WO 18-0044, Clean Revised Technical Specification Pages for License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption of Alternative Source Term

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Clean Revised Technical Specification Pages for License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption of Alternative Source Term
ML18325A145
Person / Time
Site: Wolf Creek 
Issue date: 11/15/2018
From: Mccoy J
Wolf Creek
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
WO 18-0044
Download: ML18325A145 (38)
v • d • e


Text

Jaime H. McCoy Site Vice President W _--i"~ LF CREEK

~- -~'y,

&~;

  • ,'NUCLEAR OPERATING CORPORATION November 15, 2018 WO 18-0044 U.S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, DC 20555

Reference:

1)

Letter ET 17-0001, dated January 17, 2017, from J. H. McCoy, WCNOC, to USNRC

Subject:

2)

Letter ET 17-0011, dated May 4, 2017, from J. H. McCoy, WCNOC, to USNRC

3)

Letter WO 18-0004, dated January 15, 2018, from C. 0. Reasoner, WCNOC, to USNRC

4)

Letter ET 18-0018, dated June 19, 2018, from J. H. McCoy, WCNOC, to USNRC Docket No. 50-482:

Clean Revised Technical Specification Pages for License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption ofAlternative Source Term To Whom It May Concern:

Reference 1 provided the Wolf Creek Nuclear Operating Corporation (WCNOC) application to revise the Wolf Creek Generating Station (WCGS) Technical Specifications (TS). The proposed amendment would support transition to the Westinghouse Core Design and Safety Analysis methodologies. In addition, the license amendment request (LAR) included revising the WCGS licensing basis by adopting the Alternative Source Term radiological analysis methodology in accordance with 10 CFR 50.67, "Accident Source Term." As part of the original LAR, a number of revised TS pages were submitted. Since that time, the Nuclear Regulatory Commission (NRC) staff has provided a number of requests for additional information (RAls) related to this LAR. References 2, 3, and 4 provided WCNOC's responses to RAls which included revised TS pages. In some cases, these revised TS pages were revising TS pages that were previously submitted.

The intent of this submittal is to provide all of the clean, revised TS pages associated with this LAR in their most current form. The enclosure to this submittal provides these TS pages.

P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HCNET

WO 18-0044 Page 2 of 3 The information provided in this submittal does not expand the scope of the application and does not impact the no *significant hazards consideration determination presented in Reference

1.

In accordance with 1 O CFR 50.91, "Notice for public comment; State consultation," a copy of this submittal is being provided to the designated Kansas State official.

This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4156, or Cynthia R. Hafenstine at (620) 364-4204.

Sincerely, Jaime H. McCoy JHM/rlt

Enclosure:

Wolf Creek Westinghouse Methodology and AST LAR Revised Technical Specification Pages cc:

K. M. Kennedy (NRC), w/e B. K. Singal (NRC), w/e K. S. Steves (KDHE), w/e N. H. Taylor (NRC), w/e Senior Resident Inspector (NRC), w/e

WO 18-0044 Page 3 of 3 STATE OF KANSAS

) ) ss COUNTY OF COFFEY )

Jaime H. McCoy, of lawful age, being first duly sworn upon oath says that he is Site Vice President of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the contents thereof; that he has executed the same for and on behalf of said Corporation with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.

~~~l~

By

~

/7*

Jaime~Coy Site Vice President SUBSCRIBED and sworn to before me this /S'!j day of /Vofl.trnh-er

, 2018.

GAYLE SHEPHEARD My Appointment Expires July 24, 2019

'~~~

Notary ~blic Expiration Date __

!/'--~-t-f_,_f_;}.-_o_/--'q __

ENCLOSURE TO WO 18-0044 Wolf Creek Westinghouse Methodology and AST LAR Revised Technical Specification Pages

. (34 pages)

1.1 Definitions (continued)

CHANNEL CHECK CHANNEL OPERATIONAL TEST (COT)

CORE ALTERATION CORE OPERATING LIMITS REPORT (COLR)

DOSE EQUIVALENT 1-131 Wolf Creek - Unit 1 Definitions 1.1 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.

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 AL TERA TIONS 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 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."

( continued) 1.1-2 Amendment No. 123, 170,

1.1 Definitions (continued)

DOSE EQUIVALENT XE-133 ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME LEAKAGE Wolf Creek - Unit 1 Definitions 1.1 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (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-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 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

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, wherer 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 (continued) 1.1-3 Amendment No. 123, 131, 170,

1.1 Definitions (continued)

LEAKAGE (continued)

MASTER RELAY TEST MODE OPERABLE--OPERABILITY PHYSICS TESTS Wolf Creek - Unit 1 Definitions 1.1

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.

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 USAR;

b.

Authorized under the provisions of 10 CFR 50.59; or

c.

Otherwise approved by the Nuclear Regulatory Commission.

1.1-4 (continued)

Amendment No. 123, 164, 170,

'I I

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 (SOM)

Wolf Creek - Unit 1 Definitions 1.1 The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates and the power operated relief valve lift settings and the Low Temperature Overpressure Protection (L TOP) System arming temperature, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in ac,cordance 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 the methodology for verification have been previously reviewed and approved by the NRC.

SOM 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 SOM; and

b.

In MODES 1 and 2, the fuel and moderator temperatures are changed to the hot zero power temperatures.

1.1-5 (continued)

Amendment No. 123, 170, 180,

1.1 Definitions (continued)

SLAVE RELAY TEST STAGGERED TEST BASIS THERMAL POWER TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT)

Wolf Creek - Unit 1 Definitions 1.1 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 T ADOT may be performed by means of any series of sequential, overlapping, or total channel steps.

1.1-6 Amendment No. 123, 170, I

2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs SLs 2.0 In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded:

2.1.1.1 2.1.1.2 The departure from nucleate boiling ratio (DNBR) shall be maintained

~ 1.17 for the WRB-2 DNB correlation, and ~1.13 for the ABB-NV DNB correlation, and ~ 1.18 for the WLOP DNB.

The peak centerline temperature shall be maintained:::; 5080 °F, decreasing by 58 °F per 10,000 MWD/MTU of burnup.

2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained :::; 2735 psig.

2.2 SL Violatiqns 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 2.2.2.2 Wolf Creek - Unit 1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

In MODE 3, 4, or 5, restore compliance within 5 minutes.

2.0-1 Amendment No. 1 at!, 14 4

RCS Boron Limitations < S00°F 3.1.9 3.1 REACTIVITY CONTROL SYSTEMS 3.1.9 RCS Boron Limitations < 500°F LCO 3.1.9 The boron concentration of the Reactor Coolant System (RCS) shall be greater than the all rods out (ARO) critical boron concentration.

APPLICABILITY: MODE 2 with keff < 1.0 with any RCS cold leg temperature < S00°F and with Rod Control System capable of rod withdrawal, MODE 3 with any RCS cold leg temperature < 500°F and with Rod Control System capable of rod withdrawal, MODES 4 and 5 with Rod Control System capable of rod withdrawal.

ACTIONS CONDITION A.

RCS boron concentration not within limit.

Wolf Creek-Unit 1 REQUIRED ACTION COMPLETION TIME A.1 Initiate boration to restore Immediately RCS boron concentration to within limit.

A.2 Initiate action to place the Immediately Rod Control System in a condition incapable of rod withdrawal.

OR A. 3

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

Not applicable in MODES 4 and 5.

Initiate action to increase all RCS cold leg temperatures to~ S00°F, 3.1-21 Immediately Amendment No.

SURVEILLANCE REQUIREMENTS SURVEILLANCE

~

RCS Boron Limitations < 500°F 3.1.9 FREQUENCY SR 3.1.9.1 Verify RCS boron concentration is greater than the ARO critical boron concentration.

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Wolf Creek - Unit 1 3.1-22 Amendment No.

ACTIONS (continued)

CONDITION P.

One or more Turbine Stop Valve Closure Turbine Trip channel(s) inoperable.

Q.

One train inoperable.

R.

One RTB train inoperable.

Wolf Creek - Unit 1 REQUIRED ACTION P.1 Place channel(s) in trip.

OR P.2 Reduce THERMAL POWER to < P-9.

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

One train may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing provided the other train is OPERABLE.

Q.1 Restore train to OPERABLE status.

OR Q.2 Be in MODE 3.

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

One train may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing, provided the other train is OPERABLE.

R.1 Restore train to OPERABLE status.

OR R.2 Be in MODE 3.

RTS Instrumentation 3.3.1 COMPLETION TIME 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 76 hours 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 30 hours 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 30 hours (continued) 3.3-7 Amendment No. 123, 156, I

/

ACTIONS (continued)

CONDITION S.

One or more required S.1 channel(s) inoperable.

OR S.2 T.

One or more required T.1 channel(s) or train inoperable.

OR T.2 U.

One trip mechanism U.1 inoperable for one RTB.

OR U.2 Wolf Creek - Unit 1 REQUIRED ACTION

' Verify interlock is in required state for existing unit conditions.

Be in MODE 3.

Verify interlock is in required state for existing unit conditions.

Be in MODE 2.

Restore inoperable trip mechanism to OPERABLE status.

Be in MODE 3.

RTS Instrumentation 3.3.1 COMPLETION TIME 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 7 hours 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 7 hours 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 54 hours (continued) 3.3-8 Amendment No. 123, 156, I

ACTIONS ( continued)

CONDITION V.

One channel inoperable.

w. One channel inoperable.

Wolf Creek - Unit 1 REQUIRED ACTION

NOTE--------, ----------

The inoperable channel may be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing of other channels.

V.1 Place channel in trip.

OR V.2.1 B in MODE 2 with kett <

1.0.

AND V.2.2.1 Initiate action to fully insert all rods.

AND V.2.2.2 Initiate action to_ place the Rod Control System in a condition incapable of rod withdrawal.

OR V.2.3 Initiate action to borate the RCS to greater than all rods out (ARO) critical boron concentration.

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

The inoperable channel may be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing of other channels.

W.1 Place channel in trip.

3.3-9 RTS Instrumentation 3.3.1 COMPLETION TIME 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 78 hours 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> 78 hours 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> 72 hours (continued)

Amendment No. ~.

ACTIONS (continued)

CONDITION X.

Required Action and associated Completion Time of Condition W not met.

OR Two or more channels inoperable.

SURVEILLANCE REQUIREMENTS REQUIRED ACTION X.1.1 Initiate action to fully insert*

all rods.

AND X.1.2 Initiate action to place the Rod Control System in a condition incapable of rod withdrawal.

OR X.2 Initiate action to borate the RCS to greater than all rods out (ARO) critical boron concentration.

RTS Instrumentation 3.3.1 COMPLETION TIME Immediately Immediately Immediately

. -------------------NOTE---------------------------------------------------------------

Ref er to Table 3.3.1-1 to determine which SRs apply for each RTS Function.

SR 3.3.1.1 SR 3.3.1.2 Wolf Creek - Unit 1 SURVEILLANCE Perform CHANNEL CHECK.

NOT ES--------------------------------

Not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is ~ 15% RTP.

Compare results of calorimetric heat balance calculation to power range channel output. Adjust power range channel output if calorimetric heat balance calculation results exceed power range channel output by more than + 2% RTP.

FREQUENCY 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 24 hours 3.3-10 (continued)

Amendment No. 123, 148, 156, 188,

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.1.3 SR 3.3.1.4 SR 3.3.1.5 SR,3.3.1.6 SR 3.3.1.7 Wolf Creek - Unit 1 SURVEILLANCE

NO TES-------------------------------

Not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is~ 50% RTP.

Compare results of the incore detector measurements to Nuclear Instrumentation System (NIS) AFD. Adjust NIS channel if absolute difference is~3%.

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

Th is Surveillance must be performed on the reactor trip bypass breaker for the local manual shunt trip only prior to placing the bypass breaker in service.

Perform TADOT.

Perform ACTUATION LOGIC TEST.

*----------NOTE--------------------------------

Not required to be performed until 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after achieving equilibrium conditions with THERMAL POWER~ 75 % RTP.

Calibrate excore channels to agree with incore detector measurements.

* ---NOTES-------------------------------

1.

Not required to be performed for source range instrumentation prior to entering MODE 3 from MODE 2 until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entry into MODE 3.

2.

Source range instrumentation shall include verification that interlocks P-6 and P-10 are in their required state for existing unit conditions.

Perform COT.

FREQUENCY 31 effective full power days (EFPD) 62 days on a STAGGERED TEST BASIS 92 days on a STAGGERED TEST BASIS 92 EFPD 184 days 3.3-11 (continued)

Amendment No. 123, 156, 188,

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 6)

Reactor Trip System Instrumentation FUNCTION

1. Manual Reactor Trip
2. Power Range Neutron Flux
a.

High

b.

Low

3.

Power Range Neutron Flux Rate

a.

High Positive Rate

b.

High Negative Rate

4.

Intermediate Range Neutron Flux APPLICABLE MODES OR OTHER SPECIFIED REQUIRED CONDITIONS CHANNELS 1,2 2

2 1,2 4

4 4

1,2 4

1,2 4

2 SURVEILLANCE CONDITIONS REQUIREMENTS B

SR 3.3.1.14 C

D V

w.x E

E F,G SR 3.3.1.14 SR 3.3.1.1 SR 3.3.1.2 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.16 SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.16 SR3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.16 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.16 SR 3.3.1.7 SR3.3.1.11 SR 3.3.1.16 SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 (a)

The Allowable Value defines the Limiting Safety System Setting. See the Bases for the Trip Setpoints.

(b)

With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(c)

Below the P-10 (Power Range Neutron Flux) interlock.

(d)

Above the P-6 (Intermediate Range Neutron Flux) interlock.

(e)

Below the P-6 (Intermediate Range Neutron Flux) interlock (f)

With k.ff <!: 1.0.

ALLOWABLE VALUE (a)

NA NA s 112.3% RTP S28.3% RTP s28.3% RTP S6.3% RTP with time constant

2 sec s6.3% RTP with time constant
2 sec s 35.3% RTP (continued)

(h)

With keff < 1.0, and all RCS cold leg temperatures<!: 500° F, and RCS boron concentration s the rods out (ARO) critical boron concentration, and Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(i)

With all RCS cold leg termperatures <!: 500° F, and RCS boron concentration s the ARO critical boron concentration, and Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

Wolf Creek - Unit 1 3.3-15 Amendment No. 123, 131, 132, 165,

5.
6.
7.
8.
9.
10.

(a)

(b)

(e)

(g)

Table 3.3.1-1 (page 2 of 6)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED FUNCTION CONDITIONS CHANNELS CONDITIONS Source Range Neutron 2(e) 2 l,J Flux 3(b), 4(b), 5(b) 2 J,K Overtemperature t. T 1,2 4

E Overpower t. T 1,2 4

E Pressurizer Pressure

a.

Low 1(9) 4 M

b.

High 1,2 4

E Pressurizer Water 1(9) 3 M

Level - High Reactor Coolant Flow -

1(9) 3 per loop M

Low RTS Instrumentation 3.3.1 SURVEILLANCE ALLOWABLE REQUIREMENTS VALUE(a)

SR 3.3.1.1

~ 1.6 ES cps SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.1

~ 1.6 ES cps SR 3.3.1.7 SR3.3.1.11 SR 3.3.1.1 Refer to Note 1 SR 3.3.1.3 (Page 3.3-19)

SR 3.3.1.6 SR 3.3.1.7 SR 3.3.1.10 SR3.3.1.16 SR 3.3.1.1 Refer to SR 3.3.1.7 Note 2 SR 3.3.1.10 (Page SR 3.3.1.16 3.3-20)

SR3.3.1.1

~ 1930 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16 SR 3.3.1.1

~2395 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16 SR 3.3.1.1

~ 93.9% of SR 3.3.1.7 instrument span SR 3.3.1.10 SR 3.3.1.1

~ 88.9% of SR 3.3.1.7 normalized flow SR 3.3.1.10 SR 3.3.1.16 (continued)

The Allowable Value defines the Limiting Safety System Setting. See the Bases for the Trip Setpoints.

With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

Below the P-6 (Intermediate Range Neutron Flux) interlock.

Above the P-7 (Low Power Reactor Trips Block) interlock.

Wolf Creek - Unit 1 3.3-16 Amendment No. 123, 140,

Table 3.3.1-1 (page 3 of 6)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED RTS Instrumentation 3.3.1 SURVEILLANCE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE (a)

11.

Not Used.

12.

Undervoltage RCPs

13.

Underfrequency.

RCPs

14.

Steam Generator (SG)

Water Level Low-Low (I)

15.

Not Used.

16.

Turbine Trip

a.

Low Fluid Oil Pressure

b.

Turbine Stop Valve Closure

17.

Safety Injection (SI)

Input from Engineered Safety Feature Actuation System (ESFAS)

18.

Reactor Trip System Interlocks

a.

Intermediate Range Neutron Flux, P-6

b.

Low Power Reactor Trips Block, P-7 C.

Power Range Neutron Flux, P-8 1,2 10l 10l 1,2 2(e) 2/bus 2/bus 4 per gen 3

4 2 trains 2

1 per train 4

M M

E 0

p Q

s T

T SR 3.3.1.9 SR 3.3.1.10 SR 3.3.1.16 SR 3.3.1.9 SR3.3.1.10 SR3.3.1.16 SR 3.3.1.1 SR 3.3.1.7 SR3.3.1.10 SR3.3.1.16 SR3.3.1.10 SR 3.3.1.15 SR 3.3.1.10 SR3.3.1.15 SR 3.3.1.14 SR 3.3.1.11 SR 3.3.1.13 SR 3.3.1.5 SR 3.3.1.11 SR 3.3.1.13 (a)

The Allowable Value defines the Limiting Safety System Setting. See the Bases for the Trip Setpoints.

(e)

Below the P-6 (Intermediate Range Neutron Flux) interlocks.

(g)

Above the P-7 (Low Power Reactor Trips Block) interlock.

(I)

The applicable MODES for these channels are more restrictive in Table 3.3.2-1. (See Function 6.d.)

0)

Above the P-9 (Power Range Neutron Flux) interlock.

<! 10355 Vac i 57.1 Hz

.:::,22.3% of Narrow Range Instrument Span

<! 534.20 psig

<! 1% open NA

<! 6E-11 amp NA

<;51.3% RTP contmueo Wolf Creek - Unit 1 3.3-17 Amendment No. 123, 132, I

ACTIONS (continued)

CONDITION D.

Required Action and D.1 associated Completion Time for Condition A, B AND or C not met in MODE 1, 2, 3, or 4.

D.2 E.

Required Action and E.1 associated Completion Time for Condition A, B or C not met during AND movement of irradiated fuel assemblies or during E.2 CORE ALTERATIONS.

SURVEILLANCE REQUIREMENTS CREVS Actuation Instrumentation 3.3.7 REQUIRED ACTION COMPLETION TIME Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Suspend CORE Immediately ALTERATIONS.

Suspend movement of Immediately irradiated fuel assemblies.

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

R ef er to Table 3.3.7-1 to determine which SRs apply for each CREVS Actuation Function.

SURVEILLANCE FREQUENCY SR 3.3.7.1 Perform CHANNEL CHECK.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.3.7.2 Perform COT.

92 days (continued)

Wolf Creek - Unit 1 3.3-51 Amendment No. 123,183,200,

CREVS Actuation Instrumentation 3.3.7 SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.7.3 SR 3.3.7.4 SR 3.3.7.5 SR 3.3.7.6 Wolf Creek - Unit 1 SURVEILLANCE

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

T he continuity check may be excluded.

Perform ACTUATION LOGIC TEST.

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

Verification of setpoint is not required.

Perform TADOT.

Perform CHANNEL CALIBRATION.

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

Radiation monitor detectors are excluded from response time testing.

Verify Control Room Ventilation Isolation ESF RESPONSE TIMES are within limits.

FREQUENCY 31 days on a STAGGERED TEST BASIS 18 months 18 months 18 months on a STAGGERED TEST BASIS 3.3-52 Amendment No. 123, 183,

FUNCTION

1.

Manual Initiation

2.

Automatic Actuation Logic and Actuation Relays (BOP ESFAS)

3.

Control Room Radiation-Control Room Air Intakes

4.

Containment Isolation -

Phase A CREVS Actuation Instrumentation 3.3.7 Table 3.3.7-1 (page 1 of 1)

CREVS Actuation Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE CONDITIONS CHANNELS REQUIREMENTS TRIP SETPOINT 1, 2, 3, 4, 2

SR 3.3.7.4 NA (a) and (c) 1, 2, 3, 4, 2 trains SR 3.3.7.3 NA (a) and (c)

SR 3.3.7.6 1, 2, 3, 4, 2

SR 3.3.7.1 (b)

(a) and (c)

SR 3.3.7.2 SR 3.3.7.5 SR 3.3.7.6 Refer to LCO 3.3.2, "ES FAS Instrumentation," Function 3.a, for all initiation functions and requirements.

(a)

During movement of irradiated fuel assemblies.

(b)

Trip Setpoint concentration value (µCi/cm3) is to be established such that the actual submersion dose rate would not exceed 2 mR/hr in the control room.

(c)

During CORE ALTERATIONS.

Wolf Creek - Unit 1 3.3-53 Amendment No. 123, 132,183,200,

RCS Pressure, Temperature and Flow DNB Limits

ACTIONS RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below:

a.

Pressurizer pressure is greater than or equal to the limit specified in the COLR;

b.

RCS average temperature is less than or equal to the limit specified in the COLR; and

c.

RCS total flow rate ~ 361,200 gpm and greater than or equal to the limit specified in the COLR.

MODE 1.

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

P ress u rize r pressure limit does not apply during :

a. THERMAL POWER ramp > 5% RTP per minute; or
b. THERMAL POWER step> 10% RTP.

CONDITION REQUIRED ACTION COMPLETION TIME A.

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

A.1 Not applicable to RCS total flow rate.

One or more RCS DNB parameters not within limits.

Wolf Creek - Unit 1 Restore RCS DNB parameter(s) to within limit.

3.4-1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (continued)

Amendment No. 123, 144,

RCS Pressure, Temperature and Flow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.1.2 SR 3.4.1.3 SR 3.4.1.4 Wolf Creek - Unit 1 SURVEILLANCE FREQUENCY Verify RCS average temperature is less than or equal 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to the limit specified in the COLR.

Verify RCS total flow rate is ~ 361,200 gpm and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> greater than or equal to the limit specified in the COLR.

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

N ot required to be performed until 7 days after

~ 95% RTP.

Verify by precision heat balance that RCS total flow rate is ~ 361,200 gpm and greater than or equal to the limit specified in the COLR.

18 months 3.4-4 Amendment No. 123, 144,

RCS Specific Activity 3.4.16 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

s; 60 µCi/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 1-131 to within limit.

B.

Required Action and 8.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> I

associated Completion Time of Condition A not AND met.

8.2 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 XE-133 not within limit

'OR I

DOSE EQUIVALENT 1-131 I

> 60 µCi/gm.

Wolf Creek - Unit 1 3.4-42 Amendment No. 123, 155,170,212,

SURVEILLANCE REQUIREMENTS SR 3.4.16.1 SR 3.4.16.2 SURVEILLANCE

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

0 n I y required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT XE-133 specific activity.~ 500 µCi/gm.

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

0 n ly required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity ~ 1.0 µCi/gm.

RCS Specific Activity 3.4.16 FREQUENCY 7 days 14 days 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~ 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 3.4-43 Amendment No. 123, 170,212, I

MSSVs 3.7.1 Table 3.7.1-1 (page 1 of 1)

OPERABLE Main Steam Safety Valves versus Maximum Allowable Power NUMBER OF OPERABLE MSSVs PER STEAM GENERATOR Wolf Creek - Unit 1 4

3 2

3.7-3 MAXIMUM ALLOWABLE POWER

(% RTP) 70 51 31 Amendment No. ~.

CREVS 3.7.10

3. 7 PLANT SYSTEMS 3.7.10 Control Room Emergency Ventilation System (CREVS)

LCO 3.7.10 Two CREVS trains shall be OPERABLE.

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

The control room envelope (CRE) and control building envelope (CBE) boundaries may be opened intermittently under administrative controls that ensure the building boundary can be closed consistent with the safety analysis.

APPLICABILITY:

MODES 1, 2, 3, and 4, During CORE AL TERA TIONS During movement of irradiated fuel assemblies.

ACTIONS CONDITION A.

One CREVS train A.1 inoperable for reasons other than Condition B.

B.

One or more CREVS trains B.1 inoperable due to an inoperable CRE boundary or an inoperable CBE boundary in MODES 1, 2, AND 3, or 4.

B.2 AND B.3 Wolf Creek - Unit 1 REQUIRED ACTION COMPLETION TIME Restore CREVS train to 7 days OPERABLE status.

Initiate action to Immediately implement mitigating actions.

Verify mitigating actions to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ensure CRE occupant radiological exposures will not exceed limits and CRE occupants are protected from chemical and smoke hazards.

Restore CRE boundary 90 days and CBE boundary to OPERABLE status.

3.7-26 (continued)

Amendment No. 123,134,171, 177, 179,184,200,

ACTIONS (continued)

CONDITION C.

Required Action and associated Completion Time of Condition A or B not met in MODE 1, 2, 3, or 4.

D.

Required Action and associated Completion Time of Condition A not met during movement of irradiated fuel assemblies or during CORE ALTERATIONS.

E.

Two CREVS trains inoperable during movement of irradiated fuel assemblies or during CORE ALTERATIONS.

OR One or more CREVS trains inoperable due to an inoperable CRE boundary or an inoperable CBE boundary during movement of irradiated fuel assemblies or during CORE ALTERATIONS.

Wolf Creek - Unit 1 REQUIRED ACTION C.1 Be in MODE 3.

AND C.2 Be in MODE 5.

D.1 Place OPERABLE CREVS train in CRVIS mode.

OR D.2.1 Suspend CORE ALTERATIONS.

AND D.2.2 Suspend movement of irradiated fuel assemblies.

E.1 Suspend CORE ALTERATIONS.

AND E.2 Suspend movement of irradiated fuel assemblies.

CREVS 3.7.10 COMPLETION TIME 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours Immediately Immediately.

Immediately Immediately Immediately (continued) 3.7-27 Amendment No. 123,131, 134, 171, 177, 179, 184, 200,

EES 3.7.13 3.7 PLANT SYSTEMS 3.7.13 Emergency Exhaust System (EES)

LCO 3.7.13 Two EES trains shall be OPERABLE.

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

The auxiliary building or fuel building boundary may be opened intermittently under administrative controls that ensure the building boundary can be closed consistent with the safety analysis.

APPLICABILITY:

MODES 1, 2, 3, and 4, During movement of irradiated fuel assemblies in the fuel building.

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

T he SIS mode of operation is required only in MODES 1, 2, 3, and 4. The FBVIS mode of operation is required only during movement of irradiated fuel assemblies in the fuel building.

ACTIONS

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

L CO 3.0.3 is not applicable to the FBVIS mode of operation.

. CONDITION REQUIRED ACTION COMPLETION TIME A.

One EES train inoperable.

A.1 Restore EES train to 7 days OPERABLE status.

B.

Two EES trains inoperable 8.1 Initiate actions to Immediately due. to inoperable auxiliary implement mitigating building boundary in actions.

MODE.1, 2, 3, or 4.

AND (continued)

Wolf Creek - Unit 1 3.7-33 Amendment No.123, 132,134, 171, 177, 184,200,

ACTIONS (continued)

CONDITION B.

( continued)

B.2 AND B.3 C.

Required Action and C.1 associated Completion Time of Condition A or B AND not met in MODE 1, 2, 3, or 4.

C.2 OR Two EES trains inoperable in MODE 1, 2, 3, or 4 for reasons other than Condition B.

D.

Required Action and D.1 associated Completion Time of Condition A not met during movement of irradiated fuel assemblies OR in the fuel building.

D.2 Wolf Creek - Unit 1 REQUIRED ACTION Verify mitigating actions ensure main control room occupants do not exceed 10 CFR 50 Appendix A GDC 19 limits.

Restore building boundary.

to OPERABLE status.

Be in MODE 3.

Be in MODE 5.

Place OPERABLE EES train in operation in FBVIS mode.

Suspend movement of irradiated fuel assemblies in the fuel building.

EES 3.7.13 COMPLETION TIME 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 24 hours 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours Immediately Immediately (continued) 3.7-34 Amendment No. 123,132,134,171,177, 184,200,

ACTIONS (continued)

CONDITION REQUIRED ACTION EES 3.7.13 COMPLETION TIME E.

Two EES trains inoperable E.1 Suspend movement of Immediately for reasons other than Condition B during movement of irradiated fuel assemblies in the fuel building.

SURVEILLANCE REQUIREMENTS irradiated fuel assemblies in the fuel building.

SURVEILLANCE FREQUENCY SR 3.7.13.1 Operate each EES train for~ 15 continuous minutes 31 days with the heaters operating.

SR 3.7.13.2 Perform required EES filter testing in accordance with In accordance with the Ventilation Filter Testing Program (VFTP).

the VFTP SR 3.7.13.3 Verify each EES train actuates on an actual or 18 months simulated actuation signal.

(continued)

Wolf Creek - Unit 1

3. 7-35 Amendment No. 123, 132, 134, 171, 177, 184,208,

Containment Penetrations 3.9.4 3.9 REFUELING OPERATIONS 3.9.4 Containment Penetrations LCO 3.9.4 APPLICABILITY:

Wolf Creek - Unit 1 The containment penetrations shall be in the following status:

a.

The equipment hatch closed and held in place by four bolts, or if open, capable of being closed;

b.

One door in the emergency air lock closed and one door in the personnel air lock capable of being closed; and

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

An emergency personnel escape air lock temporary closure device is an acceptable replacement for an emergency air lock door.

c.

Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:

1.

closed by a manual or automatic isolation valve, blind flange, or equivalent, or

2.

capable of being closed by an OPERABLE Containment Purge Isolation valve.

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

Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls that ensure the building boundary can be closed consistent with the safety analysis.

During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment.

3.9-5 Amendment No. 123,135,146,

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 5.5.13 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 ~ 0.1 rem TEDE 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.

Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

(continued)

Wolf Creek - Unit 1 5.0-17 Amendment No. 123, 164,

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.18 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (OBA) conditions without personnel receiving radiation exposures in excess of 5 rem TEDE for the duration of the accident. The program shall include the following elements:

a.

The definition of the CRE, CRE boundary, control building envelope (CBE), and CBE boundary.

b.

Requirements for maintaining the CRE and CBE boundary in their design condition including configuration control and preventive maintenance.

c.

Requirements for (i) determining the unfiltered air inleakage past the CRE and CBE boundaries in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

The following are exceptions to Section C.1 and C.2 of Regulatory Guide 1.197, Revision 0:

1.

The Tracer Gas Test based on the Brookhaven National Laboratory Atmospheric Tracer Depletion (ATD) Method is used to determine the unfiltered air inleakage past the CRE and CBE boundaries. The ATD Method is described in WCNOC letters '

dated February 21, 2005 (WO 05-0003), June 29, 2007 (WM 07-0057), and September 28, 2007 (ET 07-0045).

d.

Measurement, at designated locations, of the CRE pressure relative to the outside atmosphere during the pressurization mode of operation by one train of the CREVS, operating at the flow rate required by the VFTP, at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the periodic assessment of the CRE boundary.

(continued)

Wolf Creek - Unit 1 5.0-22 Amendment No. 123, 142, 152, 164, 79,

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 ln_sertion 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, F0(Z),

6.

Specification 3.2.2: Nuclear Enthalpy Rise Hot Channel Factor (FiH),

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 Li T and Overpower Li T 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:

Wolf Creek - Unit 1

1.

WCAP-11397-P-A, "Revised Thermal Design Procedure."

2.

WCAP-10216-P-A, "Relaxation of Constant Axial Offset Control -

Fa Surveillance Technical Specification."

3.

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology."

5.0-25 (continued)

Amendment No. 123,142,144,158, 159, 164, 179,

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

4.

WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)."

5.

WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON."

6.

WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology."

7.

WCAP 10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code."

8.

WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report."

9.

WCAP-8745-P-A, "Design Bases for the Thermal Power LiT and Thermal Overtemperature Li T 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 5.0-26 Amendment No. 123,142,144,158, 164,179,209,213,

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