Text

Amend 131 to License NPF-1,changing Tech Specs Re Reactor Trip Sys & ESFAS Instrumentation Requirements
	ML20209D940
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Site:	Trojan File:Portland General Electric icon.png
Issue date:	04/22/1987
From:	Knighton G; Office of Nuclear Reactor Regulation
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ML20209D912	List: ML20209D908; ML20209D940; ML20209D962;
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	NUDOCS 8704290394
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NUCLEAR REGULATORY COMMISSION 5

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PORTLAND GENEPAL ELECTRIC COMPANY THE CITY OF EUGENE, OREGON PACIFIC POWER AND LIGHT COMPANY DOCKET NO. 50-344 TROJAN NilCLEAR PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.131 License No. NPF-1 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Portland General Electric Company, et al., (the licensee) dated January 30, 1985 as revised and siipiiTemented September 23,1985, July 3, August 22, Octo'ser 17, 1986 and March 23, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied, i

l 8704290394 870422 PDR ADOCK 03000344 P

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Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPFel is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.131, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifica-tions, except where otherwise stated in specific license conditions.

3.

This license amendment is effective 30 days after its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION eorge

. Knighto, Director Projec Directorate V Division of Reactor Projects-III/IV/V and Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance:

April 22, 1987 i

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ATTACHMENT TO LICENSE AMENDMENT N0.131 TO FACILITY OPERATING LICENSE NO. MPF-1 DOCKET NO. 50-344 Revise Appendix A as follows:

Remove Pages Insert Pages TS IV TS IV 2-4 2-4 B 2-3 B 2-3 8 2-4 8 2-4 i

B 2-5 B 2-5 R 2-6 B 2-6 8 2-7 B 2-7 B 3/4 0-1 B 3/4 0-1 B 3/4 3-1 B 3/4 3-1 3/4 0-1 3/4 0-1 3/4 3-1 thru 3/4 3-32 3/4 3-1 thru 3/4 3-32

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g LIMITING CONDITIONS FOR OPERA' TION AND SURVEILLANCE REOUIREMENTS l

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SECTION Page 3/4.2 POWER OISTRIBUTION LIMITS F. ~

s n-3/4.2.l Axial Flux Dif ference...............'$..

3/4 2-1 3/4.2.2 Neat Flux Hot Channel Factor............... 3/4 2-5 3/4.2.3 RCS Flowrate and Fg.... f................ 3/4 2-8 3/4.2.4 Quadrant Power Tilt Ratio................. 3/4 2-10 3/4 2-12 3/4.2.5 DNB Parameters.....

3/4.3 INSTRUMENTATION g'

3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION............ 3/4 3-1 3/4.3.2 ENGINEEREDSAFETYFEATUREACTUATIOkSYSTEM INSTRUMENTATION..................... 3/4 3-13 l 3/4.3.3 MONITORING INSTRilMENTATION

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Radiation Monitoring Instrumentation........... 3/4 3-33 Movable Incore Detectors................. 3/4 3-37 Seismic Instrumentation........

......... 3/4 3-38 Meteorological Instrumentation.............. 3/4 3-41 1

Remote Shutdown Instrumentation.............. 3/4 3-44

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Chlorine Detect 1on Systems.........

...... 3/4 3-47

-Fire Cetection Instrumentation.............. 3/4 3-48 Decouple SJitches..................... 3/4 3-50

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Accident Monitoring Instrumentation............ 3/4 3-51 Radioactive liquid Effluent Instrumentation........ 3/4 3-54 1

Radioactive.6aseous and Process Effluent Monitoring; Instrumentation.............,... 3/4 3-59 S0 Detection Systems................... 3/4 3-68 2

3/4.4 REACTOR COOLANT SYSTEM j

.3/4.4.1 REACTOR COOLANT LOOPS................... 3/44-1 TROJAN-UNIT 1 IV Amendment No. 48, 55, 78, 99, )W,131 l

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i SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i

2.2 LIMITING SAFETY SYSTEM SETTINGS 1

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REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS Q_

2.2.1

The reactor trip system instrumentation setpoints shal-1 b' set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

ACTION:

With a reactor trip system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION state-ment requirement of Specification 3.3.1 until the channel is restored l

to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint Value.

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TROJAN-UNIT 1 2-4 Amendment No.131

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2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS

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The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip j

Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Opera-tion with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that each Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

Manual Reactor Trio The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

r Power Range. Neutron Flux The Power Range, Neutron Flux channel high setpoint provides reactor i

core protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry. The low set point provides redundant protection in the power range for a power excursion beginning from low power. The trip associated with the low setpoint may be manually bypassed when P-10 is active (two of'the four power range channels indicate a power level of above approximately 10 percent of RATED THERMAL POWER) and is automatically reinstated when P-10 becomes inactive (three of the four channels indicate a power level below approximately 10 percent of RATED THERMAL POWER).

Power Range. Neutron Flux. Mich Positive Rate l

The Power Range Positive Rate trip provides added protection against l

rapid flux increases which are characteristic of rod ejection events f rom any power level. Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection f rom partial power. No credit was taken for operation of this i

trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

Power Range. Neutron Flux. Hiah Negative Rate The Power Range Negative Rate trip provides protection to ensure that the minimum DNBR is maintained above the design DNBR value for. multiple i

control rod drop accidents. The analysis of a single control rod drop

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accident indicates a return to full power may be initiated by the l

TROJAN-UNIT 1 B 2-3 Amendment No.131

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BASES automaticcontrolsysteminresponsetoacontinuedfullpoweh'turbineload i

demand or by the negative moderator temperature feedback. Tttis transient will not result in a calculated DNBR of less than the design DNBR value, therefore single rod drop protection is not required.

Intermediate and Source Rance. Nuclear Flux i

The Intermediate and Source Range, Nuclear Flux trips can provide reactor core protection during reactor startup. These trips provide added protection to the low setpoint trip of the Power Range, Neutron Flux channels The Source Range Channels will initiate a reactor trip at about 10 hounts per second unless manually blocked when P-6 becomes 4

active. The Intermediate Range Channels will initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active. No credit was taken for operation of the trips associated with either the e

4 Intermediate or Source Range Channels in the accident analyses. However, their functional capability including the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System and to provide indication of core neutron flux to t

the operators during shutdown and startup conditions.

Overtemperature aT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays f rom the core to the temperature detectors (about 4 seconds), and pressure is within the range between the High and Low Pressure reactor trips. This setpoint includes corrections for changes in density and heat capacity of water with temperature and dynamic com-pensation for piping delays from the core to the loop ' temperature detectors. With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1.

If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.

Operation with a reactor coolant loop out of service below the 4 loop P-8 setpoint does not require reactor protection system setpoint modification because the P-B setpoint and associated trip will prevent DNB during 3 loop operation exclusive of the Overtemperature AT setpoint.

Three loop operation above the 4 loop P-8 setpoint is permissi5le af ter resetting the K1, K2 and K3 inputs to the Overtemperature AT channels and raising the P-8 setpoint to its 3 loop value.

In this mode of operation, the P-8 interlock and trip functions as a High Neutr'on Flux trip at the reduced power level.

I TROJAN-UNIT 1 B 2-4 Amendment No. #8,131

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LIMITING SAFETY SYSTEM SETTINGS BASES 2-Overpower aT

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The Overpowcr AT reactor trip provides assurance of fuel integrity, e.g., no melting, under all possible overpower conditions, limits the required range for Overtemperature AT protection, and provides a backup to the High Neutron Flux trip. The setpoint includes corrections for I

changes in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop tempera-ture detectors. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

Pressurizer Pressure The Pressurizer High and Low Pressure trips are provided to limit e

the pressure range in which reactor operation is permitted. The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig). The Low Pressure trip provides protection when above the P-7 interlock setpoint.by tripping the reactor l

in the event of a loss of reactor coolant pressure.

Pressurizer Water Level The Pressurizer High Water Level trip provides added protection against Reactor Coolant System overpressurization when above the P-7 interlock setpoint by limiting the water level to a volume sufficient to retain a steam bubble and preventing water relief through the pressurizer safety valves. No credit was taken for operation of this trip in the accident analyses; however, its functional capability, at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

Loss of Flow The loss of Flow trips provide core protection to prevent DNB in the event of a loss of one or more reactor coolant pumps.

Above approximately 10% (P-7) but below approximately 39% (P-8) of l

RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drops below 90% of nominal full loop flow. Above approxi-mately 39% (P-B) of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any single loop drops below 90% of nominal full l

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l TROJAN-UNIT 1 B 2-5 Amendment No. 48,131

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...m LIMITING SAFETY SYSTEM SETTINGS BASES loop flow. Thislattertripwillpreventtheminimumvalue[fthe DNBR from going below 1.73 during normal operational transients and anticipated transients when 3 loops are in operation and the Over-temperature AT trip setpoint is adjusted to the value specified for all loops in operation. With the Overtemperature AT trip setpoint adjusted to the value specified for 3 loop operation, the P-8 trip at 75% RATED THERMAL POWER will prevent the minimum value of the DN8R from going below 1.73 during normal operational transients and anticipated transients with 3 loops.in operation.

Steam Generator Water Level The Steam Generator Water Level Low-Low trip provides core protec-tion by preventing operation with the steam generator water level.below-the minimum volume required for adequate heat removal capacity. The specified setpoint provides allowance that there will be sufficient water e

inventory in the steam generators at the time of trip to allow for starting delays of the auxiliary feedwater system.

Steam /Feedwater Flow Mismatch and Low Steam Generator Water level The Steam /Feedwater Flow Mismatch in coincidence with a Steam Generator Low Water Level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capa-

. bility of the specified trip settings and thereby enhance the overall reliability of the Reactor Protection System. This trip provides added

. protection to the Steam Generator Water Level Low-Low trip. The Steam /

Feedwater Flow Mismatch portion of this trip gs activated when the steam flow exceeds the feedwater flow by >_1.51 x 10 lbs/ hour. The Steam l

Generator Low Water level portion of the trip is activated when the water -

level. drops below 25 percent, as indicated by the narrow range instrument.

These trip values include sufficient allowance in excess of normal operating values to preclude spurious trips but will initiate a reactor trip before the steam generators are dry. Therefore, the required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient on the Reactor Coolant System and steam generators is minimized.

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Undervoltaae and Underfreauency - Reactor Coolant PumD Busses The Undervoltage and Underfrequency Reactor Coolant Pump bus trips provide reactor core protection when above the P-7 interlock setpoint againstDNBasaresultoflossofvoltage(nominally 12.47kW)or

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TRO,1AN-UNIT 1 8 2-6 Amendment No. #5,131 l

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underfrequency (nominally 60 Hz) to more than one reactor coolant pump.-

l The specified setpoints assure a reactor trip signal is generfted before the low flow trip setpoint is reached. A 0.1 second time deley in the underfrequency trip and a 0.2 second time delay in the under~ voltage i

trip are incorporated to prevent spurious reactor trips from momentary electrical power transients.

Turbine Trio A Turbine Trip causes a direct reactor trip when operating above the P-7 interlock setpoint. Each of the turbine trips provide turbine l.

protection and reduce the severity of the ensuing transient. No' credit was taken in the accident analyses for operation of these trips. Their functional capability at the specified trip settings is required to enhance the overall reliability of the Reactor Protection Systen.

Auto Safety In.iection Input l

If a reactor trip has not already been generated by the reactor protective instrumentation, the ESF automatic actuation logic channels will initiate a reactor trip upon any signal which initiates a' safety injection. This trip is provided to protect the core in the event of a LOCA. The ESF instrumentation channels which initiate a safety injection signal are included in Table 3.3-3.

l Reactor Coolant PumD Breaker Position Trio The Reactor Coolant Pump Breaker Position Trip is an anticipatory trip which provides additional reactor core protection against DNS resulting from the opening of two or more pump breakers when above the P-7 interlock

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setpoint. The open/close position trip assures a reactor trip signal is generated before the low flow trip set point is reached. No credit was taken in the accident analyses for operation of this t' rip. The functional l

capability at the open/close position settings is required to enhance the overall reliability of the Reactor Protection System.

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TRO.1AN-UNIT 1 8 2-7 Amendment No.131

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3/4.0 APPLICABILITY SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS The specifications of this section provide the general rIquirements applicable to each of the Limiting Conditions for Operation an:d Surveil-lance Requirements within Section 3/4.

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3.0.1 This specification defines the applicability of each specification ~

in terms of defined OPERATIONAL MODES or other specified conditions and is provided to delineate specifically when each specification is applicable.

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3.0.2 This specification defines those conditions necessary to constitute compliance with the terms of an individual Limiting Condition for Operation and associated ACTION requirement.

3.0.3 This. specification delineates the ACTION to be.taken for circumstances not directly provided for in the ACTION statements and whose occurrence would violate the intent of the specification. For example, Specification 3.5.1 calls for each Reactor Coolant System accumulator to be OPERABLE and provides explicit ACTION requirements when one accumulator is inoperable. Under the terms of Specification 3.0.3, if more than one accumulator is inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months measures must be initiated to place the unit in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . As a further example, Specification 3.6.2.1 requires two Containment Spray Systems to be OPERABLE and provides explicit ACTION requirements if one spray system is inoperable: Under the terms of Specification 3.0.3, if both of the. required Containment Spray Systems are inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months measures must be initiated to place the unit in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in at least COLD SHUTDOWN in the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

It is assumed that the unit is brought to the required MODE within the required times by promptly initiating and carrying out the appropriate ACTION statement.

3.0.4 This specification provides thattentry into an OPERABLE MODE or other specified applicability condition must be made with (a) the full complement of required systems, equipment or components OPERABLE and (b) all other parameters as specified in the Limiting Conditions for Operation being met without regard for allowable deviations and out of service provisions j

contained in the ACTION statements.

The intent of this provision is to insure that facility operation is not initiated with either required equipment or systems inoperable or other specified limits being exceeded.

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TROJAN-UNIT 1 B 3/4 0-1 Amendment No. (1, 131 I

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f 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 PR'OTECTIVE AND ENGINEERED SAFETY FEATURES (ESF)

INSTRUMENTATION The OPERABILITY of the protective and ESF instrumentation systems and interlocks ensure that 1) the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combina-tion thereof exceeds its setpoint, 2) the specified coincidence logic is maintained, 3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance consistent with maintaining an appropriate level of reliability of the Reactor Protection and ESF instrumentation.

The OPERABILITY of these systems is required to provide the overal,1. _

reliability, redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The. integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.

The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards.

The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protec-tion Instrumentation Systam", and supplements to that report as approved by the NRC and documented in the SER (letter to J. J. Sheppard from Cecil 0. Thomas dated February 21,1985).

The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, over-lapping or total channel test measurements provided that such tests demonstrate the total channel response time as definede Sensor response time verfication may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times.

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that

1) the radiation levels are continually measured in the areas served

. TROJAN-UNIT 1 B.3/4 3-1 Amendment No. 131

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I-3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION h"-

3.0.l* Limiting Conditions for Operation and ACTION requirements shall be applicable during the OPERATIONAL MODES or other conditions specified for each specification.

3.0.2 Adherence to the requirements of the Limiting Condition for Operation and/or associated ACTION within the specified time interval shall constitute compliance with the specification.

In the event the Limiting Condition for Operation is restored prior to expiration of the specified time interval, completion of the ACTION statement is not required.

3.0.3 In the event a Limiting Condition for Operation and/or associated ACTION requirements cannot be satisfied because of circumstances in excess of those addressed in the specification, within one hour action shall be initiated to place the unit in a MODE in which the Specification does not apply by placing it, as applicable, in:

1.

At least HOT STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , 2.

At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and 3.

At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual Specifications.

This Specification is not applicable in MODES 5 or 6.

3.0.4 Entry into an OPERATIONAL MODE or other specified applicability condition shall not be made unless the conditions of the Limiting Condition

.for Operation are met without reliance on provisions contained in the ACTION statements unless otherwise excepted. This provision shall not

. prevent passage through OPEFATIONAL MODES as required to comply with ACTION statements.

.3.0.5 When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its nornel 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; an'd (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 at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in at least COLD SHUTDOWN i

withi.n the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . This specification is not applicable in MODES S or 6.

l TROJAN-UNIT 1 3/40-1 Amendment No. f2,131 o

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i 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION N'

3.3.1

As a minimum, the reactor trip system instrumentation -channels l

and interlocks of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

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SURVEILLANCE RE0VIREMENTS 4.3.1.1 Each reactor trip system instrumentation channel shall be l

l demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST cperations during the modes and at the frequencie: shown in Table 4.3-1.

4.3.1.2 The logic for the interlocks shall be demonstrated OPERABLE l

during the at power CHANNEL FUNCTIONAL TEST of channels affected by interlock operation. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip l

function shall be demonstrated to be within its limit et least once per 18 months. Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels" 4

column of Table 3.3-1.

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1 TROJAN-UNIT 1 3/4 3-1 Amendment No.131

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E REACTOR TRIP SYSTEM INSTRUMENTATION E

E MINIMUM 5

TOTAL NO.

CHANNELS CHANNELS APPLICABLE

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FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION i

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Manual Reactor Trip 2

1 2

1, 2 11 2

1 2

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4*, 5*

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Power Range, Neutron Flux A.

High Setpoint 4

2 3

1 2-23 B.

' Low Setpoint 4

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Power Range, Neutron Flux 4

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High Positive Rate 4.

Pos r Range, Neutron. Flux, 4

2 3

1, 2 2f HighLNegative Rate es y

5.

IntermediateRange[NeutronFlux 2

1 2

14#I, 2 3

Y 6.

Source Range, Neutron Flux 2f 4

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Startup 2

1 2

B.

Shutdown 2

1 2

3*,

4*, 5*

10 C.

Shutdowa 2

0 1

3,4,5 5

7.

Overtemperature AT Four Loop Operation 4

2 3

1, 2 6I Three Loop Operation 4

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3 1, 2 8

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Overpower AT Four Loop Operation 4

2 3

1, 2 6I Three Loop Operation 4

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3 1, 2 8

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Pressurizer Pressure - Low 4

2 3

1*

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Pressurizer Pressure - High 4

2 3

1, 2 6f 11.

Pressurizer Water Level - High 3

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CHANNELS CHANNELS APPLICAllLE

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Loss of Flow - Single Loop 3/ loop 2/ loop in 2/ loop in 1

6I f-(Above P-8) any oper-each oper-r ating loop ating loop 13.

Loss of Flow - Two loops 3/ loop 2/ loop in 2/ loop in 1

68 (Above P-7 and below P-8) two oper-each oper-ating loops ating loop j-l f

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[

14.

Steam Generator Water 3/ loop 2/ loop in 2/ loop in 1, 2 level - Low-Low any oper-each oper-i my ating loop ating loop e

3

15. Steam /Feedwater Flow 2/ loop-level 1/ loop-level 1/ loop-level 1, 2 68 l Mismatch and Low Steam and coincident and Generator Water Level 2/ loop-flow with 2/ loop-flow mismatch-1/ loop-flow mismatch or mismatch in 2/ loop-level same loop and t

1/ loop-flow mismatch 16.

Undervoltage - Reactor Coolant 4-2/ bus 1/ bus for 3

19 68 l Pumps each bus Y(

17. Underfrequency - Reactor Coolant 4-2/ bus 1/ bus for 3

19 6I l

d 3 ' ' '

g Pumps each bus 1

i 3

9

5I;

~

TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION

.^

2:

E MINIMUM Zi TOTAL NO.

CHANNELS CHANNELS APPLICABLE

-- FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTIDN 18.

Turbine Trip A.

Low Hydraulic Control 3

2 2

19 68 l

?

Oil Pressure B.

Turbine Stop Valve Closure 4-1/ valve 4-1/ valve 4-1/ valve 19 7#

l i

19. Auto Safety Injection Input 2

1 2

1, 2 1

l l

20.

Reactor Coolant Pump Breaker 4-1/ breaker 2

1/ breaker 19 9#

I' I

Position Trip per oper-ating loop 21.

Reactor Trip Breakers 2

1 2

1, 2 1

ty 2

1 2

3*,

4*, 5*

10 s-

22. Automatic Trip Logic 2

1 2

1, 2 1

2 1

2 3*, 4*, 5*

10 4

4 D

t?e a

!an.

lc ::

4

ac.

.. ~.

1 1

TABLE 3.3-1 (Continued)

TABLE NOTATION i

l

With the reactor trip system breakers in the closed position and the control rod drive system capable of rod withdrawal.

- The channel (s) associated with the protective functions derived f rom the out of service Reactor Coolant Loop shall be placed in the trip-ped condition.

The provisions of Specification 3.0.4 are not applicable.

1. When below the P-6 setpoint.

1. 1. When below the P-10 setpoint.

9 When above the P-7 setpoint.

(1) The applicable MODES and ACTION statement for these channels noted in Table 3.3-3 are more restrictive and, therefore, applicable.

ACTION STATENENTS ACTION 1 - With the number of channels OPERABLE one less than the Mini-mum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

One channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per specification 4.3;1.1 provided the other channel is OPERABLE.

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may pro-ceed provided all of the following conditions are satisfied:

a.

The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

The Minimum Channels OPERABLE requirement is met; how-ever, the inoperable channel may be bypassed for up to i

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.1.1.

c.

Either THERMAL POWER is restricted to 175% of RATED THERMAL POWER and the Power Range Neutron Flux trip setpoint is reduced to $85% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , per Specification 4.2.4.c.

TROJAN-UNIT 1 3/4 3-5 Amendment No. 33,131

/

,-.,c.

IN >

a.

.2

. u..

TABLE 3.'3-1-(Continued)

ACTION 3 - With the number of channels OPERABLE one less than the l

Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a.

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

b.

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

c.

Above 5% of RATED THERMAL POWER, power operation may continue.

ACTION 4 - With the number of OPERABLE channels one less than the Mini-mum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.

ACTION 5 - With the number of channels OPERABLE one less than the Mini-mum Channels OPERABLE requirement, verify compliance with the-SHUTDOWN MARGIN requirements of Specification 3.1.1.1 within l

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

1 ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may pro-ceed provided both of the following conditions are satisfied:

a.

The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

The Minimum Channels OPERABLE requirement is met; how-ever, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other' channels per Specification 4.3.1.1.

ACTION 7 - With the number of OPERABLE channels less than the Total Number of Channels, place the inoperable channel in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Operation may continue until performance of thV next required CHANNEL FUNCTIONAL TEST.

ACTION 8 - With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

One channel associated with an operating loop cay be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.1.1.

TROJAN-UNIT 1 3/4 3-6 Amendment No. 23,131

Wn

~ ~

~

~~

~

^^.'

~~Q~

~~

.-.a...... -...>~u TABLE 3.3-1 (Continued)

I ACTION 9 - With less than the Minimum Number of Channels OPERABLE, l

operation may continue provided the inoparable chann(1 is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

n ACTION 10 - With the number of channels OPERABLE one less than required I

by the Minimum Channels OPERABLE requirement, restor'e the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the reactor trip breakers within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .l ACTION 11 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

REACTOR TRIP SYSTEM INTERLOCKS DESIGNATION CONDITION AND SETPOINT FUNCTION P-6 With 2 of 2 Intermediate Range Prevents or defeats Neutron Flux Channels <6x10-Il the manual block of amps.

source range reactor trip.

P-7 With 2 of 4 Power Range Neutron Prevents or defeats Flux Channels E11% of RATED the automatic block of THERMAL POWER or 1 of 2 Turbine reactor trip on: Low impulse chamber pressure flow in more than one channels 166 psia, primary coolant loop, reactor coolant pump under-voltage and under-frequency, turbine trip, pressurizer low pressure, and pressurizer high level.

P-B With 2 of 4 Power Range Neutron Prevents or defeats Flux Channels 139% of RATED the automatic block of THERMAL POWER.

reactor trip on low coolant flow in a single loop.

P-10 With 3 of 4 Power Range Neutron Prevents or defeats the Flux Channels <9% of RATED manual block of: Power THERMAL POWER.

range low setpoint reactor trip, intenpediate range reactor trip, and inter-mediate range rod stops.

Provides input to P-7.

(

P-13 With 2 of 2 Turbine Impulse Provides input to P-7.

Chamber Pressure Channels

<66 psia.

1ROJAN-UNIT 1 3/4 3-7 Amendment No. 73, is, is,131

....x..._...

t.

4

(

1 v.

c.

~

(This page intentionally blank) i l

I 1

l l

l 1

1 j

i r

TROJAN-UNIT l 3/4 3-8 Amendment No.131

it l

TABLE 3.3-2

-4 5

REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMESI l

u a

FUNCTIONAL UNIT RESPONSE TIME E

i 1.

Power Range, Neutron Flux 10.5 seconds

l 2.

Power Range, Neutron Flux, 50.5 seconds

l High Negative Rate 3.

Overtemperature AT

$4.0 seconds

l 4.

Pressurizer Pressure - Low 12.0 seconds l

S.

Pressurizer Pressure - High

$2.0 seconds l

j 6.

Loss of Flow - Single Loop

$1.0 seconds l

M (Above P-8) i x-

'?

7.

Loss of Flow - Two Loops

$1.0 seconds l

(Above P-7 and below P-8) 8.

Steam Generator Water Level - Low-Low

$2.0 seconds j

51.2 seconds l

9.

Undervoltage-Reactor Coolant Pumps

10. Underf requency-Reactor Coolant Pumps 10.6 seconds l

l l

P i

it

Neutron detectors are exempt from response time testing. Response time shall be measured from detector E.

output or input of first electronic component in channel.

l R

?,

Trips are not listed for which response time ' testing is not applicable.

[

.1,v, *..

z 1

O s

i I

4

f 1

M TABLE 4.3-1 a

E REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5

?

E CHANNEL MODES IN WHICH q

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK

_ CALIBRATION TEST REQUIRED 1.

Manual Reactor Trip N.A.

N.A.

R 1, 2, 3*,

4*, 5*

l 2.

Power Range, Neutron Flux A.

High Setpoint S

D(2,4), M(3,4)

Q(11) 1, 2 Q(4,6) e 1 ##, 2 I

B.

Low Setpoint S

R(4)

S/U(1) t.

3.

Power Range, Neutron Flux, N.A.

R(4)

Q(11) 1, 2 l

,f; High Positive Rate

[:;.

4.

Power Range, Neutron Flux, N.A.

R(4)

Q(11) 1, 2 i-High Negative Rate w

5.

Intermediate Range, S

R(4)#

S/U(1) 1###, 2 I-w Neutron Flux

..o i.

6.

Source Range, Neutron Flux S

R(4,5)

S/U(1), Q(11)# 2##, 3, 4, 5(9) l i_

7.

Overtemperature AT S

R Q(11) 1, 2 l

8.

Overpower AT S

R Q(11) 1, 2 l

h

~

9.

Pressurizer Pressure - Low S

R Q(11) 19 l

10.

Pressurizer Pressure - High S

R Q(11) 1, 2 l

{

Pressurizer' Water Level - High S

R Q(11) 19 l

l f 11.

~

12.

Loss of Flow - Single Loop S

.R Q(11) 1 l

o 4

,8 (Above P-8)

U N

I

.. W

.g e

TABLE 4.3-1 (Continued)

I

{

REACTOR TRIP SYSTEN INSTRUNENTATION SURVEILLANCE REQUIRENENTS z

Cs CHANNEL NODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SdRVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED l'

13.

Loss of Flow - Two Loops S

R N.A.

1 (Above P-7 and below P-8) l r.

11

14. Steam Generator Water Level -

S R

Q(11)(12) 1, 2 l

Low-Low

15. Steam /Feedwater Flow Nismatch S

R Q(11) 1, 2 l

and Low Steam Generator Water Level lr

16. Undervoltage - Reactor Coolant N.A.

R Q(11) 19 l

m)

Pumps w

17.

Underfrequency.- Reactor Coolant N.A.

R Q(11) 19 l

Pumps

18. Turbine Trip A.

Low Hydraulic Control 011 N.A R

S/U(1,8) 19 l

j Pressure B.

Turbine Stop Valve Closure N.A.

N.A.

S/U(1,8) 19 l

l k

19. Auto Safety injection Input N.A.

N.A.

R 1, 2 l

E

?,

20. Reactor Coolant Pump Breaker N.A.

N.A.

R 19

,,,3,,,,,

l Pbsitidn* Trip z

?

21.

Reactor Trip Breaker N.A.

N.A.

N(7. 10) 1,2,3*,4*,5*l O

22.

Automatic Trip Logic N.A.

N.A.

N(7) 1,2,3*,4*,5*l 5

W :i.s.

'. w :.

TABLE 4.3-1 (Continued)

NOTATION With the reactor trip system breakers closed and the cont 1 ' rod drive system capable of rod withdrawal.

~

The provisions of Specification 4.0.4 are not applicable.

When below the P-6 setpoint.

When below the P-10 setpoint.

When above the P-7 setpoint.

(1) - If not performed in previous 31 days.

l (2) - Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel if absolute difference is >2%.

(3) - Compare incore to excore axial imbalance above 15% of RATED THERMAL POWER. Recalibrate if absolute difference >3 percent.

e (4) - Neutron detectors may be excluded from CHANNEL CALIBRATION.

l (5) - Detector plateau curves shall be obtained and evaluated.

l (6) - Incore-Excore Calibration.

l (7) - Each train tested every other month.

l (8) - Setpoint verification is not applicable.

l (9) - See Specification 3/4.9.2 for audio and visual requirements in l

MODE 6.

(10) - The reactor trip breakers shall be tested using the Automatic Trip Logic trip signal.

(11) - Each channel shall be tested at least every 92 days.on a STAGGERED TEST BASIS.

(12) - The surveillance frequency and/or MODES specified for these channels in Table 4.3-2 are more restrictive and, therefore, applicable.

TROJAN-UNIT 1 3/4 3-12 Amendment No. 131

..~.

. : 3.

n.

. a-.......

..-...a... - *

-.n y

w..,

a..

l INSTRUMENTATION f

3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION I- -

3.3.2. The Engineered Safety Feature Actuation System (ESFAS) instru-l

~

nentatitn channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5.

APPLICABILITY: As shown in Table 3.3-3.

ACTION:

a.

With an ESFAS instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With an ESFAS instrumentation channel inoperable, take the action shown in Table 3.3-3.

SURVEILLANCE REOUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel shall be demonstrated l

OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-2.

~

l 4.3.2.2 The logic for the interlocks shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by interlock operation. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.2.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per l

l 18 months.

Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" column of Table 3.3-3.

l l

i l

TROJAN-UNIT 1 3/4 3-13 Amendment No.131

.-.y y

m.

,.,m-..

--, _.y-..,sy--,,._.

3

-w,...

..---,,,,,y

,,.w=

v-y--,, -, -,..,.

l N

j TABLE 3.3-3 d

E ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATIDW E

C MINIMUM d

TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION f

~

1.

SAFETY INJECTIONT

~

a.

Actuation Logic 2

1 2

1,2,3,4 13 l

b.

Manual Initiation 2

1 2

1,2,3,4 18 l

Pressure - High l

f, c.

Containment 3

2 2

1,2,3,4 14*

p.

3 d.

Pressurizer 3

2 2

1, 2, 3#

14*

}<

Pressure - Low w

i

~

e.

Differential 1, 2, 3##

I Pressure Between Steam Lines - High Four Loops 3/ steam line~

2/ steam line 2/ steam line 14*

l' operating any steam line

-l IIII steam 2/ operating 15 Three Loops 3/ operating

/

Operating steam line line, any steam line operating E

steam line Y Also. initiates: Reactor Trip, Emergency Diesel Start, Auxiliary Feedwater, Turbine Trip, Fee &Allr Isolation, Containment Isolation, Containment Ventilation Isolation,' Control Room Isolation, Containment Cooling Fans, and f

Essential Service Water.

C 0

T

~

. f,,

e y

TABLE 3.3-3 (Continued) i-E g

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION E

~

i MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE

,i m

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION f.

Steam Flow in Two 1, 2, 34I Steam Lines - High

t Four Loops 2/ steam line 1/ steam line 1/ steam line 14*

,~ !

Operating any 2 steam lines t-

d. l -

Three Loops 2/ operating 1888/any 1/ operating 15 Operating steam line operating steam line w

steam line

i,

(

ra COINCIDENT WITH-EITHER Tavg - L N-L W 1, 2, 24I i

Four Loops 1 Tavg/ loop 2 Tavg any 1 Tavg any 14*

ii Operating loops 3 loops

! lI Three Loops 1 Tav /

1888 Tavs in I Tavg in any 15 Operating opera!!ng any operating two operating loop loop loops i

i;

.i,v.

t s

a

7

~

. pF,. -

TABLE 3.3-3 (Continued)

=

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMEN1ATION c5 MININUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE

~

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION i

DR, COINCIDENT WITH

~

Steam Line 1, 2, 38I Pressure - Low Four Loops 1 pressure /

2 pressures 1 pressure 14*

l Operating loop any loops any 3 loops Three Loops 1 pressure /

18## pressure I pressure 15 Operating operating in any oper-in any 2 loop ating loop operating loops w

l w

2.

CONTAINMENT SPRAY I

a.

Actuation Logic 2

1 2

1,2,3,4 13 l

4 b.

Manual 2

2 2

1,2,3,4 18 l

4 2

3 1,2,3,4 16 l

[

c.

Containment Pressure -

High-High 3.

CONTAINMENT ISOLATION a.

Containment Isolation Signal

[

l 1)

. Manual 2

1 2

1,' 2 ',' 3',' 4 '

18 A

2)

From Safety Injection See 1 above for initiating functions and requirements.

l if

e

- #P-

_4 TABLE 3.3-3 (Continued) 5 g

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION E[4 MININUM TOTAL NO.

CHANNELS CHANNELS APPLJCABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE N00ES ACTIDN b.

Containment Ventilation Isolation 1)

From Manual Containment See 3.a.1 above for requirements.

Isolation 2)

From Manual Containment See 2.b above for requirements.

Spray t

t 3)

From Safety Injection See 1 above for initiating functions and requirements.

l s-s

+

'T 4)

Containment

(

3 Radioactivity - High Particulate 1

1 1

1,2,3,4 17 lodine 1

1 1

High Level Noble Gas 1

1 1

g Low Level Noble Gas 1

1 1

4 4.

STEAM LINE ISOLATION a.

Actuation Logic 2

1 2

1,2,3 19 l

b.

Manual 1/ steam line 1/ steam line 1/ operating 1, 2, 3 20 l

g.

steam line et R

c.

Containment Pressure -

4 2

3 1,'2;!$

16 2

Nigh-Algh 3

i

9

.g 1

TABLE 3.3-3 (Continued _1

=8 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUNENTATION R

c5 MININUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE

+

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION d.

Steam Flow in Two Steam 1, 2, 3 l

Lines - High t

l Four Loops 2/ steam line 1/ steam line 1/ steam line 14*

Operating any 2 lines III any 1/ operating 15 Three Loops 2/ operating I

/

Operating steam line operating steam line steam line l!

1 COINCIDENT WITH EITHER w

j Tavg - Low-Low 1, 2, 3 l

w

!(

y Four Loops 1 Tavg/ loop 1 Tavg any 1 Tavg in any 14*

j

. Operating 2 loops 3 loops j

i-Three Loops 1 Tavg/oper-IIII Tavg in I Tavg in any 15 ii; Operating ating loop any operating two operating loop loops i

OR. COINCIDENT WITH

)

Steam Line Pressure - Low 1, 2, 3 l

I E

I Four Loops 1 pressure /

1 pressure 1 pressure dM" 14*

1

[

Operating loop any 2 loops any 3 loops s

Three Loops 1 pressure /-

IIII pressure I pressure in 15 E

Operating operating loop in any oper-any 2 oper-ating loop, ating loops El c

4 i

l

4

.kg t

t Y

TABLE 3.3-3 (Continued)

P.

E ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION E

MINIMUM TOTAL NO.

CHANNELS CHANNELS APPL.ICABLE i

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION i

5.

TUR8INE TRIP AND FEEDWATER ISOLATION a.

From Safety injection See 1 above for initiating functions and requirements.

l b.

Steam Generator Water Level -

3/ loop 2/ loop in any 2/ loop in each 1, 2 14*

l 1

High-High operating loop operating loop

., i'

,s t.

6.

AUXILIARY FEEDWATER PUMPS START a.

Manual Initiation (Control 2/ pump 1/ pump 2/ pump 1,2,3 21 l

tl s

Room and Panel C-160) s J

so b.

From Safety Injection See 1 above for initiating functions and requirements.

l

?

c.

Steam Generator Water Level -

3/ steam 2/any steam 2/ steam 1, 2, 3 14*

l Low-Low generator

. generator generator i

d.

Loss of Normal and Preferred 2/ bus 1/ bus 1/ bus 1, 2, 3 18 l

Power 4

O

.i,v,.

a I

n

.h 1

b L.

i l

i

~

....T'bL...'. :: :. '.~ ~ O.:

~

a.

.~

..........c.;

TABLE 3.3-3 (Continued)

TABLE NOTATION

The provisions of Specification 3.0.4 are not applicable. 'g.

Wh,e'n above the P-11 setpoint.

When above the P-12 setpoint.

1. The channel (s) associated with the protective functions derived f rom the out of service Reactor Coolant Loop shall be placed in the tripped mode.

ACTION STATEMENTS ACTION 13 - With the number of OPERABLE Channels one less that the' Total' Number of Channels, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . One channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLL.

ACTION 14 - With the number of OPERABLE Channels one less than the Total Number of Channels, operation may continue until performance of the'next required CHANNEL FUNCTIONAL TEST, provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 15 - With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in HOT SHUTDOWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; however, one channel associated with an operating loop may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1.

ACTION 16 - With the number of OPERABLE Channels one less than the otal Number of Channels, operation may continue with the i

inoperable channel bypassed, provided that the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing j

per Specification 4.3.2.1.

i ACTION 17 - With less than the Minimum Channels OPERABLE, operation may continue provided the containment ventilation valves are maintained closed.

ACTION 18 - With the number of OPERABLE Channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Trojan-Unit 1 3/4 3-20 Amendment No, 131

n

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~ ~,u..-. -,:... A.

^

~;

TABLE 3.3-3 (Continued)

ACTION STATEMENTS ACTION 19 - With the number of OPERABLE Channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT

' SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . One channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specifi-cation 4.3.2.1 provided the other channel is OPERABLE.

ACTION 20 - With the number of OPERABLE Channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the associated valve inoperable and take the ACTION required by specification 3.7.1.6.

ACTION 21 - With the number of OPERABLE Channels one less than the Total Number of Channels, declare the associated auxiliary feedwater pump inoperable and apply the requirements of Specification 3.7.1.2.

ENGINEERED SAFETY FEATURES INTERLOCKS DESIGNATION CONDITION AND SETPOINT FUNCTION P-11 With 2 of 3 pressurizer Prevents or defeats pressure channels > 1925 manual block of safety psig.

injection actuation on low pressurizer pressure.

P-12 With 2 of 4 Tavg channels Allows manual block of l

< 553*F safety injection actua-tion on high steam line flow and. low steam line pressure.

Causes steam line isolation on high steam flow. Affects steam dump blocks.

P-4 With both reactor trip Allows manual block of breakers open.

safety injection actua-tion. Causes turbine' trip. Closes feedwater valves on low Tavg and prevents their reopening if closed by safety injection or high steam generator water level.

P-14 With 2 of 3 steam generator Causes turbine trip and level channels >75% of narrow trips both.feedwater range instrument span, pumps. Closes feedwater isolation, bypass, and regulating valves.

~

am

y

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Yg_

IABLE 3.3-4 E

EhGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS*

l e

'i

~

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

~

1.

SAFETY INJECTION

,)

a.

Containment Pressure - High S 4 psig 5 4.5 psig b.

Pressurizer Pressure - Low 1 1765 psig 2 1755 psig l

c.

Differential Pressure Between Steam 5 100 psi 5 112 psi l

4, Lines - High

(.(

d.

Steam Flow in Two Steam L1nes - High i A function defined as 1 A function defined as

-[

Coincident with Tavg - Low-t w follows: 40% of full steam follows: 44% of full steam

,i w

)

or Steam Line Pressure - Low flow between 0% and 20% load flow between 0% and 20% load and then increasing linearly and then increasing linearly I

w to 110% of full steam flow to 111.5% of full steam flow at full load at full load s;

Tavo 1 553*F Tava 1 551*F i

t 600 psig steam line 1 580 psig steam line i

pressure pressure 2.

CONTAINMENT SPRAY l

{;

k a.

Containment Pressure - High-High 5 30 psig 5 32 psig E.

8 3.

CONTAINMENT ISOLATION a

z a.

Containment Isolation Signal

.i,v,*

1.

From Safety Injection See.1 above for applicable data.

er

The Actuation Logic and Manual Initiation Circuitry does not have trip setpoints or allowable values and is therefore not included in this table.

e w

,--m

-A

l TABLE 3.3-4 (Continued) y i

5 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS g

J E

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES b.

Containment Ventilation Isolation 1.

From Safety Injection See 1 above for applicable data.

l t

2.

Containment Radioactivity - High 5 2 x background 5 2 x background t

4.

STEAM LINE ISOLATION a.

Containment Pressure - High-High S 30 psig 5 32 psig g

4.

i A function defined as44%offul b.

Steam Flow in Two Steam Lines - High

$ A function defined as Coincident with Tavg - Low or Steam follows: 40% of full steam follows:

I w)

Line Pressure - Low flow between 0% and 20% load flow between 0% and 20% load i

and then increasing linearly and then increasing linearly

i. :

w4 to 110% of full steam flow to 111.5% of full steam flow t'

w at full load at full load

[

l Tava 2 553*F Tave t 551*F 2 600 psig steam t 580 psig steam line pressure line pressure 5.

TUR8INE TRIP AND FEEDWATER ISOLATION a.

From Safety Injection See 1 above for applicable data.

l j

F b.

Steam Generator Water Level -

1 75% of narrow range 1 76% of narrow range l

y High-High instrument span each steam instrument span each steam generator generator ' M' "

j g

so 1

E O

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.(

.. y TABLE 3.3-4 (Continued)

=

0 g

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E

FUNCTIONAL UNIT TRIP SETPOINT ALLOWf,BLE_ VALUES 6.

AUXILIARY FEEDWATER PUMPS START a.

From Safety Injection See 1 above for applicable data.

~

b.

Steam Generator Water Level -

1 5%'of narrow range 1 3% of narrow range l

Low-Low instrument span each instrument span each steam generator steam generator c.

Loss of Normal and Preferred Power 1 2520 volts 1 2478 volts i

e a

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w C

1 t

W 'L..

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x q, TABLE 3.3-5 l

ENGINEERED SAFETY FEATURES RESPONSE TIMEST INITIATING SIGNAL AND FUNCTION RESPONSE TIME IIf SECONOS 1.

Manual Response Time testing is not applicable to manual initiation circuitry.

2.

Containment Pressure - High a.

Safety Injection (ECCS) 5 27.0*

b.

Reactor Trip (from SI)

$ 3.0 c.

Feedwater Isolation S 8.0 d.

Containment Isolstion Signal i 18.0#/28.00##

l e.

Service Water System

$ 13.0#/48.'0#

l f.

Emergency Fan Coolers 5 10.0#/49.0 #

.l 3.

Pressurifer Pressure - low a.

Safety Injection (ECCS)

$ 13.0#/27.0*

l b.

Reactor Trip (from SI)

$ 3.0 c.

Feedwater Isolation 5 8.0 d.

Containment Isolation signal 5 18.0#

l e.

Service Water System 5 13.0#/48.0*

l f.

Emergency Fan Coolers 1 10.0#/49.0##

l 4.

Differential Pressure 8etween Steam lines - High a.

Safety Injection (ECCS)

$ 13.0#/23.0##

b.

Reactor Trip (f rom SI)

$ 3.0 c.

Feedwater Isolation 1 8.0 Trojan-Unit 1 3/4.3-25 Amendment No. U,131 i

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TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMEST T. -

INITIATING SIGNAL AND FUNCTION RESPONSETIMEISSECetOS d.

Containment Isolation Signal 5 18.0#/28.0#

l e.

Service Water System 5 13.0#/48.0 #

l f.

Emergency Fan Coolers 5 10.0#/49.0##

l S.

Steam Flow in Two Steam Lines - Hiah Coincident with Tavo - Low-Low a.

Safety Injection (ECCS)

$ 15.0#/25.0#

b.

Reactor Trip (from SI)

$ 5.0 c.

Feedwater Isciation 5 10.0 d.

Containment Isolation Signal

$ 20.0#/30.0#

l e.

Service Water System 5 15.0#/50.0 #

l f.

Steam Line Isolation 5 10.0 l

g.

Emergency Fan Coolers 5 12.0#/51.0 #

l 6.

Steam Flow it.'Two Steam Lines - Hiah Coincident

~'

with Steam Line Pressure - Low a.

Safety Injection (ECCS) 5 13.0#/23.0##

b.

Reactor Trip (from SI)

$ 3.0 c.

Feedwater Isolation 5 8.0 d.

Containment Isolation Signal 1 18.0#/28.0#

l e.

Service Water System

$ 14.0#/.48.0#

l f.

Steam Line Isolation 5 8.0 l

g.

Emergency Fan Coolers 1 10.0#/49.0 #

l Trojan-Unit 1 3/4 3-26 Amendment No.131

p,.. v.

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.. ~ ~

...... ~..

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMEST

7..

RESPONSETIMEIkSECONDS INITIATING SIGNAL AND FUNCTION 7.

Containment Pressure - Mich-High a.

Containment Spray 5 30.0 r

b.

Steam Line Isolation

$ 7.0 8.

Steam Generator Water level - Hich-High a.

Turbine Trip - Reactor Trip

$ 2.5 b.

Feedwater Isolation 5 11.0 9.

Steam Generator Water level - Low-Low a.

Auxiliary Feedwater Pumps

$ 60.0 Functions are not listed for which Response Time testing is Not Applicable.

Diesel generator starting and sequence loading delays included.

Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal. charging pumps, SI and RHR pumps.

Diesel generator starting and sequence loading delays ngi included.

Offsite power available. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps,

1. Diesel generator starting and sequence loading delays included.

Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.

Trojan-Unit 1 3/4 3-27 Amendment No. 55,131

,g

. *d q p

TABLE 4.3-2

': 4

!E ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION lE SURVEILLANCE REQUIREMENTS

!E i z_

'M

' CHANNEL MODES IN hMICH

~

CHANNEL CHANNEL FUNCTIONAL SURVE'ILLANCE FUNCTIONAL UNIT CHECK CAlTBRATION TEST (2)

REQUIRED l

i 1.

SAFETY INJECTION a.

Actuation Logic N.A.

N.A.

M(1) 1, 2, 3, 4 l

b.

Manual Initiation N.A.

N.A.

R 1,2,3,4 l

c.

Containment Pressure - High S

R M

1,2,3,4 l

i d.

Pressurizer Pressure - Low S

R M

1, 2, 3#

l i

T e.

Differential Pressure S

R M

1,2,3fi i;

,E S$

Between Steam Lines - High f.

Steam Flow in Two Steam S

R M

1, 2, 3 #

l Lines - High Coincident with Tavg - Low-Low or Steam Line Pressure - Low 7

r' l

I 2.

CONTAINMENT SPRAY a.

Actuation Logic N.A.

N.A.

M(1) 1,2,3,4 l

t b.

Manual Initiation N.A.

N.A.

R 1,2,3,4 l

c.

Containment Pressure - High-High S

R M

1,2,3,4 l

E

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TABLE 4.3-2 (Continued) y Eg ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUNENTATION 2:

SURVEILLANCE REQUIREMENTS E

Q T

CHANNEL

MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE 2,

FUNCTIONAL UNIT CHECK CAllBRATION TEST REQUIRED l

3.

CONTAINMENT ISOLATION

i a.

Containment Isolation Signal l

1) Manual N.A.

N.A.

R 1,2,3,4 b.

Containment Ventilation Isolation

[

1) From Manual Containment Isolation See 3.a above for requirements.

l m

4-g

2) From Manual Containment Spray See 2.b above.for requirements.

l

[

m tO

3) Containment Radioactivity - High 1,2,3,4 l

Particulate S

R M

Iodine S

R M

High Level Noble Gas S

R M

Low Level Noble Gas S

R M

4 4.

STEAM LINE ISOLATION a.

Actuation Logic N.A.

N.A.

M(1) 1, 2, 3 l

a

[

.b.

Manual N.A.

N.A.

R if

,'3 l

a c.

Containment Pressure - High-High S

R M

1,2,3 l

[

d.

Steam Flow in Two Steam Lines -

See 1.f above for requirements.

l High Coincident with Tavg -

Low.or Steam Line Pressure - Low U

,g

. e.9 TABLE 4.3-2 (Continued)

?.

E ENGINEERED SAFETY FEATURE ACTUATION SYSTEN INSTRUMENTATION SURVEILLANCE REQUIREMENTS c:

1 G

CHANNEL

. fl0 DES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

?

5.

TURBINE TRIP AND FEEDWATER ISOLATJON*

l a.

Steam Generator Water Level -

S R

M 1, 2 l

High-High

l' l

{,[:

6.

AUXILIARY FEEDWATER PUMPS START *

.: e a.

Manual N.A.

N.A.

R 1, 2, 3 l

i t-R b.

Steam Generator Water Level -

S R

M 1, 2, 3 l

i

,[l Low-Low

'Il Y

W c.

Loss of Normal and Preferred Power N.A.

R N.A.

1, 2, 3 l,

i I,i Ii

For Safety Injection input, see Item 1 for survelliance requirements.

'i

When above the P-11 setpoint.

1. When above the P-12 setpoint.

1 r

l 3

(1) Each train or logic channel shall be tested at least every other 31 days.

e

,, y,3..

n.

(2) The CHANNEL FUNCTIONAL TEST shall be the injection of-a simulated signal into the channel to verify the OPERABILITY of alarm, interlock, and/or trip functions which are not a part of the Reactor Trip System l

I functions.

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4 TROJAN-UNIT 1 3/4 3-31 Amendment No.131

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TROJAN-UNIT 1 3/4 3-32 Amendment No. 131

ML20209D940

Text

Attachment:

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