RA-21-0302, Application to Revise Technical Specifications to Adopt TSTF-569, Rev. 2, Revise Response Time Testing Definition
| ML22056A434 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, Harris, McGuire  |
| Issue date: | 02/24/2022 |
| From: | Gibby S Duke Energy Progress |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RA-21-0302, TSTF-569, Rev 2 | |
| Download: ML22056A434 (32) | |
Text
Shawn Gibby Vice President Nuclear Engineering 526 South Church Street, EC-07H Charlotte, NC 28202 803-701-4543 Shawn.Gibby@duke-energy.com 10 CFR 50.90 Serial: RA-21-0302 February 24, 2022 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414 / RENEWED LICENSE NOS. NPF-35 AND NPF-52 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 / RENEWED LICENSE NO. NPF-63 MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-369 AND 50-370 / RENEWED LICENSE NOS. NPF-9 AND NPF-17
SUBJECT:
APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-569, Rev. 2, REVISE RESPONSE TIME TESTING DEFINITION Pursuant to 10 CFR 50.90, Duke Energy Progress, LLC, and Duke Energy Carolinas, LLC, collectively referred to henceforth as Duke Energy, is submitting a request for an amendment to the Technical Specifications (TS) for Catawba Nuclear Station (CNS), Units 1 and 2; Shearon Harris Nuclear Power Plant (HNP), Unit 1; and McGuire Nuclear Station (MNS), Units 1 and 2.
Duke Energy requests adoption of TSTF-569, Rev. 2, Revise Response Time Testing Definition, which is an approved change to the Improved Standard Technical Specifications (ISTS), for the CNS, HNP, and MNS 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. Attachment 3 provides existing TS Bases pages marked to show the proposed changes for information only.
This letter contains no regulatory commitments.
Duke Energy requests review and approval of this license amendment request within six months of acceptance. The amendments shall be implemented within 120 days following approval.
In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated North Carolina and South Carolina state officials.
( ~ DUKE ENERGY
U.S. Nuclear Regulatory Commission RA-21-0302 Page 2 If you should have any questions regarding this submittal, or require additional information, please contact Lee Grzeck, Manager (acting) - Nuclear Fleet Licensing, at 980-373-1530.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on February 24, 2022.
Sincerely, Shawn Gibby Vice President - Nuclear Engineering NDE
Enclosure:
Description and Assessment Attachments: 1. Proposed Technical Specification Changes (Mark-up)
- 2. Revised Technical Specification Pages
- 3. Proposed Technical Specification Bases Changes (Mark-up) cc (with Attachments):
L. Dudes, USNRC Region II - Regional Administrator V. Cusumano, USNRC - Chief, TS Branch G. A. Hutto, USNRC Senior Resident Inspector - MNS J. D. Austin, USNRC Senior Resident Inspector - CNS J. Zeiler, USNRC Senior Resident Inspector - HNP A. Hon, NRR Project Manager - Fleet S. Jenkins, Chief, Bureau of Radiological Health (SC)
A. Wilson, Attorney General (SC)
W. L. Cox, III, Section Chief, North Carolina Department of Health and Human Services, RP Section (NC)
L. Garner, Manager, Radioactive and Infectious Waste Management (SC)
Chairman, North Carolina Utilities Commission
Enclosure Description and Assessment
Subject:
APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-569, Rev. 2, REVISE RESPONSE TIME TESTING DEFINITION
1.0 DESCRIPTION
2.0 ASSESSMENT
2.1 Applicability of Safety Evaluation 2.2 Variations
3.0 REGULATORY ANALYSIS
3.1 No Significant Hazards Consideration Determination 3.2 Conclusions 4.0 ENVIRONMENTAL EVALUATION
U.S. Nuclear Regulatory Commission RA-21-0302 Page 2
1.0 DESCRIPTION
Duke Energy Progress, LLC and Duke Energy Carolinas, LLC, collectively referred to henceforth as Duke Energy, requests adoption of TSTF-569, Rev. 2, Revise Response Time Testing Definition, which is an approved change to the Improved Standard Technical Specifications (ISTS), for the Catawba Nuclear Station (CNS), Units 1 and 2 Technical Specifications (TS); Shearon Harris Nuclear Power Plant (HNP), Unit 1 TS; and McGuire Nuclear Station (MNS), Units 1 and 2 TS. The proposed amendment revises the TS Definitions for Engineered Safety Feature (ESF) Response Time and Reactor Trip System (RTS)
Response Time.
2.0 ASSESSMENT
2.1 Applicability of Safety Evaluation Duke Energy has reviewed the safety evaluation for TSTF-569 provided to the Technical Specifications Task Force in a letter dated August 14, 2019 (ADAMS Accession No. ML19176A188). This review included a review of the Nuclear Regulatory Commission (NRC) staffs evaluation, as well as the information provided in TSTF-569. As described herein, Duke Energy has concluded that the justifications presented in TSTF-569 and the safety evaluation prepared by the NRC staff are applicable to CNS, Units 1 and 2; HNP, Unit 1; and MNS, Units 1 and 2 and justify the amendments for the incorporation of the changes to the CNS TS, HNP TS, and MNS TS.
2.2 Variations Duke Energy is not proposing any variations from the TS changes described in TSTF-569 or the applicable parts of the NRC staffs safety evaluation, dated August 14, 2019.
3.0 REGULATORY ANALYSIS
3.1 No Significant Hazards Consideration Determination Duke Energy Progress, LLC and Duke Energy Carolinas, LLC, collectively referred to henceforth as Duke Energy, requests adoption of TSTF-569, Rev. 2, Revise Response Time Testing Definition, which is an approved change to the Improved Standard Technical Specifications (ISTS), for the Catawba Nuclear Station (CNS), Units 1 and 2 Technical Specifications (TS); Shearon Harris Nuclear Power Plant (HNP), Unit 1 TS; and McGuire Nuclear Station (MNS), Units 1 and 2 TS. The proposed amendment revises the TS Definitions for Engineered Safety Feature (ESF) Response Time and Reactor Trip System (RTS)
Response Time.
Duke Energy has evaluated whether a significant hazards consideration is involved with the proposed amendment(s) by focusing on 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?
Response: No
U.S. Nuclear Regulatory Commission RA-21-0302 Page 3 The proposed change revises the TS Definition of RTS and ESF instrumentation response time to permit the licensee to evaluate using an NRC-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 RTS and ESF 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 RTS and ESF instrumentation are assumed to actuate their respective components within the required response time to mitigate accidents previously evaluated. Revising the TS definition for RTS and ESF instrumentation response times to allow an NRC-approved methodology for verifying response time for some components does not alter the surveillance requirements that verify the RTS and ESF instrumentation response times are within the required limits. As such, the TS will continue to assure that the RTS and ESF 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 RTS and ESF instrumentation response time to permit the licensee to evaluate using an NRC-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 RTS or ESF instrumentation, nor does it change the Surveillance Requirement to verify the RTS and ESF instrumentation response times are within the required limits. As such, the proposed change does not alter the operability requirements for the RTS and ESF 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 The proposed change revises the TS Definition of RTS and ESF instrumentation response time to permit the licensee to evaluate using an NRC-approved methodology and apply a bounding response time for some components in lieu of measurement. The
U.S. Nuclear Regulatory Commission RA-21-0302 Page 4 proposed change has no effect on the required RTS and ESF 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 RTS and ESF instrumentation. The RTS and ESF 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 RTS and ESF 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, Duke Energy 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 Commissions 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 ENVIRONMENTAL 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.
U.S. Nuclear Regulatory Commission RA-21-0302, Attachment PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)
Definitions 1.1 1.1 Definitions (continued)
(continued)
Catawba Units 1 and 2 1.1-3 Amendment Nos. 299/295 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-85, Kr-87, Kr-88, Xe-131m, 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 III.1 of EPA Federal Guidance Report No. 12, 1993, External Exposure to Radionuclides in Air, Water, and Soil.
ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE (ESF) RESPONSE when the monitored parameter exceeds its ESF actuation TIME setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC., or the components have been evaluated in accordance with an NRC approved methodology.
INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).
Definitions 1.1 1.1 Definitions (continued)
Catawba Units 1 and 2 1.1-6 Amendment Nos. 268/264 REACTOR TRIP The RTS RESPONSE TIME shall be that time interval from SYSTEM (RTS) RESPONSE when the monitored parameter exceeds its RTS trip setpoint TIME at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology..
SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
- a.
All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn.
However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck RCCA in the SDM calculation. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and
- b.
In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.
SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing each slave relay and verifying the OPERABILITY of each slave relay.
The SLAVE RELAY TEST shall include, as a minimum, a continuity check of associated testable actuation devices.
THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
TRIP ACTUATING DEVICE A TADOT shall consist of operating the trip actuating device OPERATIONAL TEST and verifying the OPERABILITY of required alarm, interlock, (TADOT) and trip functions. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the required accuracy.
SHEARON HARRIS - UNIT 1 1-3 Amendment No. 166 DEFINITIONS E - AVERAGE DISINTEGRATION ENERGY 1.12 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration (MeV/d) for isotopes, with half-lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.
ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (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, or the components have been evaluated in accordance with an NRC approved methodology.
EXCLUSION AREA BOUNDARY 1.14 The EXCLUSION AREA BOUNDARY shall be that line beyond which the land is not controlled by the licensee to limit access.
FREQUENCY NOTATION 1.15 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.
1.16 (DELETED)
IDENTIFIED LEAKAGE 1.17 IDENTIFIED LEAKAGE shall be:
- a.
Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or
- b.
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
- c.
Reactor Coolant System leakage through a steam generator to the Secondary Coolant System (primary-to-secondary leakage).
INSERVICE TESTING PROGRAM 1.17a The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).
I
SHEARON HARRIS - UNIT 1 1-6 Amendment No. 171 DEFINITIONS PROCESS CONTROL PROGRAM 1.25 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71 and State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.
PURGE - PURGING 1.26 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.
QUADRANT POWER TILT RATIO 1.27 QUADRANT POWER TILT RATIO 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. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.
RATED THERMAL POWER 1.28 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2948 MWt.
REACTOR TRIP SYSTEM RESPONSE TIME 1.29 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its 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, or the components have been evaluated in accordance with an NRC approved methodology.
REPORTABLE EVENT 1.30 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50.
SHUTDOWN MARGIN 1.31 SHUTDOWN MARGIN 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 assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn. However, with all rod cluster assemblies verified as fully inserted by two independent means, it is not necessary to account for a stuck rod cluster assembly in the SHUTDOWN MARGIN calculation. With any rod cluster assembly not capable of being fully inserted, the reactivity worth of the rod cluster assembly must be accounted for in the determination of SHUTDOWN MARGIN, and I
Definitions 1.1 1.1 Definitions (continued)
McGuire Units 1 and 2 1.1-3 Amendment Nos. 280/260 ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE (ESF) RESPONSE when the monitored parameter exceeds its ESF actuation TIME setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology..
INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).
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.
Definitions 1.1 1.1 Definitions (continued)
McGuire Units 1 and 2 1.1-5 Amendment Nos. 194/175 QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper excore RATIO (QPTR) detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.
RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 3469 MWt.
REACTOR TRIP The RTS RESPONSE TIME shall be that time interval from SYSTEM (RTS) RESPONSE when the monitored parameter exceeds its RTS trip setpoint TIME at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology..
SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
- a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck RCCA in the SDM calculation. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and
- b.
In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.
SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing each slave relay and verifying the OPERABILITY of each slave relay. The SLAVE RELAY TEST shall include, as a minimum, a continuity check of associated testable actuation devices.
U.S. Nuclear Regulatory Commission RA-21-0302, Attachment REVISED TECHNICAL SPECIFICATION PAGES
Definitions 1.1 1.1 Definitions (continued)
(continued)
Catawba Units 1 and 2 1.1-3 Amendment Nos.
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-85, Kr-87, Kr-88, Xe-131m, 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 III.1 of EPA Federal Guidance Report No. 12, 1993, External Exposure to Radionuclides in Air, Water, and Soil.
ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE (ESF) RESPONSE when the monitored parameter exceeds its ESF actuation TIME setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.
INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).
Definitions 1.1 1.1 Definitions (continued)
Catawba Units 1 and 2 1.1-6 Amendment Nos.
REACTOR TRIP The RTS RESPONSE TIME shall be that time interval from SYSTEM (RTS) RESPONSE when the monitored parameter exceeds its RTS trip setpoint TIME at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.
SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
- a.
All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn.
However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck RCCA in the SDM calculation. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and
- b.
In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.
SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing each slave relay and verifying the OPERABILITY of each slave relay.
The SLAVE RELAY TEST shall include, as a minimum, a continuity check of associated testable actuation devices.
THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
TRIP ACTUATING DEVICE A TADOT shall consist of operating the trip actuating device OPERATIONAL TEST and verifying the OPERABILITY of required alarm, interlock, (TADOT) and trip functions. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the required accuracy.
SHEARON HARRIS - UNIT 1 1-3 Amendment No.
DEFINITIONS E - AVERAGE DISINTEGRATION ENERGY 1.12 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration (MeV/d) for isotopes, with half-lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.
ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (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, or the components have been evaluated in accordance with an NRC approved methodology.
EXCLUSION AREA BOUNDARY 1.14 The EXCLUSION AREA BOUNDARY shall be that line beyond which the land is not controlled by the licensee to limit access.
FREQUENCY NOTATION 1.15 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.
1.16 (DELETED)
IDENTIFIED LEAKAGE 1.17 IDENTIFIED LEAKAGE shall be:
- a.
Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or
- b.
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
- c.
Reactor Coolant System leakage through a steam generator to the Secondary Coolant System (primary-to-secondary leakage).
INSERVICE TESTING PROGRAM 1.17a The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).
SHEARON HARRIS - UNIT 1 1-6 Amendment No.
DEFINITIONS PROCESS CONTROL PROGRAM 1.25 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71 and State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.
PURGE - PURGING 1.26 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.
QUADRANT POWER TILT RATIO 1.27 QUADRANT POWER TILT RATIO 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. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.
RATED THERMAL POWER 1.28 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2948 MWt.
REACTOR TRIP SYSTEM RESPONSE TIME 1.29 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its 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, or the components have been evaluated in accordance with an NRC approved methodology.
REPORTABLE EVENT 1.30 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50.
SHUTDOWN MARGIN 1.31 SHUTDOWN MARGIN 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 assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn. However, with all rod cluster assemblies verified as fully inserted by two independent means, it is not necessary to account for a stuck rod cluster assembly in the SHUTDOWN MARGIN calculation. With any rod cluster assembly not capable of being fully inserted, the reactivity worth of the rod cluster assembly must be accounted for in the determination of SHUTDOWN MARGIN, and
Definitions 1.1 1.1 Definitions (continued)
McGuire Units 1 and 2 1.1-3 Amendment Nos.
ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE (ESF) RESPONSE when the monitored parameter exceeds its ESF actuation TIME setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.
INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).
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.
Definitions 1.1 1.1 Definitions (continued)
McGuire Units 1 and 2 1.1-5 Amendment Nos.
QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper excore RATIO (QPTR) detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.
RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 3469 MWt.
REACTOR TRIP The RTS RESPONSE TIME shall be that time interval from SYSTEM (RTS) RESPONSE when the monitored parameter exceeds its RTS trip setpoint TIME at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.
SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
- a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck RCCA in the SDM calculation. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and
- b.
In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.
SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing each slave relay and verifying the OPERABILITY of each slave relay. The SLAVE RELAY TEST shall include, as a minimum, a continuity check of associated testable actuation devices.
U.S. Nuclear Regulatory Commission RA-21-0302, Attachment PROPOSED TECHNICAL SPECIFICATION BASES CHANGES (MARK-UP)
RTS Instrumentation B 3.3.1 BASES Catawba Units 1 and 2 B 3.3.1-56 Revision No. 8 SURVEILLANCE REQUIREMENTS (continued)
WCAP-14036-P-A Revision 1, "Elimination of Periodic Protection Channel Response Time Tests" provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time. The allocations for sensor, signal conditioning and actuation logic response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.
The response time may be verified for components that replace the components that were previously evaluated in Ref. 8 and Ref. 9, provided that the components have been evaluated in accordance with the NRC approved methodology as discussed in Attachment 1 to TSTF-569, Rev. 2, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing, (Ref.
14).
The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.16 is modified by a Note stating that neutron detectors are excluded from RTS RESPONSE TIME testing. This Note is necessary because of the difficulty in generating an appropriate detector input signal. Excluding the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response. The response time of the neutron flux signal portion of the channel shall be measured from detector output or input of the first electronic component in the channel.
REFERENCES
- 1. UFSAR, Chapter 7.
- 2. UFSAR, Chapter 6.
- 3. UFSAR, Chapter 15.
- 4. IEEE-279-1971.
- 5. 10 CFR 50.49.
RTS Instrumentation B 3.3.1 BASES Catawba Units 1 and 2 B 3.3.1-57 Revision No. 8 REFERENCES (continued)
- 6. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 7. WCAP-10271-P-A, Supplement 2, Rev. 1, June 1990.
- 8. WCAP-13632-P-A Revision 2, Elimination of Pressure Sensor Response Time Testing Requirements Jan., 1996.
- 9. WCAP-14036-P-A Revision 1, Elimination of Periodic Protection Channel Response Time Tests Oct., 1998.
10.10 CFR 50.67.
11.WCAP-14333-P-A, Rev. 1, October 1998.
12.WCAP-15376-P-A, Rev. 1, March 2003.
- 13. Technical Specification Task Force, Improved Standard Technical Specifications Change Traveler, TSTF-493, Clarify Application of Setpoint Methodology for LSSS Functions Revision 4.
- 14. Attachment 1 to TSTF-569, Rev.2, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing
ESFAS Instrumentation B 3.3.2 BASES Catawba Units 1 and 2 B 3.3.2-50 Revision No. 13 SURVEILLANCE REQUIREMENTS (continued)
WCAP-14036-P-A Revision 1, "Elimination of Periodic Protection Channel Response Time Tests" provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time. The allocations for sensor, signal conditioning and actuation logic response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.
The response time may be verified for components that replace the components that were previously evaluated in Ref. 8 and Ref. 9, provided that the components have been evaluated in accordance with the NRC approved methodology as discussed in Attachment 1 to TSTF-569, Rev. 2, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing, (Ref. 15).
The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that clarifies that the turbine driven AFW pump is tested within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 600 psig in the SGs.
SR 3.3.2.11 SR 3.3.2.11 is the performance of a COT on the NSWS Suction Transfer
- Low Pit Level.
A COT is performed on each required channel to ensure the entire channel will perform the intended Function. Setpoints must be found within the Allowable Values specified in Table 3.3.2-1. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.
ESFAS Instrumentation B 3.3.2 BASES Catawba Units 1 and 2 B 3.3.2-51 Revision No. 13 SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.2.12 SR 3.3.2.12 is the performance of an ACTUATION LOGIC TEST on the Doghouse Water Level-High High and NSWS Suction Transfer-Emergency Low Pit Level Functions.
An ACTUATION LOGIC TEST to satisfy the requirements of GL 96-01 is performed on each instrumentation to ensure all logic combinations will initiate the appropriate Function. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.
REFERENCES
- 1.
UFSAR, Chapter 6.
- 2.
UFSAR, Chapter 7.
- 3.
UFSAR, Chapter 15.
- 4.
- 5.
- 6.
10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 7.
WCAP-10271-P-A, Supplement 1 and Supplement 2, Rev. 1, May 1986 and June 1990.
- 8.
WCAP-13632-P-A Revision 2, Elimination of Pressure Sensor Response Time Testing Requirements Jan., 1996.
- 9.
WCAP-14036-P-A Revision 1, Elimination of Periodic Protection Channel Response Time Tests Oct., 1998.
- 10.
Not used.
- 11.
Not used.
- 12.
Not used.
- 13.
WCAP-14333-P-A, Revision 1, October 1998.
- 14.
Not used.
ESFAS Instrumentation B 3.3.2 BASES Catawba Units 1 and 2 B 3.3.2-52 Revision No. 13 REFERENCES (continued)
- 15. to TSTF-569, Rev. 2, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing.
3/4.3 INSTRUMENTATION BASES SHEARON HARRIS - UNIT 1 B 3/4 3-2b Amendment No. 154 channel performance. A trip setpoint may be set more conservative than the NTSP as necessary in response to plant conditions. However, in this case, the instrument operability must be verified based on the field setting and not the NTSP.
The measurement of response time at the frequencies specified in the Surveillance Frequency Control Program provides assurance that the reactor trip and the Engineered Safety Features actuation associated with each channel is completed within the time limit assumed in the safety analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable. Response time may be demonstrated by any series of sequential, overlapping, or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Response time may be verified by actual response time tests in any series of sequential, overlapping or total channel measurements; or by the summation of allocated sensor, signal processing and actuation logic response times with actual response time tests on the remainder of the channel. Allocations for sensor response times may be obtained from: (1) historical records based on acceptable response time tests (hydraulic, noise or power interrupt tests); (2) inplace, onsite, or offsite (e.g., vendor) test measurements; or (3) utilizing vendor engineering specifications. WCAP-13632-P-A, Rev. 2, "Elimination of Pressure Sensor Response Time Testing Requirements," provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the WCAP. Response time verification for other sensor types must be demonstrated by test.
WCAP 14036-P-A, Rev. 1, "Elimination of Periodic Response Time Tests," provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time. The allocations for sensor, signal conditioning, and actuation logic response times must be verified prior to placing the component into operational service and re-verified following maintenance or modification that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for the repair are the same type and value. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing element of a transmitter.
The response time may be verified for components that replace the components that were previously evaluated in Ref. 1 and Ref. 2, provided that the components have been evaluated in accordance with the NRC approved methodology as discussed in Attachment 1 to TSTF-569, Rev. 2 Methodology to Eliminate Pressure Sensor and Protection channel (for Westinghouse Plants only) Response Time Testing, (Ref.3).
The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents events, and transients. Once the required logic combination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss-of-coolant accident: (1) charging/safety injection pumps start and automatic valves position, (2) reactor trip, (3) feedwater isolation, (4) startup of the emergency diesel generators, (5) containment spray pumps start and automatic valves position (6) containment isolation, (7) steam line isolation, (8) turbine trip, (9) auxiliary feedwater pumps start and automatic valves position, (10) containment fan coolers start and automatic valves position, (11) emergency service water pumps start and automatic valves position, and (12) control room isolation and emergency filtration start.
I
3/4.3 INSTRUMENTATION BASES SHEARON HARRIS - UNIT 1 B 3/4 3-4 Amendment No. 182 P-14 On increasing steam generator water level, P-14 automatically trips all feedwater isolation valves and inhibits feedwater control valve modulation.
References
- 1.
WCAP-13632-P-A, Revision 2, Elimination of Pressure Sensor Response Time Testing Requirements, January 1996.
- 2.
WCAP-14036-P-A, Revision 1, Elimination of Periodic Protection Channel Response Time Tests, October 1998.
- 3. to TSTF-569, Rev. 2, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing.
3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS The OPERABILITY of the radiation monitoring instrumentation for plant operations ensures that:
(1) the associated action will be initiated when the radiation level monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic is maintained, and (3) sufficient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance. The radiation monitors for plant operations senses radiation levels in selected plant systems and locations and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents and abnormal conditions. Once the required logic combination is completed, the system sends actuation signals to initiate alarms or automatic isolation action and actuation of emergency systems.
3/4.3.3.2 DELETED 3/4.3.3.3 DELETED 3/4.3.3.4 DELETED 3/4.3.3.5 REMOTE SHUTDOWN SYSTEM The OPERABILITY of the Remote Shutdown System ensures that sufficient capability is available to permit safe shutdown of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50.
The OPERABILITY of the Remote Shutdown System ensures that a fire will not preclude achieving safe shutdown. The Remote Shutdown System instrumentation, control, and power circuits and transfer switches necessary to eliminate effects of the fire and allow operation of instrumentation, control and power circuits required to achieve and maintain a safe shutdown condition are independent of areas where a fire could damage systems normally used to shut down the reactor.
This capability is consistent with General Design Criterion 3, 10 CFR 50.48(a) and 10 CFR 50.48(c).
3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient
RTS Instrumentation B 3.3.1 BASES McGuire Units 1 and 2 B 3.3.1-48 Revision No. 146 SURVEILLANCE REQUIREMENTS (continued) allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the WCAP. Response time verification for other sensor types must be either demonstrated by test, or their equivalency to those listed in WCAP-13632-P-A, Revision 2. Any demonstration of equivalency must have been determined to be acceptable by NRC staff review.
WCAP-14036-P-A, Revision 1, Elimination of Periodic Protection Channel Response Time Tests provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time. The allocations for sensor, signal conditioning, and actuation logic response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.
The response time may be verified for components that replace the components that were previously evaluated in Ref. 8 and Ref. 9, provided that the components have been evaluated in accordance with the NRC approved methodology as discussed in Attachment 1 to TSTF-569, Rev. 2, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing, (Ref. 13).
The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.16 is modified by a Note stating that neutron detectors are excluded from RTS RESPONSE TIME testing. This Note is necessary because of the difficulty in generating an appropriate detector input signal. Excluding the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response. The response time of the neutron flux signal portion of the channel shall be measured from detector output or input of the first electronic component in the channel.
RTS Instrumentation B 3.3.1 BASES McGuire Units 1 and 2 B 3.3.1-49 Revision No. 146 REFERENCES
- 1.
UFSAR, Chapter 7.
- 2.
UFSAR, Chapter 6.
- 3.
UFSAR, Chapter 15.
- 4.
- 5.
- 6.
10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 7.
WCAP-10271-P-A, Supplement 2, Rev. 1, June 1990.
- 8.
WCAP 13632-P-A, Revision 2, Elimination of Pressure Sensor Response Time Testing Requirements Jan., 1996.
- 9.
WCAP-14036-P-A, Revision 1, Elimination of Periodic Protection Channel Response Time Tests Oct., 1998.
- 10.
WCAP-14333-P-A, Revision 1, October 1998.
- 11.
WCAP-15376-P-A, Revision 1, March 2003.
- 12.
Technical Specification Task Force, Improved Standard Technical Specifications Change Traveler, TSTF-493, Clarify Application of Setpoint Methodology for LSSS Functions, Revision 4.
- 13. Attachment 1 to TSTF-569, Rev. 2, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only)
Response Time Testing.
ESFAS Instrumentation B 3.3.2 BASES McGuire Unit 1 and 2 B 3.3.2-42 Revision No. 172 SURVEILLANCE REQUIREMENTS (continued) of the protection system channel response time. The allocations for sensor, signal conditioning, and actuation logic response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.
The response time may be verified for components that replace the components that were previously evaluated in Ref. 8 and Ref. 9, provided that the components have been evaluated in accordance with the NRC approved methodology as discussed in Attachment 1 to TSTF-569, Rev. 2, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing, (Ref. 12).
The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that clarifies that the turbine driven AFW pump is tested within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 900 psig in the SGs.
REFERENCES
- 1.
UFSAR, Chapter 6.
- 2.
UFSAR, Chapter 7.
- 3.
UFSAR, Chapter 15.
- 4.
- 5.
- 6.
10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 7.
WCAP-10271-P-A, Supplement 1 and Supplement 2, Rev. 1, May 1986 and June 1990.
- 8.
WCAP 13632-P-A, Revision 2, Elimination of Pressure Sensor Response Time Testing Requirements Jan., 1996.
- 9.
WCAP-14036-P-A, Revision 1, Elimination of Periodic Protection Channel Response Time Tests Oct., 1998.
ESFAS Instrumentation B 3.3.2 BASES McGuire Unit 1 and 2 B 3.3.2-43 Revision No. 172
- 10.
WCAP-14333-P-A, Revision 1, October 1998.
- 11.
WCAP-15376-P-A, Revision 1, March 2003.
- 12. to TSTF-569, Rev. 2, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only)
Response Time Testing.