ML20082E669
| ML20082E669 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 07/23/1991 |
| From: | Barrett R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20082E672 | List: |
| References | |
| NUDOCS 9108010177 | |
| Download: ML20082E669 (18) | |
Text
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'J UNITED STATES E
M. F i NUCLEAR REGULATORY COMMISSION o\\
I WASHINGTON. D C. 20566
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COMMONWEALTH EDISON COMPANY AND 10WA-ILLIN015 GAS AND ELECTRIC COMPANY DOCKET NO. 50-265 OUAD CITIES NUCLEAR POWER STATION, UNIT 2 AMENDMENT TO FACULITY OPERATING LICENSE Amendment No. 125 License No. DPR-30 1.
The Nuclear Regulatory Commission (the Comn.ission) has found that:
A.
The application for amendment ty Immawealth Edison Company (the licensee) dated April 28,19M, complies with the standards and requ,1 " rents of the Atumic Energy Act of 1954, as amended (the Act; mid the Comnrission's rules me regulations set forth in 10 CFR Chapter 2; B.
The facility will operate in confornity with the application, the provisions of the Act, and thz rules and regulations of the Commission; C.
There is reasonable assurmice (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 Commission's regulations; D.-
The issuance of this amendment will not te inimical to the coinmon 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 Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical l
Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No. DPR-30 is I
hereby amended to read as follows:
l l
l 9108010177 910723 l
PDR ADOCK 05000265
\\ol{f P
-2 B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.125, are hereby incorporated in the license.
The licensee shall cperate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the startup from refueling outage No. 11.
FOR THE NUCLEAR REGULATORY COMMISSION W
pd Jyf R14hard J. Barrett, irector Project Directorate 111-2 Division of Reactor Projects - lil/lV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
July 23,1991 l-l l
ATTACHMENT TO LICENSE AMENDMENT NO. 125 FACILITY OPERATING LICENSE NO. DPR-30 DOCKET NO. 50-265 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.
The revised pages are identified by the captioned amendment number and contain merginal lines indicating the area of change.
REMOVE INSERT 1.1/2.1-2a 1.1/2.1-2a 1.1/2.1-3 1.1/2.1-3 1.1/2.1-10 1.1/2.1-10 1.1/2.1-11 1.1/2.1-11 1.1/2.1-12 3.1/4.1-5 3.1/4.1-5 3.1/4.1-6 3.1/4.1-6 3.1/4.1-7 3.1/4.1-7 3.1/4.1-7a 3.1/4.1-7b 3.1/4.1-10 3.1/4.1-10 3.1/4.1-11 3.1/4.1-11 3.1/4.1 -11 a 3.1/4.1-14 3.1/4.1-14 3.1/4.1-15
~.;...
QUAD CITIES-DPR-30 The definitions used above for the APRM scram trip apply. -In the event of operation with a maximem fraction limiting power density (MFLPD) greater than the fraction of rated pcwer (FRP),
the setting shall be modified as follows:
S1 (.58Wo + 50)
FRP MFLPD_
1 The definitions used above for the APRM scram trip apply.
The ratto of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is-less than 1.0, in which case the actual operating value will be used.
This adjustment may e>
be performed by increasi, e
'RH gain by the inverse rt 9,
MFLPD/FRP, which accom,ishes the e
same degree of protection as reducing the trip setting by FRP/MFLPD.
C.
Reactor low water level scram setting shall-be 144 inches above the top of the active fuel
- at normal operating conditions.
D.
Reactor low water level ECCS initiation shall be > 84 inches above the top of the active fuel
- at normal operating conditions.
E.
Turbine stop valve scram shall be l'101, valve closure _from-full open.
F.
The scram for turbine control valve fast closure due to-actuation of the fast acting solenoid valve shall be J 460 psig EHC fluid pressure.
- Top of active fuel-is defined to be 360 inches above vessel zero (See Bases 3.2).
1.1/2.1-2a Amendment No. 125
QUAD CITIES DPR-30 G.
Main steamline isolation valve closure scram shall be < 10%
valve closure from full'open.
H.
Main steamline low-pressure ini-tlation of main steamline isolation valve closure shall be 1 825 psig.
I.
Turbine EHC control fluid low-pressure scram on loss of control
)
oil pressure shall be set at greater than or equal to 900 psig.
1 J.
Condenser low vacuum scram shall l
be set at 1 21 inches Hg vacuum.
l i
1.1/2.1-3 Amendment No. 1?-
QUAD CITIES DPR-30 The trip setpoint of 2 460 psig EHC fluid pressure was developed to ensure that the pressure switch is actuated prior to the closure of the turbine control valves (at approximately 400 psig EHC fluid pressure) yet assure that the system is not actuated unnecessarily due to EHC system pressure transients which may cause EHC system pressure to momentarily decrease.
G.
Reactor Coolant Low Pressure Initiates Main Steam Isolation Valve Closure The low-pressure isolation at 825 psig was provided to give protection against fast reactor depr2ssurization and the resulting rapid cooldown of the vessel. Advantage was taken of the scram feature which occurs in the Run mode when the main steamitne isolation valves are closed to provide for reactor shutdown so that operation at pressures lower than those specified in the thermal hydraulic safety limit does not occur, although operation at a pressure lower than 825 psig would not necessarily constitute an unsafe condition.
H.
Main Steamline Isolation Valve Closure Scram The low-pressure isolation of the main steamlines at 825 psig was provided to give protection against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage was taken of the scram feature in the Run mode which occurs when the main steamline isolation valves are closed to provide for reactor shutdown so that high power operation at low reactor pressures does not occur, thus providing protection for the fuel cladding integrity safety limit.
Operation of the reactor at pressures lower than 825 psig requires that the reactor mode switch be in the Startup position, where protection of the fuel cladding integrity safety limit is provided by the IRH and APRM high neutron flux scrams.
Thus, the combination of main steamline low-pressure isolation and isolation valve closure scram in the Run node assures the availability of neutren flux scram protection over the entire range of appilcability of the fuel cladding integrity safety limit, in addition, the isolation valve closure stram in the Run mode anticipates the pressure and flux transients which occur during normal or inadvertent isolation valve closure.
Hith the scrams set at 101 valve closure in the Run mode, there is no increase in neutron flux.
1.1/2.1-10 Amendment No. 125 ra mtA w N
QUAD CITIES DPR-30 I.
Turbine EHC Control fluid Low-Pressure Scram The turbine EHC control system operates using high-pressure oil.
There are several points in this oil system where a loss of oli pressure could result in a fast closure of the turbine control valves.
This fast closure of the turbine control valves is not protected by the turbine control valve fast closure scram, since failure of the oil system would not result in the fast closure solenoid valves being actuated.
For a turbina control valve fast closure, the core would be protected by the APRM and high-reactor pressure scrams. However, to provide the same margins as provided for the generator load rejection on fast closure of the turbine control valves, a scram has been added to the reactor protection system which senses failure of control oil pressure to the turbine control system.
This is an anticipatory scram and results in reactor shutdown before any significant increase in neutron flux occurs.
The transient response is very similar to that resulting from the turbine control valve fast closure scram.
The scram setpoint of 900 psig is set high enough to provide the necessary anticipatory function and low enough to minimize the number of spurious scrams.
Normal operating pressure for this system is 1250 pstg.
Finally, the control valves will not start until the fluid pressure is 600 psig.
Therefore, the scram occurs well before valve closure begins.
J.
Condenser Low Vacuum Scram Loss of condenser vacuum occurs when the condenser can no longer handle the heat input.
Loss of condenser vacuum inttlates a closure of the turbine stop valves and turbine bypass valves which eliminates the heat input to the condenser.
Closure of the turbine stop and bypass valves causesapressuretransient,neutronfluxriseandanincreaseinsurfacej heat flux.
To prevent the cladding safety limit from being exceeded if this occurs, a reactor scram occurs on turbine stop valve closure in the Run mode.
The turbine stop valve closure scram function alone is adequate to prevent the cladding safety limit from being exceeded in the event of a turbine trip transient with bypass closure.
The condenser low vacuum scram is anticipatory to the stop valve closure scram and causes a scram before the stop valves are closed and thus the resulting transient is less severe.
Scram occurs in the Run mo:le at 21-inch Hg vacuum, stop valve closure occurs at 20-inch Hg vacuum, and bypass closure at 7-inch Hg vacuum.
1.1/2.1-11 Amendment No. 125
QUAD CITIES DPR-30 References 1.
" Generic Reload fuel Application," NEDE-240ll-P-A.*
- Approved revision number at time reload analyses are performed.
I 2.
" Qualification of the One-Dimensional Core Trar. stent Model for Boiling Water Reactors", General Electric Co. Licensing Topical Report NEDO 24154 Vols. I l
and II and NEDE-24154 Volume III as supplemented by ;etter dated September 5, 1980 f rom R.li. Buchholz (GE) to P.S. Check (NRC).
1.1/2.1-12 Amendment No.125
f QUAD-CITIES DPR-30 4.1 SURVEILLANCE REQUIREMENTS BASES A.
The minimum functional testing frequency used in this specification is based on a reliability attlysis using the concepts developed in Reference 1.
This concep;.ns specifically adapted to the one-out-of-two taken twice logic of the reactor protection system.
The analysis shows that the sensors are primarily responsible for the reliability of the reactor protection system. This analysis makes use of " unsafe failure" rate experience at conventional and nuclear power plants in a reliability mcdel for the system. An " unsafe failure" is defined as one which negates channel operability and which, due to its nature, is revealed only when the channel is functionally tested or attempts to respond to a real signal.
Failures such as blown fuses, ruptured bourdon tubes, faulted amplifiers, faulted cables, etc., which result in " upscale" or "downstale" readings on the reactor instrumentation are " safe" and will be easily recognized by the operators during operation because they are revealed by an alarm or a scram.
The thannels ljsted in Tabits 4.1-1 and 4.1-2 are divided into three groups respecting functional testing.
These are:
1.
On-off sensors that provide a scram trip function (Group 1);
I 2.
Analog devices coupled with bistable trips that provide a scram l
function (Group 2);
and 3.
Devicer which serve a useful function only during ;ome restricted l
mode of operation, such as Startup/Fot Standby, Refuel, or Shutdawn, or for which the only practical test is one that can be performed at shutdown (Group 3).
l The sensors that make up Group 1 are specifically selected from among l
the whole family of industrial on-off sensors that have earned an l
excellent reputation for reliable operation. Actual history on this class of sensors operating in nuclear power plants shows four failures in 472 sensor years, or a failure rate of 0.97 X 10-6/hr. During design, a goal of 0.99999 probability of success (at the 50% confidence level) was adopted to assure that a balanced and adequate design is achieved.
The probability of success is primarily a function of the sensor failure rate and the test interval. A 3-month test interval was planned for Group 1 sensors.
This is in keeping with good operating i
practice and satisfies the design goal for the logic configuration utilized in the reactor protection system.
i 3.1/4.1-5 Amendment No. 125
m.___
QUAD-CITIES DPR-30 To satisfy the long-term objective of maintaining an adequate level of safety throughout the plant lifetime, a minimum goal of 0.9999 at the 95%
confidence level is proposed.
Hith the one-out-of-two taken twice_ logic, this' requires:that each sensor have an availability of 0.993 at the 951 confidence _ level.
This level of availability may be maintained by adjusting the test interval.as-a function of the observed failure history (Reference 1).
To facilitate the implementation of this technique, Figure 4.1-1 is provided to indicate an appropriate trend in test interval.
The procedure is a: follows:
1.
Like sensors are pooled into one group for the purpose of data acquisition.
2.
The factor M is the exposure hours and is equal to the number of sensors in a group -n, times the elapsed time T(N.nT).
3.
The accumulated number of unsafe failures is plotted as an ordinate against M as an abscissa on Figure 4.1-1.
4.
After a trend is established, the appropriate monthly test interval to satisfy the goal-will be the test interval to the lef t of the plotted points.
5.
A-test interval of 1 month will be used inttlally until a trend is established.
The turbine control valve fast acting solenoid valve pressure switches directly - 1.
measure the trip oil pressure that causes the turbine control valves to close in 1 a rapid manner.
The re' actor scram setpoint was developed in accordance with NEDC 31336-" General Electric Instrument Setpoint Methodology" dated October,1986. As-part of the calculation, a calibration pericd is. inputted to achieve a nominal trip _ point and an allowable setpoint (Technical Specification value).
The nominal setpoint is procedurally controlled.
Based on the calculation input, the calibration period is defined to be every Rei'veling Outage.
Group 2 devices utilize-an analog sensor followed by an amplifier and a b1 stable trip circuit.
The sensor'and ampilfler are active components, and a failure is
-almost always accompanied by an alarm and an indtcation of the source of trouble.
In the event of failure, repair or substitution can start immediately.
An "as-is". failure is one that " sticks" midscale and is not capable of going
'either up or down in response to an out-of-11mits input.
This type of failure
-for analog devices is a rare occurrence and is detectable by an operato; who
-observes that one signal does not track the other three.
For purposes-of j
analysis,-it is. assumed that:this rare failure will be detected within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
j 3.1/4.1-6 Amendment No.125
QUAD-CITIES DPR-30 The bistable trip circuit which is a part of the Group 2 devices can sustain unsafe failures which are ree'e et ce!v on test.
Therefore, it is necessary l
to test them periodically.
A study was conducted of the instra ~.s n >
Jn channels included in the Group 2 devices to calculate their 'uMN r fr iure rates.
The analog devices (sensors and amplifiers) are predicteu to have an unsafe failure rate of I
less than 20 X 10-6 failures / hour.
The bistable trip circuit} are predicted to have an unsafe failure rate of less than 2 X 10-0 I
failures / hour.
e idering the 2-hour nonttoring interval for the analog devices as assu' ic"e and a weekly test interval for the bistable trip circuits, the de t re:eability goal of 0.99999 is attained with ample l
margin.
The bistable devices are monitored during plant operation to record their failure history and establish a test interval using the curve of Figure 4.1-1.
There are numerous identical bistable devices used throughout the plant instrumentation system.
Therefore, significant data on the failure rates for the bistable devices should be accumulated rapidly.
The frequency of calibration of the /.PRM flow b1asing network has been estabitsbed at each refueling outage.
The flow biasing network ls functionally tested at least once per month and, in addition, cross calibration checks of the flow input to the flow-blasing network can be made l during the functional test by direct meter reading (IEEE 279 Standard for
)
Nuclear Power Plant Protection Systems, Section 4.9, September 13, 1966).
There are several instrdments which must be calibrated, and it will take several days to perform the calibration of the entire network. While the calibration is being performed, a zero flow signal will be sent to half of i
l the APRMs resulting in a half scr w and rod block condition.
Thus, if the l
calibration were performed during operation, flux shaping would not be possible.
Based on expertence at other generating stations, drift of instrument such as those in the flow blasing network is not significant; l
therefore, to avoid spurious scrams, a calibration frequency of each l
refueling outage is estabitsbed.
l Reactor low water level instruments 2-263-57A, 2-263-57B, 2-263-5BA, and l
2-263-58B have been modified to be an analog trip system.
The analog trip system consists of an analog sensor (transmitter) and a master / slave trip unit setup which ultimately drives a trip relay.
The frequency of calibration and functional testing for instrument loops of the analog trip system, including reactor low water level, has been established in Licensing Topical Report NEDO-21617-A (December 1978).
With the one-out-of-two-taken-twice logic, NE00-21617-A states that each trip unit be subjected to a calibration / functional test frequency of one month. An l
adequato calibration / surveillance test interval for the transmitter is once per operating cycle.
l l
3.1/4.1-7 Amendment No. 125
- - ~ ~
DPR-30 Group 3 devices are active only during-a given portion of the operational cycle.
For example, the IRM is active during startup and inactive during ful'-power operatton.
Thus, the only test that is meaningful is the one performed just prior to shutdown or startup, i.e., the tests that are performed just prior to use of the instrument.
Calibration frequency of the instrument channel is divided into two groups.
These are as follows:
1.
Passive type indicating devices that can be compared with like units on a continuous basis, and 2.
Vacuum tube or semiconductor devices and detectors that drift or lose sensitivity.
Experience with passive type instruments in Commonwealth Edison generating stations and substations indicate that the specified calibrations are adequate.
For those devices which employ amplifiers, etc., drift specifications call for drift to be less than 0.4%/ month, i.e., in the period of a month a drift of 0.4% would occur, thus providing for adequate margin.
The sensitivity of_LPRM detectors decreases with exposure to neutron flux at a slow and_approximately constant rate. Changes in power distribution and electronic drift also require compensation.
This compensation is accomplished by calibrating the APRM system every 7 days using heat balance data and by calibrating individual LPRMs at least every 1000 equivalent l
full-power hours using TIP traverse data. Calibration on this frequency assures plant operation at or below thermal limits.
A comparison of Tables 4.1-1 and 4.1-2 indicates that some instrument i
channels have not been incluosd in the latter table. These are mode switch in shutdown, manual scram, high water level in scram discharge volume, mala-steamline isolation valve closure, and turbine stop valve closure. All of
]
the devices or-sensors associated with these scram functions are simple on-off switches, hence calibration is not applicable, t.e., the switch is either on or off.
Further, these-swltches are mounted solidly to the device and have a very low probability of moving; e.g., the thermal-switches in-the scram discharge volume' tank.
Based on the'above, no calibration is required for these instrument channels.
- B.
The MFLPD shall-be checked ~once per day to determine if the APRM scram requires adjustment.
This may normally be done by checking the LPRM readings, TIP traces, or process computer calculations. Only a small number of-control rods are moved daily, thus the peaking factors are not expected to change significa.itly and a daily check of the MFLPD is adequate.
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3.1/4.1-7a Amendment No. '25 l
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QUAD-CITIES-DPR-30 References 1 1.
I.'M..lacobs, "Rettability of Engineered Safety Features'as a function-of. Testing Frequency," Nuclear Safety, Vol. 9 No. 4, pp. 310-312, July-August, 1968.
2.
Licensing Topical Report NEDO-21617-A (December 1978).
3.-
NEDC - 31336 " General Electric Instrument Setpoint Methodology" dated October, 1986.~
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1
-3 1/4.1-7b Amendment No. 125 1
QUAD-CITIES DPR-30 TABLE 3.1-3 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENTS RUN MODE Minimum Number of Operable,or Tripped Instrument
~ Channels per Trio SystemLil-Trio Function Trio Level Settino ActionI23 1
Mode switch in shutdown A
1 Manual scram A
APRM[3]
2 High Flux (flow biased)
Specification 2.1.A.1 A or B 2-Inoperative A-or B 2-Downscale [113 2 3/125 of full scale A or B 2
High-reactor pressure 1 1060 psig A
.2 High drywell pressure
-1 2.5 psig A
2 Reactor low water level 2 8 inchesIB)
A
~
2 (per bank) High-water leve's fn scram discharge voluwe 1 40 gallons per bank A
2 Turbine condenser low 121 inches Hg vacuum A or C vacuum 2
Main Steamline high 115 X norwal full A or C radiation [123 power background (without hydrogen addition) 4
-Main steamline isolation 1 10% valve closure A or C valve closure [6]
2 Turbine control valve fast 1450 psig [10)
A or C closure, valve trip system oil pressure low [9]
i g
2' Turbine stop valve 1 10% valve closure A or C closure [9]
i 2
Turbine EHC control fluid
- 2 900 psig A or C
(;
low pressure I9]
I 3.1/4.1
Amer.dment No. 125 l
QUAD-CITIES DPR-30 7
TABLE 3.1-4 NOTES FOR TABLES 3.1-1, 3.1-2, AND 3.1-3
[1]
There shall be two operable trip systems or one operable and one tripped system for each function.
[2]
If the first column cannot be met for one of the trip systems, that trip system shall be tripped.
If the first column cannot be met for both trip systems, the appropriate actions IIsted below shall be taken:
A.
Initiate insertion of operable rods and complete insertion of all operable rods within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
B.
Reduce power level to IRH range and place mode switch in the Startup/ Hot Standby position within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
C.
Reduce turbine load and close main steamitne isolation valves within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
[3]
An APRM will be considered inoperable if there are fewer than 2 LPRH inputs per level or there are less than 50% of the norn.11 complement of LPRH's to an APRM.
[4]
Permissible to bypass, with control rod block for reactor protection system reset.in refuel and shutdown positions of the reactor mode switch.
[5]
Not required to be operable when primary containment integrity is not required.
.[6]
The design permits closure of any one line without a scram being initiated.
[7]
Automatically bypassed when reactor pressure is < 1060 psig.
[8]
The +8-inch trip point is the water level as measured by the instrumentation outside the shroud.
The water level inside the shroud will decrease as power is increased to 100% in comparison to the level outside the shroud to a maximum of 7 inches. This is due to the pressure drop across the steam dryer.
Therefore, at 100% power, an indication of +8-inch water level will actually be +1 inch inside the shroud.
1 inch on the water level instrumentation is 1504" above vessel zero.
(See Bases 3.2).
[9]
Permissible to bypass when first stage turbine pressure is less than that which corresponds to 45% rated-steam flow. (< 400 pst)
[10]
Trip is indicative of turbine control valve fast closure (due to low EHC fluid pressure) as a result of fast acting valve actuation.
3.1/4.1-11 Amendment No. 125
CUAD-CITIES DPR-30 TABLE 3.1-4 NOTES FOR TABLES 3.1-1, 3.1-2, AND 3.1-3 (Continued)
[11)
The APRM downstale trip function is automatically bypassed when the IRH instrumentation is operable and not high.
[12)
Channel shared by the reactor protection and containment isolation
- system, i
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3.1/4.1-11a
'ndment No.125 i
QUAD-CITIES DPR-30 TABLE 4.1-2 SCRAM INSTRUMENT CALIBRATI0h HINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Instrument channel Grovo[l]
Calibration StandardIS) t[1nimun Freauencv[2]
High flux IRH C
Comparison to APRH after Every controlled heat balance shutdown [4]
High flux APRM Output signal B
Heat balance Once every 7 days Flow bias B
Standard pressure and Refueling outage voltage source LPRH B(6) lising TIP system Every 1000 equivalen full power hours High reactor pressure-A Standard pressure source Every 3 months H1gh drywell pressure A-Standard pressure source Every 3 months Reactor low water level B
Water level
[7]
Turbine condenser low vacuum A
Standard vacuum source Every 3 months Main steamlin4 high radiation 6 Appropriate radiation Refueling outage source [3]
Turbia EHC control fluid A
Pressure source Every 3 months
. low pressure Turbine control valve A
Pressure source Refueling outage fast closure Highwater level in scram A
Nater level Refueling outage j
discharge volume (dp only) fioitti:
l l
[1] A-description of the three groups is included tc the bases of this l'
specification.
f2) Calibration tests are not required when the systeer are not required to be operable or are tripied.
If tests are missed. -% y shall be performed prior to returning the systin, to an operable states'.
[3] A-current source provides an instrument channel s%ewM every 3 months.
3.1/4.1-14 Amendment 125 1
a-----
. ~.
QUAD-CITIES-DPR-30 TABLE 4.1-2 SCRAM INSTRUMEN1 CALIBRATION HINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Notes:
(Continued)
(4) Maximum calibration frequency need not exceed once per week.
[5] Response time 15 not part of the routine instrument check tnd calibration but will be checked every refueling outage.
[6.1 Does not provide scram function.
[7] Trip units are calibrated monthly concurrently with functional testing.
Transmitters are calibrated once per operating cycle.
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l 3.1/4.1-15 Amendment No. 125
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