CP-202000321, (CPNPP) - License Amendment Request (LAR) 20-001 Technical Specification (TS) 1.1, Definitions - ESP & RTS Response Time

(CPNPP) - License Amendment Request (LAR)20-001 Technical Specification (TS) 1.1, Definitions - ESP & RTS Response Time

ML20184A064
Person / Time
Site:	Comanche Peak
Issue date:	07/02/2020
From:	Thomas McCool Luminant, Vistra Operations Company
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CP-202000321, TXX-20043
Download: ML20184A064 (19)

Text

II Thomas P. McCool Comanche Peak Site Vice President Nuclear Power Plant (Vistra Operations Company LLC)

Luminant P.O. Box 1002 6322 North FM 56 Glen Rose , TX 76043 T 254 .897 .6042 CP-202000321 TXX-20043 July 2, 2020 U. S. Nuclear Regulatory Commission Ref 10 CFR 50.90 ATTN: Document Control Desk 10 CFR 50.91(b)(l)

Washington, DC 20555-0001 10 CFR 50.92(c)

Subject:

Comanche Peak Nuclear Power Plant (CPNPP)

Docket Nos. 50-445 and 50-446 LICENSE AMENDMENT REQUEST (LAR)20-001 TECHNICAL SPECIFICATION (TS) 1.1, "DEFINITIONS- ESF AND RTS RESPONSE TIME"

Dear Sir or Madam:

Pursuant to 10 CFR 50.90 and 10 CFR 50.91, Vistra Operations Company LLC (Vistra OpCo) hereby requests an amendment to the Comanche Peak Nuclear Power Plant (CPNPP) Unit 1 and Unit 2 Technical Specifications.

Vistra OpCo requests adoption of TSTF-569, "Revise Response Time Testing Definition," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the CPNPP Unit 1 and Unit 2 Technical Specifications (TS). The proposed amendment revises the TS Definitions for Engineered Safety Feature (ESF) Response Time and Reactor Trip System (RTS) Response Time.

The enclosure provides a description and assessment of the proposed changes. Attachment 1 provides the existing TS pages marked to show the proposed changes. Attachment 2 provides revised (clean) TS pages. No change to the current Technical Specification Bases is required.

Vistra OpCo has determined that the proposed change does not involve a significant hazards consideration pursuant to 10 CFR 50.92(c), and there are no significant environmental impacts associated with the change. The CPNPP Station Operations Review Committee (SORC) has reviewed the proposed license amendment. In accordance with 10 CFR 50.91(b)(l), a copy of the proposed license amendment is being forwarded to the State of Texas.

NRC staff review and approval of the proposed license amendment is requested within one year of the NRC acceptance date. Once approved, the amendment shall be implemented within 60 days.

This letter contains no new regulatory commitments regarding CPNPP Units 1 and 2.

Should you have any questions, please contact Garry W Struble at (254) 897-6628 or garry.struble@luminant.com.

TXX-20043 Page 2 of 2 I state under penalty of perjury that the foregoing is true and correct. Executed on July 2, 2020.

Enclosure:

LICENSE AMENDMENT REQUEST (LAR)20-001 TECHNICAL SPECIFICATION (TS) 1.1, "DEFINITIONS - ESF AND RTS RESPONSE TIME" DESCRIPTION AND ASSESSMENT

Attachment:

1. PROPOSED TECHNICAL SPECIFICATION CHANGES (MARKUP)

2. REVISED TECHNICAL SPECIFICATION CHANGES c (email) - Scott Morris, Region IV [Scott.Morris@nrc.gov]

Dennis Galvin, NRR [Dennis.Galvin@nrc.gov]

John Ellegood, Senior Resident Inspector, CPNPP [John.Ellegood@nrc.gov]

Neil Day, Resident Inspector, CPNPP [Neil.Day@nrc.gov]

Mr. Robert Free [robert.free@dshs.state.tx.us]

Environmental Monitoring & Emergency Response Manager Texas Department of State Health Services Mail Code 1986 P.O. Box 149347 Austin, TX 78714-9347

Enclosure to TXX-20043 Page 1of3 COMANCHE PEAK NUCLEAR POWER PLANT LICENSE AMENDMENT 20-001, TS 1.1, "DEFINITIONS - ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME AND REACTOR TRIP SYSTEM (RTS) RESPONSE TIME" DESCRIPTION AND ASSESSMENT

1.0 DESCRIPTION

Vistra Operations Company LLC (Vistra OpCo) requests adoption of TSTF-569, "Revise Response Time Testing Definition," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the Comanche Peak Nuclear Power Plant (CPNPP)

Units 1 and 2 Technical Specifications (TS). The proposed amendment revises the TS Definitions for Engineered Safety Feature (ESF) Response Time and [Reactor Trip System (RTS) Response Time.

2.0 ASSESSEMENT 2.1 Applicability of Safety Evaluation Vistra OpCo has reviewed the safety evaluation for TSTF-569 provided to the Technical Specifications Task Force in a letter dated August 14, 2019. This review included a review of the NRC staff's evaluation, as well as the information provided in TSTF-569. As described herein, Vistra OpCo has concluded that the justifications presented in TSTF-569 and the safety evaluation prepared by the NRC staff are applicable to CPNPP Units 1 and 2 and justify this amendment for the incorporation of the changes to the CPNPP TS.

2.2 Variations Vistra OpCo is not proposing any variations from the TS changes described in the TSTF-569 or the applicable parts of the NRC staff's safety evaluation dated August 14, 2019.

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Determination Vistra OpCo requests adoption of TSTF-569, "Revise Response Time Testing Definition,"

which is an approved change to the Improved Standard Technical Specifications (ISTS), into the CPNPP Units 1 and 2 Technical Specifications (TS). The proposed amendment revises the TS Definitions for Engineered Safety Feature (ESF) Response Time and Reactor Trip System (RTS) Response Time.

Vistra OpCo has evaluated whether a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Enclosure to TXX-20043 Page 2 of 3 Response: No The proposed change revises the TS Definition of ESF and RTS instrumentation response time to permit the licensee to evaluate using an NRG-approved methodology and apply a bounding response time for some components in lieu of measurement. The requirement for the instrumentation to actuate within the response time assumed in the accident analysis is unaffected.

The response time associated with the ESF and RTS instrumentation is not an initiator of any accident. Therefore, the proposed change has no significant effect on the probability of any accident previously evaluated.

The affected ESF and RTS instrumentation are assumed to actuate their respective components within the required response time to mitigate accidents previously evaluated. Revising the TS definition for ESF and RTS instrumentation response times to allow an NRG-approved methodology for verifying response time for some components does not alter the surveillance requirements that verify the ESF and RTS instrumentation response times are within the required limits. As such, the TS will continue to assure that the ESF and RTS instrumentation actuate their associated components within the specified response time to accomplish the required safety functions assumed in the accident analyses. Therefore, the assumptions used in any accidents previously evaluated are unchanged and there is no significant increase in the consequences.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed change revises the TS Definition of ESF and RTS instrumentation response time to permit the licensee to evaluate using an NRG-approved methodology and apply a bounding response time for some components in lieu of measurement.

The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). The proposed change does not alter any assumptions made in the safety analyses. The proposed change does not alter the limiting conditions for operation for the ESF and RTS instrumentation, nor does it change the Surveillance Requirement to verify the ESF and RTS instrumentation response times are within the required limits. As such, the proposed change does not alter the operability requirements for the ESF and RTS instrumentation, and therefore, does not introduce any new failure modes.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No

Enclosure to TXX-20043 Page 3 of 3 The proposed change revises the TS Definition of ESF and RTS instrumentation response time to permit the licensee to evaluate using an NRG-approved methodology and apply a bounding response time for some components in lieu of measurement. The proposed change has no effect on the required ESF and RTS instrumentation response times or setpoints assumed in the safety analyses and the TS requirements to verify those response times and setpoints. The proposed change does not alter any Safety Limits or analytical limits in the safety analysis. The proposed change does not alter the TS operability requirements for the ESF and RTS instrumentation. The ESF and RTS instrumentation actuation of the required systems and components at the required setpoints and within the specified response times will continue to accomplish the design basis safety functions of the associated systems and components in the same manner as before. As such, the ESF and RTS instrumentation will continue to perform the required safety functions as assumed in the safety analyses for all previously evaluated accidents.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, the Vistra OpCo concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

3.2 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

4.0 EVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

Attachment 1 to TXX-20043 Definitions Page 1of7 1.1 1.0 USE AND APPLICATION 1.1 Definitions NOTE The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

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

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

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

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

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

COMANCHE PEAK UNITS 1 AND 2 1.1-1 Amendment No. 150

Attachment 1 to TXX-20043 Definitions Page 2 of7 1.1 1.1 Definitions (continued)

CHANNEL OPERATIONAL A COT shall be the injection of a simulated or actual signal into TEST (COT) 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 so 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 AL TE RATION CORE ALTERATl ON shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS The COLR is the unit specific document that provides cycle REPORT (COLR) specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1131 DOSE EQUIVALENT I 131 shall be that concentration of 1131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes I 131, I 132, I 133, I 134, and I 135 actually present. The determination of DOSE EQUIVALENT 1131 shall be performed using thyroid dose conversion factors from Table Ill of TIO 14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or from Table E 7 of Regulatory Guide 1.109, Revision 1, NRC, 1977, or from ICRP 30, 1979, Supplement to Part 1, page 192 212, Table titled "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity," or 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."

COMANCHE PEAK UNITS 1 AND 2 1.1-2 Amendment No. 150

Attachment 1 to TXX-20043 Definitions Page 3 of 7 1.1 1.1 Definitions (continued)

DOSE EQUIVALENT XE 133 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 effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposu re to Radionuclides in Air, Water, and Soil" , or using the dose conversion factors from Table B 1 of Regulatory Guide 1.109, Revision 1, NRC, 1977.

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from when FEATURE(ESF)RESPONSE the monitored parameter exceeds its ESF actuation setpoint at TIME 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 measu red by means of any series of sequential, overlapping , or total steps so that the entire response time is measured . In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the license program PROGRAM that fulfills the requ irements of 10 CFR 50.55a(f).

COMANCHE PEAK UNITS 1 AND 2 1.1-3 Amendment No. 150, 168

Attachment 1 to TXX-20043 Definitions Page 4 of7 1.1 1.1 Definitions (continued)

LEAKAGE 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;

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.

COMANCHE PEAK UNITS 1 AND 2 1.1-4 Amendment No . .:t-aQ.

Attachment 1 to TXX-20043 Definitions Page 5 of 7 1.1 1.1 Definitions (continued)

OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a. Described in Chapter 14, of the FSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

PRESSURE AND The PTLR is the unit specific document that provides the reactor TEMPERATURE LIMITS vessel pressure and temperature limits, including heatup and REPORT (PTLR) cooldown rates, the power operated relief valve (PORV) lift settings and the LTOP arming temperature associated with the Low Temperature Overpressurization Protection (LTOP) System, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.6.

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

RATEDTHERMAL POWER RTP shall be a total reactor core heat transfer rate to the reactor (RTP) coolant of 3458 MWt through Cycle 13 for Unit 1 and through Cycle 11 for Unit 2. Starting with Cycles 14 and 12 of Units 1 and 2, respectively, RTP shall be 3612 MWt.

COMANCHE PEAK UNITS 1 AND 2 1.1-5 Amendment No.~

Attachment 1 to TXX-20043 Defin itions Page 6 of 7 1.1 1.1 Definitions (continued)

REACTOR TRIP SYSTEM The RTS RESPONSE TIME shall be that time interval from when (RTS) RESPONSE TIME the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage . The response time may be measured by means of any series of sequential , overlapping, or total steps so that the entire response time is measured . In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC , or the components have been evaluated in accordance with an NRC approved methodology.

SHUTDOWN MARGIN (SOM) 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.

SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay . The SLAVE RELAY TEST shall include a continuity check of associated testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping or total steps.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems ,

subsystems, channels, or other designated components in the associated function .

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

COMANCHE PEAK UNITS 1 AND 2 1.1-6 Amendment No. 4.aG

Attachment 1 to TXX-20043 Definitions Page 7 of?

1.1 1.1 Definitions (continued)

TRIP ACTUATING DEVICE A TADOT shall consist of operating the trip actuating device and OPERATIONAL TEST (TADOT) verifying the OPERABILITY of all devices in the channel required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy. The TADOT may be performed by means of any series of sequential, overlapping or total channel steps.

COMANCHE PEAK UNITS 1 AND 2 1.1-7 Amendment No. 150

Attachment 2 to TXX-20043 Definitions Page 1of7 1.1 1.0 USEANDAPPLICATION 1.1 Definitions NOTE The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

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

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

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

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

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

COMANCHE PEAK UNITS 1 AND 2 1.1-1 Amendment No. 150

Attachment 2 to TXX-20043 Definitions Page 2 of 7 1.1 1.1 Definitions (continued)

CHANNEL OPERATIONAL A COT shall be the injection of a simulated or actual signal into TEST (COT) 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 so 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 ALTE RATION CORE ALTERA TION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS The COLR is the unit specific document that provides cycle REPORT (COLR) specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1131 DOSE EQUIVALENT I 131 shall be that concentration of 1131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes I 131, I 132, I 133, I 134, and I 135 actually present. The determination of DOSE EQUIVALENT 1131 shall be performed using thyroid dose conversion factors from Table Ill of TIO 14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or from Table E 7 of Regulatory Guide 1.109, Revision 1, NRC, 1977, or from ICRP 30, 1979, Supplement to Part 1, page 192 212, Table titled "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity," or 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."

COMANCHE PEAK UNITS 1 AND 2 1.1-2 Amendment No. 150

Attachment 2 to TXX:-20043 Definitions Page 3 of7 1.1 1.1 Definitions (continued)

DOSE EQUIVALENT XE 133 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 effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil", or using the dose conversion factors from Table B 1 of Regulatory Guide 1.109, Revision 1, NRG, 1977.

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from when FEATURE(ESF)RESPONSE the monitored parameter exceeds its ESF actuation setpoint at TIME the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRG, or the components have been evaluated in accordance with an NRC approved methodology.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the license program PROGRAM that fulfills the requirements of 10 CFR 50.55a(f).

COMANCHE PEAK UNITS 1 AND 2 1.1-3 Amendment No. 15Q, 168

Attachment 2 to TXX-20043 Definitions Page 4 of 7 1.1 1.1 Definitions (continued)

LEAKAGE 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;

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.

COMANCHE PEAK UNITS 1AND2 1.1-4 Amendment No . .:t.aQ.

Attachment 2 to TXX-20043 Definitions Page 5 of7 1.1 1.1 Definitions (continued)

OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a. Described in Chapter 14, of the FSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

PRESSURE AND The PTLR is the unit specific document that provides the reactor TEMPERATURE LIMITS vessel pressure and temperature limits, including heatup and REPORT (PTLR) cooldown rates, the power operated relief valve (PORV) lift settings and the LTOP arming temperature associated with the Low Temperature Overpressurization Protection (LTOP) System, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.6.

QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper excore detector RATIO (QPTR) calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the reactor (RTP) coolant of 3458 MWt through Cycle 13 for Unit 1 and through Cycle 11 for Unit 2. Starting with Cycles 14 and 12 of Units 1 and 2, respectively, RTP shall be 3612 Mwt.

COMANCHE PEAK UNITS 1 AND 2 1.1-5 Amendment No. 4W

Attachment 2 to TXX-20043 Definitions Page 6 of7 1.1 1.1 Definitions (continued)

REACTOR TRIP SYSTEM The RTS RESPONSE TIME shall be that time interval from when (RTS) RESPONSE TIME the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.

SHUTDOWN MARGIN (SOM) 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.

SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include a continuity check of associated testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping or total steps.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

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

COMANCHE PEAK UNITS 1 AND 2 1.1-6 Amendment No. 400

Attachment 2 to TXX-20043 Definitions Page 7 of? 1.1 1.1 Definitions (continued)

TRIP ACTUATING DEVICE A TADOT shall consist of operating the trip actuating device and OPERATIONAL TEST (TADOT) 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 TA DOT may be performed by means of any series of sequential, overlapping or total channel steps.

COMANCHE PEAK UNITS 1 AND 2 1.1-7 Amendment No. 150