ML20101H460

Proposed Tech Specs Re Instrument Channel Response Time Testing

ML20101H460
Person / Time
Site:	FitzPatrick
Issue date:	06/22/1992
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20101H453	List: JPN-92-030, Forwards Application for Amend Tolicense DPR-59,revising TSs to Add Response Time Testing Requirements for RPS & MSIV Actuation Instrumentation Including Analog Transmitter Trip Sys,In Ref to Insp Rept 50-333/91-22 ML20101H457 ML20101H460 ML20101H471
References
JPTS-92-010, JPTS-92-10, NUDOCS 9206290359
Download: ML20101H460 (20)

Text

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ATTACHMENT 1 to JPN 9f.-030 PROPOSED TECHNICAL SPECIF. CATION CHANGES INSTRUMENT CHANNEL RESPONSE TIME TESTING (JPTS 92 010) i l

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l l New York Power Authority l

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 9206290359 920622 PDR ADOCK 05000333 l ~- P PDR u _. __ _ , _ _. . _ _ _ _ _ _ - . _ _ _ _ . _ _ . _ . _ . _ . _ _ _ _ _ _ .

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JAFNPP TECHNICAL SPECIFICATIONS TABLE OF CONTENTS 1.0 Definitions Pagg 1

LIMITING SAFETY SAFETY LIMITS SYSTEM SETTiNDS 1.1 Fuel Cladding intogtity 2.1 7 1.2 Reacto oolant System 2.0 27 SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REQUIREMENTS 3.0 General 4.0 30 3.1 Reactor Protection System 4.1 30f 3.2 Instrumentction 4.2 49 A. Primary Containment isolation Functions A. 49 B. Core and Containment Cooling Systems -Initiation and B. 50 Control l C. Contrel Rod Block Actuation C. 50 D. Radiation Monitoring Systems -Isolation and initiation D. 50 Functions E. Drywell Le'x DStection E. 54 F. Surveillarice Information Roadouts F. 54 G. Recircu:ation Pump Trip G. 54 H. Accident Monitoring Instrumentation H. 54

1. 4kV Emergency Bus Undervoltage Trip 54 3.3 Reactivity Control 4.3 88 A. Reactivity Umitations A. 88 B. Control Rods B. 91 C. Scram insertion Times C. 95 D. Reactivity Anomalies D. 96 3.4 Standby Liquid Control System 4.4 105 A. Normal Operation A. 105 '

B. Operation With inoperable Components B. 106 C. Sodium Pentaborate Solution C. 107 3.5 Core and Containment Cooling Systems 4.5 112 A. Core Spray and LPCI Systems A. 112 B. Containment Cooling Mode of the RHR System B. 115 C. HPCI System C. 117 D. Automatic Depressurization System (ADS) D. 119 E. Reactor Core isolation Cooling (RCIC) System E. 121 Amendment No. ff, pd, p(

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JAFNPP LIST OF TABLES Tablo Title a Pago 3.1 1 Reactor Protection System (Scram) Instrumentation Requirement 41 3.1 2 Reactor Protection System Instrumentation Response Times 43a 4.1 1 Reactor Protection System (Scram) Instrument Functional Tosts 44 4.1 2 Reactor Protection System (Scram) Instrument Calibtation 46 3.2 1 Instrumentation that Initiates Primary Containment isolation 64 ,

3.2 2 Instrumentation that Initiates or Controls the Coro and Containment 66 Cooling Systems 3.2 3 Instrumentation that Initiatos Control Rod Blocks 72 ,

3.24 (DELETED) 74 3.2 5 Instrumentation that Monitors Leakage Detection insido the Drywell 75 3.2-6 (DELETED) 76 3.2-7 Instrumentation that initiatos Recirculation Pump Trip 77 3.2-8 Accident Monitoring Instrumentation 77a 3.2 9 Primary Containment isolation System Actuation Instrumentation 770 Response Times 4.2 1 Minimum Test and Calibration Frequency for PCIS 78 4.2 2 Minimum Test and Calibration Frequency for Core and Containment 79 Cooling System 4.2 3 Minimum Test and Calibration Frequency for Control Rod Blocks 81 Actuation

. 4.2-4 (DELETED) 82 4.2-5 Minimum Test and Calibration Frequency for Drywell Leak Detection 83 4.2-6 (DELETED) 4.2 7 Minimum Test and Calibration Frequency for Recirculation Pump Trip 85 Amendment No. 26, f/I,1)d.1pf, .

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JAFNPP 1.0 (cont'd)

C. Cold Condition - Reactor coolant temperature <a2*F. nsWed CM - M insht M b a WaWe determination of acceptable operability by observation of D. Hot Standby Condition - Hot Standby condition means operation instrument behavior during operation. This detemunation with coolant temperature >212"F, the Mode Switch in Start- shall include, where possible, comparison of the up/ Hot Standby and reactor pressure < 1,005 psig. instrument with other independent instruments measuring E. Immediate - Immediate means that the required action will be sam vMe.

initiated as soon as practicab!e considering the safe operation of 5. Instrument ChanncI Functional Test - An instrument the unit and the importance of the required action. channel funct;onal test means the injection of a simulated F. Instrumentation signal into the instrument pomary sensor whera possible to verify the proper instrument channel response, alarm

1. Functional Test - A functional test is the manual operation anWa Maung adon.

or initiation of a system, subsystem, or component to 6. Primary Containment isolation Actuation instrumentation verify that it functions within design tolerances (e.g., the Response Time for Main Steam !Jne isolution is the time manual start of a care spray pump to verify that it runs and interval which begins when the monitored parameter that it pumps the required volume of water). exceeds the isolation actuation set point at the channel

2. Instrument Channel Calibration - An instrument channel 8"S" ** " #*

• sole s ge haw 06AM, M, G, & W .

calibration means the adjustment of an instrument signal output so that it corresponds, within acceptable range, and pilot solenoid relay contacts open). The response time accuracy, to a known value(s) of the parameter which the ay M nmased in one cmumous step a b ovelappng segments, with venfication that all components are tested.

instrument monitors. Calibration shall encompass the entire instrument channel including actuation. alarm or trip. 7. Logic System Function Test - A logic system functional test l means a test of relays and contacts of a logic circuit from

3. Instrument Channel - An instrument channel means an arrangement of a sensor and auxiliary equipment required sensor to activated device to ensure components are ,

to generate and transmit to a trio system a single trip oper bic per design intent. Where praaticable, action wi!I signal related to the plant para noter monitored by tha go o complebon: Le, pumps wm M sW and vah instrument channel.

opsated.

8. Protective Action - An action initiated by the Protection l System when limiting safety system setting is reacNd. A r 'ective action can be at a channel or system level.

Amendment No. / lyf, 2

JAFNPP

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P t.0 (cont'd)

13. Sensor - A sensor is that part of a channel used to detect l

9. Protective Function - a. system protocuve action which variations in a monitored variable and to provide a suitabic l

results from the protective action of the channels si to N ic monitoring a particular plant condition.

10. Reactor P'otection System Response Time is the time G. Limiting Conditions for Operation (LCO) interval which Degins when the monitored parameter The limiting conditions for operation specify the minimum exceeds the reactor protection trip set point at the acceptabic levels of system performance necessary to assure channel senscr and ends when the scram pilot valve safe start-up and operation of the facriity. When these conditions solenoids are de-energized (05A-K14 scram contactors are meti the plant can be operated safety and abnormal open). The response time may be measured in one situations can be safely contro: led.

continuous step or in overlapping segments, with verification that all components are tested, H. Limiting Safety System Setting (LSSS)

11. Simulated Automatic Actuation - Simu!ated automatic The limiting safety system settings are settings on l actuation means applying a simulated signal to the sensor instrumentation which initiate the automatic protective action at to actuate the circuit in question. a level such that the safety limits will not be exceeded. The region between the safety limit and these settings represent

12. Trip System - A in.p systern means an arrangement of margin with normal operation lying below these settings. The l instrument channel trip signals and auxiitary equipment margin has been established so that with proper operation of tne required to initiate action to accomplish a protective instrumentation safety limits will ncvcr be exceeded.

function. A trip system rnay require one or more I instrument channel trip signals related to one or more L Modes of Operation (Operational Mode) plant parameters in order to initiate trip system action. Mode - The reactor mode is established by the Mode Selection loitiation of protective action may require the tripping of a Switch. The modes include shutdown, refuel. start-up/ hot single trip system or the coincident of two top systems. standby, and run which are defined as follows:

l Amendment No. Jid ,

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JAFNPP  !

3.1 LIMITING CONDITIONS FOR OPERATION 4.1 SURVE1LL d9CE REOUIREMENTS

, 3.1 REACTOR PROTECTION SYSTEM 4.1 REACTr A PROTECTION SYSTEM

. Applicability

Applicability-Applies to the instrumentation and associated devices which initiate Applies to the survei!!ance of the instrumentation and associated the reacto: scram. devices which initiate reactor scram.

Objective: Objective:

To assure the operability of the Reactor Protection System. To soecify the type of frequency of surveillance to be applied to the i protedon instrumentation.

.v I Specification: Specification:

A. The setpoints, min. . um number of trip systems, and minimum A. Instrumentation systems shall be functionally tested and number of instrument channels that must be operable for each calibrated as indicated in Tables 4.1-1 and 4.1-2 respectively.

position of the reactor mode switch, shall be as shown in The mye th for e@ ream poMion ytm t@

g Table 3.1-1. The reactor protection system instrumentation function listed in Table 3.1-2 shall be demonstrated to be within

l response time shall be within the limits in Table 3.1-2.

the limits in the table during each 18 month test interva!. Each

. test shall include at least one channel in each trip system. A!!

! channels in both trip systems shall be tested within two test i i intervals.

B. Minimum Critical Pcwer Ratio (MCPR) B. Maximum Fraction of Umiting Power Density (MFLPD)

During reactor power operation, the MCPR operating limit shall The MFLPD sha!! be determined dail/ during reactor power ,

not be less than that shown in the Core Operating Umits Report. operation at >25% rated thermal power and the APRM high flux

. scram and Rod Block trip settings adjusted if necessary as j 1. During Reactor power operat. ion with core flow less than specified in the Core Operating Umits Report.

j 100% of ratcd, the MCPR operating limit shall be multiplied l- by the appropriate K, as specified in the Core Operating j' Umits Report.

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Amendment No. 9J,94. 55,109, Id, j i 30f  !

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JAFNPP 4.1 BASES (cont'd)

The b;-stable trip circuit which is a part of the Grcup (S) The frequency of calibration of the APRM flow biasing network devices can sustain unsafe failures which are revealed on!y on has been established as each refucting outage. The flow test. Therefore,it te necessary to test them periodically. biasing network is functionahy tested at least once/ month and, in a m, coss cabadon Ms oWow W to Mow A study was conducted of the instrumentation channels biasing network can be made during the functional test by included in the Group (B) devices to calculate their unsafe direct meter reading. There arc several instruments which failure rates. The non-ATTS (Analog Transmitter Trip System) must be calibrated and it will take several days to perform the analog devices (sensors and amplifiers)4are predicted to have cahbration of the entire network. While the calibration is being an unsafe failure rate of icss than 20x10 failures /hr. The non- p orm , m ow gnal wW M smt to ham of M Ay s l ATTS bi-stable trip circuits ara 3redicted to have unsafe faffure resulting in h if scram and rod block condition. Thus,if the  ;

4 'hr. The ATTS analog devices rate of less than 2x10 faik camahon wm Wormed dudng operabon, Uux captng (sensors), bi-stable devices 1,aaster and slave trip units) and wm nt M possbo. Bad on openew at , oh power supplies have been evafuated for reliability by Mean generating stations, drift of instruments, such as those in the Tirne Between Failure analysis or state-of-the-art qualification Iow biasing network, is not sigificant and therefore, to avoid type testing meeting the requirements of IEEE 323-1974. spurious scrams, a calibration frequency of each refueling Considering the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> monitoring interval for analog devices outage is established.

as assumed above, the instrument checks and functional tcsts as well as the analyses and/or qualification type testing of the The measurement of response time within the specified oevices, the design reliability goal for system reliability of intervals provides assurance that the Reactor Protection 0.9999 will be attained with ample margin. System trip functions are completed within the time limits assumed in the transient and accident analyses.

The bi-stable devices are monitored during plant operation to record ' heir failure history and establish a test interval using the The Reactor Protection System trip functions in Table 3.1-2 are those fuctions for which the transient and accident analyses cume o: Figure 4.1-1. There are numerous identical bi-stable devices used throughout the Plant's instrumentation system. described in CJiapter 14 of the FSAR take crcdit for the Therefore, significant data on the failure rates for the bi-stable response time o instrument channels.

devices should be accumulated rapid!y.

Amendment No. Xd 1pf, 38 n - I

JAFNPP 4.1 BASES (cont'd)

Low Reactor Water level sensor 1000 ms In terms of the transient analysis, the Standard Technical Specifications define individual trip function response time as Main Steam isolation Vafve Closure 10 ms the time interval from when the monitored parameter exceeds and Turbine Stop Vaive Closure switches its trip setpoint at the channel sensor until de-energization of the Turbine Control Valve Fast Closure 30 ms l scram pdot valve solenoids." The individual sensor response j time def:ned as " operating time" in General Electric (GE) design from the first movement of the main turbino control valves until actuation of pressure switches which j specification data sheet 22A3083AJ, note (8), is "the maximum  !

a!!owable time from when the variable being measured just detect the loss of hydraulic control oil pressure.

j exceeds the trip setpoint to opening of the trip channel sensor The 10 ms limit for the MSIV and TSV position switch response f j

contact during a transten!J A transient is defined in note (4) of time is defined by GE design specification data sheet i

the srme data sheet as "the maximum expected rate of change 22A3083AJ. It requires that after MSIV or TSV moves to the set I i I

of the variab!c for the accident or the abnormal operating point corresponding to 10% closure from fu!! cpen, the position i l condition which is postulated in the safety ana!ysis report. switch contacts shou!d open in less than or equal to 10 ms. f When the correct set point is verified by surveillance testing for j

' The individual sensor rcsponse time may be measured by the pos: tion switch, the response time requirement is considered  ;

simulating a step change of the particular parameter. This to be satisfied. The maximum permissible TCV fast closure l method provides a conservaDve value for the sensor response  ;

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time, and confirms that the instrument has retained its specified channel, logic, and scram contactor response time is 70 ms l rather than the sum of TGV fast closure logic (30 ms) and the trip l electromechanical characteristics. When sensor response time l is measured independently, it is necessary to a!so measure the logic and scram contactor response time (50 ms). This provides  !

a 10 ms margin to a!iow for uncertainty in the test method.

remaining portion of the response time in the logic train up to the time at which the scram pilot valve so!cnoids de-energize. The .The maximum permissic!c APRM channel, logic, and scram  !

channel response time must include all component delays in the contactor response time is 90 ms rather than the sum of the i response chain to the ATTS output relay plus the 50 ms design APRM cnannel response time (60 ms) and the trip logic and  !

scram contactor response time (50 ms).. - (GE design i allowance for RPS iogic system response time. A response time I

for the RPS logic retays in excese of 50 ms is acceptab!c specification data sheet 22A3083AJ), note (12). This measurement is applicab!c to both the APRM fixed high neutron I provided the overa!! response time does not exceed the response time limits of Table 3.1-2 which includes a!!owances for f!ux and the flow referenced simulated thermal power channels I and requires measuring the time delay through the LPRM cards. 1 sensors, relays, and switches as follows:

The latter case does not inc!ude the time constant of High Reactor Pressure sensor 500 ms approximatdy six seconds which is calibrated separately. The 550 ms basis for cycluding the neutron detectors from response time High DryweII Pressure sensor testing is provided by NRC Regu!atory Guide 1.118, Revision 2, section C.S.

Amendment No. p6, 39

i JAFNPP .

4.1 BASES (cont'd)

The 18 month response time testing interval is based on NRC For the APRM System, drift of electronic apparatus is not the NUREG-0123, Revision 3, " Standard Technical Specifications," only consideration in determining a calibration frequency.

surveillance requirement 4.3.1.3. Change in power distribution and loss of chamber sensitivity ddate a mation wery Mys.

Group (C) devices are active only during a given portion of the

) operational cycle. For example, the IRM is active during start- Calibration on this frequency assures plant operation at or l up and inactive during full power operation. Thus the only test below thermal limits. t

j. l that is meaningful is the one performed just prior to shutdown A comparison of Tables 4.1-1 and 4.1-2 (ndicates that two j or start-up; i.e., the tests that are performed just prior to use of '

instrument channels have not been included in the latter table.

the instrument.

These are: mode switch in shutdown and manual scram. All cf +

I Calibration frequency of the instrument channel is divided into the devices or sensors associated with these scram functions l two groups. These are as follows: are simple on-off switches and, hence calibration during

! operation is not applicable.

1. k siv? type indicating devices that can be compared  ;

! with Li<e units on a continuous basis.

B. The MFLPD is checked once per day to determine if the APRM l 2. Vacuum tube or semiconductor devices and detectors scram requires adjustment. Only a smail number of control j t

that drift or lose sensitivity. rods are moved daily and thus the MFLPD is not expected to  :

! change significantly and thus a daily check of the MFLPD is adequate.

6perience with passive type instruments in generating stations

and substations indicates that the specified calibrations are The sensitivity of LPRM detectors decreases with exposure to

adequate. For those devices which employ amplifiers, etc., neutron flux at a slow and approximately constant rate. This is - -

drift specifications call for drift to be less than 0.4 compensated for in the APRM system by calibrating twice a i percent / month; i.e., in the period of a month a maximum drift week using heat balance data and by calibrating individual j of 0.4 percent could occur, thus providing for adequate margin. LPRM's every 1000 effective full power hours, using TIP traverse data. f t

l Amendment No. jg,16, 40

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t JAFNPP l TABLE 3.1-2~~

l REACTOR PROTECTION SYSTEM INSTRUMENTATION Rt:SPONSE TIMES REACTOR TRIP SYSTEM TRIP FUNCTION RESPONSE TIME j (SecondsJ Reactor Vessel Pressure - High < 0.550 1)

(02-3PT-55A, B, C, D)

2) Drywell Pressure - High 5 0.600 t (05PT-12A, B, C, D)

< 1.050 g

3) Reactor Water Level - Low (L3) I (02-3LT-101 A, B, C, D) i

< 0.060 [

Main Steam isolation Valvt. ^losure (29PNS-80A2,82, C2. D2)

(29PNS-86A2,82, C2, D2) b) Turbine Stop Va!ve Closure 5 0'SO (94PNS-101,102,103,104)

6) Turbine Control Valve Fast C!osure 5 027]  ;

(94PS-200A, B, C, D)

$ 0.090 (2)

7) APRM Fixed (120%) Hit h Neutron Flux APRM Flow Referenced Simulated Thermal Power 5 0.090 (1) (2) 8)

Notes for Table 3.1-2:

1. Trip system response tima does not include the simuiateo thermal power time constant of approximately six seconds which is cali' orated separat

2. Trip system response time is the measured time interval frorr i sign al input to the first electronic component in the channel after the LPRM detector until the scram pilot valve solenoide de-ene " )SA-Kt 4 scram contactors open).

Amendment No.

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JAFNPP .

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TABLE 4.1-2  ;

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUEMCIES FOR REACTOR PRO 112CTION INSTRUMENT CHANNELS Instrument Channel Group (1) Calibration Minimum Frequency (2) l IRM High Flux C Ccmparison to APRM on Maximum frequency onc */ week Controlled Shutdowns APRM High Flux Output Signal B Heat Balance Daily Flow Bias Signal B Intemal Power and Flow Test Every refueling outage with Standard Pressure Source LPRM Signal B TIP System Traverse Every 1000 effective full power hours High Reactor Pressure B Standard Pressure Source Note (6)

High Drywell Pressure B Standard Pressure Source Note (6)

Reactor Low Water Level B Standard Pressure Source Note (6)

Hig, Water Levelin Scrara A Water C )lumn, Note (5) Once/ operating cycle. Note (5)

Discharge Instrument Volume High Water Levelin Scram 3 Standard Pressure Source Every 3 months Discharge Instrument Volume Main Steam Line isolation A Note (4) Note (4)

Valve Closure l Main Steam Line High Radiation B Standard Current Source (3) Every 3 months Turbine First Stage Pressure B Standard Pressure Source Note (6)

Permissive l.

Amendment No. g,'@ W M, PA,1%.

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JAFNPP ,

TABLE 4.1-2 (Cont'd)

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CAllBRATION MINIMUM CAllBRATION CREQUENCIES FOR REACTOR PROTECTION INSTRUIAENT CHANNELS

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Instrument Channel Group (1) Calibration Minimum Frequency (2)

Turbine CcMrol Valve Fast A Standard Pressure Source Once/ operating cycle Closure OC Pressure Trip j Tur'Mne Stop Valve Closure A Note (4) Note (4)

NOTES FOR TABLE 4.1-2

1. A description of three groups is included in the Bases of this Specification.

2. Calibration test is not required on the part of the system that is not required to be operabic, or is tripped, but is required prior to return to service.

3. The current source provides an instrument channel a!ignment. Calibration using a radiation source sha!! be performed each refueling outage.

4. Actuation of thesa switches by normal means will be performed during the refueling outages.

5. Calibration shall be performed ut:lizing a water column or similar device to provide assurance that damage to a float or other portions of the float assembly will be detected.

l 6. Sensor calibration once per operating cycle. Master / stave trip unit calibration once per 6 months.

Amendment No. 46, F/, Q6,1%,

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JAFNPP

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42 SURVEILLANCE REQUIREMENTS 3.2 UMITING CONDITIONS FOR OFERATION 42 INSTRUMENTATION  :

32 INSTRUMENTATION Applicability: ,

Applicability:.

Applies to the surveillance requirement of the instrumentation which Applies to the plant instrumentation which either (1) initiates and either (1) initiates and controls protective function, or (2) provides controls a protective function, or (2) provides information to aid the information to a,d n the operator in monitoring and asscssing plant status operator in monitoring and assessing plant status during normal and during normal and accident conditions.

accident conditions.

Objective:

Objec+ive:

To assure the operability of the aforementioned instrumentation. specythe type and frequency of surveilLt.,:e to be applied to the aforementioned instrumentation.

Specifications:

Specificctions: i A. Primary Containment isolation Functions A Primary Containment Isolation Functions instiumentation sha!! be functionally tested and calibrated as When primary containment integrity is requirca, the limiting conditions of operation for the instrumentauon that initiates indicaW in TaNe 424 primary containment isolation are given in Table 3.2-1. System logic shall be functionally tested as indicated in When primary containment integrity is required, the primary aW 424 .

containment isolation actuation instrumentation response time The response time of each primary cont 6. ment isolation for MSIV closure shall be within the limits in Tabic 32-9. actuation instrumentation isolation trip function listed in Tabic 3.2-9 shall be demonstrated to be within the limits in the tab!c l during each 18 month test interval. Each test shall include at r least one channel in each trip system. All channels in both trio  !

systems shall be tested within two test intervals.

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I Amendment No.1d6, ,

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JAFNPP 3.2 (cont'd) 42 (cont'd)

B. Core and Containment Cooling Systerns - Initiation and Control B. Core and Containment Cooling Systems - Initiation and Control Instmmenta ion shall be functionally tested, calibrated, and The limiting conditions for operation for the instrumentation that initiates or controls the Gore and Coniainment Cooling Systems checked as } indicated in Table 42-2.

are given in Table 3.2-2. This instrumentation must be operable System logic shail be functionally tested as indicated in when the system (s) it initiates or controls are required to be Table 4.2-2.

operable as specified in Specification 3.5.

C. Control Rod Block Actuation C. Control Rod Block Actuation

1. The limiting conditions of operation for the instrumentation instmmentagon su & "soctionally tested, calibrated, and checked as indicated in Table 42-3.

that initiates control rod block are given in Tabic 32-3.

The minimum number of operable instrument channels

%s em I g shau M funcUonaHy tesM as Mcated M ,

2.

specified in Table 3.2-3 for the rod block monitor may be aNe 424 reduced by one in one of the trip systems for mair.tenance and/c '.oting, provided that this condition does not last longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 30 day period.

D. Radiation Monitoring Systems - Isolation and initiation Functions D. Radiation IVonitoring Systems - Isolation and initiation Functions Refer to the Radiological Effluent Technical Specifications (Appendix B). Refer to the Radiological Effluent Technical Specifications (Appendix B).

l Ar endment No 96, 50

.I JAFNPP 42 BASES i

F The instrumentation listed in Tables 4.2-1 through 42-8 will be .

functionally tested ' and . calibrated at regularly scheduled '- '

r intervals. The same design reliability goal as the Reactor Whee:

Protection System is generaHy applied. Sensors, trip devices and power supplies are tested, calibrated and checked at the i o at Ween tests  !

same frequency as comparable devices in the Reactor t= the time the trip contacts are disabled from Protection System. performing their function while the test is in progress.

The response times for MS!V isolation in Table 3.2-9 include the primary sensor and all components of the logic which must r= the expected failure rate of the relays.

function to de-energize the MSIV pilot valve solenoids. To test the trip relays requires that the channel be bypassed, Electrolytic filter capacitors are installed on the i,nput to the the test made, and the system retumed to its initial state. It is main steam line flow ATTS trip units. General Electnc a talysis assumed this task requires an estimated 30 ninutes to (MDE-278-1285 December 1985) accounts for the delay complete in a thorough and workmanlike manner and that the caused by the capacitors and justifies the increase in response relays have a failure rate of 104 tailures per hour. Using this time to 2.5 seconds for the main steam line high flow actuation data and the above operation, the optimum test intermi is:

signal. With the exception of the MSIVs. response time testing

'is not required for any other primary containment isolation actuation instrumentation. The safety analyses results are not i= d N =1r10'hr.

sensitive to individual sensor response times of the logic 10*'

systems to which the sensors are connected for isolation = 40 days actuation instrumentation. For additional margin a test interval of once/ month will be used inWany.

Those instruments which, when tripped, result in a rod block have their contacts arranged in a 1 out of n logic, and all are The sensors and electronic appnatus have not been included capable of being bypassed. For such a tripping arrangement here as these are analog devicea with readouts in the control with bypass capability provided, there is an optimum test room and the sensors and electronic apparatus can be interval that should be maintained in order to maximize the checked by corraarison with other like instruments. TM l reliability of a given channel (7). This takes account of the fact checks which are made on a daily casis are adequate to asse that testing degraces reliability and the optimum interval operability of the sensors and electronic apparatus, and in>

between tests is approximately given by: test interva! given above provides for optimum testing of the relay circuits.

Amendment No. pS, tif4, Iff, 61

l JAFNPP TABLE 3.2-9 PRIMARY CONTAINMENT ISOLATION SYSTEM (syTUATION INSTRUMENTATION RESPONSE TIMES TRIP FUNCTION RESPONSE TIME (Seconds)

1) MSIV Closure - Reactor Low Water Level (L1) < 1.0 (02-3LT-57A, B and 02-3LT-58A, B)

2) MSIV Closure - Low Steam Line Pressure <

1.0 (02PT-134A, B, C, D)

3) MSIV Closure - High Steam Line Flow <_

2.5 (02 D PT- 1 16A- D, 1 17A- D, 1 18 A- D, 1 19A- D)

Note for Table 3.2-9:

The measurement of the response time interval begins when the monitored parameter exceeds the isolation actuation set point at the channel se.wr -

and ends when the Main Steam Isolation Valve pilot solenoid relay contacts open. The pilot solenoid relay contacts to be used for datermination of the end point of the response time measurement are: ,

For the inboard MSIV pilot solenoid relays: 16A-K14 (acso!enoids) .

16A-K51 (dc solenoids)

For the Outboard MSIV pilot solenoid relays: 16A-K16 (ac solenoids) 16A-K52 (dc solenoids)

Amendment No.

77e

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ATTACHMENT 11 to JPN 92-030 SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES INSTRUMENT CHANNEL RESPONSE TIME TESTING (JPTS 92-010) 4 I

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New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 l

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ATTACHMENT ll to JPN-92-030 Safety Evatustion for Proposed Changes to Technical Specifications RESPONSE TIME TESTING (JPTS 92-010)  ;

l. bESCRIPTION OF THE PROPOSED CHANGES, The proposed changes will add response time testing (RTT) requirements for the instrument channels for eleven trip functions to the definition, limiting conditions for operation, and surveillance requirements sections. Existing reactor protection system (RPS) relay logic response time testing requirements will be eliminated for those trip functions which are not

, included in the RTT requirements of the Standard Technical Specifications (STS) (Reference 1),

and for which response time is not considered in the accident and transient analyses described in the FSAR (Reference 11). To be consistent with reference 1, RTT requirements will be limited to those trip functions for which the instrument channel response time is significant (when compared to the total system response time) and for which credit for response time is taken in the transient and accider't analyses described in the FSAR. The response time limits will be increased from the prev ous 50 ms limit to accommodate the response tir a of the additional sensors, analog transm'tter trip system (ATTS) components, and other components installed in the instrument channrJs by (MOD-F1-82-053), and those components which were already existing in instrument channels and which will now, by the new definitions, be included in response time testing. The frequency of individual channel response time testing will be decreased from the existing requirements to reduce personnel radiation exposure during testing and to be consistent with reference 1.

p.gae i. Table of Contents, 3.2.8 Core and Containment Coolina Systems - Initiation and Control Replace the page number "49" with the page number "50" Section 3.2.B will be moved from page 49 to page 50 tc. provide space on page 49 for the additional response time surveillance requirements of 4.2.A.

Paae v. List of Tables. 3.1-2 and 3.2-9 Add provision for two new tables:

" 3.1 - 2 Reactor Protection System instrumentation Response Times 43a"

" 3. 2-9 Primary Containment Isolation System Actuation Instrumentation Response Times 77e" Paae 2 Definitions.1.0.F.6.

Change the number of the existing definition "6." to "7." and insert a new definition:

"6. P imary Containment isolation Actuation Instrumentation Resoonse Time for Main Steam Line isolation is the time interval which begins when the monitored parameter exceeds the isolation actuation aet point at the channel sensor and ends when the Main Steam ! solation Valve solenoHs are de-energized (16A-K14, K16, K51 & K52 pilot solenoid relay contacts open). The response time may be measured in one continuous step or in o'.erlapping segments, with verification that all components are tested."

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ATTACHMENT ll to JPN 92 030

, Safsty Evaluation for Propossd Changes to Technical Spacifications RESPONSE TIME TESTING (JPTG 92-010)

Eaoe 3. Definitions.1.0.F. 7. 8. 9.10 and 11 Change the number of the existing definition "7. Protective Action" to "8." and move the definition to the preceding page.

Change the number of the existing definition "8. Protective Function" to "9."

insert a new definition:

"10. Reactor Protection System Response Time is the time interval which begins when the monitored parameter exceeds the reactor protection trip set point at the channel sensor and ends when the scram pilot valve solenoids are de-energized (05A-K14 scram contactors open). The respcase time may be measured in one continuous step or in overlapping segments, with verification that all components are tested."

Advance by two, the numbers of the existing definitions "9. Simulated Automatic Actuation", "10. Trip System", and "11. Sensor" to the numbers 11,12, and 13, respectivelv, to reflect the insertion of the two new definitions preceding these sections.

Pace 30f. Soecifications 3.1. A 3.1.A Retain the first sentence.

Delete the existing second sentence:

"The design system response time from the opening of the sensor contact to and including the opening of the trip actuator contacts shall not exceed 50 ms."

l Add the following new sentence:

"The reactor protection system instrumentation response time shall be within the limits in Table 3.1-2."

Pace 30f. Soecification 4.1.A 4.1.A Fatain the first sentence.

Add the following paragraph:

"The response time for each reactor protection system trip function listed in Table 3.1-2 shall be demonstrated i to be within the limits in the table during ecch 18 month l test interval. Each test shallinclude at least one channel l in each trip system. All channels in both trip system shall be tested within two test intervals."

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ATTACHMENT ll to JPN 92-030

. Safety Evaluation for Proposed Changes to Technical Specifications RESPONSE TIME TESTING (JPTS-92-010)

~Peae 38. 4.1 Bases Transfer the last paragraph in the right column beginning " Group (C) devices are . . ."

and ending." . . . tests that" from page 38 to page 40. Page 40 was previously blank.

Insert tbs following two new paragraphs at the end of the right column on page 38.

" The measurement of response time within the specified intervals provides assurance that the reactor protcction system trip functions are completed within the time limit assumed in the transient and accident analyses.

The Reactor Protection System trip functions in Table 3.1-2 are those functions for which the transient and accident analyses described in Chapter 14 of the FSAR take credit for the response time of instrument channels. "

Pace 3_9. 4.1 Bases Transfer all of the existing text on page 39 to page 40 following the tvio paragraphs previously transferred from page 38. Page 40 was previously blank.

-Insert the new text which appears on page 39 provided in Attachment i to this application for amendment, Page 39 will now consist entirely of new text.

face 40. 4.1 Cases Page 40 was previously blank.

Insert, Leginning in the top lef t column, the following new paragraph:

The 18 month response time testing intervalis based on NRC NJREG-0123 Revision 3 " Standard Technical Specifications", surveillance requirement 4.3.1.3.

i Following new text above, insert the paragraph transferred from the bottom right column of the old page 38 which begins " Group (C) devices are. " Following this paragraph, insert all of the text which was previously located on page 39.

In the paragraph which begins " Group (C) devices are.." change the word "to" in the third senter.ce which reads:

"Thus the only test that is meaningful to the one performed just prior to shutdown ."

to the word "is" so that the sentence will now read:

"Thus the only test that is meaningfulis the one performed just prior to shutdown "

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1 ATTACHMENT 11 to JPN-92 030

. Safety Evaluation for Proposed Changes to Technical Specifications ,

RESPONSE TIME TESTING (JPTS-92-0101 j l

Eaae 43a. Table 3.1-2 Reactor Protectionjvstem Instrumentation Response Times Create a new page 43a and insert the new table identified above and provided in Attachment I to this application for amendment. This new table identifies the reactor trip system re.sponse time limits for seven trip functions.

Paae 46. Ta_ le 4.1.2 Reactor Protection System Instrument Calibration The existing reference note numbers, (4) and above, have teen deleted, or decreased by one, to reflect the deletion of note 4 on the following page 47, and the renumbering of the remaining notes.

On the column heading line, delete "(4)" after the column heading " Calibration."

In the " Calibration" column, change " Note (6)" to " Note (5)" and " Note (5)" to " Note (4)."

In the " Minimum Frequency" column, char.ge the four occurrences of " Note (7)" to

" Note (6)"; change " Note (6)" to " Note (5)"; and change " Note (5)" to " Note (4)."

Pane 47. Table 4.1.2 (Continued)

In the " NOTES FOR TABLE 4.1-2", delete the existing note 4. in its entirety.

"4. Response time is not a part of the routine instrument channel test but will be checked once per operating cycle."

Decrease the number of the following notes by one. Accorcingly, the existing notes 5,6, and 7, should be changed to notec 4,5, and 6 respectively.

In the table column heading line, delete the notation "(4)" after the word " Calibration." ,

in the " Calibration" column, change " Note (5)" to " Note (4)."

in the " Minimum Frequency" column, change " Note (5)" to " Note (4)."

Pace 49. Section 3.2.A. Prinary Containment Isolation Functions Add the following new sentence:

"When primary containment integrity is required, the primary containment isolation actuation instrumentation response time for MSIV closure shall be within the limits in Table 3.2-9."

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ATTACHMENT 11 to #N 92-030 Safaty Evaluation for Propossd Changes to Technical Specifications RESPONSElLME TESTING (JPTS-92-010) m Paae 49 Section 4.2.A Primary Containment isolation Functions Retain the first two sentences. Add the following paragraph:

"The . response time of each primary containment isolation actuati'en instrumentation isolation trip function listed in Table 3.2-9 shall be demonstrated to be within the limits in the tcble during each 18 month test interval. Each test shall include at least one i.,hannel in each trip system. All channels in both trip system shall be tested within two test intervals."

Pace 49 Sections 3.2.B. & 4.2.B Core and Containment Cooling Systems -Initiation and Control Delete s tions 3.2.B and 4.2.B from the page 49 and move them to the top of the following page 50. The relocation of these sections provides space for the expanded section 4.2.A.

Pace 50 Sections 3.2.8 and 4.2.B lasert at the top of the page sactions 3.2.B and 4.2.B which thould be transferred from the preceding page 49.

Pace 61. 4.2 Bases Following the first paragraph which ends with the words " . in the Reactor Protection System.", insert the following new paragraph:

"The response times for MSIV isolation in i'able 3.2-9 include the primary sensor and all components of the logic which must function te de-energize the MSIV pilot valve solenoids.

Electrolytic filter capacitors are installed on the input of the main steam line ilow ATTS trip units. General Electric analysis (MDE-278-1285 Decamber 1985)) accounts for the delay caused by the filter capacitors ar'd justifies the increase in response time to 2.5 ,

seconds for the main steam line high ficw actuction signal. With the exception of the MSIVs, response time testing is not required for any other primary containment isolation cctuation instrumentation. The safety analyses results are not sensitive to individual sensor response time of the legic systems to which the sensore are connected fcr iso'ation actuation instrumentation."

Paae 77e. Table 3.2-9 Isolation Actuation Instrumentation Resoonse Times i

i Create a new page 77e and insert the new Table 3.2-9 provided in Attachment i to this application for amendment. This table provides the response time limits from input of a trip signal to a sensor through and including the de-energizing of the pilot valve solenoids for the Main Steam Isolation Valves (MSIV) for signals from low reactor water level, low steam line pressure and high steam line flow.

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ATTACHMENT 11 to JPN-92-030 Safety Evaluation for Proposed Changes to Technical Specifications RESPONSE TIME TESTING (JPTS-92-010)

11. ' PURPOSE OF THE PROPOSED CHANGES An evaluation (Reference 2) identified those reactor protection system (RPS) and primary l containment isolation system (PCIS) trip functions for which the instrument channel response '

time is considered in the transient and accident snalyses described in the FSAR (Reference 11).

The proposed changes will expand the number of components in an instrument channel, whose response time is to be included in total channel response time, to be consistent with the response time assumptions in reference 11. The number of trip functions to be tested will be decreased to be consistent with Standard Technical Specifications (Reference 1) and reference 11. The frequency of channel RTT will be decreased to reduce personnel radiation exposure during testing and to be consistent with reference 1.

III. SAFETY IMPLICATIONS OF THE F ROPOSED CHANGES A. Background The General Electric Company prepared a licensing topical report (Reference 3) describing the ATTS. The NRC reviewed and accept 9d this report as a basis for ATTG in June 1978 (Reference 4). The Authority installed the ATTS at the James A FitzPatrick Nuclear Power Picat during 1985 (MOD-F1-82-053). The Authority submitted a proposed change to the Technical Specifications (JPTS-85-04)(Reference 5) to support operation with the ATTS and -

used Reference 3 as the basis for the change. The proposed amendment changed the surveillance and calibration requirements to accommodate the ATTS. However, the proposed amendment did not change requirements for RTT to reflect the methods described in the licensing topical report nor the Standard Technical Specifications. The NRC issued Amendment 89 to the Technical Specifications (Reference 6) based in part on Authority compliance witt. the licensing topical report.

The existing technical specification limiting condition for operation 3.1.A. states: "The design system response time from the opening of the sensor contact to and including the opening of the trip actuator contacts shallnot exceed 50 ms. " The Authority continued to measure response time through the RPS logic to meet the technical specification requirement. The

! testing method did not include the respcase time of the ATTS components (sensor, trip unit, and relay) as descdued in the General Electric ATTS licensing topical report (Reference 3).

L in April 1992, General Electric (DRG-A00-03658-1) evaluated the entire scope of response time testing requirements for both ATTS and other instrument channels, based on the transient and accident analyses described in the FSAR. The evatustion identified the specific instrument channels for which there is a basis to require RTT. The evaluation was reviewed l by the Authority in accordance with Engineering Design Control Manual (DCM) procedure 11

-(Reference 2). The DCM-11 review provided the basis for excluding specific channels from RTT testing.

l B. Increased Assurance of Safety l

The proposed changes to the Technical Specifications will improve the ability to detect instrumentation response time deficiencies. Accordingly, the proposed changes wil! provide

increased assurance of plant safety. The proposed changes are based on References 1,2, and 3. References 1 and 3 have been previously reviewed and accepted by the NRC.

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ATTACHMENT ll to JPN 92-030 Safety Evaluation for Proposed Changes to Technical Specifications RiiSPONSE TIME TESTING QPTS 92 010)

C. Increased Number of Channel Componento Subject to Response Time Testing The reactor protection system and main steam line isolation actuation instrumentation RTT willinclude all components in the channel, beginning with the sensor and including the ATTS components and RPS logic relays, through and including opening of tha contactors which de-energize the scram pilot valve solenoids, or the MSIV actuation solenoids as applicable, increasing the number of channel components subject to the response time testing will provide incraased assurance of safety.

D. Increased Time Allowar.co for Channel Response Time The original 50 ms RTT limit included only the time from opening of the sensor contacts through the RPS le'ic channel relays to opening of the scram contactors. In accordance with References 1 and 3, the response time of all components in the channel will be included, and in particular, the sensor and ATTS component response times. Accordingly, the proposed response times will be increased to include the original 50 ms for the RPS logic relays and an additional time interval for the sensor, ATTS, and other components in the channel. The response time of the neutron monitoring sensors will not be included in channel response time measurements because 6 hey are excluded by NRC Regulatory Guide 1.118 (Reference 7). The bases for the increase in' response time limits is provided in the proposed changes to Bases sections 4.1.A and 4.2.A of the Technical Specifications.

E. Reduction in Channel Testing Frequency Previously all channels of each trip function were tested on an 18 month interval. The proposed change will require that one channel of each trip function be testou ;u each trip system during an 18 month interval and that all channels of each trip function be tested within two test intervals. This reduction in channel testing frequency is based on Standard Technical Specifications (Reference 1), surveillance requirement 4.3.1.3.

F. Reduction in Number of Trip Functions Subject to RTT Prior to installation of the ATTS * ;sponse time testing was conducted to insure that the 50 ms limit was maintained for C. .p functions in the RPS logic channels. The addition of tha ATTS components, and re@ ire nents to include sensor responsa time as part of the overall channel response time, willincuase significantly the time required to perform the tusts and increase the associated radieJon exposure to personnel. However, instrument channel response time is only appropr' ate for those trip functions for which the response time is used as a significant input to the transient and accident analyses described in the FSAR. This limited application of RTT h censistent with the Standard Technical Specifications.

The General Electric licensing topical report (Reference 3) describes response time testing metFods to be applied to instrument channels containing ATTS components. The General Electric RTT basis document (DRG-A00-03658-1) states that the methods described in Reference 3, pages A-1 and A-2, were intended for application to those channels where the trip function is identified in the Technical Specifications as requiring RTT. When it is determined that RTT is not applicable to a porticular trip function there is no requiretrent to perform RTT on the ATTS components and logic for that trip function. This is consisu.;t with Standard Technical Specifications. The requirements for those ATTS components will he the normal calibration and functional test surveillance which are currently performec.

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ATTACHMENT ll to JPN 92-030 Safety Evaluation for Proposed Changes to Technical Spscifications RESPONSE TIME TESTING (JPTS-92-010)

1. Reactor Protection System The results of the transient and accident analyses described in Chapter 14 of the FSAR are potentially sensitive to response time for the eight trip functions listed in the proposed Technical Specification Table 3.1 2 " Reactor Protection System instrumentation Response Times." Therefore, restricting RTT to the parameters in the proposed Table 3.1-2. is justified and consistent with Standard Technical Specifications. The high drywell pressure trip function wi also be tested.

2, Primary Containment Isolat'an Actuation With the exception of the MSIV3, RTT is not required for any of the isolation systems and associated isolation actuation instrLmentation. The safety analyses are not sensitive to individual sensor response times of the logic systems to which the sensors are connected for isolation actuation instrumentation.

This position is suppor;ed by the following analyses which examine the basis for primary containment isolation response time with respect to the loss of coolant accident (LOCA),

and high energy line break (hD.B).

a. LOCA Considerations PCIS initiation following a LOCA is provided to minimize offsite dose effects and to minimize loss of inventory from the reactor pressure vessel Because of the conservative assumptions used in the dose calculations, consideration of the resronse times of the isolation actuation instrumentation for the containment isolation valves will not alter the results of the analyses.

RTT of MSIV isolation instrumentation will be performed based on considerations of main steam line breaks outside containment, it should be noted rapid MSIV isolction is corservatively assumed in the containment pressurization analyses in order to -

maximize drywell pressure which, in tum, maximizes the assumed containment leakage rate (Reference 12).

Review of the analytical models used for the LOCA analyses,(Reference 13), indicates }

MSIV isolation is not specifically considered because the worst-case postulated (steam line) break location would be inbocrd of the MSIVs (page 1-186 of Reference 13).

Review of the plant specific inputs to the most recent LOCA analyses, (References 8, 9, and 14), ccnfirms MSIV isolation +ime is not an input to the analyses.

b. High Energy Line Break (HELB) Con & ations The HELB analyses for JAF assur- wt break isolation is initiated by temperature serisors in the HPCI, RCIC, RWCU, ano main steam line (MSL) break protection logic.

, The HPCI, RCIC, and MSL circuits employ RTDs and the RWCU circuits utilize thermoccupies. In addition, as noted in Section 4.3.2 of Reference 34, a 2.5 second

" instrument delay time" was assumed for each HELB. As mentioned in GE document DRG A00-03658-1, sections 1 and 3.2.1, RTT has historically not been required for HELB isolation instrumentation. For example, for HPCI, RCIC, and RWCU isolation temperature detectors for Hope Creek and Perry (both STS BWR plants) are exempt fiom RTT. The nature of RTDs and thannoccuplos do not lend themselves to in-situ RTT This is consistent with the Standctd TeWnicci Specifications.

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ATTACMMENT 11 to JPN 92-030 Safety Evaluation for Proposed Changes to Technical Specifications RESPONSE TIME TESTING (JPTS-92-0_10)

3. Emergency Core Cooling System (ECCS)

Measurement of individual instrument response times for the ECCS is not required because the ECCS initiation sensors do not represent a significant portion of the particular system c.erall response time.

3.1 Core Spray (CS)

When the CS is initiated, the pump will be started within a time which includes the start-up 1: ..e of the emergency diesel generator (EDG) ar.d time for motor acceleration to full speed. The initiating instrument response time (drywell pressure) is insignificant when compared to the pump and EDG start times. Opening of the injection valve does not occur until the reactor ves ,el pressure decreases to 450 psig and the injection valve opening permissi- signal ;a received. The time necessary to reach the pressure permissive is varate depending on a specific LOCA blowdown : ate- The r"easurable response times for CS are taken from Reference 8 and given belew. Those response times provide a basis for acceptance criteria for existing surveillaxe iests. The maximum j assumed system related response times are extended in the most recent SAFER /GESTR-l LOCA analysis (Reference 9) as shown below in parentheses.

1

! Maximum Allowable Time Delay from 127 seconds initiating Signal to Pump at Rated Speed (including diesel generator start-up) M 30 seconds)

Injection Valve Stroke Time 110 seconds (115 seconds)

Response time surveillance for the CS will consist of showing compliance with these times during the specified surveillance interval.

3.2 Low Pressure Coolant injection (LPCl) System The sarne principles apply to the LPCI system as described above for CS. An overall system response time is again dependent on the specific LOCA condition, in this case, while the same reactor vessel permissive pressure applies to the LPCIinjection valve (450 psig), an additional reactor pressure permissive of less than 285 psig must be met to allow closure of the reactor recirculation pump discharge valve. The measurable response times for LPCI are also taken from Reference 8 and are given below. It follows that measurement of these times at the prescribed surveillance intervals will again provide the correct RTT for the LPCI system. The maximum assumed system related response tiroes are extended in the most recent SAFER /GESTR-LOCA analysis (Reference 9) as shown below in parentheses.

Maximum Allowable Time Delay from 122 seconds initiating Signal to Pump at Rated Speed (including diesel generator start-up) M 35 seconds)

Injection Valve Stroke Time 127 seconds M 35 seconds)

Recircuation Discharge Valve Stroke Time 133 seconds M 37 secorxis)

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ATTACHMENT 11 to JPN-92-030 Safety Evaluation for Proposed Changes to Technical Spacifications RESPONSE TIME TEST!NG (JPTS 92-010)

I 3.3 High Pressure Coolant injection (HPCI) System Unlike the two low pressure systems the HPCI system is not dependent on a reactor low pressure permissive signal to open the injection valve Consequently it is only necessary to measure the time from initiation to achievement of ated discharge flow delivered at the design pressure as t routine surveillance for RTT. Yhe system response time assumed in the Reference 8 analysis is given below. The maximum assumed system response times are extended in the most recent SAFER /GESTR LOCA analysis (Reference 9) as shown below in parentheses. Furthermore, credit is not taken for HPCI operation for worst case LOCA mitigation in References 8 and 9. Instrument channel response time is insignificant when compared to the allowed delay for start-up of the HPCI system. Accordingly itis not necessary to perform response time testing on the HPCI initiation instrumentation.

MaxLnum Allowable Time Delay from A 30 seconds initiating Signal to Rated Flow Available (160 seconds )

3.4 Automatic De-pressurization System l ADS) and Reactor Core Isolation Cooling (RCIC)

RTT is not applicable to Pe ADS or to the RCIC system. This is consistent with other existing BWR Technical Specifications and the Standard Technical Specifications. The transient and accidant analyses described in the FSAR do not take credit for instrument channel responst ' me because instrument response time is not significant when compared to the overall response time of these systems.

Instrumentation providing reactor pressure vessel water level 1 and level 3 inputs into the ADS logic can be exempted from RTT because it is in series with a nominal two minute timer, in addition, the level 3 input is merely a confirmatory input to prevent logic initiation in the event of a postulated failure of the instrument providing the level 1 input.

The ADS timers are verified by Technical Specifications to operate within a 120 i 5 second band (Reference 32). Reference 18 assumed a 125 second time delay, whereas Reference 20 considered a delay time of 140 seconde falong with a siuificant reduction in ECCS flow as well as longer LPCI and CS initiation times). The plant-specific analyses have shown a 15 second increase in ADS delay to have no signific*nt effect on peak clad temperature (PCT). By comparison, the response time of the ADS instrumentation channel is not significant and therefore response time testing is not required.

Credit is not taken for RCIC for LOCA mitigation in References 8,9 and 14. RCIC is also not credited for transient mitigation which relies solely on the RPS. Accordingly, RTT of instrumentation which initiates RCIC is not required.

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ATTACHMENT 11 to JPN-92-030

, Safety Evaluatian for Proposed Changes to Technical Specifications I

RESPONSE TIME TESTING (JPTS 92-010)

G. Response Time Testing Methodology The Authority will perform RTT for the ATTS sensors (Roseme"nt Transmitters) by verifying the transmitter time constant in accordance with the method descobed in main body of Reference 1, section 3.3.2 " Response Time." This method, which verifies the transmitter time constant, is different from the method described in EPRI report NP-267 " Sensor Response Time Verification" which is referenced on the last page (A-35/A-36) of the appendix of Reference 3.

Reference 1, includes the following as part of NRC request 032.19 on pages A-34 i, the appendix:

. Also , describe specific tests (i.e., calibration, response time, seal integrity, etcl performed as part of the surveillance test. "

Reference 3, provides the following response on Pages A-34 and A 35/A-36:

"The following tests are performed on each transmitter when it is removed from service for calibration:

(1) zero and span are set; (2) linearity of output current to input pressure is measured; (3) high and low gross failure trip points on the master trip unit are set; (4) transmitter seal integrity is observed when the transmitter is pressurized for adjustments of zero and span; and a (5) response time verification will be conducted according to the prncedures of EPRI report NP-267, " Sensor Response Time Verifical;on."

The first four items of the response are part of standard procedures for instrument calibration.

They have always been performed on all instrumentation of this type. Although item 3 is performed at the same frequency as the calibration, it is not performed as an integral part of the calibration.

Item 5 for response time verification differs from the time constant method described in the main [

body of Reference 3 in section 3.3.2. The Authority has chosen to sirify response time testing using the step change time constant method described section 3.3.2 Ler ause, when compared to the EPRI method referenced in item 5, the time constant verification method prov. des for easy setup, data collection, uses minimal test equipment, and is easier to conduct. It therefore significantly reduces the time technicians must remain in the reactor building and the corresponding personnel radiation exposure required to determine sensor response time. The reductions in time and radiation exposure resulting from the proposed time constant verification method, are supported by the actual field experierce during testirg conducted with the proposed methodology.

Informal conversations with other f acilities using the EPRI method indicate that the EPRI method requires significantly more technician time, results in correspondingly greater personnel radiation exposure and that it is difficult ta accurately reprodu;e 13st data due to the complexity of the equipment setup.

The method developed by the Authority to verify that the response time of ATTS instruments is within the proposed Technical Specification limits, as described in Attachment 111, is technically valid, reproducible and inherently conservative. This method has been reviewed by the instrument vendor (Rosemount). The vendor stated in Reference 10:

"The overall methodology, procedure and accompanying descriptions follow the same descriptions and methodology Rosemount would utilize to determine response times of pressure tr&nsmitters. "and * .themethodologyandprocedures outlinedin your transmittal are conc,1 AJas valid and correctin determining the time constants of Rosemountpressure transmi*ters.

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ATTACHMENT 11 to JPN-92-030

, Safety Evaluation for Proposed Changes to Technical Specificatione 4

RESPONSE TIME TESTING (JPTS 92-01Q]

IV. SIGNIFICANT HAZARDS CONSIDERATION Operation in accordance with the proposed amendment will not involve a significant hazards consideration as defined in 10CFR 50.92 because it will not:

1, involve a significant increase in the probability or consequences of an accident previously evaluated; The probability and consequences of previously evaluated accidents were based upon the RPS instrument channel and MSIV isolation actuation instrumentation meeting specified L reliability and response time standards. The inclusion of the complete channel in the measurement of response time increases assurance that instrument response time will be maintained within the limits assumed in the transient and accident analyses and will not increase the probability of occurrence of previously evaluated accidents.

2. create the possibility of a new or different kind of accident from any accident previously evaluated:

During the performance of 'ooth sensor calibration and RTT, the process instrument lines are isolated from the actual system. Because the sctual process system is isolated from the test signals, and because the isolation method is unchanged from existing procedures, the new test method will not create the possibility of a new or different kind of accident.

3. or involve a significant reduction in a margin of safety.

Testing of inntrumentation which was not previously subject to response time testing will not Jecrease the margin of safety. The response time limits for trip functions were increased to allow for inclusion all components in the instrumentation channel, including the ATTS components. However, the response time limits remains less than those assumed in the transient and accident analyses described in the FSAR.

V. IMPLEMENTATION OF THE PROPC L ED CHANGES j Implementation of the proposed changes will not affect the fire protection programs. The proposed changes will not impact the environment.

The new requirements to measure the response time of instrumentation will result in increased radiation exposure to personnel. The adoption of the time constant method proposed by the Authority will result in significantly lower personnel radiation exposure than would otherwise result from adoption of the methodology proposed by EPRI NP-267. The proposed increased time between requi ed channel tests will also reduce potential personnel radiation exposure.

VI. CONQLlGION These changes, as proposed, do not constitute an unreviewed %fety question as defined in 10CFR 50.59 becauta they will not:

i a. increase the probability of occurrence or the consequence,s of an accident or malfunction l of equipment importa:it to safety previously evaluated in the safety analysis report; 1

j b. create the possibility for an accident or 'nsifunction of a different type than any evaluated  ;

i previously in the safety analysis report-

c. reduce the margin of safety as defined in the basis for any technical specification.

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I ATTACHMENT 11 to JPN-92-030 l.

, Safety Evaluation for Proposed Changes to Technical Specifications RESPONSE TIME TESTING (JPTS 92 010)

Vll. REFERENCES

1. NRC NUREG-0123, Revision 3, "Standerd Technical Specifications for General Electric Boiling Water Rectors BWR/5)," issued Fall,1980.

2. NYPA DCM-11 review (Transmittal S92-27432-1 A)(JTS-92-0478 dated June 15,1992) of General Electric Nuclear Energy Regulatory and Analysis Services document DRG A00-03658-1 " Basis for Sensor Respor.se Time Testing (RTT) at the James A. FitzPatrick Nuclear Power Plant." _

3. General Electric Company Licensing Topical Report NEDO-21617 A, dated December 1978,

" Analog Transmitternrip Unit System for Engineered Safeguard Sensor Trip inputs."

Y 4. NRC letter, O. D. Parr to G.G. Sherwood (General Electric), dated June 27,1978 (MFN-279-78). Review of General Electric Topical Report NEDO-21617. " Analog Transmitter / Trip Unit System for Engineered Safeguard Sensor Trip input."

5. NYPA letter to the NRC, J.P. Bayne to D.B. Vassallo, (JPN 'iS-22) dated March 21,1985,

" Proposed Changes to the Technical Specifications Regarding Analog Transmitter Trip System (ATTS) (JPTS-85-04)."

6. NRC letter, H.l. Abieson to J.P. Bayne (NYPA), dated May 5,1985.

7. NRC Regulatory Guide 1.118, Revision 2, June 1978, Section C.S. " Periodic Testing of Electric Power and Protection Systems."

8. General Electric NEDC-31317P dated October 1986," James A. FitzPatrick Nuclear Power -

Plant, SAFER /GESTR LOCA Analysis."

9. General Electric NEDC-31317P, Revision 1, dated November 1991," James A. FitzPatrick Nuclear Power Plant SAFER /GESTR-LOCA Analysis.'

10. Rosemount letter to NYPA, T.J. Layer to J. Lazarus, dated May 29,1992,

11. James A. FitzPatrick Nuclear Power Plant, updated Final Safcty Analysis Report, Chapter 14.

12. Updated FSAR Section 14.6.1.3.3, " Primary Containment Response -Initial Conditions and Assumptions", Assumption "C", Volume 9, Page 14.6-12.

13. NEDO-20566A, " Analytical Model for Loss-of-Coolant Analysis in Accordance with 30 CFR 50, Appendix K - Volume 1", General E!ectric Co., September 1986.

14. MDE-83-0786, " Sensitivity of the James A. FitzPatrick Nuclear Power Plant Safety Systems Performance to Fundamental System Parameters", July 1986 (Refer to Table 3).

15. JAFNPP Technical Specification Table 3.2-2, item 14.

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