ML20216H479
| ML20216H479 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 04/13/1998 |
| From: | Westreich B NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20216H482 | List: |
| References | |
| NPR-30-A-125, NUDOCS 9804210149 | |
| Download: ML20216H479 (24) | |
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4 UNITED STATES g
j NUCLEAR REGULATORY COMMISSION
,8 WASHINGTON. D.C. 20666 4001 o
2
%.....,o UNION ELECTRIC COMPANY CALLAWAY PLANT UNIT 1 DOCKET NO. 50-483 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.125 License No. NPF-30 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Callaway Plant Unit 1(the facility) Facility Operating License No. NPF-30 filed by the Union Electric Company (the Company),
dated October 31,1997, as supplemented by letter dated February 27,1998, 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, as amended, the provisions of the Act, and the rules and regulations of the Commission; 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 Commission's regulations; j
D.
The issuance of this license amendment will not be inimical to the common 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 Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-30 is hereby amended to read as follows:
9804210149 980413 PDR ADOCK 05000483 P
PDR y
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.. (2) Technical Soecifications and Environmental Protection Plar)
The Technical Specifications contained in Appendix A, as revised through Amendment No.125 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective as of its date of issuance to be implemented within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION f.\\ \\,\\ n Cwv e
Barry Westreich, Project Manager Project Directorate IV-2 Division of Reactor Projects lil/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance:
April 13, 1998 i
l l
l
ATTACHMENT TO LICENSE AMENDMENT NO.125 FACILITY OPERATING LICENSE NO. NPF-30 DOCKET NO. 50-483 l-l Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain marginallines q
indicating the areas of change. The corresponding overleaf pages are also provided to j
l maintain document completeness.
1
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CALLAWAY - UNIT 1 3/4 3-11 Amendment No. 77. 34
TABLE 4.3-1 (Continued)
TABLE NOTATIONS Only if the Reactor Trip System breakers happen to be closed and the Control Rod Drive System is capable of rod withdrawal.
The specified 18 month frequency may be waived for Cycle 1 provided the surveillance is performed prior to restart following the first refueling outage or June 1. 1986. whichever occurs first.
The provisions of Specification 4.0.2 are reset from performance of this surveillance.
- Below P-6 (Intermediate Range Neutron Flux interlock) Setpoint.
- Below P-10 (Low Setpoint Power Range Neutron Flux interlock) Setpoint.
(1)
If not performed in previous 31 days.
(2) Comnarison of calorimetric to excore power indication above 15% of RATED THEIRMALPOWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
(3) Single 30 int comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of MTED THERMAL POWER.
Recalibrate if the absolute difference is greater than or equal to 2%.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.
(5) For Source Range detectors, integral bias curves are obtained, evaluated, and com)ared to manufacturer's data.
For Intermediate Range and Power Range clannels, detector plateau curves shall be obtained evaluated, and compared to manufacturer's data.
For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
(6)
Incore - Excore Calibration, above 75% of RATED THERMAL POWER.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
Determination of the loop specific vessel AT and T values l
should be made when performing the Incore/Excore quarterly,r7 calibration, under steady state conditions.
(7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the Undervoltage and Shunt Trip Attachments of the Reactor Trip Breakers.
(8) Deleted (9) Quarterly surveillance in MODES 3*, 4*
and 5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window.
Quarterly surveillance shall include verification of the Boron Dilution Alarm Setpoint of less than or equal to an increase of 1.7 times the count rate within a 10-minute period.
CALLAWAY - UNIT 1 3/4 3-12 Amendment No. 19.28.8A.87,9A,125 4
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i 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Tria Setpoint Limits specified in Table 2.2-1 are the nominal values at which the teactor trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the core and Reactor Coolant System are prevented from exceeding their Safety Limits during normal o]eration and design basis anticipated operational occurrences and to assist I
t1e Engineered Safety Features Actuation System in mitigating the consequences of accidents. The Setpoint for a Reactor Trip System or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy.
To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated.
Allowable Values for the Reactor Trip Setpoints have been specified in Table 2.2-1. Operation with Setpoints less conservative than the Tria Setpoint but within the Allowable Value is acceptable since an allowance has 3een made in the safety analysis to accommodate this error. An optional provision has been included for determining the OPERABILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value.
The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation.
In Equation 2.2-1 Z + R + S s TA. the interactive effects of the errors in the rack and the sensor, and the "as measured" values of the errors are considered.
Z, as specified in Table 2.2-1, in percent span is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement.
TA or Total Allowance is the difference in percent span, between the Trip SetJoint and the value used in the analysis for Reactor trip.
R or Rack Error is tie "as measured" deviation, in percent span, for the affected channel from the specified Tri) Setpoint. S or Sensor Error is either the "as measured" deviation of tie sensor from its calibration point or the value specified in Table 2.2-1. in percent span, from the analysis assumptions.
Use of Equation 2.2-1 allows for a sensor drift factor, an l
increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS.
l The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels.
Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties.
Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes.
Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance.
Being that there is a small statistical chance that this m il happen, an infrequent excessive drift is expected.
Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.
CALLAWAY - UNIT 1 B 2-3 Amendment No. M.125
LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)
The various Reactor trip circuits autoratically open the Reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level.
In addition to redundant channels and trains, the design approach provides a Reactor Trip System which monitors numerous system variables, therefore providing Trip System functional diversity. The func-tional capability at the specified trip setting is required for those antici-patory or diverse Reactor trips for which no direct credit was assumed in the safety analysis to enhance the overall reliability of the Reactor Trip System. The Reactor Trip System initiates-a Turbine trip signal whenever Reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive. Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.
Manual Reactor Trip The Reactor Trip System includes manual Reactor trip capability.
Power Range. Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range trip setting. The Low Setpoint trip provides protection during subcritical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels.
The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.
Power Range. Neutron Flux. High Positive Rate l
The Power Range Positive Rate trip providas protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.
Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from mid-power.
CALLAWAY - UNIT 1 8 2-4 Amendment No.56
l LIMITfNG SAFETY SYSTEM SETTINGS BASES q
Intermediate and Source Rance. Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a subtritical condition. These trips provide redundant I
ofthePowerRange,NeutronFluxchgnnels.protectiontotheLowSetpointtrip The Source Range channels will initiate a Reactor trip at about 10 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to ap3roximately 25% of RATED THERMAL POWER unless manually blocked when P-10 )ecomes active.
Overtemoerature AT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of 3ressure, power, coolant temperature, and axial power distribution, provided t1at the transient is slow with respect to piping transit delays from the core to the temperature detectors, and pressure is within the range between the Pressurizer High and Low Pressure trips. The Setpoint is automatically varied with:
(1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loo) temperature detectors. (2) pressurizer pressure, and (3) axial power distri)ution. With normal axiil 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.
AT and T', as used in the Overtemperature AT trip, represent the 100%
a RTP values as measured by the plant for each loop.
For the startup of a refueled core AT is initially assumed at a value which is conservatively o
lower than the last measured 100% RTP AT for each loop.
Upon reaching 100%
o RTP, and during each quarterly Incort Excore CHANNEL CALIBRATION thereafter.
AT and T' are adjusted to be consistent with measured values for each loop.
n This normalizes each loo)'s Overtemperature AT trip to the actual operating conditions existing at t1e time of measurement, thus forcing the trip to reflect the equivalent full power conditions as assumed in the accident analyses.
These differences in vessel AT and T can arise due to several l
factors, the most prevalent being measured RCS IUop flows greater than Minimum Measured Flow, and slightly asymmetric power distributions between cuadrants.
While RCS laop flows are not expected to change with cycle life, racial power redistribution between quadrants may occur, resulting in small changes in loop specific vessel AT and T," values. Accurate determination of the loop values should be made when performing the specific vessel AT and T,7 calibration and under steady state conditions (i.e.,
Incore/Excore quarterly r power distributions not affected by Xe or other transient conditions).
The Allowable Value, as specified in Note 2 of Table 2.2-1, is associated with the uncertainties in the process rack electronics.
The Allowable Value provides a criterion for assessing the OPERABILITY of the process rack portion of the protection channel.
Deviations in excess of the Allowable Value are indicative of instrumentathn problems and should therefore result in an investiga? ion of the Overtemperature AT process rack OPERABILITY for the affected channel.
CALLAWAY - UNIT 1 B25 Amendment No. 28.57.102.125
LIMITING SAFETY SYSTEM FETTfNGS l
BASES Overtemoerature AT (Continued)
The time constants utilized in the lag compensation of measured AT, r,
3 and measured T T, are set in the field at 0 seconds. This setting correspondsto,6e7300NLLcardsusedforlagcompensaticnofthesesignals.
Safety analyses that credit Overtemperature AT for protection must account for these field adjustable lag cards as well as all other first order lags (i.e.,
the combined RTD/thermowell) response time and the scoop transport delay and thermal lag).
The safety analyses use a total first order lag of less than or equal to 6 seconds.
Overoower AT The Overpower AT trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits tha required range for Overtemperature AT trip, and provides a backup to the High Neutron Flux trip.
The Setpoint is automatically varied with:
(1) coolant temperature to correct for temperature induced changes in density and heat capacity of water.
l and (2) rate of change of temperature for dynamic compensation for aiping delays from the core to the loop temperature detectors, to ensure tlat the allowable heat generation rate (kW/ft) is not exceeded.
The Overpower AT trip l
provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226. " Reactor Core Response to Excessive Secondary Steam Releases."
AT and T", as used in the Overpower AT trip, represent the 100% RTP o
values as measured by the plant for each loop.
For the startup of a refueled core, AT is initially assumed at a value which is conservatively lower than o
the last measured 100% RTP AT for each loop.
Upon reaching 100% RTP, and during each quarterly Incore-Excore CHANNEL CALIBRATION thereafter, AT and T" o
are adjusted to be consistent with measured values for each loop.
This l
normalizes each loop's Overpower AT trip to the actual operating conditions existing at the time of measurement thus forcing the trip to reflect the equivalent full power conditions as assumed in the accident analyses.
These differences in vessel AT and T can arise due to several factors, the most I
prevalentbeingmeasuredRCS17pflowsgreaterthanMinimumMeasuredFlow, I
0 and slightly asymetric power distributions between cuadrants. While RCS loop flows are not expected to change with cycle life, racial power redistribution between quadrants may occur, resulting in small changes in loop specific vessel AT and T values.
Accurate determination of the loo) specific vessel AT and T valuiel should be made when performing the Incore/Excore quarterly recalibrWionandundersteadystateconditions(i.e.,powerdistributionsnot affected by Xe or other transient conditions).
The Allowable Value, as specified in Note 4 of Table 2.2-1, is associated I
with the uncertainties in the process rack electronics. The Allowable Value l
provides a criterion for assessing the OPERABILITY of the process rack portion of the protection channel.
Deviations in excess of the Allowable Value are indicative of instrumentation problems and should therefore result in an investigation of the Overpower AT process rack OPERABILITY for the affected channel.
CALLAWAY - UNIT 1 B 2-6 Amendment No. 28,102.125 t
LIMITING SAFETY SYSTEM SETTINGS BASES Overoower AT (Continued)
The time constants utilized in the lag compensation of measured AT 7,
3 and measured T r
are set in the field at 0 seconds.
This setting corresponds to he 9300 NLL cards used for lag compensation of these signals.
Safety analyses that credit Overpower AT for protection must account for these field adjustable lag cards as well as all other first order lags (i.e., the combined RTD/thermowell response time and the scoop transport delay and thermal lag).
The safety analyses use a total first order lag of less than or equal to 6 seconds.
Pressurizer Pressure In each of the pressurizer pressure channels, there are two independent bistables, each with its own Trip Setting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor o)eration is permitted.
The Low Setpoint trip protects against low pressure w11ch could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure.
On decreasing power the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power. automatically reinstated by P-7.
The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system l
overpressure.
Prgssurizer Water Level r
The Pressurizer High Water Level trip is provided to prevent water relief through the pressurizer safety valves.
On decreasing power the Pressurizer High Water Level trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power. automatically reinstated by P-7.
Reactor Coolant Flow The Low Reactor Coolant Flow trips provide core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.
On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10%
of full power equivalent) an automatic Reactor trip will occur if the flow in more than one loop drops below 90% of nominal full loo) flow. Above P-8 (a power level of approximately 48% of RATED THERMAL POWE1) an automatic Reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow.
Conversely, on decreasing power between P-8 and P-7 an automatic Reactor tri) will occur on low reactor coolant flow in more than one loop and below P-7 t1e trip function is automatically blocked.
CALLAWAY - UNIT 1 B 2-6a Amendment No.38-143.125
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