ML20133B254

Proposed Tech Specs,Rectifying Inconsistencies W/Installed Instrumentation Not in Agreement W/Rev 4 to NUREG-0452 & Removing Unnecessary Numerals from Spec Numbers

ML20133B254
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	09/23/1985
From:	PORTLAND GENERAL ELECTRIC CO.
To:
Shared Package
ML20133B237	List: ML20133B233 ML20133B249 ML20133B254
References
RTR-NUREG-0452, RTR-NUREG-452 TAC-56816, NUDOCS 8510030117
Download: ML20133B254 (41)

Text

LCA 117, Rev. 1

' Attachment 1 j

Pzgs 1 of 41 i..

2. (cont.)

C. (cont.)

(2) Technical Specifications The. Technical Specifications contained in Appendices A and.B, as revised through Amendment No. 106, are hereby

[

'06 incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifica-tions except where otherwise state'd in specific license conditions.

4 (3) Less than'Four Loop Operation The' licensees shall not operate the reactor at power levels above P-7 (as defined in Table 3.3-1 of Specifica-tion 3.3.1 of Appendix A to this license) with less than l (4) reactor coolant loops in operation until safety analyses for less than four loop operation have been submitted and approval for less than four loop operation at power levels above P-7 has been granted by the Com-mission by Amendment of this license.

(4) Report on Vibration Tests The licensees shall submit within six (6) months from the date of this license a final report analyzing the results of the preoperational vibration tests of the 7..

reactor internals performed during hot functional j,.

testing at the Trojan Nuclear Plant.

' s :s W

(5) Replacement of Certain Relays The licensees shall within.six (6) months and five.(5)

-days from the.date of this license replace all those output relays of the Solid State Protection System and

/\\

Auxiliary Safeguard Cabinets (except those relays which

/7\\

the licensees have shown to the satisfaction of the NRC Staff need not be replaced) with relays that satisfy the

/

seismic qualification criteria for the Trojan Nuclear

.3 Plant, or the reactor shall be shutdown until these l'

relays are. replaced.

N5 y (6) Spent Fuel Pool Modification g

The licensee is authorized to modify the spent fuel pool as described in the application dated August.1, 1983 and amended October 31, 1983.

88 Since spent fuel is now being-stored in the spent fuel

.O pool, upon conunencement of work on either the existing Amendment No. 7, 34, 35, 37, 7 g y

.40, through 84,.86 through 104 and 105 rp Date of Issuance:

-h A

April 24,.1985 Order of Modification M

p Dated April 20, 1981 i

LCA 117, Rey, j Page 2 of 41 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 9

SECTION Page 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 Axial Flux Difference................... 3/4 2-1 3/4.2.2 Heat Flux Hot Channel Factor............... 3/4 2-5 RCS Flowrate and F.................... 3/4 2-8 3/4.2.3 R

3/4.2.4 Quadrant Power Tilt Ratio................. 3/4 2-10 3/4.2.5 DNB Parameters.....

3/4~ 2-12 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION...

3/43-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION.......

3/4 3-13 l

3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation........... 3/4 3-33 Movable Incore Detectors................. 3/4 3-37 Seismic Instrumentation.................. 3/4 3-38 Meteorological Ins'trumentation.............. 3/4 3-41 Remote Shutdown Instrumentation.............. 3/4 3-44 Chlorine Detection Systems................ 3/4 3-47 Fire Detection Instrumentation.............. 3/4 3-48 Decouple Switches..................... 3/4 3-50 Accident Monitoring Instrumentation............ 3/4 3-51 Radioactive Liquid Effluent Instrumentation........ 3/4 3-54 Radioactive Gaseous and Process Effluent Monitoring Instrumentation................ 3/4 3-59 S0 Detection Systems................... 3/4 3-68 2

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS...................

3/44-1 TROJAN-UNIT 1 IV Amendment No. #8, 55, 78, 99, 105 April 4, 1985

LCA 117, Rev. 1 Page 3 of 41 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor trip system instrumentation setpoints shall be 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.

't W

i 1RONdt-Will 1 2-4 Amendment No.

1 LCA 117, Rev. 1 Page 4 of 41 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 1

I The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the j

values at which the Reactor Trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are p'revented 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 3

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 i

capability.

Power Rance. Neutron Flux The Power Range, Neutron Flux channel high setpoint provides reactor 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 Ranae. Neutron Flux. Hiah Positive Rate l

The' Power Range Positive Rate trip provides addad protection against I

rapid flux increases which are characteristic of rod ejection events from 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 from partial power. 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.

i Power Ranae. Neutron Flux. High Neaative Rate l

The Power Range Negative Rate trip provides protection to ensure that the minimum DNBR is maintained above the design DNBR value for multiple control rod drop accidents. The analysis of a single control rod drop accident indicates a return to full power may be initiated by the TROJAN-UNIT 1 B 2-3 Amendment No.

m,

...,,__..,_-._,...,,.m

_._,m,_

y.,_,,,, _,. _, ~.,

,,y..

_.,m_, _ _.,,. _,,,,,.,,,.,, _,,

I LCA 117 Rev. 1 I

Page 5 of 41 LIMITING SAFETY SYSTEM SETTINGS BASES automatic control system in response to a continued full power turbine load demand or by the negative moderator temperature feedback. This transient 1

will not result in a calculated DNBR of less than the design DNBR value, therefore single rod drop. protection'is not required.

Interwediate and Source Range. Nuclear Flux The Intermediate and Source Range, Nuclear Flux trips can provide rea'ctor 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 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 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 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 from 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-8 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 permissible after 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 Neutron Flux trip at the reduced power level.

4

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

LCA.117, Rsv. 1 Page 6 of 41 LIMITING SAFETY SYSTEM SETTINGS BASES Overpower AT The Overpower 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 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 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 r,elief through the pressurizer safety valves.

No credit was taken for oper9 tion of this trip in the accident analyses; however, its functional cdpability at the specified

.i trip setting is required by this specificati'on to enhance the overall reliability of the Reactor Protection System.

l e

I 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 38% (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 38% (P-8) of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any single loop drops below 90% of nominal full 5

i TROJAN-UNIT 1 B 2-5 Amendment No. (8,

LCA 117, Rev.1 Page 7 of 41 LIMITING SAFETY SYSTEM SETTINGS BASES loop flow. This latter trip will prever.t the minimum value of the 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 DNBR 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 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 >l.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.

Undervoltage and Underfreauency - Reactor Coolant Pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump bus trips provide reactor core protection when above the P-7 interlock setpoint against DNB as a result of loss of voltage (nominally 12.47 kV) or TROJAN-UNIT 1 B 2-6 Amendment No. #8,

LCA 117, Rev. 1 Page 8 of 41 LIMITING SAFETY SYSTEM SETTINGS BASES underf requency (nominally 60 Hz) to more than one reactor coolant pump.

l The specified setpoints assure a reactor trip signal is generated before the low flow trip setpoint is reached. A 0.1 second time delay in

.the underfrequency trip and a 0.2 second time delay in the undervoltage 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 System.

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

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

TROJAN-UNIT 1 B 2-7 Amendment No.

. ~-

h LCA 117, R:v.1 Page 9 of 41 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 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 j'

constitute compliance with the specification.

In the event the Limiting

~

Condition for Operation is restored prior to expiration of the specified i

time interval, completion of the ACTION statement is not required.

t 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:

l 1.

At least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, P.

At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and t

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

I' Where ccrrective measures are completed that permit operation under the ACTION rtquirements, the ACTION may be taken in accordance with the j

specified time limits as measured from the time of failure to meet the Limiting Coadition for Operation.

Exceptions to these requirements are 4

stated in the individual Specifications.

This Specification is not applicable in MODES 5 or 6.

l 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 t

ACTION statements unless otherwise excepted. This. provision shall not prevent passage through OPERATIONAL MODES as reguired 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 normal power source is inoperable, it may be i

considered OPERABLE for the purpose of satisfying the requirements of I

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 at least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in at least i.

HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in at least COLD SHUTDOWN i

within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. This specification is not applicable in MODES 5 or 6.

TROJAN-UNIT 1 3/40-1 Amendment No. 53,

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

LCA 117, Rev.1 Page 10 of 41 3/4.3 -INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 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.

SURVEILLANCE REUVIREMENTS 4.3.1.1 Each. reactor trip system instrumentation channel shall be l

demonstrated 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 4.3-1.

l

4. 3.1. 2 The logic for the interlocks shall be demonstrated ~0PERABLE

'during the at power CHANNEL FUNCTIONAL TEST of channels af fected 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

[

function shall be demonstrated to be within its limit at 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" column of Table 3.3-1.

4 I

i f

TROJAN-UNIT 1 3/43-1 Amendment No.

~.

TABLE 3.3-1

.-l. '

J REACTOR TRIP SYSTEM INSTRUMENTATION-a MINIMUM t

TOTAL NO.

CHANNELS CHANNELS APPLICABLE I

'E FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES-ACTION U

j 1.

Manual Reactor Trip 2

1 2

1, 2 11 2

2 1

2 3*,

4*, 5*

10 4

2.

Power Range, Neutron Flux A.

High Setpoint 4

2 3.

1 2

2#-

B.

Low Setpoint 4'

2 3

lI##,2 2#

i 3.

Power Range, Neutron Flux-4 2

3 1, 2 2#l 5

High Positive Rate 4.

Power Range, Neutron Flux, 4

2 3

1, 2 2#l High Negative Rate 1 ##, 2 3

l u

5.

Intermediate Range, Neutron Flux 2

1 2

s u

i 6.

Source Range, Neutron Flux Y

A.

Startup 2

1 2

1. 2#'

4 l

8.

Shutdown 2

1 2

3*,

4*, 5*

10 C.

Shutdown 2

0 1

3,4,5 5

j i

7.

Overtemperature AT Four L'oop Operation 4

2 3

1, 2 6#

Three Loop Operation 4

1**

3 1, 2 8

8.

Overpower AT Four Loop Operation 4

2 3

1, 2 6#

.e > t-j Three Loop Operation 4-1**

3 1, 2 8

g " Q '.

o 6# I ~o sn0.

l 5.

9.

Pressurizer Pressure - Low 4

2 3'

19 5

"am f

k10.

Pressurizer Pressure - High 4-2 3

1,.2 6# l 4 0 2

.@ 11.

Pressurizer Water Level - High 3

2 2

19 7#l**o?

TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION N

e MINIMUM Y

TOTAL NO.

CHANNELS CHANNELS APPLICABLE g FUNCTIONAL UNIT OF' CHANNELS TO TRIP OPERABLE MODES ACTION 12.

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

7#

(Above P-8)

'any oper-each oper-

~

ating loop ating loop 13.

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

7#

(Above P-7 and below P-8) two oper-each oper-ating loops ating loop 14.

Steam Generator Water 3/ loop 2/ loop in 2/ loop in 1, 2 7#

Level - Low-Low any oper-each oper-ating loop ating loop u

2 15.

Steam /Feedwater Flow 2/ loop-level 1/ loop-level 1/ loop-level 1, 2 7# l Mismatch and Low Steam and coincident and g

Generator Water Level 2/ loop-flow with 2/ loop-flow i"

mismatch 1/ loop-flow mismatch or mismatch in 2/ loop-level same loop and 1/ loon-flow mismatch 16.

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

1/ bus 19 6# l g

Pumps each bus o

$ 17.

Underfrequency - Reactor Coolant 4-2/ bus 1/ bus for 1/ bus 19 6# l g g g

Pumps each bus g g >-

C.

a ur

-?

TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION 0

MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE -

6 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION z

Q 18. Turbine Trip A.

Low Hydraulic Control 3

2 2

19 7#

l Oil Pressure B.

Turbine Stop Valve Closure 4-1/ valve 4-1/ valve 4'1/ valve it 7#

19.

Auto Safety-Injection Input 2

1 2

1, 2 1

l 20.

Reactor Coolant Pump Breaker 4-1/ breaker 2

1/ breaker 19 9#

Position Trip per oper-ating loop u 21.

Reactor Trip Breakers 2

1 2

1, 2 1

1 2

1 2

3*, 4*, 5*

10 Y

22.

Automatic Trip Logic 2

-1 2

1, 2 1

2 1

2 3*,

4*, 5*

10 1

4 MNO t

• M-5 CEC dg

~

2 x

R

~

LCA ll7,.Rev. 1 Page 14 of 41 TABLE 3.3-1 (Continued)

TABLE NOTATION

• 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 from the out of service Reactor Coolant Loop shall be placed in the trip-ped condition.

1. The provisions of Specification 3.0.4 are not applicable.

1. 1. When below the P-6 setpoint.

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

• When above the P-7 setpoint.

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

One channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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:

The inoperable channel is placed in th'e t' ripped condition a.

within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, the inoperable channel may be taken out of the tripped condition for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for troubleshooting and repair.

b.

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

c.

Either THERMAL POWER is restricted to 575% of RATED THERMAL POWER and the Power Range Neutron Flux trip setpoint is reduced to 185% of RATED THERMAL POWER wi. thin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, per Specification 4.2.4.c.

I TROJAN-UNIT 1 3/4 3-5~

Amendment No. 33,

LCA 117, Rev.1 Page.15 of 41 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, operation may continue, provided that THERMAL POWER is expeditiously placed within the range.of the Source Range or Power Range Instrumentation.

ACTION 4 - With the number of OPERABLE channels one less than the Mini-mum Channels ~ 0PERABLE requirement, suspend all operations involving positive-reactivity changes.

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

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

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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, the inoperable channel may be taken out of the tripped condition for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for troubleshooting and repair.

b.

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

ACTION 7 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may continue until performance of the next required CHANNEL FUNCTIONAL TEST provided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 8 - With a channel associated with an operating loop inoperable, I

restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT STANDBY within~ the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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

TROJAN-UNIT 1 3/4 3-6 Amendment No. 77,

,~

LCA 117, Rev.1 Page 16 of 41 TABLE 3.3-1 (Continued)

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

operation may continue provided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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

by the Minimum Channels OPERABLE requirement, restore the inoperable channel to CPERABLE status within 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />s-or open the reactor trip breakers within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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 211% 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-f requency, turbine trip, pressurizer low pressure, and pressurizer high level.

P-8 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, intermediate 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.

TROJAN-UNIT 1 3/4 3-7 Amendment No. 33, #5, #S,

~.- _ _ _,

LCA 117. Rev. 1 Page 17 of 41 1

s 2

(This page' intentionally blank) 1ROJAN-UNIT 1 3/43-8 Amendment No.

~. -

TABLE 3.3-2 y

REACTOR TRIP SYSTEM' INSTRUMENTATION RESPONSE TIMESI

-I u

FUNCTIONAL UNIT

. RESPONSE TIME I

E 1.

Power Range, Neutron Flux 50.5 seconds *

-3 l

2.

Power Range, Neutron Flux, 50.5 seconds

• High Negative Rate 3.

Overtemperature AT

$4.0' seconds

• I 4.

Pressurizer Pressure - Low 52.0 seconds 5.

Pressurizer Pressure - High

$2.0 seconds

[

6.

Loss of Flow - Single Loop 51.0 seconds

'l (Above P-8)

N i

7.

Loss of Flow - Two Loops 51.0 seconds

[

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

Steam Generator Water Level - Low-Low

$2.0 seconds l

9.

Undervoltage-Reactor Coolant Pumps

$1.2 seconds l

10. linderfrequency-Reactor Coolant Pumps 50.6 seconds l

1

• Neutron detectors are exempt from response time testing.

Response time shall be measured from detector output or input of first electronic component in channel.

i g

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

l 2M

~

1 kh r

u r

~~<

H H

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS No CHANNEL MODES IN WHICH

?

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE c: FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED 3

H 1.

Manual Reactor Trip N.A.

N.A.

R

'1, 2, 3*,

4*,

5*~

~

2.

Power Range, Neutron Flux

~

1, 2 A.

High Setpoint S

0(2,4), M(3,4)

Q Q(4,6) 8.

Low Setpoint S

R(4)

Q 14##, 2 3.

Power Range, Neutron Flux, N.A.

R(4)

Q 1, 2 l

High Positive Rate 4.

Power Range, Neutron Flux, N.A.

R(4)

Q 1, 2 High Negative Rate U

5.

Intermediate Range, S

R(4)#

S/U(1), Q l###. 2 Neutron Flux u

e l

h 6.

Source Range, Neutron Flux S

R(4,5)

S/U(1), Q 2##, 3, 4, 5(9) 7.

Overtemperature AT S

R Q

1, 2 I

8.

Overpower AT S

R Q

1, 2 9.

Pressurizer Pressure - Low

'S R

Q 19-l l g >rgg m

g 10.

Pressurizer Pressure - High S

R Q

1, 2 o

5. 11.

Pressurizer Water Level - High S

R Q.

19' loaS" s

esu k12.

Loss of Flow - Single Loop S

R Q

19 l $,0 z

4~<

o l

r l

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE ~ REQUIREMENTS

n 0>

CHANNEL MODES IN WHICH 7

~ CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REOUIRED 13'.

Loss of Flow - Two Loops S

R N.A.

19 l

14.

Steam Generator Water Level -

S R-Q 1, 2 l

Low-Low 15.

Steam /Feedwater Flow Mismatch S

R Q

1, 2 I

and Low Steam Generator Water Level 16.

Undervoltage - Reactor Coolant N.A.

R Q

19 l

Pumps

17. Underfrequency - Reactor Coolant N.A.

R Q

19 l

l Pumps C

~18.

Turbine Trip A.

Low Hydraulic Control Oil N.A R

S/U(1,8) 19 l

Pressure B.

Turbine Stop Valve Closure N.A.

N.A.

S/U(1,8) 19 l

19.

Auto Safety Injection Input N.A.

N.A.

R 1, 2 l

l g

20.

Reactor Coolant Pump Breaker N.A.

N.A.

9 19 lw g

Position Trip gg' c.

lm>yQ[

g 21.

Reactor Trip Breaker N.A.

N.A.

M(7,10) 1, 2, 3*,

4*, 5*

s cas: N l

22.

Automatic Trip Logic N.A.

N.A.

M(7) 1, 2, 3*,

4*, 5* loE l

g mH7 M

~

~

LCA 117, Rev. 1 Page 21 of 41 TABLE 4.3-1 (Continued)

NOTATION With the reactor trip system breakers closed and the control 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.

-9 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.

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

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

l (6) - Incore-Excore Calibration.

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

I (8) - Setpoint verification is not applicable.

(9) - See Specification 3/4.9.2 for audio and visual requirements in MODE 6.

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

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

LCA 117, Rev.1 Page 22 of 41 INSTRUMENTATION 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 mentation 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 ccnservativt 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.

1 SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel shall be demonstrated

[

OPERA 8LE 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.

4.3.2.2 The logic for the interlocks shall be demonstrated OPERABLE during l 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

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.

TROJAN-UNIT l-3/4 3-13 Amendment No.

TABLE 3.3-3 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 8>

Y MINIMUM

. TOTAL NO.

CHANNELS CHANNELS APPLICABLE y FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION h

" 1.

SAFETY INJECT 10NT a.

Actuation Logic'

'2 1

2 1,.2,.3, 4

13 l

h.

Manual Initiation 2

1 2

1,2,3,4 18 j.

c.

Containment 3

2 2

1,2,3,4 14* l Pressure - High R

d.

Pressurizer 3

2 2

1, 2, 3#

14* l

^

Pressure - Low u

i e.

Differential 1, 2, 3##

Pressure Between Steam Lines - High Four Loops 3/ steam line 2/ steam line 2/ steam line.

14*

Operating any steam line

1. 1. 1. steam 2/ operating 15 Three Loops 3/ operating 1

/

Operating steam line

.line, any steam line operating steam line ENE

@ Y Also initiates: Reactor Trip, Emergency Diesel Start, Auxiliary Feedwater, Turbine Trip, Feedwater Isolation,

%g>

Containment Isolation, Containment Ventilation Isolation, Control Room Isolation, Containment Cooling Fans, and g,[,

r+

a:

Essential Service Water.

usu o $'.

meg

" ~. <

~

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION No MINIMUM Y

. TOTAL NO.

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

Steam Flow in Two 1, 2, 3##

Y, Steam Lines - High Four Loops 2/ steam line 1/ steam line 1/ steam line 14* l Operating any 2 steam 4

lines Three Loops 2/ operating l###/any 1/ operating 15 Operating steam line operating steam line steam line U

COINCIDENT WITH EITHER us h

Tavg - Low-Low 1, 2, 3##

Four Loops 1 Tavg/ loop 2 Tavg any 1 Tavg any 14* l Operating loops 3 loops 1 Tav /

l### T in 1T in any 15 Three Loops Operating operaking any operating two operating avg avg loop loop loops

]

0 E

ZIE o e' y

mng

~?

w i

6

-1 TABLE 3.3-3 (Continued)

H ENGINEERED SAFETY FEATURE ACTUATION' SYSTEM INSTRUMENTATION 2

2:

MINIMUM E:

TOTAL NO.

CHANNELS CHANNELS APPLICABLE 3

l

$ FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES-ACTION l

H l

OR, COINCIDENT WITH 1

l Steam Line 1, 2, 3###

Pressure - Low Four Loops 1 pressure /

2 pressures 1 pressure 14*

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

1### pressure 1 pressure 15 Operating operating in any oper-in any 2 loop ating loop operating loops 2.

CONTAINMENT SPRAY l

a.

Actuation Logic 2

1 2

1,2,3,4 13 b.

Manual 2

2 2

1,2,3,4 18 l

c.

Containment Pressure -

4 2

3 1,2,3,4 16 l

High-High 3.

CONTAINMENT ISOLATION f

l a: ; w; 9

.a.

Containment Isolation Signal m

1)

Manual 2

1 2

1,2,3,4 18 eg l ~ x-mgy c>

2)

From Saf.ety Injection See 1 above for initiating functions and requirements.

N{

C ".<

.a TABLE 3.3-3'(Continued)'

' ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION

'No$-

MINIMUM-

?.

TOTAL NO.

CHANNELS'.

CHANNELS APPLICABLE

_g FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE

~ MODES ACTION 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.

N Spray-3)

From Safety Injection See 1 above for initia~ ting functions and requirements.

l e-u' 4)

Containment

~

2 Radioactivity - High Particulate 1

1-1 1,2,3,4 17 g

Iodine 1

1 1

High Level Noble Gas 1

1 1

Low Level Noble Gas 1

1 1

4.

STEAM LINE ISOLATION a.

Actuation Logic 2

1 2

1, 2,'3 19 l

l j$E b.

Manual 1/ steam line 1/ steam line 1/ operating-1,2,3 20 3'

steam line g

nn>

3 c.

Containment Pressure --

4 2

3 1,2,3 16

• $~

Q-g High-High

$((

o e'

^

z x.

mg

-s s

s M-TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURE' ACTUATION SYSTEM INSTRUMENTATION g

c-.

MINIMUM-

. TOTAL NO.

CHANNELS-CHANNELS APPLICABLE 5 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION' a

d.

Steam Flow in Two Steam 1, 2, 3

.l Lines - High Four Loops 2/ steam line 1/ steam line 1/ steam line 14*

Operating any 2 lines Three Loops 2/ operating 1###/any _

1/ operating 15 Operating steam line operating steam line steam line COINCIDENT WITH EITHER u2 Tavg - Low-Low 1,2,3 l

?

Four Loops 1 Tavg/ loop 1 Tavg any 1 Tavg in any.

14*

g Operating 2 loops 3 loops Three Loops 1 Tavg/oper-l### Tavg in 1 Tavg in any 15 Operating ating loop any operating two operating loop loops OR, COINCIDENT WITH 2NE lE E #

Steam Line Pressure - Low 1, 2, 3 l

gy g

Four Loops 1 pressure /

1 pressure 1 pressure 14*

Operating loop any 2 loops any 3 loops ugy g

0 Three Loops 1 pressure /

18## pressure 1 pressure in 15 U :o Operating operating loop in any oper-any 2 oper-wd-y ating loop ating loops

a

f..

4 TABLE 3.3-3 (Continued)

ENGINEERED' SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION No

.c MINIMUM

. TOTAL NO.

CHANNELS CHANNELS APPLICABLE' 6 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION z

Q 5.

TURBINE TRIP AND FEEDWATER.

l' ISOLATION

~

m

~

a.

From Safety Injection.

See 1 bove for initiating fbactions"and requirements.

.l b.

Steam' Generator Water Level -

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

l

(

High-Highx operating loop operating loon 6.

AUXILIARY FEEDWATER PUMPS START

-}-

i a.

Manual Initiation ~(Control 2/ pump 1/ pump 7

2/ pump 1,2,3 21 l

Room and Panel C-160) b.

From Safety Injection See 1 aboveffor initiating functic,n$ and requirements.

l c.

Steam Generator Water Level -

3/ steam 2/any si.eam 2/ steam 1,2,3 14*

Low-Low generator generator generator i

i d.

Loss of Normal and Preferred

. 2/ bus 1/ bus 1/ bus 1,2,3 18' l

Power g

M>F h

1 y

• R~

g MEC l

o $'

j g

%dM

^

O C

j y

j

.m H

4

LCA 117, R5v. 1 Page 29 of 41 e

TABLE 3.3-3 (Continued)

TABLE NOTATION

' *The provisions of Specificatier. 3.0.4 are not applicable.

l

1. When above the P-11 setpoint.

-l

1. 1. When above the P-12 setpoint.

l

1. 1. 1. The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall -tne 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTOOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. One channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE.

' 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

1 ACTION 15 - With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; however, one channel associated with an operating loop may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance ' testing per Specification 4.3.2.1.

ACTION 16 - With the number of OPERABLE Chann'els one less than the Total Number of Channels, operation may continue with the inoperable channel bypassed, provided that the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.2.1.

. 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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDBY within the next-6 hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Trojan-Unit 1 3/4 3-20 Amendment No.

LCA 117, Rav. 1 Page 30 of 41 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

One channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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' hours 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

~

i 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 i

P-ll 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 i

isolation, bypass, and regulating valves.

Trojan-Unit 1 3/4 3-21 Amendment No. #1,

f i

TABLE 3.3 !-

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS*

Mog FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES z

E:

1.

SAFETY INJECTION z

U a.

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

Pressurizer Pressure - Low t 1765 psig 1 1755 psig l

c.. Differential Pressure Between Steam

$ 100 psi 1 112 psi i

Lines - High I

d.

Steam Flow in Two Steam Lines - High 5 A function defined as

$ A function defined as l

Coincident with Tavg - Low-Low follows: 40% of full steam follows: 44% of full steam 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 u1 to 110% of full steam flow to 111.5% of full steam flow at full load at full load Ty Tava 2 553*F Tava 2 551*F 2 600 psig' steam line 2 500 psig steam line.

pressure pressure 2.

CONTAINMENT SPRAY a.

Containment Pressure - High-High

$ 30 psig 5 32 psig

'e > t-3.

CONTAINMENT ISOLATION

$"9 g

om g

a.

Containment Isolation Signal u

a o..

1.

From Safety Injection See 1 above for applicable data.

og C".

y

• The Actuation Logic and Manual Initiation Circuitry does not have trip setpoints or allowable values and is g

therefore not included in this table.

h 4

4 TABLE 3.3-4 (Continuedl-y ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS e>

Z FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES c:

b.

Containment Ventilation Isolation H

w 1.

From Safety Injection See 1 above for applicable data.

l i

2.

Containment Radioactivity - High 5 2 x background 5 2 x background 4.

STEAM LINE ISOLATION a.

Containment Pressure - High-High 5 30 psig 5 32 psig l

b.

Steam Flow in Two Steam Lines - High 5 A function defined as 5 A function defined as l

Coincident _with Tavg - Low or Steam follows: 40% of full steam follows: 44% of full steam Line Pressure - Low flow between 0% and 20% load flow between 0% and 20% load s s and then increasing linearly.and then increasing linearly u

'to 110% of full steam flow to' 111.5% of full steam flow i

E at full load at full load u

Tavg 2 553*F Tavg 2 551*F 2 600 psig steam 2 580 psig steam line pressure line pressure T

5.

TURBINE' TRIP AND FEEDWATER ISOLATION a.

From Safety Injection See 1 above for applicable data.

l y

b.

. Steam Generator Water Level -

5 75% of narrow range 5 76% of narrow range lyy>E g

High-High instrument span each steam instrument span each steam gg>

o.

generator generator ns I

U$C E

o a'

• "F z

O h

I

t 4:

TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS H

$~

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES z.

6.

AUXILIARY FEEDWATER PUMPS START a.

From Safety Injection See 1 above for applicable data.

l b.

Steam Generator Water Level -

2 5% of ' narrow range 2 3% of narrow range l

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

Loss of Normal and Preferred Power 2 2520 volts 2 2478' volts 1

~

4=

6 e

a n-O u

E og-

• "?

z

~. <

g 2

LCA 117, Rev. 1 Page 34.of 41 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMEST l

INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 1.

Manual Response Time testing is not applicable to manual initiation ci rcuitry.

2.

Containment Pressure - High a.

Safety Injection (ECCS) 5 27.0*

b.

Reactor Trip (from SI) 5 3.0 c.

Feedwater Isolation 5 8.0 d.

Containment Isolation Signal 5 18.0#/28.00##

l e.

Service Water System 5 13.0#/48.0##

l f.

Emergency Fan Coolers 5 10.0#/49.0##

l 3.

Pressurizer Pressure - Low a.

Safety Injection (ECCS) 5 13.0#/27.0*

b.

Reactor Trip (irom SI) 5 3.0 c.

Feedwate Isolation 5 8.0 d.

Containment Isolation Signal 5 18.0#

l l

e.

Service Water System 5 13.0#/48.0*

f.

Emergency Fan Coolers 5 10.0#/49.0##

l l

4.

Differential Pressure Between Steam Lines - High a.

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

b.

Reactor Trip (from SI) 5 3.0 c.

Feedwater Isolation 5 8.0 Trojan-Onit1 3/4 3-25

~ Amendment No. U,

I l

LCA 117, Rev. 1 Page 35 of 41 TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMEST INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 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 Tava - Low-Low a.

Safety Injection (ECCS) 5 15.0#/25.0##

b.

Reactor Trip (from SI)

$ 5.0 c.

Feedwater Isolation 5 10.0 d.

Containment Isolation Signal 5 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

$ 12.0#/51.0##

l 6.

Steam Flow in Two Steam Lines - High Coincident with Steam Line Pressure - Low a.

Safety Injection (ECCS)

$ 13.0#/23.0##

b.

Reactor Trip (from SI)

$ 3.0 c.

Feedwater Isolation 5 8.0 d.

Containment Isolation Signal.

5 18.0#/28.0##

l e.

Service Water System 5 14.0#/48.0##

f.

Steam Line Isolation 5 8.0

.l g.

Emergency Fan Coolers 5 10.0#/49.0##

l

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

LCA 117, Rev.1 Page 36 of 41 TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMESY INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 7.

Containment Pressure - Hiah-High a.

Containment Spray 5 30.0 6.

Steam Line Isolation 5 7.0 8.

Steam Generator Water Level - High-High a.

Turbine Trip - Reactor Trip

.5 2.5 b.

Feedwater Isolation 5 11.0 9.

Steam Generator Water Level - Low-low a.

Auxiliary Feedwater Pumps 5 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.

1. Diesel generator starting and sequence loading delays not 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. 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. If,

s TABLE 4.3-2 a

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS-c N4 CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE

.i FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED 1.

SAFETY INJECTION a.

Actuation Logic N.A.

N.A.

Q 1,2,3,4 l

b.

Manual Initiation

.N.A.

N.A.

R 1,2,3,4 l

q c.

Containment Pressure - High S

R Q

1,2,3,4

[

l R

d.

Pressurizer Pressure - Low S

R

'Q 1, 2, 3#

l u

e.

Differential Pressure S

R Q

1,2,3#

l Between Steam Lines - High

=

f.

Steam Flow in Two Steam S

R Q

1,2,3#

l i

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

j 2.

CONTAINMENT SPRAY a.

Actuation Logic N.A.

N.A.

Q 1,2,3,4 l

0 m>r b.

Manual Initiation N.A.

N.A.

R 1,2,3,4 l ggg g

ou a

D, c.

Containment Pressure - High-High S

R Q

1,2,3,4 l

u Q. "

" R."

z l

?x

" ~.0 t

~

4 1

q TA8LE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ~ ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 8>

?

CHANNEL MODES'IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE g

FUNCTIONAL UNIT CHECK CALIBRATION.

TEST REQUIRED

~

3.

CONTAINMENT ISOLATION

• l a.- Containment Isolation Signal l

1) Manual N.A.

N.A.

R 1,2,3,4 l_

l l

b.

Containment Ventilation-Isolation

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

l R

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

l l

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

Particulate S

R Q

l Iodine S

R Q

High Level Noble Gas S

R Q

Low Level Noble Gas S

R Q

4.

STEAM LINE ISOLATION g

a.

Actuation Logic N.A.

N.A.

Q l,.2, 3

[

n unn b.

Manual N.A.

N.A.

R 1,2,3 l

g g >-

o ns 5

c.

Containment Pressure - High-High S

R Q

1,2,3

l. $ $ y" 2

O O

d.

Steam Flow in Two Steam Lines -

See 1.f above for requirements, m n :e.

High Coincident with Tavg -

Low or Steam Line. Pressure. Low

~

s i

?

O w

4 TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION y

pc SURVEILLANCE REQUIREMENTS E

E CHANNEL MODES IN WHICH

=

E CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE y

FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED 5.

TURBINE TRIP AND FEEDWATER ISOLATION *

~

a.

Steam Generator Water Level -

S R

Q 1, 2 l

High-High 6.

AUXILIARY FEEDWATER PUMPS START

• a.

Manual N.A.

N.A.

R 1,2,3 l

R b.

Steam Generator Water Level -

S R

Q 1,2,3 l

Low-Low us E.

c.

Loss of Normal and Preferred Power N.A.

R N.A.

1, 2, 3 o

• For Safety Injection input, see Item 1 for surveillance requirements.

1. When above the P-11 setpoint.

1. 1. When above the P-12 setpoint.

'E m>r

[

$"O a

ax-es x

o e'

• ".F m

0~"

4 LCA 117,.Rev. 1 f

=o Page 40 of 41 1

i

)

i (This page intentionally blank)

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

j LCA 117, Rev. 1

.o Page 41 of 41 4

(This page intentionally blank) 4 TROJAN-UNIT 1 3/4 3-32 Amendment No.

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