ML061280471
| ML061280471 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 05/08/2006 |
| From: | Plant Licensing Branch III-2 |
| To: | |
| Donohew J N, NRR/DORL, 415-1307 | |
| Shared Package | |
| ML061280189 | List: |
| References | |
| TAC MC8842, TAC MD0122 | |
| Download: ML061280471 (30) | |
Text
TABLE OF CONTENTS 1.0 USE AND APPLICATION................................................. 1.1-1 1.1 Definitions.................................................
1.1-1 1.2 Logical Connectors.................................................
1.2-1 1.3 Completion Times................................................
1.3-1 1.4 Frequency................................................
1.4-1 2.0 SAFETY LIMITS (SLs)................................................
2.0-1 2.1 SLs 2.0-1 2.2 SL Violations................................................
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 Control Bank Insertion Limits.................................................
3.1-13 3.1.7 Rod 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 (Fo(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 Reactor Trip System (RTS) Instrumentation...................................
3.3-1 3.3.2 Engineered Safety Feature Actuation System (ESFAS)
Instrumentation........................................................................... 3.3-21 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 Instrumentation........................................................................... 3.3-55 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........................................
3.4-5 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 RCS Loops - MODE 3.....................................................
3.4-9 3.4.6 RCS Loops - MODE 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 Pressurizer..................................................... 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................................................ 3.4-45 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.........................
............................ 3.6-1 3.6.1 Containment.....................................................
3.6-1 3.6.2 Containment Air Locks.....................................................
3.6-2 3.6.3 Containment Isolation Valves.....................................................
3.6-7 3.6.4 Containment Pressure.......................
.............................. 3.6-15 3.6.5 Containment Air Temperature.....................................................
3.6-16 3.6.6 Containment Spray and Cooling Systems........................................
3.6-17 3.6.7 Spray Additive System.......................
.............................. 3.6-20 Wolf Creek - Unit 1 ii Amendment No. 123, 131, 157, 164
TABLE OF CONTENTS 3.7 PLANT SYSTEMS..................................................... 3.7-1 3.7.1 Main Steam Safety Valves (MSSVs)................................................
3.7-1 3.7.2 Main Steam Isolation Valves (MSIVs)..............................................
3.7-5 3.7.3 Main Feedwater Isolation Valves (MFIVs)........................................ 3.7-7 3.7.4 Atmospheric Relief Valves (ARVs)...................................................
3.7-9 3.7.5 Auxiliary Feedwater (AFW) System..................................................
3.7-11 3.7.6 Condensate Storage Tank (CST).....................................................
3.7-14 3.7.7 Component Cooling Water (CCW) System......................................
3.7-16 3.7.8 Essential Service Water System (ESW)........................................... 3.7-18 3.7.9 Ultimate Heat Sink (UHS)...................................
.................. 3.7-20 3.7.10 Control Room Emergency Ventilation System (CREVS).........
3.7-22 3.7.11 Control Room Air Conditioning System (CRACS)............................
3.7-25 3.7.12 Emergency Core Cooling System (ECCS) Pump Room Exhaust Air Cleanup System - Not Used........
................. 3.7-28 3.7.13 Emergency Exhaust System (EES).........................
3.7-29 3.7.14 Penetration Room Exhaust Air Cleanup System (PREACS) -
Not Used..
3.7-33 3.7.15 Fuel Storage Pool Water Level................................................
3.7-34 3.7.16 Fuel Storage Pool Boron Concentration...........................................
3.7-35 3.7.17 Spent Fuel Assembly Storage..........................................
....... 3.7-37 3.7.18 Secondary Specific Activity................................................
3.7-39 3.8 ELECTRICAL POWER SYSTEMS.........................
3.8-1 3.8.1 AC Sources - Operating.........................
3.8-1 3.8.2 AC Sources - Shutdown........................
3.8-18 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air........................
3.8-21 3.8.4 DC Sources - Operating........................
3.8-24 3.8.5 DC Sources - Shutdown.........................
3.8-28 3.8.6 Battery Cell Parameters.........................
3.8-30 3.8.7 Inverters - Operating........................
3.8-34 3.8.8 Inverters - Shutdown........................
3.8-35 3.8.9 Distribution Systems - Operating........................
3.8-37 3.8.10 Distribution Systems - Shutdown.........................
3.8-39 3.9 REFUELING OPERATIONS.........................
3.9-1 3.9.1 Boron Concentration.........................
3.9-1 3.9.2 Unborated Water Source Isolation Valves............
............ 3.9-2 3.9.3 Nuclear Instrumentation.........................
3.9-3 3.9.4 Containment Penetrations.........................
3.9-5 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation - High Water Level.
3.9-7 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level.
3.9-9 Wolf Creek - Unit 1 iii Amendment No. 123, 132, 134, 163
TABLE OF CONTENTS 3.9 REFUELING OPERATIONS (continued) 3.9.7 Refueling Pool Water Level.......................
3.9-11 4.0 DESIGN FEATURES.......................
4.0-1 4.1 Site Location.......................
4.0-1 4.2 Reactor Core.......................
4.0-1 4.3 Fuel Storage.......................
4.0-1 5.0 ADMINISTRATIVE CONTROLS.......................
5.0-1 5.1 Responsibility.......................
5.0-1 5.2 Organization.......................
5.0-2 5.3 Unit Staff Qualifications.......................
5.0-4 5.4 Procedures.......................
5.0-5 5.5 Programs and Manuals.......................
5.0-6 5.6 Reporting Requirements.......................
5.0-22 5.7 High Radiation Area.......................
5.0-27 Wolf Creek - Unit 1 iv Amendment No. 123, 142,-158,164
Definitions 1.1 1.1 Definitions (continued)
E-AVERAGE DISINTEGRATION ENERGY (continued)
ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME LEAKAGE gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > 15 minutes, making up at least 95% of the total noniodine activity in the coolant.
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 the 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 injection or 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);
- b.
Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE; (continued)
I Wolf Creek - Unit 1 1.1-3 Amendment No. 423,131, 164
Definitions 1.1 1.1 Definitions (continued)
LEAKAGE (continued)
- 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 each 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 PHYSICS TESTS 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 USAR;
- b.
Authorized under the provisions of 10 CFR 50.59; or
- c.
Otherwise approved by the Nuclear Regulatory Commission.
(continued)
Wolf Creek - Unit 1 1.1 -4 Amendment No. 142, 164
RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:
- a.
- b.
1 gpm unidentified LEAKAGE;
- c.
10 gpm identified LEAKAGE; and I
- d.
150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).
I APPLICABILITY:
MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
RCS operational A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LEAKAGE not within limits within limits.
for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.
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 Pressure boundary LEAKAGE exists.
OR Primary to secondary LEAKAGE not within limit.
I Wolf Creek - Unit 1 3.4-31 Amendment No. 42-3, 164
RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1
- --------NOTES--------------- ----------
- 1.
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.
- 2.
Not applicable to primary to secondary LEAKAGE.
Verify RCS operational LEAKAGE is within limits by performance of RCS water inventory balance.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> I
NOTED Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.
Verify primary to secondary LEAKAGE is < 150 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> gallons per day through any one SG.
Wolf Creek - Unit 1 3.4-32 Amendment No. 4213, 164
SG Tube Integrity 3.4.17 l 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 NOT Separate Condition entry is allowed for each SG t r-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.
WolfCrek nitI 34-45Amedmen No 16 Wolf Creek - Unit I 3.4-45 Amendment No. 164
SG Tube Integrity l 3.4.17 l 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 1 3.4-46 Amendment No. 164
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:
- a.
Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the 'as found' condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The 'as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.
Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.
- b.
Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.
- 1.
Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
- 2.
Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.
Leakage is not to exceed I gpm per SG.
(continued)
Wolf Creek - Unit 1 5.0-1 1 Amendment No. 123, 15-, 164
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued)
- 3.
The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."
- c.
Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.
The following alternate tube repair criteria may be applied as an alternative to the 40% depth-based criteria:
- 1.
For Refueling Outage 14 and the subsequent operating cycle, degradation found in the portion of the tube below 17 inches from the top of the hot leg tubesheet does not require plugging. All tubes with degradation identified in the portion of tube within the region from the top of the hot leg tubesheet to 17 inches below the top of the tubesheet shall be removed from service.
- d.
Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. For Refueling Outage 14 and the subsequent operating cycle, the portion of the tube below 17 inches from the top of the hot leg tubesheet is excluded. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
- 1.
Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.
(continued)
Wolf Creek - Unit 1 5.0-1 2 Amendment No. 123, 162, 164
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued)
- 2.
Inspect 100% of the tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected.
- 3.
If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
- e.
Provisions for monitoring operational primary to secondary LEAKAGE.
(continued)
Wolf Creek - Unit 1 5.0-1 3 Amendment No. 123,16, 164
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. The program shall include:
- a.
Identification of a sampling schedule for the critical variables and control points for these variables;
- b.
Identification of the procedures used to measure the values of the critical variables;
- c.
Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage;
- d.
Procedures for the recording and management of data;
- e.
Procedures defining corrective actions for all off control point chemistry conditions; and
- f.
A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.
5.5.11 Ventilation Filter Testing Program (VFTP)
A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 2, and in accordance with the guidance specified below.
- a.
Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 1% when tested in accordance with Regulatory Guide 1.52, Revision 2 at the system flowrate specified below +/- 10%.
ESF Ventilation System Flowrate Control Room Emergency Ventilation System-Filtration 2000 cfm Control Room Emergency Ventilation System-Pressurization 750 cfm Auxiliary/Fuel Building Emergency Exhaust 6500 cfm (continued)
Wolf Creek - Unit 1 5.0-14 Amendment No. 123,15-,164 l
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (VFTP) (continued)
- b.
Demonstrate for each of the ESF systems that an inplace test of the charcoal absorber shows a penetration and system bypass < 0.05% when tested in accordance with Regulatory Guide 1.52, Revision 2 at the system flowrate specified below +/- 10%.
ESF Ventilation System Flowrate Control Room Emergency Ventilation System - Filtration 2000 cfm Control Room Emergency Ventilation System-Pressurization 750 cfm Auxiliary/Fuel Building Emergency Exhaust 6500 cfm
- c.
Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal absorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 300C and greater than or equal to the relative humidity specified below.
ESF Ventilation System Penetration RH Control Room Emergency Ventilation System (Filtration/Pressurization) 2.5%
70%
Auxiliary/Fuel Building Emergency Exhaust 5%
70%
- d.
Demonstrate at least once per 18 months for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal absorbers is less than the value specified below when tested in accordance with Regulatory Guide 1.52, Revision 2, at the system flowrate specified below +/- 10%.
ESF Ventilation System Delta P Flowrate Control Room Emergency Ventilation System - Filtration 6.6 in. W.G.
2000 cfm Control Room Emergency Ventilation System - Pressurization 3.6 in. W.G.
750 cfm Auxiliary/Fuel Building Emergency Exhaust 4.7 in. W.G.
6500 cfm (continued)
Wolf Creek - Unit 1 5.0-15 Amendment No. 123,131,139,164 l
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (VFTP) (continued)
- e.
Demonstrate at least once per 18 months that the heaters for each of the ESF systems dissipate the value specified below when tested in accordance with ANSI N510-1975.
ESF Ventilation System Wattage Control Room Emergency Ventilation System - Pressurization 5 +/- 1 kW Auxiliary/Fuel Building Emergency Exhaust 37 +/- 3 kW
- f.
Demonstrate at least once per 18 months for each of the ESF systems that following the creation of an artificial Delta P across the combined HEPA filters, the prefilters, and the charcoal absorbers of not less than the value specified below (dirty filter conditions), that the flowrate through these flow paths is with + 10% of the value specified below when tested in accordance with ANSI N510-1980.
ESF Ventilation System Delta P Flowrate Control Room Filtration System 6.6 in. W.G.
2000 cfm Control Room Pressurization System 3.6 in. W.G.
750 cfm Auxiliary/Fuel Building Emergency Exhaust 4.7 in. W.G.
6500 cfm The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.
5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks. The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP) ETSB 11-5, Revision 0, July 1981, "Postulated Radioactive Release due to Waste Gas System Leak or Failure." The liquid radwaste quantities shall be determined in accordance with Standard Review Plan, Revision 2, July 1981, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures."
(continued)
Wolf Creek - Unit 1 5.0-16 Amendment No. 123,131,164 l
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program (continued)
The program shall include:
- a.
The limits for concentrations of hydrogen and oxygen in the Waste Gas Holdup System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion);
- b.
A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than the amount that would result in a whole body exposure of 2 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents; and
- c.
A surveillance program to ensure that the quantity of radioactivity contained in the following outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.
- a.
Reactor Makeup Water Storage Tank
- b.
Refueling Water Storage Tank
- c.
Condensate Storage Tank, and
- d.
Outside Temporary tanks, excluding demineralizer vessels and the liner being used to solidify radioactive waste.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.
5.5.13 Diesel Fuel Oil Testing Proaram A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The prograi shal i nclude sampling and testing requirements, and acceptance criteria, all' in accor6dgnce with applicable ASTM Standards. The purpose of the program isito establish the following:
(continued)
Wolf Creek - Unit 1 5.0-1 7 Amendment No. 42-3, 164 l
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Diesel Fuel Oil Testing Program (continued)
- a.
Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:
- 1.
an API gravity or an absolute specific gravity within limits,
- 2.
a flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and
- 3.
water and sediment content within the limits for ASTM 2D fuel oil;
- b.
Other properties for ASTM 2D fuel oil are analyzed within 31 days following sampling and addition to storage tanks; and
- c.
Total particulate concentration of the fuel oil is < 10 mg/l when tested every 31 days in accordance with ASTM D-2276, Method A.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies.
5.5.14 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.
- a.
Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
- b.
Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:
- 1.
a change in the TS incorporated in the license; or 2..
a change to the USAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.
- c.
The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the USAR.
(continued)
Wolf Creek - Unit 1 5.0-1 8 Amendment No. 123, 138, 164 l
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Technical Specifications (TS) Bases Control Program (continued)
- d.
Proposed changes that meet the criteria of Specification 5.5.14b above shall be reviewed and approved by the NRC prior to implementation.
Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71 (e).
5.5.15 Safety Function Determination Program (SFDP)
This program ensures loss of safety function Is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following:
- a.
Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;
- b.
Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;
- c.
Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and
- d.
Other appropriate limitations and remedial or compensatory actions.
A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident ar aly s cannot be performed. For the purpose of this program, a loss of safety functikn may exist when a support system is inoperable, and:
- a.
A required system redundant to the sysem(s) supported by the inoperable support system is also inoperable; or
- b.
A required system redundant to the fy Em(s) in turn supported by the inoperable supported system is also inoperable; or
- c.
A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.
(continued)
Wolf Creek - Unit 1 5.0-1 9 Amendment No. 423, 164 l
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP) (continued)
The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.
5.5.16 Containment Leakage Rate Testing Program
- a.
A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program,' dated September 1995, as modified by the following exceptions:
- 1.
The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section Xl Code, Subsection IWL, except where relief has been authorized by the NRC.
- 2.
The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section Xl Code, Subsection IWE, except where relief has been authorized by the NRC.
- b.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 48 psig.
- c.
The maximum allowable containment leakage rate, La, at P8, shall be 0.20% of containment air weight per day.
- d.
Leakage rate acceptance criteria are:
- 1.
Containment leakage rate acceptance criterion is
- 1.0 La During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are c 0.60 La for the Type B and Type C tests and *0.75 La for Type A tests; (continued)
Wolf Creek - Unit 1 5.0-20 Amendment No. 123,142, 152, 164
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 Containment Leakage Rate Testing Program (continued)
- 2.
Air lock testing acceptance criteria are:
a)
Overall air lock leakage rate is < 0.05 La when tested at 2 Pa.
b)
For each door, leakage rate is < 0.005 La when pressurized to 2 10 psig.
- e.
The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.
- f.
The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.
5.5.17 Reactor Vessel Head Closure Bolt Integrity This program provides the requirements to support normal plant operation with one reactor vessel head closure bolt less than fully tensioned for one operating cycle. The provisions of this program shall be implemented when a head closure bolt becomes stuck In a partially inserted position such that the amount of thread engagement is not sufficient to take the tensioning loads without damage to the vessel threads or a bolt is not capable of being inserted into the bolt hole.
Prior to operation with one reactor vessel head closure bolt less than fully tensioned, the following conditions shall apply:
- a.
The circumstances associated with the less than fully tensioned closure bolt will be verified to be bounded by the analysis that was referenced in the letter dated September 15, 2000 (WO 00-0036).
- b.
A review of the results of the visual examinations performed on the closure bolts shall be performed to ensure that there is no indication of sufficient degradation of closure bolts that could affect the conclusions of Specification 5.5.17a. above.
Within 30 days following startup of the plant, a report shall be submitted to the Commission identifying the circumstances for operation with one reactor vessel head closure bolt less than fully tensioned.
Operation with the same reactor vessel head closure bolt less than fully tensioned shall be limited to one operating cycle (i.e., until the next refueling outage).
Wolf Creek - Unit I 5.0-21 Amendment No. 123,142452, 164 l
Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.
5.6.1 Not Used.
5.6.2 Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 1 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.
The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in a format similar to the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.
5.6.3 Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit during the previous year shall be submitted prior to May I of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.
The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.
5.6.4 Not Used.
(continued)
Wolf Creek - Unit 1 5.0-22 Amendment No. 123, 112, 111, 158, 164
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR)
- a.
Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
- 1.
Specification 3.1.3: Moderator Temperature Coefficient (MTC),
- 2.
Specification 3.1.5: Shutdown Bank Insertion Limits,
- 3.
Specification 3.1.6: Control Bank Insertion Limits,
- 4.
Specification 3.2.3: Axial Flux Difference,
- 5.
Specification 3.2.1: Heat Flux Hot Channel Factor, FQ(Z),
- 6.
Specification 3.2.2: Nuclear Enthalpy Rise Hot Channel Factor (FXH),
- 7.
Specification 3.9.1: Boron Concentration,
- 8.
SHUTDOWN MARGIN for Specification 3.1.1 and 3.1.4, 3.1.5, 3.1.6, and 3.1.8,
- 9.
Specification 3.3.1: Overtemperature AT and Overpower AT Trip Setpoints,
- 10.
Specification 3.4.1: Reactor Coolant System pressure, temperature, and flow DNB limits, and
- 11.
Specification 2.1.1: Reactor Core Safety Limits.
- b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 1.
WCNOC Topical Report TR 90-0025 WO1, "Core Thermal Hydraulic Analysis Methodology for the Wolf Creek Generating Station."
- 2.
WCAP-1 1397-P-A, -Revised Thermal Design Procedure."
- 3.
WCNOC Topical Report NSAG-006, "Transient Analysis Methodology for the Wolf Creek Generating Station."
(continued)
Wolf Creek - Unit 1 5.0-23 Amendment No. 123, 142,144, 158,
-145,164
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 4.
WCAP-1 0216-P-A, "Relaxation of Constant Axial Offset Control -
FQ Surveillance Technical Specification."
- 5.
WCNOC Topical Report NSAG-007, "Reload Safety Evaluation Methodology for the Wolf Creek Generating Station."
- 6.
NRC Safety Evaluation Report dated March 30, 1993, for the "Revision to Technical Specification for Cycle 7."
- 7.
WCAP-10266-P-A, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code."
- 8.
WCAP-1 1596-P-A, "Qualification of the Phoenix-P/ANC Nuclear Design System for Pressurized Water Reactor Cores."
- 9.
WCAP 10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code."
- 10.
WCAP-1 2610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report."
- 11.
WCAP-8745-P-A, "Design Bases for the Thermal Power AT and Thermal Overtemperature AT Trip Functions."
- c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
- d.
The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
(continued)
Wolf Creek - Unit 1 5.0-24 Amendment No. 123, 142,144,158, l 164 1
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
- a.
RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, hydrostatic testing, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
- 1.
Specification 3.4.3, "RCS Pressure and Temperature (P/T) Limits,"
and
- 2.
Specification 3.4.12, 'Low Temperature Overpressure Protection System."
- b.
The analytical methods used to determine the RCS pressure and temperature and Cold Overpressure Mitigation System limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 1.
NRC letter dated December 2, 1999, 'Wolf Creek Generating Station, Acceptance for Referencing of Pressure Temperature Limits Report (TAC No. MA4572)," and
- 2.
WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," January, 1996.
- c.
The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.
5.6.7 Not Used.
5.6.8 PAM Report When a report is required by Condition B or F of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate m'ethod of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
5.6.9 Not Used.
(continued)
Wolf Creek - Unit 1 5.0-25 Amendment No. 123,130, 142, 157, l 4&9,164 l
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.10 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, Steam Generator (SG) Program. The report shall include:
- a.
The scope of inspections performed on each SG;
- b.
Active degradation mechanisms found;
- c.
Nondestructive examination techniques utilized for each degradation mechanism;
- d.
Location, orientation (if linear), and measured sizes (if available) of service induced indications;
- e.
Number of tubes plugged during the inspection outage for each active degradation mechanism;
- f.
Total number and percentage of tubes plugged to date; and
- g.
The results of condition monitoring, including the results of tube pulls and in-situ testing.
Wolf Creek - Unit I 5.0-26 Amendment No. 123,142,158, 164
High Radiation Area 5.7 5.0 ADMINISTRATIVE CONTROLS 5.7 High Radiation Area As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601 (a) and (b) of 10 CFR Part 20:
5.7.1 Hich Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation:
- a.
Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.
- b.
Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.
- c.
Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.
- d.
Each individual or group entering such an area shall possess:
- 1.
A radiation monitoring device that continuously displays radiation dose rates in the area; or
- 2.
A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or
- 3.
A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or
- 4.
A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (continued)
Wolf Creek - Unit 1 5.0-27 Amendment No. 123, 142, 158, 15,l 164
High Radiation Area 5.7 5.7 High Radiation Area 5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation: (continued)
(i)
Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii)
Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, or personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.
- e.
Except for individuals qualified in radiation protection procedures, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them.
5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation:
- a.
Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition:
- 1.
All such door and gate keys shall be maintained under the administrative control of the Shift Manager/Control Room Supervisor or health physics supervision, or his or her designee.
- 2.
Doors and gates shall remain locked except during periods of personnel or equipment entry or exit.
- b.
Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.
(continued)
Wolf Creek - Unit 1 5.0-28 Amendment No. 123,142, 158, 164 l
High Radiation Area 5.7 5.7 High Radiation Area 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation: (continued)
- c.
Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.
- d.
Each individual or group entering such an area shall possess:
- 1.
A radiation monitoring device that continuously integrates the radiation rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or
- 2.
A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or
- 3.
A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i)
Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii)
Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area, or
- 3.
In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation monitoring device that continuously displays radiation dose rates in the area.
(continued)
Wolf Creek - Unit 1 5.0-29 Amendment No. 123,142,-158, 164 l
High Radiation Area 5.7 5.7 High Radiation Area 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation: (continued)
- e.
Except for individuals qualified in radiation protection procedures or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them.
- f.
Such individual areas that are within a larger area, such as PWR containment, where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.
Wolf Creek - Unit 1 5.0-30 Amendment No. 423,142,-158, 164 l