Enhancements to Technical Speciffcation Changes Given in .

Isolation Actuation Instrumentation Analysis." June 25. 1990 and supplemented '

by GE letter report GENE-A31-00001-02. " Assessment of Brunswick Nuclear Plant Isolation Actuation Instrumentation Against NEDC-31677P-A Bounding Analyses."

August 1994.  !

On December 29. 1993, the NRC staff issued Generic Letter 93-08.  !

" Relocation of Technical Specification Tables of Instrument Response Time  !

Limits." Relocation to the Updated Final Safety Analysis Report of the -  !

instrument time table associated with Technical Specification 3/4.3.2 was l approved by Amendment No. 171 to the Facility Operating License. Updated  ;

Final Safety Analysis Report Table 7.3.1-3A provides the Isolation System j instrumentation response times. The NRC staff has reviewed and accepted the i alternate methodologies of BWR Owners' Group Licensing Topical Report NEDO- l 32291-A. " System Analyses For the Elimination of Selected Response Time l Testing Requirements." As a result. the alternate methodologies provided in j NEDO-32291-A have been incorporated into the instrument response table which  ;

has been relocated to the Updated Final Safety Analysis Report and are an .

acceptable means of demonstrating instrument response times. j i

I l

s BRUNSWICK - UNIT 1 B 3/4 3-2 Amendment No.

I

INSTRUMENTATION BASES 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The. emergency core cooling system actuation instrutentation is provided to '

initiate actions to mitigate the consequences of accidents that are beyond the operator's ability to control. . This specification provides the tri) point settings that will ensure effectiveness of-the-systems to provide tie design protection. Although the instruments are listed by system, in some cases the same instrument is used to send the start signal to several systems at the same time. The out-of-service times for the instruments are consistent with the requirements of the specifications in Section 3/4.5. .

' Specified surveillance intervals and allowed out-of-service times were -

established based on the reliability analyses documented in GE reports  !

NEDC-30936P-A. Parts 1 and 2. "BWR Owners Group Technical Specification l Improvement Methodology (With Demonstration for BWR ECCS Actuation q Instrumentation)." December 1988 and RE-011. Rev.1. " Technical Specification  ;

Improvement Analysis for the Emergency Core Cooling System Actuation  :

Instrumentation for Brunswick Steam Electric Plant. Units 1 & 2." August 1994.  ;

as modified by 0G90-319-320. letter from W. P. Sullivan and J. F. Klapproth  ;

(GE) to Millard L. Wohl (NRC). " Clarification of Technical Specification  :

-Changes Given in ECCS Actuation Instrumentation Analysis." March 22, 1990.  !

On December 29. 1993. the $ ;C staff issued Generic Letter 93-08.'  !

" Relocation of Technical Specification Tables of Instrument Response Time Limits." Relocation to the Updated Final Safety Analysis Report of the instrument res onse time table associated with Technical Specification 3/4.3.3 t was approved Amendment No. 171 to the Facility Operating License. U] dated l Final Safety alysis Report Table 7.3.3-5 provides the Emergency Core Cooling l t System instrumentation res)onse times. The NRC staff has reviewed and j i accepted the alternate metlodologies of BWR Owners' Group Licensing To]ical  :

. Report NED0-32291-A. " System Analyses For the Elimination of Selected Response i

! Time Testing Requirements." As a result the alternate methodologies )rovided j

!. in NE00-32291-A have been incorporated into the instrument res)onse ta ale 8

which has been relocated to the Updated Final Safety Analysis Report and are  ;

j an acceptable means of demonstrating instrument response times. j 3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION l e

The control rod block functions are provided consistent with the  !

I recuirements of the specifications in Section 3/4.1.4. Rod Program Controls.

anc Section 3/4.2. Power Distribution Limits. The trip logic is arranged so that a trip in any one of the inputs will result in a rod block.

S i estabkecified ished based surveillance intervals on the reliability analysesand allowedinout-of-service documented GE report times were

NEDC-30851P-A. Supplement 1. " Technical. Specification Improvement Analysis for
BWR Control Rod Block Instrumentation." October 1988.

. 3/4.3.5 MONITORING INSTRUMENTATION 3/4.3 5.1 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY of the seismic monitorin instrumentation ensures that 4

sufficient capability is available to prompt y determine the magnitude of a seismic event and evaluate the response of t ose features important to safety.

This capability is required to permit comparison of the measured response to that used in the design basis for the facility.

1 BRUNSWICK - UNIT 1 B 3/4 3-2a Amendment No.

"^ s -,n-O 9 9 ENCLOSURE 8 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 NRC DOCKET NOS. 50-325 AND 50-324 OPERATING LICENSE NOS. DPR-71 AND DPR-62 REQUEST FOR EMERGENCY / EXIGENT LICENSE AMENDMENTS INSTRUMENTATION RESPONSE TIME TESTING TYPED TECHNICAL SPECIFICATION PAGES - UNIT 2 l

1

o 4

DEFINITIONS EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME l

The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint  ;

at the channel sensor until the ECCS equipment is capable of performing its  !

safety function (i.e.. the valves travel to their required positions, pump discharge pressures reach their required values; etc.). Times shall include diesel generator starting and sequence loading delays where applicable. In lieu of the methodology defined above, the EMERGENCY CORE COOLING RESPONSE TIME may also be determined using an alternate methodology that has been reviewed and approved by the NRC.

END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval to recirculation pump breaker trip from initial movement of the associated:

a. Turbine stop valves, and l
b. Turbine control valves. ,

l FREQUENCY NOTAT10N The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

GASEOUS RADWASTE TREATMENT SYSTEM l A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system  :

off-gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.  ;

IDENTIFIED LEAKAGE IDENTIFIED LEAKAGE shall be:

a. Leakage into collection systems, such as pump seal or valve .

packing leaks, that is captured and conducted to a sump or  !

collecting tank. or

b. Leakage into the containment ~ atmos)here from sources that are both i specifically located and known eitler not to interfere with the operation of the leakage detection systems or not be PRESSURE BOUNDARY LEAKAGE. ,

r BRUNSWICK - UNIT 2 1-3 Amendment No.

' DEFINITIONS ISOLATION SYSTEM RESPONSE TIME

~ The ISOLATION SYSTEM RESPONSE- TIME shall be that time interval from when the .

monitored parameter exceeds its isolation actuation setpoint at the channel '

. sensor until"the isolation valves travel to their' required positions. Times  !

shall include diesel generator-starting and secuence loading delays where 1 applicable. -In lieu of the methodology definec above, the ISOLATION SYSTEM ' l RESPONSE TIME may also be determined using an alternate methodology that has i been reviewed and; approved by the NRC.

LIMITING CONTROL ROD PATTERN -

l 3attern which results in the core A LIMITING CONTROL ROD PATTERN shall be a being on a limiting value for APLHGR or MC)R.  ;

LOGIC SYSTEM FUNCTIONAL TEST

^ LOG.LC SYSTEM FUNCTIONAL' TEST shall be a test of all relays and contacts of a

-logic circuit, from sensor output to activated device.~ to ensure that  !

Tcomponents are OPERABLE. ,

' MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE RATIO. ]

The MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION. RATE RATIO (MAPRAT) for a . .

bundle shall be-the largest value in the bundle of the ratio of the APLHGR at  !

a s'pecific height in the bundle divided by the exposure dependent APLHGR limit j for that: specific height.

MEMBER (S) 0F THE PUBLIC MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally -l '

associated with the plant. This category does not include employees of the utility. its contractors or vendors. Also excluded from this category are persons who enter. the site to service equipment or to make deliveries. This category does not include persons who use portions of the site for recreational occupational or other purposes not associated with the plant. .

MINIMUM CRITICAL POWER RATIO  ;

The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core.

ODYN OPTION A ODYN OPTION A shall be analyses which refer to minimum critical power ratio limits which are determined using a transient analysis plus an analysis uncertainty penalty.

'0DYN OPTION B ODYN OPTION B shall be ' analyses which refer to minimum critical power ratio 1

. limits determined'using a transient analysis which includes a requirement for

. 20%_ scram insertion times- to reduce the analysis uncertainty penalty,

.BRUNSWICKL- UNIT 2 1-4 Amendment No. i

,w,~ . , . ~., .-_..v .. -

b. All equipment hatches are closed and sealed. .
c. Each containment air lock is OPERABLE pursuant to Specification 3.6.1.3. -
d. The containment leakage rates are within ti:9 limits of Specification 3.6.1.2.
e. The sealing mechanism associated with each penetration (e.g. .

welds, bellows, or 0-rings) is OPERABLE. 1 PROCESS CONTROL PROGRAM (PCP)

The PROCESS CONTROL PROGRAM (PCP) shall contain the current formula. sampling, analyses, tests and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated 3rocessing of -

actual or simulated wet solid wastes will be accomplished in suc1 a way as to assure compliance with 10 CFR Part 20, 10 CFR Part 71, and Federal and State regulations and other requirements governing the, disposal of the radioactive waste. f PURGE - PURGING PURGE OR PURGING is the controlled process of discharging air or gas from a e

! confinement to maintain temperature, pressure. humidity, concentration.or other operating condition. in such a manner that replacement air or gas is required to purify the containment.

RATED THERMAL POWER RATED THERMAL POWER shall be a total reactor core heat transfer rate to the l reactor coolant of 2436 MWt.  !

REACTOR PROTECTION SYSTEM RESPONSE TIME REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. In lieu of the i methodology defined above, the REACTOR PROTECTION SYSTEM RESPONSE TIME may also be determined using an alternate methodology that has been reviewed and approved by the NRC.

REFERENCE LEVEL ZERO The REFERENCE LEVEL ZERO point is arbitrarily set at 367 inches above the vessel zero point. This REFERENCE the top fuel guide and is the single (EVEL ZERO reference is approximately for all specifications of mid-point on vessel water level.

BRUNSWICK - UNIT 2 1-6 Amendment No. l l

l

c , ~ 3/4.3' INSTRUMENTATION BASES 3/4.3 1 REACTOR PROTECTION SYSTEM INSTRUMENTATION ,

The reactor protection system automatically initiates a reactor scram

a. Preserve the integrity of the fuel cladding.
b. Preserve the integrity of the reactor coolant system.
c. Minimize the energy which must be adsorbed following a. loss-of- .

coolant accident, and prevent inadvertent criticality.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of. maintenance. When necessary, one channel may be made inoperable

'for brief intervals to conduct the required surveillance tests.

Specified surveillance intervals and allowed out-of-service times were established based on the reliability analyses documented in GE reports NEDC-30851P-A, " Technical Specification Improvement Analyses for BWR Reactor

. Protection System." March 1988 and MDE-81-0485. Rev. 1. " Technical .

Specification Im]rovement Analysis for the Reactor Protection System for Brunswick Steam Electric Plant. Units 1 and 2." August 1994, as modified by BWROG 92102. Letter from C. L. Tully (BWROG) to B. K. Grimes (NRC). "BWR Owners' Group (BWROG) Topical Reports on Technical Specification Improvement Analysis for BWR Reactor Protection Systems - Use for Relay and Solid State Plants (NEDC-30844 and NEDC-30851P)." November 4. 1992.

The reactor protection system is made up of two independent trip systems. There are usually four channels to monitor each parameter with two in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram. The system meets the intent of IEEE-279 for nuclear power plant protection systems.

The measurement of response time at the specified frequencies provides assurance that the 3rotective, isolation, and emergency core cooling functions associated with eac1 channel are completed within the time limit assumed in the accident analysis. No credit was taken for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential.

overlapping or total channel test measurements, provided such tests demonstrate the total channel res)onse time as defined. Sensor response time verification may be demonstrated Jy either 1) inplace, onsite, or offsite test

. measurements. or 2) utilizing replacement sensors with certified response

' times.

BRUNSWICK - UNIT 2- B 3/4 3-1 Amendment No. I

, ~3/4.3 INSTRUMENTATION BASES 3/4 3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION On December 29, 1993, the NRC staff issued Generic Letter 93-08.

" Relocation of Technical Specification Tables of Instrument Response Time Limits." Relocation to the Updated Final Safety Analysis Report of the instrument response time table associated with Technical Specification 3/4.3.1 was approved by Amendment No. 202 to the Facility Operating License. Updated Final Safety Analysis Report Table 7.2.1-3 provides the Reactor Protection System instrumentation response times. The NRC staff has reviewed and accepted the alternate metnodologies of BWR Omers' Group Licensing Toncal Report NEDO-32291-A. " System Analyses For the Elimination of Selected Response

. Time Testing Requirements." As a result, the alternate methodologies 3rovided in NEDO-32291-A have been incorporated into the instrument res3onse taale which has been relocated to the Updated Final Safety Analysis Report and are an acceptable means of demonstrating instrument response times.

The bases for the trip settings of the reactor protection system are discussed in the bases for Specification 2.2.

P 4

l BRUNSWICK - UNIT 2 B 3/4 3-la Amendment No. l l

,

  • INSTRUMENTATION BASES l

-3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION '

l This specification ensures the effectiveness of the. instrumentation used  !

to mitigate the consequences of accidents by prescribing the trip settings for i isolation of the reactor systems. When necessary, one channel may be - '

inoperable for brief intervals to conduct required surveillance. Some of the i trip settings have tolerances explicitly stated where both the high and low i values are critical and may have a substantial effect on safety. The  !

setpoints of other instrumentation where only the high or low end of the l setting has a direct bearing on the safety, are established at a level away  !

from the normal operating range to prevent inadvertent actuation of the  !

systems involved S ecified surveillance intervals and allowed out-of-service times were estab ished based on reliability analyses documented in GE reports l' NEDC-30851P-A Supplement 2. " Technical S)ecification Improvement Anal BWPcIsolation Instrumentation Common to R)S and ECCS Instrumentation,"ysis March for i

- 1989 and NEDC-31677P-A, " Technical Sp'ecification Improvement Analysis for BWR l Isolation Actuation Instrumentation, July 1990, as modified by OG90-579-32A.  ;

Letter to Millard L. Wohl (NRC) from W. P. Sullivan and J. F. Klapproth (GE),  !

" Implementation Enhancements to Technical Specification Changes Given in Isolation Actuation Instrumentation Anal 25. 1990 and supplemented by GE letter report GENE-A31-00001-02Assessment "ysis," June of Brunswick Nuclear Plant  !

Isolation Actuation Instrumentation Against NEDC-31677P-A Bounding Analyses," {

August 1994. i i

On December 29. 1993, the NRL staff issuea Generic Letter 93-08.  ;

'" Relocation of Technical Specification Tables of Instrument Response Time  !

Limits. " Relocation to the Updated Final Safety Analysis Report of the  :

instrument res onse time table associated with Technical S ecification 3/4.3.2 -

was approved b Amendment No. 202 to the Facility Operatin License. Updated Final Safety alysis Report Table 7.3.1-3A provides the I olation System l instrumentation response times. The NRC staff has reviewed and accepted the i~

alternate methodologies of BWR Owners' Group Licensing Topical Report NED0-32291-A,'" System Analyses for the Elimination of Selected Response Time Testing Requirements." As a result, the alternate methodologies provided in NED0-32291-A have been incorporated into the instrument response table which  ;

has been relocated to the Updated Final Safety Analysis Report and are an i acceptable means of demonstrating instrument response times. i I

i BRUNSWICK - UNIT 2 B 3/4 3-2 Amendment No. l .

1 l

_. _l

- initiate actions to mitigate the consequences of accidents that are beyond the operator's ability to control. This specification provides the tri) point

~

settings that will ensure effectiveness of the systems to provide tie design protection. Although the instruments are listed by system. in some cases the

~

same instrument is used to send the start signal to several systems at the same time. The out-of-service times for the instruments are consistent with

. the requirements of the specifications in Section 3/4 5. ,

Specified surveillance intervals and allowed out-of-service times were i established based on the reliability anal.yses documented in GE reports NEDC-30936P-A. Parts 1 and 2. "BWR Owners Group Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation .

Instrumentation)." December 1988 and RE-011. Rev. 1. " Technical Specification i Improvement Analysis for the Emergency Core Cooling System Actuation Instrumentation for Brunswick Steam Electric Plant. Units 1 & 2." August 1994. '

as modified by 0G90-319-32D. letter from W. P. Sullivan and J. F. Klapproth (GE) to Millard L. Wohl (NRC). " Clarification of Technical Specification a Changes Given in ECCS Actuation Instrumentation Analysis." March 22. 1990.  :

On December 29. 1993, the NRC staff issued Generic Letter 93-08.

" Relocation of Technical Specification Tables of Instrument Response Time Limits. " Relocation to the Updated Final Safety Analysis Report of the instrument res onse time table associated with Technical Specification 3/4.3.3 +

was approved b Amendment No. 202 to the Facility Operating License. U] dated Final Safety A alysis Report Table 7.3.3-5 provides the Emergency Core Cooling System instrumentation res]onse times. The NRC staff has reviewed and accepted the alternate metlodologies of BWR Owners' Group Licensing To)ical Report NED0-32291-A. " System Analyses For the Elimination of Selected Response .

Time Testing Requirements." As a result, the alternate methodologies 3rovided  !

in NEDO-32291-A have been incorporated into the instrument res)onse taale which has been relocated to the Updated Final Safety Analysis Report and are l an acceptable means of demonstrating instrument response times.

3/4.3 4 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the ,

recuirements of the specifications in Section 3/4.1.4. Rod Program Controls j anc Section 3/4.2. Power Distribution Limits. The trip logic is arranged so ,

that a trip in any one of the inputs will result in a rod block.  !

Specified surveillance intervals and allowed out-of-service times were i established based on the reliability analyses documented in GE report i NEDC-30851P-A. Supplement 1. " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation." October 1988.

3/4.3.5 MONITORING INSTRUMENTATION ,

3/4.3 5.1 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY of.the seismic monitorin instrumention ensures that l sufficient capability is available to prompt y determine the magnitude of a i seismic event and evaluate the response of t ose features important to l safety. This capability is required to permit comparison of the measured j response to that used in the design basis for the facility.

BRUNSWICK - UNIT 2 B 3/4 3-2a Amendment No. l 1

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