L-2022-018, Amendment 27 to Updated Final Safety Analysis Report, Chapter 1, Introduction and General Description of Plant
| ML22111A124 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 03/04/2022 |
| From: | Florida Power & Light Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML22111A137 | List:
|
| References | |
| L-2022-018 | |
| Download: ML22111A124 (175) | |
Text
UFSAR/St. Lucie - 2 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT CHAPTER 1 TABLE OF CONTENTS Section Title Page
1.0 INTRODUCTION
AND GENERAL DESCRIPTION OF PLANT ...................... 1.1-1
1.1 INTRODUCTION
............................................................................................. 1.1-1 1.2 GENERAL PLANT DESCRIPTION ................................................................. 1.2-1 1.2.1 PRINCIPAL SITE CHARACTERISTICS ....................................................... 1.2-1 1.2.2 PRINCIPAL DESIGN CRITERIA .................................................................. 1.2-1 1.2.3 MAJOR STRUCTURES AND EQUIPMENT ARRANGEMENT.................. 1.2-13 1.2.4 SHARED SYSTEMS AND INTERCONNECTIONS BETWEEN UNIT 1 AND UNIT 2 ................................................................................... 1.2-14 1.2.5 SECURITY PLAN ....................................................................................... 1.2-15 1.2.6 EMERGENCY PLAN .................................................................................. 1.2-16 1.2.7 SYMBOLS AND ABBREVIATIONS ON FIGURES .................................... 1.2-16 1.
2.8 REFERENCES
FOR SECTION 1.2 ........................................................... 1.2-17 1.3 COMPARISONS.............................................................................................. 1.3-1 1.3.1 COMPARISONS WITH SIMILAR FACILITY DESIGNS ............................... 1.3-1 1.3.2 COMPARISON OF FINAL AND PRELIMINARY INFORMATION ................ 1.3-1 1.
3.3 REFERENCES
FOR SECTION 1.3 ............................................................. 1.3-5 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS ................................. 1.4-1 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION ................. 1.5-1 1.5.1 FRETTING AND VIBRATIONS TESTS OF FUEL ASSEMBLIES ................ 1.5-1 1.5.2 DEPARTURE FROM NUCLEATE BOILING (DNB) TESTING..................... 1.5.1 1.5.3 FUEL ASSEMBLY STRUCTURAL TESTS .................................................. 1.5-2 1.5.4 FUEL ASSEMBLY FLOW MIXING TESTS .................................................. 1.5-2 1.5.5 REACTOR FLOW MODEL TESTING AND EVALUATION .......................... 1.5-2 1.5.6 FUEL ASSEMBLY FLOW TESTS ................................................................ 1.5-3 1-i Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 CHAPTER 1 TABLE OF CONTENTS (Cont'd)
Section Title Page 1.5.7 CONTROL ELEMENT DRIVE MECHANISM (CEDM) TESTS..................... 1.5-3 1.5.8 DNB IMPROVEMENT .................................................................................. 1.5-3 REFERENCES ............................................................................................. 1.5-5 1.6 MATERIAL INCORPORATED BY REFERENCE ............................................ 1.6-1 1.7 DRAWINGS ..................................................................................................... 1.7-1 1.7.1 ELECTRICAL, INSTRUMENTATION, AND CONTROL ............................... 1.7-1 DRAWINGS 1.7.2 PIPING AND INSTRUMENTATION DIAGRAMS ......................................... 1.7-1 1.8 NRC REGULATORY GUIDES ........................................................................ 1.8-1 1.9 OTHER CONCERNS AND COMMITMENTS .................................................. 1.9-1 1.9.1 TMI ACTION PLAN ...................................................................................... 1.9-1 1.9.2 UNDERGROUND CABLE REVIEW ............................................................. 1.9-1 1.9.3 REPLACEMENT STEAM GENERATORS ................................................... 1.9-1 REFERENCES ............................................................................................. 1.9-2 1.9A TMI RELATED REQUIREMENTS ................................................................ 1.9A-1 REFERENCES ........................................................................................ 1.9A-13 1-ii Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 INTRODUCTION AND GENERAL DESCRIPTION OF PLANTS CHAPTER 1 LIST OF TABLES Section Title Page 1.3-1 PLANT PARAMETER COMPARISON ..........................................................T1.3-1 1.7-1 ARCHITECT/ENGINEER SUPPLIED ELECTRICAL, ...................................T1.7-1 INSTRUMENTATION AND CONTROL DRAWINGS SAFETY RELATED 1.7-2 NSSS SUPPLIED ELECTRICAL, INSTRUMENTATION AND....................T1.7-56 CONTROL DRAWINGS SAFETY RELATED 1.7-3 ARCHITECT/ENGINEER SUPPLIED FLOW DIAGRAMS, .........................T1.7-62 PIPING AND INSTRUMENTATION DIAGRAMS SAFETY RELATED 1.7-4 NSSS SUPPLIED FLOW DIAGRAMS, PIPING AND..................................T1.7-64 INSTRUMENTATION DIAGRAMS SAFETY RELATED 1.8-1 APPLICABLE NRC REGULATORY GUIDES ...............................................T1.8-1 1.9A-1 SAFETY RELATED VALVE POSITION AND POSITION .......................... T1.9A-1 INDICATION 1-iii Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT CHAPTER 1 LIST OF FIGURES Figures Title 1.2-1 Site Plan 1.2-2 Enlarged Plot Plan 1.2-3 General Arrangement Turbine Building - Ground Floor Plan 1.2-4 General Arrangement - Turbine Building 1.2-5 General Arrangement - Turbine Building - Operating Floor Plan 1.2-6 General Arrangement Turbine Building - Sections - Sheet 1 1.2-7 General Arrangement Turbine Building - Sections - Sheet 2 1.2-8 General Arrangement Reactor Building - Floor Plans Sheet 1 1.2-9 General Arrangement - Reactor Building Floor Plans Sheet 2 and Main Steam Trestle 1.2-10 General Arrangement Reactor Building - Sections Sheet 1 1.2-11 General Arrangement Reactor Building - Sections Sheet 2 1.2-12 General Arrangement Reactor Auxiliary Building Plan Sheet 1 1.2-13 General Arrangement Reactor Auxiliary Building Plan Sheet 2 1.2-14 General Arrangement Reactor Auxiliary Building Plan Sheet 3 1.2-15 General Arrangement - Reactor Auxiliary Building 1.2-16 General Arrangement - Fuel Handling Building - Plans 1.2-17 General Arrangement - Fuel Handling Building - Sections 1.2-18 General Arrangement - Reactor Auxiliary 1.2-19 General Arrangement Reactor Auxiliary Building Miscellaneous Plans and Sections 1.2-20 General Arrangement Component Cooling Water Area and Diesel Generator Building 1.2-21 General Arrangement - Component Cooling Area 1-iv Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 CHAPTER 1 LIST OF FIGURES (Contd)
Figures Title 1.2-22 General Arrangement - Intake Structure 1.2-23 Flow Diagram Symbols 1.2-24 Control and Block Diagram 1.2-25-33 FIGURES 1.2-25 THROUGH 1.2-33 HAVE BEEN DELETED 1.2-34 Flow Diagram Auxiliary Pumps 1-v Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2
1.0 INTRODUCTION
AND GENERAL DESCRIPTION OF PLANT
1.1 INTRODUCTION
This Updated Final Safety Analysis Report (UFSAR) is submitted in accordance with the requirements of 10 CFR 50.71(e). It is based on the original FSAR, including 14 amendments, which was submitted in support of an application by Florida Power & Light Company for a license to operate a nuclear power unit designated as St. Lucie Unit 2. The unit is located on Hutchinson Island in St. Lucie County about halfway between the cities of Fort Pierce and Stuart on the east coast of Florida.
This submittal contains updated information which is accurate for the period up to six months prior to the most recent revision of this document. The updated material is of the same level of detail presented in the original FSAR. It includes changes necessary to reflect information and analysis submitted to the NRC or prepared pursuant to Commission requirements, and it includes changes describing physical modifications to the plant.
Generally, the information provided in the original FSAR where no update is required is retained for historical purposes.
The original Nuclear Steam Supply System (NSSS) is a pressurized water reactor system designed by Combustion Engineering Incorporated. The containment structure is comprised of a steel containment vessel designed by Chicago Bridge & Iron Company, and is surrounded by a reinforced concrete Shield Building designed by Ebasco Services Incorporated.
The initial rating of the NSSS thermal power level was 2570 Mwt (including a 10 Mwt net heat addition from reactor coolant pumps). 2560 Mwt was the projected initial operating power of the core and the power at which the thermal and hydraulic aspects of the core had been analyzed.
The corresponding net electrical output for the rated power level was 802 Mwe. Subsequent to the Cycle 2 reload, St. Lucie Unit 2 requested and was granted a stretch power rating of 2700 Mwt. This corresponds to a net electrical output of 830 Mwe. The UFSAR has been modified, where necessary to reflect the revisions brought about by this increased power level. The design thermal power level is 2700 Mwt, the maximum expected output of the Nuclear Steam Supply System. This is the basis for the design of the balance of plant and related facilities, including the major systems and components, the Engineered Safety Features and for site radiological release calculations.
Prior to the Cycle reload, St. Lucie Unit 2 requested an extended power rating of 3020 Mwt, comprised of an approximate 10% Extended Power Uprate (EPU) and a 1.7% Measurement Uncertainty Recapture (MUR). This represents an approximate 11.85% increase from the stretch power rating of 2700 Mwt. The UFSAR has been modified, where necessary, to reflect revisions brought about by this increased power level.
Both original steam generators (OSGs) were removed and replacement steam generators (RSGs) designed and manufactured by AREVA were installed. The effect of the RSG installation on the information provided in the UFSAR is specifically noted in the affected sections.
1.1-1 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 1.2 GENERAL PLANT DESCRIPTION 1.2.1 PRINCIPAL SITE CHARACTERISTICS The site for St. Lucie Units 1 and 2 consists of approximately 1,132 acres. The unimproved area of the site is generally flat, covered with water and has a dense vegetation characteristic of Florida coastal mangrove swamps. At the ocean shore the land rises slightly in a dune or ridge to approximately 15 ft, above mean low water.
The island and the adjoining mainland are sparsely populated. The southern most boundary of the nearest population center is the City of Fort Pierce which is 4.1 miles from the site. The City of Fort Pierce has an estimated population of 33,083 people as of a 1978 estimate. The minimum site exclusion radius is 5,100 ft. Site characteristics are given in Chapter 2.
1.2.2 PRINCIPAL DESIGN CRITERIA Principal structures, system and equipment which may serve either to prevent accidents or to mitigate their consequences are designed and erected in accordance with applicable codes to withstand the most severe earthquakes, flooding conditions, windstorms, temperature and other deleterious natural phenomena which could be reasonably assumed to occur at the site during the lifetime of the plant. Principal structures, systems and equipment are sized for the design power level of the nuclear steam supply system output.
Redundancy is provided in the reactor protective and engineered safety feature systems so that no single failure of any active component of the systems can prevent action necessary to avoid an unsafe condition. The plant is designed to facilitate inspection and testing of systems and components whose reliability are important to plant shutdown and to the protection of the public and plant personnel.
Provisions are made to minimize the probability and effect of fires and explosions, in accordance with 10 CFR 50.48(c), NFPA 805.
Systems and components which are significant from the standpoint of nuclear safety are designed, fabricated and erected to quality standards commensurate with the safety function to be performed.
Section 3.1 addresses the implementation of the NRC General Design Criteria for Nuclear Power Plants, 10 CFR Part 50, Appendix A. Chapter 17 describes the quality assurance program for the design and operation of St. Lucie Unit 2.
1.2.2.1 Reactor The reactor is of the pressurized water-type, designed to provide heat to steam generators which, in turn, provide steam to drive a turbine generator. The full power core thermal output is 3020 megawatts.
The reactor core is fueled with UO2 and UO2-Gd2 O3 and/or UO2 Er2O3 pellets enclosed in zircaloy tubes pressurized with helium and fitted with welded end plugs.
1.2-1 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 The tubes are fabricated into assemblies in which end fittings prevent axial motion and spacer grids prevent lateral motion of the tubes. Beginning with Region N, the fuel incorporates the GUARDIANTM fuel assembly design to screen and entrap debris. The GUARDIANTM design employs a redesigned bottom spacer grid that provides positive axial restraint to the rods and added screening features. Region N also includes the addition of backup arches adjacent to all cantilevered springs in the interior of the upper H1D-1L spacer grid or top Inconel grid (beginning with Region U). The backup arch limits the possible compression of the grid spring, and thereby better maintains the proper geometry between the grid support features and the fuel rod during fabrication and operation. This same feature was present in peripheral locations in each Zircaloy spacer grid for all previous St. Lucie 2 fuel batches. In these locations, the backup arches protect the grid springs that may be subject to compression during fuel handling, when peripheral fuel rods can be pressed inward as bowed fuel assemblies are slid past one another in the core. In the new upper grid design, the arches will be present at all 440 interior spring locations in the grid. The backup arches will thus limit compression of grid springs in all interior locations during fuel rod loading. The control element assemblies (CEAs) consist of inconel clad boron carbide absorber rods which are guided by zircaloy tubes located within the fuel assembly. The core consists of 217 fuel assemblies.
Beginning with Cycle 23, the feed fuel is of the AREVA CE-16 HTP' fuel design.The AREVA CE-16 HTP' fuel consists of a 16x16 assembly configuration with M5 clad fuel rods, Zircaloy-4 MONOBLOC' Corner Guide tubes, an Alloy 718 HMP' spacer at the lowermost axial elevation, Zircaloy-4 HTP' spacers in all other axial elevations, a FUELGUARD' lower tie plate, and the AREVA reconstitutable upper tie plate.
The M5 Zirconium alloy has been consistently shown to provide superior corrosion resistance and growth performance.The robust FUELGUARD' lower tie plate provides highly effective debris resistance with good flow characteristics and an acceptable pressure drop.The HTP' spacer grid design has shown improved protection against fuel rod fretting damage and increased structural integrity of fuel assemblies. The bottom HMP' spacer grid design made from Alloy 718 reduces cell relaxation during irradiation to prevent fuel rod movement.The MONOBLOC' corner guide tubes increases the wall thickness in the bottom approximately 14 inches of the guide tube, slightly increasing fuel assembly stiffness.
The AREVA CE-16 HTP' fuel has been approved by the USNRC for implementation at St. Lucie Unit 2 as described in the Safety Evaluation (SE) in Reference 4.
Minimum departure from nucleate boiling ratio (DNBR) during normal operation and anticipated operational occurrences is not less than 1.28 (cycle 1 was 1.19) using the CE-1 correlation. The maximum center line temperature of the fuel, evaluated at the design overpower condition, is below that value which could lead to fuel rod failure. The melting points of the UO2 and UO2-Gd2O3 and/or UO2-Er2O3 are not reached during routine operation and anticipated operational occurrences. For the AREVA CE16 HTP' fuel, Minimum DNBR during normal operation and anticipated operational occurrences is not less than the correlation safety limit using the HTP correlation.
The combined response of the fuel temperature coefficient, the moderator temperature coefficient, the moderator void coefficient and the moderator pressure coefficient to an increase in reactor thermal power is a decrease in reactivity. In addition, the reactor power transient remains bounded and damped in response to any expected changes in any operating variable.
1.2-2 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 Control element assemblies (CEAs) are capable of holding the core sub-critical at hot zero power conditions with margin following a trip even with the most reactive CEA stuck in the fully withdrawn position.
Fuel rod clad is designed to maintain cladding integrity throughout fuel life. Fission gas release within the rods and other factors affecting design life are considered for the maximum expected exposures.
The reactor and control systems are designed so that any xenon transients are adequately damped.
The reactor in conjunction with the Reactor Protective System is designed to accommodate safely and without fuel damage, the anticipated operational occurrences.
The reactor vessel and its closure head are fabricated from manganese molybdenum nickel steel internally clad with austenitic stainless steel. The vessel and its internals are designed so that the integrated neutron flux does not exceed 4.9 x 1019 n/cm2 (E > 1 Mev) over the 60 year design life of the vessel.
Power excursions which could result from any credible reactivity addition do not cause damage, either by deformation or rupture of the reactor vessel and do not impair operation of the Engineered Safety Features.
The internal structures include the core support barrel, the lower support structure, the core shroud, the hold-down ring and the upper guide structure assembly. The core support barrel is a right circular cylinder supported from a ring flange from a ledge on the reactor vessel. The flange carries the entire weight of the core. The lower support structure transmits the weight of the core to the core support barrel by means of vertical columns and a beam structure. The core shroud surrounds the core and limits the amount of coolant bypass flow. The upper guide structure provides a flow shroud for the CEAs and prevents upward motion of the fuel assemblies during pressure transients. Lateral motion limiters or snubbers are provided at the lower end of the core support barrel assembly. The hold-down ring acts as a shim and is set between the reactor vessel head and the upper guide structure to resist axial upward movement.
Further details concerning the reactor are given in Chapters 3 and 4.
1.2.2.2 Reactor Coolant and Auxiliary Systems The Reactor Coolant System is arranged as two closed loops connected in parallel to the reactor vessel. Each loop consists of one 42 in. ID outlet (hot) pipe, one steam generator, two 30 in. ID inlet (cold) pipes and two reactor coolant pumps. An electrically heated pressurizer is connected to the hot leg of one of the loops and a safety injection line is connected to each of the four cold legs.
The Reactor Coolant System operates at a nominal pressure of 2235 psig. The reactor coolant enters near the top of the reactor vessel, and flows downward between the reactor vessel shell and the core support barrel into the lower plenum. It then flows upward through the core, leaves the reactor vessel, and flows through the tube side of the two vertical U-tube steam generators where heat is transferred to the secondary system. Reactor coolant pumps return the reactor coolant to the reactor vessel.
1.2-3 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 The two steam generators are vertical shell and U-tube units. The steam generated in the shell side of the steam generator flows upward through moisture separators and scrubber plate dryers which reduce the moisture content to less than 0.2 percent. All surfaces in contact with the reactor coolant are either stainless steel or NiCrFe alloy in order to minimize corrosion.
The reactor coolant is circulated by four electric motor driven single suction vertical centrifugal pumps. The pump shafts are sealed by mechanical seals. Each pump motor is equipped with an antireverse mechanism to prevent reverse rotation.
Components of the Reactor Coolant System are designed and operated so that no stresses are imposed on the structural materials that result in loss of function. The necessary consideration has been given to the ductile characteristics of the materials at low temperatures.
The Reactor Coolant System is designed and constructed to maintain its integrity throughout the plant life. Appropriate means of test and inspection are provided.
See Chapter 5 for further information.
1.2.2.3 Engineered Safety Features The plant design incorporates redundant Engineered Safety Features. These systems in conjunction with the containment system, ensure that the offsite radiological consequences following any LOCA up to and including a double ended break of the largest reactor coolant pipe do not exceed the guidelines established for design basis accidents. The systems also ensure that the guidelines of 10 CFR 50 Appendix K, "Acceptance Criteria for Emergency Core Cooling Systems" are satisfied, based upon analytical methods, assumptions and procedures accepted by the NRC. The Engineered Safety Features include: (a) independent redundant systems (Containment Cooling System and Containment Spray System) to remove heat from and reduce the pressure in the containment vessel in order to maintain containment integrity, (b) a high and low pressure Safety Injection System to limit fuel and cladding damage to an amount which does not interfere with adequate emergency core cooling and to limit metal-water reactions to negligible amounts, (c) a Shield Building Ventilation System to reduce offsite consequences due EC291393 to leakage from the containment vessel, (d) a containment isolation system to minimize post-LOCA radiological effects offsite, (e) a hydrogen control system to maintain safe post-LOCA hydrogen concentration within the containment, and (f) a control room habitability system.
The Reactor Building, which is a dual containment design, is comprised of a steel containment vessel surrounded by an annular space and enclosed by a reinforced concrete Shield Building.
The containment vessel is a low leakage steel shell which is designed to confine the radioactive material that could be released from a postulated design basis, Loss-of-Coolant Accident, (LOCA). It is a cylindrical vessel with hemispherical dome and ellipsoidal bottom. The Shield Building is a medium leakage concrete structure which surrounds the annulus and steel containment vessel. It protects the containment vessel from external missiles, and provides biological shielding and a means of collecting radioactive fission products that may leak from the containment following a major hypothetical accident (see Subsection 6.2.1 for details).
The containment in conjunction with either of the associated spray and cooling systems is designed to withstand the internal pressure and coincident temperature resulting from the energy release associated with the design basis accident. The containment is equipped with two 100 percent capacity heat removal systems, each comprised of one containment spray loop and two containment cooling units.
1.2-4 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 The Containment Spray System supplies borated water to cool and reduce pressure in the containment atmosphere. The pumps take suction initially from the refueling water tank. Long term cooling is based on suction from the containment sump through the recirculation lines.
The Containment Cooling System provides containment atmosphere mixing by recirculation.
The cooling coils and fans of the Containment Cooling System are sized to provide adequate containment cooling at post-accident conditions of temperature, pressure and humidity (see Subsection 6.2.2 for details).
In the event of a LOCA, the Safety Injection System described in Section 6.3 injects borated water into the Reactor Coolant System. This provides cooling to limit core damage and fission product release, and assures adequate shutdown margin. The injection system also provides continuous long term post-accident cooling of the core by recirculation of borated water from the containment sump through the shutdown heat exchangers and back to the reactor core.
The Shield Building Ventilation System is provided to maintain a negative pressure in the annulus between the steel containment vessel and the concrete Shield Building following a LOCA. Two independent 100 percent capacity systems are provided. This system filters any radioactivity leakage from the containment vessel and therefore reduces the effects on the environment (see Subsection 6.2.3 for details). The SBVS is provided with carbon absorbers for iodine removal in the Shield Building. EC291393 A containment isolation system consisting of valves and associated actuators and controls is provided for each line penetrating the containment that must be closed to prevent a radioactivity release in the case of a loss-of-coolant accident (see Subsection 6.2.4 for details).
A hydrogen control system is provided which consists of a hydrogen sampling system. A EC295353 hydrogen purge system is provided as a non-safety, diverse system (see Subsection 6.2.5 for details).
The control room habitability system is provided to limit control room doses from airborne activity to within GDC 19 limits (see Section 6.4 for details).
1.2.2.4 Protection, Control, Instrumentation and Electrical Systems a) Reactor Protective System The reactor parameters are maintained within the acceptable limits by the inherent characteristics of the reactor, by the Reactor Regulating System, by boron in the moderator and by the operating procedures. In addition in order to preclude unsafe conditions for plant equipment or personnel, the Reactor Protective System initiates reactor trip if a selected parameter reaches its preset limit. Four independent channels normally monitor each of the selected plant parameters. The Reactor Protective System logic initiates protective action whenever the signal of any two of three channels reaches the preset limit. A fourth channel is provided as a spare and allows bypassing of one channel while maintaining a two-out-of-three system. If any two channels receive coincident signals, the power supply to the magnetic jack control element drive mechanisms is interrupted releasing the control elements to drop into the core to shutdown the 1.2-5 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 reactor. Redundancy is provided in the Reactor Protective System to assure that no single failure prevents protective action when it is required. The protective system is completely independent of and separate from the control system (see Section 7.2 for details).
b) Control System The reactor is controlled by a combination of control element assemblies (CEAs) and dissolved boric acid in the reactor coolant. Boric acid is used for reactivity changes associated with large but gradual changes in water temperature, core xenon, fuel burnup and power levels. Additions of boric acid also provide an increased shutdown margin during the initial loading and subsequent refuelings.
The boric acid solution is prepared and stored at a temperature sufficiently high to prevent precipitation. CEA movement provides changes in reactivity for shutdown or power changes. The CEAs are actuated by control drive mechanisms mounted on the reactor vessel head. The control drive mechanisms are designed to permit rapid insertion of the CEAs into the reactor core by gravity. CEA trip motion can be initiated manually or automatically.
The Reactor Regulating System (RRS) was designed to control reactivity to maintain the programmed reactor coolant temperature and power level which includes the capability to load follow. The RRS was designed to match the Nuclear Steam Supply System capability of following a ramp change from 15 percent to 100 percent power at a rate of five percent per minute and at greater rates over smaller load change increments up to a step change of 10 percent.
A RRS temperature controller is used to compare the existing average reactor coolant temperature with the value corresponding to the power called for by the temperature control program. If the temperature is different, the CEAs are manually adjusted to bring the two temperatures within the prescribed control band. Regulation of the reactor coolant temperature in accordance with this program maintains the secondary steam pressure within operating limits and matches reactor power to load demand.
The CEAs are moved through manual operation by the operator.
The pressure in the Reactor Coolant System is controlled by regulating the temperature of the coolant in the pressurizer, where steam and water are held in thermal equilibrium. Steam is formed by the pressurizer heaters or condensed by the pressurizer spray to reduce variations caused by expansion and contraction of the reactor coolant temperature changes. The pressure and water level control systems are described in Subsection 7.7.1.1.
Overpressure protection of the Reactor Coolant System is provided by power operated relief valves and spring loaded safety valves connected to the pressurizer. The discharge from the pressurizer safety and relief valves is 1.2-6 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 released under water in the pressurizer quench tank, where it is condensed and cooled. In the event the discharged steam exceeds the capacity of the tank, the tank relieves to the containment atmosphere (see Subsections 5.2.2, 5.2.6, and 5.4.13 for details).
A Turbine Control System is provided to regulate steam flow to the turbine as a function of system load. In the event of turbine trip, bypass systems are provided to release steam to the condenser and to the atmosphere. These systems are designed to reduce the sensible heat in the Reactor Coolant System, maintain the steam generator pressure during hot standby, and meet the original design basis of 45 percent steam bypass capability to mitigate challenges to the pressurizer and steam generator safety valves (see Section 7.7).
A Steam Generator Water Level Control System regulates feedwater flow to the steam generator (see Subsection 7.7.1.1). An Auxiliary Feedwater System is provided to ensure flow to the steam generators during plant startup, plant shutdown, and in the event of a plant design basis accident.
c) Instrumentation System The nuclear instrumentation includes excore and incore neutron flux detectors.
Twelve channels of excore instrumentation monitor the neutron flux and provide reactor protection and control signals during start up and power operation. Four of the channels are wide range logarithmic safety channels to measure neutron flux from source range to 200 percent of full power. Another four channels are power range safety channels to measure neutron flux linearly from one percent to 200 percent of full power. The power range safety channels are used by the reactor protection system to determine the neutron flux power and axial offset, and by the high power bypass circuitry for the high rate-of-change of power trip (see Subsection 7.2.1.1). There are two linear power range channels utilized for control purposes and two channels for startup and extended shutdown (see Subsection 7.7.1.1.9).
The original feedwater flow and temperature instrumentation consisting of venturis, differential pressure indications and Resistance Temperature Detectors (RTDs) has been supplemented by the installation of a Cameron/Caldon Leading Edge Flow Meter (LEFM) Checkplus system. This change supports the MUR 1.7% increase in core thermal power. The original feedwater flow and temperature instrumentation was retained and is used for companson monitoring of the LEFM system and as a backup feedwater mass flow measurement when needed (see Subsection 7.7.4).
The incore instrumentation consists of self-powered rhodium neutron detectors and background detectors to provide information on neutron flux distribution.
1.2-7 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 The process instrumentation monitoring includes those critical channels which are used for protective action. Temperature, pressure, flow and liquid level monitoring is provided, as required, to keep the operating personnel informed of plant conditions and to provide information from which plant processes can be evaluated and/or regulated.
Instrument signals transmitted from the containment are electric. Instrument signal transmission for the remaining plant instruments is either electric or pneumatic (see Chapter 7 for details).
The plant gaseous and liquid effluents are monitored to assure that they are maintained within acceptable radioactivity limits. Activity levels are displayed and off- normal values are annunciated. Area monitoring stations measure radioactivity at selected locations in the plant for personnel protection. A complete description of the radiation instrumentation is contained in Section 11.5 and Subsection 12.3.4.
d) Electrical System Redundant sources of offsite power are provided by four separate transmission lines.
The unit includes a 1,200 MVA, 0.9 power factor generator delivering power to a 230 kV switchyard through step-up power transformers. Auxiliary power is utilized at 6.74 kV (a 6.9 kV winding is provided for the start up transformers),
4.16 kV, 480V, and 120V ac; 125V dc systems are also provided. For emergency power, Engineered Safety Features control, and essential nuclear instrumentation, all voltages except 6.74 kV are provided.
The auxiliary load is normally supplied from two auxiliary transformers connected to the main generator bus. Start up power is supplied from two start up transformers connected to the 230 kV switchyard. Emergency power for the Engineered Safety Features is supplied by redundant diesel generator sets (see Chapter 8 for details).
1.2.2.5 Power Conversion System The power conversion system removes heat energy from the reactor coolant in two U-tube steam generators, and converts the steam into electrical energy by means of a turbine-generator. The unusable heat in the steam cycle is transferred to the main condenser for rejection by the Circulating Water System. The resulting condensate is deaerated in the condenser, then heated through feedwater heaters and returned to the steam generators as feedwater.
The turbine generator is a Siemens Energy Inc. unit. It is an 1,800 rpm tandem-compound, four-flow exhaust unit. The feedwater pumps are electric motor driven. Each of two strings of feedwater heaters consists of four low pressure and one high pressure heaters.
1.2-8 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 The Auxiliary Feedwater System contains two electric motor driven pumps and one pump driven by a noncondensing steam turbine. This system provides a source of water inventory to the steam generators during plant startup and hot standby, and during plant cooldown provides heat removal to bring the Reactor Coolant System to the shutdown cooling system activation window. (See Chapter 10 for details.)
1.2.2.6 Fuel Handling and Storage Systems The fuel handling systems provide for the safe handling of fuel assemblies and control element assemblies under all foreseeable conditions and for the required assembly, disassembly, and storage of the reactor vessel head and internals. These systems include a refueling machine located inside containment above the refueling cavity, the fuel transfer carriage, the upending machine, the fuel transfer tube, a spent fuel handling machine in the Fuel Handling Building, and various devices used for handling the reactor vessel head and internals (see Subsection 9.1.4 for details). Dry storage of spent fuel is provided as discussed in Section 1.2.2.9.
New fuel is stored dry in vertical racks in the Fuel Handling Building. The rack and fuel assembly spacing precludes criticality (see Subsection 9.1.1 for details).
The spent fuel pool is a reinforced concrete structure, stainless steel lined. Spent fuel assemblies are stored in vertical racks. Spacing between fuel assemblies is such that the effective neutron multiplication factor (keff) will remain less than 1.0 for non-accident conditions when no credit is taken for the boron in the pool water (see Subsections 9.1.2 and 9.1.3.3.2 for details). As discussed in Subsection 9.1.2.3, partial credit is taken for the negative reactivity of soluble boron in fuel pool water during certain postulated accidents.
Cooling and purification equipment is provided for the fuel pool water. This equipment may also be used for cleanup of refueling water after each fuel change in the reactor (see Subsection 9.1.3 for details).
1.2.2.7 Cooling Water and Other Auxiliary Systems a) Chemical and Volume Control System The purity level in the Reactor Coolant System is controlled by continuous purification of a bypass stream of reactor coolant. Water removed from the Reactor Coolant System is cooled in the regenerative heat exchanger. From there, the coolant flows to the letdown heat exchanger and then through a filter and a demineralizer where corrosion and fission products are removed. It is then sprayed into the volume control tank and returned to the regenerative heat exchanger by the charging pumps where it is heated prior to return to the Reactor Coolant System.
The Chemical and Volume Control System automatically adjusts the amount of reactor coolant in order to maintain a constant level in the pressurizer. This compensates for changes in specific volume due to coolant temperature changes and reactor coolant pump shaft controlled seal leakage (see Subsection 9.3.4 for details).
1.2-9 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 The Chemical and Volume Control System is capable of adding boric acid to the reactor coolant at a rate sufficient to maintain an adequate shutdown margin during Reactor Coolant System cooldown at the maximum design rate following a reactor trip. The system is independent of the CEA system.
b) Shutdown Cooling System The Shutdown Cooling System is used to reduce the temperature of the reactor coolant at a controlled rate and to maintain the proper reactor coolant temperature during refueling.
The Shutdown Cooling System utilizes the low pressure safety injection pumps to circulate the reactor coolant through two shutdown heat exchangers, returning it to the Reactor Coolant System through the low pressure injection header (see Subsection 5.4.7).
The Component Cooling System serves as a heat sink for the shutdown heat exchangers.
c) Sampling System Two sampling systems are provided; one for the reactor coolant and its auxiliary systems and one for the turbine steam and feedwater system. These systems are used for determining both chemical and radiochemical conditions of the various process fluids used in the plant (see Subsection 9.3.2).
d) Cooling Water Systems The turbine generator condenser is cooled by the Circulating Water System which takes suction from and discharges to the Atlantic Ocean.
An Intake Cooling Water System provides seawater from the Circulating Water System intake structure and serves as a heat sink for the component cooling water heat exchangers, the Turbine Closed Cooling System heat exchangers and the blowdown system open cooling water heat exchangers.
The Component Cooling Water System, consisting of three pumps and two heat exchangers, removes heat from the various auxiliary systems. Corrosion inhibited demineralized water is circulated by the system through auxiliary components of the Nuclear Steam Supply System that require cooling water.
During reactor shutdown, component cooling water is also circulated through the shutdown heat exchangers. The Component Cooling Water System provides an intermediate barrier between the Reactor Coolant System and the Intake Cooling Water System (see Subsection 9.2.2 for details).
The blowdown system closed cooling water heat exchangers remove heat from the steam generator blowdown. This heat is, in turn, removed by the intake cooling water by the open blowdown cooling water system heat exchangers.
1.2-10 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 The Turbine Closed Cooling Water System removes heat from the turbine generator oil cooler, hydrogen coolers, feed pump oil coolers, sample coolers, and other components by providing corrosion inhibited demineralized water to those components (see Section 9.2 for details).
e) Plant Ventilation Systems Separate ventilation systems are provided for the containment vessel, the control room, the Reactor Auxiliary Building, the Fuel Handling Building, Turbine Building, CCW structure, intake structure, and the Diesel Generator Building.
Two purge systems are provided for the containment atmosphere (see Section 9.4).
The annular space between the steel containment vessel and the concrete Shield Building is evacuated by the Shield Building Ventilation System utilizing charcoal filters for removal of radioactive iodine. This system is automatically put into operation upon receipt of a containment isolation actuation signal following a LOCA (see Subsection 6.2.3).
f) Plant Fire Protection System The Fire Protection System, common to St. Lucie Units 1 and 2, supplies water to fire hydrants, deluge systems and hose racks in the various areas of the plant.
Additional design features are provided throughout the plant to ensure conformance to 10 CFR 50 Appendix A, GDC 3 and 10 CFR 50.48(c),
NFPA 805. (See Subsection 9.5.1 and the Fire Protection Design Basis Document (Reference 5).)
g) Compressed Air System The Compressed Air System supplies properly conditioned compressed air required to operate pneumatic instruments and controls, operate containment isolation valves and perform normal plant maintenance. It consists of the Instrument Air System, which supplies the various air operated valves, pneumatic instruments and controls, and the Station Air System which supplies various outlets throughout the plant.
Multiple compressor units and a cross-connection are provided between the Instrument and Station Air Systems. In case of loss of instrument air, all safety related pneumatically operated devices in the plant are designed to fail in a position which would allow safe shutdown. Where safety class valves are required to operate, accumulators are provided (see Subsection 9.3.1).
h) Diesel Generator Fuel Oil Storage and Transfer System The Diesel Generator Fuel Oil System is provided to transfer diesel fuel oil from the onsite storage tanks to the day tanks which supply the emergency diesel 1.2-11 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 generator sets. Redundant subsystems are provided, capable of supplying sufficient fuel to their respective diesel generator sets, 1.2.2.8 Radioactive Waste Management System The Waste Management System provides the means for controlled handling, storage and disposal of liquid, gaseous and solid wastes. The principal design criterion is that plant personnel and the general public are protected by ensuring that all normal operating releases of radioactive material are made as low as reasonably achievable in accordance with the provisions of 10 CFR 50, Appendix I.
Reactor coolant from the Chemical and Volume Control System and from the reactor drain tank is processed by the boron management subsystem as described in Section 11.2.2.1.
Miscellaneous liquid wastes from the Reactor Auxiliary Building are collected in the equipment and chemical drain tanks and subsequently processed by the liquid waste subsystem as described in Section 11.2.2.2.
Waste gases are handled by the Gaseous Radwaste Treatment System. In this system, waste gases may be compressed and stored in the gas decay tanks which have a 30 day storage capacity or the gaseous effluent may be directly released to the plant vent if its activity level is sufficiently low. After decay, the gas in the waste gas decay tanks is sampled to ensure radioactivity levels are within acceptable limits, and is then released to the plant vent at a controlled rate.
Spent ion exchange resins and filters can be temporarily stored in high intensity containers (HICs) within the low level waste storage facility and ultimately transported in a shielded container from the plant.
Low activity wastes such as contaminated laundry, rags and paper are compacted and containerized for removal from the plant (see Chapter 11 for details).
1.2.2.9 Independent Spent Fuel Storage Installation (ISFSI)
An Independent Spent Fuel Storage Installation (ISFSI) has been constructed on the St. Lucie site to provide Unit 1 and Unit 2 spent fuel storage capacity through the current end of extended plant lives and to provide the storage required to facilitate decommissioning of the plant. The ISFSI provides the capability to store St. Lucie spent nuclear fuel, high-level radioactive waste, and reactor-related Greater Than Class C (GTCC) waste into dry storage casks.
The ISFSI is licensed under the General License provided to power reactor licensees under 10 CFR 72.210. ISFSI information is provided in References 1, 2, and 3. Therefore, only brief descriptions of the ISFSI are provided herein.
ISFSI soil improvements and construction changes have been evaluated and do not adversely affect safe plant operation. The ISFSI storm water management system limits storm water runoff to pre-construction levels. Other design and environmental effects of the ISFSI have been evaluated to ensure there are no adverse effects on safe plant operation.
1.2-12 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 1.2.2.10 Low Level Waste Storage Facility (LLWSF)
Due to the uncertainty of availability of offsite disposal options, a Low Level Waste Storage Facility (LLWSF) has been constructed on the site to provide interim low level waste storage capability for both St. Lucie units 1 and 2. Conservatively, both units produce a combined 840 cu. ft. of Class B/C low level radioactive waste (LLW) per year. This amount would fill approximately seven (7) 8-120 High Integrity Containers (HICs) per year. The LLWSF is designed to safely store five (5) years of LLW (36 HICs) within an array of concrete shields inside the precast panel concrete building.
The storage of Low Level Waste is licensed under the General License provided to power reactor licensees under 10 CFR Part 50.
The construction/implementation of the LLWSF including associated soil improvements have been evaluated and do not adversely affect safe plant operation. The existing storm water management system has the capacity to meet Florida Department of Environmental Protection requirements. Other design and environmental effects of the LLWSF have been evaluated to ensure there are no adverse effects on safe plant operation.
1.2.3 MAJOR STRUCTURES AND EQUIPMENT ARRANGEMENT Refer to the Site Plan, Figure 1.2-1, and the Enlarged Plot Plan, Figure 1.2-2, for the site general layout including the ISFSI site. The plant structures arrangement plans and sections are shown on Figures 1.2-3 through 1.2-22.
The Turbine Building is oriented parallel to State Road A1A and the shoreline of the Atlantic Ocean, with the Reactor Building located on the east, or seaward, side of the Turbine Building.
The Reactor Auxiliary Building is located perpendicular to and east of the Turbine Building, oriented in an east-west direction. The Fuel Handling Building is located east of the Reactor Building and the Reactor Auxiliary Building, oriented in a north-south direction.
The Reactor Containment Building encloses the steel containment structure, which houses the Nuclear Steam Supply System consisting of the reactor, steam generators, reactor coolant pumps, pressurizer, and other reactor auxiliaries. The containment structure is served by a polar bridge crane.
The Reactor Auxiliary Building houses the waste management facilities, Engineered Safety Features, heating and ventilating system components, electrical equipment, laboratories, offices, laundry and control room.
The Fuel Handling Building contains the spent fuel pool and new fuel storage facilities, as well as the cooling equipment for the fuel pool. The fuel is transferred from the Reactor Building to the Fuel Handling Building through the fuel transfer tube.
The Turbine Building houses the turbine generator, condensers, feedwater heaters, condensate and feedwater pumps, turbine auxiliaries and electrical switchgear assemblies and other electrical distribution systems which are non-Class 1E.
1.2-13 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 1.2.4 SHARED SYSTEMS AND INTERCONNECTIONS BETWEEN UNIT 1 AND UNIT 2 Normal plant shutdown requires the operation of several auxiliary systems, none of which are normally used by both units.
The following is a list of systems interconnected (one complete system on each unit which may, under certain conditions, be used by the other unit) between St. Lucie Units 1 and 2:
a) condensate storage tanks (AFW pump suction inter-tie),
b) diesel generator fuel oil storage and transfer system, c) station blackout cross-tie, d) liquid waste management system, e) instrument air system, f) station service air system, and g) startup transformers.
A tie between the two units has been provided from the Unit 2 condensate storage tank to the Unit 1 auxiliary feedwater pump's suction for a backup tornado missile protected water supply.
This cross-tie is normally isolated. The valve alignment assures that the minimum quantity of water required for safe shutdown is maintained at all times in both tanks.
The diesel generator fuel oil storage and transfer system has a seismic Category I interconnecting tie line between St. Lucie Units 1 and 2. Seismic Category I locked closed isolation valves assure that the tie line is opened only after administrative approval has been obtained.
In the event of a total loss of AC power, both onsite and offsite, (i.e., station blackout) power can be transferred from the non-blacked out unit's emergency diesel generator set via the station blackout tie to one of the blacked-out unit's redundant Class 1E electrical distribution trains.
Plant procedures limit the amount of the power transferred so as not to affect the non-blacked out unit's safe shutdown equipment.
The liquid waste management system is interconnected at two non-seismic, non-safety locations by normally closed valves. One interconnection allows either unit to transfer liquid wastes to the other unit's holdup tanks. The other interconnection allows the transfer of liquid waste from the aerated waste storage tank of one unit to the other.
The instrument air system is interconnected but normally isolated between units via automatically controlled valves. As instrument air pressure is lost in one unit the isolation valves automatically open to allow compressed air be provided by the other unit.
The station service air system is interconnected between units, but is isolated via normally closed valves.
1.2-14 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 The startup transformers (1A-2A, 1B-2B) are provided with a manual switching arrangement which permits paralleling 4.16kV power to St. Lucie Units 1 and 2 (see Section 8.2.1.5 for additional discussion).
St. Lucie Units 1 and 2 are designed using the "slide along" concept. The following facilities, systems and components are shared (one system which may be used by either or both units) by both nuclear units:
a) ultimate heat sink, b) steam generator blowdown treatment facility, c) makeup demineralizer regeneration (water treatment facility),
d) domestic water and fire protection system, e) switchyard, telemetering and load dispatch equipment, f) seismic instrumentation, g) site and offsite environmental monitors, h) hypochlorite system, i) turbine oil storage tank, j) carbon dioxide, nitrogen and hydrogen systems, k) auxiliary steam supply system, l) safety assessment system, and m) condensate polisher filter demineralizer system.
All facilities are constructed so that no failure can in any way preclude safe shutdown of the plant.
An accident or single failure in one unit does not affect safe shutdown of either unit. A failure in any of the share features may result in reduced load operation of either or both units, but the capability for safe shutdown is unaffected by such a failure.
The ISFSI (Section 1.2.2.9) is also shared by both units for dry storage of spent fuel.
The LLWSF (Section 1.2.2.10) is also shared by both units for the interim storage of low level waste prior to shipment off site.
1.2.5 SECURITY PLAN As discussed in Section 13.7 of the Unit 1 UFSAR, a common site security plan is provided for St. Lucie Units 1 and 2.
1.2-15 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 1.2.6 EMERGENCY PLAN As discussed in Section 13.3, a common site emergency plan is provided for St. Lucie Units 1 & 2.
1.2.7 SYMBOLS AND ABBREVIATIONS ON FIGURES Definitions of symbols and abbreviations used throughout the chapters on fluid and electrical systems are shown in detail on Figures 1.2-23 and 1.2-24. The auxiliary pumps P&I diagram is shown on Figure 1.2-34.
1.2-16 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 1.
2.8 REFERENCES
FOR SECTION 1.2
- 1. Letter from M. Rahimi (NRC) to T. Neider (Transnuclear, Inc.), Certificate of Compliance No. 1030 for the NUHOMS HD System dated January 10, 2007, including Safety Evaluation Report to Transnuclear, Inc. NUHOMS HD Horizontal Modular Storage System for Irradiated Nuclear Fuel
- 2. Appendix A to Certificate of Compliance No. 1030: NUHOMS HD System Generic Technical Specifications
- 3. Transnuclear NUHOMS HD Horizontal Modular Storage System for Irradiated Nuclear Fuel Final Safety Analysis Report
- 4. ML16063A121, St. Lucie Plant, Unit No. 2 - Issuance of Amendment Regarding Transitioning to AREVA Fuel (CAC No. MF5495)
- 5. DBD-FP-1, Fire Protection Design Basis Document 1.2-17 Amendment No. 27 (03/22)
Refer to Drawing 2998-G-058 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 SITE PLAN FIGURE 1.2-1 Amendment No. 18 (01/08)
Refer to Drawing 2998-G-059 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 ENLARGED PLOT PLAN FIGURE 1.2-2 Amendment No. 18 (01/08)
Refer to Dwg.
2998-G-060 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT TURBINE BUILDING GROUND FLOOR PLAN FIGURE 1.2-3 Amendment No. 10, (7/96)
Refer to Drawing 2998-G-061 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT TURBINE BUILDING FIGURE 1.2-4 Amendment No. 18 (01/08)
Refer to Dwg.
2998-G-062 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT TURBINE BUILDING OPERATING FLOOR PLAN FIGURE 1.2-5 Amendment No. 10, (7/96)
Refer to Dwg.
2998-G-063 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT TURBINE BUILDING SECTIONS SHEET 1 FIGURE 1.2-6 Amendment No. 10, (7/96)
Refer to Dwg.
2998-G-064 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT TURBINE BUILDING SECTIONS SHEET 2 FIGURE 1.2-7 Amendment No. 10, (7/96)
Refer to Dwg.
2998-G-065 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT REACTOR BLDG FLOOR PLANS SHEET 1 FIGURE 1.2-8 Amendment No. 10, (7/96)
Refer to Dwg.
2998-G-066 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT REACTOR BLDG FLOOR PLANS SHEET 2 AND MAIN STEAM TRESTLE FIGURE 1.2-9 Amendment No. 18 (01/08)
Refer to Dwg.
2998-G-067 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT REACTOR BUILDING SECTIONS SHEET 1 FIGURE 1.2-10 Amendment No. 10, (7/96)
Refer to Dwg.
2998-G-068 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT REACTOR BUILDING SECTIONS SHEET 2 FIGURE 1.2-11 Amendment No. 10, (7/96)
Refer to Dwg.
2998-G-069 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING PLAN SHEET 1 FIGURE 1.2-12 Amendment No. 10, (7/96)
Refer to Dwg.
2998-G-070 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING PLAN SHEET 2 FIGURE 1.2-13 Amendment No. 10, (7/96)
Refer to Dwg.
2998-G-071 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING PLAN SHEET 3 FIGURE 1.2-14 Amendment No. 10, (7/96)
Refer to Dwg.
2998-G-072 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING FIGURE 1.2-15 Amendment No. 18 (01/08)
Refer to Dwg.
2998-G-073 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT FUEL HANDLING BUILDING PLANS FIGURE 1.2-16 Amendment No. 10, (7/96)
Refer to Dwg.
2998-G-074 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT FUEL HANDLING BUILDING SECTIONS FIGURE 1.2-17 Amendment No. 10, (7/96)
Refer to Drawing 2998-G-075 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT REACTOR AUXILIARY FIGURE 1.2-18 Amendment No. 18 (01/08)
Refer to Dwg.
2998-G-076 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING MISCELLANEOUS PLANS AND SECTIONS FIGURE 1.2-19 Amendment No. 10, (7/96)
Refer to Dwg.
2998-G-077 SH 1 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT COMPONENT COOLING WATER AREA AND DIESEL GENERATOR BUILDING FIGURE 1.2-20 Amendment No. 10, (7/96)
Refer to Drawing 2998-G-077 SH 2 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT COMPONENT COOLING AREA FIGURE 1.2-21 Amendment No. 18 (01/08)
Refer to Drawing 2998-G-077 SH 3 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 GENERAL ARRANGEMENT INTAKE STRUCTURE FIGURE 1.2-22 Amendment No. 18 (01/08)
Refer to Drawing 2998-G-078 SH 100 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 FLOW DIAGRAM SYMBOLS FIGURE 1.2-23 Amendment No. 18 (01/08)
Refer to Dwg.
2998-B-276, Sheet 00-2 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 CONTROL AND BLOCK DIAGRAM FIGURE 1.2-24 Amendment No. 14 (12/01)
Figures 1.2-25 through 1.2-33 have been deleted Amendment No. 14 (12/01)
Refer to Dwg.
2998-G-078 SH 105A, B, C Amendment No. 11, (5/97)
FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 FLOW DIAGRAM AUXILIARY PUMPS FIGURE 1.2-34
UFSAR/St. Lucie - 2 1.3 COMPARISONS Comparisons contained herein were valid at the time the operating license for St. Lucie Unit 2 was issued, and are being retained in the Updated FSAR for document completeness and historical record. No present or future update of this section is required.
1.3.1 COMPARISONS WITH SIMILAR FACILITY DESIGNS Table 1.3-l presents a summary of the characteristics of St. Lucie Unit 2 as originally licensed.
The table presents comparative data for San Onofre Units 2 and 3; Arkansas Nuclear One, Unit 2; and St. Lucie Unit 1. Data was extracted from the applicable FSAR.
The San Onofre Units 2 and 3, and Arkansas Nuclear One, Unit 2 designs were selected for comparison because of the basic similarity of the reactor cores. Also they are well advanced in terms of licensing relative to St. Lucie Unit 2. St. Lucie Unit 1 was selected because it is an operating plant which is essentially the same design as St. Lucie Unit 2.
1.3.2 COMPARISON OF FINAL AND PRELIMINARY INFORMATION 1.3.2.1 General This section contains a discussion of the significant changes that have been made in the St.
Lucie Unit 2 design since submittal of the FSAR. Changes considered as significant would include changes in design bases or criteria for seismic Category I structures, and safety related systems or components, plant arrangement, mode of system operation, type of equipment, or gross changes in component or system capacity. In general, such changes further increase the safety margins and operating flexibility of St. Lucie Unit 2.
1.3.2.2 Fuel Load and Operation Dates Fuel loading was scheduled to commence in October 1982 and 100 percent power operation was expected to be reached in April 1983. The operating license was actually issued in April 1983 and 100 percent power operation was achieved in July 1983.
1.3.2.3 Deletion of Chlorine Accident Detection System A hypochlorite/biocide system has replaced the onsite use of bottled chlorine storage to control EC293500 biological fouling in the Circulating Water System (refer to Subsection 10.4.5.4). As a result, the chlorine accident detection system is not required and has been eliminated.
1.3.2.4 New and Spent Fuel Storage Racks The capacities of both the new fuel and spent fuel storage racks have increased as discussed in Subsections 9.1.1 and 9.1.2, respectively.
1.3.2.5 Construction Responsibility Florida Power & Light Company (FP&L) has assumed responsibility for construction of St. Lucie Unit 2, with Ebasco Services Incorporated providing supervision and craft labor for performance of construction as directed or required by FP&L (refer to Section 1.4).
1.3-1 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 1.3.2.6 Pipe Rupture Criteria Rupture restraint locations are selected on a "break anywhere" criteria based on Giambusso criteria which was accepted by the NRC review as delineated in the SER (November 1974).
Rupture restraints are not provided where it was shown that the broken pipe does not cause unacceptable damage to essential systems. Rupture restraints are also not provided for system pressures under 275 psig, for slot breaks in lines less than four inches, and for systems only operating during accident and/or testing conditions.
In addition, a moderate energy piping analysis has been performed based on criteria as presented in Section 3.6.
The Shutdown Cooling System, which is used as high energy fluid system for only short operational periods and as moderate energy fluid system for the major operational periods, is classified and analyzed as a moderate energy system.
1.3.2.7 Clarification of Code Commitments ACI-349 was not utilized as design criteria for St Lucie Unit 2 structures. For a clarification of the extent of use of ASME Code,Section III NF, refer to Subsections 3.8.3.2.1 and 3.9.3.4.
1.3.2.8 Containment Analysis As discussed in Subsection 6.2.1.1, the computer code utilized to determine the containment pressure/temperature results from a loss-of-coolant-accident (LOCA) or main steam line break (MSLB) was CONTRANS (rather than CONTEMPT). In addition, the main feedwater and back-up isolation valves have changed to a 4.0 second closure time.
A spectrum of small break LOCAs are also analyzed.
1.3.2.9 NOT USED EC291393 1.3.2.10 Control Room Design and Analysis The control room can support a 30 day occupancy throughout the duration of the accident without exceeding the guidelines of GDC 19. The control room is automatically isolated at the outset of the accident followed by the manual opening of an outside air intake, with filtration of the air through charcoal and HEPA filters.
The maximum temperature reached in the control room is based on having only one chiller of the Control Room Air Conditioning System available. Refer to Subsection 9.4.1 for further discussion.
1.3.2.11 Atmospheric Dump Valves and Main Feedwater Isolation Valves In lieu of two 100 percent ac controlled atmospheric dump valves, four 50 percent capacity valves are provided, two on each main steam line, with ac controlled modulation and dc control for open/close operation (refer to Subsection 10.4.9).
1.3-2 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 The backup feedwater isolation valves have been relocated immediately upstream of the main feedwater isolation valves in place of the feedwater check valves (refer to Subsection 10.4.7) and are now classified as Quality Group B, seismic Category I.
1.3.2.12 Continuous Containment Purge/Hydrogen Purge System A Continuous Containment Purge/Hydrogen System has been added, as described in Subsection 9.4.8. As a result, the Airborne Radioactivity Removal System and Containment Instrument Air Compressor inside the containment are no longer required and they have been eliminated.
1.3.2.13 Solid Waste Management System As stated in Section 11.4, when solidification is performed, in lieu of a permanent system a portable solidification system provided by an outside contractor is utilized to prepare waste material for transportation to an offsite disposal facility.
1.3.2.14 Radiation Protection The Radiation Monitoring System is a computer based digital system as described in Section 11.5 and Subsection 12.3.4.
In light of the ALARA concern, plant shielding has been improved where practicable, some of which was based on St Lucie Unit 1 experience. Some examples of improved shielding design are the shielding provided for the fuel transfer tube, and shielding for neutron streaming around the reactor vessel (refer to Subsection 12.3.1). Other changes such as a bottom-loaded filter system are provided to reduce doses to operating personnel.
1.3.2.15 Protection Logic As described in Sections 7.2 and 7.3, the Reactor Protective System and engineered safety features system logic is designed to initiate protective action whenever the signal of any two of three channels reaches the preset limit. A fourth channel is provided as a spare and allows bypassing of one channel while maintaining a two-out-of-three system.
1.3.2.16 Meteorological Data Acquisition New calculational techniques for updating the site meteorological data are used as detailed in Section 2.3.
1.3.2.17 Fire Safety Analysis Design features which conform to 10 CFR 50 Appendix A, GDC 3 and 10 CFR 50.48(c),
NFPA 805 are presented in the Fire Protection Design Basis Document. (Reference 1) 1.3.2.18 Auxiliary Feedwater System The motor-operated valves required for the operation of the turbine-driven auxiliary feedwater pump are dc controlled (refer to Subsection 10.4.9).
1.3-3 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 1.3.2.19 Chapter 15 Accident Analysis The chapter is structured around an event type/frequency matrix which categorizes the initiating events by type and expected frequency of occurrence. Only the limiting cases in each group have been quantitatively analyzed.
Incorporated into Chapter 15 is the Reload Safety Evaluation and Chapter 15 appendices.
1.3-4 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 1.
3.3 REFERENCES
FOR SECTION 1.3
- 1. DBD-FP-1, Fire Protection Design Basis Document.
1.3-5 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.3-1 PLANT PARAMETER COMPARISON St. Lucie Reference San Onofre St. Lucie Item Unit 2 (Cycle 1) Section Units 2 and 3 ANO-2 Unit 1 (Cycle 1)
Hydraulic and Thermal Design Parameters Rated core heat output, MWt 2,560 4.4 3,390 2,815 2,560 Rated core heat output, Btu/hr 8,737 x 106 4.4 11,570 x 106 9,608 x 106 8,737 x 106 Heat generated in fuel, % 97.5 4.4 97.5 97.4 97.5 System pressure, nominal, psia 2,250 4.4 2,250 2,250 2,250 System pressure, minimum steady state,psia 2,200 4.4 2,200 2,200 2,200 Hot channel factors, Heat flux, Fq 2.57 2.35 2.35 2.85 DNB ratio at nominal conditions 2.64 (CE-1) 4.4 2.07 (CE-1) 2.26 (W-3) 2.30 (W-3)
Coolant flow Minimum allowable reactor flowrate, lb/hr 139.4 x 106 4.4 148 x 106 120.4 x 106 122 x 106 Effective flowrate for heat transfer, lb/hr 134.3 x 106 4.4 142.8 x 106 116.2 x 106 117.5 x 106 Effective flow area for heat transfer, ft2 54.7 4.4 54.7 44.6 53.5 Average velocity along fuel rods, ft/sec 15.1 4.4 16.3 16.4 13.6 Average mass velocity, lb/hr-ft2 2.45 x 106 4.4 2.61 x 106 2.60 x 106 2.20 x 106 Coolant temperatures, F Nominal inlet 548 4.4 553 553.5 538.9 Design inlet 550 4.4 556 556.5 544 Average rise in vessel 48 4.4 58 58.5 55 Average rise in core 50 4.4 60 60.5 56 Average in core 573 4.4 586 583.75 572 Average in vessel 572 4.4 582 582.75 571.5 Nominal outlet of hot channel 622 4.4 642 652 640 T1.3-1 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 TABLE 1.3-1 (Cont'd)
St. Lucie Reference San Onofre St. Lucie Item Unit 2 (Cycle 1) Section Units 2 and 3 ANO-2 Unit 1 (Cycle 1)
Hydraulic and Thermal Design Parameters (Cont'd)
Heat transfer at 100% power Active heat transfer surface area, ft2 56,315 4.4 62,000 51,000 48,400 Average heat flux, Btu/hr-ft2 151,300 4.4 182,400 185,000 176,000 Maximum heat flux, Btu/hr-ft2 388,800 4.4 428,000 433,800 501,300 Average thermal output, KW/ft (Fuel Rod Only) 4.43 4.4 5.34 5.41 5.94 Maximum thermal output, KW/ft (Fuel Rod Only) 11.4 4.4 12.5 12.7 17 Maximum clad surface temperature at nominal 657.0 4.4 657.0 657 657 pressure, F Fuel center temperature, F 2,986 4.4 3,180 3,420 3,890 maximum at 100% power Core Mechanical Design Parameters Fuel assemblies Design CEA 4.2 CEA CEA CEA Rod pitch, in. 0.506 4.2 0.5063 0.5063 0.58 Cross-section dimensions, in. 7.972 x 7.972 4.2 7.972 x 7.972 7.97 x 7.97 7.98 x 7.98 Fuel weight (as UO2), lbm 204.4 x 103 4.2 223.9 x 103 183,834 207,200 Total weight, lbm 282.8 x 103 4.2 314,867 250,208 271,280 Number of grids per assembly 10 4.2 11 12 8 Fuel rods Number 49,580 4.2 49,580 40,644 36,896 Outside diameter, in. 0.382 4.2 0.382 0.382 0.44 Diametral gap, in. 0.007 4.2 0.007 0.007 0.0085 Clad thickness, in. 0.025 4.2 0.025 0.025 0.026 Clad material Zircaloy-4 4.2 Zircaloy-4 Zircaloy Zircaloy T1.3-2 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 TABLE 1.3-1 (Cont'd)
St. Lucie Reference San Onofre St. Lucie Item Unit 2 (Cycle 1) Section Units 2 and 3 ANO-2 Unit 1 (Cycle 1)
Core Mechanical Design Parameters (Cont'd)
Fuel pellets Material UO2 sintered 4.2 UO2 sintered UO2 sintered UO2 sintered Diameter, in. 0.325 4.2 0.325 0.325 0.3795 Length, in. 0.390 4.2 0.390 0.390 0.650 Control assemblies Neutron absorber (See Table 4.2-1) 4.2 (See Table 4.2-1) B4C/Ag-In-Cd B4C/SS Cladding material Inconel 625 4.2 Inconel 625 NiCrFe alloy NiCrFe alloy Clad thickness 0.035 4.2 0.035 0.035 0.040 Number of assembly, full/part-length 83/0 4.2 83/8 73/8 73/8 Number of rods per assembly 4,5/5 4.2 4,5/5 5 5 Nuclear Design Data Structural characteristics Core diameter, in. (equivalent) 136 4.2 136 123 136 Core height, in. (active fuel) 136.7 4.2 150 150 136.7 H20/UO2 Unit Cell (cold), volume ratio 1.705 4.2 1.705 1.705 1.63 Number of fuel assemblies 217 4.2 217 177 217 UO2 Rods per assembly, unshimmed/shimmed Batch A 236 4.3 236 236 176 Batch B 236/220 4.3 236/220 224 164 Batch C 236/224 or 220 4.3 236/224 or 220 224/234/233 176/164/164 Performance characteristics loading technique 3-batch mixed 4.3 3-batch mixed 3-batch mixed 3-batch mixed central zone central zone central zone central zone Fuel discharge burnup, MWD/MTU Average first cycle 13,187 4.3 12,731 12,500 12,800 T1.3-3 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 TABLE 1.3-1 (Cont'd)
St. Lucie Reference San Onofre St. Lucie Item Unit 2 (Cycle 1) Section Units 2 and 3 ANO-2 Unit 1 (Cycle 1)
Nuclear Design Data (Cont'd)
Feed enrichment, wt%
Region 1 1.71 4.3 1.87 1.93 1.93 Region 2 2.28 4.3 2.38 2.27 2.33 Region 3 2.73 4.3 2.88 2.94 2.82 Control characteristics effective multiplication (beginning of life)
Cold, no power, clean 1.170 4.3 1.170 1.195 1.170 Hot, no power, clean 1.119 4.3 1.125 1.139 1.134 Hot, full power, Xe equilibrium 1.070 4.3 1.067 1.082 1.078 Control Assemblies Total rod worth (hot), % 11.16 (EOC) 4.3 11.35 12.3 11.0 Boron concentrations for criticality:
Zero power no rods inserted, clean, ppm 901/809 4.3 899/832 1011/1001 945/935 Cold/Hot At power with no rods inserted, 715/493 4.3 719/452 881/611 820/590 clean/equilibrium xenon, ppm Kinetic characteristics, range over life Moderator temperature coefficient, /F See Table 4.3-4 4.3 See Table 4.3-4 -0.3 x 10-4 -0.4 x 10-4 to to
-2.5 x 10-4 -2.1 x 10-4 Moderator pressure Coefficient, /psi +0.6 x 10-6 4.3 +0.7 x 10-6 +0.06 x 10-6 +0.49 x 10-6 to to
+2.6 x 10-6 +2.55 x 10-6 Moderator void coefficient, /% Void -0.22 x 10-3 4.3 -0.36 x 10-3 -0.03 x 10-3 -0.26 x 10-3 to to
-1.22 x 10-3 -1.35 x 10-3 Doppler coefficient, /F See Figure 4.3-34 4.3 1.18 x 10 -1.18 x 10-5 -1.45 x 10-5 to to to 1.28 x 10-5 -1.78 x 10-5 -1.07 x 10-5 T1.3-4 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 TABLE 1.3-1 (Cont'd)
St. Lucie Reference San Onofre St. Lucie Item Unit 2 (Cycle 1) Section Units 2 and 3 ANO-2 Unit 1 (Cycle 1)
Principal Design Parameters of the Reactor Coolant System Operating pressure, psig 2,235 5.1 2,235 2,235 2,235 Operating Reactor inlet temperature, F 550 5.1 553 553.5 539.7 Operating Reactor outlet temperature, F 604 5.1 611.2 612.5 595.7 Number of loops 2 5.1 2 2 2 Design pressure, psig 2,485 5.1 2,485 2,485 2,485 Design Temperature, F 650 5.1 650 650 650 Hydrostatic test pressure (cold), psig 3,110 5.1 3,110 3,110 3,110 Principal Design Parameters of the Reactor Vessel Material See Table 5.2-3 5.2 See Table 5.2-2 SA-533, Grade B, SA-533,Grade B, Class I, low Class 1, low alloy steel, alloy steel internally clad internally clad with Type 304 with Type 304 austenitic SS austenitic SS Design pressure, psig 2,485 5.3 2,485 2,485 2,485 Design temperature, F 650 5.3 650 650 650 Operating pressure, psig 2,235 5.3 2,235 2,235 2,235 Inside diameter of shell, in. 172 5.3 172 157 172 Outside diameter across nozzles, in. 253 5.3 253 238 253 Overall height of vessel and enclosure head, 41-10-3/8 5.3 43-6-1/2 43-4-1/6 41-11-3/4 ft-in. to top of CEDM nozzle Minimum clad thickness, in. 1/8 5.3 1/8 1/8 5/16 Principal Design Parameters of the Steam Generators Number of Units 2 5.4 2 2 2 T1.3-5 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 TABLE 1.3-1 (Cont'd)
St. Lucie Reference San Onofre St. Lucie Item Unit 2 (Cycle 1) Section Units 2 and 3 ANO-2 Unit 1 (Cycle 1)
Principal Design Parameters of the Steam Generators (Cont'd)
Type Vertical U-tube 5.4 Vertical U-tube Vertical U-tube Vertical U-tube with integral with integral with integral with integral moisture separator moisture separator moisture separator moisture separator Tube material NiCrFe alloy 5.2 NiCrFe alloy NiCrFe alloy NiCrFe alloy Shell material SA-533 GR A&B, 5.2 SA-533 Gr. B SA-533 Gr. B SA-533 Gr. B Class 1 and Class 1 and Class 1 and Class 1 and SA 516, Gr. 70 SA-516, Gr.70 SA-516, Gr. 70 SA-516, Gr. 70 Tube side design Pressure, psig 2,485 5.4 2,485 2,485 2,485 Tube side design temperature, F 650 5.4 650 650 650 Tube side design flow, lb/hr 61 x 106 5.4 74 x 106 60.2 x 106 61 x 106 Shell side design pressure, psia 1,000 5.4 1,100 1,100 1,000 Shell side design temperature, F 550 5.4 560 560 550 Operating pressure, tube side, nominal, psig 2,235 5.4 2,235 2,235 2,235 Operating Pressure, shell side, maximum, psig 885 985 985 885 Maximum moisture at outlet at full load, % 0.2 5.4 0.2 0.2 0.2 Hydrostatic test pressure, tube side (cold) psia 3,110 3,110 3,110 3,110 Steam Pressure at full power, psia 815 5.4 900 900 815 Steam temperature, at full power, F 520.3 5.4 532 531.95 520.3 Principal Design Parameters of the Reactor Coolant Pumps Number of units 4 5.4 4 4 4 Type Vertical, single Vertical, single Vertical, single Vertical, single stage centrifugal stage radial flow stage centrifugal stage centrifugal with botton with bottom with bottom with bottom suction and suction and suction and suction and horizontal horizontal horizontal horizontal discharge discharge discharge discharge Design pressure, psig 2,485 5.4 2,485 2,485 2,485 T1.3-6 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 TABLE 1.3-1 (Cont'd)
St. Lucie Reference San Onofre St. Lucie Item Unit 2 (Cycle 1) Section Units 2 and 3 ANO-2 Unit 1 (Cycle 1)
Principal Design Parameters of the Reactor Coolant Pumps (Cont'd)
Design temperature, F 650 5.4 650 650 650 Operating pressure, nominal psig 2,235 5.4 2,235 2,235 2,235 Suction temperature, F 550 5.4 553 553.5 540 Design capacity, gal/min 81,200 5.4 99,000 80,000 80,000 Design head, ft 310 5.4 310 275 250 Hydrostatic test pressure (cold), psig 3,110 3,110 3,110 3,110 Motor type AC induction, AC induction, AC induction, AC induction, single speed single speed single speed single speed Motor rating, hp 6,500 9,700 6,500 6,500 Principal Design Parameters of the Reactor Coolant Piping Material See Table 5.2-3 SA-516, Gr 70 SA-516, Gr 70 SA-516, Gr 70 with nominal with nominal with nominal 7/32 SS clad 3/16 SS clad 7/32 SS clad Hot leg ID, in. 42 5.4 42 42 42 Cold leg ID, in. 30 5.4 30 30 30 Between pump and steam generator ID, in. 30 5.4 30 30 30 Engineered Safety Features High pressure safety injection pumps 2 6.3 3 3 3 Low pressure safety injection pumps 2 6.3 2 2 2 Safety injection tanks, number 4 6.3 4 4 4 Containment spray pumps 2 6.2 2 2 2 Containment fan coolers units 4 6.2 4 4 4 Air flow capacity, each at 39,600 6.2 31,000 50,000 55,800 emergency conditions, ft3 /min T1.3-7 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 TABLE 1.3-1 (Cont'd)
St. Lucie Reference San Onofre St. Lucie Item Unit 2 (Cycle 1) Section Units 2 and 3 ANO-2 Unit 1 (Cycle 1)
Engineered Safety Features (Cont'd)
Emergency power Diesel-generator unit 2 8.3 4 (for two units) 2 2 Containment System Parameters Type Steel containment 3.8.2 Steel-lined Steel-lined Steel containment vessel with prestressed post prestressed post vessel with cylin-cylindrical shell, tensioned con- tensioned con- drical shell, hem-hemispherical dome crete cylinder, crete cylinder, ispherical dome and ellipsoidal bot- curve dome roof. curved dome roof. and ellipsoidal tom - ASME Code, bottom - ASME Section III, Class MC, Code,Section III, surrounded by rein- Class B, surround-forced concrete Shield ed by reinforced Building. concrete Shield Building.
Inside Diameter, ft. 140 3.8 150 116 140 Height, ft. 232 3.8 172 207 232 Free volume, ft3 2,500,000 6.2 2,335,000 1,780,000 2,500,000 Reference accident Pressure, psig 44 3.8 60 54 44 Steel Thickness, in.
Vertical Wall 1.92 3.8 Not Applicable Not Applicable 1.91 Hemispherical Head 0.96 Not Applicable Not Applicable 0.95 Knuckles 2.125 Not Applicable Not Applicable 225 Concrete Thickness, ft.
Vertical Wall Not Applicable 3.8 4 1/3 3 3/4 Not Applicable Dome Not Applicable 3 3/4 3 1/4 Not Applicable Design Parameters - Shield Building 3.8 Not Applicable Not Applicable Inside Diameter, ft. 148 148 Height, ft. (top of foundation to top of dome) 230.5 230.5 Concrete Thickness, ft.
Vertical Wall 3 3 Dome 2.5 2.5 T1.3-8 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 TABLE 1.3-1 (Cont'd)
St. Lucie Reference San Onofre St. Lucie Item Unit 2 (Cycle 1) Section Units 2 and 3 ANO-2 Unit 1 (Cycle 1)
Containment Leak Prevention and Mitigation Systems Leak-tight pene- 6.2 Leak-tight pene- Leak-tight pene- Leak-tight pene-tration, Automatic tration, and tration, and tration, Automatic isolation where continuous steel continuous steel isolation where required. liner. Automatic liner. Automatic required.
isolation where isolation where required. required.
Gaseous Effluent Purge Discharge through 6.2 Discharge through Discharge through Discharge through vent. vent. vent. vent.
RADIOACTIVE WASTE MANAGEMENT SYSTEM Liquid Waste Processing Systems Reactor Coolant Waste Holdup Tank 11.2 (EMS)
Number 4 1/2 4 4 Capacity (Gal.),each 40,000 6,000/25,000 51,270 40,000 Concentrators Number 1 1 (For 2 units) 1 1 Capacity (gpm) 20 50 20 2 Gaseous Waste Processing Systems Waste Gas Decay Tank 11.3 Number 3 6 (For 2 units) 3 3 Capacity (ft3), each 138 500 300 144 Pressure (psig) 190 150 380 190 Hold-up time (days) 25 30 30 30 ELECTRIC SYSTEMS Number of Offsite Circuits 3 8.1 8 3 3 Number of Incoming Lines to Startup Transformers 2 8.2 2 2 2 Number of Startup Transformers 2 8.2 4 1+1(shared) 2 Number of Main Unit Transformers (Three Phase) 2 8.2 1 3 (single phase) 2 Number of 4.16 KV Engineered Safety Features System Buses 3 8.3 3 2 3 T1.3-9 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 TABLE 1.3-1 (Cont'd)
St. Lucie Reference San Onofre St. Lucie Item Unit 2 (Cycle 1) Section Units 2 and 3 ANO-2 Unit 1 (Cycle 1)
ELECTRIC SYSTEMS (Cont'd)
Number of 480V Engineered Safety Features System Buses 3 8.3 3 2 3 Number of 120V Safety Related Vital Buses 4 8.3 4 4 4 Number of Standby Diesel Generators 2 8.3 2 2 2 Diesel Generator Rating (KW) 3685 8.3 4700 2850 3500 INSTRUMENTATION SYSTEMS*
Reactor Protective System 7.2 7.2 7.2 7.2 7.2 Reactor and Reactor Coolant System 7.7.1.1 7.7.1.1 7.7.1.1 7.7.1.1 7.7.1.1 7.6.1 7.1.1.2 7.7.1.2 7.7.1.2 7.7.1.2 Steam and Feedwater Control System 7.7.1.1 7.7.1.3 7.7.1.3 7.7.1.3 7.7.1.3 Nuclear Instrumentation 7.2.1.1 7.2.1.1 7.2.1.1 7.2.1.1 7.2.1.1 7.7.1.1 Non-Nuclear Process Instrumentation 7.7.1.1 7.5.1.5 7.5.1.5 7.5.1.5 7.5.1.5 7.5.1 CEA Position Instrumentation 7.7.1.1 7.5.1.3 7.5.1.3 7.5.1.3 7.5.1.3
- This section is not suited for tabular description. SAR section numbers have been included for the location of the detailed description of each system.
T1.3-10 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS Information contained herein was valid at the time the operating license for St. Lucie Unit 2 was issued, and is being retained in the updated FSAR for document completeness and historical record.
No present or future update of this section is required.
The Florida Power & Light Company is the applicant for the operating license for St Lucie Unit 2.
Florida Power & Light Company is responsible for the design and engineering review, construction and operation of the plant.
Florida Power & Light Company has engaged Combustion Engineering, Inc. (CE) to design, manufacture and provide the Nuclear Steam Supply System and nuclear fuel for the first core and the first three core reload batches. The Nuclear Steam Supply System includes the Reactor Coolant System, reactor auxiliary system components, nuclear and certain process instrumentation, and the reactor control and protective system. In addition, CE will furnish technical assistance for erection, initial fuel loading, testing and initial startup of the Nuclear Steam Supply System.
Ebasco Services Inc. has been engaged by the Applicant for engineering and procurement services for this project and as such has performed engineering and design work for the balance - of -plant equipment, systems and structures not included under the CE scope of supply. Ebasco has also provided supervision and craft labor for performance of construction as directed or required by Florida Power & Light Company.
These and other engineering firms with approved Quality Assurance Programs may perform backfit, retrofit, maintenance and construction activities during plant operation under the auspices of Florida Power & Light Company.
1.4-1 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION Material contained herein were valid at the time the operating license for St Lucie 2 was issued, and are being retained in the Updated FSAR for document completeness and historical record.
No present or future update of this section is required.
This section provides a description of safety related technical information relevant to this application. Combustion Engineering, Inc., (CE), is conducting research and development programs relating to the requirements of this section.
The St Lucie Unit 2 reactor incorporates a 16 x 16 fuel assembly design with five guide tubes.
This design provides an increase in conservatism for loss -of -coolant accident (LOCA) considerations with a minimum change from previous CE fuel designs. Previous designs have undergone extensive testing, and operating experience is now being acquired.
The three test programs described in Subsections 1.5.1, 1.5.2, and 1.5.3 are considered necessary to confirm the adequacy of the 16 x 16 fuel assembly design.
References 1 to 6 present descriptions of development programs aimed at verifying the Nuclear Steam Supply System (NSSS) design and the anticipated performance characteristics, and at confirming the design margins. Other programs that apply to this plant are identified in Subsections 1.5.4 through 1.5.8.
1.5.1 FRETTING AND VIBRATIONS TESTS OF FUEL ASSEMBLIES Extensive autoclave vibration and dynamic flow tests have been performed to characterize fuel rod and spacer grid fretting corrosion in CE fuel assemblies.
Tests have been completed using a full sized 16 x 16 fuel assembly. This assembly is similar to the 16 x 16 five guide tube design used on the St Lucie Unit 2 reactor. This assembly was subjected to flow testing under conditions of temperature, water chemistry, pressure, and flow velocities in excess of normal reactor conditions. Further information is provided in Subsections 4.2.3.1.1 and 4.2.3.1.2. EC296160 1.5.2 DEPARTURE FROM NUCLEATE BOILING (DNB) TESTING Extensive heat transfer testing has been completed with electrically heated rod bundles representative of the CE 16 x 16 and 14 x 14 fuel assemblies. The program for each assembly geometry included tests to determine the effects on DNB of the control element assembly (CEA) guide tube, bundle heated length, and grid spacing, and lateral and axial power distributions.
Each test yielded DNB data over a wide range of conditions of interest for pressurized water reactor (PWR) design. Those data were used with the TORC subchannel analysis code to develop and to verify the CE -1 DNB correlation for predicting DNB in fuel assemblies with standard spacer grids. The CE -1 correlation, which is discussed in more detail in Subsection 4.4.4.1, is used in computing margin to DNB for St. Lucie Unit 2.
For HTP' fuel, DNBR analyses are performed using the XCOBRA-IIIC code (Reference 49) and the HTP CHF correlation (Reference 53 in Section 4.4). Details of the correlation development and testing campaigns are provided in Section 4.4 Reference 53. The Biasi correlation is used for Post-SCRAM MSLB analyses.
1.5-1 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 1.5.3 FUEL ASSEMBLY STRUCTURAL TESTS The fuel assembly structural testing program was designed to verify the structural adequacy of the fuel assembly design under normal handling, normal operation, seismic excitation, and LOCA loadings. The test program provides the structural characteristics employed in the fuel assembly structural analyses.
A series of tests were conducted on a 14 x 14 fuel assembly to determine the combined axial and lateral load deflection characteristics of the fuel assembly. Axial compression tests and axial drop tests were performed. Measurements were made of axial loads, axial deflections, lateral deflections of all spacer grids, and strains in the guide tubes and fuel rods.
A series of structural tests on the 16 x 16 fuel assembly design was also conducted. The fuel assembly was subjected to both static and dynamic tests so as to determine basic structural characteristics. In addition, several 16 x 16 spacer grids were subjected to impact tests to determine dynamic load deflection characteristics and damage limits. These tests are also discussed in Subsection 4.2.3.1.3.
1.5.4 FUEL ASSEMBLY FLOW MIXING TESTS The objective of the fuel assembly flow mixing program was to obtain information on the magnitude of coolant mixing in CE fuel assemblies. Several series of tests have been completed, and the data from these tests provide a sound basis for the treatment of coolant mixing in design thermal margin calculations.
The first series of single phase flow mixing tests was run in 1966 with a prototype CE PWR fuel assembly. The average level of coolant mixing was determined using dye injection and sampling equipment.
A second series of single phase mixing tests was conducted in 1968 with a model representing a portion of a 14 x 14 CEA type fuel assembly. Those tests, which also used dye injection and sampling techniques, are described in Reference 1.
More recently, tests were conducted in which coolant temperatures were measured in the subchannels of electrically heated rod bundles representative of the 14 x 14 or 16 x 16 fuel assemblies with standard spacer grids.
As discussed in Subsection 4.4.4.1, those data provide confirmation that the results from the previous dye sampling experiments are applicable for the fuel assembly design used in St Lucie Unit 2.
1.5.5 REACTOR FLOW MODEL TESTING AND EVALUATION The objective of the reactor flow model test programs is to obtain information on:
- a. Flow and pressure distributions in various regions of the reactor
- b. Pressure loss coefficients
- c. Hydraulic loads on certain vessel internal components This information is used for establishing or verifying design hydraulic parameters.
1.5-2 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 Flow model testing, which began in 1966, was designed to obtain those reactor hydraulic design data not amenable to direct calculation. Scale model testing possesses the advantages, relative to actual reactor tests, of:
- a. Providing the information early in the design stage
- b. Being more suitable for extensive instrumentation
- c. Being flexible so that proposed design modifications can be investigated The reactor flow models used by CE are generally 1/5 true scale models. In the first four CE flow model programs, a closed core design was used. The closed core simulates the reactor fuel assemblies with individual closed wall tubes containing orifices to provide the correct axial hydraulic resistance.
Further discussion of the CE flow model test programs is provided in Subsection 4.4.4.2.1.
1.5.6 FUEL ASSEMBLY FLOW TESTS The objectives of the fuel assembly flow test program included assessment of the effect of postulated flow maldistributions on thermal behavior and margin.
The program originated in 1967 with fuel assembly flow distribution testing. Both flow visualization and flow pattern measurements were generated on an overscale model of the lower portion of an early CE design fuel assembly.
A second test series was conducted for the CEA type fuel assembly. The second test series was designed to:
- a. Determine the effect of flow obstructions on flow distribution within the fuel assembly
- b. Determine the magnitude of the effect of the disturbed flow patterns on the thermal margin within a CEA type fuel assembly The information from these tests, described further in Reference 1, has established the effect of flow obstructions within the fuel assembly. Additional information on the effects of postulated fuel coolant channel flow blockages is presented in Subsection 4.2.3.2.14.
1.5.7 CONTROL ELEMENT DRIVE MECHANISM (CEDM) TESTS Performance testing of the magnetic jack CEDM is described in Subsection 3.9.4.4 and in EC296160 Reference 1. The program has confirmed the operability of the drive assembly in normal and misaligned conditions as well as the load carrying capability and life characteristics.
1.5.8 DNB IMPROVEMENT The DNB improvement program was initiated by CE in order to obtain empirical information on the departure from nucleate boiling (DNB) phenomenon and on other thermal and hydraulic characteristics of CE fuel assemblies. Testing has been performed with electrically heated rod bundles that correspond dimensionally to fuel rod configurations under in -reactor temperature pressure and flow conditions to obtain data on DNB, pressure drop, and coolant channel exit 1.5-3 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 temperatures. These data were employed to verify that the CE thermal hydraulic design methods conservatively predict DNB.
The DNB improvement program is described in References 1, 2, 3, and 4. It is a continuing program providing improvements in the accuracy of CE thermal and hydraulic computer programs for predicting local coolant conditions and pressure drops and confirming the applicability of currently used DNB correlations to the CE fuel design. Additional information on the program and results applicable to St Lucie Unit 2 are presented in Subsection 4.4.4.1.
1.5-4 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 SECTION 1.5: REFERENCES
- 1. "Safety Related Research and Development for Combustion Engineering Pressurized Water Reactors, Program Summaries," CENPD -87 (Proprietary), January 1973, and CENPD -87, Rev 01, (Non -Proprietary), March 1973.
- 2. "Safety Related Research and Development for Combustion Engineering Pressurized Water Reactors, Program Summaries," CENPD -143 (Proprietary) and CENPD -143, Rev 01 (Non -Proprietary), May 1974.
- 3. "Safety Related Research and Development for Combustion Engineering Pressurized Water Reactors, 1974 Program Summaries," CENPD -184 -P (Proprietary) and CENPD -184 (Non -Proprietary), May 1975.
- 4. "Safety Related Research and Development for Combustion Engineering Pressurized Water Reactors, 1975 Program Summaries," CENPD -229 -P (Proprietary) and CENPD -229 (Non -Proprietary), June 1976.
- 5. "Safety Related Research and Development for Combustion Engineering Pressurized Water Reactors, 1976 Program Summaries," CENPD -258 (Non -Proprietary), October 1977.
- 6. "Safety Related Research and Development for Combustion Engineering Pressurized Water Reactors, 1977 -1978 Program Summaries," CENPD -262 (Non -Proprietary),
December 1978.
1.5-5 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 1.6 MATERIAL INCORPORATED BY REFERENCE Topical reports incorporated by reference were valid at the time of application to the NRC, and are being retained in the updated FSAR for document completeness and historical record. No present or future update of this section is required.
The following topical reports are incorporated by reference.
Report FSAR Number Author and Title Date to NRC Section CENPD-162 Combustion Engineering, Inc. May 1975 4.4, 15.0 (with "CHF Correlation for C-E Fuel (Approved Version, Suppl. 1) Assemblies with Standard Spacer Sept.1976)
Grids-Part l; Uniform Axial Power Distribution" CENPD-168 Combustion Engineering, Inc. Oct. 1976 3.6 Rev. 1 "Design Basis Pipe Breaks for the (Approved Version, Combustion Engineering Two Loop Aug. 1977)
Reactor Coolant System" CENPD-178P Combustion Engineering, Inc. August 1981 3.9, 4.2 and 178 "Structural Analysis of the 16 x 16 Fuel Rev. 1 Assembly for Combined Seismic and Loss-of-Coolant-Accident Loadings" CENPD-115 Combustion Engineering, Inc. April 1974 3.9 Suppl. 1 "Comparison of Calvert Cliffs, Maine Yankee, and Fort Calhoun Design Parameters and Flow-Induced Structural Response" CENPD-182 Combustion Engineering, Inc. June 1977 3.10, 7.2 Rev. 1 "Seismic Qualification of C-E Instrumentation and Control Equipment" CENPD-183 Combustion Engineering, Inc. August 1975 15.3 "C-E Methods for Loss of Flow Analysis" CENPD-187 Combustion Engineering, Inc. October 1975 and 4.2 (with "Method of Analyzing Creep Collapse May 1975 Suppl. 1) of Oval Cladding" (Approved Version, April 1976)
CENPD-26 Combustion Engineering, Inc. August 1971 3.9 (with Suppl. 1 "Description of Combustion through 3) Engineering Loss of Coolant Calculational Procedures" 1.6-1 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 Report FSAR Number Author and Title Date to NRC Section CENPD-42 Combustion Engineering, Inc. August 1972 3.9 "Dynamic Analysis of Reactor Vessel Internals Under Loss of Coolant Accident Conditions with Application to C-E 800 Mwe Class Reactors" CENPD-67 Combustion Engineering, Inc. September 1973 10.3 Rev. 1, Addenda "Iodine Decontamination Factors November 1974, 1 and 2 During PWR Steam Generation and August 1975 Steam Venting" CENPD-98 Combustion Engineering, Inc. July 1973 4.4, 15.0 "Coast Code Description" (Approved Version, April 1974)
CENPD-107 Combustion Engineering, Inc. August 1974, 15.0 (with Suppl. 1 "CESEC" September 1974, through 5) Septem-1975, January 1976, June 1976 CENPD-105 Combustion Engineering, Inc. Nov. 1973 4.3 "Fast Neutron Attenuation by the ANISN-SHADRAC Analytical Method" CENPD-132 Combustion Engineering, Inc. September 1974, 6.2, 6.3, (with Suppl. 1 "Calculative Methods for the C-E Large March 1975, 15.6 and 2) Break LOCA Evaluation Model" August 1975.
CENPD-133 Combustion Engineering, Inc. September 1974, 6.2, 6.3, (with Suppl. 2) "CEFLASH-4A Fortran IV Digital March 1975 15.6 Computer Program for Reactor Blowdown Analysis" CENPD-134 Combustion Engineering, Inc. September 1974, 6.2, 6.3, (with Suppl. 1) "COMPERC-II A Program for March 1975 15.6 Emergency Refill-Reflood of the Core" CENPD-135 Combustion Engineering, Inc. September 1974, 4.2, 6.3, (with Suppl. 2, 4 "STRIKIN-II A Cylindrical Geometry March 1975, 15.6 and 5) Fuel Rod Heat Transfer Program" September 1976, May 1977 CENPD-136 Combustion Engineering, Inc. August 1974 4.2, "High Temperature Properties of 6.3,15.6 Zircaloy and UO2 for use in LOCA Evaluation Model" 1.6-2 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 Report FSAR Number Author and Title Date to NRC Section CENPD-137 Combustion Engineering, Inc. September 1974 6.3, 15.6 (with Suppl. 1) "Calculative Methods for the C-E Small Break LOCA Evaluation Model" CENPD-139 Combustion Engineering, Inc. September 1974 4.1, 4.2, (with Suppl. 1) "C-E Fuel Evaluation Model" (Approved Version, 4.3, 4.4, April 1975) 6.3, 15.6 CENPD-145 Combustion Engineering, Inc. May 1975, 4.3 "A Method of Analyzing In-Core February 1978 Detector Data in Power Reactors" CENPD-148 Combustion Engineering, Inc. September 1974 4.6, 7.2 "Review of Reactor Shutdown System (PPS Design) for Common Mode Failure Susceptibility" CENPD-153 Combustion Engineering, Inc. December 1974, 4.3 with Amendments "Evaluation of Uncertainty in the August 1977, 1 through 3 Nuclear Form Factor Measured by Self February 1978, Powered Fixed In-Core Detector April, 1979 Systems" CENPD-155 Combustion Engineering, Inc. October 1974 5.3 "C-E Procedure for Design, Fabrication, (Approved Version, Installation and Inspection of August 1975)
Surveillance Specimen Brackets Attached to Reactor Vessel Beltline Region" CENPD-161 Combustion Engineering, Inc. June 1975, 4.1, 4.2, with Amendment "TORC - A Computer Code for May 1976 4.3, 4.4,
- 1. Determining the Thermal Margin of a (Approved Version, 15.0 Reactor Core" September 1978)
CENPD-190 Combustion Engineering, Inc. January 1976 15.4 "C-E Method for Control Element (Approved Version, Assembly Ejection Analysis" August 1976)
CENPD-198 Combustion Engineering, Inc. December 1975 4.2 and Supplements "Zircaloy Growth-In-Reactor January 1978 1 and 2 Dimensional Changes in Zircaloy-4 November 1978 Fuel Assemblies" CENPD-206 Combustion Engineering, Inc. February 1977 4.4 "Comparison of TORC Code Predictions with Experimental Data" 1.6-3 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 Report FSAR Number Author and Title Date to NRC Section CENPD-207 Combustion Engineering, Inc. July 1976 4.4 "Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids, Part 2, Non-Uniform Axial Power Distributions" CENPD-213 and Combustion Engineering, Inc. February 1976, 6.3, 15.6 Suppl.1 "Application of FLECHT Reflood Heat March 1976 Transfer Coefficients to Combustion Engineering 16 x 16 Fuel Bundles" CENPD-225 Combustion Engineering, Inc. October 1976 4,2, 4.4 "Fuel and Poison Rod Bowing CENPD-199 Combustion Engineering, Inc. April 1976 4.3 "C-E Setpoint Methodology: Local Power Density and DNB LCSS and LCO Setpoint Methodology for Analog Protective System."
CENPD-188 Combustion Engineering, Inc. April 1976 4.3 "HERMITE, A Multi-Dimensional (Approved Version, Space-Time Kinetics Code for PWR September 1976 Transients" CENPD-254 Combustion Engineering, Inc. August 1977 6.3 "Post-LOCA Long Term Cooling Evaluation Model" CENPD-252P-A Combustion Engineering, Inc. July 1979 3.9 "Method for Analysis of Blowdown Forces in a Reactor Vessel" CVI-TR-7301 CVI February 1975 6.5.1 Design and Development of High Efficiency Charcoal Adsorbers and its Application in ESF Atmospheric Cleanup Systems AFF-TR-7101 American Air Filter November 1972 6.2.2 "Design and Testing of Fan Cooler Filter Systems for Nuclear Applications" WCAP-7709-L Westinghouse April 1972 6.2.5 "Electric Hydrogen Recombiners for PWR Containments" 1.6-4 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 Report FSAR Number Author and Title Date to NRC Section FPLTQAR 1-76A Florida Power & Light Co. January 1976 17.2 Revision 0 Florida Power & Light Co. June 1976 Revision 1 "Topical Quality Assurance Report" September 1976 Revision 2 January 1977 (Approved by NRC September 1977)
ETR-1002 P Ebasco Services, Inc. November 1975 3.6 "Design Considerations for Protection from Effects of Pipe Rupture - Part I -
Dynamic Analysis" 1.6-5 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 1.7 DRAWINGS Drawings contained herein were valid at the time the operating license for St Lucie 2 was issued, and are being retained in the Updated FSAR for document completeness and historical record. No present or future update of the section is required. Updated drawings are maintained at the St Lucie 2 site.
1.7.1 ELECTRICAL, INSTRUMENTATION, AND CONTROL DRAWINGS Tables 1.7-1 and 1.7-2 are lists of electrical, instrumentation and control safety-related drawings prepared by the Architect/Engineer and NSSS supplier, respectively. There are no drawings considered proprietary.
1.7.2 PIPING AND INSTRUMENTATION DIAGRAMS Tables 1.7-3 and 1.7-4 are lists of safety-related piping and instrumentation diagrams prepared by the Architect/Engineer and NSSS supplier, respectively, 1.7-1 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 TABLE 1.7-1 ARCHITECT/ENGINEER SUPPLIED ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS SAFETY RELATED Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By___ ________________Title________________
B 271 0 3/3/78 E ELECTRICAL GEN INSTALLATION NOTES G 272 5 7/23/82 E MAIN ONE LINE WIRING DIAGRAM G 274 5 7/23/82 E AUXILIARY ONE LINE WIRING DIAGRAM B 325 7 11/14/80 E BILL OF MATERIALS B 328 E CABLE & CONDUIT LIST G 332 7 10/30/82 E 480V MISC 125V DC & VITAL AC ONE LINE WIRING DIAGRAM G 332 2 0 10/30/82 E 480V MISC 125V DC & VITAL AC ONE LINE WIRING DIAGRAM SH 2 B 335 2 12/10/80 E POWER DISTRIBUTION & MOTOR DATA SHEETS B 337 2 1/4/79 E ELECTRICAL PENETRATION SCHEDULE G 340 10 1/8/83 E TURBINE BUILDING GROUND FLOOR CONDUITS, TRAYS & GRDG SH2 C 348 11 1/8/80 E MANHOLE & HANDHOLE DETAILS G 352 4 8/4/82 E ARRANGEMENT-SWITCHGEAR ROOM REACTOR AUX BLDG G 354 4 1/18/83 E CABLE TRAY ARRANGEMENT KEY PLAN G 355 7 1/11/83 E TURBINE AREA-UNDERGROUND CONDUIT& GROUNDING SH1 G 356 7 1/11/83 E TURBINE AREA-UNDERGROUND CONDUIT & GROUNDING SH2 G 358 7 1/11/83 E TURBINE AREA-UNDERGROUND CONDUIT & GROUNDING SH4 T1.7-1 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 ( Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By___ ________________Title________________
G 367 9 3/11/83 E REACTOR CONTAINMENT BLDG-COND TRAYS & GRDG PLAN-EL 62'-0 G 364 8 4/12/83 E REACTOR CONTAINMENT BLDG-COND &
GRDG-PLAN-BELOW EL 18'-0 G 364 1 6 3/11/83 E REACTOR CONTAINMENT BLDG CONDUIT LOCATION PLAN G 365 9 11/22/82 E REACTOR CONTAINMENT BLDG-COND TRAYS & GRDG PLAN EL-18'-0 G 366 8 11/22/82 E REACTOR CONTAINMENT BLDG-COND TRAYS & GRDG PLAN EL-45'-0 G 368 8 10/29/82 REACTOR CONTAINMENT BLDG-COND SECTIONS & DETAILS SH-1 G 369 1 6 10/29/82 E REACTOR CONTAINMENT BLDG-COND SECTIONS & DETAILS SH-2 G 369 2 5 10/6/82 E REACTOR CONTAINMENT BLDG-COND SECTIONS & DETAILS SH-3 G 369 3 5 10/6/82 E REACTOR CONTAINMENT BLDG-COND SECTIONS & DETAILS SH-4 G 372 1A 5 3/23/83 E
SUMMARY
SHEET CABLE TRAY SUPPORT SH-1A G 372 1B 1 8/11/78 E
SUMMARY
SHEET CABLE TRAY SUPPORT SH-1B G 372 2 4 11/18/82 E REACTOR CONT BLDG EL 18.0 CABLE TRAY SUPPORT SH-2 G 372 3 3 8/4/82 E REACTOR CONT BLDG EL 45.0 CABLE TRAY SUPPORT SH-3 G 372 4 5 8/31/81 E RCB PEN AREA EL 23-0 CABLE TRAY SUPPORT SH-4 G 372 5 5 3/23/83 E RCB PEN AREA EL 45-0 CABLE TRAY SUPPORT SH-5 T1.7-2 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By___ ________________Title________________
G 372 6 6 11/18/82 E REACTOR AUX BLDG PEN AREA CABLE TRAY SUPPORT SH-6 G 372 7 5 3/23/83 E REACTOR AUX BLDG EL-05.0 CABLE TRAY SUPPORT SH-7 G 372 8 5 1/21/83 E REACTOR AUX BLDG EL-05.0 CABLE TRAY SUPPORT SH-8 G 372 9 6 3/23/83 E REACTOR AUX BLDG EL 19.5 CABLE TRAY SUPPORT SH-9 G 372 10 4 8/31/82 E REACTOR AUX BLDG EL 19.5 CABLE TRAY SUPPORT SH-10 G 372 11 5 3/23/83 E REACTOR AUX BLDG EL 43'-0 CABLE TRAY SUPPORT SH-11 G 372 12 4 11/18/82 E REACTOR AUX BLDG EL 43'-0 CABLE TRAY SUPPORT SH-12 G 372 13 4 1/21/83 E CABLE VAULT CABLE TRAY SUPPORT SH-13 G 372 14 4 8/31/82 E REACTOR AUX BLDG EL 74.0 CABLE TRAY SUPPORT SH-14 G 372 15 2 8/31/82 E CABLE VAULT-CABLE TRAY SUPPORT SH-15 G 372 16 2 8/31/82 E PENETRATION AREA CABLE TRAY SUPPORT SH-16 G 374 1 6 11/18/82 E REACTOR AUX BLDG PENETRATION AREA-COND-TRAYS & GRDG SH-1 G 374 3 3 7/28/82 E REACTOR AUX BLDG PENETRATION AREA-SECTIONS & DETAILS G 375 1 7 10/29/82 E REACTOR CONT BLDG PEN AREA-CND, TRAYS & GRDG SH-1 G 375 3 5 1/21/83 E REACTOR CONT BLDG PEN AREA-SECTIONS & DETAILS T1.7-3 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By___ ________________Title________________
G 375 4 3 8/31/82 E REACTOR CONT BLDG PEN AREA-TRAYS - KEY PLAN G 377 1 10 3/23/83 E REACTOR AUXILIARY BUILDING UNDERGROUND COND GRDG SH-1 G 378 2 10 1/18/83 E REACTOR AUXILIARY BUILDING UNDERGROUND COND & GRDG SH-2 G 380 8 1/18/83 E OUTLYING AREA CONDUIT GROUNDING & LIGHTING G 385 8 1/11/83 E INTAKE STRUCTURE CONDUIT &
LIGHTING G 386 7 1/21/83 E INTAKE STRUCTURE-LIGHTING SECTION & DETAILS G 388 8 3/11/83 E DIESEL GENERATOR BUILDING CONDUIT, GROUNDING & LIGHTING G 390 1 9 3/23/83 E REACTOR AUXILIARY BLDG EL-0.5 CONDUIT & TRAYS SH-1 G 391 2 9 1/11/83 E REACTOR AUXILIARY BLDG EL-0.5 CONDUIT & TRAYS SH-2 G 392 1 5 1/28/83 E REACTOR AUXILIARY BLDG EL 19'-6 CONDUIT TRAYS & GRDG SH-1 G 393 2 7 1/21/83 E REACTOR AUXILIARY BLDG EL 19'-6 CONDUIT TRAYS & GRDG SH-2 G 394 1 6 1/15/83 E REACTOR AUXILIARY BLDG EL 43'-0
& 62'-0 CND TRAYS & GRDG SH-1 G 394 3 6 1/21/83 E REACTOR AUXILIARY BLDG EL 62'-0 CND & GRDG SH-3 G 395 2 6 1/21/83 E REACTOR AUXILIARY BLDG EL 43'-0
& 62'-0 CND TRAYS & GRDG SH-2 G 396 1 7 12/15/82 E REACTOR AUXILIARY BLDG EL 43'-0 SECTIONS & DETAILS SH-1 T1.7-4 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By___ ________________Title________________
G 396 2 8 1/21/83 E REACTOR AUXILIARY BLDG EL 43'-0 SECTIONS & DETAILS SH-2 G 396 3 5 11/15/82 E REACTOR AUXILIARY BLDG SECTIONS & DETAILS SH-3 G 396 4 2 11/15/82 E REACTOR AUXILIARY BLDG SECTIONS & DETAILS SH-4 G 396 5 2 11/15/82 E REACTOR AUXILIARY BLDG SECTIONS & DETAILS SH-5 G 396 6 3 3/23/83 E REACTOR AUXILIARY BLDG SECTIONS & DETAILS SH-6 G 396 7 2 11/15/82 E REACTOR AUXILIARY BLDG SECTIONS & DETAILS SH-7 G 401 1 6 3/7/83 E FUEL HANDLING BUILDING CONDUIT TRAYS & GROUNDING SH-1 G 401 2 6 3/7/83 E FUEL HANDLING BUILDING CONDUIT TRAYS & GROUNDING SH-2 G 402 7 3/7/83 E FUEL HANDLING BUILDING CONDUIT SECTIONS & DETAILS B 404 0 5/30/78 E BOX DETAILS G 407 7 1/19/83 E YARD DUCT RUNS & LIGHTING G 407X 2 7/22/82 E YARD DUCT RUNS & LIGHTING G 408 1 6 2/7/83 E YARD DUCT RUNS & LIGHTING SECTIONS & DETAILS SH-1 G 408 LS 2A 1 1/11/83 E YARD DUCT RUNS & LIGHTING SECTIONS & DETAILS SH-2A G 408 2B 3 1/11/83 E STREAM TRESTLE AREA LTG &
DETAILS T1.7-5 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By___ ________________Title________________
G 408 1B 2 7/22/82 E COMPONENT COOLING LIGHTING G 409 7 3/7/83 E TRANSFORMER YARD CONDUIT GROUNDING & LIGHTING G 409 2 4 1/11/83 E XFRMR YD-PLAN XFRMR FIRE PROT
& 5KV & 6.9KV NON-SEG PHASE BUS G 409X 5 1/11/83 E TRANSFORMER YARD CONDUIT GROUNDING & LIGHTING G 409 2 4 1/11/83 E XFMR YD-PLAN XFMR FIRE PROT &
5KV & 6-9KV NON-SEG PHASE BUS G 410 1 5 3/7/83 E CABLE VAULT TRAYS-PLAN &
SECTIONS SH-1 G 410 2 4 11/18/82 E RTG BOARDS-TRAY RISERS-PLAN G 410 3 4 12/15/82 E RTE BOARDS-TRAY RISERS-SECT G 410 6 4 3/23/83 E CABLE VAULT TRAYS - KEY PLAN G 410 7 4 3/7/83 E RAB EL 74.0 CONDUIT TRAYS & GRDG G 410 8 3 1/18/83 E RAB EL 62'-0 CONDUIT & GROUNDING 2998-G-386 2 2 11/19/82 INTAKE STRUCTURE LIGHTING SECTION & DETAILS 2998-G-415 1 4 5/6/83 RAB RADIATION MONITORING SYSTEM CONDUIT & EQUIPMENT 2998-G-415 2 4 5/6/83 RAB RADIATION MONITORING SYSTEM CONDUIT & EQUIPMENT 2998-G-415 3 4 5/6/83 RAB RADIATION MONITORING SYSTEM CONDUIT & EQUIPMENT 2998-G-415 4 4 5/6/83 RAB RADIATION MONITORING SYSTEM CONDUIT & EQUIPMENT 2998-G-415 5 4 4/18/83 RCB RADIATION MONITORING SYSTEM CONDUIT & EQUIPMENT 2998-G-415 6 4 4/18/83 RCB RADIATION MONITORING SYSTEM CONDUIT & EQUIPMENT 2998-G-415 7 4 2/22/83 RCB RADIATION MONITORING SYSTEM CONDUIT & EQUIPMENT T1.7-6 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By___ ________________Title________________
2998-G-415 8 3 11/19/82 RCB RADIATION MONITORING SYSTEM CONDUIT & EQUIPMENT 2998-G-420 1 2 1/11/83 HEAT TRACE SYSTEM CONDUIT, FLOOR PENETRATION & EQUIPMENT LOCATION 2998-G-420 2 2 1/11/83 HEAT TRACE SYSTEM CONDUIT &
TRAY SECTIONS AND DETAILS 2998-G-420 3 2 12/22/82 HEAT TRACE SYSTEM CONDUIT & TRAY 2998-G-420 4 2 12/22/82 HEAT TRACE SYSTEM CONDUIT &
TRAY SECTIONS AND DETAILS 2998-G-420 5 2 12/22/82 HEAT TRACE SYSTEM CONDUIT & TRAY 2998-G-420 6 3 12/22/82 HEAT TRACE SYSTEM THERMOCOUPLE
& POWER JUNCTION BOXES 2998-G-420 7 3 1/18/83 HEAT TRACE SYSTEM THERMOCOUPLE
& POWER JUNCTION BOXES 2998-G-420 8 2 1/18/83 HEAT TRACE SYSTEM THERMOCOUPLE
& POWER JUNCTION BOXES 2998-G-420 9 3 1/18/83 HEAT TRACE SYSTEM THERMOCOUPLE
& POWER JUNCTION BOXES 2998-G-420 10 3 1/18/83 HEAT TRACE SYSTEM THERMOCOUPLE
& POWER JUNCTION BOXES 2998-G-420 11 3 1/18/83 HEAT TRACE SYSTEM THERMOCOUPLE
& POWER JUNCTION BOXES 2998-G-420 12 3 5/23/83 HEAT TRACE SYSTEM THERMOCOUPLE
& POWER JUNCTION BOXES 2998-G-420 13 3 5/23/83 HEAT TRACE SYSTEM THERMOCOUPLE
& POWER JUNCTION BOXES 2998-G-420 14 1 7/29/82 HEAT TRACE SYSTEM THERMOCOUPLE
& POWER JUNCTION BOXES 2998-G-420 15 1 7/29/82 HEAT TRACE SYSTEM THERMOCOUPLE
& POWER JUNCTION BOXES 2998-G-420 17 0 1/24/83 HEAT TRACE SYSTEM THERMOCOUPLE
& POWER JUNCTION BOXES T1.7-7 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By___ ________________Title________________
G 224 1 R8 8/24/82 E TURBINE BUILDING INSTRUMENT ARR SH-1 G 226 1 R7 12/3/82 E REACTOR BUILDING INSTRUMENT ARR SH-1 G 226 2 R8 3/7/83 E REACTOR BUILDING INSTRUMENT ARR SH-2 G 226 3 R7 2/1/83 E REACTOR BUILDING INSTRUMENT ARR SH-3 G 226 4 R7 2/1/83 E REACTOR BUILDING INSTRUMENT ARR SH-4 G 226 5 R7 2/1/83 E REACTOR BUILDING INSTRUMENT ARR SH-5 G 226 6 R7 10/24/82 E REACTOR BUILDING INSTRUMENT ARR SH-6 G 226 7 R6 10/24/82 E REACTOR BUILDING INSTRUMENT ARR SH-7 G 227 1 R8 3/7/83 E REACTOR AUXILIARY BUILDING INSTRUMENT ARR SH-1 G 227 2 R8 3/7/83 E REACTOR AUXILIARY BUILDING INSTRUMENT ARR SH-2 G 227 3 R4 6/21/81 E REACTOR AUXILIARY BUILDING INSTRUMENT ARR SH-3 G 227 4 R9 12/3/82 E REACTOR AUXILIARY BUILDING INSTRUMENT ARR SH-4 G 227 5 R6 3/7/83 E REACTOR AUXILIARY BUILDING INSTRUMENT ARR SH-5 G 227 6 R7 12/3/83 E REACTOR AUXILIARY BUILDING INSTRUMENT ARR SH-6 G 227 6 R6 1/22/82 E REACTOR AUXILIARY BUILDING INSTRUMENT ARR SH-7 G 227 8 R6 3/7/83 E REACTOR AUXILIARY BUILDING INSTRUMENT ARR SH-8 T1.7-8 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By___ ________________Title________________
G 228 1 R6 3/7/83 E FUEL HANDLING BUILDING INSTRUMENT ARR G 229 1 R6 1/13/82 E MISCELLANEOUS INSTRUMENT ARR B 231 604 SHTS VARIOUS E INSTRUMENT INSTALLATION DETAILS G 232 4 R4 3/7/83 E REACTOR AUX BLDG ANALYZER &
SAMPLING LINES ARR 5 R3 3/7/83 E REACTOR AUX ALDG ANALYZER &
SAMPLING LINES ARR 7 R3 2/1/83 E REACTOR BLDG ANALYZER &
SAMPLING LINES ARR 8 R2 10/22/81 E REACTOR BLDG ANALYZER &
SAMPLING LINES ARR 9 R2 10/22/81 E REACTOR BLDG ANALYZER &
SAMPLING LINES ARR G 233 1 R3 12/3/82 E REACTOR AUXILIARY BUILDING LABORATORY GAS SYSTEM LAYOUT G 278 R1 11/21/79 E CONTROL & BLOCK DIAGRAM CONTAINMENT SPRAY &
RECIRCULATION SYSTEM B 326 E SCHEMATIC DIAGRAMS 103S R2 5/26/83 E OIL LIFT PUMPS FOR REACTOR COOLANT PUMP P-2A1 (2B1, 2A2, 2B2-TYP.)
139S R2 4/11/83 E PRESSURIZER LEVEL CH-L-1110 159 R2 5/23/83 E VALVES V-2505, V2510, V2511
& V-2524 163S R1 5/26/83 E VALVES FCV-2210X, FCV-2210Y
& V-2512 174S R3 4/21/83 E BORIC ACID MAKE-UP PUMP 2A 175S R3 5/26/83 E BORIC ACID MAKE-UP PUMP 2B 177S R3 4/11/83 E CHARGING PUMP 2A T1.7-9 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-l (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By___ ________________Title________________
B 326 SCHEMATIC DIAGRAMS (Cont'd) 178S R3 5/26/83 E CHARGING PUMP 2B 179S R3 5/26/83 E CHARGING PUMP 2C 187S R2 4/21/83 E CHARGING PUMPS SEAL LUBRI-CATION SYSTEM VALVES V-2627, V-2628, V-2629 201S R2 4/21/83 E COMPONENT COOLING WATER PUMP 2A 203S R2 5/26/83 E COMPONENT COOLING WATER SUCTION HDR VALVE MV-14-3 (MV-14-1, 14-2 & 14-4-TYP.)
205S R2 4/21/83 E COMPONENT COOLING WATER PUMP 2B 209S R2 4/21/83 E COMPONENT COOLING WATER PUMP 2C 237S R2 4/21/83 E HP SAFETY INJECTION PUMP 2A 238S R2 4/21/83 E HP SAFETY INJECTION PUMP 2B 249S R2 5/23/83 E SHUTDOWN COOLING ISOLATION VALVE V-3480 (V-3481, V-3651, V-3652-TYP.)
251S R2 5/23/83 E LP SAFETY INJECTION PUMP 2A 252S R2 5/23/83 E LP SAFETY INJECTION PUMP 2B 257S R2 5/23/83 E LP SAFETY INJECTION FLOW CONT VALVES (HCV-3615, 3626, 3637, 3625, 3616, 3617, 3635, 3636, 3637, 3645, 3646, 3647-TYP.)
269S R2 5/23/83 E SAFETY INJECTION TANK 2A1 ISOL VALVE V-3624 (3614, 3634, 3644-TYP.)
285S R2 5/26/83 E CONTAINMENT FAN COOLER 2-HVS-1A (-1B,-1C,-1D-TYP.)
287S R2 5/26/83 E CONTAINMENT SPRAY PUMP 2A T1.7-10 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By___ ________________Title________________
B 326 SCHEMATIC DIAGRAMS (Cont'd) 289S R2 5/26/83 E CONTAINMENT SPRAY VALVES FCV-07-1A & FCV-07-1B 290S R2 5/26/83 E CONTAINMENT SPRAY PUMP 2B 297S R1 5/28/83 E REFUELING WATER TANK VALVE MV-07-1A (07-1B-TYP.)
299S R2 5/26/83 E REACTOR SUMP VALVE MV-07-2A (07-2B-TYP) 311S R2 5/26/83 E MAIN STEAM ISOLATION BYPASS VALVE MV-08-1A (08-113-TYP) 312S R3 5/26/83 E MAIN STEAM ISOL VALVE HCV-08-1A OPENING, CLOSING
& SOL TEST 315S R3 5/26/83 E MAIN STEAM ISOL VALVE HCV-08-1B OPENING, CLOSING
& SOL TEST 411S R2 5/26/83 E REACTOR TRIP BKR. TCB-1 482S R1 6/24/83 E REACTOR CONTAINMENT & SHLD BLDG DIFF PRESS 490S R2 6/24/83 E CONTROL ROOM EMERG FILTRATION FAN 2HVE-13A (13B-TYP) 492S R2(0) 7/22/83 E CONTROL ROOM AIR COND UNIT 2-HVA/ACC-3A(-3B,-3C TYP)
SH 1 493S R2(0) 7/22/83 E CONTROL ROOM AIR COND UNIT 2-HVA/ACC-3A(-3B,-3C TYP)
SH 2 503S R2 6/24/83 E REACTOR AUX BLDG EMERG EXHAUST FAN 2HVE-9A (9B-TYP) 505S R2 6/24/83 E REACTOR AUX BLDG SUPPLY FAN 2HVS-4A (4B-TYP) 507S R1 6/24/83 E CEDM COOLING FAN 2HVE-21A T1.7-11 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By___ ________________Title________________
B 326 SCHEMATIC DIAGRAMS (Con't')
508S R1 6/23/83 E CEDM COOLING FAN 2HVE-21F 509S R2 6/24/83 E REACTOR CONTAINMENT PURGE EXHAUST FAN 2HVE-8A 510S R2 6/24/83 E REACTOR CONTAINMENT PURGE EXHAUST FAN 2HVE-8B 511S R2 6/24/83 E REACTOR CONTAINMENT PURGE ISOLATION VALVES - SH. 1 512S R2 6/24/83 E REACTOR CONTAINMENT PURGE ISOLATION VALVES - SH. 2 513S R2 6/24/83 E SHIELD BLDG VENT EXHAUST FAN 2HVE-6A 516S R2 5/24/83 E SHIELD BLDG VENT EXHAUST FAN 2HVE-6B 629S R2 7/18/83 E AUX FEEDWATER PUMP 2A 630S R2 7/18/83 E AUX FEEDWATER PUMP 2B 631S R2 7/18/83 E AUX FEEDWATER PUMP 2C TURBINE AND STM VLV MV-08-3 711 R2(0) 7/22/83 E EMERG TURBINE TRIP & TURBINE ALARMS 832S R2 7/18/83 E INTAKE COOLING WATER PUMP 2A 833S R2 7/18/83 E INTAKE COOLING WATER PUMP 2B 834S R2 7/18/83 E INTAKE COOLING WATER PUMP 2C 835S R2 7/18/83 E INTAKE COOLING WATER NON-EMERG HDR A ISOL VALVE MV-21-3 (MV-21-2-TYP) 934S R2 6/24/83 E 4160V SWGR 2A2 FDR TO BUS 2A3 (2B2 FDR TO BUS 2B3-TYP)
T1.7-12 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet RevisionPrepared
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B 326 SCHEMATIC DIAGRAMS (Cont'd) 936S R2 6/24/83 E 4160V SWGR 2A3 INCOMING EEDER FROM BUS 2A2 (2B3 FDR ROM BUS 2B2-TYP) 938S R2 6/24/83 E 4160V SWGR 2A3 FDR TO BUS 2AB 2B3 FDR TO BUS 2AB-TYP) 940S R2 6/24/83 E 4160V SWGR 2AB INCOMING FEEDER FROM BUS 2A3 (2AB FDR FROM BUS 2B 3-TYP) 949S R2 6/24/83 E 4160V SWGR 2A3 LOAD SHEDDING RELAYS 950S R2 6/24/83 E 4160V SWGR 2B3 LOAD SHEDDING RELAYS 951S R1 6/24/83 E 4160V SWGR 2AB LOAD SHEDDING RELAYS 953S R2 6/24/83 E DIESEL GENERATOR 2A BREAKER 956S R2 6/24/83 E DIESEL GENERATOR 2A LOCKOUT RELAY 957S R2 6/24/83 E DIESEL GENERATOR 2A START CKTS SH 1 959S R2 6/24/83 E DIESEL GENERATOR 2A START SOLENOIDS 963S R2 6/24/83 E DIESEL GENERATOR 2B BREAKER 966S R2 6/24/83 E DIESEL GENERATOR 2B LOCKOUT RELAY 967S R2 6/24/83 E DIESEL GENERATOR 2B START CKTS SH 1 969S R2 6/24/83 E DIESEL GENERATOR 2B START SOLENOIDS 1000S R2 6/24/83 E 125V DC BUS TRANSFER CONTROL 1170S R2 6/24/83 E CONTROL ROOM NORTH OUTSIDE AIR INSUL VA FCV-25-M T1.7-13 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet RevisionPrepared
__No.__ _No._ No Date ___By___ ________________Title________________
B 326 SCHEMATIC DIAGRAMS (Cont'd) 1176S R2 6/24/83 E SHIELD BLDG VENT COOL AIR VA FCV-25-11 (FCV-25-12 TYP) 1501S R2(0) 7/22/83 E SHUTDOWN COOLING ISOL, HEAT EXCH, WARM-UP & CONTROL VALVES (V-3545, 3664, 3665, 3456, 3457, 3517, 3658, 3536, 3539; HCV-3657, 2512, 3306, 3301) 1601S R2 6/24/83 E DIESEL GENERATOR 2A START CKTS SH 2 1602S R2 6/24/83 E DIESEL GENERATOR 2A START CKTS SH 3 1603S R2 6/24/83 E DIESEL GENERATOR 2A START CKTS SH 4 1604S R2 6/24/83 E DIESEL GENERATOR 2A START CKTS SH 5 1605S R2 6/24/83 E DIESEL GENERATOR 2A START CKTS SH 6 1611S R2 6/24/83 E DIESEL GENERATOR 2B START CKTS SH 2 1612S R2 6/24/83 E DIESEL GENERATOR 2B START CKTS SH 3 1613S R2 6/24/83 E DIESEL GENERATOR 2B START CKTS SH 4 1614S R2 6/24/83 E DIESEL GENERATOR 2B START CKTS SH 5 1615S R2 6/24/83 E DIESEL GENERATOR 2B START CKTS SH 6 T1.7-14 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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1 R14 4/1/83 INDEX 2 R14 4/1/83 INDEX 3 R13 4/1/83 INDEX 4 R14 4/1/83 INDEX 5 R14 4/1/83 INDEX 6 R14 4/1/83 INDEX 7 R14 4/1/83 INDEX 8 R14 4/1/83 INDEX 8A R14 4/1/83 INDEX 8B R14 4/1/83 INDEX 8B-1 R4 4/1/83 INDEX 8B-2 R3 4/1/83 INDEX B-327 CONTROL WIRING DIAGRAM NUCLEAR INSTRUMENTATION 8DS R4 9/9/82 E ANNUNCIATOR REFLASH MODULES SH 1 8ES R4 9/9/82 E ANNUNCIATOR REFLASH MODULES SH 2 8FS R3 9/9/82 E ANNUNCIATOR REFLASH MODULES SH 3 8HS R3 9/9/82 E ANNUNCIATOR REFLASH MODULES SH 5 8IS R3 9/9/82 E ANNUNCIATOR REFLASH MODULES SH 6 5OS R5 9/9/82 E NUCLEAR INSTR SYS WIDE RANGE LOG CH-001A, 001B 51S R2 9/9/82 E NUCLEAR INSTR SYS WIDE RANGE LOG CH-001C, 001D 54S R4 11/5/82 E NUCLEAR INSTR. SYS. PWR RANGE SAF CH-003A/004A, 003B/004B, 003C/004C 55S R5 10/21/82 E NUCLEAR INSTR. SYS.PWR RANGE SAF CH-003D/004D 56S R2 7/2/82 E NUCLEAR INSTR. SYS.FLUX INDICATORS 60S R4 9/23/82 E OUT-OF-CORE NEUTRON DETECTORS NO. 1, 2, 5 & 9 T1.7-15 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By___ ________________Title________________
B-327 CONTROL WIRING DIAGRAM (Cont'd) 61S R4 9/23/82 E OUT-OF-CORE NEUTRON DETECTORS NO. 4, 6, 10 & 11 62S R4 9/23/82 E OUT-OF- CORE NEUTRON DETECTORS NO. 3 & 7 63S R4 9/23/82 E OUT-OF-CORE NEUTRON DETECTORS NO. 8 & 12 90S R9 10/6/82 E PRESSURIZER LEVEL CHANNEL L-1110 SH. 3 91S R1 5/26/83 E MEASUREMENT CHANNELS P-1105
& P-1106 101S R13 10/20/82 E REACTOR COOLANT PUMP 2A1 103S R11 5/26/83 E OIL LIFT PUMPS FOR REACTOR COOLANT PUMP 2A1 105S R12 11/10/82 E REACTOR COOLANT PUMP 2B1 107S R11 1/24/83 E OIL LIFT PUMPS FOR REACTOR COOLANT PUMP 2B1 109S R10 10/20/82 E REACTOR COOLANT PUMP 2A2 111S R11 1/24/83 E OIL LIFT PUMPS FOR REACTOR COOLANT PUMP 2A2 113S R10 10/20/82 E REACTOR COOLANT PUMP 2B2 115S R11 5/26/83 E OIL LIFT PUMPS FOR REACTOR COOLANT PUMP 2B2 118S R6 2/26/82 E PRESSURIZER RELIEF ISOLATION VALVE V-1477 120S R6 2/26/82 E PRESSURIZER RELIEF ISOLATION VALVE V-1476 136S R9 1/8/83 E REACTOR COOLANT LOOP TEMP CHT-1111Y, T-1111X & T-1115 137S R10 1/8/83 E REACTOR COOLANT LOOP TEMP CHT-1121Y, T-1121X & T-1125 T1.7-16 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By___ ________________Title________________
B-327 CONTROL WIRING DIAGRAM (Cont'd) 139S R9 1/31/83 E PRESSURIZER LEVEL CH L-1110 SH2 140S R12 5/26/83 E MEASUREMENT CHANNELS L-1103, L-1116 & P-1103 141S R9 4/14/83 E REACTOR HEAD SEAL P-1118 &
QUENCH TANK P-1116 - PRESS CHEMICAL & VOLUME SYSTEM 146S R4 12/20/82 E CHEM & VOL CONTROL SYSTEM-BORIC ACID HEAT TRACE TRANSF 2A 147S R4 12/20/82 E CHEM & VOL CONTROL SYSTEM-BORIC ACID HEAT TRACE TRANSF 2B 150S R9 1/8/83 E MEASUREMENT CHANNELS F-2212, P-2212, P-2215, T-2229 & T-2221 154S R6 8/14/82 E MEASUREMENT CHANNELS T-2225, P-2225, L-2227 & L-2226 157S R7 11/17/81 E LETDOWN STOP VA V-2515 AND LET-DOWN CONTAINMENT ISOL VA V-2516 159S R5 8/17/82 E VALVES V-2505, V-2510, V-2511&
V-2524 161S R8 12/11/82 E VOLUME CONTROL TANK DISCHARGE VALVE V-2501 163S R6 8/17/82 E VALVES FCV-2210X, FCV-2210Y &
V-2512 165S R9 12/15/82 E BORIC ACID GRAVITY FEED VALVE V-2508 166S R8 9/18/82 E BORIC ACID GRAVITY FEED VALVE V-2509 167S R8 9/17/82 E MAKE-UP BYPASS TO CHARGING PUMPS VALVE V-2514 174S R9 2/28/83 E BORIC ACID MAKE-UP PUMP 2A 175S R9 2/28/83 E BORIC ACID MARE-UP PUMP 2B T1.7-17 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By___ ________________Title________________
B-327 CONTROL WIRING DIAGRAM (Cont'd) 176S R7 8/26/82 E CHARGING LINES 2B1 & 2A2 VA'S I-SE-02-01 & I-SE-02-02 &
RECIRC DRAIN TK VA V-3661 177S R10 12/6/82 E CHARGING PUMP 2A 178S R10 12/6/82 E CHARGING PUMP 2B 179S R10 12/6/82 E CHARGING PUMP 2C 180S R6 9/9/82 E FUEL POOL PUMP 2A FUEL POOL SYSTEM 181S R8 4/20/82 E FUEL POOL PUMP 2B 182S R8 2/10/83 E FUEL POOL PURIFICATION PUMP COMPONENT COOLING WATER SYSTEM 187S R5 8/25/82 E CHARGING PUMP SEAL LUBE SYS VALVES V-2627, V-2628 & V-2629 188S R4 7/2/82 E CHEMICAL & VOLUME CONTROL SYSTEM ANN REFLASH CIRCUITS 189S R5 1/10/83 E AUX SPRAY VALVES 1-SE-02-03, 1-SE-02-04 190S R6 12/11/82 E BORON LOAD CONTROL VALVE V-2525 192S R9 3/24/83 E MAKE-UP SYSTEM CH F-2210 194S R7 8/4/82 E LETDOWN CONTROL & CHARGING LINE ISOL VALVES V-2522 & V-2523 196S R5 9/29/82 E CHARGING PUMP 2A BYPASS VALVE V-2555 197S R6 9/29/82 E CHARGING PUMP 2B BYPASS VALVE V-2554 198S R6 1/12/83 E CHARGING PUMP 2C BYPASS VALVE V-2553 201S R7 8/25/82 E COMPONENT COOLING WATER PUMP 2A T1.7-18 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By___ ________________Title________________
B-327 CONTROL WIRING DIAGRAM (Cont'd) 202S R7 3/3/82 E NORMAL SUPPLY HDR & NORMAL RE-TURN HDR ISOL VALVES 203S R6 9/18/82 E COMPONENT COOLING WATER SUCTION HDR A VALVE MV-14-3 204S R7 9/18/82 E COMPONENT COOLING WATER DISCH HDR A VALVE MV-14-1 205S R7 11/10/82 E COMPONENT COOLING WATER PUMP 2B 206S R6 12/6/82 E CCW FROM RCP'S 207S R6 9/18/82 E COMPONENT COOLING WATER SUCTION HDR B VALVE MV-14-4 208S R7 8/27/82 E COMPONENT COOLING WATER DISCH HDR B VALVE MV-14-2 209S R7 8/21/82 E COMPONENT COOLING WATER PUMP 2C 211S R10 2/28/83 E COMPONENT COOL WTR SHUTDN HT EXCH & SURGE TANK FILL VALVES 212S R5 6/3/82 E CCW TO & FROM REACTOR COOL PUMPS HCV-14-1, 2 & HCV-14-6, 7 217S R7 8/21/82 E COMP. COOL. WTR A FLOW &
PRESSURE 218S R8 9/9/82 E COMP. COOL. WTR B FLOW &
PRESSURE 220S R7 9/18/82 E COMP. COOL. WTR TO CONT COOL UNIT 2A VALVE MV-14-9 221S R5 9/18/82 E COMP. COOL. WTR FROM CONT. COOL.
UNIT 2A VALVE MV-14-10 222S R5 9/18/82 E COMP. COOL. WTR TO CONT. COOL.
UNIT 2B VALVE MV-14-11 223S R7 9/18/82 E COMP. COOL. WTR FROM CONT. COOL.
UNIT 2B VALVE MV-14-12 224S R5 9/18/82 E COMP. COOL. WTR TO CONT. COOL.
UNIT 2C VALVE MV-14-13 T1.7-19 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By___ ________________Title________________
B-327 CONTROL WIRING DIAGRAM (Cont'd) 225S R5 9/18/82 E COMP. COOL. WTR FROM CONT. COOL.
UNIT 2C VALVE MV-14-14 226S R5 9/18/82 E COMP. COOL. WTR TO COOL UNIT 2D VALVE MV-14-15 227S R6 9/27/82 E COMP. COOL. WTR FROM CONT. COOL.
UNIT 2D VALVE MV-14-16 228S R6 2/28/83 E COMP. COOL. HDR B TO FUEL POOL HT EXCH VALVE MV-14-17 229S R6 12/11/82 E COMP. COOL. HDR A TO FUEL POOL HT EXCH VALVE MV-14-18 230S R4 9/18/82 E COMP. COOL. HDR B FROM FUEL POOL HT EXCH VALVE MV-14-19 231S R5 2/12/83 E COMP. COOL. HDR A FROM FUEL POOL HT EXCH VALVE MV-14-20 SAFETY INJECTION 233S R9 9/18/82 E HP SAFETY INJECTION TO HOT LEG 2A VALVE V-3540 234S R8 9/18/82 E HP SAFETY INJECTION TO HOT LEG 2A VALVE V-3550 235S R8 9/18/82 E HP SAFETY INJECTION TO HOT LEG 2B VALVE V-3523 236S R8 9/18/82 E HP SAFETY INJECTION TO HOT LEG 2B VALVE V-3551 237S R5 11/17/81 E HP SAFETY INJECTION PUMP 2A 238S R6 9/3/82 E HP SAFETY INJECTION PUMP 2B 239S R2 10/7/80 E 4160V SWGR 2AB SPARE 242S R8 8/17/82 E SI TANK FILL & DRAIN VALVES I-SE-03-1A, I-SE-03-1B, I-SE-03-1C, I-SE-03-1D 244S R9 12/17/82 E MINIMUM FLOW ISOLATION VALVE V-3659 T1.7-20 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By___ ________________Title________________
B-327 CONTROL WIRING DIAGRAM (Cont'd) 245S R8 12/17/82 E MINIMUM FLOW ISOLATION VALVE V-3660 246S R7 3/25/83 E SAFETY INJECTION CHANNEL A-TRIP & BLOCK 247S R0 1/29/82 E SAFETY INJECTION TANK VENT VALVES 248S R7 3/25/83 E SAFETY INJECTION CHANNEL B-TRIP & BLOCK 249S R8 1/12/83 E SHUTDOWN COOLING ISOLATION VALVE V-3480 250S R9 1/12/83 E SHUTDOWN COOLING ISOLATION VALVE V-3481 251S R4 6/30/82 E LP SAFETY INJECTION PUMP 2A 252S R4 8/17/82 E LP SAFETY INJECTION PUMP 2B 253S R9 1/12/83 E SHUTDOWN COOLING ISOLATION VALVE V-3651 254S R9 1/12/83 E SHUTDOWN COOLING ISOLATION VALVE V-3652 255S R4 8/17/82 E ISOL VALVES V-3614, V-3624, V-3634 & V-3644 POSITION INDICATORS 256S R3 8/17/82 E N2 TO SI TANK VALVES V-3612, V-3622, V-3632 & V-3642 257S R9 12/15/82 E LOW PRESS SAFETY INJECT FLOW CONT VALVE HCV-3615 258S R7 9/18/82 E HIGH PRESS SAFETY INJECT FLOW CONT VALVE HCV-3626 259S R8 1/14/83 E AUX HIGH PRESS FLOW CONT VALVE HCV-3627 260S R10 1/14/83 E LOW PRESS SAFETY INJECT FLOW CONT VALVE HCV-3625 T1.7-21 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd) 261S R8 1/14/83 E HIGH PRESS SAFETY INJECT FLOW CONT VALVE HCV-3616 262S R7 9/18/82 E AUX HIGH PRESS FLOW CONT VALVE HCV-3617 263S R9 9/18/82 E LOW PRESS SAFETY INJECT FLOW CONT VALVE HCV-3635 264S R7 9/18/82 E HIGH PRESS SAFETY INJECT FLOW CONT VALVE HCV-3636 265S R9 1/14/83 E AUX HIGH PRESS FLOW CONT VALVE HCV-3637 266S R8 9/18/82 E LOW PRESS SAFETY INJECT FLOW CONT VALVE HCV-3645 267S R8 10/14/82 E HIGH PRESS SAFETY INJECT FLOW CONT VALVE HCV-3646 268S R8 10/14/82 E AUX HIGH PRESS FLOW CONT VALVE HCV-3647 269S R6 8/5/82 E SAFETY INJECT TANK 2A1 ISOL VALVE V-3624 270S R6 8/5/82 E SAFETY INJECT TANK 2A2 ISOL VALVE V-3614 271S R6 8/5/82 E SAFETY INJECT TANK 2B1 ISOL VALVE V-3634 272S R7 2/28/83 E SAFETY INJECT TANK 2B2 ISOL VALVE V-3644 273S R10 7/20/82 E MEASUREMENT CHANNELS F-3305, P-3307, P-3308, P-3309, P-3303X
& P-3303Y 275S R0 1/29/82 E SI TANK VENT VALVES V-3736, V-3734, V-3738, V-3740 277S R7 10/14/82 E HPSI PUMP DISCHARGE VALVE V-3654 279S R7 10/14/82 E HPSI PUMP DISCHARGE VALVE V-3656 T1.7-22 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd) 280S R6 8/4/82 E SI TANK 2A2 INSTR & CHECK VA LEAKAGE DRAIN TO RWT HCV-3618 281S R8 12/15/82 E SI TANK 2A1 INSTR & CHECK VA LEAKAGE DRAIN TO RWT HCV-3628 282S R6 8/4/82 E SI TANK 2B1 INSTR & CHECK VA LEAKAGE DRAIN TO RWT HCV-3638 283S R6 8/4/82 E SI TANK 2B2 INSTR & CHECK VA LEAKAGE DRAIN TO RWT HCV-3648 284S R7 8/24/82 E HIGH PRESSURE SAFETY INJECTION FLOW & PRESSURE MONITORS CONTAINMENT COOLING 285S R5 2/18/83 E CONTAINMENT FAN COOLER 2-HVS-1A 286S R5 2/18/83 E CONTAINMENT FAN COOLER 2-HVS-1B 287S R5 6/30/82 E CONTAINMENT SPRAY PUMP 2A 288S R6 5/2/83 E IODINE REMOVAL SYSTEM INSTRU-MENTATION 289S R7 8/11/82 E CONTAINMENT SPRAY VALVES FCV-07-1A & FCV-07-1B 290S R5 8/21/82 E CONTAINMENT SPRAY PUMP 2B 291S R5 3/11/83 E HYDRAZINE SYSTEM PUMP 2A 292S R5 3/11/83 E HYDRAZINE SYSTEM PUMP 2B 293S R9 7/30/82 E CONT PRESS, SPRAY HDR A PRESS
& FLOW & REFUEL WTR TANK LEVEL 294S R9 7/30/82 E CONT PRESS, SPRAY HDR B PRESS
& FLOW & REFUEL WTR TANK LEVEL 295S R7 7/30/82 E CONT PRESSURE & REFUELING WATER TANK LEVEL - 1 T1.7-23 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd) 296S R10 4/20/83 E CONT PRESSURE, TEMP & REFUELING WATER TANK LEVEL 297S R8 12/11/82 E REFUEL WATER TANK VALVE MV-07-1A 298S R8 12/11/82 E REFUEL WATER TANK VALVE MV-07-1B 299S R6 12/11/82 E REACTOR SUMP VALVE MV-07-2A 300S R6 12/11/82 E REACTOR SUMP VALVE MV-07-2B 302S R5 12/6/82 E CONTAINMENT SPRAY & RECIRC ACTUATION CH'S A MAN RESET 303S R7 12/6/82 E CONTAINMENT SPRAY & RECIRC ACTUATION CH'S B MAN RESET 304S R6 2/18/83 E CONTAINMENT FAN COOLER 2-HVS-1C 305S R6 2/18/83 E CONTAINMENT FAN COOLER 2-HVS-1D EC291393 307S R2 8/26/82 E MCC 2A9 FDR BKR (2-HVS-1A) 308S R2 3/30/82 E MCC 2A9 FDR BKR (2-HVS-1B) 309S R2 6/3/82 E MCC 2B9 FDR 8KR (2-HVS-1C) 310S R2 4/16/82 E MCC 2B9 FDR BKR (2-HVS-1D)
CONTAINMENT ISOLATION 311S R7 12/11/82 E MAIN STEAM ISOLATION BYPASS VALVE MV-08-1A 312S R10 8/20/82 E MAIN STEAM ISOL VALVE HCV-08-1A OPENING, CLOSING
& SOL TEST 313S R7 3/25/83 E MAIN STEAM ISOL VALVE HCV-08-1A STROKE TEST &
SOLENOID TEST T1.7-24 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd) 314S R7 12/11/82 E MAIN STEAM ISOLATION BYPASS VALVE MV-08-1B T1.7-25 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd)
CONTAINMENT ISOLATION (Cont'd) 315S R8 6/3/82 E MAIN STEAM ISOL VALVE HCV-08-1B OPENING, CLOSING
& SOL TEST 316S R7 3/25/83 E MAIN STEAM ISOL VALVE HCV-08-1B STROKE TEST &
SOLENOID TEST 317S R4 10/21/82 E INSTRUMENT AIR ISOLATION VALVE HCV-18-1 319S R8 5/11/83 E STEAM GEN BLOWDOWN ISOL VALVES FCV-23, 3, 4, 5 & 6 320S R6 1/10/83 E CONTAINMENT SAMPLE ISOLATION VALVES 321 R1 3/12/82 E CONTAINMENT ISOLATION VALVE I-SE-07-5A, -5C, -5E 322 R1 3/12/82 E CONTAINMENT ISOLATION VALVE I-SE-07-5B, -5D, -5F 323 R1 3/12/82 E CONTAINMENT PRESSURE CHANNELS P-07-4A1 & P-07-4B1 324 R1 3/26/82 E CONTAINMENT WATER LEVEL L-07-13A, -13B, -14A 330S R6 12/6/82 E CONTAINMENT ISOL CH A-MAN RESET & MAIN STM ISOL VA BLOCK A 331S R7 12/6/82 E CONTAINMENT ISOL CH B-MAN RESET & MAIN STM ISOL VA BLOCK B 333 R3 2/10/82 E CONT RADIATION MONITORS DETECTOR NO. RD-26-3 & RD-26-4 334 R3 2/10/82 E CONT RADIATION MONITORS DETECTOR NO. RD-26-5 & RD-26-6 T1.7-26 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd)
SPENT FUEL POOL 335S R5 11/10/82 E AREA RADIATION MONITOR DETECTOR NO.-RD-26-7 336S R5 11/24/82 E AREA RADIATION MONITOR DETECTOR NO.-RD-26-8 337 R2 2/26/82 E AREA RADIATION MONITOR DETECTOR NO. RD-26-9 338 R2 2/26/82 E AREA RADIATION MONITOR DETECTOR NO. RD-26-10 339 R2 2/26/82 E AREA RADIATION MONITOR DETECTOR NO. RD-26-11 340 R2 2/26/82 E AREA RADIATION MONITOR DETECTOR NO. RD-26-12 REACTOR PROTECTIVE SYSTEM 369S R5 12/6/82 E STEAM GENERATORS 2A/2B PRESSURE & LEVEL 370S R5 7/30/82 E PRESSURIZER PRESSURE & LEVEL 371S R3 4/16/82 E STEAM GENERATORS 2A & 2B LEVEL 372S R6 12/6/82 E PRESSURIZER PRESSURE P-1102A MEASUREMENT LOOP 373S R6 12/6/82 E PRESSURIZER PRESSURE P-1102B MEASUREMENT LOOP 374S R7 12/6/82 E PRESSURIZER PRESSURE P-1102C MEASUREMENT LOOP 375S R6 12/6/82 E PRESSURIZER PRESSURE P-1102D MEASUREMENT LOOP T1.7-27 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd) 376S R7 7/30/82 E STEAM GENERATOR 2A LEVEL 377S R7 7/30/82 E STEAM GENERATOR 2B LEVEL 378S R6 7/30/82 E STEAM GENERATOR 2A PRESSURE 379S R7 11/10/82 E STEAM GENERATOR 2B PRESSURE 381S R6 10/21/82 E REACTOR COOLANT TEMP CH T-1112A, T-1122A 382S R6 10/21/82 E REACTOR COOLANT TEMP CH T-1112B, T-1122B 383S R6 10/21/82 E REACTOR COOLANT TEMP CH T-1112C, T-1122C 384S R6 10/21/82 E REACTOR COOLANT TEMP CH T-1112D, T-1122D 385S R7 10/21/82 E REACTOR COOLANT DELTA FLOW CH P-1101A 386S R7 10/21/82 E REACTOR COOLANT DELTA FLOW CH P-1101B 387S R7 10/21/82 E REACTOR COOLANT DELTA FLOW CH P-1101C 388S R7 10/21/82 E REACTOR COOLANT DELTA FLOW CH P-1101D 392S R3 8/7/82 E RTGB-204 120V AC & 125V DC DISTRIBUTION 393S R3 8/31/82 E RTGB-203 28V DC DISTRIBUTION 395S R4 9/9/82 E RTGB-203 120V AC DISTRIBUTION SH 2 396S R6(0) 7/25/83 E RTGB-203 125V DC DISTRIBUTION REACTOR REGULATING SYSTEM 411S R7 9/28/82 E REACTOR TRIP BKP TCR-1 T1.7-28 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd) 412S R6 9/28/82 E REACTOR TRIP BKR TCB-5 413S R7 9/28/82 E REACTOR TRIP BKR TCB-2 414S R6 9/27/82 E REACTOR TRIP BKR TCB-6 415S R8 9/27/82 E REACTOR TRIP BKR TCB-3 416S R6 9/28/82 E REACTOR TRIP BKR TCB-7 417S R7 9/27/82 E REACTOR TRIP BKR TCB-4 418S R6 9/28/82 E REACTOR TRIP BKR TCB-8 419S R8 11/10/82 E REACTOR TRIP BKR TCB-9 424S R6 9/27/82 E REACTOR TRIP SWGR & CEDMC'S 120V AC & 125V DC DISTR AREA & PROCESS RADIATION MONITORING 332 R2 2/10/82 E POST-ACCIDENT MONITORS DETECTOR NOS. RD-26-38, RD-26-39 438S R2 1/29/82 E CONTAINMENT SAMPLING VALVES SH 1 439S R2 1/29/82 E CONTAINMENT SAMPLING VALVES SH 2 440S R2 1/29/82 E CONTAINMENT SAMPLING VALVES SH 3 441S R2 1/29/82 E CONTAINMENT SAMPLING VALVES SH 4 442S R1 2/10/82 E PROCESS RADIATION MONITOR 443S R6 2/28/83 E CONTAINMENT HIGH RANGE RAD MONITORS 444S R6 4/13/83 E COMPONENT COOLING WATER RADIATION MONITORING 445S R2 11/24/82 E PLANT VENT STACK & FUEL HANDLING BLDG VENT STACK RAD MONITORING 446S R5 2/12/83 E ECCS EFFLUENT GAS & PLANT VENT GAS WIDE RAD MONITORING T1.7-29 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd) 447S R0 11/13/81 E CCS EFFLUNET GAS (VENT A) WIDE RANGE RAD MONITORS SH 2 448S R1 2/10/82 E ECCS EFFLUENT GAS (VENT B) WIDE RANGE RAD MONITORS SH 1 449S R0 11/13/81 E ECCS EFFLUENT GAS (VENT B) WIDE RANGE RAD MONITORS SH 2 452S R1 2/10/82 E CONTROL ROOM OA1 (NORTH)
RADIATION MONITORS 453S R1 2/10/82 E CONTROL ROOM OA1 (SOUTH)
RADIATION MONITORS 455S R0 12/31/81 E FUEL POOL RAD MONITORING 2-OUT-OF-3 LOGIC SH 1 456S R0 12/31/81 E FUEL POOL RAD MONITORING 2-OUT-OF-3 LOGIC SH 2 457S R4 2/26/82 E CONTAINMENT RADIATION 461S R5 12/6/82 E STEAM GEN. BLOWDOWN SAMPLE ISOL. VALVES & SNUBBER OIL RESERVOIR LEVEL HEATING & VENTILATING 462S R1 10/26/78 E AUX BLDG & ECCS SYSTEM DAMPERS SH 1 OF 6 463S R2 9/10/82 E AUX BLDG & ECCS SYSTEM DAMPERS SH 2 OF 6 464S R2 9/10/82 E AUX BLDG & ECCS SYSTEM DAMPERS SH 3 OF 6 465S R7 3/29/83 E AUX BLDG & ECCS SYSTEM DAMPERS SH 4 OF 6 466S R4 3/30/82 E AUX BLDG & ECCS SYSTEM DAMPERS SH 5 OF 6 467S R4 3/30/82 E AUX BLDG & ECCS SYSTEM DAMPERS SH 6 OF 6 468S R7 2/22/83 E ELEC EQUIPMENT ROOM FANS T1.7-30 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd) 476S R13 3/1/83 E ELEC EQUP. RM SUPPLY FAN 2HVS-5A 477S R11 2/22/83 E ELEC EQUIP. RM SUPPLY FAN 2HVS -5B 478S R4 3/20/81 E TEMP RECORDER TR-25-2A MISC T/C'S 479S R5 12/15/82 E TEMP RECORDER TR-25-2B MISC T/C'S 481S R4 12/15/82 E AIRBORNE RADIOACTIVITY & ECCS VENT SYSTEM 482S R9 3/11/83 E REACTOR CONTAINMENT & SHLD BLDG DIFF. PRESS 483S R7 8/26/82 E TEMP RECORDERS TR-25-1A MISC T/C'S 487S R6 12/20/82 E CONTAINMENT TO ANNULUS & ECCS ROOM DIFF PRESS 490S R12 4/7/83 E CONTROL ROOM EMERG. FILTRATION FAB 2HVE-13A 491S R12 4/7/83 E CONTROL ROOMEMERG. FILTRATION FAB 2HVE-13B 492S R11 1/24/83 E CONTROL ROOM AIR COND. UNIT 2-HVA/ACC-3A 494S R11 1/24/83 E CONTROL ROOM AIR COND UNIT 2-HVA/ACC-3B 496S R9 1/24/83 E CONTROL ROOM AIR COND UNIT 2-HVA/ACC-3C 499S R4 2/18/83 E CONTROL ROOM FILTER & FAN INLET DAMPERS 500S R8 1/14/83 E CONTROL ROOM O.A.I. RADIATION DETECTORS 503S R7 3/1/83 E REACTOR AUX BLDG EMER EXH FAN 2HVE-9A T1.7-31 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd) 504S R7 12/20/82 E REACTOR AUX BLDG EMERG EXH FAN 2HVE-9B 505S R7 12/20/82 E REACTOR AUX BLDG SUPPLY FAN 2HVS-4A 506S R7 12/20/82 E REACTOR AUX BLDG SUPPLY FAN 2HVS-4B 507S R8 11/10/82 E CEDM COOLING FAN 2HVE-21A 508S R9 12/22/82 E CEDM COOLING FAN 2HVE-21B 509S R8 2/10/83 E REACTOR CONTAINMENT PURGE EXHAUST FAN 2HVE-8A 510S R7 2/18/83 E REACTOR CONTAINMENT PURGE EXHAUST FAN 2HVE-8B 511S R7 7/16/82 E REACTOR CONTAINMENT PURGE ISOLATION VALVES SH 1 512S R7 6/17/82 E REACTOR CONTAINMENT PURGE ISOLATION VALVES SH 2 513S R7 11/10/82 E SHIELD BLDG VENT EXH FAN 2HVE-6A 516S R7 11/10/82 E SHIELD BLDG VENT EXH FAN 2HVE-6B 517S R4 9/17/82 E FUEL POOL DIFF PRESS & HSCP ROOM FANS 518S R2 3/20/81 E DIESEL GEN 2A BLDG FAN 2-RV-5 519S R2 3/20/81 E DIESEL GEN 2B BLDG FAN 2-RV-6 522S R6 3/1/83 E REACTOR CAVITY COOLING SYSTEM 2HVS-2A 523S R7 11/10/82 E REACTOR CAVITY COOLING SYSTEM 2HVS-2B 524S R5 3/1/83 E REACTOR SUPPORT COOLING SYSTEM 2HVE-3A T1.7-32 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd) 525S R4 5/11/82 E REACTOR SUPPORT COOLING SYSTEM 2HVE-3B 529S R6 1/24/83 E CONTAINMENT VACUUM RELIEF VALVES FCV-25-7 & FCV-25-8 WASTE MANAGEMENT & SAMPLING 532S R8 4/21/83 E SAFEGUARDS ROOM "A" SUMP PUMPS 533S R9 4/21/83 E SAFEGUARDS ROOM "B" SUMP PUMPS 536S R1 4/28/80 E DRAIN VALVES TO REACTOR AUXILIARY BUILDING SUMPS - SH 1 542S R4 9/7/82 E REACTOR DRAIN PUMP 2A 543S R5 9/10/82 E REACTOR DRAIN PUMP 2B 563S R4 9/7/82 E RDT VENT STOP & CONT ISOL VALVES V-6300, V-6341, & V-6342 564S R6 12/6/82 E WASTE GAS CONT ISOL & STOP VALVES V-6718, V-6750, & V-6565 566S R5 9/7/82 E N2 HDR CONT ISOL & DISCH STOP VALVES V-6741 & V-6728 576S R7 4/1/83 E REACTOR SUMP ISOL VALVES LCV-07-11A & LCV-07-11B AND REACTOR CAVITY LEAK DETECTORS 578S R5 7/20/82 E PRIMARY COOLANT SAMPLES VALVES V-5200 & V 5203 579S R3 4/3/81 E PRESSURIZER SURGE SAMPLE VALVES V-5201 & V-5204 580S R4 8/21/82 E PRESSURIZER STEAM SAMPLE VALVES V-5202 & V-5205 586S R1 4/28/80 E DRAIN VALVES TO REACTOR TO AUX BLDG SUMPS - SH 2 T1.7-33 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd) EC 295 FEEDWATER 353 601S R4 7/2/82 E AUX FW HDR'S A&B FLOW &
PRESSURE 602S R7 2/28/83 E AUX FW HDR C FLOW & PRESSURE &
FWP 2A & 2B FLOW 603S R9 6/24/83 E STM GEN 2A & 2B ATM STM DUMP FWP DISCH HDR PRESS SH 1 608S R8(0) 7/25/83 E AUX FWP 2A DISCHARGE TO ST.
GEN 2A MV-09-9 609S R9(0) 7/25/83 E AUX FWP 2B DISCHARGE TO ST.
GEN 2B MV-09-10 610S R9 12/11/82 E AUX FWP 2A DISCHARGE TO ST.
GEN 2B MV-09-13 611S R10 1/14/83 E AUX FWP 2B DISCHARGE TO ST.
GEN 2A MV-09-14 612S R8(0) 7/25/83 E AUX FWP 2C DISCHARGE TO ST.
GEN 2A MV-09-11 613S R7(0) 7/25/83 E AUX FWP 2C DISCHARGE TO ST.
GEN 2A MV-09-12 629S R8 3/1/83 E AUX FEEDWATER PUMP 2A 630S R8 5/26/83 E AUX FEEDWATER PUMP 2B 631S R7 5/26/83 E AUX FEEDWATER PUMP 2C - TURBINE 632S R8 5/26/83 E AUX FEEDWATER PUMP 2C - STEAM VALVE MV-08-3 T1.7-34 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd) 638S R6 3/30/82 E SG 2A/2B TO AFWP 2C WARM-UP VALVES I-SE-08-1, 2 639S R4 8/24/82 E RTGB-202 45VDC DISTRIBUTION 643S R5 8/5/82 E RTGB-202 120VAC DISTRIBUTION SH.1 645S R5 8/26/82 E RTGB-205 125VDC&120VAC DISTR.
646S R3 8/21/82 E RTGB-206 125V DC DISTRIBUTION 647S R6 9/17/82 E RTGB-206 120V AC DISTRIBUTION SH.1 648S R2 3/6/81 E RTGB-206 120V AC DISTRIBUTION SH.2 649S R4 8/5/82 E HOT SHUTDOWN CONTROL PANEL 120VAC DISTRIBUTION 652S R8 5/26/83 E SG 2A TO AFWP 2C TURBINE MV-08-13 653S R7 1/17/83 E SG 2B TO AFWP 2C TURBINE MV-08-12 654S R6 6/24/83 E STM GEN 2A&2B ATM STM DUMP FWP DISCH HDR PRESS SH.2 655S R4 2/14/83 E MAIN FEEDWATER ISOLATION VALVE HCV-09-1A 656S R4 2/14/83 E MAIN FEEDWATER ISOLATION VALVE HCV-09-113 657S R5 4/1/83 E RTGB 205, 125VDC&120VAC DISTRIBUTION SH.2 658S R3 9/2/82 E RTGB 205, 120VAC DISTRIBUTION 664S R4 8/11/82 E RTGB 206, 45VDC DISTRIBUTION 671S R5 5/11/83 E MAIN FEEDWATER ISOLATION VALVE HCV-09-2A 672S R6 5/11/83 E MAIN FEEDWATER ISOLATION VALVE HCV-09-2B T1.7-35 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd)
MAIN STEAM 695S R7 8/15/82 E AUX FWP 2C & TURB INLET PRESS
& STM GEN FLOW PRESSURE TURBINE 709S R3 1/18/82 E TURBINE TRIP STEAM GEN. HIGH-HIGH LEVEL 710S R11 1/7/83 E TURBINE AUTO-STOP-TRIP &
TURBINE ALARMS 743S R8 9/7/82 E CONDENSATE TRANSFER PUMP 744S R7 2/12/83 E AUXILIARY STEAM SJAE STM&FEED PUMP SUCT HDR PRESS COND STM TK & HOTWELL LEVEL TURBINE INSTRUMENTATION 800S R5 6/1/81 E RTGB-201, 125VDC&120VAC DISTRIBUTION TURBINE COOLING 831S R6 9/3/82 E INTAKE COOL WTR DISCH HDR PRESS PUMP 2A & PUMP 2B 832S R7 7/18/83 E INTAKE COOLING WATER PUMP 2A 833S R7 7/18/83 E INTAKE COOLING WATER PUMP 2B 834S R6 7/18/83 E INTAKE COOLING WATER PUMP 2C 835S R5 10/21/82 E INTAKE COOL WTR NON EMER HDR A ISOL VALVE MV-21-3 836S R5 8/19/82 E INTAKE COOL WTR NON EMER HDR B ISOL VALVE MV-21-2 839S R5 4/28/82 E LUBE WATER SUPPLY STRAINERS T1.7-36 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd)
PRIMARY WATER 849S R4 12/17/82 E PRIMARY WATER ISOLATION VALVE HCV-15-1 STATION AUXILIARY POWER 924S R2 1/18/82 E 4160V SWGR 2A3 DIFF. RELAY 925S R2 10/6/82 E 4160V SWGR 2B3 DIFF. RELAY 926S R3 6/9/81 E 4160V SWGR 2AB DIFF. RELAY 931S R4 6/9/81 E 4160V SWGR 2A3 AC-DC DISTR
& HEATERS 932S R4 6/9/81 E 4160 SWGR 2B3 AC-DC DISTR &
HEATERS 933S R4 6/9/81 E 4160V SWGR 2AB AC-DC DISTR
& HEATERS 934S R6 1/26/83 E 4160V SWGR 2A2 FDR TO BUS 2A3 935S R6 1/26/83 E 4160V SWGR 2B2 FDR TO BUS 2B3 936S R7 3/2/83 E 4160V SWGR 2A3 INCOMING FDR.
FROM BUS 2A2 937S R6 3/2/83 E 4160V SWGR 2B3 INCOMING FEEDER FROM BUS 2B2 938S R4 1/26/83 E 4160V SWGR 2A3 FDR TO BUS 2AB 939S R4 1/26/83 E 4160V SWGR 2B3 FDR TO BUS 2AB 940S R5 1/26/83 E 4160V SWGR 2AB INCOMING FDR FROM BUS 2A3 941S R6 4/13/83 E 4160V SWGR 2AB INCOMING FDR FROM BUS 2B3 942S R2 9/7/82 E 4160V SWGR 2AB RELAYING &
METERING 943S R8 6/24/83 E PRESS HTR TRANSF 2A3 4160V FDR BKR T1.7-37 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd) 944S R8 6/24/83 E PRESS HTR TRANSF 2B3 4160V 946S R5 4/13/83 E 480V STA SERV TRANSF 2A2 4160V FDR BKR 948S R4 9/1/82 E 480V STA SERV TRANSF 2B2 4160V FDR BKR 949S R8 12/16/82 E 4160V SWGR 2A3 LOAD SHEDDING RELAYS 950S R5 12/16/82 E 4160V SWGR 2B3 LOAD SHEDDING RELAYS 951S R4 1/11/82 E 4160V SWGR 2AB LOAD SHEDDING RELAYS EMERGENCY DIESEL GENERATOR 953S R4 12/16/82 E DIESEL GENERATOR 2A BREAKER 954S R7 12/16/82 E DIESEL GENERATOR 2A RELAYING
& METERING 955S R6 12/16/82 E DIESEL GENERATOR 2A INSTR. &
DIFF RELAYING 956S R5 6/24/83 E DIESEL GENERATOR 2A LOCKOUT RELAY 957S R8 3/2/83 E DIESEL GENERATOR 2A START CKT'S - SH.1 958S R5 3/2/83 E DIESEL GENERATOR 2A REMOTE CONTROL 959S R1 1/28/80 E DIESEL GENERATOR 2A START SOLENOIDS 960S R1 9/7/82 E DIESEL GENERATOR 2A ANNUNCIA-TOR SH.1 961S R2 12/10/82 E DIESEL GENERATOR 2A ANNUNCIA-TOR SH.2 962S R2 9/16/81 E D-G 2A ENG CYL'S TEMP & EXH DIFF TEMP MONITORING T1.7-38 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd) 963S R5 12/16/82 E DIESEL GENERATOR 2B BREAKER 964S R8 12/16/82 E DIESEL GENERATOR 2B RELAYING
& METERING 965S R6 12/16/82 E DIESEL GENERATOR 2B INSTR &
DIFF RELAYING 966S R6 6/24/83 E DIESEL GENERATOR 2B LOCKOUT RELAYS 967S R8 3/2/83 E DIESEL GENERATOR 2B START CKT'S - SH.1 968S R5 3/2/83 E DIESEL GENERATOR 2B REMOTE CONTROL 969S R2 1/8/82 E DIESEL GENERATOR 2B START SOLENOIDS 970S R2 9/7/82 E DIESEL GENERATOR 2B ANNUNCIA-TORS-SH.1 971S R2 12/10/82 E DIESEL GENERATOR 2B ANNUNCIA-TORS-SH.2 972S R3 9/16/81 E D-G 2B ENG CYL'S TEMP & EXH DIFF TEMP MONITORING 974S R2 4/16/82 E D-G 2A ENG CYC'S TEMP EXH DIFF TEMP MONITORING 480V AUXILIARY POWER 977S R3 9/7/82 E 480V SWGR 2A2 FDR 978S R4 8/3/82 E 480V SWGR 2A2 - 2AB TIE 979S R3 8/3/82 E 480V SWGR 2AB - 2A2 TIE 980S R2 1/24/83 E 480V SWGR 2B2 FDR 981S R3 8/3/82 E 480V SWGR 2B2-2AB TIE 982S R4 8/3/82 E 480V SWGR 2AB-2B2 TIE 983S R2 4/16/82 E 480V SWGR SPARE COMPARTMENTS 984S R3 3/30/82 E 480V SWGR 2A2 FDR TO FUEL HANDLING MCC 2A8 T1.7-39 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd) 985S R3 4/16/82 E 480V SWGR 2B2 FDR TO FUEL HANDLING MCC 2B8 990S R12 2/1/83 E 480V SWGR 2A2 MET REL & HTR'S 991S R4 8/25/82 E 480V SWGR 2A2 MCC FEEDERS 992S R12 2/1/83 E 480V SWGR 2B2 MET REL & HTR'S 993S R5 8/25/82 E 480V SWGR 2B2 MCC FEEDERS 994S R6 3/11/83 E 480V SWGR 2AV MEG. REL &
HTR'S 995S R2 2/20/81 E 480V SWGR 2AB FEEDER TO REACTOR AREA MCC 2AB 996S R6 3/26/82 E EMERGENCY DIESEL GEN NO. 2A LOADING LIGHTS 997S R8 3/25/83 E EMERGENCY DIESEL GEN. NO. 2B LOADING LIGHTS 998S R6 9/27/82 E EMERGENCY DIESEL GEN'S NO. 2A
& NO. 2B LOADING LIGHTS MISCELLANEOUS ELECTRICAL 999S R7 1/10/83 E BATTERY 2C & BATTERY CHARGER 2C 1000S R2 10/17/80 E 125 VDC BUS TRANSFER CONTROL 1001S R9 4/18/83 E BATTERY 2A BATTERY CHARGER 2A 1002S R8 4/14/83 E BATTERY 2B BATTERY CHARGER 2B 1003S R6 12/20/82 E BATTERY CHARGER 2AB 1004S R4 4/20/83 E ISOL CAB'S 125V DC POWER SUPPLY 1005S R4 9/1/82 E MOTOR SPACE HEATER FEEDERS 1006S R3 3/1/83 E MOTOR SPACE HEATER FEEDERS 1007S R6 4/28/82 E MISC ANNUNCIATIONS 1008S R7 3/11/83 E VITAL AC BUS POWER SUPPLY (SUPS )
T1.7-40 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd) 1009S R11 5/26/83 E INSTRUMENT BUSES & INVERTERS 2MA & 2 MC 1010S R11 5/26/83 E INSTRUMENT BUSES & INVERTERS 2MB & 2MD 1024S R1 11/10/82 E MCC 2A1,2B1,2A3,2B3,2A9,2B9, SP 2HTRS 1026S R(0) 7/25/83 E MCC 2A5,2B5,2AB,2A7,2B7 SP HTRS 1027S R2 11/10/82 E MCC 2A6,2B6,2AB,2B SP HTRS EMERGENCY DIESEL GENERATOR 1117S R2 10/6/82 E DIESEL GENERATOR 2A ANN CKT'S - SH 1 1118S R3 11/10/82 E DIESEL GENERATOR 2A ANN CKT'S - SH 2 1119S R7 6/24/83 E DIESEL GENERATOR 2A ANN CKT'S - SH 3 1120S R3 8/26/82 E DIESEL GENERATOR 2A LUBE OIL CIRC. PUMP 2A1 1121S R3 8/26/82 E DIESEL GENERATOR 2A LUBE OIL CIRC. PUMP 2A2 1126S R6 4/5/83 E DIESEL GEN FUEL OIL TRANSFER PUMP 2A 1127S R2 1/8/82 E DIESEL GENERATOR 2B ANN CKT'S SH 1 1128S R4 11/10/82 E DIESEL GENERATOR 2B ANN CKT'S SH 2 1129S R6 6/24/83 E DIESEL GENERATOR 2B ANN CKT'S SH 3 1130S R4 8/26/82 E DIESEL GENERATOR 2B LUBE OIL CIRC. PUMP 2B1 1131S R4 8/26/82 E DIESEL GENERATOR 2B LUBE OIL CIRC. PUMP 2B2 T1.7-41 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
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B-327 CONTROL WIRING DIAGRAM (Cont'd)
MISCELLANEOUS HVAC 1136S R5 4/5/83 E DIESEL GEN FUEL OIL TRANSFER PUMP 2B 1137S R3 4/7/82 E TEMPERATURE RECORDER TR-25-1B MISC THERMOCOUPLES 1138S R5 12/15/82 E HYDRAMOTOR ACTUATORS FOR FANS 2HVE-9A & 2HVE-9B 1139S R3 1/17/83 E HYDRAMOTOR ACTUATORS FOR FANS 2HVE-13A & 2HVE-13B 1140S R4 11/11/82 E SHIELD BLDG VENT SYS D-23 DAMPER CONTROL 1141S R4 11/11/82 E SHIELD BLDG VENT SYS D-24 DAMPER CONTROL 1142S R6 4/21/83 E PLANT AUXILIARIES CONTROL BOARD, ANNUNCIATOR-LA 1143S R6 4/21/83 E PLANT AUXILIARIES CONTROL BOARD ANNUNCIATOR-LB 1149S R4 3/6/82 E PLANT AUX. CONTROL BOARD ANN.
LA, LB INTER. WIRING 1150S R7 3/11/83 E SHIELD BLDG VENT SYSTEM ELECTRIC HEATING COILS 2-HVE-6A1, 6A2 1152S R7 3/11/83 E SHIELD BLDG VENT SYSTEM ELECTRIC HEATING COILS 2-HVE-6B1, 6B2 1154S R5 12/11/82 E FUEL HANDLING BLDG EMERG.
VENT VALVE FCV-25-30 1155S R4 12/11/82 E FUEL HANDLING BLDG EMERG.
VENT VALVE FCV-25-31 1156S R4 12/11/82 E SHIELD BLDG VENT SYSTEM ISOL VALVE FCV-25-32 1157S R4 12/11/82 E SHIELD BLDG VENT SYSTEM ISOL VALVE FCV-25-33 T1.7-42 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
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B-327 CONTROL WIRING DIAGRAM (Cont'd) 1158S R6 4/5/80 E CONT. CONTAIN./H2 PU DISCH.
TO SHIELD BLDG. VENT SYS.
FCV-25-29 1159S R5 10/21/82 E CONT. CONTAIN./H2 PU DISCH.
TO SHIELD BLDG. VENT SYS.
FCV-25-34 1160S R6 12/17/82 E CONT. CONTAIN./H2 PURGE ISOL VALVE FCV-25-20 1161S R5 12/17/82 E CONT. CONTAIN./H2 PURGE ISOL.
VALVE FCV-25-21 1162S R5 12/20/82 E INTAKE STRUCTURE EXHAUST FAN 2HVE-41A 1163S R5 12/20/82 E INTAKE STRUCTURE EXHAUST FAN 2 HVE-41B 1164S R4 7/20/82 E CONT. CONTAIN./H2 PURGE ISOL.
VALVE FCV-25-26 1165S R4 12/6/82 E SHIELD BLDG. HEPA FILTERS &
CHARCOAL ADSORBER DIFF PRESS 1166S R6 1/25/83 E CONTROL ROOM DAMPERS D-39 &
D-40 & DIFF PRESSURES 1167S R3 4/3/81 E CONTROL ROOM HEPA FILTER DIFF PRESSURES 1168S R3 7/1/82 E FUEL HDLG BLDG HEATING & VENT RM FAN 2HVE-17 1169S R9 12/20/82 E ROOF VENTILATORS 2RV3 & 2RV4 1170S R6 10/14/82 E CONTROL ROOM NORTH OAI ISOL VA FCV-25-14 1171S R5 10/14/82 E CONTROL ROOM SOUTH OAI ISOL VA FCV-25-15 1172S R7 10/14/82 E CONTROL ROOM NORTH OAI ISOL VA FCV-25-16 1173S R5 9/27/82 E CONTROL ROOM SOUTH OAI ISOL VA FCV-25-17 T1.7-43 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By__ ________________Title________________
B-327 CONTROL WIRING DIAGRAM (Cont'd) 1174S R3 1/29/82 E TOILET EXH FAN ISOL VA FCV-25-18 1175S R4 12/20/82 E TOILET EXH FAN ISOL VA FCV-25-19 1176S R6 9/27/82 E SHIELD BLDG VENT COOL AIR VALVE FCV-25-11 1177S R4 9/27/82 E SHEILD BLDG VENT COOL AIR VALVE FCV-25-12 1178S R6 9/27/82 E SHIELD BLDG VENT SYSTEM TIE VALVE FCV-25-13 1182S R3 4/6/82 E FUEL HANDLING BLDG DAMPERS RAD SIGNAL A 1183S R3 4/6/82 E FUEL HANDLING BLDG DAMPERS RAD SIGNAL B 1189S R1 4/6/82 1190S R2 2/20/81 E KITCHEN EXHAUST FAN ISOL VALVE FCV-25-24 1191S R2 2/20/81 E KITCHEN EXHAUST FAN ISOL VALVE FCV-25-25 1192S R0 1/29/82 1196S R4 2/28/83 E CONTAINMENT ATMOSPHERE HYDROGEN ANALYZER SH.1 1197S R4 7/26/82 E CONTAINMENT ATMOSPHERE HYDROGEN ANALYZER-2 1204S R2 7/26/82 E CONTAINMENT ATMOSPHERE HYDROGEN ANALYZER-3 1205S R4 2/28/83 E CONTAINMENT ATMOSPHERE HYDROGEN ANALYZER-4 1217S R3 7/27/81 E ANNUNCIATOR REFLASH 1219S R4 12/20/82 E BATTERY ROOM 2A ROOF VENTILA-TOR-2RV-1 1220S R4 12/20/82 E BATTERY ROOM 2B ROOF VENTILA-TOR-2RV-2 T1.7-44 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By__ ________________Title________________
B-327 CONTROL WIRING DIAGRAM (Cont'd) 1238S R2 3/20/81 E HVCB-45V DC DISTRIBUTION 1239S R3 4/6/82 E HVCB-125V DC & 120V AC DISTRIBUTION SH 1 1240S R3 10/7/80 E HVCB-120V AC DISTRIBUTION SH 2 1253S R3 9/2/82 E MOTOR OPER VALVE SPACE HEATERS FEEDERS 1254S R3 10/6/82 E MOTOR OPER VALVE SPACE HEATERS FEEDERS 1255S R6 7/26/82 E MOTOR OPER VALVE SPACE HEATERS FEEDERS 1256S R4 7/26/82 E MOTOR OPER VALVE SPACE HEATERS FEEDERS 1257S R4 7/26/82 E MOTOR OPER VALVE SPACE HEATERS FEEDERS 1260S R3 12/1/81 E MOTOR OPER VALVE SPACE HEATERS FEEDERS 1276S R1 3/28/80 E 480V SWGR SPACE COMPARTMENT 1278S R2 7/26/83 E 480V SWGR 2A2, 2B2 INTER-(0) CONNECTIONS BETWEEN CUBICLES 1279S R4 8/25/81 E 480V SWGR 2AB INTERCONNEC-TIONS BETWEEN WIRING BOXES SAFETY INJECTION & SHUTDOWN COOLING______________________
1501S R10 11/11/82 E SHUTDOWN COOLING ISOL. VALVE V-3545 1502S R7 1/12/83 E SHUTDOWN COOLING ISOL. VALVE V-3664 1503S R7 9/27/82 E SHUTDOWN COOLING ISOL. VALVE V-3665 1504S R6 9/27/82 E SHUTDOWN CLG FROM.HEAT EXCH 2A VALVE V-3456 T1.7-45 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By__ ________________Title________________
B-327 CONTROL WIRING DIAGRAM (Cont'd) 1505S R7 2/14/83 E SHUTDOWN CLG FROM HEAT EXCH.
- 2B VALVE V-3457 1506S R4 9/27/82 E SHUTDOWN CLG HEAT EXCH. - 2A INLET VALVE V-3517 1507S R4 9/27/82 E SHUTDOWN CLG HEAT EXCH.- 2B INLET VALVE V-3658 1508S R2 8/05/82 E RECORDER DISTRIBUTION MODULE INTERCONNECTION SH 1 1510S R6 9/29/82 E SHUTDN. CLG.LINE 2A WARM-UP VALVE V-3536 1511S R9 9/29/82 E SHUTDN. CLG. LINE 2B WARM-UP VALVE V-3539 1512S R7 7/18/83 E HP INJECTION TO HOT LOOP 2A FLOW & PRESS MONITORS 1513S R5 8/05/82 E HP INJECTION TO HOT LOOP 2B FLOW & PRESS MONITORS 1514S R6 9/29/82 E SHUTDN. COOLING CONTROL VALVE 2A HCV-3657 1515S R7 9/29/82 E SHUTDN. COOLING CONTROL VALVE 2B HCV-3512 1516S R8(0) 7/26/82 E SHUTDOWN COOLING & BYPASS VALVE FCV-3306 1517S R8(0) 7/26/82 E SHUTDOWN COOLING & BYPASS VALVE FCV-3301 1518S R4 4/13/83 E RECORDER DISTRIBUTION MODULE INTERCONNECTION-SH 2 1519S R7 8/21/82 E HOT LEG HPSI LINE CHECK VLV LEAK'G DRAIN VA'S V-3571, V-3572, I-SE-03-2A, I-SE-03-2B 1520S R6 7/18/83 E MINIMUM FLOW ISOLATION VALVES V-3495 & V-3496 T1.7-46 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By__ ________________Title________________
B-327 CONTROL WIRING DIAGRAM (Cont'd) 1525S R3 1/13/82 E MEASUREMENT CHANNELS T-3351 X/Y, 3352 X/Y, 3303 W/X/Y/Z 1526S R6 3/10/82 E MEASUREMENT CHANNELS P3301, P3302, P-3304, P-3307 1527S R4 10/14/82 E SI TANKS 2A1, 2A2, 2B1, 2B2 SAMPLE VA'S I-SE-05-1A, 1B, 1C & 1D 1528S R11 1/13/83 E SI TANKS SAMPLE FCV-03-1E MEASUREMENT CH'S F-3301, F-3306 1529S R2 2/26/82 E CONTAINMENT SPRAY ISOLATION VALVE MV-07-161 1530S R2 2/26/82 E CONTAINMENT SPRAY ISOLATION VALVE MV-07-164 1531S R2 2/26/82 E LPSI PUMP 2A SUCTION VALVE V-3432 1532S R2 2/26/82 E LPSI PUMP 2B SUCTION VALVE V-3444 ANNUNCIATORS 1551S R5 10/14/82 E ISOL CAB/ALC-1 INTERCONN DIAGRAM RTGB-201 ANN B SH 1 1552S R6 8/27/82 E ISOL CAB/ALC-1 INTERCONN DISGRAM RTGB-201 ANN B SH 2 1553S R6 11/5/82 E ISOL CAB/ALC-1 INTERCONN DIAGRAM RTGB-201 ANN A SH 1 1554S R6 11/5/82 E ISOL CAB/ALC-1 ,2 INTERCONN DIAGRAM RTGB-201, 204 ANN A SH 2, L 1555S R6 11/5/82 E ISOL CAB/ALC-2 INTERCONN DIAGRAM RTGB-202 ANN G 1556S R6(0) 7/26/82 E ISOL CAB/ALC-2 INTERCONN DIAGRAM RTGB-202 ANN E T1.7-47 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By__ ________________Title________________
B-327 CONTROL WIRING DIAGRAM (Cont'd) 1557S R5 9/10/82 E ISOL CAB/ALC-3 INTERCONN DIAGRAM RTGB-205 ANN M 1558S R7 9/2/82 E ISOL CAB/ALC-3 INTERCONN DIAGRAM RTGB-205 ANN N 1559S R7 5/23/83 E ISOL CAB/ALC-3 INTERCONN DIAGRAM RTGB-206 ANN S SH 1 1560S R4 6/9/81 E ISOL CAB/ALC-3 INTERCONN DIAGRAM RTGB-206 ANN S SH2 1561S R5 6/9/81 E ISOL CAB/ALC-2, 3 INTERCONN DIAGRAM RTGB-205, 206, ANN B, SH 3 1563S R3 3/6/81 E ISOL CAB/ALC-3 INTERCONN DIAGRAM RTGB-206 ANN R SH 2 1564S R4 10/21/82 E ISOL CAB/ALC-3 INTERCONN DIAGRAM RTGB-206 ANN R SH 3 1565S R5 10/14/82 E ISOL CAB/ALC-3 INTERCONN DIAGRAM RTGB-206 ANN Q SH 1 1566S R4 8/5/82 E ISOL CAB/ALC-3 INTERCONN DIAGRAM RTGB-206 ANN Q SH 2 1567S R6 2/23/83 E ISOL CAB/ALC-2, 3 INTERCONN DIAGRAM RTGB-204, 206, ANN K, Q, SH 3 1568S R7 12/15/82 E ISOL CAB/ALC-3 INTERCONN DIAGRAM RTGB-206 ANN P SH 1 1569S R4 10/11/82 E ISOL CAB/ALC-3 INTERCONN DIAGRAM RTGB-206 ANN P SH 2 1570S R4 9/9/82 E ISOL CAB/ALC-3 INTERCONN DIAGRAM RTGB-206 ANN P SH 3 1571S R5 3/6/81 E ISOL CAB/ALC-1 INTERCONN DIAGRAM HVCB ANN "T" 1572S R4 3/6/81 E ISOL CAB/ALC-1 INTERCONN DIAGRAM HVCB ANN "U" T1.7-48 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By__ ________________Title________________
B-327 CONTROL WIRING DIAGRAM (Cont'd) 1573S R6 12/20/82 E ISOL CAB/ALC-1 INTERCONN DIAGRAM HVCB ANN "V" SH 1 1574S R5 1/24/83 E ISOL CAB/ALC-1 INTERCONN DIAGRAM HVCB ANN "V" SH 2 1575S R4 3/6/81 E ISOL CAB/ALC-1 INTERCONN DIAGRAM HVCB ANN "W" 1576S R7 2/22/83 E ISOL CAB/ALC-1 INTERCONN DIAGRAM HVCB ANN "X" 1577S R5 8/30/82 E ISOL CAB/SEQ-OF EVENTS CAB INTERCONN DIAGRAM SH 1 1578S R4 6/14/82 E ISOL CAB/SEQ OF EVENTS CAB INTERCONN DIAGRAM SH 2 1580S R1 7/16/82 E ESC/ISOL CAB/ALC-3 INTER-WIRING 1583S R5 4/6/82 E BYPASS INDICATION SYSTEM A SH.2 1584S R5 4/6/82 E BYPASS INDICATION SYSTEM A SH.3 1587S R4 5/22/81 E BYPASS INDICATION SYSTEM B SH.2 1588S R4 4/6/82 E BYPASS INDICATION SYSTEM B SH.3 EMERGENCY DIESEL GENERATORS 1601S R6 6/24/83 E DIESEL GEN. 2A START CKT'S SH. 2 1602S R5 6/24/83 E DIESEL GEN. 2A START CKT'S SH. 3 1603S R1 1/28/80 E DIESEL GEN. 2A START CKT'S SH. 4 1604S R2 10/11/82 E DIESEL GEN. 2A START CKT'S SH. 5 T1.7-49 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared
__No.__ _No._ No Date ___By__ ________________Title________________
B-327 CONTROL WIRING DIAGRAM (Cont'd) 1605S R2 9/17/82 E DIESEL GEN. 2A START CRT'S SH. 6 1606S R5 10/21/82 E DIESEL GEN 2A GROUNDING
& METERING 1607S R4 1/11/82 E DIESEL GEN 2A IMMERSION HEATERS 1608S R6 3/11/83 E DIESEL GEN 2A VOLTAGE 1609S R0 7/31/81 E REGULATOR 1611S R6 6/24/83 E DIESEL GEN. 2B START CSTS SH. 2 1612S R5 6/24/83 E DIESEL GEN. 2B START CKTS SH. 3 1613S R1 1/28/80 E DIESEL GEN 2B START CKTS SH. 4 1614S R2 1/13/82 E DIESEL GEN 2B START CKTS SH. 5 1615S R3 9/17/82 E DIESEL GEN 2B START CKTS SH. 6 1616S R6 10/21/82 E DIESEL GEN 2B GROUNDING
& METERING 1617S R4 1/13/82 E DIESEL GEN 2B IMMERSION HEATERS 1618S R7 3/11/83 E DIESEL GEN 2B VOLTAGE 1619S R0 7/31/81 E REGULATOR 1621S R3 1/17/83 E ATMOS STM DUMP ISOL VA MV-08-15 1622S R4 1/17/83 E ATMOS STM DUMP ISOL VA MV-08-14 T1.7-50 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared No. No. No. Date By Title B-327 CONTROL WIRING DIAGRAM (Cont'd) 1623S R3 1/17/83 E ATMOS STM DUMP ISOL VA MV-08-17 1624S R3 1/17/83 E ATMOS STM DUMP ISOL VA MV-08-18 1625S R3 1/10/83 E STM GEN 2A ATMOS STM DUMP VA MV-08-19A 1626S R3 1/10/83 E STM GEN 2A ATMOS STM DUMP VA MV-08-18A 1627S R4 1/10/83 E STM GEN 2A ATMOS STM DUMP VA MV-08-19B 1628S R3 1/10/83 E STM GEN 2A ATMOS STM DUMP VA MV-08-18B 1629 R2 1/10/83 E RELIEF VALVE V-1474 1630 R2 1/10/83 E RELIEF VALVE V-1475 1631 R5 7/18/83 E AFWP-2A DISCH TO SG-2A I-SE-09-2 1632 R6 7/18/83 E AFWP-2B DISCH TO SG-2B I-SE-09-3 1633 R6 7/18/83 E AFWP-2C DISCH TO SG-2A I-SE-09-4 1634 R5 7/18/83 E AUX FW PUMP 2C DISCH TO STEAM GEN 2B I-SE-09-5 1635 R4 2/12/83 E FEEDWATER, HEADER PRESS 9B-9C-9D-10A-10B-10C-10D 1636 R3 12/20/82 E STEAM GEN 2A & 2B LEVEL/
PRESSURE 1637 R2 7/18/83 E REMOTE MANUAL INITIATE AFAS-1, AFAS-2 1638 R3 2/18/83 E AFAS ANNUNCIATORS SH 1 1639 R3 7/18/83 E AFAS ANNUNCIATORS SH 2 T1.7-51 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared No. No. No. Date By Title B-327 CONTROL WIRING DIAGRAM (Cont'd) 1641S R4 11/21/83 E RADIATION MONITORING 120V AC DISTRIBUTION 1642 R5 1/10/83 E 120V AC DISTRIBUTION SH 3 1643 R9 3/11/83 E 120V AC DISTRIBUTION SH 4 1648 R1 2/26/82 E LOOP NO. 2 SH 1 1649 R1 2/26/82 E LOOP NO. 2 SH 2 1650 R1 1/29/82 E LOOP NO. 2 SH 3 1653 R2 11/24/82 E LOOP NO. 3 SH 1 1654 R2 11/15/83 E LOOP NO. 3 SH 2 1655 R2 7/18/83 E LOOP NO. 3 SH 3 1656 R2 11/24/82 E LOOP NO. 3 SH 4 1657 R2 11/24/82 E LOOP NO. 3 SH 5 1658 R1 2/26/82 E LOOP NO. 4 SH 1 1659 R1 2/26/82 E LOOP NO. 4 SH 2 1668 R2 3/25/83 E RAD MONITORING LOOP 3 SH 6 1691S R3 2/14/83 E REACTOR COOLANT VENT SYSTEM-1 1692 R5 12/14/82 E REACTOR COOLANT VENT SYSTEM-2 1694 R2 12/14/82 E PLANT AUX CONTROL BOARD-2 120V AC & 125V DC DISTRIBUTION 1695 R1 3/12/82 E PLANT AUX CONTROL BOARD 45V DC DISTRIBUTION 1701S R3 2/1/83 E 480V SWGR 2A-5 MOT REL & HTRS 1702S R1 8/14/82 E 480V SWGR 2A-5 FEEDER 1703S R1 3/26/82 E REACTOR AREA MCC-2A6 T1.7-52 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared No. No. No. Date By Title B-327 CONTROL WIRING DIAGRAM (Cont'd) 1711 R4 2/1/83 E 480V SWG 2B5 METERING RELAYS
& HEATERS 1712 R2 8/15/82 E 480V SWGR 2B5 FDR 1713 R1 4/8/82 E 480V SWGR 2B5 FEEDWATER TO REACTOR AREA MCC-2B6 1751 R4 8/17/83 E QSPDS INTERCONNECTION 1755 R3 1/24/83 E REACTOR COOLANT TEMP. SAS-QSPDS INPUTS SH 1 1756 R4 3/11/83 E REACTOR COOLANT TEMP. SAS-QSPDS INPUTS SH 2 1757 R1 8/13/82 E PRESSURIZED PRESSURE ICC-INPUTS 1810 R2 8/17/83 E PRESSURIZER PRESSURE ICC-INPUTS 1829 R4 5/12/83 E PASS VALVES SH 4 1831 R3 4/12/83 E PSB-1 UNDERVOLTAGE 4160V BUS 2A3 1833 R3 4/12/83 E PSB-1 UNDERVOLTAGE 480V BUS 2A2/2A5 1834 R3 4/12/83 E PSB-1 UNDERVOLTAGE PROTECTION 480V BUS 2B2/2B5 1836 R3 4/11/83 E PSB-1 UNDERVOLTAGE PROTECTION BUS 2A3/2A2/2A5 RELAY 1837 R3 4/11/83 E PSB-1 UNDERVOLTAGE PROTECTION BUS 2B3/2B2/2B3 RELAY 1851 R2 11/30/82 E INCORE MONITOR DETECTORS L18, L20, R16, R18 1852 R2 11/30/82 E INCORE MONITOR DETECTORS C18, E13, E16, E18 1853 R2 11/30/82 E INCORE MONITOR DETECTORS C6, C13, E2, G4 T1.7-53 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared No. No. No. Date By Title B-327 CONTROL WIRING DIAGRAM (Cont'd) 1854 R2 11/30/82 E INCORE MONITOR DETECTORS G6, L4, L6, R2 1855 R2 11/30/82 E INCORE MONITOR DETECTORS R9, R15, T4, T6 1856 R2 11/30/82 E INCORE MONITOR DETECTORS T9, W4, W9, W13 1857 R2 11/30/82 E INCORE MONITOR DETECTORS R20, T20, Y8, Y14 1858 R2 11/30/82 E INCORE MONITOR DETECTORS A8, C4, E4, G2 1859 R2 11/30/82 E INCORE MONITOR DETECTORS A14, C9, E6, E9 1860 R2 11/30/82 E INCORE MONITOR DETECTORS C16, E20, G9, G13 1861 R2 11/30/82 E INCORE MONITOR DETECTORS G16, G18, L13, L16 1862 R2 11/30/82 E INCORE MONITOR DETECTORS G20, T18, W16, W18 1863 R2 11/30/82 E INCORE MONITOR DETECTORS L9, T13, T16, W6 1864 R2 11/30/82 E INCORE MONITOR DETECTORS L2, R4, R6, T2 1865 R3 3/11/83 E HEATER JUNCTION THERMOCOUPLES 1A, 2A, 3A, 4A 1866 R3 3/11/83 E HEATER JUNCTION THERMOCOUPLES 5A, 6A, 7A, 8A 1867 R3 3/11/83 E HEATER JUNCTION THERMOCOUPLES 1B, 2B, 3B, 4B 1868 R3 3/11/83 E HEATER JUNCTION THERMOCOUPLES 5B, 6B, 7B, 8B T1.7-54 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-1 (Cont'd)
Drawing Sheet Revision Prepared No. No. No. Date By Title B-327 CONTROL WIRING DIAGRAM (Cont'd)
-1692S R5 2/14/83 E REACTOR COOLANT VENT SYSTEM-2 G-878 R7 5/19/83 E HVAC - CONTROL DIAGRAMS SH. 1 G-879 2 R7 5/19/83 E HVAC - CONTROL DIAGRAMS SH. 2 G-879 3 R9 5/19/83 E HVAC - CONTROL DIAGRAMS SH. 3 T1.7-55 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-2 NSSS SUPPLIED ELECTRICAL INSTRUMENTATION AND CONTROL DRAWINGS SAFETY RELATED Drawing Sheet Revision Prepared No. No. No. Date By Title B-13172-412-330 1-9 R2 7/26/77 CE ELEMENTARY W/D REAC TRIP CKT BKR E-13172-413-130 R2 7/20/77 CE REACTOR TRIP SWITCHGEAR ARRANGEMENT E-13172-411-022 R3 10/28/82 CE NUCLEAR INSTRUMENTATION AND RPS CABINET ASSY E-13172411012 1 R6 10/27/82 CE RPS TERMINAL BLOCK WIRING DIAGRAM E-13172411012 2 R6 10/28/82 CE RPS TERMINAL BLOCK WIRING DIAGRAM SH. 2 E-13172411012 3 R6 10/28/82 CE RPS TERMINAL BLOCK WIRING DIAGRAM SH. 3 E-13172411012 4 R6 10/28/82 CE RPS TERMINAL BLOCK WIRING DIAGRAM SH. 4 805B1 R0 8/18/75 CE RTSG HEATER ELEMENTARY 805B13 R2 12/3/75 CE GE AK-2-25 CIRCUIT BREAKER ELEMENTARY & CONN DIAG 805B12 R1 3/24/76 CE RTSG DC ELEMENTARY TCB-9 8055-B10 R2 3/24/76 CE RTSG DC ELEMENTARY TCB-7 805B11 R2 3/24/76 CE RTSG DC ELEMENTARY TCB-8 805B8 R2 3/24/76 CE RTSG DC ELEMENTARY TCB-5 805B9 R2 3/24/76 CE RTSG DC ELEMENTARY TCB-6 805B6 R2 3/24/76 CE RTSG DC ELEMENTARY TCB-3 805B7 R2 3/24/76 CE RTSG DC ELEMENTARY TCB-4 805B5 R2 3/24/76 CE RTSG DC ELEMENTARYTCB-2 805B4 R2 3/24/76 CE RTSG DC ELEMENTARY TCB-1 805B3 R0 8/19/75 CE RTSG CURRENT MONITOR ELEMENTARY T1.7-56 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-2 (Contd)
Drawing Sheet Revision Prepared No. No. No. Date By Title 805B2 R1 9/26/75 CE RTSG CURRENT MONITOR ELEMENTARY 805E6 R1 3/24/76 CE RTSG WIRING DIAGRAM SECTION 04 805E7 R1 3/24/76 CE RTSG WIRING DIAGRAM SECTION 05 805E5 R1 3/24/76 CE RTSG WIRING DIAGRAM SECTION 03 805E4 R1 3/24/76 CE RTSG WIRING DIAGRAM SECTION 02 805E3 R1 3/23/76 CE RTSG WIRING DIAGRAM SECTION 01 805E2 R1 10/1/75 CE RTSG ARRANGEMENT &
DETAILS 805E1 R1 10/1/75 CE RTSG ARRANGEMENT &
DETAILS E-13172-411-071 R3 7/26/82 CE CORE PROTECT CALCULATOR NO. 1 SCHEMATIC E-13172-411-072 1 R3 1/28/83 CE CORE PROTECT CALCULATOR No. 2 SCHEMATIC E-13172-411-086 1 R2 1/28/83 CE RPS ISOLATION LOGIC &
WIRING DIAGRAM E-13172-411-013 4 R2 7/28/82 CE RPS SCHEMATIC SH4 of 4 E-13172-411-018 R2 7/28/82 CE TRIP INHIBIT MODULE WIRING DIAGRAM E-13172-411-400 R1 3/14/79 CE RPS CALIB.& INDIC. PNL.
SCHEMATIC E-13172-411-040 R3 1/28/83 CE REACTOR TRIP BYPASS SCHE-MATIC E-13172-411-325 R1 3/14/79 CE RPS MISC. SCHEMATICS E-13172-411-013 1 R2 7/28/82 CE RPS MISC. SCHEMATICS SH1 OF 4 T1.7-57 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-2 (Contd)
Drawing Sheet Revision Prepared No. No. No. Date By Title E-13172-411-013 3 R2 7/28/82 CE RPS MISC. SCHEMATICS SH3 OF 4 E-13172-411-013 2 R2 7/28/82 CE RPS MISC. SCHEMATICS SH2 OF 4 E-13172-411-401 R1 3/14/79 CE RPS CALIB & IND. PNL ASSEMBLY E-13172-411-011 R2 7/28/82 CE RPS BIN ASSEMBLIES WIRING DIAG.
E-13172-411-324 R1 3/14/79 CE AUX. LOGIC WIRING DIAGRAM E-13172-411-072 2 R2 1/28/83 CE CORE PROT. CALC NO.2 SCHEMATIC-SH2 OF 2 E-13172-411-024 R2 10/28/82 CE RPS BIN ASSEMBLY E-13172-411-015 R0 12/23/77 CE TRAC 1 WIRING & ASSEMBLY DIAGRAM E-13172-411-029 R2 1/28/83 CE TRIP TEST CABLE PNL ASSEMB.
D-13172-411-366 R2 3/14/79 CE TRIP UNIT BIN ASSEMB.
PERSPECT.
E-13172-411-085 R2 1/28/82 CE POWER RATIO SIGNAL CALC.
SCHEM.
E-13172-411-310 R2 3/14/79 CE AUX. LOGIC SCHEMATIC E-13172-411-025 R2 10/28/82 CE TRIP INHIBIT MODULE ASSY E-13172-411-034 1 R3 1/28/83 CE RPS TRIP STATUS PNL. SCHEM & W/D E-13172-411-302 R4 7/28/82 CE TRIP UNIT INTERCONN. MODULE W/D E-13172-411-021 R2 7/28/82 CE NUC. INST. RPS CAB. ASSY.
FRNT PNL LAYOUT E-13172-411-039 R2 1/28/83 CE SCHEM. INPUT SIG. CONN.TO TRIP UNITS E-13172-411-350 R1 3/14/79 CE AUX. LOGIC ASSEMBLY T1.7-58 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-2 (Contd)
Drawing Sheet Revision Prepared No. No. No. Date By Title E-13172-411-003 R2 7/28/82 CE RPS FUNCTIONAL DIAGRAM E-13172-411-033 R2 1/28/83 CE RPS TRIP STATUS PNL. ASSY E-13172-411-043 R1 12/23/77 CE LOW FLOW PROT. SYS. FUNCT.
DIAG.
E-13172-411-376 R2 3/14/79 CE TRIP UNIT INTERCONN.
MODULE ASSY D-13172-411-091 R1 3/14/79 CE BISTABLE TRIP UNIT MODULE ASSEMBLY D-13172-411-035 R2 1/28/83 CE RPS/NI INTERFACE D-13172-411-092 R1 3/14/79 CE AUXILIARY TRIP UNIT MODULE ASSEMBLY E-13172-411-034 2 R2 1/28/83 CE RPS TRIP STATUS PANEL SCHEM & WIRING DIAGRAM E-13172-411-086 2 R2 1/28/83 CE RPS ISOLATION LOGIC &
WIRING DIAGRAM AW 20 D-13172-413-412 R5 11/29/82 CE STEAM GENERATOR-B PROTEC-TIVE CHANNEL BLOCK DIAGRAM D-13172-413-411 R4 1/29/82 CE STEAM GENERATOR-A PROTEC-TIVE CHANNEL BLOCK DIAGRAM D-13172-416-214 R5 3/18/83 CE INTERC/D CHGNG PMP DISCH HDR PRES CHAN P2212 D-13172-416-121 3 R5 5/02/83 CE INTERCONN DIAG PRESS LEVEL CHANNEL L-1110 SH3 D-13172-416-121 2 R7 5/02/83 CE INTERCONN DIAG PRESS LEVEL CHANNEL L-1110 SH2 D-13172-416-121 1 R6 4/06/83 CE INTERCONN DIAG PRESS LEVEL CHANNEL L-1110 SH1 D-13172-416-112 R4 3/18/83 CE INTERC/D-PRESSURIZER PRES-SURE CHANNEL P1102 D-13172-416-131 R5 4/06/83 CE I/D-REAC COOL DELTA PRES FLOW CHANS P1101A-D T1.7-59 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-2 (Contd)
Drawing Sheet Revision Prepared No. No. No. Date By Title D-13172-416-217 R2 11/21/80 CE I/D-CHARGING PUMP SUCT PRESS CHANNEL P-2224 D-13172-416-311 1 R2 4/06/83 CE ID-HPSI,.LPSI HEADER PRESS CHANNELS P-3308,9 D-13172-416-311 2 R3 4/06/83 CE ID-HPSI,.LPSI HDR PRESS CHANNELS P3304-7 D-13172-416-103 1 R3 6/25/82 CE ID-RCS LOOP TEMP CHANNELS T1111 & 1121 SH1OF2 D-13172-416-103 2 R4 4/06/83 CE ID-RCS LOOP TEMP CHANNELS T1115 & 1125 SH2OF2 D-13172-416-104 R3 4/06/83 CE INTERCONN DIAG-RCS LOOP TEMP CHS T1112&1122 D-13172-416-113 1 R5 4/06/83 CE ID-PRESSURIZER PRESS LO RNGE CH P1103,5 SH1/2 877 D-13172-416-113 2 R5 4/06/83 CE ID-PRESSURIZER PRESS LO RNGE CH P1103,5 SH2/2 D-13172-416-115 R1 2/26/80 CE INTERCONN DIAG-RCP SEAL PRE SS CHANNELS D-13172-416-132 R1 2/26/80 CE ID-RCP CONT BLEEDOFF FLO CHS F1150,60,70,80 D-13172-416-401 R4 4/06/83 CE ID-STEAM GENERATOR STEAM PRESS CHANNEL P8013 D-13172-416-402 R4 4/06/83 CE INTERCONN DIAG-STEAM GENERATOR LEVEL CH L9013 877 D-13172-416-651 R3 11/24/81 CE I/D WMS MISC LOCAL &
REMOTE ALARMS 8770-1938 D-13172-416-470 1 R3 1/29/82 CE I/D-CONT BRD MTD NUCLEAR INST SH1 OF 4 D-13172-416-470 4 R2 7/12/81 CE I/D CONT BRD MTD NUCLEAR INST SH 4 OF 4 D-13172-416-331 R2 4/06/83 CE ID HI & LO PRES SI FLOW CHANNELS F3311,F3312 T1.7-60 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-2 (Contd)
Drawing Sheet Revision Prepared No. No. No. Date By Title D-13172-416-470 2 R4 9/27/82 CE ID CONTR BOARD MNTD NUC INSTR SH2 8770-1511-13 D-13172-416-470 3 R2 1/19/81 CE INTERCONN DIAG CONTROL BOARD MNTD INSTR SH 3 D-13172-416-105 R4 10/12/82 CE INTERCONN DIAG RCS TEMP CHANNEL T-1102 D-13172-416-222 R2 7/10/81 CE ID-BA TANKS 2A & 2B LEVEL CHS L-2206,7,8,9 T1.7-61 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-3 ARCHITECT/ENGINEER SUPPLIED FLOW DIAGRAMS, PIPING AND INSTRUMENTATION DIAGRAMS SAFETY RELATED Drawing Sheet Revision Prepared No. No. No. Date By Title G 079 1 R11 6/15/83 E FLOW DIAG - MAIN, EXTRACTION AUXILIARY STEAM & AIR EVACUATION SYSTEMS G 079 2 R10 4/12/83 E FLOW DIAG - MAIN, EXTRACTION AUXILIARY STEAM & AIR EVACUATION SYSTEMS G 080 1 R9 6/15/83 E FLOW DIAG - FDWTR & CONDENSATE SYSTEMS G 080 2 R9 6/15/83 E FLOW DIAG - FDWTR & CONDENSATE SYSTEMS G 081 1 R9 4/12/83 E FLOW DIAG - HEATER DRAIN &
VENT SYSTEMS G 081 2 R9 4/12/83 E FLOW DIAG - HEATER DRAIN &
VENT SYSTEMS G 082 R11 6/15/83 E FLOW DIAG - CRLG & INTAKE COOLING WATER SYSTEMS G 083 R11 12/27/82 E FLOW DIAG - COMPONENT COOLANT SYSTEM G 084 R11 6/15/83 E FLOW DIAG - FIREWATER DOMESTIC AND MAKEUP SYSTEMS G 085 1 R11 6/15/83 E FLOW DIAGRAM - SERVICE INSTRUMENT AIR SYSTEM G 085 2 R10 6/15/83 E FLOW DIAGRAM - INSTRUMENT INSTRUMENT AIR SYSTEM G 086 R11 6/15/83 E FLOW DIAG - MISCELLANEOUS SYSTEMS SH-1 G 087 R11 6/15/83 E FLOW DIAG - MISCELLANEOUS SYSTEMS SH-2 G 088 R11 6/15/83 E FLOW DIAG - CONTAINMENT SPRAY
& REFUELING WATER SYSTEMS T1.7-62 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-3 (Cont'd)
Drawing Sheet Revision Prepared No. No. No. Date By Title G 089 R9 10/4/82 E FLOW DIAG - TURBINE COOLING WATER SYSTEM G 090 R7 10/11/82 E REACTOR COOLANT - PRESSURE BOUNDARY DIAGRAM G 091 R9 6/15/83 E FLOW DIAG MISC SYSTEMS G 092 R7 6/15/83 E FLOW DIAG MISC SAMPLING SYSTEMS G 862 R6 11/19/82 E HVAC - AIR FLOW DIAGRAM G 863 R5 5/19/83 E HVAC - REFRIGERANT PIPING T1.7-63 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 Table 1.7-4 NSSS SUPPLIED FLOW DIAGRAMS, PIPING AND INSTRUMENTATION DIAGRAMS SAFETY RELATED Drawing Revision Prepared No. No. Date By Title E-13172-310-100 13 7/1/82 CE PIPING & INSTRUMENTATION DIAGRAM SYMBOLS E-13172-310-110 17 5/17/83 CE REACTOR COOLANT SYSTEM P&I DIAGRAM E-13172-310-111 17 7/7/83 CE REACTOR COOLANT PUMP P&I DIAGRAM E-13172-310-120 17 3/16/83 CE CHEMICAL & VOLUME CONTROL SYSTEM P&I DIAG E-13172-310-121 18 5/17/83 CE CHEMICAL & VOLUME CONTROL SYSTEM P&I DIAG E-13172-310-130 19 5/17/83 CE SAFETY INJECTION SYS P&I DIAGRAM E-13172-310-131 17 5/17/83 CE SAFETY INJECTION SYS P&I DIAGRAM E-13172-310-140 19 7/7/83 CE FUEL POOL SYS P&I DIAGRAM E-13172-310-150 18 7/7/83 CE SAMPLING SYSTEM P&I DIAGRAM E-13172-310-160 19 5/17/83 CE WASTE MANAGEMENT SYS P&I DIAGRAM E-13172-310-161 19 7/7/83 CE WASTE MANAGEMENT SYS P&I DIAGRAM E-13172-310-162 19 7/7/83 CE WASTE MANAGEMENT SYS P&I DIAGRAM E-13172-310-163 20 7/7/83 CE WASTE MANAGEMENT SYS P&I DIAGRAM (SHEET 4)
E-13172-310-164 16 5/17/83 CE WASTE MANAGEMENT SYS P&I DIAGRAM E-13172-310-105 17 7/7/83 CE AUXILIARY PUMPS P&I DIAGRAM E-13172-310-165 13 3/18/83 CE BORIC ACID CONCENTRATOR 2A P&ID T1.7-64 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.7-4 (Cont'd)
Drawing Revision Prepared No. No. Date By Title E-13172-310-166 13 3/16/83 CE BORIC ACID CONCENTRATOR B P&ID E-13172-310-167 13 3/16/83 CE RADIOACTIVE WASTE CON-CENTRATOR P&ID E-13172-310-168 11 11/22/82 CE WASTE MANAGEMENT SYS P&I DIAGRAM E-13172-310-109 13 7/7/83 CE REACTOR COOLANT SYS P&I DIAGRAM E-13172-310-122 13 7/7/83 CE CHEMICAL & VOLUME CONTROL SYS P&I DIAGRAM E-13172-310-107 06 3/16/83 CE REACTOR COOLANT SYSTEM P&I DIAGRAM E-13172-310-108 04 5/17/83 CE REACTOR COOLANT SYSTEM P&I DIAGRAM E-13172-310-153 09 7/7/83 CE SAMPLING SYSTEM P&I DIAGRAM E-13172-310-132 07 5/17/83 CE SAFETY INJECTION SYSTEM P&I DIAGRAM E-13172-310-152 08 7/7/83 CE SAMPLING SYSTEM P&I DIAGRAM E-13172-310-169 08 5/18/83 CE WASTE MANAGEMENT SYSTEM P&I DIAGRAM E-13172-310-171 04 9/23/82 CE WASTE MANAGEMENT SYSTEM P&I DIAGRAM E-13172-310-101 03 9/23/82 CE STEAM GENERATOR SUPPORT SNUBBER PIPING SYSTEM VALVE IDENTIFICATION E-13172-310-145 03 11/23/82 CE REFUELING EQUIPMENT VALVE IDENTIFICATION E-13172-310-115 06 7/7/83 CE R.C. PUMP SEAL INJECTION ADDITION P & I DIAGRAM T1.7-65 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 1.8 NRC REGULATORY GUIDES Information contained herein were valid at the time the Construction Permit for St. Lucie 2 was issued, and are being retained in the Updated FSAR for document completeness and historical record. No present or future update of this section is required.
Subject to the implementation dates therein, Regulatory Guides issued on or before May 2, 1977 (Construction Permit date for St. Lucie Unit 2) are considered to contain the recommendations that are applicable to the design of this plant. Table 1.8-1 is a listing of all such Regulatory Guides, with corresponding dates and revision numbers. Cross-references are provided in Table 1.8-1 for those regulatory guide subjects discussed in particular subsections.
In specific instances, later revisions to Regulatory Guides listed in Table 1.8-1 are addressed where following such guidance is deemed proper.
Other NRC staff requirements are discussed in Section 1.9.
1.8-1 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 TABLE 1.8-1 APPLICABLE NRC REGULATORY GUIDES Number Title Date Revision Discussion in Subsection(s) Remarks 1.1 Net Positive Suction Head for Emergency Core Cooling and 11/70 0 6.2.2.3.1 Containment Heat Removal System Pumps 6.3.4.1.1 1.2 Thermal Shock to Reactor Pressure Vessels 11/70 0 5.3.1 1.3 Assumptions Used for Evaluating the Potential Radiological 6/74 2 Not Applicable Consequences of a Loss of Coolant Accident for Boiling Water Reactors.
1.4 Assumptions Used for Evaluating the Potential Radiological 6/74 2 2.3.4 Consequences of a Loss of Coolant Accident for Pressurized Water Reactors 1.5 Assumptions Used for Evaluating the Potential Radiological 3/71 0 Not Applicable Consequences of a Steam Line Break Accident for Boiling Water Reactors 1.6 Independence Between Redundant Standby (Onsite) Power 3/71 0 8.3.1.2 Sources and Between Their Distribution Systems 1.7 Control of Combustible Gas Concentrations in Containment 11/78 2 6.2.5.3.2 Following a Loss of Coolant Accident 1.8 Personnel Selection and Training 5/77 1-R 12.5.1/12.5.3 13.1.3, 17.2 1.9 Selection of Diesel Generator Set Capacity for Standby 3/71 0 8.3.1.2 Power Supplies 1.10 Mechanical (Cadweld) Splices in Reinforcing Bars of 1/73 1 3.8.3.2 Category I Concrete Structures 1.11 Instrument Lines Penetrating Primary Reactor Containment 3/71 0 7.1.2.2 6.2.4 1.12 Instrumentation for Earthquakes 4/74 1 3.7.4 1.13 Spent Fuel Storage Facility Design Basis 12/75 1 9.1.1.3/9.1.2.3 9.1.3.3 1.14 Reactor Coolant Pump Flywheel Integrity 10/71 0 5.4.1.1 Regulatory Position C.4.6 of Revision 1 (8/75) is applicable to in-service inspections con ducted on all plants after January 1, 1976.
T1.8-1 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 TABLE 1.8-1 (Contd)
Number Title Date Revision Discussion in Subsection(s) Remarks 1.15 Testing of Reinforcing Bars for Category I Concrete 12/72 1 3.8.3.2/3.8.3.6 Structures 1.16 Reporting of Operating Information-Appendix A Technical 8/75 4 12.5.3 Specifications 1.17 Protection of Nuclear Plants Against Industrial Sabotage 6/73 1 13.6 A proprietary St Lucie Plant Security Plan is submitted under separate cover.
1.18 Structural Acceptance Test for Concrete Primary Reactor 12/72 1 Not Applicable Containments 1.19 Nondestructive Examination of Primary Containment Liner 8/72 1 Not Applicable Welds 1.20 Comprehensive Vibration Assessment Program for Reactor 5/76 2 3.9.2.4 Internals During Preoperational and Initial Startup Testing 1.21 Measuring, Evaluating, and Reporting Radioactivity in Solid 6/74 1 11.5.1.2 Wastes and Release of Radioactivity in Liquid and Gaseous 12.3.4 Effluents from Light Water-Cooled Nuclear Power Plants 1.22 Periodic Testing of Protection System Actuation Functions 2/72 0 7.2.1.1.9/7.5.2.9 7.3.1.1.1/7.6.2 7.4.2.2 1.23 Onsite Meteorological Programs 2/72 0 2.3.3 1.24 Assumptions Used for Evaluating the Potential Radiological 3/72 0 15.7.1 Consequences of a Pressurized Water Reactor Gas Storage Tank Failure 1.25 Assumptions Used for Evaluating the Potential Radiological 3/72 0 15.7.3 Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors 1.26 Quality Group Classifications and Standards for Water-Steam- 2/76 3 3.2.2 and Radio-Waste-Containing Components of Nuclear Power Plants 1.27 Ultimate Heat Sink for Nuclear Power Plants 1/76 2 9.2.5 T1.8-2 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 TABLE 1.8-1 (Contd)
Number Title Date Revision Discussion in Subsection(s) Remarks 1.28 Quality Assurance Program Requirements (Design and 6/72 0 Not Applicable This regulatory guide is applicable during the Construction) design and construction phases of nuclear power plants and as such is discussed in PSARs, not FSARs.
1.29 Seismic Design Classification 9/78 3 3.2.1 1.30 Quality Assurance Requirements for the Installation, - - 17.2 The revision and date of this document endorsed Inspection, and Testing of Instrumentation and Electric for St. Lucie Unit 2 is governed by the latest Equipment revision of the FP&L Topical Quality Assurance Report as referenced in Section 17.2 of the FSAR.
1.31 Control of Ferrite Content in Stainless Steel Weld Metal 5/77 2&3 5.2.3.4.2 Subsections 6.1.1.1 and 10.3.6.2 address 6.1.1.1 Revision 1 (6/73) of this regulatory guide also.
10.3.6.2 1.32 Criteria for Safety-Related Electric Power Systems for Nuclear 8/72 0 8.3.1.2 Power Plants 1.33 Quality Assurance Program Requirements (Operations) - - 13.5.1, 17.2 The revision and date of this document endorsed for St. Lucie Unit 2 is governed by the latest revision of the FP&L Topical Quality Assurance Report as referenced in Section 17.2 of the FSAR.
1.34 Control of Electroslag Weld Properties 12/72 0 5.2.3.3.2 5.2.3.4 1.35 Inservice Inspection of Ungrouted Tendons in Prestressed 1/76 2 Not Applicable Concrete Containment Structures 1.36 Nonmetallic Thermal Insulation for Austenitic Stainless Steel 2/73 0 5.2.3.2 6.1.1.1 1.37 Quality Assurance Requirements for Cleaning of Fluid Systems - - 6.1.1.1, 17.2 The revision and date of this document endorsed and Associated Components of Water-Cooled Nuclear Power for St. Lucie Unit 2 is governed by the latest Plants revision of the FP&L Topical Quality Assurance Report as referenced in Section 17.2 of the FSAR.
1.38 Quality Assurance Requirements for Packaging, Shipping, - - 17.2 The revision and date of this document endorsed Receiving, Storage, and Handling of Items for Water-Cooled for St. Lucie Unit 2 is governed by the latest Nuclear Power Plants revision of the FP&L Topical Quality Assurance Report as referenced in Section 17.2 of the FSAR.
T1.8-3 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 TABLE 1.8-1 (Contd)
Number Title Date Revision Discussion in Subsection(s) Remarks 1.39 Housekeeping Requirements for Water-Cooled Nuclear Power - - 17.2 The revision and date of this document endorsed Plants for St. Lucie Unit 2 is governed by the latest revision of the FP&L Topical Quality Assurance Report as referenced in Section 17.2 of the FSAR.
1.40 Qualification Tests of Continuous-Duty Motors Installed Inside 3/73 0 3.11 the Containment of Water-Cooled Nuclear Power Plants 1.41 Preoperational Testing of Redundant Onsite Electric Power 3/73 0 8.3.1.2 Systems to Verify Proper Load Group Assignments 1.42 Withdrawn 3/76 1.43 Control Stainless Steel Weld Cladding of Low-Alloy Steel 5/73 0 5.2.3.3.2 Components 1.44 Control of the Use of Sensitized Stainless Steel 5/73 0 5.2.3.4.1 6.1.1.1 10.3.6.2 1.45 Reactor Coolant Pressure Boundary Leakage Detection 5/73 0 5.2.5 Systems 1.46 Protection Against Pipe Whip Inside Containment 5/73 0 3.6.1.1/3.6.2.3.2 3.6.2.1.1 1.47 Bypassed and Inoperable Status Indication for Nuclear Power 5/73 0 7.5.2.7 Plant Safety Systems 1.48 Design Limits and Loading Combinations for Seismic Category 5/73 0 3.9.1.4 I Fluid System Components 3.9.3.1.1 1.49 Power Levels of Nuclear Power Plants 12/73 1 6.2, 15.0 The guidance provided in this regulatory guide is utilized in accident analyses performed.
1.50 Control of Preheat Temperature for Welding of Low-Alloy Steel 5/73 0 5.2.3.3.2 10.3.6.2 1.51 Withdrawn 7/75 1.52 Design, Testing, and Maintenance Criteria for Post-accident 3/78 2 6.5.1 Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Absorption Units of Light-Water-Cooled Nuclear Power Plants T1.8-4 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 TABLE 1.8-1 (Contd)
Number Title Date Revision Discussion in Subsection(s) Remarks 1.53 Application of the Single-Failure Criterion to Nuclear Power 6/73 0 7.1.2.2 Plant Protection Systems 7.2.1.2 1.54 Quality Assurance Requirements for Protective Coatings 6/73 0 6.1.2 Applied to Water-Cooled Nuclear Power Plants 1.55 Concrete Placement in Category I Structures 6/73 0 3.8.3.2/3.8.3.6 1.56 Maintenance of Water Purity in Boiling Water Reactors 6/73 0 Not Applicable 1.57 Design Limits and Loading Combinations for Metal Primary 6/73 0 3.8.2.3 Reactor Containment System Components 1.58 Qualification of Nuclear Power Plant Inspection, - - 13.1, 17.2 The revision and date of this document endorsed for St. Lucie Unit 2 is governed by the latest revision of the FP&L Topical Quality Assurance Report as referenced in Section 17.2 of the FSAR.
1.59 Design Basis Flood for Nuclear Power Plants 4/76 1 2.4.2.2/2.4.3 3.4.1 1.60 Design Response Spectra for Seismic Design of Nuclear 12/73 1 3.7.1.1 Power Plants 1.61 Damping Values for Seismic Design of Nuclear Power Plants 10/73 0 3.7.1.3 1.62 Manual Initiation of Protective Actions 10/73 0 7.1.2.2, 8.3.1.2 1.63 Electric Penetration Assemblies in Containment Structures for 10/73 0 8.3.1.2 Light-Water-Cooled Nuclear Power Plants 1.64 Quality Assurance Requirements for the Design of Nuclear - - 17.2 The revision and date of this document endorsed Power Plants for St. Lucie Unit 2 is governed by the latest revision of the FP&L Topical Quality Assurance Report as referenced in Section 17.2 of the FSAR.
1.65 Materials and Inspection for Reactor Vessel Closure Studs 10/73 0 5.3.1.7 1.66 Withdrawn 10/77 1.67 Installation of Overpressure Protective Devices 10/73 0 3.9.3.3 T1.8-5 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 TABLE 1.8-1 (Contd)
Number Title Date Revision Discussion in Subsection(s) Remarks 1.68 Initial Test Programs for Water-Cooled Nuclear Power Plants 1/77 1 Section 14.0 addressed Revision 0.
1.68.1 Preoperational and Initial Startup Testing of Feed-water and 1/77 1 Not Applicable Condensate Systems for Boiling Water Reactor Power Plants 1.68.2 Initial Startup Test Program to Demonstrate Remote Shutdown 7/78 1 Initially discussed in Section 14.0 Capability for Water-Cooled Nuclear Power Plants 1.69 Concrete Radiation Shields for Nuclear Power Plants 12/73 0 12.3.2.4 1.70 Standard Format and Content of Safety Analysis Reports for 9/75 2 Revision 3 (11/78) of this regulatory guide was Nuclear Power Plants-LWR Edition used insofar as to the extent practicable in developing the St Lucie Unit 2 Final Safety Analysis Report.
1.71 Welder Qualification for Areas of Limited Accessibility 12/73 0 5.2.3.3.2 10.3.6.2 1.72 Spray Pond Piping Made From Fiberglass-Reinforced 12/73 0 Not Applicable Thermosetting Resin 1.73 Qualification Tests of Electric Valve Operators Installed Inside 1/74 0 3.11 the Containment of Nuclear Power Plants 1.74 Quality Assurance Terms and Definitions - - 17.2 The revision and date of this document endorsed for St. Lucie Unit 2 is governed by the latest 1.75 Physical Independence of Electric Systems 1/75 1 7.1.2.2 revision of the FP&L Topical Quality Assurance 8.3.1.2 Report as referenced in Section 17.2 of the FSAR.
1.76 Design Basis Tornado for Nuclear Power Plants 4/74 0 The Design Basis Tornado for St Lucie Unit 2 is discussed in Subsection 3.3.2.
1.77 Assumptions Used for Evaluating a Control Rod Ejection 5/74 0 15.4.3 Accident for Pressurized Water Reactors 1.78 Assumptions for Evaluating the Habitability of a Nuclear Power 6/74 0 2.2.3.2 EC284401 Plant Control Room During a Postulated Hazardous Chemical Release 1.79 Preoperational Testing of Emergency Core Cooling Systems 9/75 1 6.3.4.1.1 for Pressurized Water Reactors 1.80 Preoperational Testing of Instrument Air Systems 6/74 0 14.2 T1.8-6 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 TABLE 1.8-1 (Contd)
Number Title Date Revision Discussion in Subsection(s) Remarks 1.81 Shared Emergency and Shutdown Electric Systems for 1/75 1 8.3.1.2 St Lucie Units 1 and 2 have separate and Multi-Unit Nuclear Power Plants independent onsite emergency and shutdown electric systems.
1.82 Sumps for Emergency Core Cooling and Containment Spray 6/74 0 6.2.2.2.3 Systems 1.83 Inservice Inspection of Pressurized Water Reactor Steam 7/75 1 5.4.2.2 RG 1.83 withdrawn November 2009. (Ref: NRC EC283094 Generator Tubes document NRC-2009-0488: ML13066A546) 1.84 Design and Fabrication Code Case Acceptability - ASME 3/77 9 3.9.3.1.1 Section III Division 1 1.85 Materials Code Case Acceptability - ASME Section III 3/77 9 - Materials acceptability is discussed in various Division 1 subsections of the FSAR which deal with this topic for various structures, systems, and components.
1.86 Termination of Operating Licenses for Nuclear Reactors 6/74 0 Not Applicable The regulatory guide is applicable when a licensee decides to terminate the nuclear reactor operating license.
1.87 Guidance for Construction of Class 1 Components in 6/74 0 Not Applicable Elevated-Temperature Reactors (Supplement to ASME Section III Code Classes 1592, 1593, 1594, 1595, and 1596) 1.88 Collection, Storage, and Maintenance of Nuclear Power Plant - - 17.2 The revision and date of this document endorsed Quality Assurance Records for St. Lucie Unit 2 is governed by the latest revision of the Quality Assurance FP&L Topical Report as referenced in Section 17.2 of the FSAR.
1.89 Qualification of Class IE Equipment for Nuclear Power Plants 11/74 0 8.3.1.2 3.11 1.90 In-service Inspection of Prestressed Concrete Containment 11/74 0 Not Applicable Structures With Grouted Tendons 1.92 Combining Modal Responses and Spatial Components in 12/74 0 3.7.2.6/3.7.2.7 Seismic Response Analysis 3.7.3.6/3.7.3.7 1.93 Availability of Electric Power Sources 12/74 0 8.3.1.2 The applicable recommendations of Regulatory Guide 1.93 are also a part of the Technical Specifications.
T1.8-7 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 TABLE 1.8-1 (Contd)
Number Title Date Revision Discussion in Subsection(s) Remarks 1.94 Quality Assurance Requirements for Installation, Inspection, 4/76 1 Not Applicable The revision and date of this document endorsed and Testing of Structural Concrete Structural Steel, Soils and for St. Lucie Unit 2 is governed by the latest Foundations During the Construction Phase of Nuclear Power revision of the FP&L Topical Quality Assurance Plants Report as referenced in Section 17.2 of the FSAR.
1.95 Protection of Nuclear Power Plant Control Room Operators 2/75 0 2.2.3.2 EC284401 Against an Accidental Chlorine Release 1.96 Design of Main Steam Isolation Valve Leakage Control 6/76 1 Not Applicable Systems for Boiling Water Reactor Nuclear Power Plants 1.98 Assumptions Used for Evaluating the Potential Radiological 3/76 0 Not Applicable Consequences of a Radioactive Offgas System Failure in a Boiling Water Reactor 1.99 Effects of Residual Elements on Predicted Radiation Damage 7/75 0 5.3.1.6.7 to Reactor Vessel Materials 1.101 Emergency Planning for Nuclear Power Plants 3/77 1 13.3 A St. Lucie Plant Emergency Plan is submitted under separate cover.
1.102 Flood Protection for Nuclear Power Plants 9/76 1 3.4.1 1.104 Withdrawn 8/79 1.107 Qualifications for Cement Grouting for Prestressing Tendons in 2/77 1 Not Applicable Containment Structures 1.108 Periodic Testing of Diesel Generator Units used as Onsite 8/77 1 8.3.1.2 This regulatory guide was issued after the CP Electric Power Systems at Nuclear Power Plants issuance date of May 2, 1977 on St. Lucie Unit 2.
1.109 Calculation of Annual Doses to Man from Routine Releases of 3/76 0 11.2.3/11.3.3 Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I 1.111 Methods for Estimating Atmospheric Transport and Dispersion 3/76 0 2.3.5 of Gaseous Effluents in Routine Releases from 11.2.3/11.3.3 Light-Water-Cooled Reactors 1.112 Calculation of Releases of Radioactive Materials in Gaseous 5/77 O-R 11.2.3/11.3.3 and Liquid Effluents from Light-Water-Cooled Power Reactors 1.113 Estimating Aquatic Dispersion of Effluents from Accidental and 4/77 1 11.2.3 Routine Reactor Releases for the Purpose of Implementing Appendix I T1.8-8 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 TABLE 1.8-1 (Contd)
Number Title Date Revision Discussion in Subsection(s) Remarks 1.114 Guidance on Being Operator at the Controls of a Nuclear 11/76 1 Operator training is discussed in Chapter 13.
Power Plant 1.116 Quality Assurance Requirements for Installation, Inspection, - - 17.2 The revision and date of this document endorsed and Testing of Mechanical Equipment and Systems for St Lucie Unit 2 is governed by the latest revision of the FP&L Topical Quality Assurance Report as referenced in Section 17.2 of the FSAR.
1.119 Withdrawn 6/77 1.121 Bases for Plugging Degraded PWR Steam Generator Tubes 8/76 0 Steam generator tube corrosion allowance is addressed in Subsection 5.4.4.2.
1.123 Quality Assurance Requirements for Control of Procurement of - - 17.2 The revision and date of this document endorsed Items and Services for Nuclear Power Plants for St. Lucie Unit 2 is governed by the latest revision of the FP&L Topical Quality Assurance Report as referenced in Section 17.2 of the FSAR.
1.126 An Acceptable Model and Related Statistical Methods for the 3/77 0 The subject of fuel density is discussed in Analysis of Fuel Densification Subsection 4.2.1.2.4.3.
1.127 Inspection of Water-Control Structures Associated with Nuclear 4/77 0 Not Applicable There are no water-control structures specifically Power Plants built for use in conjunction with this plant and whose failure could have radiological consequences adversely affecting the public health and safety.
1.129 Maintenance, Testing, and Replacement of Large Lead 4/77 0 The implementation section for this regulatory Storage Batteries for Nuclear Power Plants guide states that this regulatory guide is used in the evaluation of CP applicants docketed after December 1, 1977.
1.144 Auditing of Quality Assurance Programs for Nuclear Power - - 17.2 The revision and date of this document endorsed Plants for St. Lucie Unit 2 is governed by the latest revision of the FP&L Topical Quality Assurance Report as referenced in Section 17.2 of the FSAR.
1.146 Qualification of Quality Assurance Program Audit Personnel for - - 17.2 The revision and date of this document endorsed Nuclear Power Plants for St. Lucie Unit 2 is governed by the latest revision of the FP&L Topical Quality Assurance Report as referenced in Section 17.2 of the FSAR.
1.183 Alternative Radiological Source Terms for Evaluating Design 7/00 0 Provides guidance on the performance of AST Basis Accidents at Nuclear Power Reactors dose analyses for design basis accidents.
T1.8-9 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 1.9 OTHER CONCERNS AND COMMITMENTS 1.9.1 TMI ACTION PLAN Appendix 1.9A depicts those TMI Action Plan(1) requirements as described in NUREG-0737(2) for St. Lucie Unit 2. UFSAR Subsections discussing TMI "Lessons Learned" are delineated in Appendix 1.9A.
1.9.2 UNDERGROUND CABLE REVIEW Kerite insulated power and control cables have been reviewed and approved by the NRC for underground wet/dry environmental qualification.(3) 1.9.3 REPLACEMENT STEAM GENERATORS As a result of tube degradation, Florida Power & Light Company replaced the original steam generators (OSGs) with two replacement steam generators (RSGs) manufactured by AREVA.
Specific UFSAR text pertinent to the installation and operation of the RSGs was updated as necessary.
1.9-1 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 SECTION 1.9: REFERENCES
- 1. NUREG - 0660, May 1980 "NRC Action Plan Developed as a Result of the TMI-2 Accident.
- 2. NUREG - 0737, Letter dated October 31, 1980, D G Eisenhut (NRC) to all Licensees of Operating Plants and Applicants for Operating Licenses and Holders of Construction Permits, Subject "Post-TMI Requirements."
- 3. Letter dated January 31, 1978, K Kniel (NRC) to R E Uhrig (FP&L), "Use of Kerite Insulated Cable."
1.9-2 Amendment No. 24 (09/17)
UFSAR/St. Lucie - 2 APPENDIX 1.9A 1.9A TMI RELATED REQUIREMENTS The following item numbers correspond to those listed in NUREG-0737 "Clarification of TMI Action Plan Requirements" (October, 1980)(1) . NRC staff documented reviews and approval of these TMI related requirements are given by references to the Safety Evaluation Report(2) and its supplements.(3-6)
I.A.1.1 SHIFT TECHNICAL ADVISOR Florida Power & Light Co (FP&L) programs in response to this requirement have been developed for St. Lucie Unit 1 (Docket No. 50-335) and are also applicable to St. Lucie Unit 2.
I.A.1.2 SHIFT SUPERVISOR ADMINISTRATIVE DUTIES FP&L programs in response to this requirement have been developed for St. Lucie Unit 1 (Docket No. 50-335) and are also applicable to St. Lucie Unit 2.
I.A.1.3 SHIFT MANNING Procedures reflecting the requirements of NUREG-0737 and Generic Letter 82-16 in limiting overtime, hours of work and minimum shift complement have been generated for St. Lucie Unit 1 and also apply to St. Lucie Unit 2.
I.A.2.1 IMMEDIATE UPGRADING OF OPERATOR AND SENIOR OPERATOR TRAINING AND QUALIFICATIONS Unit Staff qualifications are delineated in the plant Technical Specifications Section 6.3.
I.A.2.3 ADMINISTRATION OF TRAINING PROGRAMS Training is covered by Section 6.4 of the plant Technical Specifications.
I.A.3.1 REVISE SCOPE AND CRITERIA FOR LICENSING EXAM FP&L initial and requalification training program revisions to address the increased scope of the license exams have been developed for St. Lucie 1 (Docket No. 50-335) and are also applicable to St. Lucie Unit 2.
I.B.1.2 EVALUATION OF ORGANIZATION AND MANAGEMENT The FP&L organization is provided in the FPL Quality Assurance Topical Report discussed in Section 17.2. The principal function of the Independent Safety Engineering Group as indicated by NUREG-0737 is assessment of operating experience. This function is the responsibility of the Engineering Manager and the Performance Improvement Manager.
I.C.1 SHORT TERM ACCIDENT ANALYSIS AND PROCEDURE REVISION The Combustion Engineering (CE) Owners' Group revised analysis and guidelines contained in CEN-152(7) were reviewed. Meetings were held with representatives of the CE Owners' Group in 1.9A-1 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 Bethesda, Maryland, on June 23, 24, and 29, 1982 to discuss NRC's preliminary comments on the analysis and guidelines. At a follow-up meeting in Bethesda on August 20, 1982, a revised CEN-152 was submitted which addressed a majority of the NRC staff concerns discussed at the June meetings. This revised document is now under review. Until the revised analysis and guidelines are approved, CEN-117 and CEN-128 are being used as interim technical bases for the St. Lucie Plant Unit No. 2 emergency operating procedures.
Based on their review of selected emergency operating procedures and their observation of these procedures being exercised on a simulator and in a control room walk-through, as described in Item I.C.8, NRC has concluded that the interim guidelines have been adequately incorporated into the procedures. Further revision to the procedures is expected to be necessary when the revised analysis and guidelines are approved. This satisfies the requirements of Item I.C.1, as per SER Supplement 4.(5)
I.C.2 SHIFT RELIEF AND TURNOVER PROCEDURES The FP&L program in response to this requirement has been developed for St. Lucie 1 (Docket No. 50-335) and also is applicable to St. Lucie Unit 2.
I.C.3 SHIFT SUPERVISOR RESPONSIBILITIES The FP&L program in response to this requirement has been developed for St. Lucie 1 (Docket No. 50-335) and also is applicable to St. Lucie Unit 2.
I.C.4 CONTROL ROOM ACCESS The FP&L program in response to this requirement has been developed for St. Lucie 1 (Docket No. 50-335) and also is applicable to St. Lucie Unit 2. Access limitations are also addressed in the site Security Plan.
I.C.5 PROCEDURES FOR FEEDBACK OF OPERATING EXPERIENCE TO PLANT STAFF Procedures have been generated to reflect the requirements of NUREG-0737. Administrative controls are addressed in the FPL Quality Assurance Topical Report discussed in Section 17.2.
I.C.6 VERIFY CORRECT PERFORMANCE OF OPERATING ACTIVITIES Performance and procedures currently in effect at St. Lucie Unit 1 reflect the requirements of NUREG-0737. This requirement is also met at St. Lucie Unit 2. Reviews and audits are covered in the FPL Quality Assurance Topical Report discussed in Section 17.2.
I.C.7 NSSS VENDOR REVIEW OF PROCEDURES The NRC reviewed selected emergency operating procedures as described in SER Supplement 2 and concluded that the NSSS vendor's comments have been acceptably incorporated into the selected emergency operating procedures.
1.9A-2 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 I.C.8 PILOT MONITORING OF SELECTED EMERGENCY PROCEDURES FOR NTOL APPLICANTS Any deficiencies identified by an NRC audit were corrected.
I.C.9 LONG TERM PROGRAM PLAN FOR UPGRADING PROCEDURES Generic Letter 82-33(8) requests that each licensee and applicant develop and submit to the NRC by April 15, 1983 its own plant-specific schedule for completion of the upgrading and implementation of Emergency Operating Procedures (EOPs). FP&L has upgraded and implemented the EOPs.
I.D.1 CONTROL ROOM DESIGN Generic Letter 82-33 requests that each licensee and applicant develop and submit to the NRC by April 15, 1983 its own plant-specific schedule for submittal of the Control Room Design Review Program Plan and of the Summary Report. FPL has submitted the Summary Report of the Detailed Control Room Design Review (DCRDR), dated October 1983. The history and methodology of the DCRDR is presented in UFSAR Section 7.7.3.
I.D.2 PLANT SAFETY PARAMETER DISPLAY SYSTEM The Safety Assessment System (SAS)/Emergency Response Data Acquisition And Display System (ERDADS) (refer to Appendix 7.5A) provides the Safety Parameter Display System (SPDS) and all other data required in the control room. The ERDADS system also provides data to Technical Support Center (TSC), Emergency Offsite Facility (EOF) and the Nuclear Data Link (NDL) through the PI server.
Generic Letter 82-33 requests that each licensee and applicant develop and submit to the NRC by April 15, 1983 its own plant-specific schedule for completion of the SPDS and submittal of the SAR and SPDS Implementation Plan. By letter L-83-238 dated April 15, 1983, FP&L indicated the following:
I.G.1 TRAINING DURING LOW - POWER TESTING This training is in accordance with Robert L. Tedesco, Assistant Director for Licensing to Dr.
Robert E. Uhrig letter dated June 12, 1981. (Subject, TMI-2 Action Plan Item I.G.1). Since testing was accomplished at a comparable prototype plant, SONGS-2, only the training required by this letter need be accomplished.
II.B.1 REACTOR COOLANT SYSTEM VENTS A description of the Reactor Coolant System Vents is provided in Subsection 9.3.7.
1.9A-3 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 II.B.2 PLANT SHIELDING A design review was conducted to evaluate the radiological environment of the plant following an accident in which significant core damage has occurred. The evaluation provides for access to vital areas and equipment needed for post-accident operations. A detailed description and results of this design review is provided in Appendix 12.3A.
For environmental qualification of safety-related equipment for post-accident conditions refer to Section 3.11.
II.B.3 POST-ACCIDENT SAMPLING A description of the Post-Accident Sampling System is provided in Subsection 9.3.6.
II.B.4 TRAINING FOR MITIGATING CORE DAMAGE Training criteria are discussed in Section 6.4 of the plant Technical Specifications.
II.D.1 RELIEF AND SAFETY VALVE TEST REQUIREMENTS The design and testing of these valves are summarized in Table 5.4-9 and Subsection 5.4.13.
FP&L's letter of March 22, 1983 from Mr. R Uhrig, FP&L to Mr. D Eisenhut, NRC, references two Combustion Engineering Topical Reports (9, 10) as documentation as to how the EPRI/NSAC test results are applicable to the St. Lucie 2 relief and safety valves.
The staff finds that the general approach in the reports of using the EPRI test results to demonstrate plant specific operability of the relief and safety valves is acceptable (see SER Supplement 3).
NUREG-0737 required utilities to evaluate the functional performance capabilities of PWR safety, PORV, and block valves and to verify the piping systems for normal, transient, and accident conditions. Reference 18 documents the NRC review and acceptance of performance capabilities of pressurizer safety valves, PORVs, and block valves. Qualification of the plant specific piping systems by performing the appropriate analyses is still under evaluation.
II.D.3 RELIEF AND SAFETY VALVE POSITION INDICATION Acoustic flow monitors are used for the indication of pressurizer safety relief and power operated relief valve position. Design information is presented in Subsection 7.6.3.10.
II.E.1.1 AUXILIARY FEEDWATER SYSTEM RELIABILITY EVALUATION
- a. A standard deterministic type of safety review has been performed using as principal guidance the acceptance criteria specified in Standard Review Plan 10.4.9 "Auxiliary Feedwater System" (R1) and Branch Technical Position ASB 10-1, "Design Guidelines for Auxiliary Feedwater System Pump Drive and Power Supply Diversity for PWR Plants" (R0). The results of this review are provided in Appendix 10.4.9A.
1.9A-4 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2
- b. The guidelines of Enclosure 2 of NRC letter to pending OL applicants dated March 10, 1980(11) has been addressed to describe the design basis accident and transients and the corresponding acceptance criteria for Auxiliary Feedwater System in Appendix 10.4.9A.
- c. Event tree and fault tree logic techniques have been conducted as part of a reliability analysis to determine dominant failure modes and assess Auxiliary Feedwater System reliability levels. The results of this reliability evaluation are provided in Appendix 10.4.9B.
II.E.1.2 AUXILIARY FEEDWATER INITIATION AND INDICATION Safety Grade Auxiliary Feedwater Flow indication and automatic initiation is implemented for St.
Lucie Unit 2 and is described in Subsections 10.4.9, 7.3.1.1.8, and 7.5.
II.E.3.1 EMERGENCY POWER SUPPLY FOR PRESSURIZER HEATERS St. Lucie Unit 2 employs a Combustion Engineering (CE) pressurized water nuclear steam supply system. An analysis performed by CE for St. Lucie Unit 2 has determined that 150 kilowatts of pressurizer heater capacity is needed to maintain hot standby conditions when offsite power is lost. CE recommends this minimum pressurizer heater capacity be available within two hours following loss of offsite power.
The St. Lucie Unit 2 design provides two heater banks each rated 200 kilowatts which are connected to separate 400-volt emergency power trains. The emergency power trains are energized from separate and independent diesel generators upon loss of offsite power. Each of the two heater banks has access to only one Class 1E diesel generator and their controls are likewise supplied from separate safety-grade power supplies. The pressurizer heaters are automatically shed from the Class 1E power system upon the occurrence of a Safety Injection Actuation Signal (SIAS). Procedures for manually loading the pressurizer heaters onto the emergency power sources following an SIAS are available to the operator, and identify under what conditions selected loads can be shed from the emergency bus to prevent overloading of the diesel generators when the pressurizer heaters are connected. The connection of the pressurizer heater elements and controls to the Class 1E buses is through safety-grade circuit breakers.
Based on NRC review, the staff concludes that the power supplies for pressurizer heaters are capable of being powered from both offsite and onsite emergency power systems. This is consistent with the staff positions and clarifications and is acceptable, as per the Safety Evaluation Report.
II.E.4.1 DEDICATED HYDROGEN PENETRATIONS This requirement is not applicable to St. Lucie Unit 2. EC295353 II.E.4.2 CONTAINMENT ISOLATION DEPENDABILITY The following items address corresponding NRC positions contained in NUREG-0737:
- a. As discussed in Subsection 7.3.1.1 the containment isolation actuation signal (CIAS) is initiated upon high containment pressure, high containment radiation or 1.9A-5 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 on SIAS actuation. Therefore, the CIAS complies with the recommendation in Standard Review Plan 6.2.4 "Containment Isolation System" (R1) with respect to diversity in the parameters sensed for initiation of containment isolation.
- b. Using the definition in Appendix A to the Branch Technical Position APCSB 3-1 (11/24/75) (attached to Standard Review Plan 3.6.1), essential system and components are defined as those systems and components required to shutdown the reactor and mitigate the consequences of an accident.
Table 6.2-52 identifies the essential penetrations as ESF penetrations. As indicated in Subsection 6.2.4, containment penetrations associated with nonessential systems are either administratively locked closed or automatically isolated upon a CIAS. Penetrations for systems like post-accident monitoring instrumentation and RCS sampling however are provided with manual override of the CIAS to enable the operator to open the containment isolation valves and activate the systems as necessary.
- c. The St. Lucie Unit 2 containment isolation system complies with General Design Criteria (GDC) 55, 56, and 57. A CIAS is used to isolate nonessential systems.
GDC 57 permits the use of one containment isolation valve located outside containment which is capable of automatic or remote manual operation and does not require closure on a CIAS. The penetrations that fall into this category are main steam and feedwater which are automatically isolated upon receipt of a MSIS. However, with the diversity of high containment pressure or low steam generator pressure, an MSIS is generated and isolates the main steam isolation valves and Main Feedwater isolation valves. The component cooling water lines to and from the reactor coolant pump fall under the requirements of GDC 56. An SIAS isolates these penetrations and is initiated by diverse parameters: low pressurizer pressure or high containment pressure.
- d. The present design of control systems for automatic containment isolation valves is such that resetting the isolation signal does not result in the automatic reopening of containment isolation valves. Certain valves (e.g., post-accident sampling, instrument air) which are required to open during an accident are provided with the capability of manually overriding the automatic isolation signal.
Reopening of these containment isolation valves requires deliberate operator action, and is accomplished only on a valve-by-valve basis. The containment isolation design does not utilize ganged control switches for containment isolation valves.
- e. A review of the operating history of containment pressure for St. Lucie Unit 1 was performed. (St. Lucie Units 1 & 2 have similar containment volumes and thermal power ratings). Pressure increases of up to two psi can be expected to occur from time to time during plant operation. The instrument loop error, including setpoint variances, effects of line voltage fluctuations, temperature effects and instrument drift is incorporated in the plant Technical Specification setpoint values.
1.9A-6 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2
- f. The containment purge valves comply with the operability criteria provided in Branch Technical Position CSB 6-4 (R1) and are maintained and surveyed pursuant to the plant Technical Specifications.
The 48 inch purge valves are verified to be closed at least every 31 days.
- g. The continuous containment purge valves close on a CIAS which, as stated in Item 1, is initiated upon a high radiation or high pressure inside containment.
II.F.1 ADDITIONAL ACCIDENT MONITORING INSTRUMENTATION Generic Letter No. 82-33 requests that each licensee and applicant develop and submit to the NRC by April 15, 1983 its own plant specific schedule for submittal of the Regulatory Guide (RG) 1.97 Evaluation Report describing how RG 1.97 has been met. FP&L submitted this material in Letter L-83-573.
For a discussion of the FPL compliance with RG 1.97 see Section 7.5.2.9 II.F.2 INSTRUMENTS FOR CORE COOLING Description of Instruments for Core Cooling is provided in Subsections 3.9.5.1.5 and 7.5.4.
II.G.1 EMERGENCY POWER FOR PRESSURIZER EQUIPMENT The description of the operation of the PORV and PORV block valves is found in Subsection 5.2.6.
The PORVs are powered from safety-related 125V dc buses 2A and 2B and are available continuously. The PORV block valves are powered from safety-related 480V ac motor control centers which are powered through the onsite distribution system. Upon loss of offsite power, the diesel generator is started and powers the onsite system (refer to Section 8.3). Therefore, the PORV block valves receive reliable power in the event they are required to operate during a loss of offsite power. The design is acceptable to NRC as per the SER.
II.K.1 IE BULLETINS ON MEASURES TO MITIGATE SMALL-BREAK LOCAS AND LOSS OF FEEDWATER ACCIDENTS As per the requirements of NUREG-0737, only two concerns under this item are applicable to St. Lucie Unit 2. These concerns are addressed below.
II.K.1.5 REVIEW ESF VALVES All safety-related valve positions, positioning requirements, and positive controls were reviewed, and documented in Table 1.9A-1, to assure that valves remain positioned (open or closed) in a manner to ensure the proper operation of engineered safety features.
The provision of complete display of instrumentation is an integral part of the design of systems required for safe shutdown and accident mitigation. A major component of the display information provided in the control room is position indication for valves and HVAC dampers.
Table 1.9A-1 lists all active valves and dampers that may be required to operate to achieve safe shutdown or mitigate the consequences of an accident. For most valves and dampers position indicating lights are provided on control panels in the control room. For all other valves and 1.9A-7 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 dampers whose failure might have adverse consequences, sufficient information is available for position determination in the control room (refer to Table 1.9A-1).
The related procedures for maintenance, testing, plant and system startup and supervisory periodic surveillance require that these valves are returned to their correct positions following necessary manipulation and are maintained in their proper position during all operational modes. These procedures have been developed in response to this NUREG-0737 requirement for St. Lucie Unit 1 (Docket No. 50-335) and also are applicable to St. Lucie Unit 2.
II.K.1.10 OPERABILITY STATUS FP&L programs in response to this requirement have been developed for St. Lucie Unit 1 (Docket No. 50-335) and also are applicable to St. Lucie Unit 2. As indicated in NUREG-0660 (not clarified by NUREG-0737) for units applying for operating licenses, this item is addressed in Items I.D.2 and I.C.6 above.
II.K.2.13 THERMAL MECHANICAL REPORT-EFFECT OF HIGH PRESSURE FP&L is participating in CE Owners Group generic efforts to evaluate the effect of high pressure safety injection on reactor vessel integrity in response to Item II.K.2.13 of NUREG-0737 (see Subsection 5.3.3.8). FP&L concurs with the CE Owners Group evaluation as reported in CEN 189 and CEN 189 Appendix F, December 1981. Staff review of this item is covered in their Unresolved Safety Issues program, issue A-49, Pressurized Thermal Shock." See SER Supplement 2.
II.K.2.17 POTENTIAL FOR VOIDING IN THE REACTOR COOLANT SYSTEM DURING TRANSIENTS II.K.2.
17.1 DESCRIPTION
In the event a void formation is identified in the Reactor Coolant System the operators are trained to implement a procedure to mitigate voiding. The NSSS vendor has completed an extensive analysis of voiding in the Reactor Coolant System. The results show that rapid refill and drain of the reactor vessel head does not cause stress levels in excess of those occurring during a normal cooldown at 100°F/hour. The results of this analysis for St. Lucie Unit 1, which is applicable to St. Lucie Unit 2, are provided in Appendix 5.2C.
Reactor Coolant System Cooldown rate is addressed in Amendment 4, Subsection 5.4.7.5.
FP&L is also participating in the CE Owners Group effort to address item II.K.2.17; FP&L concurs with the evaluation as reported in CEN-199(12).
II.K.2.19 SEQUENTIAL AUXILIARY FEEDWATER FLOW ANALYSIS As indicated by the NRC (letter from R A Clark, Chief Operating Reactor Branch 3, Division of Licensing to R E Uhrig, Vice President, Florida Power & Light Co., dated July 2, 1981), this item is not applicable to CE supplied steam generators which utilize inverted U tubes.
1.9A-8 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 II.K 3.1 INSTALLATION AND TESTING OF AUTOMATIC PORV ISOLATION SYSTEM II.K.3.2 REPORT ON OVERALL SAFETY EFFECT OF PORV ISOLATION SYSTEM FP&L has participated in CE Owners Group activities conducted since the Three Mile Island accident to address various aspects of PORV design and operation. These activities have included review of operating experience with PORVs on CE reactors, development of input to the EPRI program for testing these valves, review of requirements for emergency power to the PORVs and the associated block valves, development of a recommendation for PORV position indication, review and updating of emergency procedure guidelines to assure PORV operation is adequately addressed, and development of associated operator training materials. The requirements of Action Plan Item II.K.3.2 have also been addressed as a CE Owners Group activity in CEN-145(13).
It has been concluded based on the CE Owners Group activities that the addition of an automatic PORV isolation system on St. Lucie Unit 2 to further decrease the probability of a small-break loss-of-coolant accident caused by a stuck-open PORV is not necessary. This conclusion is based on the following considerations. First, the design of the PORV actuation logic is such that the valves are only actuated coincident with the high pressurizer pressure trip of the reactor. The PORV cases are not used prior to the Reactor Protection System actuation in an attempt to avoid the reactor trip. Thus, challenges to the PORVs are reduced because the margin between the normal operating pressure and the high pressure reactor trip is maximized.
The success of this design approach is evident based on the operating experience compiled to date which has only 19 challenges to the PORVs in 29 reactor-years of operation on CE plants (data from a recent survey of the CE Owners Group). It should be noted that 11 of these 19 challenges were caused by a turbine runback feature which has been removed. The PORVs successfully reclosed in each case where they were challenged.
The second consideration for not needing an automatic PORV isolation system is that various actions have been taken which significantly improve the reliability of the PORVs and associated block valves. The elimination of the turbine runback feature mentioned previously, and the provision of a direct reliable means for indicating PORV position to the operator reduce the recurrence frequency of a small break LOCA due to PORV failure by an estimated factor of 15.
Improved operator training programs, improved emergency procedures, and the provision of emergency power to the PORVs and block valves reduce the small break LOCA recurrence frequency further although the exact magnitude has not been quantified.
The final consideration for not needing an automatic PORV isolation system is that the recurrence frequency of a small break LOCA due to PORV failure has been substantially reduced by the actions mentioned previously to an estimated value which falls well within the uncertainty band of the recurrence frequencies for a LOCA due to a small pipe rupture estimated in WASH-1400.
Thus, the recurrence frequency is now at an acceptably low value. The incorporation of an automatic PORV isolation system would further increase PORV system reliability. However, this action is not considered to be necessary since the recurrence frequency of PORV system failures without this feature is small.
1.9A-9 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 II.K.3.3 REPORTING SAFETY VALVE AND PORV FAILURE AND CHALLENGES FP&L assures that any failure of a PORV or safety valve to close is reported to the NRC promptly. All challenges to the PORVs or safety valves are documented in a Special Report.
II.K.3.5 AUTOMATIC TRIP OF REACTOR COOLANT PUMPS DURING A LOCA FP&L is a member of the CE Owners Group. The CE Owners Group has selected an operational strategy which will close out TMI Action Plan Item II.K.3.5, "Automatic Trip of Reactor Coolant Pumps." Following a current review of several possible strategies, the strategy chosen is to trip two pumps initially followed by the trip of the remaining two pumps at the same time a LOCA has been diagnosed. The "trip two, leave two" strategy has been discussed in the past as a preferred approach. Based on the currently available information, it remains the preference of CE, the CE Owners Group, and FP&L. A program is being developed whose goal will be to provide information which both meets the NRC guidelines stated in the reference letter and provides the operational requirements for participating utilities to use in developing emergency operating procedures and conducting training. The expectation is that the selected operational strategy for the RCPs will make use of manual operator actions. The operational strategy currently in use on St. Lucie Unit No. 2 is to trip all RCPs during the initial phase of a depressurization transient followed by pump restart when it is confirmed that the event is not a LOCA. This strategy will remain in effect until replaced by the new approach which will be implemented with supported by appropriate documentation and operator training.
The NRC, in Reference 17, has concluded that the CE Owner's Group methodology significantly improves reactor safety. The adoption and implementation of this methodology resolves TMI Action Plan Item II.K.3.5 satisfactorily.
II.K.3.17 REPORT ON OUTAGES OF EMERGENCY CORE-COOLING SYSTEMS LICENSEE REPORT AND PROPOSED TECHNICAL SPECIFICATION CHANGES Reports on ECCS outages will follow the guidelines of 10 CFR 50.73 for the development and content of License Event Reports which will document any significant problems with the ECCS equipment. Other ECCS equipment failures are reported via the Institute of Nuclear Power Operations (INPO) Equipment Performance and Information Exchange System (EPIX), formally known as Nuclear Plant Reliability Data System (NPRDS). These two methods provide an on-line reporting system which satisfies the requirements of NUREG-0737, Item II.K.3.17.
These methods were accepted by the NRC in Reference 16.
II.K.3.25 EFFECT OF LOSS OF AC POWER ON PUMP SEALS FP&L has conducted a test of RCP seals under simulated loss of ac power conditions of full temperature and pressure. After approximately 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> at coolant conditions of 550°F and 2250 psig, the RCP seal cartridge still performed satisfactorily with the pump idle. Some seal damage was observed during the post-test inspection; however, the maximum seal leakage during the test was only 16 gph (
Reference:
FP&L letter L-81-107, March 10, 1981).
PCM 98021 replaced the RCP SU mechanical seals with N-9000 seals. An aged N-9000 seal has been rigorously tested by Flowserve (OEM) in a test fixture to simulate the conditions 1.9A-10 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 imposed by a station blackout for an eight (8) hour period. During this test downward shaft movements and pressure changes were imposed.
II.K.3.30 REVISED SMALL BREAK LOCA METHODS TO SHOW COMPLIANCE WITH 10 CFR PART 50, APPENDIX K NRC Generic Letter 83-10b(14) documents NRC evaluation of the analyses of LOFT Test L3-6 performed by the CE Owners Group and concludes that the evaluations acceptably predict the test results, and finds the currently approved CE evaluation model for small LOCAs in continued conformance with 10 CFR 50 Appendix K for the case of limited RCP operation after reactor trip, and for the range of licensed CE reactor designs.
II.K.3.31 PLANT SPECIFIC CALCULATIONS TO SHOW COMPLIANCE WITH 10 CFR PART 50.46 See Item II.K.3.30 of NUREG-0737.
III.A.1.1 UPGRADE EMERGENCY PREPAREDNESS The St. Lucie Plant Emergency Plan discussed in Section 13.3 incorporates the requirements of this task.
III A.1.2 UPGRADE EMERGENCY SUPPORT FACILITIES FP&L programs in response to this requirement have been or are being developed for St. Lucie Unit 1 (Docket No. 50-335) and also are applicable to St. Lucie Unit 2.
Generic Letter No. 82-33 requests that each licensee and applicant develop and submit to the NRC by April 15, 1983 its own plant-specific schedule for completion of the Emergency Response Facilities (ERFs). By letter L-83-238 dated April 15, 1983, FP&L indicated the ERFs schedule is as follows:
The TSC is operational.
- b. Operational Support Center (OSC)
The OSC is operational.
- c. Emergency Operation Facility (EOF)
The EOF is operational.
III.D.1.1 INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT LIKELY TO CONTAIN RADIOACTIVE MATERIAL In the unlikely event of an accident, the Containment Isolation Actuation Signal (CIAS) isolates all non-essential systems, thereby eliminating all large radioactive leakage paths from containment. The only means of leakage into the Reactor Auxiliary Building is through ESF system components (i.e., pump seals, valve leakage, etc.) and post-accident monitoring sample lines. Liquid leakages collected in the ECCS room sumps are normally routed to the equipment drain tank in the Waste Management System (WMS). The normal operational mode of the 1.9A-11 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 ECCS room sump pumps has not been modified. On high sump water level, the pumps discharge to the equipment drain tank. To prevent radioactive contaminants from entering the WMS, the ESF Leakage Collection and Return System (see Subsection 9.3.5) provides operators with a method to direct ESF leakage to the containment. This system eliminates highly radioactive liquid from entering normally "Low activity" waste hold-up tanks. Likewise, all sources of high activity sample gas (e.g., hydrogen sampling) are re-routed to the containment, thus eliminating contamination of the Waste Gas System. The above described design precludes the use of Liquid and Gaseous Waste Management Systems during an unlikely event of an accident.
The following systems contain high activity fluid during a postulated accident:
- a. Shutdown Cooling System
- b. High Pressure Safety Injection (Recirculation Phase)
- c. Containment Spray (Recirculation Phase)
- d. Sampling System.
Periodic integrated leak testing, at intervals not to exceed each refueling cycle, is established for these systems. A program is established to evaluate results and initiate leakage reduction measures as appropriate.
III.D.3.3 IN PLANT RADIATION MONITORING FP&L programs in response to this requirement have been developed for St. Lucie Unit 1 (Docket No. 50-335) and are applicable to St. Lucie Unit 2. The Health Physics procedures address detailed radioiodine assessment. These are generally described in Subsection 12.5.3.
Training is an integral part of the non-licensed training program is covered in the plant Technical Specifications.
III.D.3.4 CONTROL ROOM HABITABILITY Potential hazards in the vicinity of the site have been identified and evaluated to confirm that operators in the control room are adequately protected (refer to Section 2.2). In addition, radioactive releases have been analyzed for their effects on control room operators (refer to Section 6.4). Liquid source terms from within the Reactor Auxiliary Building, although not factored into the dose rate to the operators presented in Section 6.4, would have insignificant impact in terms of doses because the control room itself is located on top of the Reactor Auxiliary Building and is well separated from liquid source terms.
1.9A-12 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2
REFERENCES:
APPENDIX 1.9A
- 1. U.S. Nuclear Regulatory Commission, "Clarification of TMI Action Plan Requirements,"
USNRC Report NUREG-0737, October, 1980.
- 2. NUREG-0843, Safety Evaluation Report related to the operation of St. Lucie Plant, Unit No. 2, Docket No. 50-389; October 1981.
- 3. NUREG-0843, Supplement No. 1; December 1981.
- 4. NUREG-0843, Supplement No. 2; September 1982.
- 5. NUREG-0843, Supplement No. 3; April 1983.
- 6. NUREG-0843, Supplement No. 4; June 1983.
- 7. CEN-152, "Combustion Engineering Emergency Procedure Guidelines," dated November 22, 1982.
- 8. NRC Generic Letter 82-33, Supplement 1 to NUREG-0737 - Requirements for Emergency Response Capability, dated December 17, 1982.
- 9. CEN-227, "Summary Report on the Operability of Pressurizer Safety Valves in CE Designed Plants," December 1982.
- 10. CEN-213, "Summary Report on the Operability of Powered Operated Relief Valves,"
July 1982.
- 11. Letter from D. F. Ross Jr., NRC to All Pending Operating License Applicants of Nuclear Steam Supply Systems Designed by Westinghouse and Combustion Engineering,
Subject:
Actions Required from Operating License Applicants of Nuclear Steam Supply Systems Designed by Westinghouse and Combustion Engineering Resulting from the NRC Bulletins and Orders Task Force Review Regarding the Three Mile Island Unit 2 Accident, dated March 10, 1980.
- 12. CEN-199, "Effects of Vessel Head Voiding During Transients and Accidents in CE-NSSS's," March 1982.
- 13. CEN-145, "PORV Failure Reduction Methods-Final Report," December 1980.
- 14. NRC Generic Letter No.83-10b, Resolution of TMI Action Items II.K.3.5, "Automatic Trip of Reactor Coolant Pumps," dated February 8, 1983.
Eisenhut, "SPDS Implementation Plan and Parameter Selection Report."
- 16. Letter, from E. G. Tourigny (NRC) to W. F. Conway (FPL), "Emergency Core Cooling Systems (ECCS) Outages, 5-Year Report - St Lucie Plant Unit No. 2," dated May 11, 1988.
1.9A-13 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2
- 17. Letter, from J. A. Norris/G.E. Edison (NRC) to W.F. Conway (FPL), "Closing of Multiplant Action G-01-Reactor Coolant Pump Trip (NUREG-0737 Item II.K.3.5)," dated March 15, 1989 18 Letter, from J. A. Norris (NRC) to C.O. Woody (FPL), "NUREG-0737 Item II.D.1 Performance Testing of Relief and Safety Valves," dated May 11,1989.
- 19. NRC letter, J. A. Norris to J. H. Goldberg, "Instrumentation To Follow The Course Of An Accident (Regulatory Guide 1.97)", dated April 1, 1992
- 20. FPL Letter to NRC, L-92-194, "St. Lucie Unit 1 and 2 Docket No. 50-335 and 50-389 Regulatory Guide 1.97", dated July 14, 1992
- 22. NRC letter, J. A. Norris to J. H. Goldberg, "ST. Lucie Units 1 And 2 - Proposed Modifications Related To Regulatory Guide 1.97 (TAC Nos. 64333 And 64334)", dated November 12, 1991 1.9A-14 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.9A-1 SAFETY RELATED VALVE POSITION AND POSITION INDICATION ACTUATION NORMAL VALVE ACCIDENT VALVE(a) FAILURE METHOD OF(b)
SYSTEM VALVE FUNCTION TYPE OPERATOR SIGNAL POSITION POSITION MODE POSITION INDICATION Reactor Coolant V1474 LTOP Ang. Glb. Sol. --- Closed Closed Closed 1 V1475 LTOP Ang. Glb. Sol. --- Closed Closed Closed 1 V1476 LTOP Isol. Gate Motor --- Open Closed As Is 1 V1477 LTOP Isol. Gate Motor --- Open Closed As Is 1 V1460 RV Head Vent Glb. Solenoid --- Closed Closed Closed 1 V1461 RV Head Vent Glb. Solenoid --- Closed Closed Closed 1 V1462 RV Head Vent Glb. Solenoid --- Closed Closed Closed 1 V1463 RV Head Vent Glb. Solenoid --- Closed Closed Closed 1 V1464 RV Head Vent Glb. Solenoid --- Closed Closed Closed 1 V1465 RV Head Vent Glb. Solenoid --- Closed Closed Closed 1 V1466 RV Head Vent Glb. Solenoid --- Closed Closed Closed 1 Chemical and V2522 Cont. Isol. Glb. Pneu. CIAS Open Closed Closed 1 Volume Control V2508 BAMT. Isol. Gate Motor SIAS Closed Open As Is 1 V2509 BAMT. Isol. Gate Motor SIAS Closed Open As Is 1 V2514 BAMP. Disch. Gate Motor SIAS Closed Open As Is 1 V2525 PMW Supply Gate Motor SIAS Closed Closed As Is 1 V2504 RWT Supply Gate Motor --- Closed Open As Is 1 V2515 Cont. Isol. Glb. Pneu. SIAS Open Closed Closed 1 V2516 Cont. Isol. Glb. Pneu. SIAS/CIAS Open Closed Closed 1 SE-02-3 Aux. Spray Glb. Sol. --- Locked Closed Open Closed 1 SE-02-4 Aux. spray Glb. Sol. --- Locked Closed Open Closed 1 SE-02-1 Charging Glb. Sol. --- Open Open Open 1 SE-02-2 Charging Glb. Sol. --- Open Open Open 1 V2553 Charg. Bypass Glb. Motor --- Open © Closed As Is 2 V2554 Charg. Bypass Glb. Motor --- Open © Closed As Is 2 V2555 Charg. Bypass Glb. Motor --- Open © Closed As Is 2 V2523 Charg. Isol. Glb. Pneu. --- Locked Open Open Open 1 FCV-2210Y BAMT Supply Glb. Pneu. SIAS Closed Closed Closed 1 V2524 Cont.Isol. Glb. Pneu. CIAS Open Closed Closed 1 V2505 Cont.Isol. Glb. Pneu. CIAS Open Closed Closed 1 V2501 VCT Isol. Gate Motor SIAS Open Closed As Is 1 V2650 BAMT Recirc Glb. Pneu. SIAS Open Closed Closed 1 V2651 BAMT Recirc Glb. Pneu. SIAS Open Closed Closed 1 Safety FCV-3301 SDC BFY Motor --- Locked Open Open As Is 1 Injection FCV-3306 SDC BFY Motor --- Locked Open Open As Is 1 HCV-3512 SDC BFY Motor --- Locked Closed Open As Is 1 HCV-3657 SDC BFY Motor --- Locked Closed Open As Is 1 V3456 SDC Isol. Gate Motor --- Locked Closed Open As Is 1 V3457 SDC Isol. Gate Motor --- Locked Closed Open As Is 1 V3517 SDC Isol. Gate Motor --- Locked Closed Open As Is 1 V3658 SDC Isol. Gate Motor --- Locked Closed Open As Is 1 V3540 Hot Leg Inj. Glb. Motor --- Locked Closed Open As Is 1 V3550 Hot Leg Inj. Glb. Motor --- Locked Closed Open As Is 1 T1.9A-1 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.9A -1(Contd)
ACTUATION NORMAL VALVE ACCIDENT VALVE(a) FAILURE METHOD OF(b)
SYSTEM VALVE FUNCTION TYPE OPERATOR SIGNAL POSITION POSITION MODE POSITION INDICATION Safety V3523 Hot Leg Inj. Glb. Motor --- Locked Closed Open As Is 1 Injection(Cont'd) V3551 Hot Leg Inj. Glb. Motor --- Locked Closed Open As Is 1 V3656 HPSI Isol. Gate Motor --- Locked Open Open As Is 1 V3654 HPSI Isol. Gate Motor --- Locked Open Open As Is 1 SE-03-2A Cont. Isol. Glb. Sol. SIAS/CIAS Closed Closed Closed 1 SE-03-2B Cont. Isol. Glb. Sol. SIAS/CIAS Closed Closed Closed 1 V3659 Recirc. Gate Motor RAS Locked Open Closed As Is 1 V3660 Recirc. Gate Motor RAS Locked Open Closed As Is 1 V3495 Recirc. Glb. Sol. RAS Locked Open Closed Closed 1 V3611 SIT Drain Glb. Pneu. SIAS Closed Closed Closed 1 V3621 SIT Drain Glb. Pneu. SIAS Closed Closed Closed 1 V3631 SIT Drain Glb. Pneu. SIAS Closed Closed Closed 1 V3641 SIT Drain Glb. Pneu. SIAS Closed Closed Closed 1 V3496 Recirc. Glb. Sol. RAS Locked Open Closed Closed 1 HCV-3615 Inj. Glb. Motor SIAS Closed Open As Is 1 HCV-3625 Inj. Glb. Motor SIAS Closed Open As Is 1 HCV-3635 Inj. Glb. Motor SIAS Closed Open As Is 1 HCV-3645 Inj. Glb. Motor SIAS Closed Open As Is 1 HCV-3616 Inj. Glb. Motor SIAS Closed Open As Is 1 HCV-3626 Inj. Glb. Motor SIAS Closed Open As Is 1 HCV-3636 Inj. Glb. Motor SIAS Closed Open As Is 1 HCV-3646 Inj. Glb. Motor SIAS Closed Open As Is 1 HCV-3617 Inj. Glb. Motor SIAS Closed Open As Is 1 HCV-3627 Inj. Glb. Motor SIAS Closed Open As Is 1 HCV-3637 Inj. Glb. Motor SIAS Closed Open As Is 1 HCV-3647 Inj. Glb. Motor SIAS Closed Open As Is 1 V3480 SDC Isol. Gate Motor --- Locked Closed Open As Is 1 V3481 SDC Isol. Gate Motor --- Locked Closed Open As Is 1 V3651 SDC Isol. Gate Motor --- Locked Closed Open As Is 1 V3652 SDC Isol. Gate Motor --- Locked Closed Open As Is 1 V3545 SDC X-Tie Gate Motor --- Locked Open Open As Is 1,8 V3664 SDC Isol. Gate Motor --- Locked Closed Open As Is 1 V3665 SDC Isol. Gate Motor --- Locked Closed Open As Is 1 V3536 SDC Warmup Glb. Motor --- Locked Closed Open As Is 1 V3539 SDC Warmup Glb. Motor --- Locked Closed Open As Is 1 V3614 SIT Isol. Gate Motor SIAS Locked Open Open As Is 1,8 V3624 SIT Isol. Gate Motor SIAS Locked Open Open As Is 1,8 V3634 SIT Isol. Gate Motor SIAS Locked Open Open As Is 1,8 V3644 SIT Isol. Gate Motor SIAS Locked Open Open As Is 1,8 SE-03-1A SIT Drain Glb. Sol. SIAS Closed Closed Closed 1 SE-03-1B SIT Drain Glb. Sol. SIAS Closed Closed Closed 1 SE-03-1C SIT Drain Glb. Sol. SIAS Closed Closed Closed 1 SE-03-1D SIT Drain Glb. Sol. SIAS Closed Closed Closed 1 HCV-3618 CV Leakage Glb. Pneu. SIAS Closed Closed Closed 1 HCV-3628 CV Leakage Glb. Pneu. SIAS Closed Closed Closed 1 T1.9A-2 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.9A -1(Contd)
ACTUATION NORMAL VALVE ACCIDENT VALVE(a) FAILURE METHOD OF(b)
SYSTEM VALVE FUNCTION TYPE OPERATOR SIGNAL POSITION POSITION MODE POSITION INDICATION Safety HCV-3638 CV Leakage Glb. Pneu. SIAS Closed Closed Closed 1 Injection(Cont'd) HCV-3648 CV Leakage Glb. Pneu. SIAS Closed Closed Closed 1 V3571 Inj. Relief Glb. Pneu. SIAS Closed Closed Closed 1 V3572 Inj. Relief Glb. Pneu. SIAS Closed Closed Closed 1 V3444 RWT Isol Gate Motor --- Locked Open Closed As Is 1 V3432 RWT Isol Gate Motor --- Locked Open Closed As Is 1 Sampling SE-05-1A Cont. Isol. Glb. Sol. CIAS Closed Closed Closed 1 SE-05-1B Cont. Isol. Glb. Sol. CIAS Closed Closed Closed 1 SE-05-1C Cont. Isol. Glb. Sol. CIAS Closed Closed Closed 1 SE-05-1D Cont. Isol. Glb. Sol. CIAS Closed Closed Closed 1 SE-05-1E Cont. Isol. Glb. Sol. CIAS Closed Closed Closed 1 V5200 Cont. Isol. Glb. Sol. CIAS Closed Closed Closed 1 V5201 Cont. Isol. Glb. Sol. CIAS Closed Closed Closed 1 V5202 Cont. Isol. Glb. Sol. CIAS Closed Closed Closed 1 V5203 Cont. Isol. Glb. Pneu. CIAS Closed Closed Closed 1 V5204 Cont. Isol. Glb. Pneu. CIAS Closed Closed Closed 1 V5205 Cont. Isol. Glb. Pneu. CIAS Closed Closed Closed 1 SIT Vent Valves V3733 SIT Vent to Atm. Glb. Sol. --- Closed Closed Closed 1 V3734 SIT Vent to Atm. Glb. Sol --- Closed Closed Closed 1 V3735 SIT Vent to Atm. Glb. Sol. --- Closed Closed Closed 1 V3736 SIT Vent to Atm. Glb. Sol. --- Closed Closed Closed 1 V3737 SIT Vent to Atm. Glb. Sol. --- Closed Closed Closed 1 V3738 SIT Vent to Atm. Glb. Sol. --- Closed Closed Closed 1 V3739 SIT Vent to Atm. Glb. Sol. --- Closed Closed Closed 1 V3740 SIT Vent to Atm. Glb. Sol. --- Closed Closed Closed 1 Waste V6341 Cont. Isol. Diaph. Pneu. CIAS Open Closed Closed 1 Management V6342 Cont. Isol. Diaph. Pneu. CIAS Open Closed Closed 1 V6718 Cont. Isol. Diaph. Pneu. CIAS Open Closed Closed 1 V6750 Cont. Isol. Diaph. Pneu. CIAS Open Closed Closed 1 V6741 Cont. Isol. Glb. Pneu. CIAS Open Closed Closed 1 Main HCV-08-1A Cont. Isol Glb. Pneu. MSIS Open Closed As Is 1 Steam HCV-08-1B Cont. Isol Glb. Pneu. MSIS Open Closed As Is 1 MV-08-1A Warmup Glb. Motor MSIS Closed Closed As Is 1 MV-08-1B Warmup Glb. Motor MSIS Closed Closed As Is 1 MV-08-18A ADV Glb. Motor --- Closed Open As Is 1 MV-08-18B ADV Glb. Motor --- Closed Open As Is 1 MV-08-19A ADV Glb. Motor --- Closed Open As Is 1 MV-08-19B ADV Glb. Motor --- Closed Open As Is 1 MV-08-12 Aux. Stm Gate Motor AFAS Closed Open As Is 1 MV-08-13 Aux. Stm Gate Motor AFAS Closed Open As Is 1 T1.9A-3 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.9A - 1(Contd)
ACTUATION NORMAL VALVE ACCIDENT VALVE(a) FAILURE METHOD OF(b)
SYSTEM VALVE FUNCTION TYPE OPERATOR SIGNAL POSITION POSITION MODE POSITION INDICATION Main MV-08-3 Aux.Stm Glb. Motor --- Locked Open Open As Is 1 Steam (Cont'd) MV-08-14 ADV Isol. Gate Motor --- Open Open As Is 1 MV-08-15 ADV Isol. Gate Motor --- Open Open As Is 1 MV-08-16 ADV Isol Gate Motor --- Open Open As Is 1 MV-08-17 ADV Isol. Gate Motor --- Open Open As Is 1 Main Feed HCV-09-1A Cont. Isol. Gate Hyd. MSIS/AFAS Open Closed* As Is 1 Water HCV-09-1B Cont. Isol. Gate Hyd. MSIS/AFAS Open Closed* As Is 1 HCV-09-2A Cont. Isol. Gate Hyd. MSIS/AFAS Open Closed* As Is 1 HCV-09-2B Cont. Isol. Gate Hyd. MSIS/AFAS Open Closed* As Is 1 MV-09-9 Aux. Feed Glb. Motor AFAS Closed Open/Closed As Is 1 MV-09-10 Aux. Feed Glb. Motor AFAS Closed Open/Closed As Is 1 MV-09-11 Aux. Feed Glb. Motor AFAS Closed Open/Closed As Is 1 MV-09-12 Aux. Feed Glb. Motor AFAS Closed Open/Closed As Is 1 MV-09-13 Aux. Feed Gate Motor --- Closed Open As Is 1 MV-09-14 Aux. Feed Gate Motor --- Closed Open As Is 1 SE-09-2 Aux. Feed Isol. Glb. Sol. AFAS Closed Open/Closed Closed 1 SE-09-3 Aux. Feed Isol. Glb. Sol. AFAS Closed Open/Closed Closed 1 SE-09-4 Aux. Feed Isol. Glb. Sol. AFAS Closed Open/Closed Closed 1 SE-09-5 Aux. Feed Isol. Glb. Sol. AFAS Closed Open/Closed Closed 1 Intake MV-21-2 Sys. Isol. BFY Motor SIAS Open Closed As Is 1 Cooling MV-21-3 Sys. Isol. BFY Motor SIAS Open Closed As Is 1 Water Component HCV-14-8A Sys. Isol. BFY Pneu. SIAS Open Closed Closed 1 Cooling HCV-14-8B Sys. Isol. BFY Pneu. SIAS Open Closed Closed 1 Water MV-14-17 FP. Isol. BFY Motor SIAS Open Closed As Is 1 MV-14-18 FP. Isol. BFY Motor SIAS Closed Closed As Is 1 MV-14-19 FP. Isol. BFY Motor --- Open Closed As Is 1 MV-14-20 FP. Isol. BFY Motor --- Closed Closed As Is 1 MV-14-9 Fan Isol. BFY Motor --- Open Open As Is 1 MV-14-10 Fan Isol. BFY Motor --- Open Open As Is 1 MV-14-11 Fan Isol. BFY Motor --- Open Open As Is 1 MV-14-12 Fan Isol. BFY Motor --- Open Open As Is 1 MV-14-13 Fan Isol. BFY Motor --- Open Open As Is 1 MV-14-14 Fan Isol. BFY Motor --- Open Open As Is 1 MV-14-15 Fan Isol. BFY Motor --- Open Open As Is 1 MV-14-16 Fan Isol. BFY Motor --- Open Open As Is 1
- The AFAS maybe overridden and the valve re-opened by the control room operator only during 2-EOP-06, Total Loss of Feedwater.
T1.9A-4 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.9A - 1 (Contd)
ACTUATION NORMAL VALVE ACCIDENT VALVE(a) FAILURE METHOD OF(b)
SYSTEM VALVE FUNCTION TYPE OPERATOR SIGNAL POSITION POSITION MODE POSITION INDICATION Component HCV-14-1 RCP Isol. BFY Pneu. SIAS Open Closed Closed 1 Cooling Water HCV-14-2 RCP Isol. BFY Pneu. SIAS Open Closed Closed 1 (Cont'd) HCV-14-6 RCP Isol. BFY Pneu. SIAS Open Closed Closed 1 HCV-14-7 RCP Isol. BFY Pneu. SIAS Open Closed Closed 1 HCV-14-9 Sys. Isol. BFY Pneu. SIAS Open Closed Closed 1 HCV-14-10 Sys. Isol. BFY Pneu. SIAS Open Closed Closed 1 HCV-14-3A SDC HX BFY Pneu. SIAS Closed Open Open 1 HCV-14-3B SDC HX BFY Pneu. SIAS Closed Open Open 1 MV-14-1 CCW Pump Isol. BFY Motor --- Open (1) Open (1) As Is 1 MV-14-2 CCW Pump Isol. BFY Motor --- Closed (2) Closed (2) As Is 1 MV-14-3 CCW Pump Isol. BFY Motor --- Open (1) Open (1) As Is 1 MV-14-4 CCW Pump Isol. BFY Motor --- Closed (2) Closed (2) As Is 1 Primary Water HCV-15-1 Cont. Isol. Glb. Pneu. CIAS Closed Closed Closed 1 Instr. Air HCV-18-1 Cont. Isol. Glb. Pneu. CIAS Open Closed Closed 1 Station Air HCV-18-2 Cont. Isol. Glb. Pneu. CIAS Closed Closed Closed 1 Steam FCV-23-3 Cont. Isol. Gate Pneu. CIAS Open Closed Closed 1 Generator FCV-23-5 Cont. Isol. Gate Pneu. CIAS Open Closed Closed 1 Blowdown FCV-23-7 Cont. Isol. Glb. Pneu. CIAS Open Closed Closed 1 FCV-23-9 Cont. Isol. Glb. Pneu. CIAS Open Closed Closed 1 Diesel Oil SE-59-1A1 Oil Supply Glb. Sol. --- Closed Open Closed 3 SE-59-1A2 Oil Supply Glb. Sol. --- Closed Open Closed 3 SE-59-1B1 Oil Supply Glb. Sol. --- Closed Open Closed 3 SE-59-1B2 Oil Supply Glb. Sol. --- Closed Open Closed 3 HVAC FCV-25-1 Cont. Isol. BFY Pneu. CIAS Closed Closed Closed 1 FCV-25-2 Cont. Isol. BFY Pneu. CIAS Closed Closed Closed 1 FCV-25-3 Cont. Isol. BFY Pneu. CIAS Closed Closed Closed 1 FCV-25-4 Cont. Isol. BFY Pneu. CIAS Closed Closed Closed 1 FCV-25-5 Cont. Isol. BFY Pneu. CIAS Closed Closed Closed 1 FCV-25-6 Cont. Isol. BFY Pneu. CIAS Closed Closed Closed 1 FCV-25-20 Cont. Isol. BFY Pneu. CIAS Open Closed Closed 1 FCV-25-21 Cont. Isol. BFY Pneu. CIAS Open Closed Closed 1 FCV-25-26 Cont. Isol. BFY Pneu. CIAS Open Closed Closed 1 FCV-25-36 Cont. Isol. BFY Pneu. CIAS Open Closed Closed 1 FCV-25-7 Vac. Relief BFY Pneu. Cont. Press Closed Open Closed 1 FCV-25-8 Vac. Relief BFY Pneu. Cont. Press Closed Open Closed 1 FCV-25-29 SBVS Isol. BFY Motor --- Locked Closed Closed As Is 1 FCV-25-30 Cont. Isol. BFY Motor CIAS Open Closed As Is 1 FCV-25-31 Cont. Isol. BFY Motor CIAS Open Closed As Is 1 FCV-25-32 Cont. Isol. BFY Motor CIAS Closed Open As Is 1 FCV-25-33 Cont. Isol. BFY Motor CIAS Closed Open As Is 1 FCV-25-34 SBVS Isol. BFY Motor --- Locked Closed Closed As Is 1 T1.9A-5 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.9A - 1(Contd)
ACTUATION NORMAL VALVE ACCIDENT VALVE(a) FAILURE METHOD OF(b)
SYSTEM VALVE FUNCTION TYPE OPERATOR SIGNAL POSITION POSITION MODE POSITION INDICATION HVAC (Cont'd) FCV-25-14 CRECS Isol. BFY Motor CIAS Open Closed As Is 1 FCV-25-15 CRECS Isol. BFY Motor CIAS Open Closed As Is 1 FCV-25-16 CRECS Isol. BFY Motor CIAS Open Closed As Is 1 FCV-25-17 CRECS Isol. BFY Motor CIAS Open Closed As Is 1 FCV-25-18 CRECS Isol. BFY Motor CIAS Open Closed As Is 1 FCV-25-19 CRECS Isol. BFY Motor CIAS Open Closed As Is 1 FCV-25-24 CRECS Isol. BFY Motor CIAS Open Closed As Is 1 FCV-25-25 CRECS Isol. BFY Motor CIAS Open Closed As Is 1 FCV-25-11 SBVS Isol. BFY Motor Diff. Pres. Closed Open As Is 1 FCV-25-12 SBVS Isol. BFY Motor Diff. Pres. Closed Open As Is 1 Containment MV-07-1A RWT Isol. BFY Motor RAS Open Closed As Is 1 Spray MV-07-1B RWT Isol. BFY Motor RAS Open Closed As Is 1 MV-07-2A Sump Isol. BFY Motor RAS Closed Open As Is 1 MV-07-2B Sump Isol. BFY Motor RAS Closed Open As Is 1 FCV-07-1A Cont. Isol. BFY Pneu. CSAS Closed Open Open 1 FCV-07-1B Cont. Isol. BFY Pneu. CSAS Closed Open Open 1 LCV-07-11A Cont. Isol. Glb. Pneu. CIAS/SIAS Closed Closed Closed 1 EC LCV-07-11B Cont. Isol. Glb. Pneu. CIAS/SIAS Closed Closed Closed 1 291 MV-07-3 Cont. Spray Isol. Gate Motor --- Open Open As Is 1 393 MV-07-4 Cont. Spray Isol. Gate Motor --- Open Open As Is 1 SE-07-5A thru 5D Cont. Pressure Globe Sol. --- Open Open Open 1 SE-07-5E, 5F Cont. Pressure Globe Sol. --- Open Open Open 1 Containment FCV-26-1 Cont. Isol. Glb. Pneu. CIAS Open Closed Closed 1 Air FCV-26-2 Cont. Isol. Glb. Pneu. CIAS Open Closed Closed 1 Monitoring FCV-26-3 Cont. Isol. Glb. Pneu. CIAS Open Closed Closed 1 FCV-26-4 Cont. Isol. Glb. Pneu. CIAS Open Closed Closed 1 FCV-26-5 Cont. Isol. Glb. Pneu. CIAS Open Closed Closed 1 FCV-26-6 Cont. Isol. Glb. Pneu. CIAS Open Closed Closed 1 Hydrogen FSE-27-8 Cont. Isol. Glb. Sol. --- Closed Open Closed 1 Sampling FSE-27-9 Cont. Isol. Glb. Sol. --- Closed Open Closed 1 FSE-27-10 Cont. Isol. Glb. Sol. --- Closed Open Closed 1 FSE-27-11 Cont. Isol. Glb. Sol. --- Closed Open Closed 1 FSE-27-12 Cont. Isol. Glb. Sol. --- Closed Open Closed 1 FSE-27-13 Cont. Isol. Glb. Sol. --- Closed Open Closed 1 FSE-27-14 Cont. Isol. Glb. Sol. --- Closed Open Closed 1 FSE-27-15 Cont. Isol. Glb. Sol. --- Closed Open Closed 1 FSE-27-16 Cont. Isol. Glb. Sol. --- Closed Open Closed 1 FSE-27-17 Cont. Isol. Glb. Sol. --- Closed Open Closed 1 FSE-27-18 Cont. Isol. Glb. Sol. --- Closed Open Closed 1 T1.9A-6 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.9A - 1(Contd)
ACTUATION NORMAL VALVE ACCIDENT VALVE(a) FAILURE METHOD OF(b)
SYSTEM VALVE FUNCTION TYPE OPERATOR SIGNAL POSITION POSITION MODE POSITION INDICATION HVAC D-17A Cont. Room N/A Motor CIAS (d) Closed Open Open 1 D-17B Cont. Room N/A Motor CIAS (d) Closed Open Open 1 D-18 Cont. Room N/A Motor CIAS (d) Closed Open Open 1 D-19 Cont. Room N/A Motor CIAS (d) Closed Open Open D-29 FHB Isol. N/A Motor High Rad Open Closed Closed 1 D-30 FHB Isol. N/A Motor High Rad Open Closed Closed 1 D-31 FHB Isol. N/A Motor High Rad Open Closed Closed 1 D-32 FHB Isol. N/A Motor High Rad Open Closed Closed 1 D-33 FHB Isol. N/A Motor High Rad Open Closed Closed 1 D-34 FHB Isol. N/A Motor High Rad Open Closed Closed 1 D-35 FHB Isol. N/A Motor High Rad Open Closed Closed 1 D-36 FHB Isol. N/A Motor High Rad Open Closed Closed 1 D-23 SBVS Cont. N/A Motor Diff. Pres. Open Open Open 4 D-24 SBVS Cont. N/A Motor Diff. Pres. Open Open Open 4 HVAC D-1 RAB Isol. N/A Motor SIAS Open Open Open 5 D-2 RAB Isol. N/A Motor SIAS Open Open Open 5 D-3 RAB Isol. N/A Motor SIAS Open Open Open 5 D-4 RAB Isol. N/A Motor SIAS Open Open Open 5 D-9A RAB Isol. N/A Motor SIAS Open Closed Closed 1 D-9B RAB Isol. N/A Motor SIAS Open Closed Closed 1 D-12A RAB Isol. N/A Motor SIAS Open Closed Closed 1 D-12B RAB Isol. N/A Motor SIAS Open Closed Closed 1 D-7A RAB Isol. N/A Motor SIAS Open Closed Closed 6 D-7B RAB Isol. N/A Motor SIAS Open Closed Closed 6 D-8A RAB Isol. N/A Motor SIAS Open Closed Closed 6 D-8B RAB Isol. N/A Motor SIAS Open Closed Closed 6 D-5A RAB Isol. N/A Motor SIAS Open Closed Closed 1 D-5B RAB Isol. N/A Motor SIAS Open Closed Closed 1 D-6A RAB Isol. N/A Motor SIAS Open Closed Closed 1 D-6B RAB Isol. N/A Motor SIAS Open Closed Closed 1 D-13 RAB Isol. N/A Motor SIAS (d) Open Open Open 1 D-14 RAB Isol. N/A Motor SIAS (d) Open Open Open 1 D-15 RAB Isol. N/A Motor SIAS (d) Open Open Open 1 D-16 RAB Isol. N/A Motor SIAS (d) Open Open Open 1 L-7A RAB Isol. N/A Motor SIAS (d) Open Open Open 7 L-7B RAB Isol. N/A Motor SIAS (d) Open Open Open 7 T1.9A-7 Amendment No. 27 (03/22)
UFSAR/St. Lucie - 2 TABLE 1.9A-1 (Cont'd)
Notes:
a) Accident Valve Position The designation "open" or "closed" indicates the position as a result of an ESFAS signal or a position that may be manually selected as part of a post accident procedure.
b) Method of Position Indication in the Control Room
- 1) Position Indicating Lights.
- 2) Failure of valve to close would result in low flow indication by flow transmitter FIA-22l2.
- 3) Failure of valve to open would result in a low-low alarm for Diesel Generators Day tank.
- 4) Failure of damper to open would result in low flow indication by flow transmitter FIS-25-20A1 or 20B1 for D23 and D24 respectively.
- 5) Failure of damper to open would result in high differential pressure indication by pressure transmitter PDIS-25-16A or 16B.
- 6) Each damper is backed up by redundant counterpart. Failure of one damper to close would result in no adverse consequence.
- 7) Failure of damper to open would result in low flow indication by flow transmitter FIS-25-21A1 or 21B1 for 2L-7A and 7B respectively.
- 8) Analog Position indicator; Indicator power separate from control power.
c) Valve position is dependent on charging pump running status. See section 9.3.4.2.2g for details.
d) Damper is actuated to its accident position by the start signal of its associated fan.
e) Normal Valve Position
T1.9A-8 Amendment No. 27 (03/22)