ML20141L660
| ML20141L660 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 05/04/2020 |
| From: | Florida Power & Light Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20141L672 | List:
|
| References | |
| L-2020-043 | |
| Download: ML20141L660 (37) | |
Text
LIST OF EFFECTIVE PAGES CHAPTER 13 CONDUCT OF OPERATIONS PAGE AMENDMENT 13-1 29 13-i 28 13-ia 29 13-ii 28 13.1-1 22 13.2-1 25 13.3-1 22 13.4-1 22 13.5-1 22 13.6-1 22 13.7-1 2 13.8-1 22 13.8-2 22 13.8-2a 25 13.8-3 29 13.8-4 22 13.8-4a 23 13.8-4b 25 13.8-5 28 13.8-5a 28 13.8-5b 29 13.8-5c 28 13.8-6 22 13.8-6a 26 13.8-7 21 13.8-8 22 13.8-9 15 13.8-10 15 13.8-11 19 13.8-12 15 13.8-12a 26 13.8-12b 28 13.8-13 26 13.8-14 15 13.8-15 18 13.9-1 29 UNIT 1 13-1 Amendment No. 29 (10/18)
TABLE OF CONTENTS CHAPTER 13 CONDUCT OF OPERATIONS SECTION TITLE PAGE 13.1 ORGANIZATIONAL STRUCTURE 13.1-1 13.2 TRAINING PROGRAM 13.2-1 13.3 EMERGENCY PLANNING 13.3-1 13.4 REVIEW AND AUDIT 13.4-1 13.5 PLANT PROCEDURES 13.4-1 13.6 PLANT RECORDS 13.6-1 13.7 INDUSTRIAL SECURITY 13.6-1 13.8 LICENSEE-CONTROLLED TECHNICAL SPECIFICATION REQUIREMENTS 13.8-1 13.8.1 TECHNICAL SPECIFICATION REQUIREMENTS 13.8-1 13.8.1.1 Seismic Instrumentation 13.8-1 13.8.1.2 Incore Detectors 13.8-2 13.8.1.3 Meteorological Instrumentation 13.8-3 13.8.1.4 Explosive Gas Monitoring Instrumentation 13.8-4 13.8.1.5 Crane Travel - Spent Fuel Storage Pool Building 13.8-4a 13.8.1.6 Spent Fuel Cask Crane Travel 13.8-4a 13.8.1.7 Combustible Gas Control 13.8-4b 13.8.1.8 Leading Edge Flow Meter (LEFM) 13.8-5 13.8.1.9 Reactor Coolant Chemistry 13.8-5 13.8.1.10 Communications 13.8-5a 13.8.1.11 Manipulator Crane Operability 13.8-5a 13.8.1.12 DOST Sediment Cleaning Surveillance 13.8-5b Requirements UNIT 1 13-i Amendment No. 28 (05/17)
TABLE OF CONTENTS CHAPTER 13 CONDUCT OF OPERATIONS SECTION TITLE PAGE 13.8.2 LINE-ITEM TECHNICAL SPECIFICATION REQUIREMENTS 13.8-5b 13.8.2.1 Reactor Protective Instrumentation Response Times 13.8-5b 13.8.2.2 Engineered Safety Features Actuation Systems Instrumentation Response Times 13.8-5c 13.8.2.3 Flood Protection 13.8-6 13.8.2.4 ESF Pump Acceptance Criteria and Test Methods 13.8-6a 13.9 NRC TS AMENDMENT RELATED 13.9-1 COMMITMENTS 13.9.1 Amendment 234 TSTF-422 Surveillance 13.9-1 Requirements UNIT 1 13-ia Amendment No. 29 (10/18)
LIST OF TABLES CHAPTER 13 CONDUCT OF OPERATIONS SECTION TITLE PAGE 13.8.1-1 Seismic Monitoring Instrumentation 13.8-7 13.8.1-2 Seismic Monitoring Instrumentation 13.8-8 Surveillance Requirements 13.8.1-3 Meteorological Monitoring Instrumentation 13.8-9 13.8.1-4 Meteorological Monitoring Instrumentation 13.8-10 Surveillance Requirements 13.8.1-5 Explosive Gas Monitoring Instrumentation 13.8-11 13.8.1-6 Explosive Gas Monitoring Instrumentation 13.8-12 Surveillance Requirements 13.8.1-7 LEFM Calorimetric Instrumentation 13.8-12a 13.8.1-8 Reduced Power Limits Applicable to 13.8-12a Inoperable LEFM Calorimetric Instrumentation 13.8.1-9 Reactor Coolant System Chemistry Limits 13.8-12b 13.8.1-10 Reactor Coolant System Chemistry Limits 13.8-12b Surveillance Requirements 13.8.2-1 Reactor Protective Instrumentation Response 13.8-13 Times 13.8.2-2 Engineered Safety Features Response Times 13.8-14 UNIT 1 13-ii Amendment No. 28 (05/17)
CHAPTER 13 CONDUCT OF OPERATIONS 13.1 ORGANIZATIONAL STRUCTURE The Florida Power and Light Company Organization is provided in the FPL Quality Assurance Topical Report discussed in Section 17.2.
13.1-1 Amendment No. 22 (05/07)
13.2 TRAINING PROGRAM The Florida Power and Light St. Lucie facility training program is maintained under the direction of the training manager, and meets or exceeds the requirements and recommendations of Section 5.5 of ANSI/ANS-3.1 1978. St. Lucie plant training programs are accredited through the National Nuclear Accrediting Board (NNAB). National Academy for Nuclear Training (NANT) guidelines, which are endorsed by the NNAB, are utilized at St. Lucie.
13.2-1 Amendment No. 25 (04/12)
13.3 EMERGENCY PLANNING The St. Lucie Plant Radiological Emergency Plan is a separate document which has been previously submitted to the NRC for St. Lucie Unit 1. The Radiological Emergency Plan describes Florida Power and Light Company's plans for responding to all foreseeable emergencies which have resulted in, or have the potential to result in, the accidental release of radiation to the environment. The plan has been prepared to meet the requirements of 10 CFR 50.34(b)(6)(V), 10 CFR 50.47, 10 CFR 50.72 and 10 CFR 50 Appendix E. The purpose of this plan is to define and assign authority and responsibility in order to protect the health and safety of the public and plant personnel. This document is updated periodically.
13.3-1 Amendment No. 22 (05/07)
REFERENCES FOR SECTION 13.3
- 1. Florida Power and Light Company, St. Lucie Plant Radiological Emergency Plan, Revision 49.
13.3-2 Amendment No. 22 (05/07)
13.4 REVIEW AND AUDIT Conduct of reviews and audits of operating phase activities is provided in the FPL Quality Assurance Topical Report described in Section 17.2.
13.4-1 Amendment No. 22 (05/07)
13.5 PLANT PROCEDURES Written plant procedures are established, implemented and maintained covering the applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978, those required for implementing the requirements of NUREG-0737 and plant activities including refueling operations, surveillance and test activities of safety related equipment and the Fire Protection Program implementation, as outlined in Section 6.8, Procedures and Programs, of the plant Technical Specifications.
13.5-1 Amendment No. 22 (05/07)
13.6 PLANT RECORDS A recorded history of the plant is maintained in accordance with 10 CFR 50, Appendix B, Section XVII.
Complete records are retained as prescribed in the FPL Quality Assurance Topical Report described in Section 17.2, to assure the ability to reconstruct significant events and satisfy any statutory requirements which apply.
13.6-1 Amendment No. 22 (05/07)
13.7 INDUSTRIAL SECURITY The St. Lucie Plant Security Plan(1) has been previously submitted to the NRC pursuant to 10 CFR 2.790(d). The security plan outlines methods and procedures for the prevention or mitigation of consequences of industrial sabotage and other acts of vandalism, arson and civil disturbance which are inimical to the safe operation of the plant and the health and safety of the public. Due to the sensitive nature of this information the latest Plant Security Plan is available for NRC review but withheld from public disclosure.
REFERENCES FOR SECTION 13.7
- 1. Florida Power and Light Company, St. Lucie Plant Security Plan submitted to NRC.
13.7-1 Am. 2-7/84
13.8 LICENSEE-CONTROLLED TECHNICAL SPECIFICATION REQUIREMENTS This section of the FSAR contains ACTION STATEMENTS, LIMITING CONDITIONS OF OPERATION (LCOs), and SURVEILLANCE REQUIREMENTS for technical specifications that have little or no impact on the prevention or mitigation of design basis accidents.
In accordance with the NRC final policy statement on Technical Specifications improvements for nuclear power reactors, these "supplemental" technical specifications remain a condition of the facility operating license, but in light of their low risk assessment, do not require prior NRC approval for revision or changes. Requirements relocated to the FSAR will be controlled through 10 CFR 50.59.
Line items or segments of current plant Technical Specifications that can be changed without NRC approval are also listed in this section. The identification of these "line-item" technical specifications is based on guidance contained in the NRC final policy statement.
The technical specification requirements contained in this section were removed from St. Lucie Unit 1 Technical Specifications by means of approved License Amendments. All surveillance test intervals described below can be extended by up to 25%.
13.8.1 TECHNICAL SPECIFICATIONS REQUIREMENTS 13.8.1.1 Seismic Instrumentation Plant operating restrictions associated with the seismic instrumentation were removed from the facility Technical Specifications by License Amendment No. 135 and NRC Safety Evaluation Report issued April 25, 1995.
Operability of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix A of 10 CFR Part 100. The instrumentation is shared between Units 1 and 2 and is consistent with the recommendations of Regulatory Guide 1.12, "Instrumentation for Earthquakes", April 1974. Seismic instrumentation is also discussed in Section 3.7.4.
13.8.1.1.1 Limiting Condition for Operation The seismic monitoring instrumentation shown in Table 13.8.1-1 shall be OPERABLE.
APPLICABILITY: At all times.
ACTION:
- a. With the number of OPERABLE seismic monitoring channels less than required by Table 13.8.1-1, restore the inoperable channel(s) to OPERABLE status as soon as practical.
13.8-1 Amendment No. 22 (05/07)
13.8.1.1.2 Surveillance Requirements Each of the above seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown on Table 13.8.1-2.
Each of the above seismic monitoring instruments actuated during a seismic event (greater than or equal to 0.01g) shall be restored to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and a CHANNEL CALIBRATION performed within 5 days. Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion.
13.8.1.2 Incore Detectors Plant operating restrictions associated with the Incore Detectors were removed from the facility Technical Specifications by License Amendment No. 136 and NRC Safety Evaluation Report issued June 6, 1995.
The operability of the incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core. The NRC Safety Evaluation Report issued June 6, 1995 with Technical Specification License Amendment #136 removing the incore detector operating restrictions from the Technical Specifications and placing them in the UFSAR recommends that any safety evaluation to reduce the minimum number of operable incore detectors address certain issues. These issues are specified below:
- 1) How an inadvertent loading of a fuel assembly into an improper location will be detected;
- 2) How the validity of the tilt estimates will be ensured;
- 3) How adequate core coverage will be maintained;
- 4) How the measurement uncertainties will be assured and why the added uncertainties are adequate to guarantee that measured peak linear heat rates, peak pin powers radial peaking factors, and azimuthal power tilts will meet Technical Specification limits; and
- 5) How the incore detector system will be restored to full (or nearly full) service before the beginning of each cycle.
13.8.1.2.1 Limiting Conditions for Operation
- 1. The incore detection system shall be operable using the BEACON code with:
- a. The minimum requirement for detector availability at the beginning of each cycle is 75% for startup testing misload verification. For the rest of the cycle 50% detector availability is needed, and 13.8-2 Amendment No. 22 (05/07)
- b. There is no requirement for detector symmetric locations; however, there is a minimum requirement of operable detectors for each quadrant for core tilt surveillance and excore detector calibration. This requirement is as follows:
10 Detectors/Quadrant 4 Detectors/Top-Half Quadrant 4 Detectors/Bottom-Half Quadrant, and
- c. A measurement-calculational uncertainty factor applied in accordance with JPN-PSLSEFJ-96-022, "Evaluation of the Best Estimate Analyzer for Core Operations
- Nuclear (BEACON)."
APPLICABILITY: When the incore detection system is used for:
- a. Recalibration of the excore axial flux offset detection system;
- b. Monitoring the AZIMUTHAL POWER TILT;
- c. Calibration of the power level neutron flux channels; or
- d. Monitoring the linear heat rate.
ACTION:
- a. With the incore detection system inoperable, do not use the system for the above applicable monitoring or calibration functions.
13.8-2a Amendment No. 25 (04/12)
13.8.1.2.2 Surveillance Requirements The incore detection system shall be demonstrated OPERABLE:
- a. By performance of a CHANNEL CHECK within 7 days prior to its use and when required for:
- 1. Recalibration of the excore axial flux offset detection system;
- 2. Monitoring the linear heat rate pursuant to Technical Specification 4.2.1.4;
- 3. Monitoring the AZIMUTHAL POWER TILT; or
- 4. Calibration of the Power Level Neutron Flux Channels.
- b. At least once per 18 months by the performance of a CHANNEL CALIBRATION operation which exempts the neutron detectors but includes all electronic components.
The neutron detectors are calibrated prior to installation in the reactor core.
The incore detector monitoring system may be used for monitoring the core power distribution by verifying that the incore detector Local Power Density alarms are adjusted to satisfy the requirements of the core power distribution map and when the setpoint for these alarms include allowances, set in the conservative directions, for (1) a measurement-calculation uncertainty factor applied consistent with Section 13.8.1.2.1.1.c; (2) an engineering uncertainty factor of 1.03; (3) a THERMAL POWER measurement uncertainty factor of 1.02.
13.8.1.3 Meteorological Instrumentation The Meteorological Instrumentation System is shared between St. Lucie Units 1 and 2. The EC246531 Meteorological system shares the same Tower and dataloggers (one at the Met Tower, one at the Control Room). Meteorological Instrumentation operation restrictions were removed from the facility Technical Specifications by License Amendment No. 147 and NRC Safety Evaluation Report issued August 20, 1996.
Operability of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating the potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972.
13.8.1.3.1 Limiting Condition for Operation The meteorological monitoring instrumentation shown in Table 13.8.1-3 shall be OPERABLE. The emergency power source may be inoperable in Modes 5 or 6.
APPLICABILITY: At all times.
ACTION:
- a. With the number of OPERABLE meteorological monitoring channels less than required by Table 13.8.1-3, suspend all release of gaseous radioactive material from the radwaste gas decay tanks until the inoperable channel(s) is restored to OPERABLE status.
13.8-3 Amendment No. 29 (10/18)
13.8.1.3.2 Surveillance Requirements Each of the above meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and CHANNEL CALIBRATION operation at the frequencies shown in Table 13.8.1-4.
13.8.1.4 Explosive Gas Monitoring Instrumentation The Explosive Gas Monitoring Instrumentation operation restrictions were removed from the facility Technical Specifications by License Amendment No. 147 and NRC Safety Evaluation Report issued August 20, 1996.
The explosive gas monitoring instrumentation is provided to monitor the concentrations of potentially explosive gas mixtures in the waste gas holdup system. The OPERABILITY and use of the instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.
13.8.1.4.1 Limiting Condition for Operation The explosive gas monitoring instrumentation channels shown in Table 13.8.1-5 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Technical Specification 3.11.2.5 are not exceeded.
APPLICABILITY: As shown in Table 13.8.1-5.
ACTION:
- a. With the explosive gas monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above Specification declare the channel inoperable.
- b. With less than the minimum number of explosive gas monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 13.8.1-5.
13.8.1.4.2 Surveillance Requirements Each explosive gas monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 13.8.1-6.
13.8-4 Amendment No. 22 (05/07)
13.8.1.5 Crane Travel - Spent Fuel Storage Pool Building The Crane Travel - Spent fuel Storage Pool Building load restriction was removed from the facility Technical Specifications by NRC approval of License Amendment No. 190 and NRC Safety Evaluation issued April 28, 2004.
The restriction on movement of loads in excess of the nominal weight of a fuel assembly and CEA over irradiated fuel assemblies ensures that no more than the contents of one fuel assembly will be ruptured in the event of a fuel handling accident. This assumption is consistent with the activity release assumed in the accident analyses.
13.8.1.5.1 Limiting Condition for Operation Loads in excess of 2000 pounds shall be prohibited from travel over irradiated fuel assemblies in the storage pool.
APPLICABILITY: With fuel assemblies in the storage pool unless the following conditions are met:
- a. fuel assemblies are in a spent fuel transfer cask in the cask pit area and,
- b. the Spent Fuel Cask Crane main hook is being used to place the cask lid on the cask.
ACTION: With the requirements for above specification not satisfied, place the crane load in a safe condition. The provisions of Specification 3.0.3 are not applicable.
13.8.1.5.2 Surveillance Requirements Crane interlocks and physical stops which prevent crane travel with loads in excess of 2000 pounds over fuel assemblies shall be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereafter during crane operation.
13.8.1.6 Spent Fuel Cask Crane Travel The Spent Fuel Cask Crane load restriction was removed from the facility Technical Specifications by NRC approval of License Amendment No. 190 and NRC Safety Evaluation issued April 28, 2004.
The maximum load which may be handled by the spent fuel cask crane main hook is limited to 150 tons.
This restriction is provided to prevent the load exceeding the single-failure-proof design load limit.
The minimum vertical clearance below the load while the load is in transit is required to be 3.0 inches (3.0-inch minimum). This restriction is provided to prevent impact in the unlikely event of an uncontrolled load movement.
When moving the spent fuel transfer cask shield plug, the rigging shall ensure that the lid remains horizontal and cannot tilt. This restriction is provided to prevent the shield plug from contacting spent fuel in the spent fuel transfer cask.
13.8-4a Amendment No. 23 (11/08)
13.8.1.6.1 Limiting Condition for Operation The maximum load which may be handled by the spent fuel cask crane main hook shall not exceed 150 tons.
The minimum vertical clearance below a load when the load is in transit shall be at least 3.0 inches.
The cask lid rigging shall prevent the cask lid from tilting.
APPLICABILITY: Whenever loads are moved with the cask crane main hook.
ACTION: With the requirements for above specification not satisfied, place load in a safe condition.
The provisions of Specification 3.0.3 are not applicable.
13.8.1.6.2 Surveillance Requirements The main hook load shall be verified to not exceed 150 tons prior to attaching it to the spent fuel cask crane.
The vertical clearance below the load shall be verified to be at least 3.0 inches before moving the load horizontally.
13.8.1.7 Combustible Gas Control Plant operating restrictions associated with hydrogen analyzers and electric hydrogen recombiners - W were removed from the facility Technical Specifications by License Amendment No. 204 and NRC Safety Evaluation Report issued February 22, 2008.
The operability of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions.
The containment fan coolers are used in a secondary function to ensure adequate mixing of the containment atmosphere following a LOCA. This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit.
13.8.1.7.1 13.8.1.7.1.a Two independent containment hydrogen analyzers shall be OPERABLE.
APPLICABILITY: MODES 1 AND 2.
ACTION:
With one hydrogen analyzer inoperable, restore the inoperable analyzer to OPERABLE status within 30 days or demonstrate within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that the grab sample system of the inoperable hydrogen analyzer has the capability to draw a sample of the containment atmosphere into the grab sample canister. Verify this capability of the grab sample system at least once per 30 days thereafter. If no Hydrogen Analyzers are operable, take the actions above for one inoperable hydrogen analyzer, initiate a Condition Report to document the condition and the plans to restore at least one Hydrogen Analyzer to operable status.
13.8-4b Amendment No. 25 (04/12)
13.8.1.7.1.b Each hydrogen analyzer shall be demonstrated OPERABLE by the performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days, and at least once per 92 days on a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gases containing:
13.8.1.8 Leading Edge Flow Meter (LEFM)
The PSL Unit 1 Extended Power Uprate (EPU) raised the licensed maximum power level to 3020 MWt.
The EPU change to the maximum rated thermal power (RTP) included a 1.7% Measurement Uncertainty Recapture (MUR) based on installation of the LEFM. The use of LEFM for determination of feedwater temperature and feedwater mass flow, results in an overall calorimetric uncertainty of 0.3%. The MUR uprate of 1.7% results from the difference between the original 2% power determination uncertainty (required by 10CFR50 Appendix K) and the LEFM based calorimetric uncertainty of 0.3%.
Operability of the LEFM instrumentation is required to support an overall calorimetric uncertainty of 0.3%.
Various LEFM system failure modes and resulting action statements are considered based on the use of independent LEFM instrumentation of feedwater headers A & B, and also based on redundancy within each LEFM sub-system. Existing feedwater flow (Venturis) and temperature (RTD) instrumentation have been retained and are used as calorimetric instrumentation if needed.
13.8.1.8.1 Limiting Condition for Operation The LEFM instrumentation shown in Table 13.8.1-7 shall be OPERABLE.
APPLICABILITY: Mode 1 at greater than 98.3% Reactor Power.
ACTION:
a) With the number of OPERABLE LEFM / Calorimetric instrument channels less than required by Table 13.8.1-7, restore the inoperable channels to OPERABLE status or be in compliance with the reduced power limits of Table 13.8.1-8 within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
b) If the plant experiences a power change of greater than 10% while operating on the Venturis during the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period, then power level will be restricted to less than or equal to 2968 MWt until the LEFM system is fully OPERABLE.
c) If LCV-9006 is in operation at greater than 98.3% Reactor Power, then the Calorimetric Input shall be swapped to the Venturis and the 48-Hour LCO shall be commenced.
13.8.1.9 Reactor Coolant System Chemistry Plant operating restrictions associated with reactor coolant system chemistry were removed from the facility Technical Specifications by License Amendment No. 225 and NRC Safety Evaluation Report issued August 14, 2015.
The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride and UNIT 1 13.8-5 Amendment No. 28 (05/17)
fluoride limits are time and temperature dependent. While poor chemistry control can lead to a more rapid degradation of the primary materials, this type of degradation is a long-term process; furthermore, poor Reactor Coolant System chemistry is a cause of, not a detector or indicator of, Reactor Coolant System degradation. The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.
13.8.1.9.1 Limiting Condition for Operation The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 13.8.1-9.
APPLICABILITY: All MODES ACTION:
MODES 1, 2, 3 and 4:
a) With any one or more chemistry parameter in excess of its Steady State Limit but within its transient Limit, restore the parameter to within its Steady State Limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b) With any one or more chemistry parameter in excess of its Transient Limit, restore the parameter to be in excess of its Steady State Limit but within its Transient Limit within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
MODES 5 and 6:
With the concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady State Limit for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or in excess of its Transient Limit, reduce the pressurizer pressure to less than or equal* to 500 psia, if applicable, and perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation prior to increasing the pressurizer pressure above 500 psia or prior to proceeding to MODE 4.
13.8.1.9.2 Surveillance Requirements The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in Table 13.8.1-10.
13.8.1.10 Communications The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during core alterations.
13.8.1.10.1 Limiting Condition for Operation Direct communications shall be maintained between the control room and personnel at the refueling station.
APPLICABILITY: During CORE ALTERATIONS.
ACTION:
When direct communications between the control room and personnel at the refueling station cannot be maintained, suspend all CORE ALTERATIONS. The provisions of Specification 3.0.3 are not applicable.
13.8.1.10.2 Surveillance Requirements Direct communications between the control room and personnel at the refueling station shall be demonstrated within one hour prior to the start of and in accordance with the Surveillance Frequency Control Program during CORE ALTERATIONS.
13.8.1.11 Manipulator Crane Operability The OPERABILITY requirements of the cranes used for movement of fuel assemblies ensures that: 1) each crane has sufficient load capacity to lift a fuel element, operations.
UNIT 1 13.8-5a Amendment No. 28 (05/17)
- 2) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.
13.8.1.11.1 Limiting Condition for Operation The manipulator crane shall be used for movement of CEAs or fuel assemblies and shall be OPERABLE with:
- a. A minimum capacity of 2000 pounds, and
- b. An overload cut off limit of < 3000 pounds APPLICABILITY: During movement of CEAs or fuel assemblies within the reactor pressure vessel.
ACTION:
With the requirements for crane OPERABILITY not satisfied, suspend use of any inoperable manipulator crane from operations involving the movement of CEAs and fuel assemblies within the reactor pressure vessel.
13.8.1.11.2 Surveillance Requirements The manipulator crane used for movement of CEAs or fuel assemblies within the reactor pressure vessel shall be demonstrated OPERABLE within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to the start of such operations by performing a load test of at least 2500 pounds and demonstrating an automatic load cut off before the crane load exceeds 3000 pounds.
13.8.1.12 DOST Sediment Cleaning Surveillance Requirements St. Lucie Technical Specification Amendment 233 dated July 28, 2016, removed Surveillance Requirement 4.8.1.1.2.g.1 related to fuel oil storage tank cleaning from the Technical Specifications and include fuel oil storage tank cleaning in the Updated Final Safety Analysis Report for St. Lucie Unit 1 which the licensee is required to control by the provisions set forth in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.59, "Changes, tests, and experiments."
The surveillance requirement for demonstrating the FUNCTIONALITY of the fuel oil systems for EC288179 standby diesel generators follows the guidance of Regulatory Guide 1.137, "Fuel Oil Systems for Standby Diesel Generators."
Applicability: At all times Surveillance Requirement:
Each diesel generator shall be demonstrated FUNCTIONAL by draining each Diesel Oil Storage EC288179 Tank, removing the accumulated sediment and cleaning the tank using an appropriate cleaning compound, with a frequency set at 15 years.
13.8.2 LINE-ITEM TECHNICAL SPECIFICATION REQUIREMENTS 13.8.2.1 Reactor Protective Instrumentation Response Times The response time surveillance test acceptance criteria for Technical Specification 3.4.3.1 was relocated by License Amendment No. 128 and NRC Safety Evaluation dated July 12, 1994.
The measurement of reactor protective instrumentation response times at the specified frequencies provides assurance that the protective function associated with each channel is completed within the time limit assumed in the accident analyses. No credit is taken in the analyses for those channels with response times indicated as not applicable.
UNIT 1 13.8-5b Amendment No. 29 (10/18)
13.8.2.1.1 Response Time Surveillance Test Requirements and Acceptance Criteria Reactor protective instrumentation response times may be demonstrated by any series of sequential, overlapping, or total channel test measurements, including allocated sensor response time, provided that such tests demonstrate total channel response time is within the limits specified in Table 13.8.2-1.
CEOG Topical Report CE NPSD-1167, and FPL No Significant Hazards Evaluation PSL-ENG-SEIS-03-043 provide the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in these documents.
The allocated sensor response time must be verified prior to placing a new component in operation and re-verified after maintenance that may adversely affect the sensor response time (e.g., replacement of a transmitter DP cell or variable damping circuits). Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times. The CEOG topical report and FPL evaluation only cover certain sensor model numbers.
If sensors are replaced with types not previously evaluated, then response time testing (RTT) for the new sensor must either be performed and the appropriate changes made to plant procedures, or an additional request for RTT elimination must be submitted and approved. If, however, the replacement sensor is one for which RTT elimination has been approved, then FPL may modify the plant procedures, using an allocated response time based upon a vendor-supplied response time value, or upon statistical analysis of historical data for that transmitter type and model.
13.8.2.2 Engineered Safety Features Actuation Systems Instrumentation Response Times The response time surveillance test acceptance criteria for Technical Specification 3.4.3.2 was relocated by License Amendment No. 128 and NRC Safety Evaluation dated July 12, 1994.
The measurement of engineered safety features actuation systems instrumentation response times at the specified frequencies provides assurance that the ESF action function associated with each channel is completed within the time limit assumed in the accident analyses. No credit is taken in the analyses for those channels with response times indicated as not applicable.
UNIT 1 13.8-5c Amendment No. 28 (05/17)
13.8.2.2.1 Response Time Surveillance Test Requirements and Acceptance Criteria Engineered safety features actuation system instrumentation response times may be demonstrated by any series of sequential, overlapping, or total channel measurements, including allocated sensor response time, provided that such tests demonstrate total channel response time is within the limits specified in Table 13.8.2-2.
CEOG Topical Report CE NPSD-1167, and FPL No Significant Hazards Evaluation PSL-ENG-SEIS-03-043 provide the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in these documents.
The allocated sensor response time must be verified prior to placing a new component in operation and re-verified after maintenance that may adversely affect the sensor response time (e.g., replacement of a transmitter DP cell or variable damping circuits). Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times. The CEOG topical report and FPL evaluation only cover certain sensor model numbers.
If sensors are replaced with types not previously evaluated, then response time testing (RTT) for the new sensor must either be performed and the appropriate changes made to plant procedures, or an additional request for RTT elimination must be submitted and approved. If, however, the replacement sensor is one for which RTT elimination has been approved, then FPL may modify the plant procedures, using an allocated response time based upon a vendor-supplied response time value, or upon statistical analysis of historical data for that transmitter type and model.
13.8.2.3 Flood Protection Plant operating restrictions associated with the flood protection were removed from the facility Technical Specifications by License Amendment No. 142 and NRC Safety Evaluation Report issued April 11, 1996.
The flood control provisions (dune and slope protection) shall be designed and maintained in accordance with the original design provisions contained in Section 2.4.2.2 of the FSAR.
The beach dunes, old beach road, and mangrove swamp provide additional assurance that safety-related structures are adequately protected during design basis flooding events.
13.8.2.3.1 Surveillance Requirements A visual inspection shall be conducted by a qualified engineer after every hurricane and at a minimum a visual inspection every 5 years if there has been no hurricane in that period. If the visual inspection finds the beach dune and the old beach road to have been breached, the inspection will be expanded to include an evaluation of the mangrove swamp and SR-A1A embankment adjacent to the site. If the expanded visual inspection finds significant erosion requiring repair of the SR-A1A roadbed or embankment, then FPL will consider the effects of the storm damage as part of the hurricane recovery activities.
13.8-6 Amendment No. 22 (05/07)
13.8.2.4 ESF Pump Acceptance Criteria and Test Methods The performance criteria values and test procedural details for Surveillance Requirements (SR) 4.1.2.5, "Boric Acid Pumps - Shutdown," 4.1.2.6, "Boric Acid Pumps - Operating," 4.5.2.f, "ECCS Subsystems -
Operating," and 4.6.2.1.b, "Containment Spray and Cooling Systems," were removed from the facility Technical Specifications by License Amendment No. 194 and NRC Safety Evaluation Report issued October 6, 2004.
The performance criteria values and test procedural details are relocated below.
13.8.2.4.1 Surveillance Requirements 4.1.2.5, "Boric Acid Pumps - Shutdown" and 4.1.2.6, "Boric Acid Pumps - Operating" The boric acid pump required by LCO 3.1.2.5 shall be demonstrated OPERABLE by verifying that the pump develops the specified discharge pressure tested pursuant to the Inservice Testing Program.
13.8.2.4.2 Surveillance Requirements 4.5.2.f, "ECCS Subsystems - Operating" Each ECCS subsystem required by LCO 3.5.2 shall be demonstrated OPERABLE:
- f. By verifying that each of the following pumps develops the required total developed head on recirculation flow when tested pursuant to the Inservice Testing Program.
- 1. High-Pressure Safety Injection pumps
- 2. Low-Pressure Safety Injection pumps 13.8.2.4.3 Surveillance Requirements 4.6.2.1.b, "Containment Spray and Cooling Systems" Each containment spray system required by LCO 3.6.2.1 shall be demonstrated OPERABLE:
- b. By verifying that each spray pump develops the required total developed head on recirculation when tested pursuant to the Inservice Testing Program.
13.8-6a Amendment No. 26 (11/13)
TABLE 13.8.1-1 SEISMIC MONITORING INSTRUMENTATION (Instrumentation located in St. Lucie Unit 1)
MINIMUM MEASUREMENT CHANNELS INSTRUMENT CHANNEL SENSOR LOCATION RANGE OPERABLE
- 1. STRONG MOTION TRIAXIAL ACCELEROGRAPHS
- a. SMR-42-1 R.B. Elev. 23.0' 0-2 g 1(Note 1)
- b. SMR-42-2 R.B. Elev. 62.0' 0-2 g 1
- c. SMR-42-3 R.A.B. Elev. -0.5' 0-2 g 1
- d. SMR-42-4 R.A.B. Elev. 43.0' 0-2 g 1
- e. SMR-42-5 Yard Elev. 19.5 0-2 g 1 NOTE:
- 1. With St. Lucie Unit 2 reactor control room alarm.
13.8-7 Amendment No. 21 (12/05)
TABLE 13.8.1-2 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS (Instrumentation located in St. Lucie Unit 1)
CHANNEL INSTRUMENTS AND CHANNEL CHANNEL FUNCTIONAL SENSOR LOCATIONS CHECK CALIBRATION TEST
- 1. STRONG MOTION TRIAXIAL ACCELEROGRAPHS
- a. SMR-42-1 M R SA
- b. SMR-42-2 M R SA
- c. SMR-42-3 M R SA
- d. SMR-42-4 M R SA
- e. SMR-42-5 M R SA 13.8-8 Amendment No. 22 (05/07)
TABLE 13.8.1-3 METEOROLOGICAL MONITORING INSTRUMENTATION INSTRUMENT MINIMUM MINIMUM CHANNELS INSTRUMENT AND ELEVATION ACCURACY OPERABLE
- 1. WIND SPEED a) Nominal Elev (10 meters) +/- 0.5 mph (NOTE 1) 1 (NOTE 3) b) Nominal Elev (57.9 meters) +/- 0.5 mph (NOTE 1) N.A.
- 2. WIND DIRECTION a) Nominal Elev (10 meters) +/- 5º 1 (NOTE 4) b) Nominal Elev (57.9 meters) +/- 5º N.A.
- 3. AIR TEMPERATURE (Delta T) a) Nominal Elev (10 meters) +/- 0.1ºC (NOTE 2) 1 (NOTE 5) b) Nominal Elev (57.9 meters) +/- 0.1º C (NOTE 2) 1 (NOTE 5) c) Nominal Elev (33.5 meters) +/- 0.1ºC (NOTE 2) N.A.
NOTES:
- 1. Starting speed of anemometer shall be <1 mph.
- 2. T measurement channels only.
- 3. The 57.9 meter channel may be substituted for the 10 meter wind speed for up to 30 days in the event the 10 meter channel is inoperable. Wind speed data from the 57.9 meter elevation should be adjusted using the wind speed power law:
S 10 meters = S57.9 meters (0.1727)n Where:
S = wind speed in mph n = 0.25 For Pasquill Vertical Stability Classes A, B, C, and D.
n = 0.50 For Pasquill Vertical Stability Classes E, F, and G.
1.727 x 10-1 = constant = 10 meters/57.9 meters.
- 4. The 57.9 meter channel maybe substituted for the 10 meter wind direction channel for up to 30 days in the event the 10 meter channel is inoperable.
- 5. The 33.5 meter channel may be substituted for one of the 10 meter or 57.9 meter temperature channels for up to 30 days if one of the channels is inoperable. The data should always be normalized to ºC/100 meters to determine the vertical stability class.
13.8-9 Amendment 15, (1/97)
TABLE 13.8.1-4 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION
- 2. WIND DIRECTION a) Nominal Elev. (10 meters) D SA b) Nominal Elev. (57.9 meters) D(NOTE 1) SA(NOTE 1)
- 3. AIR TEMPERATURE (Delta T) a) Nominal Elev. (10 meters) D SA b) Nominal Elev. (57.9 meters) D SA c) Nominal Elev. (33.5 meters) D(NOTE 1) SA(NOTE 1)
NOTES:
- 1. Required only if these channels are being substituted for one of the Minimum Channels OPERABLE per Table 13.8.1-3.
13.8-10 Amendment 15, (1/97)
TABLE 13.8.1-5 EXPLOSIVE GAS MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION
- 1. Waste Gas Decay Tanks Explosive Gas Monitoring System
- a. Oxygen monitors 1 (NOTE 1) 1 ACTION 1- With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, operation of the waste gas holdup system may continue, provided samples of oxygen are analyzed by the lab gas partitioner at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
NOTE:
- 1. During waste gas holdup system operation.
13.8-11 Amendment No. 19 (10/02)
TABLE 13.8.1-6 EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED
- 1. WASTE GAS TANKS EXPLOSIVE GAS MONITORING SYSTEM
- a. Oxygen Monitor D Q(1) M (NOTE 1)
- b. Oxygen Monitor D Q(1) M (NOTE 1)
(alternate)
(1) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
NOTE:
- 1. During waste gas holdup system operation.
13.8-12 Amendment 15, (1/97)
TABLE 13.8.1-7 LEFM CALORIMETRIC INSTRUMENTATION Functional Unit Total No. of Minimum Channels Operable Channels LEFM CPU 2 1 LEFM Meter Section (Path 1-4, 5-8) 4 4 Calorimetric Section of DCS 1 1 TABLE 13.8.1-8 REDUCED POWER LIMITS APPLICABLE TO INOPERABLE LEFM CALORIMETRIC INSTRUMENTATION Maximum Maximum % of Total Power Description of Inoperable LEFM Calorimetric Instrument MWt 3020 MWt Uncertainty 3015 99.84 0.46% Either Header A or Header B of LEFM in Check Mode (1) 2968 98.3 2.0% Both Header A and Header B of LEFM in Check Mode (1) 2968 98.3 2.0% Any of the two LEFM Meters Fail Mode (1) 2968 98.3 2.0% Calorimetric Section of DCS is Out Of Service Note (1): LEFM Check and Fail Modes are automatically determined within the LEFM system and are annunciated and illustrated on the LEFM CPU display screens.
13.8-12a Amendment No. 26 (11/13)
TABLE 13.8.1-9 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS STEADY STATE TRANSIENT PARAMETER LIMIT LIMIT DISSOLVED OXYGEN 0.10 ppm* 1.00 ppm*
CHLORIDE 0.15 ppm 1.50 ppm FLUORIDE 0.10 ppm 1.00 ppm
- Limit not applicable with Tavg 250°F.
TABLE 13.8.1-10 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS MINIMUM MAXIMUM TIME PARAMETER SAMPLING FREQUENCIES BETWEEN SAMPLES DISSOLVED OXYGEN 3 times per 7 days* 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CHLORIDE 3 times per 7 days 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> FLUORIDE 3 times per 7 days 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
- Not required with Tavg 250°F.
UNIT 1 13.8-12b Amendment No. 28 (05/17)
TABLE 13.8.2-1 REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME
- 1. Manual Reactor Trip Not Applicable
- 2. Power Level - High 0. 40 seconds(NOTES 1 & 2) and 8.0 seconds(NOTE 3)
- 3. Reactor Coolant Flow - Low 1.025 seconds
- 4. Pressurizer Pressure - High 0.90 seconds
- 5. Containment Pressure - High 1.40 seconds
- 6. Steam Generator Pressure - Low 0.90 seconds
- 7. Steam Generator Water Level - Low 0.90 seconds
- 8. Local Power Density - High 0.40 seconds(NOTES 1 & 2) and 8.0 seconds(NOTE 3)
- 9. Thermal Margin/Low Pressure 0. 90 seconds(NOTES 1 & 2) and 8.0 seconds(NOTE 3) 9a. Steam Generator Pressure Difference - High 0.90 seconds
- 10. Loss of Turbine - Hydraulic Fluid Pressure - Low Not Applicable
- 11. Wide Range Logarithmic Neutron Flux Monitor Not Applicable NOTES:
- 1. Neutron detectors are exempt from response time testing. Response time shall be measured from detector output or input of first electronic component in channel.
- 2. Response time does not include contribution of RTDs.
- 3. RTD response time only. This value is equivalent to the time interval required for RTDs output to achieve 63.2% of its total change when subjected to a step change in RTD temperature.
13.8-13 Amendment No. 26 (11/13)
TABLE 13.8.2-2 ENGINEERED SAFETY FEATURES RESPONSE TIMES RESPONSE TIME INITIATING SIGNAL AND FUNCTION IN SECONDS
- 1. Manual
- a. SIAS Safety Injection (ECCS) Not Applicable Containment Fan Coolers Not Applicable Feedwater Isolation Not Applicable Containment Isolation Not Applicable
- b. CSAS Containment Spray Not Applicable
- c. CIS Containment Isolation Not Applicable Shield Building Ventilation System Not Applicable
- e. MSIS Main Steam Isolation Not Applicable Feedwater Isolation Not Applicable
- f. AFAS Auxiliary Feedwater Actuation Not Applicable
- 2. Pressurizer Pressure-Low
- a. Safety Injection (ECCS) 30.0(NOTE 1)/19.5(NOTE 2)
- b. Containment Isolation(NOTE 3) 30.5(NOTE 1)/20.5(NOTE 2)
- c. Containment Fan Coolers 30.0(NOTE 1)/17.0(NOTE 2)
- d. Feedwater Isolation 60.0
- 3. Containment Pressure-High
- a. Safety Injection (ECCS) 30.0(NOTE 1)/19.5(NOTE 2)
- b. Containment Isolation(NOTE 3) 30.5(NOTE 1)/20.5(NOTE 2)
- c. Shield Building Ventilation System 30.0(NOTE 1)/14.0(NOTE 2)
- d. Containment Fan Coolers 30.0(NOTE 1)/17.0(NOTE 2)
- e. Feedwater Isolation 60.0 13.8-14 Amendment 15, (1/97)
TABLE 13.8.2-2 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES RESPONSE TIME INITIATING SIGNAL AND FUNCTION IN SECONDS
- 4. Containment Pressure-High-High
- a. Containment Spray 30.0(NOTE 1)/18.5(NOTE 2)
- 5. Containment Radiation-High
- a. Containment Isolation(NOTE 3) 30.5(NOTE 1)/20.5(NOTE 2)
- b. Shield Building Ventilation System 30.0(NOTE 1)/14.0(NOTE 2)
- 6. Steam Generator Pressure-Low
- a. Main Steam Isolation 6.9
- b. Feedwater Isolation 20.0
- 7. Refueling Water Storage Tank-Low
- a. Containment Sump Recirculation 91.5
- 8. Steam Generator Level-Low
- a. Auxiliary Feedwater 205(NOTE 2), 305(NOTE 1)
- 9. Auxiliary Feedwater Isolation
- 10. Loss of Power
- a. 4.16 kV Emergency Bus Undervoltage (Loss of Voltage) Not Applicable
- b. 4.16 kV Emergency Bus Undervoltage (Degraded Voltage) Not Applicable
- c. 480 V Emergency Bus Undervoltage (Degraded Voltage) Not Applicable NOTES:
- 1. Diesel generator starting and sequence loading delays included.
- 2. Diesel generator starting and sequence loading delays not included. Offsite power available.
- 3. Not applicable to containment isolation valve I-MV-18-1.
13.8-15 Amendment No. 18, (04/01)
13.9 NRC TS Amendment Related Commitments 13.9.1 Amendment 234, TSTF-422, Risk-Inform Requirements Regarding Selected Required Action End States The NRC SER for Amendment 234 required the following commitments be incorporated into the UFSAR:
- 1. On an ongoing basis, the licensee will follow the guidance established in Section 11 of NUMARC 93-01, Nuclear Management and Resource Council, Revision 4A, April 2011.
- 2. Upon implementation of the approved TS amendments, when TS required ACTION end state remains within the applicability of the TS, the licensee will follow the guidance established in WCAP-16364-NP, Revision 2, dated May 2010, with the exception that Section 11 of NUMARC 93-01, Revision 4A, will be utilized to meet 10 CFR 50.65(a)(4) requirements in lieu of NUMARC 93-01, Revision 3.
Both of these commitments are embodied in Procedure ADM-17.16, Implementation of the Configuration Risk Management Program.
UNIT 1 13.9-1 Amendment No. 29 (10/18)