ML063490274
| ML063490274 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 12/18/2006 |
| From: | Donohew J NRC/NRR/ADRO/DORL/LPLIV |
| To: | |
| Donohew J N, NRR/DORL/LP4, 415-1307 | |
| Shared Package | |
| ML066610193 | List: |
| References | |
| TAC MD1814 | |
| Download: ML063490274 (12) | |
Text
Operating License Revision 050 (4)
UE, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source of special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)
UE, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level UE is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal (100% power) in accordance with the conditions specified herein.
(2)
Technical Specifications and Environmental Protection Plan*
The Technical Specifications contained in Appendix A, as revised through Amendment No. 178 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)
Environmental Qualification (Section 3.11, SSER #3)--
Deleted per Amendment No. 169 Amendments 133, 134, &135 were effective as of April 30, 2000 however these amendments were implemented on April 1, 2000.
- The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.
Amendment 178 A140.0001
TABLE OF CONTENTS 3.3 INSTRUMENTATION (continued) 3.3.4 Remote Shutdown System 3.3-51 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation..........................................................................
3.3-54 3.3.6 Containment Purge Isolation Instrumentation 3.3-56 3.3.7 Control Room Emergency Ventilation System (CREVS)
Actuation Instrumentation 3.3-61 3.3.8 Emergency Exhaust System (EES) Actuation Instrumentation 3.3-66 3.3.9 Boron Dilution Mitigation System (BDMS) 3.3-71 3.4 REACTOR COOLANT SYSTEM (RCS)................................................
3.4-1 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits...................................................
3.4-1 3.4.2 RCS Minimum Temperature for Criticality.......................................
3.4-3 3.4.3 RCS Pressure and Temperature (P/T) Limits..................................
3.4-4 3.4.4 RCS Loops - MODES 1 and 2.........................................................
3.4-6 3.4.5 RCS Loops - MODE 3.....................................................................
3.4-7 3.4.6 RCS Loops - MODE 4.....................................................................
3.4-10 3.4.7 RCS Loops - MODE 5, Loops Filled................................................
3.4-12 3.4.8 RCS Loops - MODE 5, Loops Not Filled.........................................
3.4-15 3.4.9 P ressurizer.......................................................................................
3.4-17 3.4.10 Pressurizer Safety Valves................................................................
3.4-19 3.4.11 Pressurizer Power Operated Relief Valves (PORVs)...................... 3.4-21 3.4.12 Cold Overpressure Mitigation System (COMS)...............................
3.4-25 3.4.13 RCS Operational LEAKAGE...........................................................
3.4-30 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage
................................ 3.4-32 3.4.15 RCS Leakage Detection Instrumentation 3.4-36 3.4.16 RCS Specific Activity............................................
........................... 3.4-40 3.4.17 Steam Generator (SG) Tube Integrity..............................................
3.4-42 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)........................... 3.5-1 3.5.1 Accumulators...................................................................................
3.5-1 3.5.2 ECCS - Operating
........ 3.5-3 3.5.3 ECCS - Shutdown............................................................................
3.5-6 3.5.4 Refueling Water Storage Tank (RWST) 3.5-8 3.5.5 Seal Injection Flow...........................................................................
3.5-10
-3.6 ---.....
CONTAINMENT SYSTEMS
..:: :3.6-3.6.1 Containment 3.6-1 3.6.2 Containment Air Locks.....................................................................
3.6-3 CALLAWAY PLANT 2
Amendment 178
Definitions 1.1 1.1 Definitions (continued)
CHANNEL OPERATIONAL TEST (COT)
CORE ALTERATION CORE OPERATING LIMITS REPORT (COLR)
DOSE EQUIVALENT 1-131 A COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY such that the setpoints are within the necessary range and accuracy. The COT may be performed by means of any series of sequential, overlapping, or total channel steps.
CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel.
Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
The COLR is the unit specific document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.
DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131, 1-132,1-133,1-134, and 1-135 actually present.
The determination of DOSE EQUIVALENT 1-131 shall be performed using thyroid dose conversion factors from:
- 1. Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or
- 2. Table E-7 of Regulatory Guide 1.109, Revision 1, NRC, 1977, or
- 3. International Commission on Radiological Protection (ICRP) Publication 30, "Limits for Intakes of Radionuclides by Workers," Supplement to Part 1, pages 192-212, Table titled "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity," 1979, or
- 4. Table 2.1 of EPA Federal Guidance Report No. 11, EPA-
-520/1-88-020, "Limiting Values of Radionuclide Intake -
and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1988.
(continued)
Amendment 178 CALLAWAY PLANT 1.1-2
Definitions 1.1 1.1 Definitions (continued)
DOSE EQUIVALENT XE-133 ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME LEAKAGE DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuies per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-87, Kr-88, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity.
The determination of DOSE EQUIVALENT XE-133 shall be performed using the effective dose conversion factors for air submersion listed in Table I11.1 of EPA Federal Guidance Report No. 12, EPA-402-R-93-081, "External Exposure to Radionuclides in Air, Water, and Soil", 1993.
The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.
LEAKAGE shall be:
- a. Identified LEAKAGE
- 1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water leakoff), that is captured and conducted to collection systems or a sump or collecting tank;
- 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or'not to be pressure boundary LEAKAGE; or
- 3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);
(continued)
CALLAWAY PLANT 1.1-3 Amendment 178 CALLAWAY PLANT 1.1-3 Amendment 178
Definitions 1.1 1.1 Definitions LEAKAGE (continued)
- b. Unidentified LEAKAGE MASTER RELAY TEST MODE All LEAKAGE (except RCP seal water leakoff) that is not identified LEAKAGE;
- c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.
A MASTER RELAY TEST shall consist of energizing all master relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required master relay. The MASTER RELAY TEST shall include a continuity check of each associated required slave relay. The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.
A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:
- a. Described in Chapter 14 of the FSAR;
- b. Authorized under the provisions of 10 CFR 50,59; or
- c. Otherwise approved by the Nuclear Regulatory Commission.
OPERABLE - OPERABILITY PHYSICS TESTS (continued)
CALLAWAY PLANT 1.1-4 Amendment 178 I
Definitions 1.1 1.1 Definitions (continued)
PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
QUADRANT POWER TILT RATIO (QPTR)
RATED THERMAL POWER (RTP)
REACTOR TRIP SYSTEM (RTS) RESPONSE TIME SHUTDOWN MARGIN (SDM)
The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and REPORT (PTLR) cooldown rates, the power operated relief valve (PORV) lift settings, and the Cold Overpressure Mitigation System (COMS) arming temperature, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.6.
QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.
RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3565 MWt.
The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.
SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
- a. All rod duster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and
- b. In MODES 1 and 2, the fuel and moderator temperatures
-- --are changed to the hot zero-power temperatures.-
(conunuea)
Amendment 178 CALLAWAY PLANT 1.1-5 I
Definitions 1.1 1.1 Definitions (continued)
SLAVE RELAY TEST STAGGERED TEST BASIS THERMAL POWER TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT)
A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.
A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.
THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
A TADOT shall consist of operating the trip actuating device and verifying the OPERABILITY of all devices in the channel required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy. The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps.
CALLAWAY PLANT 1.1-6 Amendment 178 I
RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 APPLICABILITY:
RCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133 specific activity shall be within limits.
MODES 1, 2, 3, and 4.
ACTIONS COM PLETION CONDITION REQUIRED ACTION TIME TIME A.
DOSE EQUIVALENT NOTE 1-131 not within limit.
LCO 3.0.4.c is applicable.
A.1 Verify DOSE Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> EQUIVALENT 1-131
_< 60 ipCi/gm.
AND A.2 Restore DOSE 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> EQUIVALENT 1-1031 to within limit.
B.
DOSE EQUIVALENT NOTE XE-133 not within limit.
LCO 3.0.4.c is applicable.
B.1 Restore DOSE 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> EQUIVALENT XE-133 to within limit.
(continued)
CALLAWAY PLANT 3.4-40 Amendment No. 178
RCS Specific Activity 3.4.16 ACTIONS (continued)
COM PLETION CONDITION REQUIRED ACTION TIME TIME C.
Required Action and C.1 Be in MODE 3.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A or B AND not met.
C.2 Be in MODE 5.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR DOSE EQUIVALENT 1-131
> 60 pCi/gm.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 NOTE Only required to be performed in MODE 1.
Verify reactor coolant DOSE EQUIVALENT XE-133 7 days specific activity < 225 1iCi/gm.
NOTE Only required to be performed in MODE 1.
Verify reactor coolant DOSE EQUIVALENT 1-131 14 days specific activity < 1.0 RCi/gm.
AND Between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a THERMAL POWER change of Ž_ 15% RTP
-within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> -
period CALLAWAY PLANT 3.4-41 Amendment No. 178
SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 SG tube integrity shall be maintained.
AND All SG tubes satisfying the tube repair criteria shall be plugged in accordance with Steam Generator Program.
APPLICABILITY:
MODES 1 2, 3, and 4.
ACTIONS
.-.---------------.....------------- NOTE Separate Condition entry is allowed for each SG tub CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more SG tubes A.1 Verify tube integrity of 7 days satisfying the tube repair the affected tube(s) is criteria and not plugged in maintained until the next accordance with the Steam refueling outage or Generator Program.
inspection.
AND A.2 Plug the affected tube(s)
Prior to entering in accordance with the MODE 4 following Steam Generator the next refueling Program.
outage or SG tube inspection (continued)
CALLAWAY PLANT 3.4-42 Amendment No. 178 I
SG Tube Integrity 3.4.17 ACTIONS (continued)
COMPLETION CONDITION REQUIRED ACTION TIME B.
Required Action and B.1 Be in MODE 3.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.
B.2 Be in MODE 5.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR SG tube integrity not maintained.
CALLAWAY PLANT 3.4-43 Amendment No. 178 I
SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with the Steam In accordance with Generator Program.
the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that satisfies the Prior to entering tube repair criteria is plugged in accordance with the MODE 4 following Steam Generator Program.
a SG tube inspection CALLAWAY PLANT 3.4-44 Amendment No. 178 I