Text

Proposed Tech Specs Extending Functional Test Intervals & Allowable out-of-service Times,Removing APRM Downscale Scram Function & Response Time Values & Incorporating Miscellanous Changes
	ML20073L094
Person / Time
Site:	FitzPatrick
Issue date:	10/03/1994
From:	; POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	;
Shared Package
ML20073L065	List: JPN-94-050, Forwards Application for Amends to License DPR-59,proposing Changes to TS by Extending Functional Test Intervals & Allowable out-of-service Times,Removing APRM Downscale & Response Time Values & Incorporating Miscellaneous Changes; ML20073L077; ML20073L094;
References
	NUDOCS 9410120312
	Download: ML20073L094 (150)
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ATTACHMENT I to JPN-94-050 REVISED TECHNICAL SPECIFICATION PAGES FOR .

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New York Power Authority )

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59 I

9410120312 941003 ,

PDR ADOCK 0500033a p PDR

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9 LIST OF PAGE~ CHANGES P_toposedlechnicaLSpecification JPTS-90-010 Revise Appendix A as follows:

Remove.Pa9es itisertEages Be_ move Pages inserLEases i i 61 61 v v 62 62 vii vii 63 63 5 5 64 64 30a 30a 65 65 30d 30d New 65a 30e 30e 66 66 30f 30f 67 67 New 30g 68 68 31 31 69 69 32 32 70 70 36 36 70a Deleted 37 37 70b Deleted 38 38 70c Deleted 39 39 71 71 40 40 New 71a 41 41 72 72 41a Deleted 73 73 41b Deleted 74 74 42 42 76 76 43 43 77 77 43a 43a 77e Deleted 44 44 78 78 45 45 79 79 45a Deleted 80 80 48 48 81 8 49 49 82 82 50 50 84 84 55 55 85 85 56 56 87 87 57 57 285 285

JAFNPP IECHNLCAkSPECIElCAT10HS IABLERE.CDEIENIS Eage 1.0 Definitions 1 LIMITING SAFETY SAEEIY._LIMlIS S1SIEM_SETTlNGS 1.1 Fuel Cladding Integrity 2.1 7 1.2 Reactor Coolant System 2.2 27 SURVEILLANCE LIMITING CDRDITIONS FOR OPERAI1QN REQUIREMENTS 3.0 General 4.0 30 3.1 Reactor Protection System 4.1 30g l 3.2 Instrumentation 4.2 49 A. Primary Containment isolation Functions A 49 B. Core and Containment Cooling Systems - B 50 Initiation and Control C. Control Rod Block Actuation C 50 D. Radiation Monitoring Systems - Isolation D 50 and initiation Functions E. Drywell Leak Detection E 54 F. DELETED F 54 G. Recirculation 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 Limitations 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 B 115 l System i C. HPCI System C 117 l D. Automatic Depressurization System (ADS) D 119 l E. Reactor Core Isolation Cooling (RCIC) E 121 i System l

Amendment No. [, 1%.1[4,1)/3,1)l0 1

r JAFNPP ,

LIS10FIABLES Iable litte Eage 3.1-1 Reactor Protection System (Scram) Instrumentation Requirement 40 3.1 -2 (DELETED) 4.1-1 Reanor Protection System (Scram) mstrument Functional Tests 44 4.1-2 Reactor Protection System (Scram) instrurnent Calibration 46 3.2-1 Primary Containment isolation System Instrumentation Requirement 62 3.2-2 Core and Containment Cooling System initiation and Control 66 Instrumentation Operability Requirements 3.2-3 Control Rod Block Instrumentation Requirements 72 3.2-4 (DELETED) 3.2-5 Instrumentation that Monitors Leakage Detection inside the Drywell 75 3.2-6 (DELETED) 3.2-7 ATWS Recirculation Pump Trip instrumentation Requirements 76 3.2-8 Accident Monitoring Instrumentation 77a l 3.2-9 (DELETED) 3.2-10 Remote Shutdown Capability Instrumentation and Controls 77f 4.2-1 Primary Containment Isolation System Instrumentation Test and 78 Calibration Requirements 4.2-2 Core and Containment Cooling System Instrumentation Test and 80 Calibration Requirements 4.2-3 Control Rod Block Instrumentation Test and Calibration Requirements 82 4.2-4 (DELETED) 4.2-5 Minimum Test and Calibration Frequency for Drywell Leak Detection 83 4.2-6 (DELETED) 4.2-7 ATWS Recirculation Pump Trip instrumentation Test and Calibration 85 Requirements Amendment No.dJ , p,1)D,1)if1,1[p$,1)K), {/6

g-JAFNPP LIST OF FIGMES Eigules Iille Page 4.1-1 (Deleted) 4.2-1 (Deleted) 3.4-1 Sodium Pentaborate Solution (Minimum 34.7 B-10 Atom % Enriched) 110 Volume-Concentration Requirements 3.4-2 Saturation Temperature of Enriched Sodium Pentaborate Solution 111 3.5-1 Thermal Power and Core Flow Limits of Specifications 3.5.J.1,3.5.J.2 134 and 3.5.J.3 3.6-1 Reactor Vessel Pressure - Temperature Limits Through 12 EFPY 163 Part 1 3.6-1 Reactor Vessel Pressure - Temperature Limits Through 14 EFPY 163a Part 2 3.6-1 Reactor Vessel Pressure - Temperature Limits Through 16 EFPY 163b ;

Part 3 4.6-1 Chloride Stress Corrosion Test Results at 500 F 164 6.1-1 (Deleted) l 6.2-1 (Deleted) l l

Amendment No. /, %, 96,M J , %,4J , ps, ps,1J4,1/J,1/6,1/7,1/1,1/7,1/d, }g2 vii I

JAFNPP 1.0 (cont'd) opened to perform necessary operational activities. deficiency subject to regulatory review.

2. At least one door in each airiock is closed and sealed. S. Secondary _ Containment integrity - Secondary containment integrity means that the reactor building is intact and the following conditions

3. All automatic containment isolation valves are operable are met:

or de-activated in the isolated position.

1. At least one door in each access opening is closed.

4. All blind flanges and manways are closed.

2. Tne Standby Gas Treatment System is operable.

N. Rateo_eower - Rated power refers to operation at a reactor power of 2.436 MWt. This is also temied 100 percent power 3. All automatic ventilation system isolation valves are operable and is the maxir.um power level authorized by the operating or secured in the isolated position.

license. Rated team flow. rated coolant flow, rated nuclear systern pressur : refer to the values of these parameters T. Survaillance Frequencyllotalions / InteLvals when the reauor is at rated power.

The surveillance frequency notations / intervals used in these O. Beactor Power._ Operation - Reactor power operation is any specifications are defined as follows:

operation with the Mode Switch in the Startup/ Hot Standby or Run position with the reactor critical and above 1 percent Notations !rttervals Erenurn.lcy rated thermal power.

D Daily At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months P. Beactor Ves_sfl Pressure - Unless otherwise indicated. W Weekly At least once per 7 days reactor vessel pressures listed in the Technical M Monthly At least once per 31 days Specifications are those measured by the reactor vessel O Quarterly or At least once per 92 days steam space sensor. every 3 months SA Semiannually or At least once per 184 days O. Refueling _ Outage - Refueling outage is the period of time every 6 months between the shutdown of the urit prior to refueling and the A Annually or Yearly At least once per 366 days startup of the Plant subsequent to that refueling. R Note 1 At least once per 18 months (550 days)

R. Safely _Umits - The safety limits are limits within which the S/U Prior to each reactor startup reasonable maintenance of the fuel cladding integrity and the NA Not applicable reactor coolant system integrity are assured. Violation of such a limit is cause for unit shutdown and review by the Note 1: "Once each operating cycle," "once per operating cycle,"

l Nuclear Regulatory Commission before resumption of unit "each refueling outage " "at least once during each operating cycle."

operation. Operation beyond such a limit may not in itself "once each operating cycle not to exceed 18 months", or similar result in serious consequences but it indicates an operational phrases are equivalent to the definition for frequency notation "R" 5

Amendment No. [1/.1%

JAFNPP 3.0 Continued 4.0 Continued D. Entry into an OPERATIONAL CONDITION (mode) or other that a Survei!!ance Requirement has not been performed. The specified condition shall not be made when the conditions for the ACTION requirements may be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to Limiting Condition for Operation are not met and the associated permit the completion of the surveillance when the allowable ACTION requires a shutdown if they are not met within a outage time limits of the ACTION requirements are less than 24 specified time interval. Entry into an OPERATIONAL hours. Surveillance requirements do not have to be performed CONDITION (mode) or specified condition may be made in on inoperable equipment.

accordance with ACTION requirements when conformance to them permits continued operation of the facility for an unlimited D. Entry into an OPERATIONAL CONDITION (mode) shall not be period of time. This provision shall not prevent passage through made unless the Surveillance Requirement (s) associated with the OPERATIONAL CONDITIONS (modes) required to comply with Limiting Condition for Operation have been performed within the ACTION requirements. Exceptions to these requirements are applicable surveillance interval or as otherwise specified. This stated in the individual specifications. provision shall not prevent passage through or to Operational Modes as required to comply with ACTION Requirements.

E. When a system, subsystem, train. component or device is detennined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding normal or emergency power source is OPERABLE: and (2) all of its redundant system (s), subsystem (s), train (s), component (s) and device (s) are OPERABLE, or likewise satisfy the requirements of this specification. Unless both conditions (1) and (2) are satisfied. the unit shall be placed in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This specification is not applicable when in Cold Shutdown or Refuel Mode.

F. Equipment removed from service or declared inoperable to comply with required actions may be retumed to service under administrative control solely to perform testing required to demonstrate its operability or the operability of other equipment.

This is an exception to LCO 3.0.B.

Amendment No. % 1)f4,1)lf8 30a

JAFNPP 3.0 Bas _es - Continued F. LCO 3.0.F establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with required actions. The sole purpose of this Specification is to provide an exception to LCO 3.0.B to allow testing to demonstrate: (a) the operability of the equipment being retumed to service; or (b) the operability of other equipment.

The administrative controls ensure the time the equipment is retumed to service in conflict with the requirements of the required actions is limited to the time absolutely necessary to perform the allowed testing. This Specification does not provide time to perform any other preventive or corrective maintenance.

An example of demonstrating the operability of the equipment being retumed to service is reopening a containment isolation valve that has been closed to comply with the required actions and must be reopened to perfomi the testing.

An example of demonstrating the operability of other equipment is taking an inoperabie channel or trip system out of the tripped condition to prevent the trip function from occurring during the performance of testing on another channel in the other trip system. A similar example of demonstrating the operability of other equipment is taking an inoperable channel or trip system out of the tripped condition to permit the logic to function and indicate tne appropriate response during the performance of testing on another channel in the same trip system.

Amendment No. {l0.1/8 30d

. . = .

JAFNPP 4.0 BASES A. This specification provides that surveillance activities C. Continued necessary to insure the Limiting Conditions for Operation are met and will be performed during the OPERATIONAL interval, defined by the provisions of Specification 4.0.B. as a CONDITIONS (modes) for which the Limiting Conditions for condition that constitutes a failure to meet the OPERABILITY Operation are applicable. Provisions for additional surveillance requirements for a Limiting Condition for Operation. Under the activities to be performed without regard to the applicable - provisions of this specification, systems and components are.

OPERATIONAL CONDITIONS (modes) are provided in the assumed to be OPERABLE when Surveillance Requirements -

individual Surveillance Requirements. have been satisfactorily performed within the specified time interval. However, nothing in this provision is to be construed B. Specification 4.0.B establishes the limit for which the specified as implying that systems or components are OPERABLE when time interval for Surveillance Requirements may be extended, they are found or known to be inoperable although still-it permits an allowable extension of the normal surveillance meeting the Surveillance Requirements. This specification also ,

interval to facilitate surveillance scheduling and consideration - clarifies that the ACTION requirements are applicable when of plant operating conditions that may not be suitable for Surveillance Requirements have not been completed within the..

conducting the surveillance (e.g., transient conditions or other allowed surveillance interval and that the time limits of the ongoing surveillance or maintenance activities). It also , ACTION requirements apply from the point in time it is provides flexibility to accommodate the length of a fuel cycle identified that a surveillance has not been pe formed and'not .

for surveillances that are performed at each refueling outage at the time that the allowed surveillance was exceeded.

and are specified with an 18 month surveillance interval. It is Completion of the Surveillance Requirement within the not intended that this provision be used repeatedly as a - allowable outage time limits of the ACTION requirements convenience to extend surveillance intervals beyond that restores compliance with the requirements of Specification specified for surveillances that are not performed during - 4.0.C. However, this does not negate the fact that the failure refueling outages. The limitation of this specification is based to have performed the surveillance within the allowed on engineering judgement and the recognition that the most surveillance interval, defined by the provisions of Specification probable result of any particular surveillance being performed is .4.0.B, was a violation of the OPERABILITY requirements of a the verification of conformance with the Surveillance Limiting Condition for Operation tt'at is subject to enforcement Requirements. The limit on extension of the normal action. Further, the failure to perform a surveillance within the surveillance interval ensures that the reliability confirmed by provisions of Specification 4.0.B is a violation of a Technical -

surveillance activities is not significan 'aduced below that Specification requirement and is, therefore,'a reportable event obtained from the specified surveillance interval. under the requirements of 10 CFR 50.73(a)(2)(i)(B) because it is a condition prohibited by the plant Technical Specifications.

C. This specification establishes the failure to perform a Surveillance Requirement within the allowed surveillance Amendment No. p6,1/8,1p6 f 30e

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-. o 4

JAFNPP 4.0 BASES - Continued C. Continued C. Continued if the allowable outage time limits of the ACTION requirements Surveillance Requirements do not have to be performed on are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or a shutdown is required to comply inoperable equipment because the ACTION requirements define with ACTION requirements, a 24-hour allowance is provided to the remedial measures that apply. However, the Surveillance -

permit a delay in implementing the ACTION requirements. This Requirements have to be met to demonstrate that inoperable provides an adequate time limit to complete Surveillance equipment has been restored to OPERABLE status.

Requirements that have not been performed. The purpose of this allowance is to permit the completion of a surveillance before a shutdown is required to comply with ACTION D. This specification establishes the requirement that all requirements or before other remedial measures would be applicable surveillances must be met before entry into an required that may preclude completion of a surveillance. The OPERATIONAL CONDITION or other condition of operation basis for this allowance includes consideration for plant specified in the Applicability statement. The purpose of this -

conditions, adequate planning, availability of personnel, the specification is to ensure that system and component time required to perform the surveillance and the safety OPERABILITY requirements or parameter limits are met before significance of the delay in completing the required entry into an OPERATIONAL CONDITION or other specified surveillance. This provision also provides a time limit for the condition associated with plant shutdown as well as startup.

completion of Surveillance Requirements that become applicable as a consequence of OPERATIONAL CONDITION Under the provisions of this specification, the applicable (mode) changes imposed by ACTION requirements and for Surveillance Requirements must be performed within the completing Surveillance Requirements that are applicable when specified surveillance interval to ensure that the Limiting.

an exception to the requirements of Specification 4.0.C is Conditions for Operation are met during initial plant startup or allowed. If a surveillance is not completed within the 24-hour following a plant outage.

allowance, the time limits of the ACTION requirements are applicable at that time.' When a surveillance is performed When a shutdown is required to comply with ACTION t within the 24-hour allowance and the Surveillance requirements, the provisions of this specification do not apply.

Requirements are not met, the time limits of the ACTION ' because this would delay placing the facility in a lower requirements are applicable at the time the survcillance is CONDITION of operation.

terminated.

Amendment No. p4, jild, yd,1/3,1) ilk,1)6 30f

JAFNPP 3.1 LljyllTING_C_QNDLTIONS FOR OPERAILO.N 4.1 SURVELLLANCE REQUIREMF,NIS 3.1 BEACTOR PROTECTIOlLSYSTEM 4.1 BEACTOR PROTECTION SYSTEM 6pplicability; applicability; Applies to the instrumentation and associated devices which Applies to the surveillar.ce of the instrumentation and associated initiate the reactor scram. devices which initiate reactor scram.

ObjeClive; O_bjeclive; To assure the operability of the Reactor Protection System. To specify the type of frequency of surveillance to be applied to the protection instrumentation.

Specificatim Spacificatim l A. The setpoints and mir,imum number of instrument channels A. Instrumentation systems shall be functionally tested and per trip system that must be operable for each position of the calibrated as indicated in Tables 4.1-1 and 4.1-2 reactor mode switch, shall be as shown in Table 3.1-1. respectively.

The response time of the reactor protection system trip functions listed below shall be demonstrated to be within its limit at least once per 18 months. Neutron detectors are exempt from response time testing. Each test shall include at least one channel in each trip system. All channels in both trip systems shall be tested within two test intervals.

1. Reactor High Pressure (02-3PT-55A, B, C, D)

2. Drywell High Pressure (05PT-12A, B, C, D)

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

4. Main Steam Une Isolation Valve Closure (29PNS-80A2, B2, C2, D2)

(29PNS-86A2, B2, C2, D2)

5. Turbine Stop Valve Closure (94PNS-101,102,103.104)

6. Turbine Control Valve Fast Closure (94PS-200A, B, C, D)

7. APRM Fixed High Neutron Flux

8. APRM Flow Referenced Neutron Flux Amendment No.

30g l

JAFNPP 3.1 (cont'd) 4.1 (cont'd)

B. Minimum _ Critical Power Ratip_LMGPR) B. Maximum Fraction of Limiting Power Density (MFLPD)

During reactor power operation. the MCPR operating limit The MFLPD shall be determined daily during reactor power shall not be less than that shown in the Core Operating operation at 225% rated thermal power and the APRM high Limits Report. flux scram and Rod Block trip settings adjusted if necessary as specified in the Core Operating Limits Report.

1. During Reactor power operation with core flow less than 100% of rated, the MCPR operating limit shall be C. MCPR shall be determined daily during reactor power multiplied by the appropriate K, as specified in the Core operation at 225% of rated thermal power and following any Operating Limits Report. change in power level or distribution that would cause operation with a limiting control rod pattem as described in

2. If anytime during reactor operation at greater than 25% the bases for Specification 3.3.B.S.

of rated power it is determined that the operating limit MCPR is being exceeded, action shall then be initiated D. Verification of the MCPR operating limits shall be performed within fifteen (15) minutes to restore operation to within as specified in the Core Operating Umits Report.

the prescribed limits. If the MCPR is not retumed to within the prescribed limits within two (2) hours, an orderly reactor power reduction shall begin immediately.

The reactor power shall be reduced to less than 25% of rated power within the next four hours, or until the MCPR is retumed to within the prescribed limits.

Amendment No [4, 'd 7 6, 518, 98,199,1/7,1)&

31

JAFNPP 3.1 BASES A. The reactor protection system automatically initiates a reactor The outputs of the subchannels are combirred in a 1 out of 2 scram to: logic; i.e., an input signal on either one or both of the subchannels will cause a trip system trip. The outputs of the

1. Preserve the integrity of the fuel cladding. trip systems are arranged so that a trip on both systems is required to produce a reactor scram.

2. Preserve the integrity of the Reactc. Coolant System.

This system meets the intent of IEEE-279 (1971) for Nuclear

3. Minimize the energy which must be absorbed following a Power Plant Protection Systems. The syste.n has a reliability loss of coolant accident, and prevent inadvertent greater than that of a 2 out of 3 system and somewhat less criticality. than that of a 1 out of 2 system.

This specification provides the iimiting coeditions for operation With the exception of the average power range monitor necessary to preserve the ability of the sysum to perform its (APRM) channel the intermediate range monitor (IRM) intended function even during periods when eastrument channels, the scram discharge volume, the main steam channels may be out of service because of maticenance. isolation valve closure and the turbine stop valve closure, each When necessary, one channel may be made inoperable for subchannel has one instrument channel. When the minimum brief intervals to conduct required functional tests and condition for operation on the number of operable instrument calibrations. The basis for the allowable out-of-service times is channels per untripped protection trip system is met or if it provided in GE Topical Report NEDC-30851P-A, " Technical cannot be met and the affected protection trip system is Specification improvement Analysis for BWR Reactor placed in a tripped condition, the effectiveness of the Protection System," March 1988. protection system is preserved.

The Reactor Protection System is of the dual channel type Three APRM instrument channels are provided for each (Reference subsection 7.2 FSAR). The System is made up of protection trip system. APRM's A and E operate contacts in two independent trip systems, each having two subchannels one subchannel and APRM's C and E operate contacts in the of tripping devices. Each subchannel has an input from at other least one instrument channel which monitors a critical parameter.

1 Amendment No. }4, pd, ~)6, lyt) 32

JAFNPP 4.1 BASES l is not capable of going either up or down in response to an A. The channels listed in Tables 4.1-1 and 4.1-2 are divided into out-of-limits input. This type of failure for analog devices is a three groups for functional testing. These are: rare occurrence and is detectable by an operator who observes that one signal does not track other channel (s).

l Group A: On-off sensors that provide a scram trip function.

The bi-stable trip circuit which is a part of the Group (B) l Group B: Analog devices coupled with bi-stable trips that devices can sustain failures which are revealed only during provide a scram function. testing. Therefore, it is necessary to functionally test them periodically. A three month surveillance interval has been Group C: Devices which only serve a useful function during determined in accordance with NEDC-30851P-A, " Technical l

some restricted mode of operation, such as Specification improvement Analyses for BWR Reactor startup or shutdown, or for which the only practical Protection System."

test is one that can be performed at shutdown.

Group (C) devices are active only during certain modes of The sensors that make up Group (A) are on-off sensors. The operation. For example, the IRM is active during start-up probability of success is primarily a function of the sensor and inactive during full power operation. Thus the only test failure rate and the test interval. The basis for a three-month that is meaningful is the one performed just prior to shutdown functional test interval for group (A) sensors is provided in or start-up; i.e., the tests that are performed just prior to use NEDC-30851P-A, " Technical Specification improvement of the instrument.

Analysis for BWR Reactor Protection Systems."

Calibration frequency of the instrument channel is divided into Group (B) devices utilize an analog sensor coupled with a bi- two groups. These are as follows:

stable trip (either the solid-state analog transmitter trip system ,

(ATTS) or the more conventional arrangement of instrument 1. Passive type indicating devices that can be compared amplifier and bi-stable). The sensor and amplifier are active with like units on a continuous basis.

components and a failure is almost always accompanied by an alarm and an indication of the source of trouble. An as-is 2. Vacuum tube or semiconductor devices and detectors failure is one that sticks mid-scale and that drift or lose sensitivity.

Amendment No. )4 36

JAFNPP 4.1 13AS_ES (cont'd)

For the APRM System, drift of electronic apparatus is not the The measurement of response time provides assurance that only consideration in determining a calibration frequency. the Reactor Protection System trip functions are completed Change in power distribution and loss of chamber sensitivity within the time limits assumed in the transient and accident dictates a calibration every 7 days. Calibration on this analyses.

frequency assures plant operation at or below themial limits.

In terms of the transient analysis, the Standard Technical The frequency of calibration of the APRM flow biasing Specifications (NUREG-0123, Rev.3) define individual trip network has been established as each refueling outage. The function response time as "the time interval from when the.

flow biasing network is functionally tested at least once every monitored parameter exceeds its trip setpoint at the. channel .

three months and, in addition, cross calibration checks of the . sensor until de-energization of the scram pilot valve flow input to the flow biasing network can be made during the solenoids." The indivioual sensor response time defined as functional test by direct meter reading _ There are several " operating time" in General Electric (GE) design specification instruments which must be calibrated and it will take several data sheet 22A3083AJ,. note (8), is "the maximum allowable days to perform the calibration of the entire network. White time from when the variable being measured just exceeds the the calibration is being performed, a zero flow signal will be trip setpoint to opening of the trip channel sensor contact sent to half of the APRM's resulting in a half scram and rod during a transient." A transient is defined in note (4) of the block condition. Thus, if.the calibration were performed same data sheet as "the maximum expected rate of change during operation, flux shaping would not be possible. Based of the variable for the accident or the abnormal operating :

on experience at other generating stations,- drift of condition which is postulated in the safety analysis report.

instruments, such as those in the flow biasing network, is not significant and therefore, to avoid spurious . scrams, a calibration frequency.of each refueling outage is established.

Amendment No.

37

JAFNPP 4.1 BASES (cont'd)

The individual sensor response time may be measured by B. The MFLPD is checked once per day to determine if the simulating a step change of the particular parameter. This APRM scram requires adjustment. Only a small number of method provides a conservative value for the sensor response control rods are moved daily and thus the MFLPD is not time, and confirms that the instrument has retained its specified expected to change significantly and thus a daily check of the electromecha, sal characteristics. When sensor response time MFLPD is adequate.

is measured independently, it is necessary to also measure the remaining portion of the response time in the logic train up to The sensitivity of LPRM detectors decreases with exposure the time at which the scram pilot valve solenoids de-energize. to neutron flux at a slow and approximately constant rate.

The channel response time must include all component delays This is compensated for in the APRM system by calibrating in the response chain to the ATTS output relay plus the design twice a week using heat balance data and by calibrating allowance for RPS logic system response time. A response individual LPRM's every 1000 effective full power hours, time for the RPS logic relays in excess of the design allowance using TIP traverse data.

is acceptable provided the overall response time does not exceed the response time limits specified in the UFSAR. The basis for excluding the neutron detectors from response time testing is provided by NRC Regulatory Guide 1.118, Revision 2 section C.S.

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

Two instrument channels in Table 4.1-1 ha* e not been included in Table 4.1-2. These are: mode switch in shutdown and manual scram. All of the devices or sensors associated with these scram functions are simple on-off switches end, hence, calibration during operation is not applicable.

Amendment No.[jld,#i t ,1)d 38

4 4 JAFNPP THIS PAGE IS INTENTIONALLY BLANK Amendment pd,1)8 39

- - - - - _ - - - - - - - - - - - - - _ - - _ - - - - - - - - - - - - _ .-. L -- - - - - - --

JAFNPP TABLE 3.1-1 BEACIOR_P_RO_TE_CTION SYSTEh1 (SCR_AM) INST 8UMENTATION_REQUIREMEfRS Minimum No. of Mode in Which Function Operable Instrument Must be Operable Total Number of Channels Per Instrument Channels Trip System Refuel Startup Run Provided by Design (Notes 1 and 2) Trip Function Trip Level Setting (Note 7) for Both Trip Systems Action (Note 3) 1 Mode Switch in X X X 1 Mode Switch A Shutdown 1

1 Manual Scram X X X 2 A 3 IRM High Flux s 96 % (120/125) X X 8 A of full scale 3 IRM Inoperative X X 8 A 2 APRM Neutron Flux- s 15% Power X X 6 A '

Startup (Note 15) 2 APRM Flow Referenced (Note 12) X 6 A or B Neutron Flux (Not to exceed 117%) (Notes 13 and 14) 2 APRM Fixed High s 120% Power X 6 A or B Neutron Flux (Note 14) 2 APRM inoperative (Note 10) X X X 6 A or B Amendment No. /, }$,1/3 40

w 1 _

JAFNPP TABLE 3.1-1 (conrd)

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENTS Minimum No. of Mode in Which Function Operable Instrument Must be Operable Total Number of t Channels Per instrument Channels Trip System Refuel Startup Run Provided by Design ,

(Notes 1 and 2) Trip Function Trip Level Setting (Note 7) for Both Trip Systems Action (Note 3) 2 Reactor High Pressure s 1045 psig X X~ X 4 A (Note 9) 2 Drywell High Pressure s 2.7 psig X X X 4 A (Note 16) (Note 8) (Note 8) 2 Reactor Low Water - g 177 in. above TAF X X X 4 A Level (Note 16) 3 High Water Level in s 34.5 gallons per X X X 8 A

! Scram Discharge Volume Instrument Volume (Note 4) 4 Main Steam Line s 10% valve closure X 8 A lsolation Valve Closure (Note 6) 2 Turbine Control 500 < P < 850 psig X 4 A or C Valve Fast Closure Control oil pressure (Note 5) between fast closure solenoid and disc dump valve

, 4 Turbine Stop - s10% valve X 8 A or C Valve Closure closure (Notes 5 & 6)

Amendment No. )$, %,16, %, jd, %,'1/4,1)lf2 41

_ _ _ _ _ _ . ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . . _ _ . . _ . . _ . _ ~ - . _ _. - . _ . . _ _ _ . . , ~ _ . - _ _ . _ -

JAFNPP TABLE 3.1-1 (cont'd)

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENTS NOTES OF TAB _LE 3.1-1

1. There sha!! be two operable or tripped trip systems for each Trip Function, except as provided for below-

a. For each Trip Function with one less than the required minimum number of operable instrument channels, place the inoperable instrument channel and/or its associated trip system in the tripped condition

within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Otherwise, initiate the ACTION required by Table 3.1-1 for the Trip Function.

b. For each Trip Function with two or more channels less than the required minimum number of operable instrument channels:

1) Within one hour, verify sufficient instrument channels remain operable or tripped

to maintain trip capability in the Trip Function, and

2) Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , place the inoperable instnament channel (s) in one trip system and/or that trip system" in the tripped condition *, and

3) Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , restore the inoperable instrument channel (s) in the other trip system to an operable status, or place the inoperable instrument channel (s) in the trip system and/or that trip system in the tripped condition *.

If any of these three conditions cannot be satisfied, initiate the ACTION required by Table 3.1-1 for the affected Trip Function.

An inoperable instrument channel or trip system need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, if the inoperable hstrument channel is not restored to operable status within the required time, the ACTION required by Table 3.1-1 for that Trip Function shall be taken.

This action applies to that trip system with the greatest number of inoperable instrument channels. If both systems have the same number of inoperable instrument channels, the ACTION can be applied to either trip system.

2. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into associated Umiting Conditions For Operation and required actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Trip Function maintains RPS trip capability.

Amendment No. f$, pf, 1/2,1/4,1/2 42

. ,,5 I

JAFNPP TABLE 3.1-1 (cont'd)

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENTS NOTES OF TABLE 3.1-1 (cont'd)

3. Action Statements:

A. Insert all operable control rods within four hours.

B. Reduce power level to IRM range and place Mode Switch in the Startup position within eight hours.

C. Reduce power level to less than 30 percent of rated within four hours.

4. Permissible to bypass, if the Reactor Mode Switch is in the Refuel or Shutdown position.

5. Bypassed when turbine first stage pressure is less than 217 psig or less than 30 percent of rated power.

6. The design permits closure of any two lines without a scram being initiated.

7. When the reactor is subcritical and the reactor water temperature is less than 212 F, only the following trip functions need to be operable:

A. Mode Switch in Shutdown. "

B. Manual Bram. .

C. High Flux IRM .

D. Scram Discharge Volume High Level when any control rod in a control cell containing fuel is not fully inserted.

E. APRM 15% Power Trip.

l 8. Not required to be operable when primary containment integrity is not required.

l

9. Not required to be operable when the reactor pressure vessel head is not bolted to the vessel.

10. An APRM will be considered operable if there are at least 2 LPRM inputs per level and at least 11 LPRM inputs of the normal -

complement.

l 11. (Deleted)

Amendment No. f, QQ, l14, j#,98, %, %,1pD,1/7,1/J,1[,2/f7

-43

JAFNPP TABLE 3.1-1 (contd)

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENTS

12. The APRM Flow Referenced Neutron Flux Scram setting shall be less than or equal to the limit specified in the Core Operating Limits Report.

13. The Average Power Range Monitor scram function is varied as a function of recirculation flow (W). The trip setting of this function must be maintained as specified in the Core Operating Limits Report.

14. " 3 APRM flow biased high neutron flux signal is fed through a time constant circuit of approximately 6 seconds. The APRM fixed h.si neutron flux signal does not incorporate the time constant, but responds directly to instantaneous neutron flux.

15. This Average Power Range Monitor scram function is fixed point and is increased when the reactor mode switch is placed in the Run position.

16. Instrumentation common to PCIS.

Amendment No. ijd 43a

JAFNPP TABLE 4.1-1 BEACT_OR PROTECIlOXSYSTEM_{ SCRAM)1NSTRUMENTAILON TEST REQU BEMEN_TS Functional Test Trip Function Group (Note 2) Functional Test Frequency (Note 3) Instrument Check Mode Switch in Shutdown A Place Mode Switch in Shutdown R NA Manual Scram A Trip Channel and Alarm O NA RPS Channel Test Switch A Trip Channel and Alarm W (Note 1) NA IRM High Flux C Tnp Channel and Alarm (Note 4) S/U and W (Note 5) NA IRM inoperative C Trip Channel and Alarm (Note 4) S/U and W (Note 5) NA APRM High Flux B Trip Output Relays (Note 4) O NA Inoperative B Trip Output Relays (Note 4) O NA Flow Biased High Flux B Trip Output Relays (Note 4) O NA High Flux in Startup or Refuel C Trip Output Relays (Note 4) S/U and W (Note 5) NA Reactor High Pressure B Trip Channel and Alarm (Note 4) O D Drywell High Pressure B Trip Channel and Alarm (Note 4) O D Reactor Low Level B Trip Channel and Alarm (Note 4) O D High Water Level in Scram A Trip Channel O (Note 6) NA Discharge Instrument Volume High Water Level in Scram B Trip Channel and Alarm (Note 4) O D Discharge Instrument Volume Amendment No. SJ , jd,78. Ed 1/d 44

JAFNPP TABLE 4.1-1 (Cont'd)

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION TEST REQU. IREMENTS Functional Test Trip Function Group (Note 2) Functional Test Frequency (Note 3) Instrument Check Main Steam Une Isolation Valve Closure A Trip Channel and Alarm O NA Turbine Control Valve Fast Closure A Trip Channel and Alarm O NA Turbine First Stage Pressure Permissive B Trip Channel and Alarm (Note 4) O D Tu bine Stop Valve Closure A Trip Channel and Alarm O NA NOTES FOR TABLE 4.1-1

1. The automatic scram contactors shall be exercised once every week by either using the RPS channel test switches or performing a functional test of any automatic scram function. If the contactors are exercised using a functional test of a scram function, the weekly test using the RPS channel test switch is considered satisfied. The automatic scram contactors shall also be exercised after maintenance on the contactors.

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

3. Functional tests are not required on the part of the system that is not required to be operable or are tripped. If tests are missed on parts not required to be operable or are tripped, then they shall be performed prior to retuming the system to an operable status.

4. This instrumentation is exempted from the instrument channel test definition. This instrument channel functional test will consist of injecting a simulated electrical signal into the instrument channels.

5. Weekly functional test required only during refuel and startup mode.

l

! 6. The functional test shall be performed utilizing a water column or similar device to provide assurance that damage to a float l or other portions of the float assembly will be detected.

Amendment No. /, ps, Q$ ,1%, pd7 1

K N

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L L

A N

P O P I T

N N 8 F E 4

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JAFNPP 3.2 LLMITINGXQNDITIONS FOB _QP_ERATIQN 4.2 SUEVElLLANCE REQU1BEMENTS 3.2 1.NSITBUMENTATLON _ 4.2 INSTRUMENTATIQN 6pplicability- 6pplicabiEty:

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

QbleClive: Qbiectiv3:

To assure the operability of the aforementioned instrumentation. To specify the type and frequency of surveillance to be applied to the aforementioned instrumentation.

Sp_ecificatioCS: Specifications:

A. Primary _Canlainment Isolation Funct!Ons A. Prima.ry_CDDiainment Isolation Fun 31 ions When primary containment integrity is required, the limiting instrumentation shall be functionally tested and calibrated as conditions of operation for the instrumentation that initiates indicated in Table 4.2-1. System logic shall be functionally primary containment isolation are given in Table 3.2-1. tested as indicated in Table 4.2-1.

The response time of the main steam isolation valve actuation instrumentation isolation trip functions listed below shall be demonstrated to be within their limits at least once per 18 months. Each test shall include at least one channel in each trip system. All channels in both trip systems shall be tested within two test intervals.

1. MSIV Closure - Reactor Low Water Le' vel (L1)

(02-3LT-57A,B and 02-3LT-58A.8)

2. MSIV Closure - Low Steam Une Pressure (02PT-134A.8,C,D)

3. MSIV Closure - High Steam Une Flow (02 D PT- 1 16 A-D, 1 17A-D, 1 18 A-D, 1 19 A-D)

Amendment No.1/0,113 /

49

JAFNPP 3.2 (cont'd) 4.2 (cont'd)

B. Core and Containment Coolina Systems - Initiation and B. Core and Containment Cooling Systems - Initiation and Contgg Control The limiting conditions for operation for the instrumentation Instrumentation shall be functionally tested, calibrated, and that initiates or controls the Core and Containment Cooling checked as indicated in Table 4.2-2.

Systems are given in Table 3.2-2. This instrumentation must be operable when the system (s) it initiates or controls are System logic shall be functionally tested as indicated in required to be operable as specified in Specification 3.5. Table 4.2-2.

C. Control Rod Block Actuation C. Control Rod Block Actuation The limiting conditions of operation fo.- the instrumentation that initiates control rod block are given in Table 3.2-3. Instnamentation shall be functionally tested, calibrated, and checked as indicated in Table 4.2-3.

System logic shall be functionally tested as indicated in Table 4.2-3.

D. Radiation Monitorina Systems - Isolation and initiation

~ ~

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

Amendment No. M 1)8 50

JAFNPP 3.2 BASES Besides reactor protection instrumentatron which initiates a reactor The instrumentation which initiates primary containment isolation is scram. additional protective instrumentaton is also provided. This connected in a dual bus (two trip systems) arrangement. Main protective instrumentation initiates action to mitigate the Steam Line isolation Valve (MSIV) isolation utilizes a one-out-of-two-consequences of accidents which are beyond the operator's ability to taken-twice logic arrangement which closes the four inboard and four control or terminates operator errors before they result in serious outboard MSIVs. Other penetrations which have both inboard and consequences. This set of specifications provides the limiting outboard automatic isolation valves (except for the primary conditions of operation for the primary system isolation function, containment hydrogen and oxygen concentration sample, and the initiation of the Core Cooling Systems. Control Rod Block and gaseous and particulate radioactivity sample systems) utilize logic Standby Gas Treatment Systems. The objectives of the arrangements in which one trip system closes inboard isolation specifications are to assure the effectiveness of the protective valves and the other trip system closes outboard isolation valves.

instrumentation when required, even during periods when portions of The primary containment hydrogen and oxygen concentration such systems are out of service for maintenance and to prescribe sample supply and return lines, as well as the gaseous and the trip settings required to assure adequate performance. When particulate sample supply and retum lines, utilize inboard and necessary, one channel may be made inoperable for brief intervals outboard isolation valves that are both closed by a sing!e trip to conduct required functional tests and calibrations. system. Hydrogen and oxygen concentration sample supply and retum isolation valve control circuits are provided with the capability Some of the settings on the instrumentation that initiate or control to override automatic isolation to allow sampling during and following core and containment cooling have tolerances explicitly stated where an accident. Penetrations which are isolated by a single automatic the high and low values are both critical and may have a substantial isolation valve (and a remote manual or check valve) utilize a single effect on safety. The set points of other instrumentation, where only trip system to effect closure of the automatic isolation valve.

the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent The low water level instrumentation set to trip at 177 in. above the inadvertent actuation c! the safety system involved and exposure to top of the active fuel closes all isolation valves except those in abnormal situations. Group 1. Details of the isolation valve grouping are given in Section 7.3 of the updated FSAR. For valves which isolate at this level, this Actuation of primary containment valves is initiated by protective trip setting is adequate to prevent uncovering the core in the case of instrumentation shown in Table 3.2-1 which senses the conditions a break in the largest line.

for which isolation is required. Such instrumentation must be available whenever primary containment integrity is required. The low-low reactor water level instrumentation is set to trip when reactor water level is 126.5 in. above the top of active fuel. This trip Amendment No. '02.'10,?,202 55

JAFNPP 3.2 BASES (cont'd) initiates the HPCI and RCIC systems and trips the recirculation Venturis are provided in the main steam lines as a means of pumps. The low-low-low reactor water level instrumentation is set measuring steam flow and also limiting the loss of mass inventory to trip when the water level is 18 in. above the top of active fuel. from the vessel during a steam line break accident. The primary This trip activates the remainder of the ECCS subsystems, closes function of the instrumentation is to detect a break in the main steam the main steam isolation valves, main steam line drain valves and line. For the worst case accident, main steam line break outside the reactor water sample line isolation valves, and starts the drywell, a trip setting of 140 percent of rated steam flow in conjunction emergency diesel generators. These trip level settings were with the flow limiters and main steam line valve closure, limits the chosen to be high enough to prevent spurious actuation but Iow mass inventory loss such that fuel is not uncovered. fuel temperature enough to initiate ECCS operation and primary system isolation so peak at approximately 1,000 F and release of radioactivity to the that post-accident cooling can be accomplished and the guidelines environs is below 10 CFR 100 guidelines. Reference Section 14.6.5 of 10 CFR 100 will not be exceeded. For large breaks up to the of the updated FSAR.

complete circumferential break of a 24 in. recirculation line and with the trip setting given above. ECCS initiation and primary The main steam line high temperature isolation function utilizes 16 system isolation are initiated in time to meet the above criteria. sensors (instrument channels), with 4 sensors located at each of 4 Reference paragraph 6.5.3.1 of the updated FSAR. different areas in the vicinity of the main steam lines. The 4 instrument channels associated with each of the 4 areas are arranged The high drywell pressure instrumentation is a diverse signal for in a 1-out-of-2-taken-twice logic. Thus a main steam line break in any malfunctions to the water level instrumentation and in addition to of the 4 areas will effect closure of all 8 main stearn line isolation initiating ECCS, it causes isolation of Groups B and C isolation valves.

valves. For the breaks discussed above, this instrumentation will generally initiate ECCS operation before the low-low-low water

!evel instrumentation; thus the results given above are applicable here also. Details of the isolation valve closure group are given in Section 7.3 of the updated FSAR. The water level instrumentation initiates protection for the full spectrum of loss-of-coolant accidents.

I Amendment No. 22,18,52,102.'72,20?

56 -

JAFNPP I

3.2 BASES (cont'd)

High radiation monitors in the area of the main steam lines The RCIC high flow and temperature instrumentation are ]

have been provided to detect gross fuel failure as in the arranged the same as that for the HPCI. The trip settings of I control rod drop accident. A trip setting of 3 times normal approximately 300 percent for high flow or 40*F above full-power background is established to close the main steam maximum ambient for temperature are based on the same line drain valves, the recirculation loop sample valves, the criteria as the HPCI.

mechanical vacuum pump isolation valves, and trip the pumps, i to limit fission product release. For changes in the Hydrogen The HPCI high temperature isolation function utilizes 16 sensors Water Chemistry hydrogen injection rate, the trip setpoint may (instrument channels) located in the vicinity of the HPCI be adjusted based on a calculated value of the expected equipment and piping. The 16 instrument channels provide radiation level. Hydrogen addition will result in an increase in inputs into two trip systems, eight instrument channels per trip the N-16 carryover in the main steam. system. One trip system is associated with the inboard isolation valve and the other trip system is associated with the outboard Pressure instrumentation is provided to close the main steam isolation valves. Trip logic for each trip system is one-out-of-isolation valves in the run mode when the main steam line eight-taken-once logic for the high temperature isolation pressure drops below 825 psig. The reactor pressure vessel function. The logic for the RCIC high temperature isolation thermal transient duo to an inadvertent opening of the turbine function is the same as the HPCI logic, except 8 instrument bypass valves when not in the run mode is less severe than channels,4 per trip system provide input to the high the loss of feedwater analyzed in Section 14.5 of the FSAR, temperature isolation logic circuits.

therefore, closure of the main steam isolation valves for thermal transient protection when not in the run mode is not The reactor water cleanup system high temperature required. instrumentation are arranged similar to that for the HPCI. The trip settings are such that uncovering the core is prevented and The HPCI high flow and temperature instrumentation are fission product release is within limits.

provided to detect a break in the HPCI steam piping. Tripping of this instrumentation results in actuation of HPCI isolation The instrumentation which initiates ECCS action is arranged in a valves. Tripping logic for the high flow is a 1 out of 2 logic. dual bus system. As for other vital instrumentation arranged in The trip settings of approximately 300 percent of design flow this fashion, the specification preserves the effectiveness of the for this high flow or 40*F above maximum ambient for high system even during periods when maintenance or testing is temperature are such that uncovering the core is prevented being performed. An exception to this is when logic functional and fission product release is within limits. testing is being performed.

The cont ul rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to the Safety Limit. The trip Amendment No. [,si6,%,%,pd,1/4,1/7,2[7 57

n JAFNPP 4.2 BA_SES ,

The instrumentation listed in Tables 4.2-1 through 4.2-8 will be 5. GE Topical Report NEDC-30936P-A, Parts 1 and 2, "BWR functionally tested and calibrated at regularly scheduled Owners Group Technica: Specification Improvement -l l intervals. The same design reliability goal as the Reactor Methodology (With Demonstration for BWR ECCS Protection System is generally applied. Sensors, trip devices Actuation instnamentation)," December 1988.

and power supplies are tested, calibrated and checked at the same frequency as comparable devices in the Reactor Protection 6. GE Topical Report GENE-770-06-1-A, " Bases for Changes System. to Surveillance Test Intervals and Allowed Out-Of-Service Times For Selected instrumentation Technica!

The sunteillance test interval for the instrumentation channel Specifications," December 1992.

functional tests are once/three months for most instrumentation.

, This surveillance intervalis based on the fol!owing NRC 7. GE Topical Report GENE-770-06-2-A, " Addendum to Bases f approved licensing topical reports: for Changes to Surveillance Test intervals and Allowed Out - !

Of-Service Times For Selected Instrumentation Technical -

1. GE Topical Report NEDC-30851P-A, " Technical Specifications," December 1992.

Specification improvement Analysis for BWR Reactor l

Protection System," March 1988.

The measurement of the response time interval for the Main

=

i

2. GE Topical Report NEDC-30851P-A, Supplement 1 Steam isolation Valve (MSIV) isolation actuation instrumentation !

" Technical Specification improvement Analysis for BWR begins when the monitored parameter exceeds the isolation Control Rod Block Instrumentation," October 1988. actuation setpoint at the channel sensor and ends when the ;

MSIV pilot solenoid relay contacts open. With the exception of ,

GE Topical Report NEDC-30851P-A, Supplement 2- the MSIVs, response time testing is not required for any other 3.

" Technical Specification improvement Analysis for BWR primary containment isolation actuation instrumentation. The :

Isolation Instrumentation Common'to RPS and ECCS safety analyses results are not sensitive to individual sensor Instrumentation," July 1986. response times of the logic systems to which the sensors are connected for isolation actuation instrumentation

4. GE Topical Report NEDC-31677P-A, " Technical Specification Improvement Analysis for BWR lsolation Actuation Instrumentation, " July 1990.

Amendment No. jl$,1/4,1/1,1)fd _

61- !

- - , ---.~v .-< . . - - . ~ . . ,,-,,-,v-e,,.-n +- - -- , . - . < - - - c~u - -- .+----r, .-r -n-.,- r. . . - .+ ~ n.nn,~ ~<.n- - -,+- ,-

JAFNPP TABLE 3.2-1 EBIMARY CONTAINMENT ISOLATION SYSTEM INSTRUMENTATION REQUIREMENTS Minimum No. of Operable Instrument Total Number of Channels Per Instrument Channels Trip System Provided by Design (Notes 1 and 2) Trip Function Trip Level Setting for Both Trip Systems Action (Note 3) 2 Reactor Low Water Level (Notes 4.& 7) a 177 in. above TAF 4 A 2 Reactor Low Water Level (Notes 7 & 8) a 177 in. above TAF 2 A 1 Reactor High Pressure s 75 psig 2 D (Shutdown Cooling Isciation)

2 Reactor Low-Low-Low Water Level 218 in. above the TAF 4 A 2 Drywell High Pressure (Notes 4 & 7) s 2.7 psig 4 A 2 Drywell High Pressure (Notes 7 & 8) s 2.7 psig 2 A 2 Main Steam Une Tunnel s 3 x Normal Rated 4 E .

High Radiation Full Power Background 2 Main Steam Une Low Pressure 2 825 psig 4 B (Note 5) 2 Main Steam Une High Flow s 140% of Rated Steam Flow - 4 G 8 Main Steam Une Leak s 40 F above max ambient 16 B Detection High Temperature 4 Reactor Water Cleanup System s 40 F above max ambient 8 C Equipment Area High Temperature 2 Condenser Low Vacuum (Note 6) 2 8" Hg. Vac 4 8 i

Amendment No. 62

".. ~,

1 JAFNPP IABLE 3.2-1 (Cont'd)

PRIMARY _ CONTAINMENT ISOLAI!ON_SXSIEM_LNSIEUMENTATION REQU1BEMENTS Minimum No. of Operable instrument Total Number of i Channels Per instrument Channels Trip System Provided by Design (Note 1 aN 2) Trip Function Trip Level Setting for Both Trip Systems Action (Note 3) 1 HPCI Turbine Steam < 160 in H2O dp 2 F Une High Flow 1 -HPCI Steam Line 100 > P > 50 psig 2-. F- ;

Low Pressure 1 HPCI Turbine High s 10 psig 2 F Exhaust Diaphragm Pressure 8 HPCI Steam Line/ s 40'F above '16 F Area Temperature max. ambient i

1 RCIC Turbine Steam s 282 in H2O dp 2 F .

Une High Flow 1 RCIC Steam 'Une 100 > P > 50 psig 2 .F Low Pressure s

1 'RCIC Turbine High s 10 psig 2 F !

~

Exhaust Diaphragm Pressure .1

! 4 RCIC Steam Une/. s 40"F above 8 F

Area Temperature max. ambient Amendment No./, Q -$'

63' ,

L___=.___________..

_ ____u_____._..._.. _.__ ._ .. _ . _ . . _ . ..c. . . , . . _ . . . _ . . . , . . . _ . _.. .__ _ ._ _ __ _. . , _ _._

JAFNPP

- TABLE 3.2-1 (Cont'd)

!. PRIMARY CONTAINMENT ISOLATION SYSTEM INSTRUMENTATION REQUIREMENTS NQIES FOR TABLE 3.2-1 I

1. Whenever Primary Containment integrity is required by Specification 3.7.A.2, there shall be two operable or tripped trip systems for

<- each Trip Function, except as provided for below-

a. For each Trip Function with one less than the required minimum number of operable instrument channels, place the inoperable instrument channel and/or its associated trip system in the tripped condition

within:

1) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for trip functions common to RPS instrumentation, and

2) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for trip functions not common to RPS instrumentation, i

or, initiate the ACTION required by Table 3.2-1 for the affected trip function.

b. For each Trip Function with two or more channels less than the required minimum number of operable instrument channels: !

1) Within one hour, verify sufficient instrument channels remain operable or tripped

to maintain trip capability in the Trip Function, and

2) Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , place the inoperable instrument channel (s) in one trip system and/or that trip system ** in the tripped condition *,.

and

3) Restore the inoperable instmment channel (s) in the other trip system to an operable status, or place the inoperable instrument channel (s) in the trip system and/or that trip system in the tripped condition

within:

(a) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for trip functions common to RPS instrumentation, and j

(b) . 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for trip functions not common to RPS instrumentation.

If any of these three conditions cannot be satisfied, initiate the ACTION required by Table 3.2-1 fo'c the affected Trip Function.

~ Asterisk shown on next page -

Amendment No. %, [, d p,96 ,1/3, Ifb,1/2,1gB,2/3, S d7-64 .

_ .- . . . . ~ . - . . . . . . . - . . - = . . . ....-. ,.. - . . . . . . . . . - . . , . . .

' ~

v .gi JAFNPP IARLE 3.2-1 (Cont'd)

PRIMARY CONTAINMENT ISOLATION SYSTEM INSTRUMENTATION REQUIREMENTS NOTES FOR TABLE 3.2-1 (cont'd)

An inoperable instrument channel or trip system need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, if the inoperable instrument channel is not restored to operable status within the required time, the ACTION required by Table 3.2-1 for that Trip Function shall be taken.

i

" This action applies to that trip system with the greatest number of inoperable instrument channels. If both systems have the same number of inoperable instrument channels, the ACTION can be applied to either trip system.

2. When a channel, and/or the affected primary containment isolation valve, is placed in an inoperable status solely for performance of required instrumentation surveillances, entrv into associated Umiting Conditions for Operation and required actions may be delayed as follows:

a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Trip Functions utilizing a two-out-of-two-taken-once logic; or b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for the remaining Trip Functions provided the associated Trip Function maintains PCIS initiation capability for at least one containment isolation valve in the affected penetration.

3. Actions:

A. Piace the reactor in the colt ,;ndition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B. Isolate the main steam lines within eight hours.

C. Isolate Reactor Water Cleanup System within four hours.

D. Isolate shutdown cooling within four hours.

E. Isolate the main steam !ine drain valves, the recirculation loop sample valves, and the mechanical vacuum pump, within eight hours.

F. ' isolate the affected penetration flow path (s) within one hour and declare the affected system inoperable.

G. Isolate the affected main steam line within eight hours.

Amendment No./,4 8 , jf/, !)6,1jlk,1/J,1)/2,387 7

65 i

. _ , - - . < ~ . . . - . _ . - - . _-. .. - - . , . . .. .- - - - . . - - ~ . . . . - , . . . , . - _ _ . , - - - . . - . _ . _ _ . . _ _ _ _ . _ _ _ . , _

JAFNPP..

TABLE 3.2-1 (Cont'd)

PRIMARY CONTAINMENT ISOLATION SYSTEM INSTRUMENTATION REQUIREMENTS NOTES FOR TABLE 3.2-1 (cont'd)

4. These signals also stait SGTS and initiate secondary containment isolation.

5. Only required in run mode (interiocked with Mode Switch).

6. Only required in the run mode and turbine stop valves are open.

7. Instrumentation common to RPS. ,

8. Trip Function utilizes a two-out-of-two-taken-once logic for isolation of both primary containment isolation valves on the hydrogen and 1 oxygen sample, and gaseous and particulate sample supply and return lines.

t' Amendment No.

65a !

JAFNPP TABLE 3.2-2 CORE AND CONTAINMENT COOLING SYSTEM INITIATION A3D_

CONTROL INSTRUMENTATION OPERABILITY REQUIREMENIS Minimum No. of Operable Instrument Total Number of Channels Per Instrument Channels item Trip System Provided by_ Desiga N o. (Notes 1 and 2) Trio Function Trio Level Settino for Both Trio Systems Remarks 1 2 Reactor Low-Low 2 126.5 in. above TAF 4 (HPCI & RCIC) Initiates HPCI, RCIC, and +

Water Level SGTS.

2 2 Reactor Low-Low- 2 18 in. above TAF 4 (Core Spray & RHR) Initiates Core Spray, RHR (LPCI),

Low Water Level- and Emergency Diesel Generators.

l 4 (ADS) Initiates ADS (if not inhibited by ADS override switches), in conjunction with Confirmatory Low Level,120 second delay and RHR (LPCI) or Core Spray pump discharge pressure interlock.

3 2 Reactor High Water s 222.5 in. above TAF 2 (Note 8) Trips HPCI turbine.

Level 4 2 Reactor High Water s 222.5 in. above TAF 2 (Note 8) Closes RCIC steam suppply Level valve.

5 1 (Note 9) Reactor Low Level 2 0 in. above TAF 2 Prevents inadvertent (inside shroud) operation of containment spray during accident condition. t 6 2 Containment High 1 < p < 2.7 psig 4 Prevents inadvertent operation Pressure of containment spray during accident condition.

i Amendment No. }& OJ , jf/, %,1/9 66

JAFNPP IABLE 3.2-2 (Contd)

CORE AND CONTAINMF,NT_CO_OLING_S_YST_EM IN!TIATIQN AND CONTROL _INSTRU_MENTA_TIOfLO_P_ER_AB_lLITY REQUIREMENTS Minimurn No. of Operable Instrument Total Number of Channels Per instrument Channels item Trip System Provided by Design No. (N21esland 2) Trio Fun _ction Trip _ Level _ Salting for Both Trio Systems Remarks 7 1 (Note 9) Reactor Low Level 2177 in. above TAF 2 Confirmatory low water level for ADS actuation.

8 2 Drywell High s 2.7 psig 4 initiates Core Spray, RHR Pressure (LPCI), HPCI and SGTS.

9 2 Reactor Low Pressure 2 450 psig 4 Permits opening Core Spray and RHR (LPCI) injection valves.

10 1 (Note 9) Reactor Low Pressure 50 s p s 75 psig 2 Permits closure of RHR (LPCI) injection valves while in shutdown cooling in conjunction with PCIS signal.

11 1 Core Spray Pump 11 0.6 sec. 1 (Note 8) Initiates starting of (Notes 3 & 9) Start Timer core spray pump.

(each loop) (each loop) 12 1 RHR (LPCI) Pump (Notes 3 & 9) Start Timer 1st Pump (A Loop) 1.0 + 0.5 (-) O sec. 1 (Note 8) Starts 1st Pump (A Loop) 1st Pump (B Loop) 1.0 + 0.5 (-) O sec. 1 (Note 8) Starts 1st Pump (B Loop) 2nd Pump (A Loop) 6.0 0.5 sec. 1 (Note 8) Starts 2nd Pump (A Loop) 2nd Pump (B Loop) 6.0 0.5 sec. 1 (Note 8) Starts 2nd Pump (B Loop) l Amendment No. jd, $$ , F/,94,1/J

67 l

. a -

Y JAFNPP TABLE 3.2-2 fCont'd)

CORE AND CONTAINMENT COOLING SYSTEM INITIATION AND CONTROL INSTRUMENTATION OPERABILITY REQUIREMENTS Minimum No. of Operable Instrument Total Number of Channets Per Instrument Chanrels item Trip System Provided by Design No. (Notes 1 and 2) Trio Function Trio Level Settina for Both Trio Svstems Remarks 13 1 (Note 9) Auto Blowdown Timer 120 sec. 5 sec. 2 initiates ADS (if not inhibited by ADS override switches),

. 14 4 RHR (LPCI) Pump 125 psig 20 psig 8 Permits ADS actuation.-

Discharge Pressure +

Interlock 15 2 Core Spray Pump 100 psig 10 psig 4 Permits ADS actuation.

Discharge Pressure Interlock 16 * (Note 9) RHR (LPCI) Trip Loss of Voltage 2 Monitors availability System Bus Power of power to logic systems.

Monitor 17 1 (Note 9) Core Spray Trip Loss of Voltage 2 - Monitors availability.

System Bus Power of power _to logic systems.

Monitor .

Amendment No. /, %, %,1/4 f

'G8

___._______.m_______m . . ~. . -_.._..,...,,~c- . . , - . - _ _ _ . s, - ,-- ,,. ,-...-._..m. . - , - , m . - _.--_,.. _ __ % ...-__ ,, . .._m. . . . , ~ , _ . . ..._.. .~m

.?

1

'JAFNPP

+

TABLE 3.2-2 (cont'd)

CO.RE AND CONTAINMENT COOLING SYSTEM INITIATION AND CONTROL INSTRUMENTATION OPERABILITY REQUIREMENTS 1 Minimum No. of Operable Instrument Total Number of Channels Per instrument Channels item Trip System Provided by Design No. (Notes 1 and 2) Trio Function Trio Level Setting for Both Trio Systems Remarks 18 1 (Note 9) ADS Trip System Loss of Voltage 2 Monitors availability Bus Power Monitor of power to logic systems.

19 1 (Note 9) HPCI Trip System Loss of Voltage 2 Monitors availability Bus Power Monitor of power to logic systems.

20 1 (Note 9) RCIC Trip System Loss of Voltage 2 Monitors availability Bus Power Monitor of power to logic. systems. 1 21 1 (Note 9) Core Spray Sparger s 0.5 psid 2 Alarms to indicate l to Reactor Pressure Core Spray sparger Vessel d/p pipe break.

22 2 Condensate Storage 2 59.5 in. above 2 (Note 8) Transfers RCIC pump Tank Low Level tank bottom . suction to suppression

(= 15,600 gal. avail) chamber.

23 2 Condensate Storage ' 2 59.5 in. above tank 2 (Note 8) Transfers HPCI pump Tank Low Level bottom suction to suppression

. (=15,600 gal avail) - chamber.

24 2 Suppression Chamber s 6 in. above normal 2 (Note 8) Transfers HPCI pump High Level level suction to suppression chamber.

25 1 (Note 9) LPCI Cross-Connect NA 1 (Note 8) Alarms when valve :

Valve Position is not closed.

t Amendment No. Jf , jd, jd, jl4-69 !

G-

=__ - -- ___-_ - _ ___

. .e JAFNPP j IABLEl2-2 (conLd)

COBE_AND_CO_NIAINMENT COOLING SYSTEM INITIAllO_N_AND N CONIROL INSTRUMENTA_IlON_OPEBABILLTY_REQULREMENIS Minimum No. of Operable Instrument Total Number of Channels Per Instrument Channels item Trip System Provided by Design blo. 1 Notes 1 annd_2) Trip Fun.ction Trip LevetSetting for Both Trio _ Systems Remarks 26 (1 per 4kV bus) 4kV Emergency Bus 110.6 1.2 2 initiates both akV (Note 9) Undervoltage Relay secondary volts Emergency Bus Undervoltage (Degraded Voltage) Timers. (Degraded Voltage LOCA and non-LOCA)

(Notes 4 and 6) 27 (1 per 4kV bus) 4kV Emergency Bus 9.0 1.0 sec. 2 (Note 5)

(Note 9) Undervoltage Timer (Degraded Voltage LOCA) 28 (1 per 4kV bus) 4kV Emergency Bus 45 5.0 sec. 2 (Note 5)

(Note 9) Undervoltage Timer (Degraded Voltage non-LOCA) 29 (1 per 4kV bus) 4kV Emergency Bus 85 4.25 2 Initiates 4kV Errergency Bus (Note 9) Undervoltage Relay secondary volts Undervoltage Loss of -

(Loss of Voltage) Voltage Timer.

(Notes 4 and 7) 30 (1 per 4kV bus) 4kV Emergency Bus 2.50 0.05 sec. 2 (Note 5)

(Note 9) Undervoltage Timer (Loss of Voltage) 31 2 Reactor Low Pressure 285 to 335 psig 4 Permits closure of recirculation pump discharge valve.

Amendment No. /, <)8 70

JAFNPP TABLE 32-2 (Cont'd)

CORE AND CONTAINMENT COOLING SYSTEM INITIATION AND CONTROL INSTRUMENTATION OPERABILITY REQUIREMFJ{[$

NOTES FOR TABLE 3.2-2

1. Whenever any ECCS subsystem is required by Specification 3.5 to be operable, there shall be two operable or tripped trip systems (or in the case of single trip system instrument logics, one operable trip system), except as provided for below:

a. For each Trip Function with one less than the required minimum number of operable instrument channels, place the inoperable instrument channel in the tripped condition

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Otherwise, declare the associated ECCS inoperable.

b. For each Trip Function with two or more channels less than the required minimum number of operable instrument channels:

1) Within one hour, verify sufficient instrument channels remain operable or tripped

to maintain trip capability in the Trip Function, and

2) Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , place the inoperable instrument channel (s) in one trip system" in the tripped condition *, and

3) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , restore the inoperable instrument channel in the other trip system to an operable status.

If any of these three conditions cannot be satisfied, declare the associated ECCS inoperable.

An inoperable instrument channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, if the inoperable instrument channel is not restored to operable status within the required time, declare the associated ECCS inoperable.

" This action applies to that trip system with the greatest number of inoperable instrument channels. If both systems have the same number of inoperable instrument channels, the ACTION can be applied to either trip system.

2. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into associated Umiting Conditions For Operation and required actions may be delayed as follows: (a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for single channel Trip Functions; or (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for the remaining Trip Functions provided the associated Trip Function maintains ECCS initiation capability.

Amendment No. p8, jid, ipb,1pD,1$D 71

JAFNPP' Table 3.2-2 (Cont'd)

CORE AND CONTAINMENT COOLING SYSTEM INITIATION AND CONTROL INSTRUMENTATION OPERABILITY REQUIREldENIS

3. Refer to Technical Specification 3.5 for Limiting Conditions for Operation. Failure of one (1) instrument channel disables .

automatic initiation of one (1) pump.

4. Tripping of 2 out of 2 sensors is required for an undervoltage trip. With one operable sensor, operation may continue with the inoperable sensor in the tripped condition.

5. The 4kV Emergency Bus Undervoltage Timers (degraded voltage LOCA, degraded voltage non-LOCA, and loss-of-voltage) initiate the following: starts the Emergency Diesel-Generators; trips the normal / reserve tie breakers and trips all 4kV motor breakers (in conjunction with 75 percent Emergency Diesel-Generator voltages); initiates diesel-generator breaker close permissive (in conjunction with 90 percent Emergency Diesel-Generator voltages) and; initiates sequential starting .of vital loads in conjunction with low-low-low reactor water level or high drywell pressure.

6. A secondary voltage of 110.6 volts corresponds to approximately 93% of 4160 volts on the bus.

I

7. A secondary voltage of 85 volts corresponds to approximately 71.5% of 4160 volts on the bus.

8. Only one trip system.

9. Single channel trip systems.

J 1

Amendment No.

71a

= _ - - _ _ _ _ _ _ - - - - - . . _ - - .-. -. - ,_. .. _ .- - . . . . -_. . -.. ....- ...- - -. - - - . ._

2

, JAFNPP

- TABLE 3.2-3 CONTROL _F10_D_B_ LOCK INSTRUMENTAllON BEQUEE_M_ENTS. .

Minimum No. of

- Operable instrument Channels Per Total Number of -

Trip Function instrument Channels (Notes _ Land 3) Trip Function Trip _ Level Setting Provided By Design Action _(N_ ole _2) 4 APRM Flow Referenced (Note 9) -6 A Neutron Flux 4 APRM Neutron Flux-Start-up s 12% 6 A 4 APRM Downscale 2 2.5 indicated on scale 6 A 2 (Note 7) Rod Block Monitor (Flow Biased) (Note 9) 2 B 2 (Note 7) Rod Block Monitor (Downscale) 2'2.5 indicated on scale 2 B 6 IRM Detector not in Start-up (Note 8) 8 A Position 6 IRM Upscale - s 86.4% (.108/125) of full scale 8 A 6 IRM Downscale (Note 4) 2 2% (2.5/125) of full scale 8 -A 3 SRM Detactor not in Start-up (Note 5) 4 A Position 5

3 (Note 6) SRM Upscale s 10 counts /sec 4 A-2 Scram Discharge Instrument s 26.0 gallons per .2 C (Note 10) . -

Volume High Water Level instrument volume Amendment No. 3d, S$, /$,98,1/2 72 -

. . _ _ _ . . - _ . _ . . . . . _ - . . _ _ . . . . - . . _ - . _ - . - . _ . - . . _ . . . _ . - . _ _ . . . _ ~ . . . . _ , . _ . -

- - - - - _ _ . _ - _.....-.___-u..-- --

_ s JAFNPP IABLE 3.2-3 (Cont dj CONTROLlOD_B_L_O_CK INSTRUMENTATIORBEQUlBEMENTS EOTES_ EOR TABLE 12-3

1. The trip functions shall be operable in the Startup and Run modes except as follows:

a) SRM and IRM: Startup mode only.

b) RBM: Run mode and > 30% reactor power only.

c) APRM Neutron Flux-Startup: Startup mode only.

d) APRM Flow Referenced Neutron Flux: Run mode only.

2. Actions:

Action A: If the number of operable instrument channels is:

a) one less than the required minimum number of operable instrument channels per trip function, restore the inoperable instrument channel to operable status within 7 days, or place the inoperable instrument channel in the tripped condition within the next hour.

b) two or more channels less than the required minimum number of operable instrument channels per trip function place at least one inoperable instrument channel in the tripped condition within one hour.

A_clinn B: If the number of operable instrument channels is:

a) one less than the required minimum number of operable instrument channels per trip function, verify that the reactor is not operating on a Umiting Control Rod Pattem, and within 7 days restore the inoperable instrument channel to operable status; otherwise, place the inoperable instrument channel in the tripped condition within the next hour. See Specification 3.3.B.5.

b) two channels less than the required minimum number of operable instrument channels per trip function, place at least one inoperable instrument channel in the tripped condition within one hour. See Specification 3.3.B.S.

Action Cl if the number of operable instrument channels is less than the required minimum number of operable instrument channels per trip function, place the inoperable instrument channel in the tripped condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Amendment No.6$ , f/, /2, %,1%,1/2 73

4 JAFNPP Iabhd2-3_LCantd)

C O NIB O L_PQD_B LQC K_IN STR U M E N_TATlO_N_R_ E Q UIR E M E NTS NOTE _S_FOR TABLE _3.2-3_(Con 'd) i

3. When a channel is placed 'n an inoperable status solely for performance of required surveillances, entry into associated Limiting Conditions for Or>eration and required actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Trip Function maintains CRB initiation capability. I

4. IRM downscale is bypassed when it is on its lowest range.

5. This function is bypassed when the count rate is 2100 cps.

6. This SRM Function is bypassed when the IRM range switches are on range 8 or above.

7. RBM is required when reactor power is greater than or equal to 30%.

8. This function is bypassed when the Mode Switch is placed in Run.

9. The APRM Flow Referenced Neutron Flux and Rod Block Monitor trip level setpoint shall be less than or equal to the limit specified in the Core Operating Umits Report.

10. When the reactor is subcritical and the reactor water temperature is less than 212 F, the control rod block is required to be operable only if any control rod in a control cell containing fuel is not fully inserted.

1 Amendment No. %, <)d, R$, p$,1$

74

, e JAFNPP IABLE_3.2-7 ADVS_RECIR_CULATION PUMP _ TRIP _!NSIRUMENTAIlONREQU!BEMENIS

~

Minimum Number of Operable Instrument Channels Per Trip System (Notes 1 & 2) Trip Function Trip Level Setting Applicable Modes 2 Reactor Pressure - High s 1120 psig Run 2 Reactor Water Level - Low Low > 126.5 in. Run aboue TAF NOTES __F_OR TABLE 3.2-Z See next page for Notes 1 and 2.

l Amendment No. [. %. [. jd. f/. If,1[.1/1 76

JAFNPP TABLElP-7_Icontdj ATWS_ RECIRCULATION PUMP TRIP INSTRUMENTAT_ ION REQUIRlMENTS NO_TES_EO R_ TABLE _3.2-Z

1. There shall be two operable or tripped trip systems for each Tnp Function, except as provided for below:

a. For each Trip Function with one less than the required minimum number of operable instrument channels, place the inoperable instrument channel and/or its associated trip system in the tripped condition

within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Otherwise, place the reactor in the start-up/ hot standby mode within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. For each Trip Function with two or more channels less than the required minimum number of operable instrument channels:

1) Within one hour, verify sufficient instrument channels remain operable or tripped

to maintain trip capability in the Trip Function, and

2) Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . place the inoperable instrument channel (s) in one trip system and/or that trip system" in the tripped condition *,

and

3) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . restore the inoperable instrument channel in the other trip system to an operable status.

If any of these three conditions cannot be satisfied. place the reactor in the start-up/ hot standby mode within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

An inoperable instrument channel or trip system need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, if the inoperable instrument channel is not restored to operable status within the required time, place the reactor in the start-up/ hot standby mode within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

This action applies to that trip system with the greatest number of inoperable instrument channels. If both systems have the same number of inoperable instrument channels, the ACTION can be applied to either trip system. !

2. When a channel is placed in an inoperable status solely for performance of required sunteillances, entry into associated Limiting Conditions for Operation and required actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Trip Function maintains ATWS RPT initiation capability.

Amendment No.[, p0.1[.1/4.1/2 77

1 JAFNPP TABLE 4.2-1 PEIMARY CONTAINMENT ISOLATION SYSTEM INSTRUMENTATION TEST AND CAllBRATION REQUIREMENTS i

instrument Channel (Note 8) Instrument Functional Test Calibration Frequency Instrument Check (Note 4)

1) Reactor High Pressure O O NA (Shutdown Cooling isolation)

2) Reactor Low-Low-Low Water Level O (Note 5) . R (Note 15) D ,

l

3) Main Steam High Temperature O (Note 5) R (Note 15) D

4) Main Steam High Flow O (Note 5) R (Note 15) D

5) Main Steam Low Pressure O (Note 5) R (Note 15) D

6) RWCU Area High Temperature . O O (Note 16) NA

7) Condenser Low Vacuum O (Note 5) R (Note 15) D

8) Main Steam Une High Radiation O (Note 5) O/R (Note 11) D

. 9) HPCI & RCIC Steam Une High Flow O (Note 5) R (Note 15) D

10) HPCI & RCIC Steam Une/ O (Note 5) R (Note 15) D Area High Temperature

11) HPCI & RCIC Steam Une Low Pressure O (Note 5) R (Note 15) D

12) HPCI & RCIC High Exhaust Diaphragm O O NA Pressure NOTE: See notes following Table 4.2-5.

Amendment No. [, jib,1%,1/1,1%,1pb,2/7 78

, =;

JAFNPP IA_BLE_412-1 (Cont'd)

EBIMARY CANTAINMENT ISOLATION _S1STEMJMSlBi1 MENTATION IE.ST AND_CALIBBAllON REQt11BEMEN_TS

)

Logic System Functional Test (Notes 7 & 9) Frequency

1) Main Steam Une Isolation Valves SA Main Steam Une Drain Valves Reactor Water Sample Valves

2) RHR - Isolation Valve Control SA

- Shutdown Cooling Valves

3) Reactor Water Cleanup isolation SA

4) Drywell isolation Valves SA TIP Withdrawal Atmospheric Control Valves

5) Standby Gas Treatment System . SA Reactor Building Isolation ;

6) HPCI Subsystem Auto Isolation SA

7) RCIC Subsystem Auto isolation SA NOTE: See notes following Table 4.2-5.

Amendment No. /,48, jfd, Q4,196,1%,1/0,1/1,1)fD 79

JAFNPP TABLE 4.2-2 CORE AND CONTAINMENT COOLING SYSTEM INSTRUMENTAILQB TEST AND CALim Instrument Channel Instrument Functional Test Calibration Frequency Instrument Check (Note 4)

1) Reactor Water Level Q (Note 5) SA / R (Note 15) D 2a) Drywell Pressure (non-ATTS) Q Q NA 2b) Drywell , 'ressure (ATTS) Q (Note 5) SA / R (Note 15) _ D 3a) Reactor Pressure (non-ATTS) O O NA' 3b) Reactor Pressure (ATTS) Q (Note 5) SA / R (Note 15) D

4) Auto Sequencing Timers NA R NA

5) ADS - LPCI or CS Pump Disch. Q Q NA

6) Trip System Bus Power Monitors O NA NA

7) Core Spray Sparger d/p Q Q D

8) HPCI & RCIC Suction Source Levels Q O NA.

9) 4kV Emergency Bus Under-Voltage R R NA (Loss-of-Voltage, Degraded Voltage LOCA and non-LOCA) Relays and Timers.

10) LPCI Cross Connect Valve Position R NA NA NOTE: See notes fc!!owing Table 4.2-5.

l Amendment No. /, %, 1/9,1/1,2/1 $

80

6 JAFNPP-TABLE 4.2-2 (Cont'd)

CORE AND CONTAINMENT COOLING SYSTEM INSTRUMENTATION TEST AND CAllBRATION REQUIREMENTS f

Logic System Functional Test Frequency.

1) - Core Spray Subsystem SA (Notes 7 & 9).

2) Low Pressure Coolant injection Subsystem SA (Notes 7 & 9) !

3) Containment Cooling Subsystem SA

4) HPCI Subsystem SA (Notes 7 & 9)

5) - ADS Subsystem SA (Notes 7 & 9)

NOTE: See notes following Table 4.2-5.

t Amendment No. /, %, j8.1/1 81

JAFNPP TABLE 4.2-3 CONTROL ROD BLOCK INSTRUMENTION TEST AND CALIBRATION REQUIREMENTS Instrument Functional Instrument Instrument Channel Test (Note 5) Calibration Check (Note 4)

1) APRM - Downscale Q Q D

2) APRM - Upscale Q Q D -

3) IRM - Upscale S/U (Note 2) Q (Notes 3 & 6) D

4) IRM - Downscale S/U (Note 2) Q (Notes 3 & 6) D

5) IRM - Detector Not in Startup Position S/U (Note 2) NA NA

6) RBM - Upscale Q Q D

7) RBM - Downscale Q Q D

8) SRM - Upscale S/U (Note 2) O (Notes 3 & 6) D

9) SRM - Detector Not in Startup Position S/U (Note 2) NA NA ;

10) Scram Discharge Instrument Volume - Q Q D High Water Level (Group B Instruments) i Logic System Function Test (Notes 7 & 9) Frequency

1) System Logic Check - SA NOTE: See notes following Table 4.2-5.

Amendment No. /,90, p6 4 82 7

.. :s JAFNPP NOTES FOR TABLES 4.2-1 THROUGH 4.2-5 1 Initially once every month until acceptance failure rate data are 8. Reactor low water level, and high drywell pressure are not available; thereafter, a request may be made to the NRC to included on Table 4.2-1 since they are listed on Table 4.1-2.

change the test frequency. The compilation of instrument failure rate data may include data obtained from other boiling water . . 9. The logic system functional tests shall include a calibration of reactors for which the same design instruments operate in a time delay relays and timers necessary for proper functioning -

environment similar to that of JAFNPP. of the trip systems. ,

2. Functional tests are not required when these instnaments are not 10. (Deleted) required to be operable or are tripped. Functional tests shall be l

performed within seven (7) days prior to each startup. 11. Perform a calibration once per operating cycle using a radiation source. Perform an instrument channel alignment

3. Calibrations are not required when these instruments are not once every 3 months using the built-in current source.

required to be operable or are tripped. Calibration tests shall be performed within seven (7) days prior to each startup or prior to 12. (Deleted) a pre-planned shutdown.

13. (Deleted)

4. Instrument checks are not required when these instruments are not required to be operable or are tripped. 14. (Deleted)

5. This instrumentation is exempt from the functional test definition. 15. Sensor calibration once per operating cycle. Master / slave trip The functional test will consist of injecting a simulated electrical unit calibration once per 6 months.

signal into the measurement channel.

16. The quarterly calibration of the temperature sensor consists

6. These instrument channels will be calibrated using simulated of comparing the active temperature signal with a redundant electrical signals once every three months. temperature signal.

7. Simulated automatic actuation shall be performed once each operating cycle.

Amendment No. %, d jid, 17,'1/1,2/7 6 84

e. ..

JAFNPP TABLE 4.2-7 ATWS RECIRCULATION PUMP TRIP INSTRUMENTATION TEST AND CALIBRATION REQUIRE.MENTS FUNCTION CHANNEL CHANNEL TRIP UNIT CHANNEL . SIMULATED AUTO ACTUATION CHECK FUNCTIONAL CAllBRATION CAllBRATION & LOGIC FUNCTIONAL TEST TEST Reactor Pressure-High D Q SA R R Reactor Water Level-Low Low D Q SA R R Amendment No./,fd ,)d,)4. jd1 85

JAFNPP THIS PAGE IS INTENTIONALLY BLANK Amendment No.

87

e- s-JAFNPP

7.0 REFERENCES

(1) E. Janssen, " Multi-Rod Bumout at Low Pressure," ASME Paper (9) C.H. Robbins, " Tests of a Full Scale 1/48 Segment of the 62-HT-26 August 1962. Humbolt Bay Pressure Suppression Containment," GEAP-3596, November 17,1960.

(2) K.M. Backer, " Burr.out Conditions for Flow of Boiling Water in Vertical Rod Clusters," AE-74 (Stockholm, Sweden), May 1962. (10) " Nuclear Safety Program Annual Progress Report for Period Ending December 31,1966, Progress Report for Period Ending (3) FSAR Section 11.2.2. December 31,1966, ORNL-4071."

(4) FSAR Section 4.4.3. (11) Section 5.2 of the FSAR.

(5) 1.M. Jacobs, " Reliability of Engineered Safety Features as a (12) TID 20583, " Leakage Characteristics of Steel Containment Function of Testing Frequency," Nuclear Safety, Vol. 9, No. 4, Vessel and the Analysis of Leakage Rate Determinations."

July-August 1968, pp 310-312.

(13) Technical Safety Guide, " Reactor Containment Leakage Testing Deleted and Surveillance Requirements," USAEC, Division of Safety l(6) Standards, Revised Draft, December 15,1966.

(7) 1.M. Jacobs and P.W. Mariott, APED Guidelines for Determining Safe Test Intervals and Repair Times for Engineered Safeguards (14) Section 14.6 of the FSAR.

- April 1969.

(15) ASME Boiler and Pressure Vessel Code, Nuclear Vessels, (8) Bodega Bay Preliminary Hazards Report, Appendix 1, Docket Section lit. Maximum allowable intemal pressure is 62 psig.

50-205, December 28,1962.

(16) 10 CFR 50.54, Appendix J, " Reactor Containment Testing Requirements."

(17) 10 CFR 50, Appendix J, February 13,1973.

1 Amendment No.1)d 285~

g 1 ATTACHMENT 11 to JPN-94-050 SAFETY EVALUATION FOR EBQEOS ED_IECHNLCAkSEEC1BCallDNSH ANG ES INSTRUMENTATION SURVEILLANCE TEST INTERVALS, ALLQWABLEQ11T-OF-SERVICE TIMESJND OTHER CBANGES JEIS-R0:010 l

i l

New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59 i

___J

it

. Attachment 11 to JPN-94-050 SAFETY EVALUAllON FOR PROPOSED TECHNICAL SPECIFICATION CHANGES :

INSTRUMENTATION SURVEILLANCE TEST INTERVALS, ALLOWABLE l QUT-OF-SERVICEllMESJWD_DIHER.CBARGES (JPTS-20-010) f

1. DESCRIPTIOfLAND PURPOSE OF THE PROPOSED CHANGES This application requests the following changes to the James A. FitzPatrick Technical- !

Specifications:

i

1. Incorporate the results of the General Electric Licensing Topical Reports, prepared under the direction of the BWR Owners Group, supporting an increase of the ,

sumeillance test intervals (STI), and the repair and testing allowable out-of-service times (AOT), for most of the instrumentation listed in the Technical Specifications. l

2. Remove instrument response time limits. ,

3. Remove the Average Power Range Monitor (APRM) downscale scram function.

4. Make additional miscellaneous changes to the instrumentation sections.

I Because of the extensive nature of the changes, the exact wording of the proposed changes to the Technical Specifications (TS) are not provided in the following ;

description. The proposed changes are described in sufficient detail so that, when j

reviewed in conjunction with the revised TS pages in Attachment I and the marked-up TS pages in Attachment lit, a clear understanding of the changes to each specification '

and the referenced tables is provided. i Minor changes in format, such as type font, margins or hyphenation, are not described in j this submittal. These changes are typographicalin nature and do not affect the content !

of the Technical Specifications. ,

The proposed changes to the James A. FitzPatrick Technical Specifications are grouped [

into four categories. These categories and the intended purpose of the changes are as ;

follows

1. Incorocrate STI and AOT improvements - Category _1 i This application proposes an extension of the surveillance test intervals (STis) from weekly or monthly to quarterly for the functional tests for most of the instrumentation in the Technical Specifications. Additionally, allowable out-of-service times (AOT) are proposed for the instrumentation. These times, specified separately for both repair and test situations, represent the time that the instrument may be inoperable before entry into its Limiting Condition For Operation action statement. The bases for these changes are presented in seven NRC approved ,

General Electric Licensing Topical Reports (References 1 through 7), prepared l under the direction of the BWR Owners Group.

I I

l

.- l i

. Attachment li to JPN-94-050 SAFETY EVALUATION Page 2 of 39 The STI and AOT revisions will enhance plant safety by reducing the potential for l test related plant scrams, excessive test cycles on equipment, and operator errors. j l

2. flelacation ollnstwment Responselirne Limits - CategoIy_2 This application incorporates a line-item TS improvement that implements the guidance of Generic Letter 93-08, " Relocation of Technical Specification Tables of Instrument Response Time Limits" (Reference 24). The NRC guidance permits the relocation of the instrument response times from the TS to the Updated FSAR ,

(UFSAR). Maintaining the response times in the UFSAR permits changes under the regulatory provisions of 10 CFR 50.59 without the need to use the license amendment process. The change does not alter the surveillance requirements for these limits.

The change to the TS involves the deletion of the two TS tables that specify instrument response time limits (Table 3.1-2, " Reactor Protection System Instrumentation Response Times," and Table 3.2-9, " Primary Containment isolation System Actuation Instrumentation Response Times"). The response time limit requirements will be relocated to the UFSAR in the July 1995 annual update.

3. Delete APRM Downsca!e Scram - CateQQIy.3 This application proposes a change to delete the APRM downscale scram function i listed in Technical Specification Table 3.1-1, " Reactor Protection System (Scram) ir.strumentation Requirements." This function does not independently initiate a scram. The APRM downscale scram actually serves as an Intermediate Range Monitor (IRM) high flux scram interlock that bypasses the IRM high scram when both of the following conditions exist: (1) Reactor is in the RUN mode, and (2)

APRM power level is above the APRM downscale trip setpoint (above 2.5%). This interlock feature is described in section 7.5.5.4 and Table 7.5-4 of the UFSAR.

The change will permit removal of the APRM downscale scram from the design of

. the RPS system. Removal of this interlock feature will defeat the IRM high scram in the RUN mode irrespective of the APRM power level. The rod block and i' annunciator functions associated with the APRM downscale trip will remain.

The change will permit any one IRM channel per trip system and any one APRM 3 channel per trip system to be simultaneously bypassed, as intended by the plant !

design (UFSAR 7.5.5.3 and 7.5.7.4), avoiding the need to operate the plant in the

" half scram" condition. Due to the different number of APRM and IRM channels (six vs. eight) wme IRM channels share the same APRM channel. Consequently, some combinations of bypassed IRM and APRM channels results in less than the minimum number of required operable APRM downscale scrams. For these combinations, one of the failed channels cannot be bypassed, leaving the plant in a

" half scram" condition. The change will provide the IRM and APRM channel i bypass flexibility intended by the original plant design without the need to operate l the reactor in a " half scram" condition. j i

___--__-__-__-__-______l

. Attachm:mt 11 to JPN-94-050 SAFETY EVALUATION Page 3 of 39

4. MiscellaDegus_ Changes - CategoIy_4 Changes that include editorial revisions, clarification improvements, relocation of material, and revisions to conform the TS to the actual plant design as described in the FSAR. .

The following describes the changes and their purpose. For ease of review, the discussion has been structured to parallel the order of presentation in the Technical Specifications. The page numbers specified in the heading of each change description pertain to the location of the subject material in the current Technical Specifications.

These page numbers are the same for the location of the subject material in the revised Technical Specifications unless otherwise noted in the discussion. The category of each change (CAT 1, 2, 3, or 4) is specified in each change description.

A. Table of Contents. List of Tables / Figures. Pages i. v and_yji

1. Page i: Revise the Table of Content page numbers to reflect the re-distribution of text. (CAT 4)

2. Page v: Delete Tables 3.1-2 and 3.2-9 from List of Tables. The change reflects the removal of instrument response time limits in accordance with NRC Generic Letter 93-08. Revise the table titles and page numbers to reflect changes described later in this section. (CAT 2 and 4)

3. Page vii: Delete Figures 4.1-1 and 4.2-1 from List of Figures. The change reflects the addition of new STI / AGTs based on several GE Licensing Topical Reports as discussed later in the safety evaluation. (CAT 1)

B. Specification 1.0. Definitions. Pagej

1. Page 5: Add a new definition: "T. Instrument Surveillance Frequency

, Notations / Intervals," that defines surveillance frequency notations and intervals.

The definition conforms with the definition in the Standard Technical Specifications (Reference 17), Table 1.1, page 1-7. The new definition permits the use of notations for surveillance intervals on the instrumentation tables subject to changes in this amendment application, and relates all surveillance intervals to a consistent and precise time pericd.

Further, the definition clarifies "once each operatic.g cycle", and simllar phrases, by relating the interval to the definition of the frequency notation "R". This will apply the definition of "R" to test frequencies based on operating cycle intervals as used throughout the Technical Specifications, in addition to the "once per operating cycle" requirement used on pages subject to this application. In this manner, a consistent and precise time period will be established for all operating cycle requirements in the TS. (CAT 4) 1

.~ _ . - _ . - - . _ . .

4 - -

Attachment 11 to JPN-94-050 l SAFETY EVALUATION ,

Page 4 of 39 !

2. Page 5: Change " Atomic Energy Commission" to " Nuclear Regulatory Commission." (CAT 4)

C. Spacification 3.0. General Limiting Conditionsfor_Qperation. Pages 30a. 30d. and !

. 30e., 4 The changes appear on revised pages 30a, 30d, 30e, and 30f. _j Add new Specification 3.0.F (revised page 30a) tc permit equipment removed from service, or declared inoperable to satisfy a TS action statement, to be retumed to >

service in order to perform testing to demonstrate its operability or the operability of ,

other equipment. Also add Bases 3.0.F on revised page 30d. These additions s conform with NUREG-1433, LCO 3.0.5, page 3.0-2, and Bases LCO 3.0.5, pages B .l 3.0-6 and 7 (Reference 18). Current pages 30d and 30e are renumbered 30e and :

30f, respectively, to accommodate a redistribution of text. The revised text appears ;

on revised pages 30a and 30d. .

New Specification 3.0.F envelops a current LCO presented in the second sentence ,

of Specification 4.1.D which states: "The trip system containing the unsafe failure may be placed in the untripped corJition during the period in which surveillance 4 testing is being performed on the other RPS channels". The addition of the new specification as proposed will permit deletion of this sentence. (CAT 4) '

D. Specification 3.1/4.1. Reactor Protection SysleHLfages 30f and 31 f

The revise text appears on revise pages 30g and 31.

1, Page 30f: Delete the reference to the table that specifies RPS instrument response time limits. The table is removed in accordance with NRC Generic Letter f 93-08. The information in the table regarding response time limits will be relocated ,

to the UFSAR. The changes involve: (1) removal of the last sentence in the first l paragraph of Specification 3.1.A, (2) removal of the Table reference in the second 1 paragraph of Specification 4.1.A, (3) the insertion of the sentence, " Neutron l detectors are exempt from response time testing," into Specification 4.1.A, and (4) l add a list of the RPS trip functions subject to the response time surveillance [

requirement of Specification 4.1.A (list conforms to the trip functions listed on i deleted Table 3.1-2). The change regarding neutron detectors is consistent with !

note 2 on deleted Table 3.1-2 (page 43a), and with the guidance of Generic Letter 93-08. (CAT 2)

2. Page 30f; Make editorial changes to the first sentence of Specification 3.1.A to i conform with the contents of Table 3.1-1. Re-distribute text associated with {

Specification 3.1.B and 4.1.8 to page 31. (CAT 4) j l

3. Page 31: Delete the first sentence of Specification 4.1.D that reads "When it is determined that a channel has failed in the unsafe condition, the other RPS j channels that monitor the same variable shall be functiona!!y tested immediately j before the trip system containing the failure is tripped." This provision is

' -- Attachment 11 to JPN-94-050 SAFETY EVALUATION Page 5 of 39 unnecessary based on the absence of a similar provision in the General Electric Ucensing Topical Report NEDC-30851P-A (Reference 1) that provides the basis for an extension in the STl/AOTs for RPS instrumentation, and the amendment approved for Duane Arnold (Reference 23). (CAT 1)

4. Page 31: Deiste the second sentence of Specification 4.1.D to reflect the addition of new Specification 3.0.F on revised page 30a. See change C. (CAT 4)

5. Page 31: Renumber Specification 4.1.E as 4.1.D. (CAT 4)

E. Bases 3.1. Page 32 Revise the second paragraph to reference the NRC approved evah'ation used as the bases for the new AOTs. (CAT 1)

F. Bases _4J. Paces 36. 37. 38. 39. and 4D The changes appear on revised pages 36, 37, and 38. Page 39 is blank.

The changes to the Bases 4.1, (RPS surveillance requirement) accomplish the following objectives:

(1) Remove text that discusses previous instrument reliability evaluations that are obsolete, and insert the reference to the NRC approved evaluations.

i (2) Establish consistency with the new instrument repair and test AOTs.

(3) Clarify the current design of the RPS instrumentation. 1 (4) Reflect the removal of the instrument response time limits.

(5) Reorganize the text according to subject matter.

The changes that accomplish these objectives are as follows: 1

1. Page 36: Delete the first paragraph since it discusses an instrument reliability evaluation superseded by the NRC approved Licensing Topical Report (LTR).

(CAT 1)

2. Page 36: Revise the second paragraph to: (1) add a reference to the GE Licensing Topical Report NEDC-30851P-A (Reference 1) used as the bases for the change to a quarterly functional test interval, and (2) delete text discussing instrument reliability criteria unrelated to the reliability data used in the newly referenced GE evaluation. (CAT 1) ,

i

3. Page 37: Delete the first paragraph since it discusses an instrument reliability evaluation that has been superseded by the NRC approved LTR. (CAT 1)

/

m _ _ . _ .__ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _

,. m. _ . , . _ . . _= . . _ _

.- Attachmsnt il to 'JPN-94-050 SAFETY EVALUATION Page 6 of 39

4. Page 37: Revise the first sentence in the second paragraph to clarify the design of toe instrumentation. Delete the third sentence in this paragraph since this may not be possible under all circumstances. Delete the last sentence in this paragraph since it is not applicable to the newly referenced GE evaluation. The revised text appears on revised page 36. (CAT 1 and 4) ;

- 5. Page 38: Delete the word " unsafe" from the first sentence since it is a term not !

used in the TS. Also, revise the last sentence of the first paragraph to add l

" functionally," and add a reference to the GE Ucensing Topical Report in a new ;

sentence. The revised material appears on revised page 36. (CAT 1) ~ ;

6. Page 38: Delete the second and third paragraph since they discuss a previous [

instrument reliability evaluation that has been superseded by the NRC approved .

LTR. (CAT-1)

7. Page 38: Replace "once/ month" with "once/every three months" in the second ;

sentence of the fourth paragraph to reflect the new AOTs. The revised material !

appears on revised page 37. (CAT 1) .i

8. Page 38: . Delete the last paragraph to maintain consistency with the changes regarding the removal of the RPS response time limits in accordance with the guidance of Generic Letter 93-08 (Reference 24). (CAT 2)

9. Page 39: Add a reference for the Standard Technical Specifications mentioned in the first paragraph and move to revised page 37. Revise the second paragraph and move to revised page 38. Delete the third and fourth (except for the last- t sentence) paragraphs to remove text containing specific values for the instrument response times. These changes maintains consistency with the changes -

associated with Generic Letter 93-08 (Reference 24), regarding removal of the ,

response time limits. (CAT 2 and 4)

10. Page 40: Move the first paragraph to revised page 38. Move the second ;

paragraph (with a minor editorial change) and the third paragraph to revised page ,

36 so that the discussion of Group C devices follows the discussion of Group B devices. (CAT 4)

11. Page 40: Delete the fourth paragraph since drift specifications may change as a !

result of the modification process. '(CAT 4)

12. Page 40: Move the first and second paragraphs in the right column to the top of l revised page 37 to consolidate discussion of APRM calibration in one area of the bases. (CAT 4)

13. Page 40: Make editorial changes to the first sentence of the third paragraph in i the right column and move paragraph to revised page 38. (CAT 4) t

14. Page 40: Move Bases 4.1.B to revised page 38. (CAT 4) j i

.,.n , - ._. _.,--__ __ . - _ _ __ .- _-_ E

Attachment 11 to JPN-94-050 SAFETY EVALUATION Page 7 of 39 G. Table 3.1-1. Reactor Protection System Instrumentation. Pages 41. 41a. 41b. 42 acIL43 Revised Table 3.1-1 is redistributed onto revised pages 40,41,42,43, and 43a.

Pages 41a and 41b are deleted.

The changes accomplish the following objectives:

(1) Incorporates the 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months repair AOT and the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months test AOT based on GE Licensing Topical Report NEDC-30851P-A (Reference 1).

(2) Consolidates all of the LCOs statements into the notes to Table 3.1-1.

(3) Addresses a concern identified by the NRC that a loss of scram function may occur if two or more channels are inoperable during the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AOT.

(4) Deletes the APRM Downscale Scram trip function.

1. Pages 41,41a,41b,42,43: Renumber and add notes as follows. These changes appear on revised pages 40,41. 42, and 43. (CAT 1) a) Change note for the ACTION column from 1 to 3.

b) Add note 2 to the " Minimum No. of Operable Instrument Channels Per Trip System" column heading.

c) Renumber note 2 as note 4 with some minor editorial changes.

d) Relocate action statements from note 1 to note 3.

e) Renumber notes 4 through 8 as notes 5 through 9, respectively.

2. Page 41: Delete the mode switch annotation "(4 Selections)" since this information is not applicable to the operating requirements of Table 3.1-1. (CAT 4)

3. Page 41: Add the IRM High Flux trip level setting percentage that corresponds to 120/125. (CAT 4)

4. Pages 41, 41b, and 42: Delete obsolete note 16 from the " Mode in Which Function Must be Operable" column heading to reflect a previous amendment (amendment 207). (CAT 4)

5. Page 41a: Delete the APRM downscale trip function. The change provides the IRM and APRM channel bypass flexibility, as intended by the plant design (UFSAR 7.5.5.3 and 7.5.7.4), avoiding the need to operate the plant in a " half scram" condition under certain inoperable IRM/APRM channel combinations. (CAT 3)

Attachment 11 to JPN-94-050 SAFETY EVALUATION Page 8 of 39

6. Page 41a and 43: Add note 16 to the Drywell High Pressure and Reactor Low Water Level trip functions to identify the RPS instrumentation common to PCIS.

(CAT 1)

7. Page 42: Add the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months repair AOT as note 1 to Table 3.1-1. The AOT value conforms to General Electric Licensing Topical Report NEDC-30851P-A (Reference 1), pages 5-33; and NRC Safety Evaluation Report (Reference 8): Enclosure 2, page 1. The Limiting Condition for Operation (LCO) used in note 1 of Table 3.1-1 addresses an NRC concem (Reference 16) that a loss of scram function may occur if two or more channels are inoperable during the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AOT. The LCO uses the text, with some minor editorial changes, as that approved in an operating license amendment for the Duane Amold Energy Center (Reference 23). Failure of multiple channels in an RPS trip function may result in a loss of scram function if tripping the channels is delayed by the AOT. Note 1 requires confirmation within one hour of RPS functional capability after two or more channels become inoperable. Further, the LCO limits the inoperable channels in one of the trip systems to a 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AOT. (CAT I)

8. Page 42: Add the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AOT as note 2 to Table 3.1-1. The AOT value conforms to General Electric Licensing Topical Report NEDC-30851P-A (Reference 1), pages 5-34. The wording of the AOT conforms to NUREG-1433, Specification 3.3.1.1, and provides assurance that the associated trip function will remain ;

operational following entry into the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months test AOT. (CAT 1) !

9. Page 43: Delete note (9) since it pertains to the deleted APRM downscale trip l function. (CAT 3) l

10. Page 3: Delete note 11 since it is not currently used, nor has it been previously used, in the Technical Specifications. The note references another section of the i TS, and its deletion does not impact any TS requirement. (CAT 4) ;

H. Table 3.1-2. Reactor _&QbCliRELSystem Instrumentation ReERODse Times.

Page 43a Page 43a: Delete Table 3.1-2 in its entirety to reflect the guidance of NRC Generic Letter 93-08 (Reference 24) permitting the transfer of these limits to the UFSAR.

(CAT 2)

1. Tables 4.1-1 and 4.1-2. Reactor Protection System Instrumentation Surveillance BeguiremenLPages 44. 45. 45a Revised Table 4.1-1 appears on revised pages 44, and 45. Page 45a is deleted.

1. Revise the frequency of functional tests to conform with Licensing Topical Report NEDC-30851P-A (Reference 1), pages 5-35,5-36; and NRC Safety Evaluation Report (Reference 8), Enclosure 2, page 2 and 3. The specific changes are:

- Attachm:nt il to JPN-94-050 SAFETY EVALUATION Page 9 of 39

a. Page 44/45: Revise the frequency of channel functional tests from weekly or monthly to quarter 1y for the following:

APRM High Flux

APRM Inoperative

APRM Flow Biased High Flux

Reactor High Pressure

Drywell High Pressure

Reactor Low Level

Scram Discharge Instrument Volume - High Water Level (Group A & B sensors)

Main Steam Line Isolation Valve Closure

Turbine Control Valve Fast Closure

Turbine First Stage Pressure Permissive

. Turbine Stop Valve Closure The change is not applicable to the IRM high flux and APRM high flux in startup or refuel modes.

The turbine first stage pressure permissive listed in Table 4.1-1, page 45, is not listed in the STS used for the markup TS page in LTR NEDC-30851P-A.

However, the oermissive provides an interlock feature that bypasses the turbine control valve fast closure, and turbine stop valve closure, when turbine first stage pressure is below a point corresponding to 30% rated thermal power (FSAR 7.2.3.8), and therefore should be considered an integral part of the scram features subject to the quarteriy functional test interval.

The notations, as defined in proposed definition "T" on page 5, are used for all surveillance frequencies specified on the table. (CAT 1)

b. Page 44: Revise the frequency of the functional test for the RPS Channel Test Switch from "every refueling outage" to " weekly." The FitzPatrick RPS design uses separate scram contactors for the automatic scram logic and the manual scram logic (FSAR 7.2.3.5). Consequently, the manual scram switches do not actuate the automatic scram contactors. The RPS Channel Test Switches are provided to manually actuate the automatic scram contactors. Accordingly, the RPS Channel Test Switches should be subjected to the " weekly" functional test frequency in accordance with Licensing Topical Report NEDC-30851P-A ,

I (Reference 1). The weekly testing of these switches is acknowledged in the last paragraph on page 5-21 of the referenced LTR. (CAT 1)

c. Page 44/45a: Delete obsolete text from note 1, and soMitute a provision that )

weekly functional tests of a scram function may be usev . Itead of the RPS :

Channel Test Switches. Also transfer the test require" at after maintenance l from the table to note 1. Reference note 1 at the fun @ 11 test notation for the '

RPS Channel Test Switch. (CAT 1)

2. Page 44: Delete the APRM downscale trip function. The change provides the IRM I I

and APRM channel bypass flexibility, as intended by the plant design, avoiding the need to operate the plant in a " half scram" condition under certain inoperable i

I

Attachment 11 to JPN-94 G50 SAFETY EVALUATION Page 10 of 39 IRM/APRM channel ccmbinations. (CAT 3)

3. Make the following editorial chariges (CAT 4):

a. Page 44: Insert missing reference to note 2 to the " Group" heading.

b. Page 44/45: Change "APRM-Flow Bias" to "APRM-Flow Bias High Flux."

Change " Calibrate Flow Bias Signal" to " Trip Output Relays" for the APRM Flow Bias (Table 4.12 identifies calibration requirements). Revise the table title, and change the " Minimum Frequency" heading to " Functional Test Frequency."

c. Page 44/45: Delete note 7, and add a column to the table specifying the .

instrument check requirements. ,

d. Page 44/45a: Move the operating mode requirements for the weekly testing of the IRM High Flux, IRM Inoperative, and APRM High Flux in Startup or Refuel, scram functions to a new note (note 5).

e. Page 44/45/45a: Renumber note 7 as note 6.

f. Redistribute text from page 45a to page 45. Delete page 45a.

g. Page 45: Change the nomenclature for the " turbine control valve EHC oil pressure" to " turbine control valve fast closure" to be consistent with the nomenclature in Table 3.1-1.

J. Eigure 4.1-1. PageJB Delete Figure 4.1-1 since it is referenced in text deleted by change F.6. (CAT 1) i K. EageA9.EpucificatioDS_12A / 4.2. A. Primaty_ContaintueDt isolation Functions ,

Delete the references to Table 3.2-9 which specifies main steam isolation valve actuation instrumentation response time limits. The table will be removed in accordance with NRC Generic Letter 93-08 (Reference 24) permitting the transfer of these limits to the UFSAR. The changes involve (1) the removal of the second paragraph of Specification 3.2.A, (2) the removal of the reference to Table 3.2-9 in the last paragraph of Specification 4.2.A, and (3) editorial changes to Specification 4.2.A that identifies the specific trip functions subject to the response time surveillance requirement. (CAT 2)

L. Eage.50lontrolBodElechActuatino Delete 3.2.C.2 regarding the test AOT for the rod block monitor. A revised test AOT will be added to Table 3.2-3, instrumentation that Initiates Control Rod Blocks, as note 3, that conforms to GE Licensing Topical Report GENE-770-06-1 A (Reference 6), Appendix A, page A-40; and NRC Safety Evaluation Report i

l

Attachment 11 to JPN-94-050 SAFETY EVALUATION Page 11 of 39 (Reference 14), attachment: Technical Evaluation Report, page 16 and 17; Enclosure 1, Table 2; and Enclosure 2, page 3-51. (CAT 1).

M. Bases 3.2. Page155JifLand.51

1. Page 55: Add a paragraph that describes the trip logic for the isolation valves.

(CAT 4)

2. Page 56: Add a paragraph that describes the trip logic for the main steam line high temperature isolation function. (CAT 4)

3. Page 57: Add a paragraph that describes the trip logic for the HPCI and RCIC

'high temperature isolation function. (CAT 4) -

N. Bases 4.2. Pages_flLJP,LaDd33 Revised Bases 4.2 appears on revised page 61.

1. Pages 61,62, and 63: Revise Bases 4.2 as follows to reflect the STI / AOT changes based on the GE Licensing Topical Report (References 1 through 7).

(CAT 1)

a. Add a new second paragraph that identifies the GE Ucensing Topical Reports that provide the bases for the instrument AOTs and STis.

b. Delete the remainder of Bases 4.2 except for the first paragraph, and the last two sentences of the second paragraph, on page 61. The deleted materialis superseded by the bases provided in the referenced Licensing Topical Reports.

2. Page 61: Delete the first three sentences of the second paragraph on page 61 to reflect the guidance of NRC Generic Letter 93-08 (Reference 24) permitting the transfer of response time limits from the TS to the UFSAR. The deleted text involves a reference to deleted Table 3.2-9 and discussion of specific response times. The remaining text is presented as part of the last paragraph of revised Bases 4.2 (Page 61). Transfer the definition of response time interval from deleted Table 3.2-9 to the last paragraph on revised page 61. (CAT 2)

O. Iable 3.2-1. Instalmentation That initiates Primarv_ContainmenLLEolation. PagesE_4 andE5 The revised table and associated notes are redistributed to revised pages 62,63, 64, and 65, and new page 65a.

1. Pages 64,65: Make the following editorial changes. (CAT 4)

a. Relocate the action statement in note 2 to note 3.

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l l

Attachment 11 to JPN-94-050 :

SAFETY EVALUATION j Page 12 of 39 !

b. Delete note 5 since it does not appear on the table. j

c. Renumber notes 6 through 8 as notes 4 through 6, respectively. .}

d. Change the " Instrument". column heading to " Trip Function" and revise the title of Table 3.2-1. [

t

e. Make minor editorial changes to the trip function nomenclatures, j

'i

f. ' Move re-numbered notes 4,5, and 6 to the more appropriate location next to i the trip function notations. .

j' g'. Delete re-numbered note 5 from the Condenser Low Vacuum function since re-numbered note 6 addresses the run mode requirement.' i

h. Add new page 65a to accommodate a re-distribution of text i

2. Revised Page 63: Relocate eight instruments from Table 3.2-2 to Table 3.2-1 since these instruments perform an isolation function, not an ECCS function. This: I will establish consistency with the STS (Reference 17), and resolves an NRC concem expressed in inspection Report No. 50-333/88-01, item 10 (Reference 19).

The NRC inspector was concemed that the absence of these instruments from Table 3.2-1 may result in the inoperability of these instruments when primary ;

containment integrity is required by Specification 3.7.A.2. The relocation subjects i these instruments to the operability requirements of Specification 3.7.A.2. These ;

instruments are:

. HPCI turbine steam line high flow

HPCI steam line low pressure

HPCI turbine high exhaust diaphragm pressure e HPCI steam line/ area temperature

RCIC turbine steam line high flow ,

RCIC steam line low pressure l

RCIC turbine high exhaust diaphragm pressure j

RCIC steam line/ area temperature, l

Add action statement F to the table for these valve isolation functions. Action 1 statement F is described in renumbered note 3. With less than the minimum !

number of instrument channels operable, action statement F requires that the. l affected penetration flowpath be isolated within one hour, and the affected system i declared inoperable. Closing an isolation valve on the penetration assures that the -l associated system (HPCI or RCIC) is isolated when a portion of the pipe break l protection is in a degraded condition.- Declaring the HPCI or RCIC system j inoperable, limits continued plant operation to the seven day AOT specified for an ;

inoperable HPCI and RCIC system in TS 3.5.C.1.a and TS 3.5.E.1, respectively.

This action statement is similar to the action statement in the STS (Reference 17),

page 3-12. and 3-14, Action 22, for RCIC systems. .The portion of the action ;

statement regarding the requirement to " isolate the affected penetration flow path" l conforms with required action statement F.1 on page 3.3-2 of NUREG-1433 l (Reference'16). (CAT 4)

I

Attachm:nt 11 to JPN-94-050 -

SAFETY EVALUATION Page 13 of 39

3. Revised Page 63: For the HPCI Steam line Area Temperature function, revise the total number of instrument channels provided by design for both trip system, and the minimum number of operable instrument channels per trip system, from 2 and 1, to 16 and 8, respectively. For the RCIC Steam Une Area Temperature function, revise the total number of instrument channels provided by design for both trip system, and the minimum number of operable instrument channels per trip system, from 2 and 1, to 8 and 4, respectively. The change reflects the actual design of the trip system. The instrument channels provide inputs to two trip systems. If any one of the channels becomes inoperable, tne required action must be taken. This change does not impact the operability requirements for this trip function. (CAT 4)

4. Page 64: For the Main Steam Line Detection High Temperature function, revise the total number of instrument channels provided by design for both trip system, and the minimum number of operable instrument channels per trip system, from 4 and 2, to 16 and 8, respectively. The changes reflects the actual design of the trip system as the plant was licensed (FSAR 7.3.4.2). There are four temperature sensors located at each of fuur different areas in the vicinity of the steam lines for a total of 16 temperature sensors. If any one of the 16 channels becomes inoperable, the required action must be taken. This change does not impact the operability requirements for this trip function. (CAT 4)

5. Pages 64 and 65: Revise the action statement for the Main Steam Line High Flow isolation trip function to " Isolate the affected main steam line within eight hours" by adding Action G to the table and revised note 3. There are a total of four high flow ,

instrument channels, one for each of the four main steam lines. Loss of an instrument channel affects leakage detection capability for only its associated main steam line. Therefore, the appropriate action in response to an inoperable instrument channel is to isolate the main steam line associated with the high flow instrument channel. (CAT 4)

6. Pages 64 and 65: Add two trip functions to reflect the design of the isolation logic of the isolation valves on the primary containment hydrogen and oxygen '

concentration sample, and the gaseous and particulate radioactivity sample supply and return lines. These additional trip functions appears as the " Reactor Low Water Level" and "Drywell High Pressure" functions with notes 7 and 8. Note 8 is added to identify the isolation logic of the isolation valves on these systems as a '

two-out-of-two-taken-once logic. (CAT 4)

7. Page 65: Add the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing AOT as note 2 to Table 3.2-1, Instrumentation That Initiates Primary Containment Isolation. This AOT value conforms to: (1) GE Licensing Topical Reports NEDC-30851P-A, Supplement 2 (Reference 3), (2) GE LTR NEDC-31677P-A (References 5), page D-3, (3) NRC Safety Evaluation Report (Reference 10), Enclosure 1, Table 2; and Enclosure 2, page 3-16, and (4) NRC Safety Evaluation Report (Reference 13), Enclosure 1, page 2, and Enclosure 2, page 3-14. The wording of the AOT conforms, except as noted below, to NUREG-1433, Specification 3.3.6.1, and provides assurance that the associated trip function will remain operational following entry into the 6 hor fest AOT. The i exceptions are as follows:

a) Maintaining PCIS initiation capability is clarified to mean that "the associated

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Attachment 11 to JPN 94-050 SAFETY EVALUATION Page 14 of 39 Trip Function maintains PCIS initiation capability for at least one containment isolation valve on the affected penetration. (CAT 1) b) The 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AOT is also applied to the primary containment isolation valve affected by the test of the instrument channel. This permits the affected isolation valve to be deactivated in the open position for the duration of the test of its associated isolation logic. Placing the valve in this configuration is necessary on systems whose isolation valves are normally open to support plant operation (e.g. RWCU, HPCI, RCIC). This provision supersedes the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months AOT requirement of Specification 3.7.D.2 for inoperable isolation valves, and applies only when the valve is deactivated in the open

asition to support testing of its instrumentation. (CAT 1).

c) The STS (Reference 18) permits an unconditional 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing AOT for single channel trip systems. The presumption is that if the isolation logic is designed such that isolation capability is disabled when one instrument channel is removed from service for testing, then an unconditional 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AOT is necessary to effect testing of the instrumentation. For the same reason, an -

unconditional 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AOT is applied in note 2 to the trip functions designed as a two-out-of-two-taken-once logic This logic is utilized for the isolf. tion valves on the primary containment hydrogen and oxygen concentration snmple, and the gaseous and particulate sample supply and retum lines. Note 8 is added to ;

identify the two trip functions subject to this unconditional AOT. (CAT 1)

8. Page 65: Delete the first sentence of note 2 to Table 3.2-1, and a6d the longer repair AOTs to note 1. The AOT values conform to: (1) GE Licensing Topical Reports NEDC-30851P-A, Supplement 2 (Reference 3), (2) GE LTR NEDC-31677P-A (Reference 5), D-1 and D-2; (3) NRC Safety Evaluation Report (Reference 10), Enclosure 1, Table 2 and Enclosure 2, page 3-9, and (4) NRC Safety Evaluation Report (Reference 13), Enclosure 1, page 2, and Enclosure 2, page 3-9.

The change adds a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AOT for isolation instrumentation common to RPS and/or ECCS instrumentation, and a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AOT for isolation instrumentation not common to RPS. The proposed AOT uses text that is similar to the text proposed for the RPS AOT. The AOT is permitted for multiple channel failures only when their inoperability does not prevent the PCIS trip function capability. Note 1 ,

requires confirmation within one hour of PCIS functional capability after two or '

more channels become inoperable, otherwise entry into the action statement is required. (CAT 1)

9. Page 65: Add note 7 to identify instrumentation common to the RPS instrumentation. The instrumentation common to both the RPS and PCIS trip functions are the Reactor Low Water Level and Drywell High Pressure Instruments.

This note is provided to assist in the interpretation of the new AOT in note 1.

(CAT 1)

10. Page 65: Reword note 6 to read as a requirement rather than a design feature.

This does not change the operating requirement for this isolation function. (CAT 4)

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Attachment 11 to JPN 94 050 l SAFETY EVALUATION l Page 15 of 39 P. Iatlle 3.2-2. Insimmentation That InitiatesgCookols the Core and Containment C00liDg3ystems. Pages_.f}6. 67. 68. 69. 70. 70a. 70b. 70c. and 71.

The revised table and associated notes appear on revised pages 66,67, 68, 69, 70,71. Pages 70a,70b, and 70c are deleted.

1. Pages 66 through 71: Make the following editorial changes. (CAT 4)

a. Add a note 2 notation to the Table 3.2-2 heading for " Minimum Number of Operable Instrument Channels Per Trip System."

b. Delete " inst. channel (s)" in the column for " Total Number of Instrument Channels Provided by Design for Both Trip Systems" to eliminate unnecessary redundancy.

c. Add a second Reactor High Water Level trip function to page 66 to reflect the fact that the HPCI and RCIC trip features are controlled by independent trip systems, Renumber sequentially the remaining trip functions.

d. Make several editorial revisions to the trip function nomenclature, and the text in the " Remarks" column, to improve clarity. Revise the title of Table 3.2-2.

2. Page 66 through 71: Add a note 8 to identify functions with only one trip systems.

Add a note 9 to identify single channel trip systems. These notes assist in the interpretation of the revised repair and test AOTs. (CAT 1)

3. Page 67: For Drywell High Pressure, specify that the total number of instrument channels provided by design for both trip systems is "4" to reflect the plant design as described in UFSAR section 7.4.3.2.2. (CAT 4)

4. Page 69: Change the total number of instrument channels provided by design for both trip systems, and the minimum number of operable instrument channels per trip system, for the RHR (LPCI) Pump Discharge Pressure interlock, from 4 and 2, ,

to 8 and 4, respectively. The change reflects the actual design of the trip logic. If any one of the 4 channels becomes inoperable, the required action must be taken.

This change does not impact the operability requirements for this trip function.

(CAT 4)

5. Page 71: Add a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months repair AOT for the instrumentation listed in Table 3.2-2 that initiates or controls the emergency core cooling systems. Revised note 1 reflects the new 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AOT. The change for all instruments except the RCIC system conforms to: (1) GE Licensing Topical Report NEDC-30936P-A (Reference 4), Part 2, page A-18, (2) NRC Safety Evaluation Report (Reference 11), and (3)

NRC Safety Evaluation Report (Reference 12), pages 3 and 4. The change for the RCIC system instruments conforms to GE Licensing Topical Report GENE-770 2-A (Reference 7): Appendix C, page C-4-4; and NRC Safety Evaluation Report (Reference 15), Enclosure 1, pages 3 and 4, and Enclosures 2. (CAT 1)

The proposed AOT uses text that is similar to the text proposed for the RPS AOT.

The AOT is permitted for multiple channel failures only when their inoperability

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Attachment li to JPN-94-050 SAFETY EVALUATION Page 10 of 39 does not prevent ECCS trip function capability. Note 1 requires confirmation within one hour of ECCS functional capability after two or more channels become inoperable, otherwise entry into the action statement is required. ]

The NRC approved 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months repair AOT in NEDC-30936P-A, Part 2, Table A-9, references the action statement of Table 3.3.3-1 of the STS (Reference 17).

Accordingly, the new note 1 to Table 3.2-2 is revised to conform with the following features of the action statement in Table 3.3.3-1 of the STS (Reference 17):

a. The current action statement (note 1) requires that the trip system be placed in the tripped condition if at least one of the channels is inoperable. This is not the intent of the action statement, since placing a trip system in the tripped condition would initiate ECCS. The revised action statement (note ia) specifies a trip of the inoperable channel as the appropriate response to an inoperable condition. Tripping the inoperable channel will complete the safety function of the channel and permits the operable portion of the logic to function as designed (initiate ECCS in response to an actuation signal). (CAT 4)

b. The current action statement requires the reactor to be placed in the cold condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if the LCO is not satisfied. The current requirement can be interpreted in a manner that is inconsistent with the requirements of Technical Specification 3.5.F which identifies ECCS subsystem operability requirements when in cold shutdown. The revised action statement removes this inconsistent action requirement by declaring the " associated ECCS inoperable" when an inoperable instrument channel is not placed in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Declaring an ECCS inoperable assures that the appropriate LCO in Section 3.5 (ECCS LCOs) of the TS is applied. (CAT 4)

6. Page 71: Add a 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months test AOT for the instrumentation listed in Table 3.2-2 that initiates or controls the emergency core cooling systems. New note 2 reflects the new 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AOT. The test AOT value for all instruments, except the RCIC system, conforms to (1) GE Ucensing Topical report NEDC-30936P-A (Reference 4). Part 2, page A-17; (2) NRC Safety Evaluation Report (Reference 11); and (3) NRC Safety Evaluation Report (Reference 12), pages 3 and 4. The test AOT value for the RCIC system instruments conforms to GE Ucensing Topical Report GENE-770-06-2-A (Reference 7), Appendix C, page C-4-3; and NRC Safety Evaluation Report (Reference 15), Enclosure 1, pages 3 and 4, and Enclosure 2. The wording of the AOT conforms to NUREG-1433, Specification 3.3.5.1, and provides assurance that the associated trip function will remain operational following entry into the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months test AOT. (CAT 1)

7. Page 71: Current note 3 requires the LCO of Specification 3.5.A to be implemented when one of the Core Spray or Residual Heat Removal (RHR) pump timers is inoperable. This action statement inadvertently excludes consideration of the operability of these systems when in the cold condition. The revised note 3 will reference all Core Spray and RHR system requirements in TS 3.5 when their pump timers are inoperable. Additionally, the note is revised to explain that the consequences of an inoperable instrument (timer) is to disable the automatic initiation function for the pump. (CAT 4)

- Attachrunt 11 to JPN-94-0E0 SAFEW EVALUATION Page 17 of 39

8. Pages 70a,70b, and 70c: Relocate eight instruments from Table 3.2-2 to Table 3.21 since these instruments perform a primary containment isolation function, not an ECCS initiation or control function. This change will establish consistency with the STS, and resolves an NRC concem as discussed in item O.2. These instruments are: HPCI Turbine steam line high flow, HPCI steam line low pressure, HPCI Turbine high exhaust diaphragm pressure, HPCI steam line/ area temperature, RCIC Turbine steam line high flow, RCIC steam line low pressure, RCIC Turbine high exhaust diaphragm pressure, and RCIC steam line/ area temperature. (CAT 4)

O. Table 3.2-3. Instrumentation That Initiates Control Rod Bl0Chs. Pages 72 and 73 The revised Table 3.2-3 and associated notes appears on revised Pages 72,73, and 74.

1. Page 72: Revise the table to specify the " minimum operable channels per trip function." rather than the " minimum operable channels per trip system." The minimum number of operable channel requirements listed in the column are doubled to reflect this change. Also remove the phrase "for both channels" from the column for total number of instrument channels. The change reflects the as-built configuration of the control rod block initiation logic (UFSAR 7.7.4.3). The control rod block (CRB) logic is designed as a "1 out of n" logic, where n is the total number of CRB channels. Only one input need be in the trip condition on either logic to effect a rod block signal. For example, the APRM rod block trip logic is based on one out of six, not one out of three taken twice logic. The change conforms to STS (Reference 17), Table 3.3.6-1, page 3-51 and 3-52. (CAT 4)

2. Page 72: Add notes 1 and 3 notations to column heading titled: " Minimum No. of Operable instrument Channels Per Trip Function." Add note 2 notation to column heading titled: " Action." Make several editorial changes to the trip function i nomenclature to establish consistency with Table 3.1-1, and revise the title of Table

]

3.2-3. (CAT 1)

3. Page 72: Add the setpoint to full scale ratio to the trip level settings of the IRM ;

Upscale and IRM Downscale trip function. The ratio corresponds to the current setpoint specified in the table. (CAT 4)

4. Page 72 and 73: Delete current note 4 and change the minimum number of operable instrument channels per trip system requirement for the SRM trip functions from 4 to 3. This change reflects the capability as designed to bypass i any one of the four SRM channels (FSAR 7.5.4.2). (CAT 4) i

5. Pages 72 and 73: Clarify the notes for Table 3.2-3 by adopting features of the STS. The required actions for the control rod blocks are currently split between l notes 1 and 10. Further, the current required actions in note 1 appears after the I operating mode requirements for the CRBs. This presentation makes it difficult for the reader to locate the operability requirements. The following changes will improve the presentation of this material:

l

Attachment il to J' PN-94-050 SAFETY EVALUATION Page 18 of 39 Split the current note 1 into two notes (notes 1 and 2). Note 1 presents the operating mode requirements for the CRBs. This change does not change the operability requirement of the current specification. Note 2 presents the AOTs and action statements for the CRBs in a manner that conforms to the STS (Reference 17): pages 3/41-18, 3/4 3-51, and 3/4 3-52, except that the 7 day AOT for the Action B instruments (RBM) is retained since this is a plant specific value. Some minor editorial. changes have been incorporated into the action statements in note 2 to establish consistency with the required action language used for the other tables in TS 3.2. New note 2 replaces the AOTs in existing notes 1 and 10 (note 10 is therefore deleted). The change is also consistent with the NRC Safety Evaluation Report (Reference 14). The new AOTs / A,ction Statements differ in substance from the current TS as follows:

a) The required action for Action A instruments is the same as the current TS except for (1) the deletion of accelerated testing requirements for the operable system when less than the minimum nunsber of operable channel are operable, and (2) adds a one hour time period to implement a required trip. The changes conform to the STS (Reference 17).

The accelerated testing requirement is unnecessary based on the absence of a similar provision in the General Electric Licensing Topical Report GENE-770 1-A, and STl/AOT amendments approved for other plants (Reference 23).

Frequent testing will reduce instrument reliability due to increased out-of-service time to perform the testing, and increases the risk of test-induced trips. (CAT 1) b) The required action for Action B instruments (RBM) requires that the reactor not be operating in the Umiting Control Rod Pattern, which is consistent with TS 3.3.B.5 on page 94.' (CAT 1) c) The required action for Action C instruments conforms with the referenced Licensing Topical Report as described in change 0.7 below. (CAT 1)

6. Page 73: Add a 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months test AOT for the control rod block instruments as note 3 to Table 3.2-3. The AOT value conforms to GE Ucensing Topical Report GENE-770-06-1-A (Reference 6): Appendix A, page A-40; and NRC Safety Evaluation Report (Reference 14): attachment titled " Technical Evaluation Report," page 16 and 17; Enclosure 1, Table 2, and Enclosure 2, page 3-51. The change permits deletion of Specification 3.2.C.2 on page 50. The wording of the AOT conforms to NUREG-1433, Specification 3.3.2.1, and provides assurance that the associated trip function will remain operational following entry into the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months test AOT. (CAT 1) >

7. Page 73: Establish a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months repair AOT for the Scram Discharge Instrument Volume High Water Level control rod block. An AOT for this function is not currently specified (see current note 10 to Table 3.2-3). This AOT will appear in

" Action C" of revised note 2 to Table 3.2-3. Note 10 is deleted. The change conforms to GE Licensing Topical Report GENE-770-06-1-A (Reference 6):

Appendix A, pages A-41 and A-42; and NRC Safety Evaluation Report (Reference ,

14): attachment titled " Technical Evaluation Report," page 16 and 17; and Enclosure 1, Table 2. (CAT 1)

Attachment 11 to JPN-94-050 SAFETY EVALUATION Page 19 of 39

8. Page 73: Make the following editorial changes, (CAT 4)

a. Reword current note 6 to read as an operability requirement rather than a design requirement.

b. Make minor editorial changes to the nomenclature of instruments specified in current note 8 to be consistent with the table.

c. Renumber current notes 2 through 9 as notes 4 through 10, respectively.

d. Page 74 accommodates a redistribution of text.

R. Iable_3.2-7. ATWS Recirculation Pumo Trio Actuation Instrumentation. Page 77 The text for this table is redistributed onto revised pages 76 and 77.

1. Add notes 1 and 2 notation to " Minimum Number of Operable Channels per Trip System" column to reflect the repair and test AOTs. (CAT 1)

2. Change the AOT for multiple channel failures from one hour to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The change applies to the situation where there are two or more inoperable channels.

The AOT value is based on GE Licensing Topical Report GENE-770-06-1-A (Reference 6), Appendix A, page A-15; and NHC Safety Evaluation Report (Reference 14), Enclosure 1, Table 2. The proposed AOT appears as note 1, and uses text that is similar to that proposed for the RPS AOT. The AOT is permitted fr multiple channel failures only when their inoperability does not prevent the RPT inp function capability. Note 1 requires confirmation within one hour of RPT functional capability after two or more channels become inoperable, otherwise entry into the action statement is required. The AOT value for a single channel failure, and the action statement, remains unchanged. (CAT 1)

3. Add a 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months test AOT as note 2 to Table 3.2-7. The AOT value is based on GE Licensing Topical Report GENE-770-06-1-A (Reference 6), Appendix A, page A-17; and NRC Safety Evaluation Report (Reference 14), Enclosure 1, Table 2. The wording of the AOT conforms to NUREG-1433, Specification 3.3.4.2, and provides assurance that the associated trip function will remain operational following entry into the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months test AOT. (CAT 1)

4. Make minor editorial changes to the column headings and in the action statements to establish consistency with the other instrumentation tables. Also, some minor format changes are made to the table, and the title of the table is revised.

Rearrange the order of the columns to conform with the other tables. (CAT 4)

S. Iable_3.2-9. Primarv Containment _lsalatiOD3yHiem ActuationJDsimmentation Besponse Times. Page 77e Delete page 77e to remove the response time limits for the main steam isolation valve closure actuation instrumentation. The change conforms with the guidance of

. Attachm:nt il to JPN-94-050 SAFETY EVALUATION Page 20 of 39 NRC Generic Letter 93-08 regarding the transfer of response time limits from the TS to the UFSAR. (CAT 2)

T. Table 4.2-1. Niinimum Test and Calihration Frequency for PC!1 EagelB The revised Table 4.2-1 is redistributed onto revised pages 78 and 79.

1. Relocate eight instruments from Table 4.2-2 to Table 4.2-1 since these instruments perform an isolation function, not an ECCS function. This will establish consistency with the STS, and resolves an NRC concern as described previously in changes O.2 and P.8. These instruments are: HPCI Turbine steam line high flow, HPCI steam line low pressure, HPCI Turbine high exhaust diaphragm pressure, HPCI steam line/ area temperature, RClO 'urbine steam line high flow, RCIC steam line low pressure, RCIC Turbine high exhaust diaphragm pressure, and RCIC steam line/ area temperature. (CAT 4)

2. Revise the frequency of the functional tests from monthly to quarterly to conform with: (1) GE Licensing Topical Report NEDC-30851P-A, Supplement 2, (Reference 3), Enclosure 2; (2) GE Licensing Topical Report NEDC 31677P-A (References 5),

Appendix D, pages D-4 through D-8; (3) NRC Safety Evaluation Report (Reference 10), Enclosure 1, page 3; and Enciosure 2; and (4) NRC Safety Evaluation Report (Reference 13), Enclosure 1, page 2, and Enclosure 2. The change also applies to the eight instruments relocated from Table 4.2-2. The notations, as defined in proposed definition "T" on page 5, are used for all surveillance frequencies specified on the table. The PCIS functions for which the test frequencies are changed from monthly to quarterly are: (CAT 1) l e Reactor High Pressure

Reactor Low-Low-Low Water Level

Main Steam High Temperature

Main Steam High Flow

Main Steam Low Pressure

RWCU High Temperature

Condenser Low Vacuum ,

Main Steam Line High Radiation a HPCI / RCIC Steam Line High Flow

HPCI / RCIC Steam Line/ Area High Temperature ;

e HPCI / RCIC Steam Line Low Pressure l

HPCI / RCIC High Exhaust Diaphragm Pressure )

l

3. Add a notation for new note 16 that defines the quarterly calibration for the temperature sensors for the RWCU Area High Temperature PCIS trip. (CAT 4)

4. Make the following editorial changes. (CAT 4)

a. Correct a nomenclature error in the Reactor High Pressure trip function by changing " permissive" to " isolation." This trip function closes the shutdown ;

cooling isolation valves. j l

i

f Attachmen211 to JPN-94-050 SAFETY EVALUATION Page 21 of 39

b. The notations, as defined in proposed definition "T" on page 5, are used for the j surveillance frequencies specified on the page,

c. Revise the title of the table. I

d. Redistribute text to page 79.

U. Iable_41.-2.lAinimum Test.and Calibration Frequency For Cors and Containment C00110glystems. 79 and &Q The revised Table 4.2-2 is redistributed onto revised pages 80 and 81.

1. Relocate eight instruments from Table 4.2-2 to Table 4.2-1 since these instruments perform an isolation function, not an ECCS function. This will establish consistency with the STS, and resolves an NRC concem previously described in change 0.2, P.8, and T.1. These instruments are: HPCI & RCIC Steam Une High Flow, HPCI

& RCIC Steam Line/ Area High Temperature, HPCI & RCIC Steam Line Low Pressure, and HPCI & RCIC Exhaust Diaphragm Pressure High. The logic system function tests for the HPCI and RCIC Auto isolation are also moved from Table 4.2-2 to Table 4.2-1. (CAT 4)

2. Revise the frequency of the instrument functional tests from monthly to quarterly for the following instruments that initiate or control the emergency core cooling systems:

Reactor Water Level e Drywell Pressure

. Reactor Pressure

ADS - LPCI or CS Pump Disch.

HPCI & RCIC Suction Source Levels

Trip System Bus Power Monitors

Core Spray Sparger d/p The testing frequency for the other instruments on the table remain unchanged.

The notations, as defined in proposed definition "T" on page 5, are used for all surveillance frequencies specified on the table. The change for all instruments, except the RCIC system, conforms to GE Ucensing Topical Report NEDC-30936P-A, Part 2 (References 4): pages A-15 and A-16: and NRC Safety Evaluation Report

. (Reference 10): Enclosure 1, page 3. The change for the RCIC system instruments conforms to GE Licensing Topical Report GENE-770-06-2-A (Reference 8): Appendix C, page C-4-6; and NRC Safety Evaluation Report (Reference 15): Enclosures 1, page 3, and Enclosure 2. (CAT 1)

Three of the functions for which the functional test frequency is extended from monthly to quarterly, do not appear on the marked-up Technical Specification pages in the GE Licensing Topical Report referenced in the preceding paragraph. '

Two of the functions are the Trip System Bus Power Monitors and the Core Spray Sparger d/p alarm. They are absent from the mark-up pages since these instruments do not appear in the STS (Reference 17). The function of these

e l

1 Attachment 11 to JPN-94-050 SAFETY EVALUATION Page 22 of 39 instruments is limited to a monitoring function; i.e., they do not initiate an ECCS '

actuation. However, both instruments are part of the generic models analyzed in the referenced GE Licensing Topical Report. The Bus Power Mon? tors are shown in NEDC-30939P-A, Part 2, Appendix E, page E-4, E-5, and E-22. The Core Spray Sparger d/p instrument is shown in NEDC-30936P-A, Part 1, Appendix M, page M-

16. The third function is the Reactor Low Level (inside shroud) trip function. This function prevents diversion of LPCl flow to the containment sprays if there is insufficient reactor water level. Basic event " Flow from pump loop B (A) diverted" on the fault tree shown in NEDC-30936P-A, Part 1, Appendix B, page B-20 considers failure of this function. Accordingly, the quarteriy functional test frequency is applicable to these functions. (CAT 1)

3. Delete the logic system functional test for the ADS Relief Valve Bellow Pressure Switch since this trip function was eliminated when the relief valves were replaced with a valve of improved design during a previous modification. (CAT 4)

4. Revise the title of the table. (CAT 4)

V. Iable 4.2-3Jdinimum Testandfalibration FregueDry For Control Rod Block Actuations. PageJ1 The revised Table 4.2-3 is renumbered page 82.

1. Revise the freqcency of the functional testing to quarterly for the APRM, Rod Block Monitor, and Scram Discharge Volume control rod blocks to conform to GE ,

Licensing Topical Report NEDC-30851P-A, Supplement 1 (Reference 2), page A-4; ;

and NRC Safety Evaluation Report (Reference 9): Enclosure 3. The notations, as defined in proposed definition T" on page 5, are used for all surveillance frequencies specified on the table. (CAT 1)

2. Delete the requirement to perform a calibration of the SRM-Detector Not in Startup Position and IRM-Detector Not in Startup Position control rod blocks since a calibration is not applicable because these functions do not utilize analog devices.

The instrument functional test of these position switches assures operability of their associated CRB function. The change conforms to the STS in enclosure 3 of NRC '

Safety Evaluation Report (Reference 9). (CAT 4)

3. Revise the title of the table. (CAT 4) i W. _Page 84. Notes For Tables 4.2-1 Through_4 m 2-5

1. Add note 16 to define the method used to calibrate the temperature sensors for the .

RWCU Area High Temperature PCIS function. (CAT 4)

2. Delete notes 10,13. and 14, since they are not currently used in the TS. Note 10 pertains to a sampling requirement that does not exist in the Technical Specifications (Appendix A). Notes 13 and 14 notations were deleted by Amendment 181 which deleted Table 4.2-6, Surveillance instrumentation, and i

s Attachment il to JPN-94-050 SAFETY EVALUATION Page 23 of 39 incorporated Table 4.2-8, Accident Monitoring Instrumentations. That amendment deleted the SRM and IRM instruments from both the surveillance and accident monitoring categories. Consequently, the reference in note 13 to the SRM/lRM surveillance requirements in Tables 4.1-1,4.2-1, and 4.2-3 is unnecessary. Note 14 requires the safety / relief valves, listed on deleted Table 4.2-6, Surveillance Instrumentation, to be functionally tested once each operating cycle. The note is unnecessary since this testing requirement was transferred to Table 4.2-8, Accident Monitoring instrumentation, by Amendment No.181. Deletion of these notes does not impact current TS requirements. (CAT 4) ,

3. Delete the last sentence in note 7 which reads " Where possible all logic system functional tests will be performed using the test Jacks." This level of detail regarding the technique to be used to perform a test is not appropriate in the TS.

i (CAT 4)

X. Table 4.2-7. Minimum Test and Calibration Frequency For ATWS Recirculation '

Pump Trip Actuation Instrumentation. Page35_

1. Revise the frequency of the channel functional test requirement from monthly to quarterly. This change is based on GE Licensing Topical Report GENE-770-06 A (Reference 6), Appendix A, page A-19; and NRC Safety Evaluation Report (Reference 14), Enclosure 1. Table 2; and Enclosure 2, page 3-40. The notations, as defined in proposed definition "T" on page 5, are used for all surveillance frequencies specified on the table. (CAT 1)

2. Revise the titie of the table. (CAT 4)

Y. Eigure 4.21. Test intervab vs. Probability of System Unavailability. Page 87 Page 87: Delete Figure 4.2-1 since it is referenced in text deleted by change N.1.

(CAT 1)

Z. Section.LQHeleIences._Eage_285 Delete reference 6 since it is used in text deleted on page 36, and does not appear elsewhere in these TS. (CAT 1)

- Attachment 11 to JPN-94-050 SAFETY EVALUATION Page 24 of 39

11. SAEEIUMP11CAllDNS_0E.IHE PROPOSED _CBANGES The proposed changes to the James A. FitzPatrick Technical Specifications are grouped into four categories. The safety implications associated with each of these change categories are as follows:

1. Incorporate STI aud.AOT Improvatnents_ .CategoIy 1 The amendment will extend the Reactor Protection System (RPS), Primary Containment isolation System (PCIS), Emergency Core Cooling System (ECCS),

Control Rod Block (CRB), and Recirculation Pump Trip (RPT) instrumentation functional test intervals from weekly or monthly to quarterly as described in the previous section. The interval for the functional test of the automatic scram contactors will change from "every refueling outage" to " weekly" using either the channel functional test or the RPS Channel Test Switches. The bases for this change is documented in General Electric Licensing Topical Report NEDC-30851P-A (Reference 1) which concludes that a common cause failure of the scram contactors is a major contributor to RPS unavailability.

Additionally, allowable out-of-service times, as described in the previous section, are specified for the instrumentation These times, specified separately for both repair and test situations, represent the time that the instrument may be rendered inoperable before entry into its associate Limiting Condition For Operation action statements.

The bases for these changes are presented in generic evaluations developed by the BWR Owners Group, submitted to the NRC in seven GE Licensing Topical Reports (References 1 through 7), and approved by NRC Safety Evaluation Reports (References 8 through 15). These generic evaluations are applicable to the FitzPatrick design as documented in References 26 through 31.

The generic evaluations utilize fault tree modeling to estimate the impact of the proposed STI and AOT changes on the instrument system failure frequency. The acceptance criteria used in the analyses for the proposed changes is based on a net change in risk. The net change in risk is the difference between the increase in risk that results from the longer STI / AOT, and the decrease in risk that results from the reduced likelihood of inadvertent scrams due to a reduction in testing. The generic evaluations concluded that the net change in risk is negligible. Further, the evaluations concluded that the overall effect of the proposed changes provides a net increase in plant safety when all factors are considered. This increase in plant safety is achieved by reducing the potential for: (a) unnecessary plant scrams (reduced challenges to plant shutdown systems and improved plant availability); (b) excessive test cycles on equipment (reduced wear-out potential); and (c) diversion of plant personnel and resources on unnecessary testing (potential safety and operational improvement).

The NRC Safety F. valuations Reports (References 8 through 15) concluded that the GE Licensing Topical Reports provide an acceptable generic basis for supporting plant-specific Technical Specification changes. The SER further states that each applicant for a license amendment must confirm the plant-specific applicability of the

I Attachment 11 to JPN-94-050 SAFETY EVALUATION Page 25 of 39 generic evaluations. A plant specific evaluation was performed for each instrumentation category. The evaluations compared the instrumentation design configurations and surveillance requirements to the generic model used in the GE Ucensing Topical Reports. Differences were identified and assessed as to their impact on instrumentation reliability calculated for the generic models. Additionally, the setpoint drift expected under the extended STis was evaluated to determine the acceptability of current setpoint calculations. Provided below is a summary of the plani specific evaluations performed by the Authority (References 26 through 31) that address the proposed changes.

SjtrDmary_of the Resujts of the Plant Soe.cific_EYaluation for the RP_S_ System

a. SER Requirement Confirm the applicability of the generic analysis for NEDC-30851P-A to the plant.

Bant-Specific _ Evaluation The FitzPatrick plant is a BWR/NSSS with a RPS relay instrumentation design which is very similar to the base case model used in the generic RPS reliability evaluation (NEDC-30851P-A).

b. S.ER Requiremeal Demonstrate by use of current drift information provided by the equipment vendor or plant-specific data, that the drift characteristics for instrumentation used in the RPS channels in the plant are bounded by the assumptions used in l NEDC-30851-P-A when the functional test interval is extended from monthly to quarterly.

The NRC staff provided additional guidance in a letter from C. E. Rossi (NRC) to R. F. Janecek (BWR Owners Group), dated April 27,1988, which states that:

"... licensees need only confirm that the setpoint drift which could be expected under the extended STls has been studied and either (1) has been shown to remain within the existing allowance in the RPS and ESFAS instrument setpoint calculation of (2) that the allowance and setpoint have been adjusted to account for the additional expected drift."

Baat-Specific Response ,

Actual plant setpoints are set conservatively with respect to the Technical Specification limits to accommodate for the uncertainty associated with each particular instrument channel and associated test interval. Trip settings are adjusted, as necessary, when the instrument channel is calibrated and/or functionally tested. At the FitzPatrick plant, RPS calibrations, except for the LPRMs, occur at three months or longer intervals. Since the setpoints are set to accommodate the longer calibration intervals, the setpoint drift will be

l Attachment il to JPN-94-050 SAFETY EVALUATION Page 26 of 39 acceptable between the quarterly functional test. The LPRM output is checked daily using a heat balance and adjusted as necessary to maintain an acceptable level of setpoint drift.

Confirmation that postulated setpoint drift associated with the longer functional test interval is within existing safety margins was established by reviewing the instrument calibration and functional test results (Reference 26).

c. SER Requitement Confirm that the differences between the parts of the RPS that perform the trip functions in the plant and those of the base case plant were included in the analysis for the plant using the procedures of Appendix K of NEDC-30851-P-A, or provide plant-specific analyses to demonstrate that there is no appreciable change in RPS availability or public risk.

Plant -Specdic_ Evaluation NEDC-30851P-A concludes that RPS failure frequency is predominately a function of common cause failures of the scram contractors. There are no significant variations between the FitzPatrick RPS scram contractor configuration and the generic model. The FitzPatrick RPS differs from the generic model in sensor logic, sensor relays, scram valve design, backup and manual scram actuation, and technical specification requirements. These differences were identified, and evaluated in accordance with NEDC-30851P-A, Appendix K, procedures (Reference 26). This evaluation found the design l differences to have a negligible effect on RPS failure frequency, based on the case studies presented in NEDC-30851P-A, Appendix K.

l Summary of the Results of the PlantSpecific Evaluations for the Othetlnstruments !

a. SER Regulrement l Confirm tk applicability of the generic analyses to the plant.

Plant-Specific ResppDSB ,

The FitzPatrick plant is a BWR/4 NSSS with an ECCS, PCIS, CRB, and RPT l relay instrumentation design which is very similar to the BWR 3/4 base case I model used in the generic instrumentation reliability analyses (NEDC-30851P-A, l Supplements 1 and 2, NEDC-30936P-A, NEDC-31677P-A, GENE-770-06-1, and GENE-770-06-2-A).

The FitzPatrick ECCS design was compared to the design considered in the generic evaluation in accordance with the procedure provided in Appendix F of l NEDC-30936P-A, Part 2 (References 27 & 28). The differences identified are ;

minor in nature and have a negligible impact on water injection failure )

frequency. The differences are within the boundary conditions of the generic !

model or are bounded by envelope case 4A of the generic evaluation. ,

l i

Attachment 11 to JPN-94-050 SAFETY F. VALUATION Page 27 of 39 There are no significant differences between the generic model (NEDC-31677P-A) and the FitzPatrick design for PCIS instrumentation common to the RPS.

The only significant differences between the generic model and the FitzPatrick design for non-common PCIS instrumentation are the number of sensor variables that isolate RWCU, and the logic for actuating Secondary Containment isolation. These differences were evaluated using the case studies contained in NEDC-31677P-A and were found to have an insignificant effect on isolation functiori failure frequency (Reference 29).

Their are no significant differences between the generic model and the FitzPatrick design for CRB instrumentation. The generic evaluations (References 1 and 7) for the CRB function are applicable to plants with either Rod Block Monitors (RBM) or Rod Withdrawal Limiters (RWL) since they provide equivalent protection. The Fitzpatrick plant uses the Rod Block Monitor (RBM) for the CRB function (Reference 30).

The BWR/4 RPT system design considered in the generic evaluated (GENE-770-06-1), is a two-out-oi-two taken once logic per trip system for either reactor low water level or reactor high pressure. The Fitzpatrick RPT system is initiated by a one-out-of-two logic taken twice, from the same signals. The Fitzpatrick logic design is bounded by the generic evaluation since it is inherently more reliable (Reference 31).

b. SER Reauiremer11 Confirm that any increase in instrument drift due to the extended STis is ;

properly accounted for in the setpoint calculation methodology. (For additional information on this issue, see letter from C. E. Rossi to R. F. Jancek, dated April 27,1988)

Bant-Specific Response Actual plant setpoints are set conservatively with respect to the Technical Specification limits to accommodate for the uncertainty associated with each particular instrument channel and associated test interval. Trip settings are adjusted as necessary when the instrument channel is calibrated and/or .

functionally tested. The FitzPatrick plant calibrates the subject equipment et intervals of three months or longer. Since the setpoints are set to accommodate the longer calibration intervals, the setpoint drift will be acceptable between the quarterly functional tests. Therefore, changing the functional test interval from weekly or monthly to quarterly does not impact the :

TS trip settings.

Confinnation that postulated setpoint drift associated with the longer functional test interval is within existing safety margins was established by reviewing the instrument calibration and functional test results (References 27, 29, 30, and 31).

l l

- Attachment 11 to JPN-94-050 )

SAFETY EVALUATION ,

Page 28 of 39 Mowable Out-of-Service Time Bases The test allowable out-of-service times (AOT) provide a reasonable period of time to perform testing on an instrument channel made inoperable to perform required surveillance. An instrument channel removed from service (made inoperable) to ,

perform required surveillance must be restored to service (made operable) in less than or equal to the six hour AOT, unless the instrument channel is found to be in need of repair At that time, the instrument channel is declared inoperable and the repair AOT is entered.

When the repair AOT is entered for an inoperable instrument channel, a verification that sufficient instrument channels remain operable or tripped to maintain trip capability for that trip function is performed (if one or more additional instrument channels for the same trip function are also inoperable) and, within the time period specified in the repair AOT, the instrument channel: a) must be restored to service in an operable status or, b) must be placed in a tripped status or, c) the associated trip system must be tripped. If the required actions are not completed within the time-period specified by the repair AOT, then the specified action for that trip function must be taken within the time period stated in the ACTION (e.g., insert all operable control rods within four hours; isolate the main steam lines within eight hours; declare the associated ECCS inoperable; place the reactor in the startup/ hot standby mode within the next six hours).

When the repair AOT is entered, the entire time period specified by the AOT is available, even though it might become apparent that the repair cannot be completed within the allowed time, because analysis performed in support of the Licensing Topical Reports assumed that the entire time period specified in the AOT was used in each case. While it is expected that the entire time period specified by the AOT will seldom be needed, allowing the entire time period specified to be used in determining when required actions must be completed avoids situations where interpretation is necessary, and simplifies training in the use of the new AOTs.

The values for the proposed test ano repair AOTs are based on the Ucensing Topical Reports previously referenced. The AOT text proposed for the five instrument groups (RPS, PCIS, ECCS, CRB, and ATWS RPT) differ to reflect: (1) each instrument group's unique AOT value as presented in the Topical Reports , (2) the action statements associated with the function of the instrument group, and (3) distinctions in the trip logic of each instrument group. The objective of the proposed Technical Specifications is to optimize consistency in the AOT text for the different instrument groups to the extent possible. Compliance with this objective minimizes the complexity of the TS which will improve operator understanding of the AOTs, reduce the potential for a TS violation resulting from a misinterpretation, and simplify operator training in the use of the TS.

Accordingly, a similar repair AOT text was applied to four of the instrument groups (all except the CRB). The text conforms to the AOT approved for the RPS AOT for Duane Amold (Reference 23), and Nine Mile Point 1 and 2 (References 33 and 34),

except for some minor editorial changes. The AOT recognizes the impact of multiple channel failures on trip function operability. The LCO for the AOT precludes the use '

of an AOT in excess of one hour for those situations where the multiple channel

Attachmsnt 11 to JPN-94-050 SAFETY EVALUATION Page 29 of 39 failures reder the trip function inoperable. Only when the failure does not prevent trip function capebility will the full AOT recommended by the Topical Reports be permitted. In this nonner, the application of the AOT is maximized without jeopardizing the operat>My of the trip function. Additionally, this approach minimizes the potential for operation with an instrument channel or trip system in the tripped condition which increases the potential for a plant transient. The LCO for the AOT meets these objectives by requiring, within one hour of a multiple instrument channel failure, confirmation of trip function capability. If confirmation cannot be satisfied within the one hour period, the action statement is entered. This AOT concept is not applicable to the CRB instrument group due to its unique design; i.e., only one trip system and no action required beyond placing the channel in the tripped condition.

The restrictions on the use of the repair AOT for multiple channel failures conforms to the STS in NUREG-1433 (Reference 18). The STS limits the AOT to one hour if multiple channel failures results in loss of trip capability. Refer to the following sections of NUREG-1433.

RPS LCO 3.3.1.1.C PCIS LCO 3.3.6.1.B ATWS-RPT LCO 3.3.4.2.B ECCS LCO 3.3.5.1.B, C, D, E, F, and G All of the instrument groups use the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months test AOT based on the Topical Reports previously referenced. The text of the test AOT conforms to the text used in NUREG-1433 (Reference 18), and approved for Duane Amold (Reference 23) and NMP-2, except as noted below. For multiple channel trip systems, the AOT is permitted only when the removal of the channel from service to test does not prevent trip function capability. Only single channel trip systems, and trip systems based on a two-out-of-two-taken-once logic, as specified in the AOT statement, are excluded from this restriction. Several deviations from the AOT text used in NUREG-1433 are proposed for the PCIS trip functions. These deviations are described, along with their bases, in change description O.7.

2. Belocation of Instrument Response Time Limits - CategoIy_2 The change deletes the RPS and MSIV isolation instrumentation response time limits from the Technical Specifications. The July 1995 UFSAR update will incorporate the response time limit requirements. These changes conform with the NRC guidance presented in Generic Letter 93-08 (Reference 24). The surveillance requirement in ,

TS 4.1.A and TS 4.2.A that confirms the response time limits every eighteen months remains unchanged. The plant surveillance procedures for response time testing (ISP-102 through ISP-111) include acceptance criteria that reflects the response time limits in Tables 3.1-2 and 3.2-9.

Incorporation of these requirements into the UFSAR will assure that the NRC is maintained informed of this design feature. Any changes to the response time limits contained in the UFSAR will be evaluated under 10 CFR 50.59.

E

Attachm:nt il to JPN-94-050 SAFETY EVALUATION Page 30 of 39 :

There are no safety considerations associated with this change since it does not involve any changes to the response time limits, surveillance requirements, or procedural changes that impact plant operations.

3. Relele APRM Downscale Scram Function - Category]

Although the APRM downscale trip is listed as an RPS scram (Tables 3.1-1 and 4.1- l 1), it does not directly initiate a reactor trip. The trip performs an interlock function associated with the IRM high flux scrarn. The interlock prevents the IRM high flux scram function from being defeated in the Run operating mode until the APRM downscale setpoint has cleared (power above 2.5%).

The bases for deleting the APRM downscale scram function is presented in a :

General Electric evaluation performed for the Dresden and Quad Cities plants ,

(Reference 22) and is repeated in the following paragraphs. This evaluation is applicable to the FitzPatrick plant based on similarity to the Dresden design. This function existed on several early plants but has been deleted from later plant designs. It is no longer required by the STS and has been deleted from the Technical Specifications of several plants that originally included it in their design. '

References 20 and 21 identify two license amendments that removed this scram feature from other BWR plants.

Deletion of the APRM downscale scram function will permit all available combinations of inoperable IRM and APRM channels to be simultaneously bypassed, as intended by the plant design (UFSAR 7.5.5.3 and 7.5.7.4). Due to the different number of APRM and IRM channels (six vs. eight), some IRM channels share the same APRM channel in the APRM downscale scram logic. Consequently, some bypass combinations of inoperable IRM and APRM channels would result in less than the minimum number of required operable APRM downscale scram trips, precluding bypass capability for one of the inoperable channels. Under these circumstances, the plant must remain in a " half scram" condition. Removal of this 1 trip function will avoid the need to operate the plant in the " half scram" condition, with l the associated risk of a plant transient, for certain inoperable IRM/APRM i combinations. l l

Removal of the APRM downscale scram function from the Technical Specifications is not a safety concem for the following reasons-

a. The design basis accident in this region of operation is the control rod drop accident (CRDA). The only scram function that the FSAR takes credit for in the mitigation of the CRDA is the APRM 15% power fixed high neutron flux scram (startup mode) which is assumed to occur at 120% power (UFSAR 14.6.1.2).

b. If the mode switch is changed to the Run mode prematurely during startup, or if the reactor power is reduced too far before changing the rnode switch to the !

Startup mode, the control rod block associated with the APRM downscale trip l will activate, precluding further control rod withdrawal. This control rod block feature is required by the Technical Specifications (Table 3.2-3), and is not ,

altered by the requested change. I 1

Attachm:nt 11 to JPN-94-050 SAFETY EVALUATION Page 31 of 39

c. During cold plant startups, prematurely changing to the Run mode will likely result in MSIV closure and consequent scram due to insufficient steam pressure.

Currently, the plans are to implement a modification to remove the APRM downscale scram function, contingent on NRC approval of this amendment request, during refueling outage Reload 12/ Cycle 13 (scheduled to start in early 1997). The surveillance requirements currently specified for the APRM downscale scram function will continue during the period between the issuance of the amendment and the completion of the modification removing this function.

4. Miscellaneous Changes - CategofyJ The miscellaneous changes are grouped into four subcategories and evaluated as follows:

a. Editorial Changes Editorial changes include the relocation of text, renumbering of table notes, and corrections in nomenclature. These changes are necessitated by other text changes in this application, and do not change any Technical Specification requirement. Changes: A.1, A.2, B.2 D.2, D.5, F.9, F.10, F.11, F.12, F.13, F.14, G.2, G.4, G.10,1.3, 0.1, 0.10, P.1, Q.3, O.8, R.4, T.4, U.4, V.3, W.2 and X.2.

b. Clarifications The clarification changes either (1) establish consistency with the actual design of the plant instrumentation as described in the UFSAR, (2) establish consistency between different TS requirements, or (3) define TS requirements more explicitly. By enhancing the accuracy and clarity of the text, the changes will assure a correct interpretation of the TS instrumentation requirements, essential to the safe operation of the plant. The changes in this group are as follows:

(1) Use of notations for the surveillance frequencies assures a precise and cleariy defined time interval. The change conforms to the STS (Reference 17). Change: B.I.

(2) Establish consistency among the Limiting Conditions For Operation (LCO) for related systems by conforming to the more conservative LCO.

Changes: P.S.b, and P.7.

(3) Changes to reflect the actual design of instrumentation described in the UFSAR.

This includes the change that relates the requirement for the minimum number of operable control rod block channels to " trip function" rather

Attachment il to JPN-94-050 SAFETY EVALUATION Page 32 of 39 than " trip system." While the rod block logic circuitry is arranged as two similar logic circuits, (i.e., one-half of the sensors provide inputs to one logic circuit, that other half to the other logic circuit), either logic circuit independently provides a rod block signal to inhibit rod withdrawal. The current Technical Specifications splits the channels into two groups for each CRB function for the purpose of establishing the minimum number of operable instrument channels per trip system. This prohibits the bypass of more than one channel per group. Under the proposed change, the total number of channels within a CRB function that may be bypassed will remain at two. However, any two of the channels could be bypassed without consideration of its grouping. Therefore, the reliability of the CRB function will not be compromised by the change.

The change will permit the bypass of all available combinations of two IRM channels or two APRM channels using their associated bypass switches, as intended by the plant design (UFSAR 7.5.5.3 and 7.5.7.4).

As wired, certain combinations of bypassed IRM or APRM channels are not permitted by the current CRB Table 3.2-3. For these combinations, inoperable channels resulting in a " half-scram" condition cannot be bypassed. This constraint on the ability to bypass all available combinations of two IRM channels, or two APRM channels, increases the potential for an unnecessary plant transient. The change is consistent with the design of the control rod block system (UFSAR 7.7.4.3), and with the STS (Reference 17), Table 3.3.6-1, page 3-51.

Changes: F.4, G.3, M.1, M.2, M.3, 0.3, 0.4, 0.5, 0.6, P.3, P.4, P.5.a, 0.1, 0.4, U.3, V.2 and W.3.'

(4) Limiting Condition for Operation 3.0.F is added to clearly define the conditions under which inoperable equipment, or equipment removed from service, may be returned to service for the sole purpose to perform testing to demonstrate its operability or operability of other equipment. The LCO requires the application of administrative controls to limit the time to that necessary to perform the testing. Use of this provision permits the repair ,

and retum of equipment to service following the appropriate testing, and i reduces the potential of plant transients that result from operation with instrument channels in the tripped condition. Changes: C, D.4.

5 (5) A note is added to the quarterly calibration requirement for the RWCU Area High Temperature instrument channel in Table 4.2-1 to describe the i

~ '

calibration method for the temperature sensors. The method consists of comparing the active temperature signal with a redundant temperature signal.

Injection of a known temperature signal into RTD's and thermocouples to perform a calibration cannot be performed from a remote location due to I the nature of these devices. The temperature sensors for most instrument and controls systems are calibrated by removing the sensor and placing it into a known temperature bath. However, this method of calibration for the eight RWCU Area High Temperature sensors conflicts with ALARA 1

! )

7 Attachmant il to JPN-94-050 '

SAFETY EVALUATION -

.Page 33 of 39 7

practices since entry into high radiation areas (ranges from 100.t0 500 E mr/hr in the general area of the temperature sensor) would be required to remove, test, and replace the sensors. For this reason, a comparison method of calibration is used. This involves a comparison of the output of the temperature sensors used for the trip function with another temperature sensor that monitors the same parameter. The acceptance p criteria currently used is s 10 F differential. Change: T.3 and W.1

c. Relocate Instrumentation Requirements From the ECCS Section to the PCIS Section.

This change relocates the operability and surveillance requirements for the HPCI and RCIC isolation actuation instrumentation from the ECCS tables (Tables 3.2-2 and 4.2-2) to the Primary Containment Isolation tables (Tables 3.2-1 and 4.2-1). The proposed location is the appropriate location for these -

instruments, and conforms to the STS (Reference 17). The change resolves an NRC concem expressed in Reference 19 that the current location does not assure that the instruments are operable when primary containment integrity is _

required. While the Authority interprets the current TS as requiring the operability of the HPC! / RCIC isolation instruments when containment integrity-is required, the change will enhance the clarity of this requirement. Changes:

0.2, P.8, T.1, and U.1.

Ill. EVALUATION OF SIGNIELCANT HAZARQS_C_QNSIDERATION Operation of the FitzPatrick plant in accordance with the proposed Amendment would not involve a significant hazards consideration as defined in 10 CFR 50.92, since it would not:

1. involve a significant increase in the probability or consequences of an accident previously evaluated because:

a. Incoroorate STI and AOT Improvement - Categorv 1 The proposed changes are limited to an extension of the surveillance testing intervals and allowable out-of-service times of plant instrumentation. The changes do not introduce any new modes of plant operation, make any physical changes, or alter any operational setpoints. Therefore, the changes do not degrade the performance of any safety system assumed to function in the accident analysis. Consequently, there is no effect on the probability of occurrence of an accident.

Regarding the consequences of an accident, the GE Ucensing Topical Reports (References 1 through 7) concluded that the proposed extensions in the STI and AOT for the safety system instrumentation results in an insignificant change in the core damage frequency. The extension of the STI / AOTs results in a slight increase in the unavailability of the safety functions. However, this effect is offset by a reduction in the probability of inadvertent plant trips due to

Attachment il to JPN-94-050 SAFETY EVALUATION Page 34 of 39 reduced testing. The overall effect on the probability of an accident is negligible. While the effects of reducing unnecessary cycles on safety system instrumentation is not quantifiable, the effect will be to further reduce the core damage frequency. The NRC concurred in their SERs (References 8 through

15) with these conclusions. Consequently, there is not a significant increase in the consequences of an accident.

b. Helocation of the InstrumeD1_BBSp0Dse_ lima. Limits _ Cittegorv 2 The change involves the use of an attemate regulatory process for controlling the instrument response time limits. The change does not introduce any new modes of plant operation, make any physical changes, alter any operational setpoints, or change the surveillance requirements.

c. Delete APRM Downscale_StrlLm - CategoIyj The design basis accident applicable to the startup power region is the Control Rod Drop Accident (CRDA). The FSAR does not credit the APRM downscale scram in the termination of this accident. Accident mitigation is provided by the APRM fixed high neutron flux scram. Therefore, elimination of this scram function has no adverse affect on previously evaluated accidents.

d. MiscellaneousStianges - CategoryJ The changes do not introduce any new modes of plant operation, make any physical changes, or alter any operational setpoints. The changes involve enhancements that clarify the Technical Specification requirements. ,

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

a. Incorocrate STI and AOT improvements - Categorv_1 The proposed changes do not introduce any new accident initiators or failure mechanisms since the changes do not introduce any new modes of plant operation, make any physical changes, or alter any operational setpoints. The changes reduce the probability of accidents initiated by test-induced plant trips.

b. Belocation_of the Response Time Limits - Categoty_2 - j J

The change involves the use of an attemate process for controlling the )

instrument response time limits. The change does not introduce any accident initiators since it does not involve any new modes of plant operation, make any physical changes, alter any operational setpoints, or change the surveillance requirements.

c. Delete APRM Downscale ScIam - C.ategolyl Scram functions are intended to shutdown the reactor following transients or l

1

Attachment il to JPN-94-050 SAFETY EVAL.UATION Page 35 of 39 accidents and their removal does not introduce an accident initiator. The limiting accident evaluated in the FSAR for the startup power region is the control rod drop accident. This accident is assumed to occur irrespective of the scram functions provided to terminate this accident.

d. Miscellaneous _ Changes - Category _4 The changes do not introduce any new accident initiators or failure mechanisms since the changes do not alter the physical characteristics of any plant system or component. The changes involve enhancements that clarify the Technical Specification requirements.

3. involve a significant reduction in the margin of safety because:

a. IDcD_rporate STI and AOT improvements _CategoIy_1 The proposed changes do not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The affected instrumentation setpoints already account for the effects of drift and include sufficient allowance for an extension in the STis. The evaluations presented in the referenced Licensing Topical Reports concluded that the overall effect of the proposed changes provides a net increase in plant safety.

The improvement is achieved by reducing the potential for: (a) test related plant scrams (reduced challenges to plant shutdown systems and improved plant availability); (b) excessive test cycles on equipment (reduced wear-out potential); (c) operator errors (AOT provides reasonaole time for making repairs and tests); (d) scrams that occur when inoperable channels are tripped because insufficient repair time exists; and (e) diversion of plant personnel and resources on unnecessary testing (potential safety and operational improvement).

b. Belocation of the Response Time Limits - Category 2 _

The change involves the use of an attemate regulatory process for controlling the instrument response time limits. The change does not introduce any new

, modes of plant operation, make any physical changes, alter any operational setpoints, or change the surveillance requirements.

c. Relete APRM Downscale Scram - Categoly_3 The only scram function that the UFSAR takes credit for in the mitigation of the limiting accident (control rod drop accident) is the APRM 15% power fixed high neutron flux scram. This scram function, as well as the IRM high flux scram function in the startup mode which could also terminate this accident, are not affected by this change. Only the APRM downscale scram, for which the UFSAR takes no credit in the termination of any analyzed event, is eliminated by this change. The APRM downscale control rod block is not affected by this change. Elimination of the APRM downscale scram will avoid the need to operate the plant in a " half scram" condition for certain IRM/APRM channel bypass combinations, therefore eliminating the potential for an inadvertent plant

'* - Attachment 11 to JPN-941 050 ]

- SAFETY EVALUATION Page 3G of 39 '

transient. For these reasons, the change does not involve a significant reduction in the safety margin.

d. Miscellan'eous Changes CateQQIyl The changes assure compliance with the Technical Specifications by improving its clarity and accuracy. For these reasons the changes will improve the plant's .

margin of safety.

IV. IMPLEMENTATION OF THE PROPO_ SED _CtLANGES Implementation of the proposed changes will not adversely affect the ALARA or Fire Protection Program at the FitzPatrick plant, nor will the changes impact the environment.

V. COECLQSlQN

' The changes, as proposed, do not constitute an unreviewed safety question as defined in 10 CFR 50.59. That is, they:

1.- will not increase the probability nor the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report;

2. will not increase the possibility of an accident or malfunction of a type different from any previously evaluated in the Safety Analysis Report; and

3. will not reduce the margin of safety as defined in the basis for any technical specification.

The changes involve no significant hazards consideration, as defined in 10 CFR 50,92.

VI. - BEEEBENCES

1. GE Topical Report NEDC-30851P-A, " Technical Specification improvement Analyses for BWR Reactor Protection System," March _1988.

2. - GE Topical Report NEDC-30851P-A', Supplement 1 " Technical Specification improvement Analyses for BWR Control Rod Block Instrumentation," October 1988.

3. GE Topical Report NEDC-30851P-A, Supplement 2 " Technical Specification Improvement Aaalyses for BWR lsolation instrumentation Common to RPS and ECCS Instrumentation," March 1989.

4. GE Topical Report NEDC-30936P-A , Parts 1 and 2, "BWR Owners Group Technical Specification improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation)," December 1988.

Attachm:nt 11 to JPN-94-050 l SAFETY EVALUATION Page 37 of 39

5. GE Topical Report NEDC-31677P-A , " Technical Specification improvement Analysis for BWR lsolation Actuation Instrumentation)," July 1990.

6. GE Topical Report GENE-770-06-1-A, " Bases for Changes to Surveillance Test intervals and Allowed Out-Of-Service Times For Selected Instrumentation Technical '

Specifications," December 1992.

7. GE Topical Report GENE-770-06-2-A, " Addendum tc Bases for Changes to Surveillance Test Intervals and Allowed Out-Of-Service Times For Selected Instrumentation Technical Specifications," December 1992. <

8. NRC Safety Evaluation Report, iettcr from Ashok C. Thadani, NRC to T. A.

Picken, BWR Owners Group, " General Electric Co. Topical Reports NEDC-30844, BWR Owners Group Response to NRC Generic Letter 83-28, and NEDC-30851P, Technical Speci!! cation Improvement Analysis for BWR RPS," July 15,1987.

9. NRC Safety Evaluation Report, letter from Charles E. Rossi, NRC to D. N. Grace, BWR Owners Group, " General Electric Company Topical Report NEDC-30851P, Supplement 1, Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation," September 22,1988.

10. NRC Safety Evaluation Report, letter from Charles E. Rossi, NRC to D. N, Grace, BWR Owners Group, " General Electric Company Topical Report NEDC-30851P-A, Supplement 2, Technical Specification improvement Analysis for BWR isolation !

Instrumentation Common to RPS and ECCS Instrumentation," January 6,1989.

11. NRC Safety Evaluation Report, letter from A. Thadani, NRC to D. N. Grace, BWR Owners Group, " General Electric Company Topical Report NEDC-30936P, BWR Owners Group Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation), Part 1," December 9, )

1988.

12. NRC Safety Evaluation Report, letter from Charles E. Rossi, NRC to D. N. Grace, BWR Owners Group, " General Electric Company Topical Report NEDC-30936P, BWR Owners Group Technical Specification Improvement Methodology (with I Demonstration for BWR ECCS Actuation Instrumentation), Part 2," December 9. l 1988.

l

13. NRC Safety Evaluation Report, letter from Charles E. Rossi NRC to S. D. Floyd, BWR Owners Group, " General Electric Company Topical Report NEDC-31677P, Technical Specification Improvement Analysis for BWR isolation Actuation Instrumentation", June 18,1990.

14. NRC Safety Evaluation Report, letter from Charles E. Rossi, NRC to R. D. Binz, BWR Owners Group, " General Electric Company Topical Report GENE-770-06-1, Bases for Changes to Surveillance Test intervals and Allowed Out-of-Service Times for Selecteci instrumentation Technical Specifications," July 21,1992.

15. NRC Safety Evaluation Report, letter from Charies E. Rossi, NRC to G. J. Beck, BWR Owners Group, " General Electric Company Topical Report GENE-770-06-2, !

e Attachment 11 to JPN-94-050 SAFETY EVALUATION Page 38 of 39 Addendum to Bases for Changes to Surveillance Test intervals and Allowed Out-of-Service Times for Selected instrumentation Technical Specifications (BWR RCIC ;

Instrumentation), July 30,1992. ,

16. NRC letter, C. Rossi, NRC to G. Beck, BWROG, dated July 26,1992.

17. NUREG-0123 " Standard Technical Specifications in General Electric Boiling Water Reactors (BWR/5)," Revision 3, dated Fall 1980.

18. NUREG-1433, " Standard Technical Specifications for General Electric Boiling ;

Water Reactors (BWR/4)", Revision 0, dated September 1992.

19. NRC Inspection 50-333/88 January 12 to March 7,1988 - Routine Inspection of Plant Activities, dated March 29,1988.

20. NRC letter, D. C. Scaletti, NRC to D. M. Musolf, Northem States Power Co.,

regarding issuance of Amendment 100 to Monticello, dated August 26,1987.

21. NRC letter, B. Siegal, NRC to H. E. Bliss, Commonwealth Edison Co., regarding issuance of Amendment 50 to Dresden, dated August 24,1988.

22. GE letter, J. A. Miller, Services Project Manager, to E. D. Enigenburg, Dresden Nuclear Station, dated August 26,1987.

23. NRC letter, R. M. Pulsifer, NRC to L. Liu, Iowa Electric Light and Power Co.,

regarding issuance of Amendment 193 for Duane Arnold Energy Center, dated April 14,1993. .

24. NRC Generic Letter 93-08,

Relocation of Technical Specification Tables of Instrument Response Time Limits," December 29,1993.

25. James A. FitzPatrick Nuclear Power Plant Updated Finci Safety Analysis Report, Chapter 7.

26. JAFNPP Plant Specific Evaluation: "RPS Reliability Based Surveillance Test improvements," Report No. JAF-RPT-RPS-01384, Rev.1, dated May 6,1994.

27. JAFNPP Plant Specific Evaluation: "ECCS Actuation instrumentation Reliability Based Surveillance Test improvements," Report No. JAF-RPT-MULTI-01426, Rev.

1, dated March 28,1994.

28. JAFNPP Plant Specific Evaluation: " Miscellaneous Instrumentation Surveillance Test improvements," Report No. JAF-RPT-MISC-01477, Rev. O, dated March 28, 1994.

29. JAFNPP Plant Specific Evaluation: "PCIS Reliability Based Surveillance Test improvements," Report No. JAF-RPT-PC-01425, Rev. O, dated January 19,1994.

30. JAFNPP Plant Specific Evaluation: "CRB Instrumentation Surveillance Test improvements," Report No. JAF-RPT-MULTI-01420, Rev. O, date- J January 12,1994.

Is Attachment ll to JPN-94-050 SAFETY EVALUATION Page 39 of 39

31. JAFNPP Plant Specific Evaluation: " Recirculation Pump Trip Instrumentation Reliability Based Surveillance Test improvements," Report No. JAF-RPT-RWR-01434, Rev.1, dated March 29,1994.

32. NRC letter, C. E. Rossi to R. F. Janecek, BWR owners Group, " Staff Guidance For Licensee Determination That The Drift Characteristics For Instrumentation Used in RPS Channels Are Bounded By NEDC-3085P Assumptions When The Functional Test Interval is Extended From Monthly To Quarterly," dated April 27,1988.

33. NRC letter, D. S. Brinkman to B. R. Sylvia, Niagara Mohawk Power Co., regarding issuance of Amendment 139 for Nine Mile Point Nuclear Station Unit No.1, dated February 24,1993.

34. NRC letter, J. E. Menning to B. R. Sylvia, Niaaara Mohawk Power Co., regarding issuance of Amendment 41 for Nine Mile f aint Nuclear Station Unit No. 2, dated May 11,1993.

<g

. c, J.

ATTACHMENT 111 to JPN-94-050 MARKED-UP TECHNICAL SPECIFICATION PAGES FOR !

EBOP_OSEDlECHNICAkSEEnlEICAT10N_ CHANGES INSTRUMENTATION SURVEILLANCE TEST INTERVALS, ,

ALLOMABLE_ QUI-DE-SERVJCEllMESJLND OTHER_ CHANGES ;

i JEIS-90-010 i i

t t

l 1

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t New York Power Authority i

JAMES A. FITZPATRICK NUCLEAR POWER PLANT. i Docket No. 50-333 ,

DPR-59 !

t

[

I

e Attachment lll to JPN-94-050 INSERTS FOR MARKED-UP TECHNICAL SPECIFICATION PAGES InserLA Survei!!ance_EIeguencyl40tations / Intervals The surveillance frequency notations / intervals used in these specifications are defined as follows:

Notations lateIvals Eleguency D Daily At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months W Weekly At least once per 7 days M Monthly At least once per 31 days O Quarterly or At least once per 92 days every 3 months SA Semiannually or At least once per 184 days every 6 months A Annually or Yearly At least once per 366 days R Note 1 At least once per 18 months (550 days)

S/U Prior to each reactor startup NA Not applicable Note 1: "Once each operating cycle," "once per operating cycle," "each refueling outage," "at least once during each operating cycle," "once each operating cycle not to exceed 18 months", or similar phrases, are equivalent to the definition for frequency notation "R" inseith PRIMARY CONTAINMENT ISOLATION SYSTEM INSTRUMENTATION REQUIREMENTS InseILG The response time of the reactor protection system trip functions listed below shall be demonstrated to be within its limit at least once per 18 months. Neutron detectors are exempt from response time testing.

InseILD The basis for the allowable out-of-service times is provided in GE Topical Report NEDC-30851P-A, " Technical Specification Improvement Analysis for BWR Reactor Protection System? March 1988.

Page 1 of 11

a inse1LE' The basis for a three-month functional test interval for group (A) sensors is provided in NEDC-30851P-A, " Technical Specification improvement Analysis for BWR Reactor Protection Systems."

IDSerLE Group (B) devices utilize an analog sensor coupled with a bi-stable trip (either the solid-state :

analog transmitter trip system (ATTS) or the more conventional arrangement of instrument )

amplifier and bi-stable).

A three month surveillance interval has been determined in accordance with NEDC-30851P-A. i

" Technical Specification improvement analyses for BWR Reactor Protection System."

10SftrLB The channel response time must include al1 component delays in the response chain to the .

ATTS output relay plus the design allowance for RPS iogic system response time. A response time for the RPS logic relays in excess of the design allowance is acceptable provided the overall response time does not exceed the response time limits specified in the UFSAR. 5 IDSf1Ll 1'

1. There shall be two operable or tripped trip systems for each Trip Function, except as provided for below:

a. For each Trip Function with one less than the required minimum number of operable instrument channels, place the inoperable instrument channel and/or its associated trip system in the tripped condition

within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Otherwise, initiate the ACTION l

required by Table 3.1-1 for the Trip Function.

b. For each Trip Function with two or more channels less than the required minimum number of operable instrument channels:

1) Within one hour, verify sufficient instrument channels remain operable or tripped

to ;

maintain trip capability in the Trip Function, and

2) Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , place the inoperable instrument channel (s) in one trip system and/or that trip system" in the tripped condition *, and l

3) Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , restore the inoperable instrument channel (s) in the other trip system to an operable status, or place the inoperable instrument channel (s) in the Page 2 of 11 1

I trip system and/or that trip system in the tripped condition *,

if any of these three conditions cannot be satisfied, initiate the ACTION required by i Table 3.1-1 for the affected Trip Function.

An inoperable instrument channel or trip system need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, if the inoperable instrument channel is not restored to operable status within the required time, the ACTION required by Table 3.1-1 for that Trip Function shall be taken.

This action applies to that trip system with the greatest number of inoperable ,

instrument channels. If both systems have the same number of inoperable j instrument channels, the ACTION can be applied to either trip system.

2. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into associated Limiting Conditions For Operation and required actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the assnciated Trip Function maintains RPS trip capability.

3. Action Statements:

A. Insert all operable control rods within four hours.

B. Reduce power level to IRM range and place Mode Switch in the Startup position within eight hours.

C. Reduce power level to less than 30 percent of rated within four hours.

insed.J The automatic scram contactors shall be exercised onca every week by either using the RPS channel test switches or performing a functional test of any automatic scram function. If the contactors are exercised using a functional test of a scram function, the weekly test using the RPS channel test switch is considered satisfied. The automatic scram contactors shall also be exercised after maintenance on the contactors. .

IDSelLK The response time of the main steam isolation valve actuation instrumentation isolation trip functions listed below shall be demonstrated to be within their limit at least once per 18 months.

InSedl.-

The surveillance test interval for the instrumentation channel functional tests are once/three months for most instrumentation. This surveillance interval is based on the following NRC -

approved licensing topical reports:

1. GE Topical Report NEDC-30851P-A. " Technical Specification improvement Analysis for ;

Page 3 of 11

BWR Reactor Protection System," March 1988.

2. GE Topical Report NEDC-30851P-A, Supplement 1, " Technical Specification improvement Analysis for BWR Control Rod Block Instrumentation," October 1988. 1

3. ' GE Topical Report NEDC-30851P-A, Supplement 2, " Technical Specification improvement Analysis for BWR isolation instrumentation Common to RPS and ECCS Instrumentation," July 1986.

4. GE Topical Report NEDC 31677P-A. " Technical Specification Improvement Analysis for BWR lsolation Actuation Instrumentation, " July 1990.

5. GE Topical Report NEDC-30936P-A, Parts 1 and 2, "BWR Owners Group Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation)," December 1988.

6. GE Topical Report GENE-770-06-1-A. " Bases for Changes to Surveillance Test Intervals and Allowed Out-Of-Service Times For Selected Instrumentation Technical Specifications," December 1992. ;

7. GE Topical Report GENE-770-06-2-A, " Addendum to Bases for Changes to Surveillance Test Intervals and Allowed Out-Of-Service Times For Selected Instrumentation Technical Speci'ications," December 1992.

The measurement of the response time interval for the Main Steam isolation Valve (MSIV) isolation actuation instrumentation begins when the monitored parameter exceeds the isolation actuation setpoint at the channel sensor and ends when the MSIV pilot solenoid relay contacts open.

!nSEILM

1. Whenever Primary Containment integrity is required by Specification 3.7.A.2, there shall be two operable or tripped trip systems for each Trip Function, except as provided for below;

a. For each Trip Function with one less than the required minimum number of operable instrument channels, place the inoperable instrument channel and/or its associated trip system in the tripped condition

within:

1) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for trip functions common to RPS instrumentation, and

2) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for trip functions not common to RPS instrumentation, or, initiate the ACTION required by Table 3.2-1 for the affected trip function.

b. For each Trip Function with two or more channels less than the required minimum number of operable instrument channels:

1) Within one hour. verify sufficient instrument channels remain operable or tripped

to maintain trip capability in the Trip Function, and Page 4 of 11

i

2) .Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , place the inoperable instrument channel (s) in one trip system and/or that trip system". in the tripped condition *, and

3) Restore the inoperable instrument channel (s) in the other trip system to an operable status, or place the inoperable instrument channel (s) in the trip system and/or that trip system in the tripped condition

within:

(a) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for trip functions common to RPS instrumentation,' and (b) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for trip functions not common to RPS instrumentation.

If any of these three conditions cannot be satisfied, initiate the ACTION required by i Table 3.2-1 for the affected Trip Function.

Asterisk shown on next page .

An inoperable instrument channel or trip system need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, if the inoperable instrument channel is not restored to operable status within the required time, the ACTION required by Table 3.2-1 for that Trip Function shall be taken.-

- This action applies to that trip system with the greatest number of inoperable instrument channels. If both systems have the same number of inoperable instrument channels, the ACTION can be applied to either trip system.

2. When a channel, and/or the affected primary containment isolation valve, is placed in an ,

inoperable status solely for performance of required instrumentation surveillances, entry into associated Limiting Conditions for Operation and required actions may be delayed as follows:

a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Trip Functions utilizing a two-out-of-two-taken-once logic; or b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for the remaining Trip Functions provided the associated Trip Function maintains PCIS initiation capability for at least one containment isolation valve in the affected penetration.

3. Actions:

A. Place the reactor in the cold condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B. Isolate the main steam lines within eight hours.

C. Isolate Reactor Water Cleanup System within four hours. !

D. Isolate shutdown cooling within four hours. !

E. Isolate the main steam line drain valves, the recirculation loop sample valves, and

. the mechanical vacuum pump, within eight hours. ,

F. Isolate the affected penetration flow path (s) within one hour and declare the affected l system inoperable. !

G. Isolate the affected main steam line within eight hours.

Page 5 of 11 l

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^1nsmt.N :

Instrumentation common to RPS. ;

10Sf!LQ ,

1. Whenever any ECCS subsystem is required by Specification 3.5 to be operable, there shall be two operable or tripped trip systems (or in the case of single trip system instrument :

logics, one operable trip system), except as provided for below:

i

a. For each Trip Function with one less than the required minimum number of operable instrument channels, place the inoperable instrument channel in the tripped condition

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Otherwise, declare the associated ECCS inoperable,

b. For each Trip Function with two or more channels less than the required minimum number of operable instrument channels: ,

1) Within one hour, verify sufficient instrument channels remain operable or tripped

to maintain trip capability in the Trip Function, and ,

i

2) Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , place the inoperable instrument channel (s) in one trip system" in the tripped condition *, and

3) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , restore the inoperable instrument channel in the other trip system

to an operable status.

If any of these three conditions cannot be satisfied, declare the associated ECCS ;

inoperable.

An inoperable instrument channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, if the ,

inoperable instrument channel is not restored to operable status within the required time, declare the associated ECCS inoperable.

This action applies to that trip system with the greatest number of inoperable instrument channels. If both systems have the same number of inoperable instrument channels, the ACTION can be applied to either trip system.

2. - When a channel is placed in an inoperable status solely for performance of required l surveillances, entry into associated Limiting Conditions For Operation and required actions !

may be delayed as follows: (a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for single channel Trip Functions; or (b) l for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for the remaining Trip Functions provided the associated Trip Function ,

maintains ECCS initiation capability. l

3. Refer to Technical Specification 3.5 for Limiting Conditions for Operation. Failure of one I (1) instrument channel disables automatic initiation of one (1) pump. .

1 Page 6 of 11 l

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a l

t InheILE !

1. The trip functions shall be operable in the Startup and Run modes except as follows:

a) SRM and IRM: Startup mode only. 7 b) RBM: Run mode and ;> 30% reactor power only.

c) APRM Neutron Flux-Startup: Startup mode only.

d) APRM Flow Referenced Neutron Flux: Run mode only.

2.- Actions:

ActionA: If the number of operable instrument channels is:

i a) one less than the required minimum number of operable instrument channels per trip !

function, restore the inoperable instrument channel to operable status within 7 days, or !

place the inoperable instrument channel in the tripped condition within the next hour. l b) two or more channels less than the required minimum number of operable instrument I channels per trip function, place at least one inoperable instrument channel in the tripped condition within one hour.

Action 3 If the number of operable instrument channels is:

a) one less than the required minimum number of operable instrument channels per trip function, venfy that the reactor is not operating on a Limiting Control Rod Pattem, and within 7 days restore the inoperable instrument channel to operable status; otherwise, >

place the inoperable instrument channel in the tripped condition within the next hour.

See Specification 3.3.B.5. >

b) two channels less than the required minimum number of operable instrument channels per trip function, place at least one inoperable instrument channel in the tripped condition within one hour. See Specification 3.3.B.S.

Action.C; If the number of operable instrument channels is less than the required minimum number of operable instrument channels per trip function, place the inoperable instrument channel in the tripped condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . !

3. When a channel is placed in an inoperable status solely for performance of required i surveillances, entry into associated Limiting Conditions for Operation and required actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Tnp Function maintains CRB ,

. initiation capability.

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i Page 7 of 11 1

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Insert O

1. There shall be two operable or tripped trip systems for each Trip Function, except as e provided for below: l

a. For each Trip Function with one less than the required minimum number of operable instrument channels, place the inoperable instrument channel and/or its associated trip i system in the tripped condition

within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Otherwise, place the reactor in the start-up/ hot standby mode within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. For each Trip Function with two or more channels less than the required minimum number of operable instrument channels: -

1) Within one hour, verify sufficient instrument channels remain operable or tripped

to maintain trip capability in the Trip Function, and i

2) Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , place the inoperable instrument channel (s) in one trip system and/or that trip system ** in the tripped condition *, and

3) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , restore the inoperable instrument channel in the other trip system to ,

an operable status.

If any of these three conditions cannot be satisfied, place the reactor in the start-up/ hot ,

standby mode within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . ,

An inoperable instrument channel or trip system need not be placed in the tripped i condition where this would cause the Trip Function to occur. In these cases, if the inoperable instrument channel is not restored to operable status within the required time, place the reactor in the start-up/ hot standby mode within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

This action applies to that trip system with the greatest number of inoperable instrument channels. If both systems have the same number of inoperable '

instrument channels, the ACTION can be applied to either trip system.

2. Wher, a channel is placed in an inoperable status solely for performance of required surveillances, entry into associated Limiting Conditions for Operation and required actions l may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Trip Function maintains ATWS j RPT initiation capability. l l

InsetLB 1

CORE AND CONTAINMENT COOLING SYSTEM INITIATION AND CONTROL INSTRUMENTATION OPERABILITY REQUIREMENTS Page 8 of 11

I I

InseILS CONTROL ROD BLOCK INSTRUMENTATION REQUIREMENTS I

loseill l ATWS RECIRCULATION PUMP TRIP INSTRUMENTATION REQUIREMENTS lasetLU PRIMARY CONTAINMENT ISOLATION SYSTEM INSTRUMENTATION TEST AND CAllt3 RATION REQUIREMENTS lasetLV CORE AND CONTAINMENT COOLING SYSTEM INSTRUMENTATION TEST AND CALIBRATION REQUIREMENTS 10SHILW_

CONTROL ROD BLOCK INSTRUMENTATION TEST AND CALIBRATION REQUIREMENTS inseILX ATWS RECIRCULATION PUMP TRIP INSTRUMENTATION I

TEST AND CALIBRATION REQUIREMENTS inseILY Equipment removed from service or declared inoperable to comply with required actions may i be retumed to service under administrative control solely to perform testing required to I demonstrate its operability or the operability of other equipment. This is an exception to LCO 3.0.B. ,

i InSe1LZ LCO 3.0.F establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with required actions. The sole purpose of this Specification is to provide an exception to LCO 3.0.B to allow testing to demonstrate: (a) the operability of the equipment being retumed to service; or (b) the operability of other equipment.

Page 9 of 11

The administrative controls ensure the time the equipment is returned to service in conflict with the requirements of the required actions is limited to the time absolutely necessary to ;

perform the alle.ved testing. This Specification does not provide time to perform any other preventive or corrective maintenance.

An example of demonstrating the operability of the equipment being returned to service is reopening a containment isolation valve that has been closed to comply with the required i actions and must be reopened to perform the testing. ;

An example of demonstrating the operability of other equipment is taking an inoperable ;

channel or trip system out of the tripped condition to prevent the trip function from occurring ;

during the performance of testing on another channelin the other trip system. A similar example of demonstrating the operability of other equipment is taking an inoperable channel or trip system out of the tripped condition to permit the logic to function and indicate the !

appropriate response during the performance of testing on another channel in the same inp ;

system.

[0Sf11A6

1. _ Reactor High Pressure (02-3PT-55A, B, C, D) [

2. Drywell High Pressure (05PT-12A, B, C, D) j

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

4. Main Steam Line isolation Valve Closure !

(29PNS-80A2, B2, C2, D2)

(29PNS-86A2, B2, C2, D2) ,

S. Turbine Stop Valve Closure (94PNS-101,102,103.104) .

6. Turbine Control Valve Fast Closure (94PS-200A, B, C, D)

~

7. APRM Fixed High Neutron Flux

8. APRM Flow Referenced Neutron Flux ,

InseILBB :

Initiates ADS (if not inhibited by ADS override switches), in conjunction with Confirmatory Low Level,120 second delay and RHR (LPCI) or Core Spray pump discharge pressure interlock. ,

i insetLC_C The quarterly calibration of the temperature sensor consists of comparing the active temperature signal with a redundant temperature signal. ;

Inse!1DD i

The instrumentation which initiates primary containment isolation is connected in a dual bus j (two trip systems) arrangement. Main Steam Line Isolation Valve (MSIV) iso!ation utilizes a j one-out-of-two-taken-twice logic arrangement which closes the four inboard and four outboard !

MSIVs. Other penetrations which have both inboard and outboard automatic isolation valves ,

i Page 10 of 11

i

(except for the primary containment hydrogen and oxygen concentration sample, and the gascoas and particulate radioactivity sample systems) utilize logic arrangements in which one trip system closes inboard isolation valves and the other trip system closes outboard isolation valves. The primary containment hydrogen and oxygen concentration sample supply and '

' return lines, as well as the gaseous and particulate sample supply and return lines, utilize inboard and outboard isolation valves that are both closed by a single trip system. Hydrogen and oxygen concentration sample supply and return isolation valve control circuits are :'

provided with the capability to override automatic isolation to allow sampling during and following an accident. Penetrations which are isolated by a single automath icolation valve (and a remote manual or check valve) utilize a single trip system to effect dosure of the automatic isolation valve.

10se1EE The main steam line high temperature isolation function utilizes 16 sensors (instrument channels), with 4 sensors located at each of 4 different areas in the vicinity of the rnain steam lines. The 4 instrument channels associated with each of the 4 areas are arranged in a 1-out-of-2-taken-twice logic. Thus a main steam line break in any of the 4 areas will effect closure of all 8 main steam line isolation valves.

lasert FF ,

i The HPCI high temperature isolation function utilizes 16 sensors (instrument channels) located in the vicinity of the HPCI equipment and piping. The 16 instrument channels provide inputs into two trip systems, eight instrument channels per trip systemc One trip system is 1 associated with the inboard isolation valve and the other trip system is associated with the outboard isolation valves. Trip logic for each trip system is one-out-of-eight-taken-once logic for the high temperature isolation function. The logic for the RCIC high temperature isolation ,

function is the same as the HPCI logic, except 8 instrument channels,4 per trip system provide input to the high temperature isolation logic circuits.

IDSe!LGG 1

The measurement of the response time interval for the Main Steam isolation Valve (MSIV) isolation actuation instrumentation begins when the monitored parameter exceeds the isolation actuation setpoint at the channel sensor and ends when the MSIV pilot solenoid relay contacts open.

InselLBB Trip Function utilizes a two-out-of-two-taken-once logic for isolation of both primary l containment isolation valves on the hydrogen and oxygen sample, and gaseous and particulate sample supply and retum lines.

l Page 11 of 11

I

, JAFNPP TECHNICAL SPECIFICATIONS TABLE OF CONTENTS EA21 l 1.0 Definitions 1

{

LIMITING SAFETY SAFETY LIMITS SYSTEM SETTINGS l 1.1 Fuel Cladding Integrity 2.1 7 l 1.2 Reactor Coolant System 2.2 27 SURVEILLANCE L!MITING CONDITIONS FOR OPERATION REQUIREMENTS !

3.0 General 4.0 30 l 3.1 Reactor Protection System 4.1

.iMH '50 3 3.2 Instrumentation 4.2 49-A. Primary Containment Isolation Functions A. 49 B. Core and Containment Cooling Systems - B. 50 Initiation and Control C. Control Rod Block Actuation C. 50 l D. Radiation Monitoring Systems -Isolation D. 50 and initiation Functions E. Drywell Leak Detection E. 54 I F. DELETED F. 54 l !

G. Recirculation Pump Trip G. 54 i H. Accident Monitoring instrumentation H. 54 1

1. 4kV Emergency Bus Undervoltage Trip 54 ;

1 3.3 Reactivity Control 4.3 88 l A. Reactivity Limitations A. 88 i B. Control Rods B. 91 !

C. Scram insertion Times C. 95 !

D. Reactivity Anomalies D. 96 l 3.4 Standby Liquid Control System 4.4 105 l A. Normal Operation A. 105 !

8. 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 l B. Containment Cooling Mode of the RHR B. 115 System C. HPCI System C. 117 D. Automatic Depressurization System (ADS) D. 119 l E. Reactor Core isolation Cooling (RCIC) E. 121 1 System Amendment No. ,1 0,1 4,1 l3, if 0 !

1 JAFNPP LIST OF TABLES P.P412

.Iabla 1 i11 2 3.1-1 Reactor Protection System (Scram) Instrumentation Requirement 'f o 44-3.1-2 lileactor Protection system Instrumentation Response Time -40e-e FI .

Reactor Protection System (Scram) Instrument Functional Tests 44 4.1-1 46 4.1-2 Reactor Protection System (Scram) Instrument Calibratio gg 3.2-1 Gnstrumentation that initiates Primary Containment isolatinID 4A 64-3.2-2 , 66 (Instrumentation that initiates or Controls the Core and Containmed LCoolino Systemsf (

3.2-3 gnstrumentation that initiates Control Rod Blnnket 72 3.2-4 (DELETED)

-M 3.2-5

Instrumentation that Monitors Leakage Detection inside the Drywell 75 i 3.2-6 g (DELETED) 3.2-7 Onstrumentation that initio+^r De9c@4n Anmn Trf b5fd Y 7(, "

3.2-8 Accident Monitoring Instrumentation 77a 3.2-9 % Primary Containment Isolation System Actuation Instrumentati 47e.

' Response Tima#

3.2-10 Remote Shutdown Capability Instrumentation and Controls 77f 4.2-1 LMinimum Test and Calibration Frequency for PCIOd:--ZwssVf f) 78 4.2-2 Minimum Test and Calibration Frequency for Core and Containm,e$e B 0-79--

[1 Cooling Systemsf L 4.2-3 TMinimum Test and Calibration Frecuency for Control Rod Bloc @ B A I lu:tuatiorf (

4.2-4 es-(DELETED) 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 Tn) 85 L9ff )(

Amendment No. A,pa" 130',Jaf,,1pa', per, pf v

JAFNPP UST OF FIGURES

N Z g sgr{ "pg yg) "

4.1-1 Graphic ,*id in the Selection of an Adequate Interval Between Tests @f L 4.2-1 Test Interval vs. Probability of System Unavailability [ g 3.41 Sodium Pentaborate Soluti 110 Concentration Requirements .7 B-10 Atom % Enrichedfolume-

h. .immum 3.4-2 Saturation Temperature of Enriched Sodium Pentaborate Solution 111 3.5-1 Thermal Power and Core Flow Umits of Specifications 3.5.J.1,3.5.J.2 and 134 3.5.J.3 N 3.6-1 Reactor Vessel Pressure - Temperature Umits Through 12 EFPY 163 y Part 1 , , ._.

3.6-1 Reactor Vessel Pressure - Temperature Umits Through 14 EFPY 163a Part 2 3.6-1 Reactor Vessel Pressure - Temperature Umits Through 16 EFPY 163b Part 3 4.61 Chloride Stress Corrosion Test Resui% at 500 F 164 6.1 1 (Deleted) -

6.2 1 (Deleted) i I

l 1

Amendment No.14, K 44r,94,2f,74, as,9( 18G, tM,1.M trf, W, t7f,156, %

vil 1

l

JAFNPP 1.0 (cont'd) opened to perform necessary operational activities.

MEMr kplAfoy he's:

R. Safety Limits - The sal y limits are hmits within which

2. the reasonable mainte nce of the fuel claddeng integrity At least one door in each airlock is closed and and the reactor cool t system integrity are assured.

sealed. Violation of such a it is cause for unit shutdown and review by tholmtomic :neray t,ommesseorf befare

3. All automatic containment isolation valves are resumption of unit operation. Operation beyond such a operable or de-activated in the isolated position. limit may not in itself result in serious consequences but it !

^

4. indicates an operational deficiency subsect to regulatory All blind flanges and manways are closed. review.

N. Rated Power - Rated power refers to operation at a reactor S. Secondarv Containment inteority - Secondary containment power of 2,436 MWt. This is also termed 100 percent integrity means that the reactor building is intact and the power and is the maximum power level authorized by the following conditions are met:

operating bconse. Mated steem flow, rated coolant flow, rated nuclear system pressure, refer to the values of these 1. At least one door in each access opening is closed.

parameters when the reactor is at rated power.

2. The Standby Gas Treatment System is operable.

O. Reactor Power Onoraten - Reactor power operation is any operation with the Mode Switch in the Startup/ Hot 3. All automatic ventilation system isolation valves are Standby or Run position with the reactor critical and above operable or secured in ttw isolated position.

1 percent rated thermal power.

T. (DeletedW ja P. Reactor Vessel Pressure - Unless otherwise indicated, l

' reactor vessel pressures listed in the Technical (

Specifications are those measured by the reactor vessel steem space sensw.

1 h5drk [

O. Refuehne Outane - Refuehng outage is the period of time between the shutdown of the unit prior to refuehng and the startup of the Plant =i:7_=.i to that refuehng. '

Amendment No. f, if4, lf8 5 .

m-- _ . r- - w _, .-- - - _ _ _ _ _m - .e. . . . . . _ _ _ _ _ _ . _ .

a~rNPP 3.0. Continued 4.0 Continued N I D. Entry into an OPERATIONAL CONDITION (model or other that a Surveillance Requirement has not been performed. The specified condition shall not be made when the conditions for ACTION requirements may be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Limiting Condition for Operation are not met and the permit the completion of the surveillance when the allowable associated ACTION requires a shutdown if they are not met outage time limits of the ACTION requirements are less than .

within a specified time interval. Entry into an OPERATIONAL 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Surveillance requirements do not have to be CONDITION (mode) or specified condition may be made in performed on inoperable equipment. (

accordance with ACTION requirements when conformance to them permits continued operation of the facility for an D. Entry into an OPERATIONAL CONDITION (mode) shall not be -

unlimited period of time. This provision shall not prevent made unless the Surveillance Requirement (s) associated with passage through OPERATIONAL CONDITIONS (modes) the Limiting Condition for Operation have been performed required to comply with ACTION requirements. Exceptions to within the applicable surveillance interval or as otherwise %q these requirements are stated in the individual specifications specified. This provision shall not prevent passage through or to Operational Modes as required to comply with ACTION E. When a system, subsystem, train, component or device is Requirements. j" determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requireme".is of its applicable Limiting Condition for Operation provided: (1) its corresponding normal or emergqncy power source is OPERABLE: and (2) all of its reocandant system (s),

subsystem (s), traints), componends) and device (s) are OPERABLE, or likewise satisfy the r0quirements of this specification. Unless both conditions (1) and (2) are satisfied, the unit shall be placed in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This specification is not applicable when in Cold Shutdown or Refuel Mode.

F. ,

ng y Amendment No. d J ,1[4, [8 30s

" E M M Eimi >

U f ,

JAFNPP 4.0 BASES 0

A. This specification provides that surveillance activities C. Continued necessary to insure the Limiting Conditions for Operation are -

met and will be performed during the OPERATIONAL interval, defined by the provisions of Specification 4.0.B. as a CONDITIONS (modes) for which the Limiting Conditions for condition that constitutes a failure to meet the OPERABILITY Operation are applicable. Provisions for additional surveillance requirements for a Limiting Condition for Operation. Under the activities to be performed without regard to the applicable provisions of this specification, systems and components are OPERATIONAL CONDITIONS (modes) are provided in the assumed to be OPERABLE when Surveillance Requirements individual Surveillance Requirements. have been satisfactorily performed within the specified time interval. However, nothing in this provision is to be construed B. Specification 4.0.B establishes the limit for which the specified as implying that systems or components are OPERABLE when time interval for Surveillance Requirements may be extended. they are found or known to be inoperable although still it permits an allowable extension of the normal surveillance meeting the Surveillance Requirements. This specification also interval to facilitate surveillance scheduling and consideration clarifies that the ACTION requirements are applicable when of plant operating conditions that may not be suitable for Surveillance Requirements have not been completed within the conducting the surveillance (e.g., transient conditions or other a!! owed surveillance interval and that the time limits of the ongoing surveillance or maintenance activities). It also ACTION requirements apply from the point in time it is provides flexibility to accommodate the length of a fuel cycle identified thet a so;emance has not been performed and not for surveillances that are performed at each refueling outage at the time that the allowed surveillance was exceeded.

and are specified with an 18 month surveillance interval. It is Completion of the Surveillance Requirement within the not intended that this provision be used repeatedly as a allowable outage time limits of the ACTION requirements convenience to extend surveillance intervals beyond that restores compliance with the requirements of Specification specified for surveillances that are not performed during 4.0.C. However, this does not negate the fact that the failure refueling outages. The limitation of this specification is based to have performed the surveillance within the allowert on engineering judgement and the recognition that the most surveillance interval, defined by the provisions of Specification probable result of any particular surveillance being performed 4.0.B. was a violation of the OPERABILITY requirements of a is the verification of conformance with the Surveillance Limiting Condition for Operation that is subject to enforcement Requirements. The limit on extension of the normal action. Further, the failure to perform a surveillance within the surveillance interval ensures that the reliability confirmed by provisions of Specification 4.0.B is a violation of a Technical surveillance activities is not significantly reduced below that Specification requirement and is, therefore, a reportable event otitained from the specified surveillance interval. under the requirements of 10 CFR 50.73fal(2)(i)(3) because it is a condition prohibited by the plant Technical Specificatiens. y v

C. This specification establishes the failure to perform a i Survoillance Requirement within the allowed surveillance J

1 8 Amendment No. ,3% 3cg 40+ _

t

JAFMPP N 4.0 BASES - Continued C. Continued C. Continued i

If the allowable outage time limits of the ACTION requirements Surveillance Requirements do not have to be performed on are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or a shutdown is required to comply inoperable equipment because the ACTION requirements with ACTION requirements, a 24-hour allowance is provided to define the remedial measures that apply. However, the permit a delay in implementing the ACTION requirements. Surveillance Requirements have to be met to demonstrate that This provides an adequate time limit to complete Surveillance inoperable equipment has been restored to OPERABLE status.

3 Requirements that have not been performed. The purpose of i

-i this allowance is to permit the completion of a surveillance

! before a shutdown is required to comply with ACTION D. This specification establishes the requirement that all '

requirements or before other remedial measures would be applicable surveillances must be met before entry into an required that may preclude completion of a surveillance. The OPERATIONAL CONDITION or other condition of operation basis for this allowance includes consideration for plant specified in the Applicability statement. The purpose of this conditions, adequate planning, availability of personnel, the specification is to ensure that system and component time required to perform the survedlance and the safety OPERABILITY requirements or parametes limits are met before significance of the delay in completing the required entry into an OPERATIONAL CONDITION or other specified ,

'i surveillance. This provision also provides a time limit for the condition associated with plant shutdown as well as startup.

completion of Survedlance Requirements that become applicable as a consequence of OPERATIONAL CONDITION Under the provisions of this specification, the applicable (mode) changes imposed by ACTION requirements and for Surveillance Requirements must be performed within the completing Surveillance Requirements that are applicable when specified surveillance interval to ensure that the Limiting an exception to the requirements of Specification 4.0.C is Conditions for Operation are met during initial plant startup or allowed. If a survedlance is not completed within the 24-hour following a plant outage.

allowance, the time limits of the ACTION requirements are

! applicable at that time. When a survedlance is performed When a shutdown is required to comply with ACTION I

within the 24-hour allowance and the Surveilla9ce requirements, the provisions of this specification do not apply Requirements are not met, the time limits of the ACTION because this would delay placing the facility in a lower requirements are applicable at the time the surveillance is CONDITION of operation. ,

j terminated.

[ b h at W N 4 9 6 4 ,85:101,1 4 :18 3 Amendment No. ya,155, 19lD f '

34 -

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JAFNPP 3.1 (cont'd) 4.1 (cont'd)

[ 2. If anyterne during reactor operation at greater than 25% of C. MCPR shall be determined daily during reactor power operation rated power it is determined that the operating limit MCPR at >25% of rated thermal power and following any change in <

is bemg excaartart action shall then be instiated withm power level or distribution that would cause operation with a fifteen (15) minutes to restore operation to within the hrrutog control rod pattom as AM in the bases for g preecnbod bmits. If the MCPR is not retumed to withm the Specificidicii 3.3.B.S. 7 pr two ,an w % reacta fD. When it is determmed that a channel has failed in the unsafe Power reduction M hewn @... -- _ ,. The reactor condition, the other RPS channels that monitor the same N ' N variable shall be functionally 1ested immediately before the trip to ,

system cordainog the failure is tripped The trip system

&hWh '

contammg the unsafe failure may be placed in the untripped v conditKyi dunng the period in which surveillance testing is being t performed on the other RPS channels. J T)[ Verification of the MCPR operating limits shall be performed as specified in the Core Operating Umits Report N

L i

Amendment No. p,74,74,116,9tf,196,17/ %

31

. . . , . . . . . . . . . , , -- . , - > - - - . . . - . , s , , .-.,e... , .. . -, .,

y

. . ..- i 7 JAFNPP 3.1 BASES i The reactor protection system automatically initiates a fA. reactor scram to:

The outputs of the subchannels are combined in a 1 out of 2 logic; i.e., en input signal on either one or both of the subchannels will cause a trip system trip.

1. Preserve the integrity of the fuel claddmg. The outputs of the trip systems are arranged so that a trip on both systems is required to produce a

2. Preserve the integrity of the Reactor Coolant reactor scram.

System.

This system meets the intent of IEEE-279 (1971) for

3. Minimize the energy which must be absorbed Nuclear Power Plant Protection Systems. The following a loss of coolant accident, and prevent system has a reliability greater than that of a 2 out inadvertent criticalsty. of 3 system and somewhat less then that of a 1 out of 2 system.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system With the exception of the average power range .

to perform its intended function even dureng periods when monitor (APRM) channel the intermediate range instrument channels may be out of service because of monitor llRM) channels, the scram discharge volume, maintenance. When necessary, one channel may be made the main steem isolation valve closure and tte moperable for brief intervals to conduct required functional turbine stop valve closure, each subchannel has one tests and calibrations. instrument channel. When the mmemum condition for b SFPD operation on the number of operable instrument i

The Reactor Protection System is o channel type channels per untripped protection trip system is met l (Reference subsection 7.2 FSARI. The System is made up or if it cannot be met and the affected protection trip of two independent trip systems, each having two system is placed in a tripped condition, the subchannels of tripping devices. Each subchannel has an effectiveness of the protection system is preserved.

input from at least one instrument channel which monitors a critical parameter. .Three APRM instrument channels are provided for each protection trip system. APRM's A and E operate contacts in one subchannel and APRM's C 'i and E operate contacts in the other ~ I i

l i

l Ameeximent No. , , ,

If0 e .- - - _ . . .-. .- . - - - - - -

t L

JArilPP 4 .1 - BASES divided into three groups for funct.ional testing. These are A. FThe minimum functional testing) frequency used in this specificat. ion hreesp A. On-off sensors that. provide a is based on a reliability analysis scram trip tunction.

using the concepts developed in Reference (6) This concept was (yr,g, 11 Analog devices coupled with specifically adapted to the 1 out of I bi-stable trips that provide a 2X2 logic of the Iteactor Prutection sc ram function.

System. The anhlysis shows that the are primarily restonnible Devices which only serve a senso rs fi.yengC. use ful function during some for the reliability of the heactor Prot.ection Sy s t e:n . This analysis restricted mode of operation, makes use of unsafe t'a il u re rate stich as starttip or shtitdown, experience at conventional and or for which the only practical nuclear power plants in a test _is one that can he per-reliability model for t.he sy u't.ein. formed at shutdown.

An unsafe failure is defined as one sennorn that inake un Groun IM which negates channel operability and which, due to its nature, is The are peci t l eal.17 selectest trom among' y i revealed only when the channel is ne whole ~t .on11y of industrial functionally tested or at teing ts to on--o t t uunnot :0 t.ha t. have earned an respond to a:real signal. l'a i lu re s . rexce21ent reputation for reliablu such as blown fuses, ruptured I operation.. Durirai desion, a goal of '

bourdon tubes, f aulted a:nplif iers, 1 0.99999 pt obabilit y of sucetsu (a t and Iaulted cables, which result in the 50 percent. confidence 1evel) waa upscain or downscale readings on the adopt ed to annure t. h a t. a talanc.d react.or iAstrumentation .are s.af e and cand adoerna e e ib. ii .tn is och'iaved) q8 will be easily recognized by the The p robabi l i ty o f success _ i s p ri saa ri ly operators during operation be cause a funct ion of Ihe sensor failure rat e they are revealed by an alarm or a) and the test i nt e rva l .fA th ree-mont h N Lscram. f

~

ftest interval is planned for group (h)

. s ens o rs . This is in keeping with good The channels listed in and it.1-2 are c1ierat ing I3ract i ces , and sat is fies the Tables 13 . 1 - 1 ..

. design goal f.or the logie conf..i guraty (

AmendmentNo.% Reptue wth Lua e 36

( l t

4.1 DASES (cont 8d) JAFNPP utilized in the Reactor Protection it . After a trend is established, System. the appropriate monthly test interval to satisfy the goal To satisty the long-term objective will be the test interval to of maintaining on adequate level of the left of the pictted points.

salcty throughout the plant lif e-time, a minimiva goal of 0.9999 5. A test interval of 1 month will t at the 95 percent confidence level be used initially until a trend is proposed. With the 1 out of 2X2 is established, which is based logic, this reauires that each on system availability analysis sensor have an availability of 0.993 and uood en9 i neering judgment at the 95 percent confidence level, plus operating experience.

This level of availability may be maintained by adjusting the test oup (3) devices utilize an analog interval as a function of the sensor followed by an amplifier and Q' observed failure history facilitate the implementation (6). To of a ht-stable trip circuit.

sensor and amp 11 tier are active 9M this technique, Figure is .1- 1 is components and a failure is almost 1HS,A F provided to indicate an appropriate always accompanied by an alarm and trend in test interval. The an indication of the source of procedure is as follows:

trouble.orfinsubstitutinn the event of canfailure, frepair sta

1. Like sensors are poolal into Limmediatniv. fan as-is failure is one group for the purpose of one that sticks mid-scale and is not data acquisition. capable of going either up or down in response to an out-of-limits

2. The tactor H is the exposure input. This type of failure for hours and is equal to the analog devices is a rare occurrence

~ number of sensors in a group, and is detectable by an operator who n, times the elapsed time, T observes that_ one sional does not (tt = nT). CbMf I 'trnck 44+e otherTheseew. [For purposel of analysis, it is assumed that this ,

3. The accumulated number of rare failure will 1e detected within unsate failures is plotted as f

2 hr. j an ordinate against H as an abscissa on Figure 4.1-1.

37.

I JAFNPP duri Iddth*g q II ' N 4.1 BASES (corW'd) i every flirce ma=ns m i

fuu.henah @

The tn-statAe trip circuit which a part of the Group 3) The frequency ci calibration of the APRM Sow tiesing network l devices can sustainwesefe I wtuch are revealed has been estabbshed as each refueNng outage. ; @ flow l

4eet Therefore,it is necessary est them periodicaNy. Inserf y tested at least once/su netwnd, twasing network is functiongip A study was conducted of the instrumentation chand - - = b ah, cross caNbrah checks of Wie Row W k tw Wow included in the Group (B) devices to calculate their unsale) % N can be e dur% the W W by failure rates. The rmn-ATTS (Analog Transnutter Trip System) drect noter readng There me several kwirurnerts which analog devices (sensors and ampMiers)4are predicted to have M k @aW W N W take mal days b W h an unsale failure rate of less then 20x10 failures /hr. The non- Nh d Wu enHre 6 h the cabah k W

, a mo h sgnal wW be sort to hen of We Na ATTS bi-et=Na trip ciraAts we predcted to have unsafe failure rate of less than 2x104 failures /hr. The ATTS analog devices resolung b a haN scram and rM block com9 ten. Thus, if #w l 42- - (sensors), bi-et=Na devices (mester and slave trip units) and maHon were W during operadon, Atm shaping I

power suppues have been evalueled for reliabshty by Mean M nd be posable. Based on W at oWar Time h FaRura ansfysis or state-of-the-st quaMicWion generating sianons, drNt of instruments, sudt as those in the type testing meeting the requirements of IEEE 323-1974. Now tdasing work, h not sW and Umrefore, to avoid Considering Wie 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months monitoring interval for angiog devices spurious scrams, a cahbration frequency of each refueling i as assumed above, the instrument cliecks and functional tests mAage WN. ,

l as well as the analyses and/or qusMicahon type testing of the The measurement of response time whbin the specined devices, the design reEnNuly goel for system rehabihty of intervals provedes assurance thei the Reactor Protection 0.9999 wlE be attained with ample margin. System trip functions se completed within the time limits The td-stable devices are monitored dunng plant operation to transierW and W a@.

record their faNure history and metat*=h a test interval using the FThe Reactor Protection System trip functions in Table 3.1-2 a2 curve of Figure 4.11. There are numerous identical bi-stable # those fuctions for which the transient and accident analyses devices used throughout the Plant's instrumerWation system.

Therelore, signlAcant data on the fature rates for the bi-stable described in Chapter 14 of the FSAR take credit for thye response time of instrument channels.f devices should be accumulated rapidly. '

i Amendment No. A([M )M, Jd 8

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JAFNPP ;

TABLE 3.1-1 AJJ REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT $

Minimum No. Modein Which Function of Operable Must be Operable Total Number of instrument instrument Channels Channels Per Refuel Startup Run Provided by Design Action Trip System (1){i) Trip Function Trip I.svel Setting pg ,

for Both Trip Systems ,py(3) 9 1 Mode Switchin X X X 1 Mode Switch A Shutdown -

(tS;';tc.4 1 Manual Scram o gef, X X X 2 Instrument Channels A 3 IRM High Rux _

25 f X X 8 Instrument Ctennels A 3 IRMInoperative X *X 8 Instrument Channels A 2 APRM Neutron Flux- < 15% Power X X 6 Instrument Channels A Startup (15) l 2 APRM Row Referenced (12) X 6 Instrument Channels A or B Neukon Rux (Not to exceed 117 %) (13)(14)

{

2 APRM Rxed High < 120% Power X 6 Instrument Channels A or B Neutron Rux (14) 2 APRMinoperative (10) X X X 6 Instrument Channels A or B l

Amendment No. [14,30,46, M, IW,96,134 % .

b Re76< wA i+,is,les y,

- _ - _ - - - . - - - - - . - - - - - _ . . . - , , . , w ., e r - ~ ,,

JAFNPP TAN r 3.1-1 (cont'dl g i

REACTOR PROTECTION SYSTEM (SCfWWil.lNSTRUMENTATION REQUMEMENT T Minimum No. Modes in Which Function of Operable Must be Operable Total Number of i

instrument .

Instrument Channels Channels per Refuel Startup Run Provided by Design Action Trip System (1(1). Trip Function Trip Level Setting' for Both Trip Systems 441' @

[2 APRM Downscale a: 2.5 indicated on scale (9)

X 6 Instrument Channels Ao 2 gReactofPressure s 1045 psig Xys) X X 4 Instrument Channels A 1 el % essure s 2.7 psig X X ) X A 4 Instrument Channels 2 eactor L Water a: 177 in, above TAF X X X 4 Instrument Channels A Level (Nehr :4 3 High Water Level in s 34.5 gallons per X 8 Instrument Channels A Scram Descharge Volume Instrument Volume XJ(d 4 X 4 Mein Steam Line s 10% valve closure X(, ) 8 Instrument Channels A lsolation Valve Closure i

1 Am. nom.nt no. cji . p. p. p. v. W. its. in 22D L rom w A Ig,3o,c,3r, e7 9, 40 s IPC,143.

JAFNPP TAE4E_b l:1_icont*tl)

BEACIDRJROTECTION. SYSTEtLISCH AM) ! NSTRUM_ENTATION_ REQtjl REMENT Minimum No. . Modes in Which Total of Operable Trip Level Function Must be Number of Instrument Trip Function Setting Operable Instrument Action Channels _ ._ Channels (1) per Trip Refuel Startup Run Provided '

System (1) (6) by Design (16) for Both Trip Systems V

2 Turbine Control 500 <P( 850 psig X(/) 4 Instrument A or C Valve Fast Closure Control oil pressure Channels between fast closure solenoid and disc dump valve hove. +ed h r e vos el ,ome 4I AmendmentNo.pI, . I4 41h ,

JAFNPP TABLE 3.1-1 (cont'd) '

REACTOR PROTECTION SYSTEM (SCRAMI INSTRUMENTATION REQUIREMENT Minimum No.

of Operable Modes in Which Function Must be Operable Total Number of Instrument '

Channels per Instrument Charmels Refuel Stortup Run Provided by Design Trip System (1) Trip Function Trip Level Setting Action (6) 11 6) for Both Trip Systems (1) hovt. f0 g 4 Turbine Stop s10% volve XMK5) revscM Velve Closure cloeure 8 instrument Channels A or C ftM NOTES OF TABLE 3.1-1 Regime Jin Inse.+ I fi. There shall be two operable or tdW trip systems for each function, except es specified in 4.1.D.

From and after the time that the ,

minimum number of operable instrument channel for a trip system cannot be met, that offacted trip system shall be placed (tripped) condition, or the appropriate actions listed below shall be taken.

A. Insert all operable control rods within four hours.

( C. Reduce w- -~ ' -~' M '::: iB. Reduce power level to IRM range and place Mode Switch in the Startup Position within ei

- ---af=2 4 winni four hours.

- I'$ IA He

' 5) / Permissible to bypass,ifRefueleniLShutdown positions ^5,6 Reactor Mode Switc?.

.3.-Geloted. P 8F i ny /. Bypassed when turbine first stage pressure is less then 217 psig or less then 30 percent of rated f, $. The design permits closure of any two lines without a scram being initiated. ' f6ve5r.

p/ 7 p. operable I

When # the reactor is soberitical and the reactor water temperature is less then 212*F o lf A. Mode Switch in Shutdown.

~

( B. Manuel Screm.

Move +o rev 04 pqe 93 Amendment No. 1[.%,[

42

JAFNPP TABLE 3.1-1 (contd)

REACTOR PROTECTION SYSTEM iSCRAMI INSTRL 4TATION REtM=_E=ENT

' NOTES OF TABLE 3.1-13(fcont'd)

C. High Flux IRM. ,

D.

Scram Discharge Volume High Level when any control rod in a control cell containing fuel is not fully inserted.

E. APRM 15% Power Trip.

8/. Not required to be operable when primary containment integrity is not required.

1 p. Not required to be operable when the reactor pressure vessel head is not bolted to the vessel.

9. (The APRM dowr6c.eks trip is automaticeRy tW when the IRM Instrumentation is operable and not iAm

10. An APRM will be considered operable if there are at least 2 LPRM inputs per level and at least 11 LPRM inputs of the norm

11. E : C::^i .1' ? *.' D el d d

12. The APRM Flow Referenced Neutron Flux Scram setting shall be less than or equal to the limit specifi~11n the Core Oper

13. The Average Power Range Monitor scram function is varied as a function of recirculation flow (W). The trip setting of this function maintained as specified in the Core Operating Umits Report.

j ,14.neutron The APRM flow biased high neutron flux signalis fed through a time constant circuit of flux signal does not incorporate the time constant, but responds directly to instantaneous neutron flux.

approximately 6 seconds. T

15. This Average Power Range Monitor scram function is fixed point and is increased when the reactor mode switch is pla 1

h. Indrummlaisd comm,gfg pcg5 N McM h paqir 43 A .

I am oe.om. 9 9.v.v.v.p.v. 29. ip. ip. ip #

43-(

___ _ _ _ _ _ _ . _ _ . _ _-_-.--_____u __,-+u

, - e ,,-cm r, e . w. . . - r vm , <. e n.

vm-=u- + -~, sm,- .....w-., . - , , - ,, .e, , e- ,

.-m . , , ~ ,

1 JAFMPP TABLE 3.1-2 -

REACTOR PROTECTION SYSTEM INSTRLASENTATION RESPONSE TIMES t -

TRIP FUNCTION REACTOR TRIP sMM RESPONSE TIME .

(Seconds)

1) MWesef Presouro-High

' 1 0.550 (02-3FT-55A, B, C,0) 4

@ Drywat Pressure-ligh < 0.000 (05FT 12A, B, C, D) -

@ Reactor VMor Level- Low (L3) 5 1.050

$X2 3LT-101A, B, C,0)

4) Main!NeemlordssaariWfve Closure ' ~"

pSPNS4042, B2,C2, D@

psPNSasA2,B2,C2, De D E L E-TE D

5) Turbine StopWivo Gosure 5 0.000 (94PNS-101,102,103,104)

6) Turbine ControlWlve Fast Closure 5 0.070 (94PS-200A, B, C,0)

7) APRM Fixed (120%)lepiNeutron Flux 1 0.090 (2)

Q APfte Flow Referenced Smidened Thermal Power 5 0.090 (t) (2)

Noteslor Tatdo 3.12-

1. Trip system response Mme does not incdude the simuisled thermal power time constant of approximately six seconds which is calibrated separateh.

2. Tetp system response Ome le tie measured time interval from trip signal input to the Hrst electronic component in the channel after the LPRM denar4ar unII Wie screm prot valve solenoids de-energize (05A-K14 scrarn contactors open).

Amendment No. p6 43a

- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - - + - - . - -

t , .

JAFNPP Table 4.1-1 REACIDLPRDTECTIDK_EYSTEILISCEAMI INSTRUMEarrAT)oed TWST REGyIEeMgrwg FuncbEnl Ted g7 T'rip Fh ,+ sed

_1 ;'

Grous[2D Functiemal Test

_ .Frn-mcw i3)O Mode Switch la Shutdows A Place Mode Switch in Shutdown Qach refueling out @ 8 Manual Scram A Trip Channel and Alarm [Every 3 months M Q BPS Chamael Test Switch A Trip Channel and Alare Every refuelleg outage or after chamael malatemance g(

I 1RM IOnce per week during refueling Nigh Flum C Trip Chamael and Alarm (4)

,J or startup and before each

'**** N 5/UandW(N)

. ' IBM Trip Channel and Alarm (4) F0nce per week during refuellag Isoperative C or startup and before each 4

J""'W s/o -,1 w (i)

APRM (Dece/ week)---% Q Migh Flux B Trip Output Relays (4)

' 3 Trip Output Relays (4) IP- =/-- " W Q I :;:rative a Trip Output Relays (4) Once/ wee d "

b ; scale

- - - (4) @ ace / math (1&- Y h Flow Bl. sed N(ak FINx - - - - - - - - - - -

Trip Output Relays (4)- ce per week during refueli Migh Flus'in*St&rtup or Refuel C 6orstartupandbeforeeachea,tu,yn

{ UYly Oaput Rel.y (@

3 Trip Chamael and Alarm (4) (Sace/ month. (t)(a) M Q g ReactoWPressure B Trip Chamael and Alarm (4) (Osce/mooth. (1)(0 M Q .

n f (Enghprywel19 Pressure Beactor Lew Level B Trip Chassel and Alarm (4) (Once/ month. (1)(S M Q ,

Trip Chamael month. 17 D- Q ((o)

High Mater Level la Scram A Discharge lastrument Volume High Mater Level in Scram B Trip Chamael and Alarm (4) %ce/ month. (1)(S b Q Discharge Instrument Volume

A . N Regime ude g wAkn D ma

< Amendment No. . . ,

, 8 44 g g w g p geg ggg

JAFNPP TABLE 4.1-1 (Cont'd)

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT A TscN

^";; ^^2^ r_""

. .r= C0T TT20-".=20 ren : w; ::;;- ;^. :; ;_-- 7EST;;;- REQWlR gg

nre

n Mgyrg TM Fm+rn . .

Raat r, +

s tr.-.;.,; C: m: C Groun din (LldeO Furrruvial Test =f

..._--Xeouency (JnfAndt 3) Th5frunitidChk n

Main Steam Une Isolation Valve Closure A Trip Channel and Alami Fa3+ clowre Nee / month.f1M Q Nk Turbine Control Valve 3 C l0 ^" Pi-m A Trip Channel and Alarm COnce/montit}-* G HA Turbine First Stage Pressure Permissive B Trip Channel and Alarm (Dnce/ month.(1)(8_)}--!P h p Turbine Stop Valve Closure A Trip Channel and Alarm (Once/ month.(1))- -* Q NA NOTES FOR TABLE 4.1-1 bfflAst W h IMSWE d

1. (Sitially once every month until acceptable failure rate data are available; thereafter, a request may be made to the NRC to change frequency. The compilation of instrument failure rate data may include data obtained from other boiling water reactors for which the same (design instrument operates in an environment similar to that of JAFNPP.f

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

3. Functional tests are not required on the part of the system that is not required to be operable or are tripped.

If tests are missed on parts not required to be operable or are tripped, then they shall be performed prior to retuming the system to an operable status.

4.

This instrumentation is exempted from the instrument channel test definition. This instrument channel functional test will consist of injecting a simulated electrical signal into the instrument channels Amendment No. l}4. 6/,1)d,J3tf 45 l

JAFNPP Table 4.1-1 (Cont'd)

REACIQR_fRQIECIION SYSTEM (SCRAM) INSTRUMENT FUNCIlONAL TEST MINIMJM FUNCTIONAL TEST FREQUENCIES FOR SAFETT INSTRUMENT AND CONTROL CIRCUITS NOTES FOR TABLE 4.1-1 fcont'd)_

5. h " Weekl y 4H'h0 4l csd rt' fredGulyduvid def M $}.y )gg}g

_he m . ,:==.

TO '

kg3M b [. The functional test shall be performed utilising a water coluna or similar device to provide assurance that damage

, = to a float or other portions of the float assembly will be detected.

I w~ 7l. ==. -: e =' --- :-

\

Amendment No. . . . ,

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JAFNPP' 32 LIMITING CONDITIONS FOR OPERATION 42 SURVEILLANCEREQUIREMENTS 3.2 INSTRUMENTATION 4.2 INSTRUMENTATION Appilcabihty: Wa8*y:

Appiles to the plant instrumentation wtSch either (1) initiales and Mh r@ement d the hadon which controts a protective function, or (2) provides information lo aid the M and Wrd prdecHve W,or M provides operator in monitoring and assessang plant status durinD normal and b am opera % moruknng and a% plert atmus accident constions. during normal and armiant conditions. .

p. Objective:

To assure the operability of the aforementioned instrumentWion. To specify the type and frequency of sunellance to be apphed to the aforementionedinstrumentation.

Speciscations: Speconcations:

l A. Primary Containment isolation Functions A. Primary Containment isolWlon Functions When primary containment integrity is required, the limiting M W *aW as conditions at operasion for the instrumentahon that initimes -

primary containment aardanian are given in Table 3.2-1. System logic shen be fur-A ~; tested as indcated in pnmary containment integrity is required, the primary a M 24. ,

containment lenimainn areamann instrumentation response time (The response time of each primary containment lentasann .) .

for MSIV closure shen be within the limits in Table 3.2 9. actuation instrumentation inniasing trip furstion listed in Table s

32 9 shen be demonstrated to be within the limits in the table; bply < "5 durtre each 18 monsh teet irWervd./Each led shsE include at win seesi one channs in each trip system AR channes in both Irlp 4g systems shaR be tested within two test intervals.

I. H sIV c togur e - Reador Lo lader Les,alluY '

Jj or-si.r- r7A n-r cr_ n '

z - (ge,,v,._gy,,;5 < - ~ ao e- . .c< Pr,, o <<

3. N iv clos - ,g ,L4, 4 m _ t,a , g io .

(.ot o vr - n os, A -o, so , A .o , p,s p.,, ,,4 g ,)

Amendment No. Ulti, y6

49 ,

-w,,-- m--%,-,-.., - , .---4v- ,-.-,my,,..-- , -- ,r-.- ,-e--, .. , - -+ , -,--,---,-w, .we,--. --,.----r - - , - - , - - - , . , . - 2-,-.,-

JM NPP 1

42 W 4 32 kont'4 B. Core and Containment Coonna Systems - Initiation and Control

& N and N W W-Mmeuon and W instrumentation shaN be hinctionsAy tested, caibrated, and h

ma W m for W b h h Wiet Wie N and W W W checked as indicated in Table 42-2.

are given in TatWe 3.2-2. This instrumeriamaimi must be operable System logic shaN be functionsAy tested as hdicated !- .

when Wie systemp) it intilstes or conircis are required to be Table 42-2.

operable as specilledin W 35.

l c. conemt nod stock Me c. contros Rod esock Actuanon

- um=*sim she be amenoney tested. cobraied, and Q The emang condmons of op emon ior 9,o ins n,ne,. anon Med as Wed in Table 424.

(that inilletes conIrci rod block are given in Table 32-3.

System logic shall be functioneRy tested as inc5cated in

2. The minimum number of operstdo inegrument channels Table 42-3.

speclRed in Table 323 for Wie rod teock manlior may be r=*M by ano in one of the trip systems lor maintenance and/or tesung, provided that Wils contStan does not last m longer then 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 30 day period.

D. Radiaton neonnoring Systems - % and inmenon Functions D. RatSetion 4Aor14toring Systems - Isolation and ini:lanon htb 6 Refer to Wie Re:Sological Eleuent TN SpecmceHons (AppentAx B). Re8er to the Radiological EfRuont Technical SpecMcations (AppentSx B).

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B t i

gS n emgi t i g ni s a m D ot l n esn e d n

s mt eu a nn3 hut f a e er r u nt o .i r i buii, t ai d nptoh heinCooier rct e5 cs oClot imt t el iv e 2 t u

i eae haiae tl ss cit c nla eic i

n v gi m 3 i Tt r d e E c n c. s a 8 T m i S in a A l l l l lV m

- 4 ! '

L g

, 1 ..

JAFNPP '

3.2 BASES (cont'd) '

r High radiation monitors in the area of the main steam lines The trip settings of approximately 300 percent of design flow have been provided to detect gross fuel failure as in the for this high flow or 40*F above maximum ambient for high control rod drop accident. A trip setting of 3 times normal temperature are such that uncovering the core is prevented and full-power background is established to close the main steam fission product release is within limits. t line drain valves, the recirculation loop sample valves, the -

mechanical vacuum pump isolation v'alves, and trip the pumps, The RCIC high flow and temperature instrumentation are to limit fission product release. For changes in the Hydrogen arranged the same as that for the HPCI. .The trip settings of Water Chemistry hydrogen injection rate, the trip setpoint may approximately 300 percent for high flow or 40'F above be adjusted based on a calculated value of the expected maximum ambient for temperature are radiation level. Hydrogen addition will result in an increase in sed on %same the N-16 carryover in the main steam.

criteria as the HPCl.

-c eed FFj T. '

The reactor water cleanup system h pM Pressure instrumentation is provided to close the main steam instrumentation are arranged similar to that for the HPCI. The

' isolation valves in the run mode when the main steam line trip settings are such that uncovering the core is prevented and pressure drops below 825 psig. The reactor pressure vessel

~

fission product release is within limits.

thermal transient due to an inadvertent opening of the turbine bypass valves when not in the run mode is less severe than The instrumentation which initiates ECCS action is arranged in a -

the loss of feedwater analyzed in Section 14.5 of the FSAR, dual bus system. As for other vital instrumentation arranged in therefore, closure of the main steam isolation valves for this fashion, the specification preserves the effectiveness of the -

thermal transient protection when not in the run modo is not - system even during periods when maintenance or testing is required.

bemg performed. An exception to this is when logic functional testing is being performed, The HPCI high flow and temperature instrumentation are -

provided to detect a break in the HPCI steam piping. Tripping The control rod block functions are provided to prevent of this instrumentation results in actuation of HPCI isolation excessive control rod withdrawal so that MCPR does not valves. Tripping logic for the high flow is a 1 out of 2 logic. decrease to the Safety Limit. The trip Amendment No. 1/,3/,:)d.p,p,1/4,1[7 2[

57 4

_ _ _ _ _ _ . _ . _ - _________________.___.__m_m - ._ m_ _.__ __._ . _--._- -- m.,_.-- n -..-2m . . . , . . , . . . . - , - - . , , . - , , . . . , _ -

. ~ . . . _ _m ,

1 hj i{~

siiIPb 1 11

!I!!Nli l9i l!!!i i

5"!!M$!D x /ix .

.i ii si, hlg;jd;l

,lilh 1Co i

it!!iintill 0!Il i

JAFNPP 4.2 BASES (cont'd) calculated test interval must be bypassed during testing. The The above minimization of the availability is each individual channel, optimizes it to be independent of all illustrated by curve No. 1 of considering 4.2-1 which assumes that a As an example, assume that Fig.

others. channel has a failure rate of are two channels with an there O.1 x 10-6/hr and that 0.5 hr is individual technician assigned to each. required to test it. The unavailability Each technician tests his but channel the at thetwo is a minimum at a test interval i, of optimum frequency, -

are not allowed to 3.16 x 103 hr.

technicians the communicate so that one can advise If two similar channels are used intest a1 other that his channel is under test. out of 2 configuration, the Under these conditions, it is possible interval for minimum unavailability for both channels simultaneously. Now, toassume be under that the test changes as a function of the rules for testing. The simplest case is to test technicians are required to communicate each one independent of the other. In and that two channels are never tested this case, there is assumed to be a at the same time, finite probability that both may be by-passed at one time. This case is shown Forbidding simultaneous testing improves by Curve No. 2. Note that the unavail-the availability of the system over that ability is lower as expected for a which would be achieved by testing each redundant system and the minimum occurs channel independently. These 1 out of n Thus, if the at the same test interval. independently, trip systems will be tested one at a two channels are tested time in order to take advantage of this the equation above yields the test inherent improvement in availability. interval for minimum unavailability.

Optimizing each channel independently A more usual case is that the testing is may not truly optimize the system not done independently. If both considering the overall rules of system system channels are bypassed and tested at the operation. Ilowever, true same time, the result is shown in Curve The optimization is a complex problem.

not sharp, and No. 3. Note that the minimum occursior at optimums are broad, about 40,000 hr., much longer than optimizing the individual channels is Cases 1 and 2. Also, the minimum is not generally adequate for the system. nearly as low as Case 2 which indicates that this method of testing does not The formula given above minimizes the take full advantage of the redundant unavailability of a single channel which N

. t e r I 1 [ hN

_ 62

i I

JAFNPP !

4.2 Bases (cont'd) channel. Hypassing both channels for #

2 More than one channel should simultaneous testing should be avoided. not be bypassed for testing at The most likely case would be to #

stipulate that one channel be bypassed, -

tested, and restored, and then i immediately following, the second ,

j channel be bypassed, tested, and restored. This is shown by Curve No. 4.

Note that there is no true minimum. The curve does have a definite knee and very {

1 little reduction in system unavailability ',

is achieved by testing at a shorter :

interval than computed by the equation for a single channel. l .

l

{

The best test procedure of all those ;

examined is to perfectly stagger the -

I tests. That is, if the test interval is The automatic pressure relief 5 '

four months, test one or the other ,

channel every two months. This is shown instrumentation can he considered to be a 1 out of 2 logic system and the bases in Curve No. S. The difference between given above for the rod blocks apply i Cases 4 and 5 is negligible. There may here also and were used to arrive at i be other arguments, however, that more J l qthe functional testing frequency. l ;

strongly support the perfectly staggered jr tests, including reduction in human error.

i The conclusions to be drawn are these:

1. A 1 out of a system may be treated the same as a single channel in terms of choosing s%, a test intervals and AmendmentNo./.[ 63 ,

JAFNPP TABLE 3.2-1 [wge,,f S

[JNSTRt *"""MTATION THAT INITIATES PP".RY CONTA"2""cNT ISOLATION)

Minimum No.

of Operable Total Number of instrument .

Instrument Channels Channels Per Tr[p Iymeh ,g Provided by Design J Trip System (t)(f) Anstrument Trip Level Setting ior Both Trip Systems Action /)

2 eactor Low Water Level 7) h 177 in. above TAF IMS#rk ,P L Caeker Lene WaffeL(tvdI(hjekl7th 2177 eb . alpsvaA4 T'A F 4 A

1 Reactor High Pressure s 75 psig 2 D (Shutdown Cooling isolation) 2 - Reactor Low-Low-Low Water Level 2 18 in. above the TAF 4 A p[ 2 2,g 2

[gh)Drywelpressure(7)

Drv tism iamahonlNiai un mala Pecosurd(k;dfth s 2.7 psig f 2 7 psie' 4

2. A A

s Normal R'ated' 4 E Steam Line T

, Full PowerJgckground j 2 (Low PraceEd Main Steam UneN 2 825 psig 4 B L$fd 2 CHiah Flow) Main Steam LineW s 140% of Rated Steam Flow 4

)lf 6 g/ Main Steam Une Leak s 40*F above max ambient M/ B Detection High Temperature 4 React eanup System Equipment s 40 F above max ambient 8 C Area Hfgh Temperature ,

2 kaai)Condensbacuum b z 8" Hg. Vac 4 B w

Y Amendment Nojif, 77, $, f/, gh,1/3,1/9,1/2, IIS, 2$3M 6 2.

4+-

a

. ~o JAFNPP -

TABLE 3.2-1 (Conrd)

[MSFN $

gSTRUMENTATION THAT INITIATES PRIMARY CONTAINMENT ISOL&I1Q]M -

NOTES FOR TABLE 3.2-1 g Igged y (9hyh CA rM$dd A$t.S't f Whenever Primary Containment integrity is required by Section 3.7, there shall be two operable or tripped trip systems for each funct 2.

L From and action appropriate after the listedtime below it is found shall that the first column cannot be met for one of the trip systems, that trip system shall be trip be taken.

A. Place the reactor in the cold condition within 24 houG.

B. Isolate the main steam lines within eight hours.

C. Isolate Reactor Water Cleanup System within four hours.

D. Isolate shutdown cooling within four hours.

E. Isolate the main steam line drain valves, the recirculation loop sample valves, I and the mechanical vacuum pumps, within eight hours.

(3. Deleted f 6 Delet

[5._

Two required for each steam line[

q jf. These signals also start SfGTS and initiate secondary containment isolation.

=y /. Only required in run mode (interlocked with Mode Switch).

M cJ. QMN TA4**'[88kl$Ikt.I#5_Mf _?.dbf3"(f 5 hf V815

T 8E__****

y

,n,...~. .. - .. . ~ . . . - . - - . . . - - . , _

'7, Lsert N w ._

NhY l

j Amendment No. )I. [8, f f.1[2, If9, If2 7 65

. . _ - - . __.____ a

JAFMPP TABLE.3.2-2 -

b5fd 8 O-nmiaanrrATION THAT ]MITI ATES OR CONTROLS THE_ CORE AND_COMIAIMMENT)

~

C00LIIHLEYSTEC -

Minimum No. Total of Operable Number of Instru-Instrument j[M ment Channels Pro-Item Channels Per K vided by Design for

_Ho. Trio System ill[d Trim F - tlam Trig _ Level Settina Both Trig _ Systema Remarks 1 2 Reactor Imw-Low 1126.5 in, above TAF Initiates HPCI, Water I4 vel 4(HPCI (Inst. Channel& RCIC1RCIC [ & SGTS.

2 2 Reactor Low-Low- 118 in. above TAF 4(Core Spray & RNR) p Initiates Core Spray,RglR Low Water Level (Instrument ChannelaT (Diesel Generators.LPCf.and 4(ADS) natrumen finitiatesADS'iaconjuncT

-1s f- tion with confirmatory low level, 120 second ,

y time delay and LPCI or

h$f d S S Core Spray pump discharge pressure interlock if nei g 4 gg " inhibited by ADS override (switches. ;

hf _ _

3 2 Reactor Nigh Water (222.5 in. above TAF 2 inst. Channels' [ Trips FPCI urbine.eme.

L.v.1 _:____ r:_ n= sim

->= .

E rr:nf r- ;11__.

/f 1 (Ngtgg Reactor Low Level 10 la. above tar 2pst. Channels Prevents inadvertent (laside shroud) operation of contain-ment spray during I i accident condition.

H- 7-Re-+.< Q Wahr e zu.u.a,,,rv t.(Hote e),

L. tvd C.lcses Re w. d em SEgly VAbf .

i,ena ent No. w , yr, 69, 54 /'9 66-

_. - _ _ _ - . . . _ _ . _ . . _ . _ _. . _ . . . _ . . _ . - _ . . _ . . ~ _ .- _ _ .._.__ _ __ . _ _ . _ _ . _ - . . . _ _ . _ _ _ , _

JAFNPP TABLE 3.2-2 (cont'd) UU pSTRUNRMTATIONTHATINITIATESORCONTROLSTHECOREANDCONTAINMEN COOLING SYSTEMSf Minimum No. *

" Total of Operable Instrument Number of Instru-Nh3 l AnM L ment Channels Pro-Item Channels Per No. Trip System Trip Ftanction vided by Design for Trip Level Settinst Both Trip Systems Remarks h/ 2 Containment High 1(p<2.7 peig 4f inst. Channels Pressure Prevents inadvertent operation of containement spray during accident condition.

Reador.,

7/ 1 ghq C-.. : . . Low d117 in. above TAF 2 Inst. Channels t"^

Level .~.-.-- '. - '

'-'j- -ti - "l'h i

"--'_. L. 1. .. _ _

o_.__ .___,

g/ 2 Dryweli k [2.7psig y PCI Inst. Channels k

Ini Pressure lates Core Spray, M (

(LPCI HPCI and SGTS.

q[ 2 Reactor Low Pres- 1450psig 4 Inkst. Channels j / L Persh ' ~opeR*R sure -- .-

. ing f Core Spray an %PCI

" "ir: L.. va10es.

j lu{tsHow

- NirmMory l8W Water ,

l levd ter APs uhdies Amendment No. 14, AG , Ff , 74, 9 i

61

,i._,.-

r ,ws,- =r,c, v-v- =,,----,-w,- -

-r- r . , , , . - . , - - -- .mm_- ._-_ m - * - - - - - - - - _ - - - - _ _ - - - - - - - - - - - . - - - - - - - - - - - - - - - -

Q . .

l JAFHPP TABLE _3.2-2 lcont'd1 h* N

[

[ESTEllM E NTAI l Q!LTl! AI_ I N [Il hT E S_QR_CQ NT R O LS_Tij E_f0RS_M D_

N OQLING. SYSTEMS -

Minimum No.

of Operable Total Instrument Item Channels Per Ak * )\jgh l,g" Number of Instru-ment Channels Pro-

_Fo. Trip _SysLg M [TriDFunstign vided by Design for Ttip_Lgygl_Sgtling Bath _Tri1LSystems

[lO 1 hohg4 Reactor Low Pres-Remeds sure 50 f p _4 75 psig 2[ Inst. Channe (Inconjunctionwit}

LPCIS slanalfpermits closure of RHR (LPCI) io lajection valves ru1S 1res rarturionia.r er.xxx 7 L

i J9, skas,wn

( 11 ruxSrreninrearrouxLLrsting 1 g Core Spray Pump (See Notes 3}g Start Timer 11 1 0.6 sec. 1 Inst. Channe Initiates starting of (each loop) corespraypump/.

NC (each loop)

Amendment No. [ , [ [ . %

68 l -_-_

s JAFNPP TABLE 3.2-2 (Cont'd) U NSTRUMENTATION TilAT INITIATES OR CONTROLS Tite CORE AND COtRAINMEtR]

COOLING SYSTEMS f Minienum No. Total of Operable u n Number of Instru-Ins t rument kh*J l AntM 1 ment Channels Pro-Item Channels Per vided by Design for No. Trip System ( Trip Function Trip Leve' '*ing Both Trip Systems Hemarks IL M (LPc0 Ril Pump Start Timer (Nde 6) 1 E$

Ist Pump (A Loop) 1.0 + 0.5 (-) O sec. 11 Inst. Channel j Starts lot Pump (A Loop)

(See Note 53) let Pump (B Loop) 1.0 + 0.5 (-) O sec. 1 Inst. Channel Starts 1st Pump (B loop) 2nd Pump (A Loop) 6.0 + 0.5 sec. 1 Inst. Channel Starts 2nd Pump (A Inop) 2nd Pump (B Loop) 6.0 + 0.5 sec. 1 Inst. Channel Starts 2nd Pump (B Loop) 1 M Auto Blowdown Timer 120 sec + 5 sec. 2 Inst. Channels Initiates ADS

? . . . . . _ .

am - . . , ,

ihnot inhib d by ADS >verride i

I switche(s)

W P n.'+

/4 ""a ('"c') ""=e 225 ysi, + 20 gsie g/ inst. Chenneig .e s.sADS actuetion.

Discharge Pressure i. q 7n-Interlock af ' . - _ ,. _ _ _ _

M Amendment No. / M f ( , [ 69

+ ,.

JAfHPP TABLE 3.2-2 (Cont'd) [MWF @

[STRUMENTATION THAT COOLINGINITI ATES SYSTEMS p OR CONTROLS THE CORE AND CONTAINM Total Minimum No. Number of Instru-of Operable ment Channels Pro- 'c__,,

M

Nhd ad[

Instrument vided by Design for "F'i; "~-g Itse Channels Per Trip runction Trip Level Setting Both Trip Systems Remarks rS N h risj Ho. Trip System Wlf 2 Core Spray Pump 100 psig + 10 psig 4 ( . Channels _ ,,

ADh ac ;uation, pending -

Discharge Pressure ' ~ " ' " ~ ~ ~ ' A 27~2 i

Interlock ZZ;

-^ '"

" , '2z :;;- _Moe 2

Q Monitors availability Inst.' Channels EU 1 M MIR (LPCI) Trip loss of Voltage 2 of power to logic System bua power systema.

monitor loss of Voltage 2 Inst. Channels Monitor availability hel$ 1(Ngtt Core Spray Trip System bus power of power ) logic systems.

monitor Ioss of Voltage 2 Inst. Channels Monitors availability W l$ 1 Nok'n ADS Trip System of power to logic bus power monitor systems.

Inst. Channels Monitors availability nect Trie Sr=ta= to== of volta 9e 2 N>-lg 1(Nhe)t bus power monitor of power to logic systems.

Monitors availability Btj$ 1 Ndeh RCIC Trip System Ioss of Voltage 2 Inst. Channely of power to logic

' / bus power monitor systems.

4 Amendment No. Y, /$ 70

4 J AFNPP --

IABLE 3.2-2 ( cong 'sil 1hWrh-%

@TRUMENTATIONTHATINITIATESORCONTROLSTHEC'OREANDCONTAINMEN)T

. M OOLING SYSTEMS 7 Minimum No. Total of Operable Number of Instru-Item Instrument Channels Per h ment Channels' Pro-vided by Design for No. Trio System (ld,d, Trio F= -tion Trip Level Settino Both Trio Systems [ Remarks mR 2 Condensate Storage 159.5 inches above Channels Transfers I<CIC pump

! Tank Iow Imvel tank bottom % g [ Inst. suction to suppres-(= 15,600 gal. avail) sion chamber P

(3 This Item Intentionally Blan_lQ b

G4 This Item Intentionally Blank)

IO Mdadak L X 7.,\ l car

8erar Sear 9er 4_- o.5 esta 2 In=t- Channels (Noies)i to Reactor Pressure Alar =5*.3*o*m.c core spray sparger pipe vessel d/p break.

X L} 2 . Condensate storage A___59.5 in. above tank 2 Inst. Channels Transfers HPCI pump Tank Low Level bottom f suction to suppression

(=15,600 gal avail) Ncka chamber.

k l M E4 2 Suppression Chamber S 6 in. above normal 2 Inst. Channelsi Transfers HPCI pump High Level- level -

d suction to suppression chamber.

! 26- 1 BCIC Turbine Steam 4_282 in. H 2O dp '2 Inst. Channels c'--- '- '-' - " ' --

Line High Flow k -3P ' 20:0 :- , ^ --

1

% %e h Table 3 z-I -

Amendment No. , O ', 4 70a t

JAFNPP b

TABLE 3 2-2 (Cont'd)

[b9

[lNSTRUMENTATION THAT INfTIATES OR CONTROLS LTHE CORE AND CONTAr~NT COOLING SYSTEMS Mnmum No.

of Operable g instrument M Total Number of item Channels Per instrument Channels No. Provuled by Design TripSystem (1)(a Trip Funchon Trip LevelSettsng for Both Trip Remarks

.er i KC Steam Une/ <<rF Above (^^

- iinst. Channels 1 ": ':-.'.'_' n AreaTernparature rnax. amtnent in W " ' , _ .. -^> F W 1 BCIC Steam Une 100> P>50 psig 2 inst. Channels f'*- :; J m . 'J_^-

Low Pressure inJ4GIGeabsystem "I M- l1 HPCI Turtune Steam F

< 160in H2 O dp e F'I 2 Inst. Channels N Une High Row W 5"S ^',f_.,

e i1 RCIC Turtune High < 10 psig 2 lost. Channels Exhaust Daphragm Clees.4eeleesseValves gp

{ ' T'O G _- p . . 1 Dressure w 1 HPCI Turtzne High ~< 10 psig 2 fnst. Channels Exhaust Daphragm GlosoJoelassenMaiwee Pressure in Hpr % 7 F 1,y LPCI Cross-Connect NA (N N Alavos 1(No}e1) Valve Posetson 1 Inst. Channels aswesetee enmanesateen.

when valve is not closed,

& fV HPCl Steam Une 100 > P > SO psig 2 inst. Channels g}1 Low Pressure CaesoJoelelsenMetre

,..vi - r=="=

p l

W 1 HPCI Steam Une/ < MrF l q Area Temperature above max. ambient 2 Inst. Channe3 N ,,_:9F

'- : "J': d , ;t.T.

. , ,f W r Amendment No. 4,46,66, #4 ' % % gg yyg.p g M hfs

T4 e/thedet/T 1 O Yet he 3* bl C x A

.vh A -+ *

+

JAFNPP TABLE 3.2-2 (Cont'd) InM R I

INSTRUMENTATKJN THAT INIIIATES OR CONTROLQ LTHE CORE AND CONTAINEWENT COOLING SYSTEMS j Minimum No. of Total Number of

Operable Instrument Inshument Channels ltem Channels Per bid Provided by Design No. Trip System (1){ Trip Function Trip Level Setting for Both Trip Systems Remarks Z.6 .47 (1 per 4kV bus) 4kV E..-gency Bus 110.621.2 2 Inst. Channeld Undervoltage Relay f. Initiates both 4kV l t

(Nefe secerdery volts Emergency Bus Undervoltage '

(Degraded Voltage) Twners. (Degraded Voltage LOCA

/. (pnd Notesnon-LOCA) 4 and 6) 1"! . Gee - (1 per 4kV bus) 4kV Emergency Bus 9.0 1.0 sec. 2 Inst. Channett

age Tm ' f.(Note 5) l

{He$t 1) (Degraded waage tOCA)

I ,

t6 49b (1 per 4kV bus) 4kV Emergency Bus 45 5.0 sec. 2 inst. Channel 5 W T""*' f.hote5) l (Nefe9) (Degraded Voltage non-LOCA) h 11 -89 (1 por 4kV bus) 4kV Emergency Bus - 85 4.25 2 Inst. Channeln f. Initiatesi V Emergency Bus (y,9 4 Undervoltage Relay (Loss of Voltage) secondary volts Undervoltaw Loss of Witage Timer

f. iNotes 4 and j 3* * (1 per 4kV bus) 4kV Emergency Bus 2.50 0.05 sec. 2 inst. Channelt mage Ta [.(Note 5)

(Nefe1 (tass of Voltage) 1I W 2 Reactor Low Pressure 285 to 335 psig PerEds closered !

4 Lnst. Channejl ";. . _: ; fx :' f-g recirculation pump discharge valves J

nmendment No. f,g.1f,,A 1/0 70c

~r-. _ _ , - . - . , . , - - -- . . . - - ._m, . ~ m . , , . ..m.. .. - . . - . . . - -_ _ _ - - .

i l JAFNPP l

TABLE 32-2 (Cont'd)

INSTRUMENTATION THAT INITIATES OR CONTROLS

((MN 8 l

THE CORE AND CONTAINMENT COOLING SYSTEMS i

l NOTES FOR TABLE 32 2

[

l (1. Whenever any ECCS subeystem is required by specsfication 3.5 to be operable, there shall be two operable trip systems. From and after the time it is found tuit the Erst column cannot be rnet for one d the trip systems, that trip system shall be placed in the tripped I condtion or the reactor sher be placed in the cold condibon withen 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

Deleted"

3. i%fer to Technical Specification 3.5.A for limiting conditions for operation, failure of one (1) instrument channel disables one (1)

M l

( pump-47 Tripping of 2 out of 2 sensors is required for an undervoltage trip. With one operable sensor, operation may continue with the inoperable eenoor in the tripped condtion. [

5. The 48# Emergency Bus U A ciiise Timers (dograded voltage LOCA, degraded voltags rmLOCA,and loss-of-voltage) initiate the following starts the Emergency Diesel-Generators; trips the normal / reserve tie breakers and trips all 4kV motor breakers (in cor4 unction with 75 percent Emergency Dieselfmnerator voltages); initiales diesel-generator breaker close permissive (in corjunction with 90 percent Emergency Diesel-Generator voltages) and; initiates sequential starting of vital loads in conjunction with lowdow4aw reactor water level or high drywell pressure.

6. A secondary voltage of 110.6 volts corresponds to approximately 93% of 4100 volts on the bus.

l

7. A secondary voltage of 85 volts corresponds to approximately 71.5% of 4100 volts on the bus.

h*Il~=>8. 04 ou +,.;p sp+ew Add-9 4 Sd gle c.kaneel brip rysA .

b A-d-,No.H.+.+,10 71 i

I

- --. - - . . - - - + , . ,, . _ -. _ - - _ _ - - - - _ _ - _ _ _ . _ _ .

, , mi JAFNPP TABLE 3.2-3 IM *N 3 (INSTRUMENTATION THAT INITIATES CONTROL ROD BLOCKS Minimum No.

of Operable Total Number of instrument F=eA4 instrument channeis Channels Per) Provided by Design Trippyelesa bu)a, Trip Level Setting or Both ChannaFO Action (q Neby Ia.J A 4astrument Trift--J Flow W +<,Reice.sn...k A

/% th

%gl APRMg'; :i 74.. C':: "; '

~

r M- (ggg 6Ns 34 A '

yF APRMjf;: Nah11 @; Fhe-5t=49 w h) 5 12% 6 lee 6 Ghannels %A 42 APRM Downscale ,

> 2.5 indicated on scale 6 lae6. Ghennels %-M l L #M (f4 #fC'T Rod Block Monitor (Row Biased) * (Nete h - 2 '~' C'1-' :

%B 2,#$17(W}ch Rod Block Monitor (Downscale) > 2.5indicat on 2 lael.Ghamaels, m (, y

%B IRM Downscale[ [pj, pg > scal 8 '~

C"1M

(, #

%d IRM Detector not in Start-up Position W (g{g G lack Cha...i

-/ k 64 IRM Upscale 5 86.4 ull scale E 8 5 :. C h a ..,-

. J/)A f g% SRM Detector not in Start-up Position M (hje)t5) 4 '~' Ch=i RM Upscale 5

[

3 / % % (t4ct 510 counts /sec 4 leel Ghanne'*

ty Scram Discharge Instrument PA 5 26.0 gallons per 2 lae6.-Ghannele "" " 2 Volume High Water Level instrument volume c.fg}c o)

NOTES FOR TABLE 32-3 (C For SRM the Start-up and Run

( The and IRM block need positions of the not be operable in Reactor run mode,Mode and _ Selector Switch, there shall be two operable or tripped trip syste Adf Wafh b seed f4aggg,yg h e 97, Amendment No. 44, GE,7$,96 1)d 72

. , _ . , -.- - .. . - . . . . . ~ , .. ._ __ . _-._____.______.___._____m___. -__

d JAFNPP ,

TABLE 32-3 (Cont'd) - LSed 3 (1NSTRUMENTATION THAT INITIATES CONTROL ftOD BLOCKM i

NOTES FOR TABLE 32.-3 i

the IBM rod block need not be operable in start-up mode. When the reactor is in the start-up mode, the APRM upscale (start-up mode) rod) '

- block sheE be operable. When the reactor is in to run modo, the APRM upscale (flow biased) and APRM downscale rod blocks shall be j-operable. From and aAer the time It is found that the first column cannot be met for one of the two trip systems, this condition may exist for up to esven days provided that during ept time to operable system is fw-%C; tested immediately and daily thereafter; if this condition lasts longer then seven days, the system ehes be tripped. From and after the time it is found that the first column cannot be met for both trip (systems, the systems sheE be tripped.

(

i In e lush Tase-+ 'P

.2y IRM downecele is bypeseed when it is on its lowest range.

45' This function is bypassed when the count rate is > 100 cps. _,

' / . NON

?. L h C . 2 2 " ^ ; T :; & ; 7 -i g

}o_

R$ " SRM Function is when the IRM range switches are on range 8 or above.

ay 3 ,

Nt sleen

,~.._. _A_1&_6+er 98F'PJ3.1Mdf* __ _-

b er 4 e 30 /a, I N

y.r;g is tg' wheny Mode Switch is pieced in Run.

b trip level setpoint shen be less than or equal to the limit specified in the Core Operati i i Ilg.e:

.. . . _w. _

Report.

g.9:' When the reactor is subcntical and the reactor water temperature is less than 212"F, the control rod block is required to be operable only if any ,

control rod in a control cell containing fuel is not fully inserted.

O. When one of the instruments with scram decharge instrument volume high water rod blocks is not operable, the trip system shall tv tripped.

i i Amendment No. AG, se,7df,31f,146 3/2 73

=w-n -,m c -= e-m. r- e 4 ,,__=-_-$-r --&_- *---= - - <_ -- __

. + - - _ _ _ _ -=_s_== --m -m--. m .-.__ # &-.__- .___ m_ _ . _ _ _._-_--______.m_m _m_____

.- i.

j JAFNPP TABLE 3.2-7 i

ATWS RECORCUUmON PtNW TRIP AOftNWI996INSTRUWENTATION k re ti Ms r- 1 Apphcable Modes' Trip Aellose opor 1

per Trip A&M (4th ,

1 r%+n.med

! N h Reactor Pressure - Higt Run 2 < 1120pelg .

^

2. ; O 2- Reactor Water Level- Low Low Run 2 > 126.5 in. ^ 2. r C sove TAF Ndes for Talle 3 t-7 r ~

8<c neaf pec %er Nehldt son A.

[ (I8'*fe h revi,M ge 77 ;

When the number of operable channels is one less then the required number of operable channels per trip system for one or boti trip systems, restore the inopersbee channel to en operable condman wnhin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months N not restored wnhin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , pleos the inoperetse channel in e tripped condmon within one hour. N piecing the inopersbee diennel in the artpped condman woosd reeun in a redrcadeson pump trip, tehe Action C.

Action B.

When the twJrnber of operable channels is two less then the required msnber of operable channels per trip system for one or bott trip systems, elther restore at loest one channel per trip system to en operable status within one hour or place the inoperable channels in the tripped contseon within one ta r. N piedng the inoperable c'1ennel in tie tripped condelson would result in recirnasenri pump trip, tehe Action C.

Action C.

Q Action A or D is not completed within the ellowed time, be in the start-up/ hot standby mode within the next six hours r , -

Amendment No{rf,96, I)@,1jp4. Nt Wkl 136,46,57 41,64,/11,iW, leg

_ _ _ . _ _ _ . . . . _ _ =

4 JAFNPP TABLE 32-9 PRIMARY CONTAINMENT ISOLATION SYSTEM ACTUATION INSTRUMENTATION RESPONSE TIMES TFWP FUNCTION RESPONSE TIME (Seconds)

1) MSIV Qoeurs - Roscoor Low Water Level (t.1) 1.0 (02-3LT-57A, B and 02-3LT-58A, B) 5 i

@ MStV Closure- Low Stoem Une Pressure i 1.0 (02PT-134A, B, C,0) 3p MSIV Ooeure-High Steam Une Flow 2.5 (02 OPT-116A D,117A-0,118A-D,119A-D) 1 Notelor Table 3.2 The measurement of me roeparas time intervel begine when the monflored parameter exceeds the bolation actuation set point al the channel sensor and ends when the Main Steam tenasano Valve pilot solenoid relay contacts open. The pliot solenoid relay contacts to be used for determination of the end point of the response time measurement are:

For the inboard MSfV prot solenoid releys: 16A-M14 (acsolenoids) 16A-K51 (dc solenoids)

For me OWhoord MSfV pilot solenoid relays: 16A-K16 (acsolenoids) 16A-K52 (de soienoids) fle d To.bih -

Amendment No. 183 77e '

-, n . .

JAFNPP TABLE 4.2-1 L,pd Q C"~JM TEST AND CA mRATION FRFOliFNCY FOR PCN >

Instrument Channel Instrument Functional Test Calibration Frequency Instrument CheckM(Neta q}

1) Reactor High Pressuren,y ,g (Shutdown Cooling N... ~..

G%,y AdM MS n n Q _Orn'2 me...:.s N A Nene-1

2) Reactor Low-Low-Low Water Level Q % ,5) I'( (TS) T) Orn' ;-

3) Main Steam High Temp. Q .(5) R (15) D Orn'ii-

4) Main Steam High Flow Ug" Q !(5) g (15) D ^~r' ;

5) Main Steam Low Press Q i(5) D Onn'i,

6) "n;;;; '.".';;;: Cunc igh Temp. Q l Q, Ons/U .(.. ...~gNA(15) bleesp- blatel6)

7) Condenser Low Vacuu G (5) R (15) O Onn'i, M m Line High Radiation Q (5) Q/g (11) p On;;'i j Logic System Functional Test (Netes M 189 7

Frequency Main Steam Line isolation Valves SA Cn"/0 me :-

1)

Main Steam Line Drain Valves f Reactor Water Sample Valves

2) RHR - Isolation Valve Control 6 A Cnx!; m.i; =

N CVt 50 Shutdown Cooling Valves paqc 79

3) Reactor Water Cleanup Isolation sA n~c'e -- :: "c

4) Drywell isolation Valves sAO 'e =; 'h:

TIP Withdrawal Atmospheric Control Valves

5) Standby Gas Treatment System 6A Onn'S .- S:

R 1.wsed&eactor Buildng Isolation NOTE: See notes following Table 4.2-5.

Amendment No. [ f,1[6, If1, If, If0 h7 78

l - ., ,

.. v JAFNPP TABLE 4.2-2 rnardu V

hNIMUM TEST AND cat HERATION FREQUENCY FOR CORE AND CONTAINMENT COOLING Instrument Channel Instrument Functional Test Calibration Frequency Instrument Check 444 64ste')f

1) Reactor Water Level Q6,Gi(Nefef) SA/R Me-(Wekish D c :: "'.

2a) Drywell Pressure (non-ATTS) Q44+ Q Onee@weemns H A Nea+-

2b) Drywell Pressure (ATTS) Q 44996t (nan') R 96t (weie iv) o Oc= :;

3a) Reactor Pressure (non-ATTS) G 444- Q Omoe@mem%s NA tiene i

3b) Reactor Pressure (ATTS) () 199999-(N et' 'l D 0 ::!2 .

SA/R N (ucte153 R ~::;1_: w -ie

4) Auto Sequencing Timers NA mne. N A mene

5) ADS - LPCI or CS Pump Disch. Q (4+- Q Oneedamenehe. N 4 None-

6) Trip System Bus Power Monitors k 14+ N A usa

MA maa-Q ttt p e rr_" ; _

7f) Core Spray Sperger d/p G eneeAteenehs 1F- 9) { Steam Line High Flow [HPCI & RCIC) Q (4946t h sff I @ 4446. (No Ft 15) D Cn.., J. 7 O 10) kSteam Line/ Area High Temp.lHPCI & Rcig R M69- (.Wele 15) O cx::ffr, Q MH6t- (Nett f)

V llMt HPCI & RCIC Steam Line Low Pressure Q64446).(MottTD R n&6- (Nde 15D D en::! :,-

S 49) HPCI & RCIC Suction Source Levels Q 199- Q Qaeed&menths N A None

$ 44) 4kV Emergency Bus Under-Voltage R On::f:;_._;.., e g n_<:7- -g 7 e n yA u (t.oss-of-Voltage, Degraded Voltage ,

LOCA and non-LOCA) Relays and T'wners.

Q )2. 469 HPCI & R bt Diephragm Q 444- Q Oncedaeemehn. N A mene.

Pressur I' ;

10 4M L*Cl/ Cross Connect Velve Position ~ R 0=::!:::- : -; : f- N A None- NA Mene.

cr x -

NOTE: See notes following Table 4.2-5. Te irfheak4

Move twdrum#d and iesh'l u t. L,4 O .- pp va 1

Amendment No./I74. 49. Ep, IP,196,1 $),1(0,1$1,19)0 s __ m m D3dplmC w% 3,e'9,l6h iiSI,Lol #h

JAFNPP TABLE 4.2-2 (Cont'd) In 3M \/

F-^^^- ma TEST "D C_AI HMIATION FREQUENCY FOR CORE AND CONTA:N*ENT COOLING SYSTEMD Logic System Functional Test Frequency

1) Core Sprey Subsystem "7; '0; Oneo#reenths SA (Neks 7

dil Low Pressure Cocient injection Subsystem 97t+3) 2)

enoodGeesahs sA (N Fe -

3) Conteinment Cooling Subsystem Omoed&menths 54 Ws SA ( Nofr> 7# i\

~

4) HPCI Subsystem M M% HPCl Subsystem Auto isolation 499 OneedG.smesubs S4 Uefe7)

ADS Subsystem 0; ;;; "-'8 ---*s M ()Jefry 7 4 i\

5 /)

49F Once46 snenehs 5 A (N'fC7\

M f RCIC Subsystem Auto isolation 4aHe& ~ -'~ " , ~ i- ( p ,F ,, 7 4 13 l e .e f:"f1: : :_: m:r: c: ^^

NOTE: See notes following Table 4.2-5. & Mg g 7 4

-- s r .F(D 4

Amendment Noh 75,1EB. Ill % -

f_ Replace w(th 3,(ot. 81, Itot

8I ,

JAFNPP TABLE 42-3 M$ rrt W _

(MINIMUM TEST AND rw mRATION FREQUENCY FOR CONTROL ROD BLOCKS ACTUATIOR) instrument Funcuonal Instrument instrument Channel Test (5) Calitxation Check (4)' ;

1) APRM- Downscale (4- Q P l : / 2 .T 2 .- Q T> Oci-/t; '

n APRM- Upecale MQ &c:/ cc- : Q P Onoefday

3) IRM- Upscale 42) f/d (Nde th. Z M k(Neh JI6) -D ^ ::/f:1

4) 1RM- Downscale MkU(Nde2}. Z g Q(Notes 346 D O neopsisy

-@ RBM - Upecale 44).- Q Cn::l0 .-- 2__ g p Oncefder i 7jl) ROM - Downscale MQ P ::/0 n 2i- Q ' D C-/1,

[$/) SRM - Upecale (ES- T[V(Nok th M.--(Ei)Q(Mofc3Mg), 9 Gaie/88er 9p) SRM - Detector Not in Startup Position M-5[U(Nd,t) M- M NA NA Mene i (5p) IRM - Detector Notin Startup Posdion MS/p(p,g} W M NA -NA None

10) Scram Discharge instrument Volume- WQ WQ "D C z:/f:;

High Water Level (Group B instruments) M i

Logm System Function Test M Frequency

. (Mcfes74 9h

1) System Logic Check Oca;/S n- 29 5 A NOTE: See notes following Table 42-5.

f Beglue wi+41, en,9s Amendment No.GT, gr 88, 1EB g g, w

~_ . - .. - .-- -- . . .. . . , - - . . . . .. - - . -

.: t -h JAFNPP NOTES FOR_ TABLES _42-LIIStOUGH 4.2-5

1. Initially once every month until acceptance failure rate data are 7. Simulated automatic actuation shall be performed once each -

avestable; thereafter, a request may be made to the NRC to operating cycle. Where possible, all logic system functional change the test frequency. The compeletson of instrument tests will be performed using the test jacks.

failure rete data may include data obtained from other boiling water reactors for which the same design instruments operate 8. Reactor low water level, and high drywell pressure are not in a environment similar to that of JAFNPP. e mcluded on Table 4.2-1 since they are listed on Table 4.1-2.

2. Functional tests are not required when those instruments are 9. The logic system functional tests shall include a calibration of not required to be operable or are tripped. Functional tests time delay relays and timers necessary for proper functioning of shall be performed within seven (7) days prior to each startup. the trip systems.

3. Calibrations are not required when these instruments are not 10.(At least one (1) Main Stack Dilution Fan is required to be in required to be operable or are tripped. Calibration tests shelly---sp\s,,eist;cs in order to isokinetically sample the Main Stack.

be performed within seven (7) days prior to each startup or prior to a pre-planned shutdown. 11. Perform a calibration once per operating cycle using a radiation source. Perform an instrument channel alignment once every 3

4. Instrument checks are not required when these instruments months usmg the built-in current source.

are not required to be operable or are tripped.

12. (Deleted) g

5. This instrumentation is exempt from the functional test /

definition. The functional test will consist of injectog a 13(Calibration and instrument check surveillance for SRM and IRM l ;

simulated electrical signal into the measurement channel. et8---9 __ L Instruments are as specified in Tables 4.1 -1_ 4.1 -2. 4.2-3. J

6. These instrument chenriels will be cahbrated using semulated 14. / Functional test is performed once each operating cycle electrical signals once overy three months. WM

15. Sensor calibration once per operating cycle. Master / slave trip unit calibration once per 6 months.

L**Mc.

4

, g ,, gg .

Amendment No. p. %,5/, p,1/1 N 84

. , d ?_

JAFNPP TABLE 4.2-7 8"k X MINIMUM TEST AND CAUBRATION FREQUENCY FOR ATWS RECIRCtRAW pisup TRIP ACTUATION INSTRUMENTATION FUNCTION CHANNEL CHANNEL TRIP UNIT CHANNEL SIMULATED AUTO ACTUATION CHECK FUNCTIONAi. CAUBRATION CAUBRATION & LOGIC FUNCTIONALTEST TEST

[1 Reactor Pressure-High r

C ;/t/

Q C. a /0; i p 5A O;;/O ; . .; ;.; Ccs&

,4_ ::..; t;;/&gg 7.;: p D Q sA R J2 Reactor We'er Level-Low Low L Occe/1, &c:/? ? d , e 072 i

, , _ . _ n . _,>~v

_- . ,- , m W

l i

Amendment No.%,36,pf,JMT, /1 _.

I

- 1

(

uh - _ f) a-

- $=

Y T

I L

I On.

/ %

A B

A L

I A

V A

N U .

n M

E T

S Y

S

-p /

N 7 8

C F i.

O .

Y .

TI L

I

,e B

A B e.

O 9

R P

0t 3N =

l W (ne 5_ ,, 5 - _" O*

mmh 5Hm 1yCm f*,

_i :, EOD r~q< f g mZ Z ! ; E H<-

L H 2 C

JAFNPP 7.0 flEFERENCES (9) C.H. Robbins, " Tests of a Full Scale 1/48 Segment of (1) E. Jansson, " Multi-Rod Bumout at Low Pressure," ASME the Humbolt Bay Pressure Suppression Conteenment,"

Paper 62-HT-26, August 1962. GEAP-3596, November 17,1960.

(2) K.M. Becker, "Bumout Conditions for Flow of Boiling (10)

Nuclear Safety Program Annual Progress Report for Water in Vertical Rod Clusters," AE-74 (Stockholm, Period Endmg December 31,1966, Progress Report Sweden), May 1962. for Period Endmg December 31,1966, ORNL-4071."

(3) FSAR Section 11.2.2. (11) Section 5.2 of the FSAR.

(4) FSAR Section 4.4.3. (12) TID 20583, " Leakage Characteristics of Steel Containment Vessel and the Analysis of Leakage Rate (5) 1.M. Jacobs, " Reliability of Engineered Safety Features as Determenetions."

a Function of Testing Frequency," Nuclear Safety, Vol.

9, No. 4. July-August 1968, pp 310-312. (13) Technical Safety Guide. " Reactor Contamment Leakage Testing and Surveillance Requirements,"

(6) 'songemen Epstein, Albert Shiff, UCRL-50451, improveng] USAEC, Division of Safety Standards, Revised Draft, Availability and Reediness of Field Equipment Through December 15,1966.

Periodoc inspection, July 16,1968, p.10, Equation (24),

(Lawrence Redsetion Laboratoryy (14) Section 14.6 of the FSAR.

k ASME Booler and Pressure Vessel Code, Nuclear (7) 1.M. Jacobs and P.W. Mariott, APED Guedelenes for Determining Safe Test intervals and Repair Times for p g g (15)

Vessels, Section Ill. Maximum allowable internal [l; Engineered Safeguards - April 1969. pressure is 62 poig. lf Wlh (

(8) n 10 CFR 50.54, Appendix J, " Reactor Containment I Bodega Docket 50-205, Boy Proleminary Herards Report, Appendix 1. nDClded (16)Testing Requwements."

December 28,1962.

(17) 10 CFR 50. Appendix J. February 13,1973. l Amendment No. 1[ 285 u- .-__ _ _ _ _ _ . . _ _ _ _ _ _ . _ _ _ _ _ _ _

__ - __-- . - . _ . _ . _ . _ _ . _. _ _ _ _ .

ML20073L094

Text

7.0 REFERENCES

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