ML063050256

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Technical Specifications, Revision to Reactor Coolant System Specific Activity
ML063050256
Person / Time
Site: Wolf Creek Wolf Creek Nuclear Operating Corporation icon.png
Issue date: 10/31/2006
From: Donohew J
NRC/NRR/ADRO/DORL/LPLIV
To:
Wolf Creek
Donohew J N, NRR/DORL/LP4, 415-1307
Shared Package
ML062790358 List:
References
TAC MC8819
Download: ML063050256 (12)
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TABLE OF CONTENTS 1.0 US E A ND A P P LIC A TIO N ................................................................................. 1.1-1 1.1 Defi n itio n s .................................................................. .............................. 1 .1-1 1.2 Logical C onnectors ................................................................................. 1.2-1 1.3 C om pletio n T im es ..................................................................................... 1.3-1 1 .4 F re q u e n cy ................................................................................................. 1.4 -1 2.0 S A FETY LIM IT S (S Ls) ...................................................................................... 2.0-1 2 .1 S Ls .................................................................................................... 2 .0 -1 2 .2 S L V io la tio n s ............................................................................................. 2 .0 -1 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY ............... 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY ............................... 3.0-3 3.1 REACTIVITY CONTROL SYSTEMS ........................................................ 3.1-1 3.1.1 SHUTDOWN MARGIN (SDM) ......................................................... 3.1-1 3.1.2 Core Reactivity ..................................... 3.1-2 3.1.3 Moderator Temperature Coefficient (MTC) ...................................... 3.1-4 3.1.4 Rod Group Alignment Limits .............................................................. 3.1-7 3.1.5 Shutdown Bank Insertion Limits ....................................................... 3.1-11 3.1.6 C ontrol Bank Insertion Lim its ............................................................ 3.1-13 3.1.7 R od Position Indication ..................................................................... 3.1-16 3.1.8 PHYSICS TESTS Exceptions - MODE 2 ........................................ 3.1-19 3.2 POWER DISTRIBUTION LIMITS ............................................................. 3.2-1 3.2.1 Heat Flux Hot Channel Factor (FQ(Z))

(FQ Methodology) ................................. 3.2-1 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FXH) ............................ 3.2-6 3.2.3 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC) Methodology) .................................................... 3.2-9 3.2.4 QUADRANT POWER TILT RATIO (QPTR) ..................................... 3.2-10 3.3 INSTRUMENTATION ....................... ....... ...... 3.3-1 3.3.1 Reactori Trip System (RTS) Instrumentation .................................... 3.3-1 3.3.2 Engineered Safety Feature Actuation System (ESFAS)

Instrum entation ........................................................................... 3.3-2 1 3.3.3 Post Accident Monitoring (PAM) Instrumentation ............................. 3.3-37 3.3.4 Remote Shutdown System ............................................................... 3.3-41 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation .................................... 3.3-44 Wolf Creek --Unit 1 - - i Amendment No.-123-

TABLE OF CONTENTS 3.3 INSTRUMENTATION (continued) 3.3.6 Containment Purge Isolation Instrumentation .................................. 3.3-46 3.3.7 Control Room Emergency Ventilation System (CREVS)

Actuation Instrumentation ........................... 3.3-50 3.3.8 Emergency Exhaust System (EES) Actuation Instrum entation ........................................................................... 3.3-55 3.4 REACTOR COOLANT SYSTEM (RCS) ................................................... 3.4-1 3.4.1 RCS Pressure, Temperature, and Flow Departure from.

Nucleate Boiling (DN B) Lim its .................................................... 3.4-1 3.4.2 RCS Minimum Temperature for Criticality ........................................ 3.4-5 3.4.3 RCS Pressure and Temperature (P/T) Limits ................................. 3.4-6 3.4.4 RCS Loops - MODES 1 and 2 .......................................................... 3.4-8 3.4.5 RC S Loops - MO D E 3 ....................................................................... 3.4-9 3.4.6 RCS Loops - MO DE 4 ....................................................................... 3.4-12 3.4.7 RCS Loops - MODE 5, Loops Filled ................................................. 3.4-14 3.4.8 RCS Loops - MODE 5, Loops Not Filled .......................................... 3.4-17 3.4 .9 P ressurizer ........................................................................................ 3.4-19 3.4.10 Pressurizer Safety Valves .................................................................. 3.4-21 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) ....................... 3.4-23 3.4.12 Low Temperature Overpressure Protection (LTOP) System ........... 3.4-26 3.4.13 RCS Operational LEAKAGE ............................................................. 3.4-31 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage .................................. 3.4-33 3.4.15 RCS Leakage Detection Instrumentation ......................................... 3.4-37 3.4.16 RCS Specific Activity ................................. 3.4-41 3.4.17 Steam Generator (SG) Tube Integrity ................................................ 3.4-43 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) .............................. 3.5-1 3.5.1 Accumulators .... . .. ......... 3.5-1 3.5.2 EC CS - O perating ............................................................................. 3.5-3 3.5.3 EC CS - S hutdow n ............................................................................. 3.5-6 3.5.4 Refueling Water Storage Tank (RWST) .................... 3.5-8 3.5.5 S eal Injection Flow ............................................................................ 3.5-10 3.6 CONTAINMENT SYSTEMS ..................................................................... 3.6-1 3.6.1 Containment ........................... ............ 3.6-1 3.6.2 Containm ent A ir Locks..................................................................... 3.6-2 3.6.3 Containment Isolation Valves .... ..................

.... . .... .... 3.6-7 3.6.4 Containm ent Pressure ...................................................................... 3.6-14 3.6.5 Containment Air Temperature .......................................................... 3.6-15 3.6.6 Containment Spray and Cooling Systems ........................................ 3.6-16 3.6.7 Spray Additive System ...................................................................... 3.6-19 Wolf Creek - Unit 1 ii Amendment No. 123, 131, 157, 614, 167,

. -170

Definitions 1A1 1.0 USE AND APPLICATION

.1.1 Definitions

--------------------NOTE --------------------------------

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

Term Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

ACTUATION LOGIC TEST An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state required for OPERABILITY of a logic circuit and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices.

AXIAL FLUX DIFFERENCE AFD shall be the difference in normalized flux signals (AFD) between the top and bottom halves of an excore neutron detector.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps.

(continued)

Wolf Creek --Unit 1 1.1 -1 Amendment No. 123

Definitions 1.1 1.1 Definitions (continued)

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

CHANNEL OPERATIONAL A COT shall be the injection of a simulated or actual signal TEST (COT) 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 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.

CORE OPERATING LIMITS The COLR is the unit specific document that provides cycle REPORT (COLR) 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 DOSE EQUIVALENT 1-131 shall be that concentration of I-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 (continued)

WO-If Creek- Unit -1 1.1 - nAmendment No. --2-3,170

Definitions 1.1 1.1 Definitions (continued)

DOSE EQUIVALENT 1-131 3) ICRP 30, 1979, page 192-212, Table titled, "Committed (continued) Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity," or

4) Table 2.1 of EPA Federal Guidance Report No. 11, 1988, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion."

DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-1 33 shall be that concentration of Xe-1 33 (microcuries 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-1 33m, Xe-133, Xe-135m, Xe-135, and Xe-1 38 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 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil," or using the dose conversion factors from Table B-1 of Regulatory Guide 1.109, Revision 1, NRC, 1977.

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE (ESF) RESPONSE when the monitored parameter exceeds its ESF actuation TIME 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 the methodology for verification have been previously reviewed and approved by the NRC.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE
1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal (continued)

Wolf -Creek ---Unit 1- 1.1i-3 fAmendment No. 123, 13!, 170

Definitions 1.1 1.1 Definitions (continued)

LEAKAGE water injection or leakoff), that is captured and (continued) 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);
b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or 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.

MASTER RELAY TEST 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 eabh associated required slave relay. The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.

MODE 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.

OPERABLE--OPERABILITY 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 _ _

.........- -e-rf6fnihg theirreiat-ed support function(s).

(continued)

Wolf-Creek - Unit 1 -1_.1 _Ame nd m-e at No.- !A, 4 6 1 70

Definitions 1A 1.1 Definitions (continued)

PHYSICS TESTS 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 USAR;
b. Authorized under the provisions of 10 CFR 50.59; or
c. Otherwise approved by the Nuclear Regulatory Commission.

PRESSURE AND The PTLR is the unit specific document that provides the TEMPERATURE LIMITS reactor vessel pressure and temperature limits, including REPORT (PTLR) heatup and cooldown rates and the power operated relief valve lift settings and the Low Temperature Overpressure Protection (LTOP) System 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. Plant operation within these operating limits is addressed in LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," and LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System."

QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper excore RATIO (QPTR) 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.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 3565 MWt.

REACTOR TRIP The RTS RESPONSE TIME shall be that time interval from SYSTEM (RTS) RESPONSE when the monitored parameter exceeds its RTS trip setpoint TIME 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 the methodology for verification have been previously reviewed and approved by the NRC.

(continued)

Wolf Creek - Unit-1 1 .I -5 WAmendment No. 4-223, 170 I

Definitions 1.1 1.1 Definitions (continued)

SHUTDOWN MARGIN (SDM) 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 cluster 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.

SLAVE RELAY TEST 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.

STAGGERED TEST BASIS 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 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TRIP ACTUATING DEVICE A TADOT shall consist of operating the trip actuating device OPERATIONAL TEST (TADOT) 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.

Wolf Creek-- Unit 1 1-.1'6 Wo~f~rek~

mendment Uit 1 .1-6No. 12,70it

RCS Specific Activity 3.416 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 RCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133 specific activity shall be within limits.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT 1-131 ------------- NOTE ---------

not within limit. LCO 3.0.4c. 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 pCi/gm. I 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 1-131 to within limit.

B. DOSE EQUIVALENT ----------------- NOTE ---------

XE-133 not within limit. LCO 3.0.4c. 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-1 33 to within limit.

(continued)

Wolf Creek -- Unit 1 3.4-41 WAmrnendment No. 123, 155, 170

RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION 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 < 500 pCi/gm.

SR 3.4.16.2 ------------------- NOTE -------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 1-131 14 days specific activity < 1.0 pCi/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 Wolf CreeK --Unit 1- -3.4-42 -Aniendment No. 4-2-, 170

SG Tube Integrity 34.17 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 the Steam Generator Program.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS

NOTE----.

Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity of the 7 days satisfying the tube repair affected tube(s) is criteria and not plugged in maintained until the next accordance with the Steam refueling outage or SG Generator Program. tube inspection.

AND A.2 Plug the affected tube(s) in Prior to entering accordance with the Steam MODE 4 following Generator'Program. the next refueling outage or SG tube inspection 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.

Wolf Creek--- Unit-1 -a-.4-43 -Amu-ndmenrNo.-464,170

SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with the In accordance Steam Generator Program. with 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 Wolf Creek - Unit 3.4-44 AoUend4m nrft No. 46641170 -

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