ML20141L646
| ML20141L646 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 05/04/2020 |
| From: | Florida Power & Light Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20141L672 | List:
|
| References | |
| L-2020-043 | |
| Download: ML20141L646 (117) | |
Text
LIST OF EFFECTIVE PAGES CHAPTER 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT SITE CHARACTERISTICS Page Amendment Page Amendment 1-1 30 F1.2-10 15 1-2 30 F1.2-11 15 F1.2-12 15 1-i 28 F1.2-13 15 1-ii 23 F1.2-14 15 1-iii 26 F1.2-15 15 1-iv 23 F1.2-16 15 F1.2-17 15 1.1-1 26 F1.2-18 15 1.1-1a 26 F1.2-19 15 1.2-1 26 F1.2-19A 25 1.2-2 29 F1.2-20 22 1.2-3 29 F1.2-21 0 1.2-4 29 F1.2-22 23 1.2-5 29 F1.2-23 22 1.2-5a 29 1.2-6 22 F1.2-24 23 1.2-7 26 F1.2-25 22 1.2-8 26 F1.2-26 23 1.2-9 30 F1.2-27 23 1.2-10 22 F1.2-28 23 1.2-11 26 F1.2-29 23 1.2-12 22 F1.2-30 0 1.2-13 29 1.2-14 26 1.3-1 0 1.2-15 4 1.3-2 0 1.2-16 4 1.3-3 0 1.2-17 4 1.3-4 0 1.2-18 4 1.3-5 0 1.2-19 4 1.3-6 0 1.2-20 4 1.3-7 0 1.2-21 4 1.3-8 0 1.2-22 22 1.3-9 0 1.2-23 28 1.3-10 30 1.2-24 26 deleted 1.3-11 0 1.3-12 0 F1.2-1 15 1.3-13 0 F1.2-2 15 1.3-14 0 F1.2-3 15 F1.2-4 15 1.4-1 15 F1.2-5 15 F1.2-6 15 1.5-1 0 F1.2-7 15 1.5-2 0 F1.2-8 15 1.5-3 0 F1.2-9 15 1.5-4 0 UNIT 1 1-1 Amendment No. 30 (05/20)
LIST OF EFFECTIVE PAGES (Cont'd)
CHAPTER 1 (Contd)
Page Amendment 1.5-5 0 1.5-6 0 1.5-7 0 1.5-8 0 1.5-9 0 1.5-10 0 1.5-11 0 1.5-12 0 1.5-13 0 1.6-1 1 1.6-2 0 1.6-3 0 F1.6-2 16 1.7-1 0 1.7-2 0 1.7-3 0 1.7-4 0 1.7-5 0 1.7-6 0 1.7-7 0 1.7-8 20 1.7-9 0 1.7-10 0 1.7-11 0 1.7-12 0 1.7-13 0 1.7-14 0 1.7-15 0 1.7-16 0 1.7-17 0 1.7-18 0 1.7-19 17 1.7-20 0 1.7-21 0 1.7-22 0 UNIT 1 1-2 Amendment No. 30 (05/20)
INTRODUCTION AND GENERAL DESCRIPTION OF PLANT CHAPTER 1 TABLE OF CONTENTS Section Title Page
1.1 INTRODUCTION
1.1-1 1.2 GENERAL PLANT DESCRIPTION 1.2-1 1.2.1 PRINCIPAL SITE CHARACTERISTICS 1.2-1 1.2.2.1 PRINCIPAL DESIGN CRITERIA 1.2-1 1.2.2.2 Reactor 1.2-1 1.2.2.3 Reactor Coolant and Auxiliary Systems 1.2-3 1.2.2.4 Containment Structure 1.2-3 1.2.2.5 Engineered Safety Features 1.2-4 1.2.2.6 Instrumentation and Control 1.2-4 1.2.2.7 Electrical Systems 1.2-5 1.2.2.8 Waste Management and Radiation Protection 1.2-5 1.2.2.9 Fuel Handling and Storage 1.2-5 1.2.2.10 Power Conversion 1.2-5 1.2.2.11 Independent Spent Fuel Storage Installation (ISFSI) 1.2-5a 1.2.2.12 Low Level Waste Storage Facility 1.2-5a 1.2.3 OPERATING CHARACTERISTICS AND SAFETY CONSIDERATIONS 1.2-6 1.2.3.1 Nuclear Steam Supply System 1.2-6 1.2.3.2 Engineered Safety Features and Emergency Systems 1.2-6 1.2.3.3 Protection, Control, Instrumentation and Electrical Systems 1.2-9 1.2.3.4 Power Conversion System 1.2-11 1.2.3.5 Fuel Handling and Storage Systems 1.2-11 1.2.3.6 Cooling Water and Other Auxiliary Systems 1.2-12 1.2.3.7 Radioactive Waste Management System 1.2-13 1.2.4 MAJOR STRUCTURES AND EQUIPMENT ARRANGEMENT 1.2-14 1.2.5 (Deleted) 1.2-16 1.2.6 SHARED SYSTEMS AND INTERCONNECTIONS BETWEEN UNIT 1 AND 1.2-22 UNIT 2 UNIT 1 1-i Amendment No. 28 (05/17)
CHAPTER 1 TABLE OF CONTENTS (Cont'd)
Section Title Page 1.2.7 SYMBOLS AND ABBREVIATIONS ON FIGURES 1.2-23 1.
2.8 REFERENCES
FOR SECTION 1.2 1.2-23 1.3 COMPARISONS 1.3-1 1.3.1 COMPARISON WITH SIMILAR FACILITY DESIGNS 1.3-1 1.3.2 COMPARISON OF PRELIMINARY AND FINAL DESIGN 1.3-2 1.3.2.1 General 1.3-2 1.3.2.2 Site Characteristics 1.3-2 1.3.2.3 Design Criteria 1.3-2 1.3.2.4 Reactor 1.3-3 1.3.2.5 Reactor Coolant System 1.3-3 1.3.2.6 Engineered Safety Features 1.3-3 1.3.2.7 Instrumentation and Control 1.3-4 1.3.2.8 Electrical Systems 1.3-4 1.3.2.9 Auxiliary Systems 1.3-4 1.3.2.10 Steam and Power Conversion System 1.3-5 1.3.2.11 Radioactive Waste Management 1.3-5 1.3.2.12 Radiation Protection 1.3-5 1.3.2.13 Conduct of Operations 1.3-5 1.3.2.14 Initial Tests and Operations 1.3-5 1.3.2.15 Accident Analyses 1.3-5 1.3.2.16 Technical Specifications 1.3-6 1.3.2.17 Quality Assurance 1.3-6 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4-1 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5-1 1.6 MATERIALS INCORPORATED BY REFERENCE 1.6-1 DESIGN COMPLIANCE WITH AEC SAFETY GUIDES, INFORMATION GUIDES 1.7 1.7-1 AND CODE OF FEDERAL REGULATIONS CROSS REFERENCES 1-ii Amendment No. 23 (11/08)
INTRODUCTION AND GENERAL DESCRIPTION OF PLANT CHAPTER 1 LIST OF TABLES Table Title Page 1.3-1 Plant Parameter Comparison 1.3-7 1.6-1 Materials Incorporated By Reference 1.6-2 1.7-1 Design Compliance 1.7-2 UNIT 1 1-iii Amendment No. 27 (04/15)
INTRODUCTION AND GENERAL DESCRIPTION OF PLANT CHAPTER 1 LIST OF FIGURES Figure 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 Mezzanine Floor Plan 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 Plans, Sheet 1 1.2-9 General Arrangement - Reactor Building Plans, Sheet 2 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 Sections, Sheet 1 1.2-16 General Arrangement - Reactor Auxiliary Building Sections, Sheet 2 1.2-17 General Arrangement - Reactor Auxiliary Building Miscellaneous Plans and Sections 1.2-18 General Arrangement - Fuel Handling Building Plans 1.2-19 General Arrangement - Fuel Handling Building Sections 1.2-19A Erection Equipment Location Plan 1.2-20 Flow Diagram Symbols 1.2-21 Control and Block Diagram 1.2-22 Deleted 1.2-23 Instrument Type 1.2-24 Deleted 1.2-25 System Number 1.2-26 Deleted 1.2-27 Deleted 1.2-28 Deleted 1.2-29 Deleted 1.2-30 Equipment Symbols - Sheet 2 1.6-2 Control Wiring Diagram Index 1-iv Amendment No. 23 (11/08)
CHAPTER 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT This chapter was originally prepared to describe the St. Lucie Unit 1 Structure, Systems, and Components (SSC) during the initial fuel cycle. Since that time changes have been made to reflect the uprating of the unit to a stretch power level of 2700 MWt and subsequent Extended Power Uprate (EPU) to 3020 MWt. Where applicable, changes have been made to reflect the uprating of the unit to an Extended Power Uprate (EPU) power level of 3020 MWt. Where information associated with the higher power level is not available, the existing information is identified as pre-EPU.
1.1 INTRODUCTION
This Updated Final Safety Analysis Report is submitted in accordance with the requirements of 10 CFR 50.71(e). It is based on the original FSAR, including 57 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 the St. Lucie Plant. 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 update information which is accurate for the period up to six months prior to the most recent revision of this document. The update information 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 record. Exceptions include Chapters 13, 14 and 17.
The nuclear steam supply system (NSSS) is a pressurized water reactor designed by Combustion Engineering Incorporated. The containment structure is comprised of a steel containment vessel surrounded by a reinforced concrete shield building and was designed by Ebasco Services Incorporated.
St. Lucie Unit 1 was originally licensed and operated at 2560 MWt. Florida Power and Light (FPL) submitted an application to the NRC in support of St. Lucie Unit 1 Stretch Power Uprate (SPU) (Reference 1). The NRC approved the application in License Amendment No. 48 dated November 23, 1981 (ML013530273) which allowed operation of the Unit up from 2560 MWt to 2700 MWt; an increase of approximately 5.5%.
FPL submitted an application to the NRC for an Extended Power Uprate (EPU) (Reference 2) which provided the basis to increase the licensed thermal power of St. Lucie Unit 1 from 2700 MWt to 3020 MWt for an increase of approximately 11.85% which includes a 10% EPU and a 1.7% measurement of uncertainty recapture (MUR).
The net increase from the current core thermal power level is calculated as follows:
(2700 MWt x 1.10) x 1.017% 3020 MWt (3020 - 2700 MWt) / 2700 MWt = 11.85%
The proposed EPU core power level represents an increase of approximately 18% from the original licensed thermal power level of 2560 MWt.
The core power level of 3020 MWt represents 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 evaluation of certain postulated accidents (see Chapter 15 for details). The corresponding net electrical output for the core thermal power level is approximately 1052 MWe during the winter months and 1032 MWe during the summer months.
1.1-1 Amendment No. 26 (11/13)
REFERENCES FOR SECTION 1.1
- 1. R. E. Uhrig (FPL) to D. G. Eisenhut; U. S. Nuclear Regulatory Commission (L-80-381), Application for Stretch Power, November 14, 1980.
- 2. R. L. Anderson (FPL) to U. S. Nuclear Regulatory Commission (L-2010-259), License Amendment Request for Extended Power Uprate, November 22, 2010, Accession No. ML103560419.
1.1-1a Amendment No. 26 (11/13)
1.2 GENERAL PLANT DESCRIPTION 1.2.1 PRINCIPAL SITE CHARACTERISTICS The site for the St. Lucie Plant 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 feet above mean low water.
The island and the adjoining mainland are sparsely populated. The nearest population center is the City of Fort Pierce, which is eight miles from the site and has 29,721 people as of the 1970 census. The minimum site exclusion radius is 5,100 feet. Site characteristics are given in detail in Chapter 2.
1.2.2 PRINCIPAL DESIGN CRITERIA Principal structures, systems and equipment which may serve either to prevent accidents or to mitigate their consequences are designed and are 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 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 the protection of the public and plant personnel.
Provisions are made to minimize the probability and effect of fires and explosions.
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.
1.2.2.1 Reactor The following design criteria apply to the reactor:
a) 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 core thermal output is 3020 megawatts.
1.2-1 Amendment No. 26 (11/13)
b) The reactor fuel is slightly enriched uranium dioxide contained in Zircaloy tubes.
c) Minimum departure from nucleate boiling ratio (DNBR) during normal operation and anticipated transients is not below that value which could lead to fuel rod failure. 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 point of the UO2 is not reached during routine operation and anticipated transients.
d) Fuel rod clad is designed to maintain cladding integrity throughout fuel life. Fission gas release within the rods and other factors affecting design life is considered for the maximum expected exposures.
e) The reactor and control systems are designed so that any xenon transients will be adequately damped.
f) The reactor is designed to accommodate safely and without fuel UNIT1 1.2-2 Amendment No. 29 (10/18)
damage, the anticipated operational occurrences, backed up by automatic and redundant reactor trips.
g) The reactor coolant system is designed and constructed to maintain its integrity throughout the plant life. Appropriate means of test and inspection are provided.
h) Power excursions which could result from any credible reactivity addition will not cause damage, either by deformation or rupture of the reactor vessel and will not impair operation of the engineered safety features.
i) Control element assemblies (CEAs) are capable of holding the core subcritical at hot zero power conditions with margin following a trip, even with the most reactive CEA stuck in the fully withdrawn position.
j) 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.
k) 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.
Further details concerning the reactor are given in Chapter 4.
1.2.2.2 Reactor Coolant and Auxiliary Systems Heat removal systems are provided which can safely accommodate core heat output. Each of the heat removal systems is designed to provide reliable operation under all normal and expected transient circumstances. (See Chapters 5 and 9).
Components of the reactor coolant system are designed and will be operated so that no excessive pressure or thermal stress will be imposed on the structural materials. The necessary consideration has been given to the ductile characteristics of the materials at low temperatures.
1.2.2.3 Containment Structure The containment structure, including the access openings and penetrations, is designed to contain the pressures and temperatures resulting from a loss of coolant accident (LOCA) in which: (a) a range of reactor coolant pipe breaks, up to a double-ended break of the largest reactor coolant UNIT 1 1.2-3 Amendment No. 29 (10/18)
pipe, is assumed and the governing energy release into the containment therefrom is calculated; (b) there is a simultaneous loss of external electric power with coincident loss of a diesel generator; (c) additional heat is transferred from the reactor to containment by water supplied from the safety injection system; (d) at least one containment spray pump and one train of fan coolers (i.e., two units) functions; and (e) the engineered safety features are actuated within 30 seconds following an accident, except that the safety injection tanks operate when the system pressure drops below the tank pressure.
Means are provided for pressure and leak rate testing of the entire containment system including provisions for leak rate testing of individual piping and electrical penetrations that rely on gasketed seals, sealing compounds, or expansion bellows.
Integrity of the containment is assured against postulated missiles from plant equipment failures and against postulated missiles from external sources. (See Chapters 3 and 6 for details).
1.2.2.4 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, will not exceed the guidelines established for design basis accidents. The systems ensure that the Final Acceptance Criteria for Emergency Core Cooling Systems for Light Water Power Reactors (Federal Register Vol 39, Jan. 4, 1974) 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 will 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 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, (f) control room habitability system, and (g) an Iodine Removal System to remove radioactive iodine from the containment atmosphere to minimize post-LOCA effluents. (See Chapter 6 for details).
1.2.2.5 Instrumentation and Control Interlocks and automatic protective systems are provided along with administrative controls to ensure safe operation of the plant.
A reactor protective system is provided which initiates reactor trip if the reactor approaches prescribed safety limits.
Sufficient redundancy is installed to permit periodic testing of the reactor protective system so that failure or removal from service of any one protective system component or portion of the system will not preclude reactor trip or other safety action when required.
UNIT 1 1.2-4 Amendment No. 29 (10/18)
The protective system is separated from the control instrumentation systems so that failure or removal from service of any control instrumentation system component or channel does not inhibit the function of the protective system. (See Chapter 7 for details.)
St. Lucie Unit 1 was originally designed with feedwater (FW) flow and temperature instrumentation consisting of venturis, differential pressure transmitters and resistance temperature detectors (RTDs) for each FW header. Since then, improvements have occurred in FW flow and temperature measurement instrumentation and the associated power calorimetric uncertainty values. Based on the installation of new FW flow/FW temperature instrumentation and the associated reduction in reactor core power uncertainty values, FPL in conjunction with the EPU application provided the basis to increase the core rated thermal power by 1.7%.
The NRC defines power uprates less than 2% which are achieved by implementing enhanced techniques for calculating reactor power as Measurement Uncertainty Recapture (MUR).
Modifications required for the MUR portion of the EPU uprate include installation of the Cameron/Caldon Leading Edge Flow Meter (LEFM) CheckPlus system. Existing FW flow and temperature instrumentation is retained and used for comparison monitoring of the LEFM system and as a backup FW mass flow measurement when needed.
1.2.2.6 Electrical Systems Redundant sources of off-site power are provided by four separate transmission lines to ensure that no single failure of any active component can prevent a safe and orderly shutdown. These off-site power circuits are diverse, as three of the lines are overhead and the fourth is run underground and under the Indian River waterway.
Redundant sources of emergency on-site power are provided by two diesel generators to ensure safe shutdown in the event of complete loss of off-site power. (See Chapter 8 for details.)
1.2.2.7 Waste Management and Radiation Protection The waste management system is designed to provide controlled treatment 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 practical in accordance with the provisions of 10 CFR 50, Appendix I. (See Chapter 11 for details.)
Units 1 and 2 are provided with separate control rooms, housed in their respective reactor auxiliary buildings, each having adequate shielding to permit occupancy during all postulated accidents involving radiation releases.
The radiation shielding in the plant, in combination with plant radiation control procedures, ensures that operating personnel do not receive radiation exposures in excess of the applicable limits of 10 CFR 20 during normal operation and maintenance. (See Chapter 12 for details.)
1.2.2.8 Fuel Handling and Storage New and spent fuel handling and storage facilities (Section 9.1) are provided for the safe handling and storage of fuel and are designed to preclude accidental criticality. Dry storage of spent fuel is provided as discussed in Section 1.2.2.11.
1.2.2.9 Power Conversion The steam and power conversion system is provided to generate steam in two steam generators for energy transfer to a regenerative cycle turbine generator which generates electricity. After expanding in the turbine, the steam is condensed in the main condenser. The condensate is collected and returned to the steam generators and reject heat is dissipated in the circulating water system. (See Chapter 10 for details.)
UNIT 1 1.2-5 Amendment No. 29 (10/18)
1.2.2.10 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 site 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.
1.2.2.11 Low Level Waste Storage Facility 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.
1.2-5a Amendment No. 29 (10/18)
1.2.3 OPERATING CHARACTERISTICS AND SAFETY CONSIDERATIONS 1.2.3.1 Nuclear Steam Supply System The reactor core is fueled with uranium dioxide pellets enclosed in Zirconium alloy tubes with welded end plugs.
The tubes are fabricated into assemblies in which end fittings prevent axial motion and grids prevent lateral motion of the tubes. The control element assemblies (CEAs) consist of Inconel clad boron carbide absorber rods which are guided by Zirconium alloy tubes located within the fuel assembly. The core consists of 217 fuel assemblies loaded with multiple U-235 enrichments.
The reactor vessel and its closure head are fabricated from manganese moly steel internally clad with stainless steel. The vessel and its internals are designed so that the integrated neutron flux (E > 1 Mev) at the vessel wall will be less than 4.7 x 1019 n/cm2 over a 60-year period.
The internal structures include the core support barrel, the core support plate, the core shroud, 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 core support plate 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 minimizes 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 reactor coolant system is arranged as two closed loops connected in parallel to the reactor vessel. Each loop consists of one 42-inch ID outlet (hot) pipe, one steam generator, two 30-inch ID inlet (cold) pipes and two pumps.
An electrically heated pressurizer is connected to the hot leg of one of the loops and a safety injection line is con-nected to each of the four cold legs.
The reactor coolant system operates at a nominal pressure of approximately 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.
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 shaft leakage is minimized by mechanical seals. Each pump motor is equipped with an anti-reverse mechanism to prevent reverse rotation of any pump that is not in operation.
1.2.3.2 Engineered Safety Features and Emergency Systems Engineered safety features systems protect the public and plant personnel in the highly unlikely event of an accidental release of radioactive fission products from the reactor system, particularly as the result of a LOCA.
The safety features function to localize, control, mitigate, and terminate such accidents to hold exposure levels below applicable limits.
UNIT 1 1.2-6 Amendment No. 27 (04/15)
The engineered safety features are:
a) The safety injection system (including high pressure and low pressure safety injection pumps and the safety injection tanks) b) The containment system c) The containment spray system d) The containment cooling system e) The shield building ventilation system f) The containment isolation system g) The hydrogen control system h) The control room habitability system 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 systems also provide 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 containment is comprised of a steel containment vessel surrounded 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 hypothetical accident resulting in substantial core meltdown and release of fission products as defined in 10 CFR 50. It is a cylindrical vessel with hemispherical dome and ellipsoidal bottom. The shield building is a medium leakage concrete structure which surrounds the 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 (MHA). (See Section 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 worst postulated LOCA at a nominal core power level of 3020 Mwt.
The containment is equipped with two spray systems and two cooling systems for cooling the containment atmosphere following the postulated LOCA. (See Section 6.2.2 for details.)
1.2-7 Amendment No. 26 (11/13)
The containment sprays supply borated water to cool and reduce pressure in the containment atmosphere. The system is designed so that with one spray pump, one set of spray nozzles, and one shutdown cooling heat exchanger in operation, adequate cooling is provided to cool 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 is designed to provide 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.
The shield building ventilation system is provided to maintain a negative pressure in the annulus between the steel containment building 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 Section 6.2.3 for details.)
An 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 Section 6.2.4 for details.)
The hydrogen sampling system analyzes the post-LOCA containment atmosphere for potential hydrogen buildup by use of seven sample probe locations and an automatic hydrogen analyzer. If, through sampling, it is determined that the maximum allowable hydrogen concentration limits are being approached, the hydrogen purge system exhausts the containment atmosphere through charcoal filters. The hydrogen control system is discussed in UFSAR sections 6.2.5.
The control room habitability system is provided to limit control room doses due to airborne activity to within GDC 19 limits. The system is discussed in detail in Section 9.4.1.
The iodine removal system is provided which is designed to operate in conjunction with the containment spray system to remove radio-iodines from the containment atmosphere following a loss of coolant accident (LOCA).
This greatly reduces the quantity of radio-iodines released to the environment in the event of post-LOCA leakage from the containment vessel.
1.2-8 Amendment No. 26 (11/13)
1.2.3.3 Protection, Control, Instrumentation and Electrical Systems a) Reactor Protection 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 is designed to initiate protective action whenever the signal of any two of four channels reaches the preset limit. If any two of these four channels receives coincident signals, the power supply to the magnetic jack control element drive mechanism is interrupted releasing the control elements to drop into the core to shut down the reactor. Redundancy is provided in the reactor protective system to assure that no single failure will prevent protective action when it is required. The protective system is completely independent of and separate from the control system.
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 motion is initiated manually. EC291158 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 5 percent per minute and at greater rates over smaller load change increments up to a step change of 10 percent.
A temperature controller is used to compare the existing average reactor coolant temperature with the EC291158 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 EC291158 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. EC291158 The CEAs are moved through manual operation by the operator. EC291158 1.2-9 Amendment No. 30 (05/20)
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 due to reactor system temperature changes. The pressure and water level control systems are described in Section 7.7.1.2.
Overpressure protection of the reactor coolant system is provided by power operated relief valves and spring loaded safety valves connected to the pressurizer and designed in accordance with Section III of the ASME code.
In addition, overpressure protection of the reactor coolant system during low temperature, solid system mode (i.e.,
startup), is provided by the overpressure Mitigation System described in Section 5.2.2.6. The discharge from the pressurizer safety and relief valves is 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.
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 permit turbine trip without opening the steam generator safety valves when the condenser is available.
A steam generator water level control system regulates feedwater flow to the steam generator. (See Section 7.7.2.3) An auxiliary feedwater system is provided to ensure flow to the steam generators in the event the main feedwater supply is out of service.
c) Instrumentation System The nuclear instrumentation includes out-of-core and incore neutron flux detectors. Ten channels of out-of-core instrumentation monitor the neutron flux and provide reactor protection and control signals during start-up and power operation. Four of the channels monitor the neutron flux through the start-up range, and six channels monitor the neutron flux from zero power through the full power range. Of the latter, four are used for reactor protection and two for reactor control. (See Section 7.5.2.)
The incore instrumentation consists of self-powered rhodium and vanadium neutron detectors and thermocouples to provide information on neutron flux distribution and temperature in the core.
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. The boron concentration in the reactor coolant water is also monitored and the concentration is displayed in the control room.
Instrument signals penetrating the containment are electric. Instrument signal transmission for the remaining plant instruments is either electric or pneumatic. (See Chapter 7 for details.)
1.2-10 Amendment No. 22 (05/07)
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 instrumentation is contained in Chapter 7.
d) Electrical System The plant includes a 1,200 mva, 0.90 power factor generator delivering power to a 230 kv switchyard through step-up power transformers. Auxiliary power is utilized at 6.9 kv, 4.16 kv, 480 v, and 120 v ac; 125 v dc systems are also provided for emergency power, engineered safety features control, and essential nuclear instrumentation.
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.)
1.2.3.4 Power Conversion System 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 heater.
The auxiliary feedwater system contains two 325 gpm electric motor driven pumps and one 600 gpm pump driven by a non-condensing steam turbine. This system is designed to provide emergency heat removal capacity. (See Chapter 10 for details.)
1.2.3.5 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 tilting machines, 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 Section 9.1.4 for details.)
New fuel is stored dry in vertical racks in the fuel handling building. Room is provided for storing one third of a core. The rack and fuel assembly spacing precludes criticality. (See Section 9.1.1 for details.)
The fuel pool is a reinforced concrete structure, stainless steel lined, which contains high density spent fuel pool storage racks consisting of individual cells with 8.65 inch by 8.65 inch (nominal) square cross-section. Each cell can accommodate a single fuel assembly from St. Lucie Unit 1. A total of 1706 cells are arranged in 17 distinct modules of varying size in two regions. Region 1 is designed to store fuel assemblies with enrichments up to 4.6 weight-percent U-235 that have not achieved adequate burnup for Region 2. The Region 2 cells are capable of accommodating small quantities of fresh fuel in specialized storage arrangements; more generally, Region 2 can accommodate fuel assemblies with various initial enrichments which have accumulated minimum burnups with an acceptable bound. Additionally, a Region 1 fuel storage rack (cask pit rack) can be installed in the cask pit area of the Spent Fuel Pool. The single module cask pit rack consists of 143 individual storage cells each with an 8.58 inch by 8.58 inch (nominal) square cross section. The cask pit rack, when installed, can store fresh or spent fuel assemblies, regardless of burnup history, with enrichments up to 4.6 weight-percent U-235.
1.2-11 Amendment No. 26 (11/13)
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 Section 9.1.3 for details.)
1.2.3.6 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 Section 9.3.4 for details.)
b) Shutdown Cooling System The shutdown cooling system is used to reduce the temperature of the reactor coolant at a controlled rate from 325°F to a refueling temperature of approximately 135°F 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. (Section 9.3.5)
The component cooling system serves as a heat sink for the shutdown heat exchangers.
c) Component Cooling System The component cooling 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 all 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 system provides an intermediate barrier between the reactor coolant system and the intake cooling water system. (See Section 9.2.2 for details) d) 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.
1.2-12 Amendment No. 22 (05/07)
These systems are used for determining both chemical and radiochemical conditions of the various process fluids used in the plant.
e) 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 steam generator open blowdown heat exchangers.
The turbine cooling system removes heat from the main turbine lube oil cooler, main generator hydrogen coolers, main feed pump oil coolers, sample coolers, and other components by providing corrosion inhibited demineralized water to those components.
f) Plant Ventilation Systems Separate ventilation systems are provided for the containment vessel, the control room, the reactor auxiliary building, the fuel handling building and the diesel generators building. A purge system is provided for the containment vessel atmosphere.
The annular space between the steel containment vessel and the concrete shield building is ventilated by the shield building ventilation system utilizing charcoal filters. This system is automatically put into operation following EC a postulated LOCA. (See Section 6.2.3 for details.) 288994 g) Plant Fire Protection System The fire protection system supplies water to fire hydrants, deluge systems and hose racks in the various areas of the plant.
Noncombustible and fire resistant materials are used throughout the facility, particularly in areas containing critical portions of the plant such as the containment structure, control room, cable spreading room, and rooms containing components of the engineered safety features systems.
A number of portable fire extinguishers are placed at key locations for use in extinguishing limited fires. (See Section 9.5 for details.)
1.2.3.7 Radioactive Waste Management System The waste management system provides the means for controlled handling, storage and disposal of liquid, gaseous and solid wastes.
Reactor coolant from the chemical and volume control system and from the reactor drain tank is processed by the Liquid Waste Management system, which is comprised of pumps, filters, flash tank, ion exchangers, and holdup tanks. The effluent is sampled and may be discharged to the circulating water system if the radioactivity is within specified limits.
1.2-13 Amendment No. 29 (10/18)
Miscellaneous liquid wastes from the reactor auxiliary building are collected in the equipment and chemical drain tanks and subsequently processed by filtration, ion exchange and concentration. The distillate enters the waste condensate tank. If the radioactivity level of the liquid in the condensate tank is found to be high, the waste can be recycled through the waste ion exchanger. The liquid in the holding tank is sampled to ensure radioactivity levels are within the acceptable limits prior to discharge to the circulating water system.
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 integrity 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 drummed for removal from the plant. (See Chapter 11 for details).
1.2.4 MAJOR STRUCTURES AND EQUIPMENT ARRANGEMENT The turbine building for the St. Lucie Plant is oriented parallel to State Road A1A and the shoreline of the Atlantic Ocean, with the reactor containment structure located on the east, or seaward, side of the turbine building. The reactor auxiliary building is located perpendicular and close to the turbine building, oriented in an east-west direction. The fuel handling building is located next to the reactor containment building and the reactor auxiliary building, oriented in a north-south direction. The service building is located north of the turbine building.
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 in Figures 1.2-3 through 1.2-19.
The location of erection equipment (cranes) during construction of Unit 2 is shown in Figure 1.2-19a. This figure is maintained for historical purposes.
The containment structure houses the nuclear steam supply system (NSSS), consisting of the reactor, steam generators, reactor coolant pumps, pressurizer, and some of the other reactor auxiliaries. The containment structure is served by a polar bridge crane.
The auxiliary building houses the waste management facilities, engineered safety features components, heating and ventilating system components, switchgear, laboratories, offices, laundry and control room.
1.2-14 Amendment No. 26 (11/13)
The fuel handling building contains the spent fuel pool and new fuel storage facilities, as well as the cooling and purification equipment for the fuel pool. The fuel is transferred from the reactor containment 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 certain of the switchgear assemblies.
The service building provides offices, shop and warehouse space, and is located next to the turbine building unloading bay.
1.2-15 Am. 4-7/86
(DELETED) 1.2-16 Am. 4-7/86
(DELETED) 1.2-17 Am. 4-7/86
(DELETED) 1.2-18 Am. 4-7/86
(DELETED) 1.2-19 Am. 4-7/86
(DELETED) 1.2-20 Am. 4-7/86
(DELETED) 1.2-21 Am. 4-7/86
1.2.6 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 on-site and off-site, (i.e., station blackout) power can be transferred from the non-blacked out units emergency diesel generator set via the station blackout tie to one of the blacked-outs redundant Class 1E electrical distribution trains. Plant procedures limit the amount of the power transferred as not to affect the non-blacked out units 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 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.
The startup transformers (1A-2A, 1B-2B) are provided with a manual switching arrangement which permits paralleling 4.16 kV power to St. Lucie Units 1 and 2 (see section 8.2.1.3 for additional discussion).
1.2-22 Amendment No. 22 (05/07)
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 single 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 the other unit. A failure in any of the shared 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 1.2.2.11) is also shared by both units for dry storage of spent fuel.
The Low Level Waste Storage Facility (Section 1.2.2.12) is also shared by both units for onsite interim storage of low level waste prior to shipment offsite.
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 1.2-20, 1.2-21, 1.2-23, 1.2-25 and 1.2-30.
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 of 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 UNIT 1 1.2-23 Amendment No. 28 (05/17)
Refer to drawing 2998-G-058 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 SITE PLAN FIGURE 1.2-1 Amendment No. 15 (1/97)
Refer to drawing 2998-G-059 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 ENLARGED PLOT PLAN FIGURE 1.2-2 Amendment No. 15 (1/97)
Refer to drawing 8770-G-060 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT TURBINE BUILDING -
GROUND FLOOR PLAN FIGURE 1.2-3 Amendment No. 15 (1/97)
Refer to drawing 8770-G-061 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT TURBINE BUILDING -
MEZZANINE FLOOR PLAN FIGURE 1.2-4 Amendment No. 15 (1/97)
Refer to drawing 8770-G-062 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT TURBINE BUILDING -
OPERATING FLOOR PLAN FIGURE 1.2-5 Amendment No. 15 (1/97)
Refer to drawing 8770-G-063 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT TURBINE BUILDING -
SECTIONS SHEET 1 FIGURE 1.2-6 Amendment No. 15 (1/97)
Refer to drawing 87708-G-064 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT TURBINE BUILDING -
PLAN & SECTIONS SH2 FIGURE 1.2-7 Amendment No. 15 (1/97)
Refer to drawing 8770-G-065 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT REACTOR BUILDING -
FLOOR PLANS SH1 FIGURE 1.2-8 Amendment No. 15 (1/97)
Refer to drawing 8770-G-066 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT REACTOR BUILDING -
FLOOR PLANS SH 2 FIGURE 1.2-9 Amendment No. 15 (1/97)
Refer to drawing 8770-G-067 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT REACTOR BUILDING -
SECTIONS SHEET 1 FIGURE 1.2-10 Amendment No. 15 (1/97)
Refer to drawing 8770-G-068 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGMEENT REACTOR BUILDING --
SECTIONS SHEET 2 FIGURE 1.2-11 Amendment No. 15 (1/97)
Refer to drawing 8770-G-069 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING PLAN SHEET 1 FIGURE 1.2-12 Amendment No. 15 (1/97)
Refer to drawing 8770-G-070 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING PLAN SHEET 2 FIGURE 1.2-13 Amendment No. 15 (1/97)
Refer to drawing 8770-G-071 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING PLAN SHEET 3 FIGURE 1.2-14 Amendment No. 15 (1/97)
Refer to drawing 8770-G-072 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT REACTOR AUXILIARY BUIDLING SECTIONS SHEET 1 FIGURE 1.2-15 Amendment No. 15 (1/97)
Refer to drawing 8770-G-075 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGMENT REACTOR AUXILIARY BUILDING SECTIONS SHEET 2 FIGURE 1.2-16 Amendment No. 15 (1/97)
Refer to drawing 8770-G-076 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING MISCELLANEOUS PLANS FIGURE 1.2-17 Amendment No. 15 (1/97)
Refer to drawing 8770-G-073 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT FUEL HANDLING BUILDING PLANS FIGURE 1.2-18 Amendment No. 15 (1/97)
Refer to drawing 8770-G-074 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGMENT FUEL HANDLING BUILDING SECTIONS FIGURE 1.2-19 Amendment No. 15 (1/97)
Historical information only.
FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 ERECTION EQUIPMENT LOCATION PLAN FIGURE 1.2-19A Amendment No. 25 (04/12)
Refer to drawing 8770-G-078 Sheet 100 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 FLOW DIAGRAM SYMBOLS FIGURE 1.2-20 Amendment No. 22 (05/07)
LOGIC DIAGRAM LEGEND SYMBOL DESIGNATION ACTION TRIP SIGNAL LINE
~
GIVES OUTPUT ONLY IF AND GATE ALL INPUTS ARE PRESENT START OR PERMISSIVE SIGf'>AL l..INE
--~----------------------~ TIEBACK GIVES OUTPUT IF
=1&- OR GATE ANY INPUT IS PRESENT GIVES OUTPUT ONLY PNEUMATIC OR HYDRAULIC SIGNAL LINE
~ NOT GATE IF IN PUT IS NOT PRESENT PRODUCES OUTPUT SAME TIME DELAY
~ UNIT AS INPUT AFTER DELAY
~SEC.)
EB
A DEVICE WHICH RETAINS SWITCH BLOCK MEMORY ITS STATE UNTIL RE- SERVICE UNIT QUESTED TO .ALTER ITS ST .ATE TAG NO. I. I :-1 LOCATION FUNCTIONAL DESCRIPTION OUT PUT CHANGES ON
~
TIME DELAY
~
APPLICATION OF INPUT (WIPED OUT)
UNIT AND IS WIPED OUT AFTER SECS. I ~
POSITION OR SWITCH CONDITION 0 INDICATING LAMP-RED fND!CATING LAMP-GREEN
@ PUSH BUTTON MATCHING SYMBOLS EQUIPMENT BLOCK EQUIPMENT DESCRIPTION & NO.
~LETTER
~
DESIGNATION
~ SH.NO. EQUIPMENT CONDITIO~:
I OUTPUT ~RUN STOP)
MAIN CONTROL ROOM ANNUNCIATOR H - HIGH
~xx-xx:<- PANEL NO.
~L-LOW DWG. NO. !1770-6-276 LOCAL PANEL ANNUNCIATOR FLORIDA POWER !!. liGHT COMPANY H - HIGH St. Lucie Plant c:;:) XX - ) ( X X - P.ANEL NO.
L-LOW CONTROL AND BLOCK DIAGRA"'
FIGURE 1.2-2l
DELETED FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 FIGURE 1.2-22 Amendment No. 23 (11/08)
Refer to drawing 8770-B-270C Sheet I-8, I-9 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 INSTRUMENT TYPE FIGURE 1.2-23 Amendment No. 22 (05/07)
DELETED FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 FIGURE 1.2-24 Amendment No. 23 (11/08)
Refer to drawing 8770-B-270C Sheet I-10 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 SYSTEM NUMBER FIGURE 1.2-25 Amendment No. 22 (05/07)
DELETED FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 FIGURE 1.2-26 Amendment No. 23 (11/08)
DELETED FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 FIGURE 1.2-27 Amendment No. 23 (11/08)
DELETED FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 FIGURE 1.2-28 Amendment No. 23 (11/08)
DELETED FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 FIGURE 1.2-29 Amendment No. 23 (11/08)
l9U I PMEIIT II II HEAT EXCHANGER i MIXER
--Cs- POSIT I YE DISPLACEMENT PUMP 0 MANWAY OR
~CCE SS HATCH
--EY CEll TR I FUGAl PUMP -p EDUCT OR
-0r FAll.
BlOWER OR
_[_ VUT C0+4PRESSOR
-ET CANNED
- EIITR I FUGAl PUMP y EJECTOR ctJ TANK STRAINER
- I\' I Y-STRAINER
-G- IN-liNE STRAINER
-HJ- TEMPORARY STRII MER
-- fiLTER
---§- IMPUlS£ TRAP
--- lUCl£1 If lOA I TRAP DWG. NO. 8770-B-270 FLORIDA POWER & LIGHT COMPANY St. Lucie Plant EQUIPMENT 5YIIBOL5 - SHEET 2 FIGURE 1. 2-)(1
1.3 COMPARISONS Comparisons contained herein were considered valid at the time the operating license for St Lucie 1 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 COMPARISON WITH SIMILAR FACILITY DESIGNS Table 1.3-1 presents a summary of the characteristics of the St. Plant. The table includes similar data for Calvert Cliffs Units 1 and 2, Turkey Point Units 3 and 4, and Palisades Unit 1, as published in the FSAR for those units.
The parameters as listed in the Hutchinson Island PSAR are shown in parentheses for ease of comparison.
The Calvert Cliffs Units 1 and 2 design was selected for comparison because of the basic similarity of the reactor core and the nuclear steam supply system and to enable the reviewer to apply to the St. Lucie Plant the information gained from the review of Calvert Cliffs since the FSAR submittal in early 1971. Palisades was selected for comparison because of the similarity of the reactor coolant system and because it was the first large pressurized water reactor plant designed by Combustion Engineering to be licensed for operation. The Turkey Point station is included as representing a contemporary plant designed by another supplier.
The principal differences to be observed between the parameters listed for Calvert Cliffs and those for St. Lucie include the range of boron concentration, the number of control element assemblies (CEA's), and the reactor vessel minimum clad thickness. These differences have no effect on plant safety.
The difference in boron concentration and the number of CEA's reflects the evolution of the core design, with improved calculations providing a more accurate representation of the reactivity control capability and requirements of the CEA's and the worth of dissolved boron.
The containment concept used for this plant is similar to that which has received a construction permit at Prairie Island (Docket 50-282 and 50-306) and Kewaunee (Docket 50-305) except that the diameter is larger and the thickness is greater for this plant. Since the shell thickness is greater than 1 1/2 inches, field stress relieving of the complete vessel was necessary. Extensive experience exists in field stress relieving of large vessels, including reactor pressure vessels.
1.3-1
1.3.2 COMPARISON OF PRELIMINARY AND FINAL DESIGN Comparisons contained herein were considered valid at the time the operating license for St. Lucie Unit 1 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.
1.3.2.1 General This section contains a discussion of all significant changes that have been made in the plant design since submittal of the PSAR. Changes considered as significant include changes in design bases or criteria for safety related structures, systems or components, plant arrangement, mode of system operation, type of equipment or gross changes in component or system capacity.
1.3.2.2 Site Characteristics No significant site characteristic changes have been brought to light that would reflect a design change since submittal of the PSAR.
1.3.2.3 Design Criteria a) AEC General Design Criteria The PSAR contained a comparison of plant design to the Proposed (70) AEC General Design Criteria. It was on the basis of these GDC that the preliminary plant design was formulated and presented in the PSAR. Since the receipt of the construction permit (July 1, 1970) the (64) AEC General Design Criteria have been published as Appendix A to 10CFR50 (July 7, 1971). As far as was practicable, depending on engineering, construction and equipment procurement schedules, the criteria given in Appendix A were reflected in the plant design. A comparison of the final design with the present GDC is given in Section 3.1.
b) Design Codes It was stated in the PSAR that the design code for plant piping systems would be USAS B31.7 Nuclear Power Piping Code for reactor coolant system piping and USAS B31.1 Power Piping Code for other safety related piping. Procurement schedules allowed the use of B31.7 for all safety related plant piping and this code was applied in the design.
Also, although the PSAR contains no reference to its application, the Draft ASME Nuclear Pump and Valve Code was used in the design of safety related pumps and valves. The application of design codes used in the final design is discussed in Section 3.2.
c) Design Tornado Criteria It was stated in the PSAR that the design tornado windspeed for this site would be 337 mph. This value was changed at the request of the AEC before the construction permit was issued, and the design tornado used in the design of structures, systems and components consisted of 300 mph translational and 60 mph rotational wind speeds with a 3 psi pressure differential. The tornado design criteria are discussed in Section 3.3.2.
1.3-2
1.3.2.4 Reactor The following changes have been made to the reactor to improve the operating characteristics.
a) Control Element Drive Mechanisms Magnetic jack drive mechanisms are provided for positioning the control element assemblies instead of rack and pinion drive mechanisms. The magnetic jack control element drive mechanism is completely sealed by a pressure boundary, eliminating the need for seals. Motion of the control element drive shaft is accomplished by sequencing five solenoid coils located around the pressure boundary. The control element drive mechanisms are described in Section 4.2.3.1.
Combustion Engineering, Inc., has supplied identical control element drive mechanisms on previous plants, including Maine Yankee (AEC Docket No. 50-309) and Calvert Cliffs Units 1 and 2 (AEC Docket No. 50-317 and 50-318).
b) Number of Control Element Assemblies and Drive Mechanisms There are 81 CEA in the reactor compared to 85 CEA shown in the PSAR design while the number of drive mechanisms has changed from 65 to 69. This results in an increase in the number of single CEA (45 to 57.
including 8 part-length CEA) and a reduction in the number of dual CEA (40 to 24), thereby providing greater flexibility for optimization of CEA programming and fuel management.
c) Burnable Poison Shims Burnable poison shims have been added to the fuel assemblies, replacing some fuel. These shims permit lowering of the initial boric acid concentration in the coolant. This provides additional assurance that the moderator temperature coefficient, at power at beginning of life, will not be positive.
1.3.2.5 Reactor Coolant System No significant changes have been made to the reactor coolant system design since issuance of the PSAR.
1.3.2.6 Engineered Safety Features No significant changes have been made to the safety injection system design since issuance of the PSAR.
Other than design code changes as discussed in Section 1.3.2.3, there have been no significant changes in the design of the containment, containment spray system, containment cooling system or shield building ventilation systems from that described in the PSAR.
Changes have been made in the final design of the containment hydrogen purge system from that described in the PSAR, primarily in the location 1.3-3
of the containment isolation valves and purge line piping. The PSAR (Amendment 4) hydrogen purge report showed the location of the isolation valves and penetrations inside the shield building amulus at the top of the containment vessel with the purge lines inside the annulus. The present arrangement has the isolation valves located outside the shield building in the HVAC equipment room in the reactor auxiliary building. The purge lines are inside the containment vessel and penetrate the containment vessel at elevation 43'. The present design offers improved system reliability due to greater accessibility of the isolation valving and the reduced length of exposed piping outside the containment. In addition two (2) hydrogen recombiners have been added inside containment. The final design of the containment hydrogen control system is discussed in Section 6.2.5.
1.3.2.7 Instrumentation and Control The PSAR stated that the actuating signals for the Shield Building Ventilation System (SBVS) and Containment Cooling System would be Containment Spray Actuation Signal (CSAS). The actuating signal for the SBVS was subsequently changed to Containment Isolation Signal (CIS) and the actuating signal for the containment cooling system was changed to Safety Injection Actuation Signal (SIAS). These signals offered greater diversity and were more directly related to the functions of the systems being actuated. The final design of the engineered safety feature actuation system is discussed in Section 7.3.
1.3.2.8 Electrical Systems The diesel generator sizing criteria stated in answer to Question 5.6, Amendment 6 of the PSAR stated that that the maximum post-accident load on the diesel generator could exceed the continuous rating of the unit. These criteria were subsequently changed to ensure that the maximum load was within the continuous rating. The final sizing criteria are discussed in Section 8.3.1.1.7.
The PSAR stated that the on-site diesel oil fuel storage capacity would be sufficient for thirty (30) days full load operation of one diesel generator. This capacity was subsequently reduced to a 8.0 day supply based on post-LOCA load profile. This time is sufficient to allow resupply in the event of an emergency condition requiring prolonged operation of the emergency diesel generators. The oil storage system design and capacity are discussed in Section 9.5.4.
1.3.2.9 Auxiliary Systems a) Circulating Water System The PSAR stated that cooling water for normal operation and emergency plant conditions would be taken from the Atlantic Ocean through a diked area of Big Mud Creek. This area was to have served as the ultimate heat sink in providing a source of emergency cooling water for a long term residual heat removal following a LOCA. This design was subsequently changed due to the restriction in access to and use of Big Mud Creek which would be caused by erection of a dike at the Indian River inlet to Big Mud Creek. The present circulating water system design provides for 1.3-4
a canal to take water directly from the Atlantic Ocean through intake pipes under the beach. In the event or failure of the intake canal or pipes, emergency cooling water will be taken from Big Mud Creek through redundant emergency cooling water pipes extending from the area in front of the intake structure to Big Mud Creek. The final design of the circulating water system and ultimate heat sink are discussed in Section 9.2.7.
b) ECCS Area Ventilation System The PSAR contains no requirements for filtering of leakage from engineered safety feature components which contain recirculating containment sump water following a LOCA. The ECCS area ventilation system was subsequently included in the plant design to provide for control and filtration of airborne radioactivity from area of the reactor auxiliary building containing such components. The system design is discussed in Section 9.4.
1.3.2.10 Steam and Power Conversion System There are no significant changes in the final design of the steam and power conversion system from that described in the PSAR.
1.3.2.11 Radioactive Waste Management There are no significant changes in the final design of the radioactive waste management system from that described in the PSAR.
1.3.2.12 Radiation Protection There are no significant changes in plant design concerning shielding and radiation protection from that described in the PSAR.
1.3.2.13 Conduct of Operations There are no significant changes in conduct of operations affecting plant design from that described in the PSAR.
1.3.2.14 Initial Tests and Operations There are no significant changes in initial tests and operations affecting plant design from that described in the PSAR.
1.3.2.15 Accident Analyses The only change in accident analyses since the submittal of the PSAR is the more extensive refinement of computer codes, analytical investigations and tests to demonstrate compliance with the AEC acceptance criteria for emergency core cooling systems for light water cooled nuclear power reactors.
The off-site accident doses present in the PSAR were calculated using the atmospheric dispersion factors (/Q) based on preliminary meteorological data. The dose analyses presented in Chapter 15 are based on 1.3-5
/Q values obtained from the onsite meteorological monitoring program described in Section 2.3. The onsite data demonstrates the conservatism of the values used in the PSAR accident analyses.
1.3.2.16 Technical Specifications Subject coverage is in accordance with NRC standard technical specifications format/content as revised for the St. Lucie Plant.
1.3.2.17 Quality Assurance An independent Quality Assurance Department has been organized by Florida Power & Light Company to review and audit the activities of the Engineering and the Power Resources Departments.
There are no significant changes in the Ebasco Quality Assurance Program from that described in the PSAR.
1.3-6 Rev. 51 - 10/7/75
TABLE 1.3-1 PLANT PARAMETER COMPARISON ST. LUCIE REFERENCE CALVERT CLIFFS* PALISADES* TURKEY POINT*
HYDRAULIC AND THERMAL DESIGN PARAMETERS Unit 1 Section Units 1 and 2 Unit 1 Units 3 and 4 Rated Core Heat Output, Mwt 2560 (2440)*** 4.4 2560 2200 2200 Rated Core Heat Output, Btu/hr 8737xl06 (8330x106) 4.4 8740x106 7509x106 7479x106 Heat Generated in Fuel, % 97.5 4.4 97.5 97.5 97.4 Maximum Overpower, % 12 4.4 12 12 12 System Pressure, Nominal, psia 2250 4.4 2250 2100 2250 System Pressure, Minimum Steady State, psia 2200 4.4 2200 2050 2220 Hot Channel Factors, Overall Heat Flux, Fq 2.85 (3.10) 3.00 3.8 3.23 Enthalpy Rise, FH 2.02 (2.14) 4.4 1.65 2.51 1.77 DNB Ratio at Nominal Conditions 2.30 (2.15) 4.4 2.18 2.00 1.81 Coolant Flow Total Flow Rate, lb/hr 122x106 4.4 122x106 125x106 101.5x106 Effective Flow Rate for Heat Transfer, lb/br 117.5x106 6 (119.5x10 ) 4.4 117.5x106 121.25x106 97.0x106 Effective Flow Area for Heat Transfer, ft2 53. 5 4.4 53.5 58.7 41.8 Average Velocity along Fuel Rods, ft/sec 13.6 (13.8) 4.4 13.6 12.7 14.8 Average Mass Velocity, lb/hr-ft2 2.20x106 (2.23x106) 4.4 2.20x106 2.07x106 2.32x106 Coolant Temperatures °F Nominal Inlet 538.9 (550) 4.4 543.4 545 546.2 Design Inlet 544 4.4 548 548 550.2 Average Rise in Vessel, °F 55 (52) 4.4 52 46 55.9 Average Rise in Core, °F 56 (53) 4.4 54 47 58.3 Average in Core, °F 572 (576.5) 4.4 570.4 568.5 575.4 Average in Vessel 571.5 (576) 4.4 569.5 568 574.2 Nominal Outlet of Hot Channel 640 (652) 4.4 643 642.8 642 Average Film Coefficient, Btu/hr-ft2-F 5300 (5100) 4.4 5240 4860 5400 Average Film Temperature Difference, °F 35 (32) 4.4 33.5 30 31.8 Heat Transfer at 100% Power Active Heat Transfer Surface Area, ft2 48,400 (50,240) 4.4 48,416 51,400 42,460 Average Heat Flux, Btu/hr-ft2 176,000 (162,000) 4.4 176,000 142,400 171,600 Maximum Heat Flux, Btu/hr-ft2 501,300 (501,000) 4.4 527,900 541,200 554,200 Average Thermal Output, kw/ft 5.94 (5.6) 4.4 5.94 4.63 5.5 Maximum Thermal Output, kw/ft 17** (17.4**) 4.4 18** 17.6** 17.9 Maximum Clad Surface Temperature at Nominal 657 (658) 4.4 657 648 657 Pressure, °F Fuel Center Temperature, °F Maximum at 100% Power 3890 (3775) 4.4 4170 4040 4030 Maximum at Overpower 4060 (4155) 4.4 4460 4350 4300 Thermal Output, kw/ft at Maximum Overpower 19.1** (19.4**) 4.4 20** 19.7 20.0**
- The values listed for these plants were taken from public documentation.
- Based on total heat output of the core rather than heat generated in the fuel alone.
- PSAR Data 1.3-7
TABLE 1.3-1 (Contd)
ST. LUCIE Reference CALVERT CLIFFS PALISADES TURKEY POINT CORE MECHANICAL DESIGN PARAMETERS Unit 1 Section Units 1 and 2 Unit 1 Units 3 and 4 Fuel Assemblies Design CEA 4.2 CEA Cruciform RCC Rod Pitch, in. 0.58 4.2 0.58 0.550 0.563 Cross-Section Dimensions, in. 7.98 x 7.98 4.2 7.98 x 7.98 8.1135 x 8.1135 8.426 x 8.426 Fuel Weight (as UO2), pounds 207,200 (217,600) 4.2 207,269 210,524 176,200 Total Weight, pounds 271,280 (284,000) 4.2 282,570 295,800 226,200 Number of Grids per Assembly 8 8 8 7 Fuel Rods Number 36,896 (38,192) 4.2 36,896 43,168 32,028 Outside Diameter, in. 0.44 4.2 0.44 0.4135 0.422 Diametral Gap, in. 0.0085 (0.0065) 4.2 0.0085 0.0065 0.0065 Clad Thickness, in. 0.026 4.2 0.026 0.022 0.0243 Clad Material Zircaloy or M5 4.2 Zircaloy or M5 Zircaloy or M5 Zircaloy Fuel Pellets Material UO2 Sintered 4.2 UO2 Sintered UO2 Sintered UO2 Sintered Diameter, in. 0.3795 (0.3815) 4.2 0.3795 0.359 0.367 Length, in. 0.650 (0.600) 4.2 0.650 0.600 0.600 Control Assemblies Cd-In-Ag (5 Cd-In-Ag (5 Neutron Absorber B4C/SS (B4C) 4.2 B4C 80%) Cruciform 80%)
304 SS Tubes, 304 SS - Cold Cladding Material NiCrFe Alloy 4.2 NiCrFe Alloy E.B. welded to Worked 13.5 in. span Clad Thickness 0.040 4.2 0.040 0.016 0.019 Number of Assembly, full/part length 78/8 (85) 4.2 77/8 41/4 53 117 Tubes per Number of Rods per Assembly 5 4.2 5 20 Rod Core Structure Core Barrel I.D./O.D., in. 148/151.5 (148/151) 148/149.75 149.75/152.5 133.875/137.875 Thermal Shield I.D./O.D., in. 156.75/162.75 (156/162) None None 142.625/148.0 NUCLEAR DESIGN DATA Structural Characteristics Core Diameter, inches (Equivalent) 136 4.2 136 136.71 119.5 Core Height, inches (Active Fuel) 136.7 (137) 4.2 136.7 132 144 Reflector Thickness and Composition Top - Water plus steel 10 10 10 10 Bottom - Water plus steel 10 10 10 10 Side - Water plus steel 15 15 15 15 H2O/U, Unit Cell (Cold) 1.63 (3.35) 4.3 3.44 3.50 4.18 Number of Fuel Assemblies 217 4.2 217 204 157 UO2 Rods per Assembly, unshimmed/shimmed 176/164 (176) 4.2 176/164 212/208 204 Batch A 176 4.2 -- -- 176 Batch B 164 4.2 -- -- 164 Batch C (176/164/164) 4.2 -- -- (176/164/164)
Performance Characteristics 3 Batch Mixed 3 Batch Mixed 3 Batch Mixed 3 Regions Loading Technique Central Zone (3 Batch) 4.3 Central Zone Central Zone Central Zone Fuel Discharge Burnup, MWD/MTU Average First Cycle 12,800 (11,900) 4.3 13,775 10,180 13,000 First Core Average 22,000 22,550 17,600 24,500 UNIT 1 1.3-8 Amendment No. 27 (04/15)
TABLE 1.3-1 (Contd)
ST. LUCIE Reference CALVERT CLIFFS PALISADES TURKEY POINT NUCLEAR DESIGN DATA (Continued) Unit 1 Section Units 1 and 2 Unit 1 Units 3 and 4 Feed Enrichment w/o Region 1 1.93 (1.80) 4.3 2.09 1.65 1.85 Region 2 2.33 (2.48) 4.3 2.51 2.08/2.54 2.55 Region 3 2.82 (2.93) 4.3 2.99 2.54/3.20 3.10 Control Characteristics Effective Multiplication (beginning of life)
Cold, No Power, Clean 1.170 (1.288) 4.3 1.169 1.212 1.180 Hot, No Power, Clean 1.134 (1.229) 4.3 1.129 1.175 1.38 Hot, Full Power, Xe Equilibrium 1.078 (1.153) 4.3 1.081 1.111 1.077 Control Assemblies Material B4C/SS (B4C) B4C Cd-In-Ag Cd-In-Ag (5-15-80%) (5-15-80%)
Number of Control Assemblies 81 (85) 4.3 85 45 Cruciform 53 117 Tubes Number of Absorber Rods per RCC (or CEA) 5 5 Welded to Form 20 Assembly 13.5 in. span Total Rod Worth (Hot), % 11.0 (9.0) 4.3 11.0 8.6 7 Boron Concentrations for Criticality:
Zero Power no rods inserted, clean, ppm Cold/Hot 945/935 (1400/1600) 4.3 985/991 1180/1210 1250/1210 At Power, with no Rods inserted, clean/equilibrium 820/590 (1400/1100) 4.3 885/650 1070/830 1000/670 xenon, ppm Kinetic Characteristics, Range Over Life Moderator Temperature Coefficient -0.4x10-4 to (0 to ) 4.3 -0.32 x 10-4 to -0.08 x 10-4 to -0.32 x 10-4 to p/oF -2.1x10-4 (-2x10-4) -1.96 x 10-4 -2.25 x 10-4 -1.96 x 10-4 Moderator Pressure Coefficient +0.49x10-6 to (0 to ) 4.3 +0.65 x 10-6 +0.10 x 10-6 -1.0 x 10-6 to p/psi +2.55x10-6 (+2x10-6) +2.39 x 10-6 +1.7 x 10-6 +3.0 x 10-6 to
+0.65 x 10-6 to
+2.39 x 10-6 Moderator Void Coefficient -0.26x10-3 to (0 to ) 4.3 -0.41 x 10-3 to -0.06 x 10-3 to +0.5 x 10-3 to p/% Void -1.35x10-3 (11.6x10-3) -1.43 x 10-3 -1.0 x 10-3 -2.0 x 10-3 to
-0.41 x 10-3 to
-1.43 x 10-3 Doppler Coefficient * -1.45 x 10-5 to (-1x10-5 to) -1.46 x 10-5 to -1.56 x 10-5 to -1.0 x 10-5 to p/oF -1.07 x 10-5 (-1.8x10-5) -1.02 x 10-5 -1.08 x 10-5 -2.0 x 10-5 to
-146 x 10-5
-1.02 x 10-5 1.3-9
TABLE 1.3-1 (Contd)
REACTOR COOLANT SYSTEM - CODE ST. LUCIE Reference CALVERT CLIFFS PALISADES TURKEY POINT*
REQUIREMENTS Unit 1 Section Units 1 and 2 Unit 1 Units 3 and 4 Component Reactor Vessel ASME III class A 5.2 ASME III class A ASME III class A ASME III class A Steam Generator Tube Side ASME III class A 5.2 ASME III class A ASME III class A ASME III class A Shell Side ASME III class A 5.2 ASME III class A ASME III class A ASME III class A Pressurizer ASME III class A 5.2 ASME III class A ASME III class A ASME III class A Pressurizer Relief (or Quench) Tank ASME III class C 5.2 ASME III class C ASME III class C ASME III class C Pressurizer Safety Valves ASME III 5.2 ASME III ASME III ASME III EC Reactor Coolant Piping USAS B 31.7 (USAS B 31.1) 5.2 USAS B 31.7 USAS B 31.1 USAS B 31.1 287 661 PRINCIPAL DESIGN PARAMETERS FOR THE COOLANT SYSTEM Operating Pressure, psig 2235 5.1 2235 2085 2235 Reactor Inlet Temperature, oF 539.7 (549) 5.1 544.5 545 546.2 Reactor Outlet Temperature, oF 595.1 (601) 5.1 599.4 591.1 602.1 Number of Loops 2 5.1 2 2 3 Design Pressure, psig 2485 5.1 2484 2485 2485 Design Temperature, oF 650 5.1 650 650 650 Hydrostatic Test Pressure (cold), psig 3110 3110 3110 3110 Total Coolant Volume, cu. Ft 11,101 11,101 10,809 9088 PRINCIPAL DESIGN PARAMETERS of THE REACTOR VESSEL SA-302, Grade SA-533, Grade B, SA-302, Grade B, B, Class I, low
austenitic SS.
Design Pressure, psig 2485 5.4 2485 2485 2485 Design Temperature, oF 650 5.4 650 650 650 Operating Pressure, psig 2235 2235 2085 2235 Inside Diameter of Shell, in. 172 5.4 172 172 172 Outside Diameter across Nozzles, in. 253 253 254 236 Overall Height of Vessel and Enclosure Head, ft. - in. to 41-11-3/4 (42 1) 5.4 41-11-3/4 40-1-13/16 41-6 top of CRD Nozzle Minimum clad thickness, in. 5/16 5.4 1/8 3/16 5/32 PRINCIPAL DESIGN PARAMETERS OF THE STEAM GENERATORS Number of Units 2 5.5 2 2 3 Vertical U-Tube Vertical U-Tube with Vertical U-Tube Vertical U-Tube with integral moisture with integral Type 5.5 integral moisture with integral separator moisture separator moisture separator separator
- Operating Level
- Replacement RVCH is a one piece forging SA-508 Class 3 low alloy steel, internally clad with Type 308L/309L austentic stainless steel.
1.3-10 Amendment No. 30 (05/20)
TABLE 1.3-1 (Contd)
PRINCIPAL DESIGN PARAMETERS OF THE STEAM ST. LUCIE Reference CALVERT CLIFFS PALISADES TURKEY POINT GENERATORS (Continued) Unit 1 Section Units 1 and 2 Unit 1 Units 3 and 4 Tube Material NiCrFe Alloy 5.5 Inconel Inconel Inconel Shell Material SA-533 Gr. B. SA-533 Gr.B Class 1 Carbon Steel SA-302 Class 1 and and SA 1516 gr70 SA1516 gr 70 Tube Side Design Pressure, psig 2485 5.5 2485 2485 2485 Tube Side Design Temperature, oF 650 5.5 650 650 650 Tube Side Design Flow, lb/hr 61 x 106 5.5 61 x 106 62.5 x 106 33.93 x 106 Shell Side Design Pressure, psig 985 5.5 985 985 1085 Shell Side Design Temperature, oF 550 5.5 550 550 556 Operating Pressure, Tube Side, Nominal, psig 2235 5.5 2235 2085 2235 Operating Pressure, Shell Side, Maximum, psig 885 5.5 885 885 1020 Maximum Moisture at Outlet at Full Load, % 0.2 5.5 0.2 0.2 1/4 Hydrostatic Test Pressure, Tube Side (cold), psig 3110 3110 3110 3107 Steam Pressure, psia, at full power 815 5.5 850 770 745 Steam Temperature, oF, at full power 520.3 5.5 525.2 513.8 510 PRINCIPAL DESIGN PARAMETERS OF THE REACTOR COOLANT PUMPS Number of Units 4 5.5 4 4 3 Type Vertical, single 5.5 Vertical, single stage Vertical, single Vertical, single stage centrifugal centrifugal with stage radial flow stage radial flow with bottom bottom suction and with bottom with bottom suction and horizontal discharge suction and suction and horizontal horizontal horizontal discharge discharge discharge Design Pressure, psig 2485 5.5 2485 2485 2485 Design Temperature, oF 650 5.5 650 650 650 Operating Pressure, nominal psig 2235 5.5 2235 2085 2235 Suction Temperature, oF 540 (550) 5.5 543.4 545 546.5 Design Capacity, gpm 80,000 (81,200) 5.5 81,200 83,000 89,500 Design Head, ft. 250 (290) 5.5 300 260 260 Hydrostatic Test Pressure, (cold), psig 3110 3110 3110 3107 Motor Type A-C Induction A-C Induction A-C Induction A-C Induction Single Speed Single Speed Single Speed Single Speed Motor Rating, hp 6500 7200 6250 6000 PRINCIPAL DESIGN PARAMETERS OF THE REACTOR COOLANT PIPING SA516 - gr 70 SA516 - gr 70 with SA212B clad Material with nominal nominal 7/32 SS Austenitic SS with SS 7/32 SS clad clad Hot Leg - I.D., in. 42 5.5 42 42 29 Cold Leg - I.D., in. 30 5.5 30 30 27 1/2 Between Pump and Steam Generator - I.D., in. 30 30 30 31 1.3-11
TABLE 1.3-1 (Contd)
ST. LUCIE Reference CALVERT CLIFFS PALISADES TURKEY POINT CONTAINMENT SYSTEM PARAMETERS Unit 1 Section Units 1 and 2 Unit 1 Units 3 and 4 Type Steel containment vessel with cylindrical shell, Steel-lined prestressed post Steel-lined prestressed Steel-lined hemispherical dome and ellipsoidal bottom - tensioned concrete cylinder, post tensioned concrete prestressed post ASME Code,Section III, Class B, surrounded curve dome roof. cylinder, curved dome tensioned concrete by reinforced concrete shield building roof. cylinder, shallow domed roof.
Design Parameters - Containment Inside Diameter, ft. 140 130 116 116 Height, ft. 232 181 2/3 190 1/2 169 Free volume, ft3 2,500,000 2,000,000 1,640,000 1,550,000 Reference accident Pressure, psig 44 (max. allow) (40) 50 55 59 Steel Thickness, in. Not Applicable Not Applicable Not Applicable Vertical Wall 1.91 (2)
Hemispherical Head 0.95 (1)
Knuckles 21/4 Concrete Thickness, ft. Not Applicable Vertical Wall 33/4 3 33/4 Dome 31/4 21/2 31/4 Design Parameters - Shield Building Not applicable Not applicable Not applicable Inside Diameter, ft. 148 Height, ft (top of foundation to top of dome) 230.5 (241)
Concrete Thickness, ft.
Vertical Wall 3 Dome 21/2 Leak-tight penetration and Leak-tight penetration Leak-tight penetration continuous steel liner. and continuous steel and continuous steel Leak-tight penetration. Automatic isolation Containment Leak Prevention and Mitigation Systems Automatic isolation where liner. Automatic liner. Automatic where required.
required. The exhaust from isolation where isolation where penetration rooms to vent. required. required.
Gaseous Effluent Purge Discharge through vent. Discharge through vent. Discharge through Through particulate vent. filter and monitors part of main exhaust system.
Engineered Safety Features 6.3 Safety Injection System No. of High Head Pumps 3 3 3 4 (shared)
No. of Low Head Pumps 2 2 2 2 Containment Fan Coolers 6.2.2 No. of Units 4 4 4 3 Air Flow Capy, each, at emergency condition, cfm 55,800 55,000 30,000 65,000 Post Accident Filters No. of Units None None None None Type None None None None Containment Spray 6.2.2 No. of Pumps 2 2 2 2 Emergency Power 2 total for both Diesel Generator Units 2 8.2 3, total for both units 2 units Safety Injection Tanks - Nuclear 4 6.3 4 4 3 1.3-12
TABLE 1.3-1 (Contd)
Reference TURKEY ST. LUCIE CALVERT CLIFFS PALISADES Section POINT*
Unit 1 Units 1 and 2 Unit 1 RADIOACTIVE WASTE MANAGEMENT SYSTEM Units 3 and 4 Liquid Waste Processing Systems Reactor Coolant Waste Receiver Tnak 11.2.2 Number 4 2 4 1 Capacity (gal.), each 40,000 90,000 50,000 34,300 Degasifier 11.2.2 Number 2 1 2 Capacity (gpm) 0-120 0-160 0-25 Waste Evaporators 11.2.2 Number 1 3 Not Applicable 1 Capacity (gpm) 2 20 3 Gaseous Waste Processing Systems Waste Gas Decay Tank 11.3.2 Number 3 3 3 6 Capacity (ft3), each 144 610 100 525 Pressure 190 150 120 150 Held up Time (days) 30 60 30 45 INSTRUMENTATION SYSTEMS*
Reactor Protective System 7.2 7.2 7.2 7.2 Reactor and Reactor Coolant Control System 7.7.1.1, 7.7.1.2 7.4 7.3 7.2, 7.3 Steam and Feedwater Control System 7.7.1.3 7.4 7.3 9.11, 10.2.2 Nuclear Instrumentation 7.2.1.1 7.5-2, 7.5.4 7.4.1, 7.4.2 7.4, 7.6 Non-Nuclear Process Instrumentation 7.5.1.5 7.5.1 7.4.1, 7.4.2 7.5 CEA Position Instrumentation 7.5.1.3 7.5.3 7.4.1, 7.4.2 7.3.2
- This section is not suited for tabular description. SAR section numbers have been included for the location of the detailed description of each system.
1.3-13
TABLE 1.3-1 (Contd)
ST. LUCIE Reference CALVERT CLIFFS PALISADES TURKEY POINT ELECTRIC SYSTEMS Unit 1 Section Units 1 and 2 Unit 1 Units 3 and 4 Number of Offsite Circuits 3 - 240 KV 8.2.1.1 2 - 500 KV 4 - 345 KV 7 - 240 KV Number of Incoming Lines to Startup Transformers 2 8.2.1.3 2 1 2 Number of Startup Transformers 2 8.2.1.3 2 2(1) 2 Number of Unit Transformers 2 Fig. 8.3-1 4 2 2 Number of 4.16 KV Engineered Safety Features System 3(2) 8.3.1.1.3 4 2 - 2300V 4 Buses Number of 480V Engineered Safety Features System 3(2) 8.3.1.1.4 8 2 8 Buses Number of 120V Vital AC Buses 4 8.3.1.1.5 8 4 8 Number of 125V DC Engineered Safety Features System 3(2) 8.3.2 2 2 2 Buses Number of Standby Diesel Generators 2 8.3.1.1.7 3 2 2 Diesel Generator Continuous Rating (KW) 3500 8.3.1.1.7 2500 2500 2750 (1) Only one is used for engineered safety features systems.
(2) One bus will swing from other two sources; can be connected to only one bus at anytime.
1.3-14
1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS Information contained herein was valid at the time the operating license for St. Lucie Unit 1 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 Hutchinson Island Nuclear Power Unit 1. Florida Power & Light Company is responsible for the design, engineering review, construction and operation of the plant.
Florida Power & Light Company has engaged Combustion Engineering, Inc. to design, manufacture and deliver the Nuclear Steam Supply System and nuclear fuel for the first core and three reload batches to the site. 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, C-E will furnish technical assistance for erection, initial fuel loading, testing and initial startup of the Nuclear Steam Supply Systems.
Ebasco Services Incorporated has been engaged by the Owners as the Engineer-Constructor 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 C-E scope of supply. Ebasco has been engaged to perform onsite construction of the entire plant with technical advice for installation of the reactor components to be provided by C-E.
Following issuance of the facility operating license, Ebasco Services Incorporated was retained as Engineer-Constructor for St. Lucie Unit 1 plant backfit, retrofit and maintenance activities, under the direction of Florida Power & Light Company.
Presently, Bechtel Power Corporation has been engaged to perform site backfit construction activities under the direction of Florida Power & Light Company.
1.4-1 Amendment 15, 1/97
1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION Information contained herein was valid at the time the operating license for St. Lucie Unit 1 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 letter from Dr. P. A. Morris to G. Kinsman of FP&L dated September 8, 1970 requested further information on several items relating to design and construction which were unresolved at the time of issuance of the Construction Permit. The status of these items is as follows:
Item 1: The dynamic analysis to be performed in design of Class I equipment. An outline of the scope of the information we will need is given on pages 81-83 of our Hutchinson Island Safety Evaluation dated April 13, 1970 (pages 8-10 of the report by our seismic design consultant, John A. Blume &
Associates, Engineers).
Status: The information requested was submitted with PSAR Amendment 11 (11/27/71) and is also contained in Section 3.7.
Item 2: The design of the ventilation system to be added in the vicinity of ECCS lines and pumps outside the containment so that leakage from ECCS equipment which must function for long time periods following an accident would be released to the environment via a charcoal filtration system thus limiting offsite doses.
Status: The design of the ECCS area ventilation system is discussed in Section 9.4.3.
Item 3: The design of a charcoal filtration system or equivalent for the fuel handling area, so that in the event of a fuel handling accident, the doses calculated would be well within the guidelines of 10CFR Part 100.
Status: The need for a fuel handling building charcoal filtration system to reduce offsite doses to within the guidelines of 10 CFR 100 is discussed in Section 15.4.3. An elemental iodine charcoal filter will be installed as described in Section 9.4.6.
Item 4: The results of the design analysis of the hot penetration guard pipe performed by an independent consultant.
Status: An independent analysis of the hot penetration guard pipe design has been performed as described in Section 3.8.2.
Item 5: The criteria and design details relative to physical separation of redundant elements of your instrumentation and control system which are essential to safety.
Status: The information requested is presented in Sections 7.2 and 7.3.
Item 6: The Preoperational Vibration Testing Program and In-Service Vibration Monitoring Program for the primary system.
Status: The status of the Preoperational Vibration Monitoring Program for St. Lucie Plant Unit 1 is detailed in Section 3.9.1.3.
1.5-1
Item 7: The final loads and ratings of the emergency diesel generators so that we can be assured that the ratings comply with our current criteria.
Status: The information requested was submitted with PSAR Amendment 10 (4/26/71) and is also contained in Section 8.3.1.
Item 8: The analysis of the relative humidity in the shield building annulus after accidents to assure that it would be reduced below 70 percent in a short time and thus assure an acceptable efficiency of the shield building annulus filter system.
Status: The analysis of annulus relative humidity is presented in Section 6.2.1.
Item 9: The design of equipment which would be the primary means to control hydrogen buildup in the containment following a loss-of-coolant accident in addition to the purging system which is presently proposed.
Status: The design of the containment hydrogen control system is discussed in Section 6.2.5.
Item 10: The results of studies of means to prevent common failure modes from negating scram action.
Status: In conformance with the requirements of WASH-1270 "Technical Report on Anticipated Transients Without Scram", CE has analyzed the reactor protective system to identify areas that may be particularly vulnerable to common mode failures. Topical reports CENPD-149 "Review of Reactor Shutdown System (RPS Design) For Common Mode Failure Susceptibility" describes the common mode failure review of the reactor shutdown system.
Item 11: The results of studies of consequences of failure to scram during anticipated transients and design features which would make such failure tolerable.
1.5-2
Status: Combustion Engineering has analyzed the response of pressurized water reactors which are of the St Lucie type to demonstrate the response of the plant to anticipated transients without scram (ATWS) The Combustion Engineering report, entitled "Topical Report on Anticipated Transients Without Scram (Proprietary)" was submitted to the AEC on January 10, 1972 Evaluations are performed in this report based upon the incredible assumption that no CEA's are inserted into the core during the course of the following transients: CEA withdrawal CEA drop, idle loop start-up, loss of flow, boron dilution, excess load, loss of load, loss of feedwater, sample line break, and pressurizer safety valve failure. The transient resulting from loss of normal on-site and off-site power is also analyzed, but with a conservative one percent negative reactivity insertion assumed following reactor trip signal generation, since for this case the failures which initiate the transient would also remove power from the CEDM, allowing the CEA to insert. Applicant's letter of 3/31/75 to the NRC refer-enced CENPD-158, 'Anticipated Transients Without Reactor Trip," December, 1974, as the analysis applicable to St. Lucie Unit 1 except where qualified in the 3/31/75 letter. The generic solutions to NRC concerns for ATWS will be implemented on St. Lucie Unit 1. A schedule for implementation will be provided to the NRC Staff following their review and acceptance of the CE report.
Item 12 The results of studies of the consequences of secondary system accidents particularly relative to fuel clad damage during a steamline break accident and to multiple steam generator tube failures Status: The results of the studies of the consequences of secondary system accidents is presented in Sections 15.4.3 and 15.4.5.
Item 13 The analysis of the consequences to turbine failure including (1) the ability of the structures and systems which are important to safety to withstand the effects of turbine missiles inview of the results of the recent turbine missile calculations and (2) radiological consequences of a turbine missile entering the fuel storage pool and damaging fuel Status: The analysis of the potential turbine missiles is presented in Section 3.5 Item 14: The status of the Research and Development items identified on Pages 45-50 of our Hutchinson Island Safety Evaluation dated April 13, 1970.
Status: The items referred to appear in two groups in the AEC Safety Evaluation: Section 15.1, Paragraphs (a) through (e), and Section 15.2, Paragraphs (a) through (i). A statement of each item follows:
1.5-3
15.1.a Fuel Assembly Flow Tests The Staff Safety Evaluation states: "Tests are being conducted on fuel assemblies (1) to verify flow mixing factors and (2) to establish the characteristics of fuel rods in axial flow".
1.5-4
Status: The Flow Mixing Tests have been completed and are described in Section 4.4.4.4 and Division 1, Section 1, of Reference 1.
The results of the hot flow tests at reactor flow conditions using full length partial cross section fuel assemblies are discussed in Section 4.4.3.6 and Division 1, Section 1, of Ref. 1.
15.1.b Mechanical Testing of Control Element Assemblies (CEA's)
The Staff Safety Evaluation states: "A Series of tests have been completed on single CEA's demonstrating functional feasibility of the CEA concept under all possible combinations of misalignments, dynamic loading, bowing and thermal effects."
Status: A series of tests have been completed on both single and dual CEA's in a cold water, low pressure facility to satisfy the following objectives:
a) Determine the mechanical and functional feasibility of the CEA-type control rod concept.
b) Experimentally determine the relationship between CEA drop time and CEA drop weight, annular clearance between CEA fingers and guide tubes and coolant flow rate within the guide tube.
c) Experimentally determine the relationship between flow rate and pressure drop within the guide tube as a function of CEA axial position and of finger-to-guide tube clearance.
d) Determine the effects on drop time of adding a flow restriction or of plugging the flow holes in the lower portion of a guide tube (as might occur under incident conditions).
e) Determine the effects of misalignment within the CEA guide tube system on drop time.
The results of these tests were used as the basis for selecting the final CEA and guide tube geometries. The tests have demonstrated that the five finger CEA concept is mechanically and functionally feasible and that the CEA design has met the criteria established for drop time under the most adverse conditions. The testing has also verified that the analytical model used for predicting the drop times gives uniformly conservative results.
The effects on drop time of all possible combinations of frictional restraining forces in the CEDM, angular and radial misalignment of the CEDM bowing of the guide tubes and misalignments of the CEA have been experimentally investigated and defined. The conditions tested simulated all the effects of tolerance buildup, dynamic loadings and thermal effects. The tests demonstrated that misalignments and distortions in excess of those expected from tolerance buildup or any other anticipated cause would still result in acceptable drop times.
1.5-5
15.1.c Performance of Control Element Drive Mechanisms (CEDMs)
The Staff Safety Evaluation states: "A prototype CEDM was subjected to accelerated life testing under cold flow conditions. The program demonstrated that the mechanism met the established specifications and requirements. A long term program for testing the CEDM at operating temperature and pressure is in progress."
Status: The St.Lucie plants use magnetic jack control element drives, rather than the rack and pinion drives described in the Unit 1 PSAR. The Combustion Engineering magnetic jack together with the associated research and development program has been described in detail in Section and Division 1, Section 5, of Reference 1.
15.1.d Reactor Vessel Flow Investigation Program The Staff Safety Evaluation states: "Model tests will be conducted to determine the hydraulic performance for normal operation and for operation with one or more inactive pumps."
Status: The results of the reactor flow model tests are discussed in Section 4.4.3.1 ; Division 1, Section 2, of Ref. 1; and Sec. 2 of Ref. 2.
15.1.e In-Core Instrumentation The Staff Safety Evaluation states: "Hot and cold functional tests on in-core thermocouples and in-core flux detectors will be performed to detect vibration, wear or fretting."
Status: Tests on in-core thermocouples and flux detectors have been performed to ensure that the instrumentation will perform as expected at the temperatures to be encountered and that it does not vibrate excessively and cause excessive wear or fretting. Cold flow testing has been completed on a similar detector cable; no adverse vibrations or wear effects were encountered. Hot flow testing is also complete. After 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> at 590 F and 2100 psig in a test loop, no breach of mechanical integrity was observed. The tests are described in Division 1 Section 4, of Reference 1 and in Section 6 of Reference 2.
Mechanical tests of the insertion and removal equipment and instrumentation were performed on thimbles of the same approximate configuration as those used on St. Lucie Unit 1. The top entry in-core instrumentation design provides a means of eliminating the need of handling instrument assemblies separately, to accommodate three in-core instrumentation thimble assemblies. Major components and subassemblies of the mockup included:
a) An in-core instrumentation test assembly, including the upper guide structure support plate, three thimble guide sleeves, fuel alignment plate, three fuel bundle guide tubes and the core support plate.
1.5-6
b) A thimble assembly consisting of the instrument plate, three in-core instrumentation thimbles and the lifting sling.
c) An upper guide tube with the guide tube attached to the thimble extension and the detector cable partially inserted in the guide tube.
Insertion and withdrawal tests were performed to determine the frictional forces of a multitude instrument thimble assembly during insertion and withdrawal from a set of fuel bundles. This test simulated the operation that will be performed during the refueling of the reactor. To determine whether jamming of the thimbles would occur during this operation, bending loads were applied to the thimble assembly by tilting the instrument plate in 0.5 degree increments up to a total of five degrees horizontal. Guide tubes were filled with water. The assembly was raised and lowered approximately five times for each tilt setting. Results showed no discernible difference in the friction forces for the various tilt settings. The tests demonstrated that the repeated insertions and withdrawal of in-core instrumentation thimble assemblies into the fuel bundle guides can be accomplished with reasonable insertion forces.
Life cycle tests were performed to determine if the frictional forces increase as a result of 40 insertions and withdrawals. An automatic time was installed in the crane electrical circuitry to automatically cycle the thimble assembly between the fully inserted and withdrawn position. The instrument plate was set for five degrees tilt and the assembly was cycled 60 times. The insertion and withdrawal forces were measured during the first and last five cycles. No discernible difference was noticed.
An off-center lift test was performed to determine if the thimble assembly could be withdrawn from the core region while lifting the assembly from an extreme off-center position. For a lifting point 11 inches off-center, insertion was accomplished without incident. The flexibility of the thimble is such that jamming of the assembly due to off-center lifting does not occur.
Cable insertion tests were performed to determine the forces required to completely insert and withdraw a detector cable from the in-core instrumentation thimble assembly. The guide tube routing included typical bends equal to, or worse than those found in the reactor. The detector cable was passed through the guide tubing and into a thimble. In all cases, the insertion and withdrawal forces were reasonable for hand insertion.
15.2.a Effect of Fuel Rod Failure on ECCS Performance 1.5-7
Status: C-E has conducted experimental and analytical investigations of fuel rod failures under simulated LOCA conditions. The analytical work provides indications of the actual conditions to be expected in the core during a transient, in terms of potential clad heating rates, internal pressures and transient duration. The experimental work, described in Section 16 of Reference 2, applies these parameters in various combinations to establish the nature of fuel rod deformation which might occur under accident conditions. This subject has been covered comprehensively in the Statement of Affirmative Testimony and Evidence of Combustion Engineering in the Matter of Rulemaking Hearing for the Acceptance Criteria for Emergency Core Cooling Systems for Light-Water-Cooled Nuclear Power Reactors, Docket No. RM-50-1. Further details are given in Section 15.4.1.2.
15.2.b Effects of Fuel Bundle Flow Blockage Status: That portion of the program responding to the ACRS concern related to flow blockage during operation has now been completed and the results are summarized in Section 4.4.3.6 and in ASME paper 68-WA/HT-34 presented at the December 1968 Winter Annual Meeting.
15.2.c Verification of Fuel Damage Limit Criterion Status: The basis for C-E design is described in Section 4.2.1. The staff notes (Page 46) that-C-E does not have an experimental program directed towards establishing these limits. Since the Staff's Safety Evaluation Report was issued, CE has conducted a series of fuel irradiation tests in the Battelle Research, MZFR, GETR and Halden Reactors to determine the densification characteristics of CE fabricated fuel. The results of the tests conducted through 1973 are summarized in Section 12 of Reference 2.
15.2.d Effects of Blowdown Forces on Primary System Components Status: The dynamic response of reactor internals resulting from hydro-dynamic blowdown forces under a postulated LOCA condition is discussed in a proprietary Combustion Engineering Topical Report, CENPD-42 which was submitted on August 1, 1972. This report contains a complete description of the theoretical basis for methods of analysis for the various reactor components, as well as documentation of computer program and the respective analytical structural models.
Reactor vessel internal structures will be analyzed to ensure the required structural integrity during abnormal operating conditions, including the effects of blowdown, pressure drop and buckling forces.
For the LOCA, the CEFLASH-4 computer program is used to define the flow transient and the WATERHAMMER program determines the corresponding dynamic pressure load distribution. The dynamic response of the reactor vessel internals to the space and the time-dependent pressure loads will be obtained through the use of a number of structural 1.5-8
dynamic analysis codes. Lateral and vertical dynamic response of the internals will be considered, as well as the transient response and dynamic buckling of a core support barrel in shell modes. Both the CEFLASH-4 and WATERHAMMER models are being evaluated against the LOFT program results.
1.5-9
The loads resulting from the LOCA will be added to the loads resulting from normal operation and the design basis earthquake for each critical component and the component deflections and stresses analyzed to ensure compliance with the criteria specified in Section 3.9.1.
15.2.e ECCS Thermal Effect on Rods Status: ACRS has asked that information be developed to show that the "--melting and subsequent disintegration of a portion of fuel assembly-- will not lead to unacceptable conditions." They refer specifically to the "--effects in terms of fission product release, local high pressure production and the possible initiation of failure in adjacent fuel elements--".
Inquiry has been made as to whether that accident conditions might occur which cause clad temperatures to reach such high temperatures that embrittlement occurs and whether subsequent quenching operations will cause the embrittled portions to disintegrate and thereby prevent a sufficient flow of emergency core coolant to the remainder of the core.
Fuel damage of the magnitude indicated is prevented by the inherent nuclear and thermal characteristics of the UO2 core and by the provision of engineered safety features.
With regard to the nonexcursion mechanisms leading to the conditions described by ACRS, the following might be conjectured:
a) Fuel bundle inlet flow blockage during fuel power operation and subsequent overheating of the coolant starved fuel, or b) Loss of reactor coolant.
Condition (a), inlet flow blockage during fuel power operation and subsequent overheating and melting of the fuel, is not considered possible because open (non shrouded) fuel bundles are used, thereby providing cross flow to the flow starved channel even if some of the inlet holes were blocked.
Details and conclusions of the tests performed at C-E on the influence of inlet geometry on flow in the entrance region are presented in ASME paper 68-WA/HT-34 delivered at the December 1968 Winter Annual Meeting. Further analysis of these tests showed that if a group of four flow holes in the core support plate at the base of the fuel bundle were blocked, the subchannels above the blocked region would have an inlet velocity about 21 percent of the core average bulk inlet velocity. Because of cross flow from the surrounding nonblocked regions, the net effect of this flow shortage, using conservative calculations, is to increase the enthalpy rise of the blocked region by a maximum of 35 percent. At nominal conditions, the hot channel DNB ratio would drop from 2.0 to 1.4, assuming that the blockage occurred directly below the design hot channel.
1.5-10
Condition (b) has been covered comprehensively in the Statement of Affirmative Testimony and Evidence of Combustion Engineering in the matter of Rulemaking Hearing for the Acceptance Criteria for Emergency Core Cooling System for Light Water Cooled Nuclear Power Reactors, Docket No.
RM-50-1. The emergency core cooling system is designed to remove the decay heat from the core for the necessary period of time following a LOCA. Core power distributions and LOCA temperature-time histories indicate that for peak clad temperatures below 2300 F, the total clad oxidation will be significantly less than 1 percent.
15.2.f Verification of the Ability of the Fuel to Perform Under Transient Conditions at End of Life Status: The fuel cladding is designed to limit the transient stresses to two-thirds of the unirradiated value of the yield stress even during a depressurization transient near the end of life, when the internal gas pressure is highest, as stated in Section 4.2.1.
Experimental verification of the maximum linear heat generation rate employed in the St. Lucie design is discussed in Section 4.2.1. Numerous irradiation tests, which bracket the design of the fuel used have been performed, including those in the Westinghouse Test Reactor, the Shippingport blanket irradiations, the mixed oxide irradiations in the Saxton reactor, the zirconium clad UO2, fuel rod evaluations in the Vallecitos boiling water reactor, the large seed blanket reactor rod irradiations, the center melting irradiations in Big Rock, Peach Bottom 2 irradiations, and NRX irradiations (AECL-Canada). In these tests, fuel rods similar to those employed in the design of the core were successfully irradiated to fuel burnups varying from very short term tests up to 60,000 MWD/MTU and at linear heat rates ranging from 5.6 up to 27.0 KW/ft.
As noted in the Staff Safety Evaluation, no further work is planned.
15.2.g ECCS Thermal Effects on the Reactor Vessel Status: Sufficient emergency core cooling water is available to flood the core region in the event of a LOCA.
The St. Lucie design uses a section of each of the reactor coolant system cold legs to conduct the water from the safety injection system to the reactor vessel. This water then flows into the downcomer annulus and into the lower plenum of the reactor vessel before flooding the core.
Analytical investigations were performed to provide assurance that the resultant cooling of the irradiated inner surface of the thick-walled reactor vessel will not induce or propagate cracks sufficient to cause the reactor vessel to fail.
A detailed analysis of the reactor response to thermal shock was performed.
1.5-11
Further work was performed to refine the surface heat transfer coefficient and the brittle fracture model used in the evaluation. Details of these analyses are summarized in Section 6.3.3.5.
The results confirm the conservation of the approach used by C-E and verify that cracks in the vessel will not grow during the thermal shock transient associated with emergency core cooling operation.
As noted in the Staff Safety Evaluation, it has been concluded that reactor vessel integrity will not be jeopardized as a result of thermal shock induced by ECCS operation.
CE is also participating in an on-going research and development program to verify the integrity of the reactor vessel against thermal shock from ECCS.
15.2.h Failed Fuel Detection The process radiation monitor is described in Section 11.4.2.1.
Status: Early detection of the gross failure of fuel permits early application of action necessary to limit the consequences.
Based on a study of the expected fission and corrosion product activities in the reactor coolant, it was concluded that the gross gamma plus specific isotope monitor provides a simple and reliable means for early detection of fuel failures.
The design bases of the detection system include the following:
a) Trends in fission product activity in the reactor coolant system are used as an indication of fuel rod cladding failures. The minimum detectable activity is specified as 10-4 Ci/cc I-135.
b) There is a time delay of less than five minutes before the activity, emitted from a fuel rod cladding failure, is indicated by the instrumentation. This time delay is a function of the location of the monitor.
c) The information obtained from this system will not be used for automatic protective or control functions or detection of the specific fuel assembly (or assemblies) which has failed.
d) The high activity alarm will be supplemented with radio-chemical analysis of the reactor coolant for fission products to provide positive identification for a fuel rod failure.
15.2.i Hydrogen Control Status: The discussion of this subject is contained in Section 6.2.5.
1.5-12
REFERENCES FOR SECTION 1.5
- 1. CENPD 87 - Safety related research and development for Combustion Engineering pressurized water reactors, January 1973.
- 2. CENPD 143 - Safety related research and development for Combustion Engineering pressurized water reactors. Program summarizes January 1973 thru February 1974.
1.5-13
1.6 MATERIALS INCORPORATED BY REFERENCE Topical reports listed in Table 1.6-1 are used and/or referenced as part of this application. Table 1.6-1 also includes documents submitted to the AEC in other applications that are incorporated in this application 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.
1.6-1 Am. 1-7/83
TABLE 1.6-1 MATERIALS INCORPORATED BY REFERENCE FSAR Topical Report Title CE Report No. Reference Issue Date Section Seismic Qualification of Category I CENPD-61 December 8, 1972 3.10 Electric Equipment for Nuclear Suppl 1 1/23/73 Steam Supply Systems Suppl 2 3/14/73 Suppl 3 5/10/73 Suppl 4 7/9/73 Suppl 5 8/24/73 Dynamic Analysis of Reactor Vessel CENPD-42 August 1972 3.9.1.3 Internals Under Loss of Coolant 3.9.1.4.3 Conditions with Application of Analysis to CE 800 MWE Class Reactors Coast Code Description CENPD-98A July 5, 1973 15.1 15.2 Densification of Combustion CENPD-118P September 3, 1974 4.2.1.4.15 Engineering Fuel CENPD-118 Calculational Methods for the CE CENPD-132-P August 1974 6.3.3.6 Large Break LOCA Evaluation Model CENPD-132 Rev 01 CEFLASH 4A, A Fortram IV Digital CENPD-133-P August 1974 6.3.3.6 Computer Program for Reactor CENPD-133 Rev 01 Blowdown Analysis CENPD-133 Pm Suppl 1 8/74 CENPD-133 Suppl 1 CENPD-133 P Suppl 2 CENPD-133 Suppl 2 1.6-2
TABLE 1.6-1 (Continued)
FSAR Topical Report Title CE Report No. Reference Issue Date Section COMPERC-II, A program for Emergency CENPD-134 P September 1974 6.3.3.6 Refill -Reflood of the core CENPD-134 Rev 01 CENPD-134 P Suppl 1 2/75 CENPD-134 Suppl 1 STRIKEN II, A Cylindrical Geometry CENPD-135 P September 1974 6.3.3.6 Fuel Rod Heat Transfer Program CENPD-135 Rev 01 CENPD-135 P Suppl 2 2/75 CENPD-135 Rev 01, Suppl 2 High Temperature Properties of CENPD-136 P August 1974 6.3.3.6 Zircaloy and UO2 for Use in the CENPD-136 Rev 01 CE LOCA Evaluation Model Calculational Methods for the CE CENPD-137-P September 1974 6.3.3.6 Small Break LOCA Evaluation Model CENPD-137 Rev 01 PARCH, A Fortran IV Digital Computer CENPD-138-P September 1974 6.3.3.6 Program to Evaluate Pool Boiling, CENPD-138 Rev 01 Axial Rod and Coolant Heatup. CENPD-138 P Suppl 1 2/75 CENPD-138 Rev 01 Suppl 1 CE Fuel Evaluation Model Topical Report CENPD-139 P-A September 1974 4.2.14 CENPD-139-A CE Procedures for Design, Fabrication, CENPD-155 P-A September 1974 5.4.4 Installation and Inspection of Surveil- CENPD-155-A lance Specimen Brackets 1.6-3
REFER TO DRAWING 8770-B-327, Sheets 1, 2, 3, 4, 5, 6, 7, 8 & 8A FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 CONTROL WIRING DIAGRAM INDEX Figure 1.6-2 Amendment No. 16, (1/98) 1.6-4
1.7 DESIGN COMPLIANCE WITH AEC SAFETY GUIDES, INFORMATION GUIDES AND CODE OF FEDERAL REGULATIONS CROSS REFERENCES Information contained herein was valid at the time the operating license for St. Lucie 1 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.
Conformance with the AEC General Design Criteria is discussed in Section 3.1. The criteria requirements and cross references indicated in the AEC Safety Guides and Information Guides are presented in Table 1.7-1.
1.7-1
TABLE 1.7-1 DESIGN COMPLIANCE Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section Safety Guide 1 - 1. Emergency core cooling and containment 6.2.2.2; Net Positive Suction heat removal system pumps should have 6.3.2 Head for Emergency adequate NPSH at maximum expected Core Cooling and pumped fluid temperatures assuming no Containment Heat increase in containment pressure over that Removal System present prior to a postulated LOCA.
Pumps - GDC 41 Safety Guide 2 - 1. Data collection and research on the 5.4 Thermal Shock to properties of reactor pressure vessel Reactor Pressure material should be continued to verify that Vessels - GDC 35 non-brittle behavior can be assured throughout the vessel lifetime under postulated accident conditions.
- 2. No significant changes in the approved core or reactor pressure vessel designs will occur. This negates the need to review in this individual case the potential thermal shock problem.
- 3. The-vessel design does not preclude annealing should the safety margin to brittle failure appear to become unacceptable under the assumption of cooling operation.
Safety Guide 3 - 1. Not applicable.
Safety Guide 4 - 1. The radioactive material release 15.4 Assumptions Used for atmosphere diffusion and dose conversion Evaluating the Potential assumptions are followed in evaluating the Radiological design basis LOCA.
Consequences of a LOCA for Pressurized Water Reactors 1.7-2 1.7-1
TABLE 1.7-1 (Continued)
Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section Safety Guide 5 - 1. Not applicable.
Safety Guide 6 - 1. AC and DC safety loads should be 8.2.2; Independence between separated into redundant load groups such 8.3.1.2; Redundant Standby that loss of one group will not prevent the 8.3.2.2 (On-Site) Power minimum safety functions to be performed.
Sources and between
- 2. Each AC load group should have a their Distribution connection to a preferred offsite power Systems - GDC 17 source and to a standby on-site power source, which should have no automatic connection to a redundant load group.
- 3. Each DC load group should be energized by a battery-battery charger combination, which should have no automatic connection to a redundant load group.
- 4. Redundant load groups and redundant standby sources should be independent to this extent:
- a. No automatic source paralleling,
- b. No automatic load transfer between sources,
- c. No automatic load group connecting to another load group, and
- d. Manual paralleling of load groups protected from standby power source paralleling by an interlock.
- 5. Multiple prime movers driving a single generator should be shown to have equivalent reliability to a single prime mover, including consideration of common mode and random single failures.
Safety Guide 7 - 1. The capability should be provided for: 6.2.5.3; Supplement - Control a. Measuring H2 concentration, 15.4 of Combustible Gas b. Mixing the containment atmosphere Concentrations in c. Controlling gas concentrations without Containment following reliance on purging; equipment LOCA-GDC 41 available from off-site on appropriate time notice.
1.7-3
TABLE 1.7-1 (Continued)
Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section
- 2. Gas control systems should meet design quality assurance, redundancy, energy source and instrumentation of an engineered safety feature.
- 3. Controlled purge through a fission product removal system should be provided.
- 4. The gas generation model in Safety Guide Table 1 should be followed in evaluating the design basis, LOCA.
- 5. Practical limitation of materials in the containment that would produce H2 due to the emergency coolant and spray solutions.
- 6. Backfitting on item 1 should occur prior to operation unless the purge dose is shown to be negligible following submittal of justification.
Safety Guide 8 - 1. ANSI N18.1, Proposed Standard in 13.2; 14.2 Personnel Selection Selection and Training of Personnel for and Training - Nuclear Power Plants, (6/22/72), should 10CFR50.34 (b.) (6.) be followed for selection and training of (i.) nuclear power plant personnel.
Safety Guide 9 - 1. The predicted loads should not exceed the 8.3 Selection of Diesel smaller of the 2000-hr. or 90% of the 30-Generator Set Capacity min. rating.
for Standby Power
- 2. During preoperational testing, the predicted Supplies - GDC 17.
loads should be verified by tests.
- 3. Diesel generator dynamics should exceed:
- a. Loading frequency transient > 95%;
voltage transient > 75%; disconnection of largest single load - speed increase
< 75% of overspeed trip set point or 115% of nominal speed whichever is lower; voltage 1.7-4
TABLE 1.7-1 (Continued)
Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section restored within 90% and frequency within 98% in < 40% of load sequence time interval.
- 4. Prototype qualification data and preop tests should be confirmed on each diesel.
Safety Guide 10 - 12.1.2
- 1. Each member of the splicing crew should Mechanical (Cadweld) prepare two qualification splices using the Splices in Reinforcing same materials as those used in the Bars of Concrete structure. The qualification and production Containments - GDC 1 splices should meet the requirements specified by the designer and approved by the licensee.
- 2. The qualification and production splices should be inspected at both ends of the splice sleeve and at the center tap hole in accordance with the requirements specified by the designer and approved by the licensee.
- 3. Qualification and production splice samples for tensile testing should test equal to or in excess of 125 percent of the minimum yield strength according to ASTM Standards, and the average tensile strength of each group of 15 consecutive samples should equal or exceed the guaranteed ultimate strength specified for the reinforcing bar.
- 4. Tensile Test Frequency should following the Guide schedule for production splices and sister splices, if provided.
- 5. The Procedure for Sub-Standard Tensile Test Results should follow the Guide schedule for retesting, cessation of mechanical splicing, independent laboratory analysis, balance of production rejection and reduced strength acceptability.
1.7-5
TABLE 1.7-1 (Continued)
Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section Safety Guide 11 - 1. Protection system sensing (pipe) lines for 6.2.4 Instrument Lines instruments should:
Penetrating Primary a. Meet redundancy, independence, Reactor Containment - quality assurance, and testability GDC 55, 56 criteria.
- b. Have minimum leakage consistent with their function.
- c. If ruptured, will maintain integrity and function of secondary containment systems,
- d. Rupture, will not cause offsite exposure in excess of 10 CFR 100.
- e. Have an automatic or remote operation isolation valve as close as practical to the outside containment wall, which valve position remains as is on loss of power, and its position is indicated in the control room,
- f. Be located, separated and protected to minimize accidental rupture, but are accessible for visual in-service inspection.
- 2. Sensing lines not part of the protection system should meet items 1.b. through 1.f.
and have one isolation valve inside and one outside the containment wall.
Safety Guide 12 - 1. One strong motion triaxial accelerograph 3.7.4 Instrumentation for should be installed:
Earthquakes - 10 CFR a. In the reactor containment structure 50.36(c.) basement, and
- b. Another one at a higher elevation of the reactor containment structure, such that the
- c. Vertical separation of accelerographs is a significant fraction of the containment height.
1.7-6
TABLE 1.7-1 (Continued)
Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section
- d. Axes orientations orthogonal.
- e. Locations in vertical line.
- f. Locations accessible.
- g. Rigidly attached to the containment structure.
- 2. Peak deflection accelerographs should be installed on other selected category I structures, the need to be evaluated on a case basis.
- 3. If different soil conditions underlie independent foundations containing Category I components, additional instrumentation should be provided, the need to be evaluated on a case by case basis. If needed, a free-field accelerograph should be installed.
- 4. The peak acceleration level in the reactor containment basement should be available to the control room operator a few minutes after a postulated earthquake.
- 5. The accelerographs should be designed to function in the expected range of environment conditions.
- 6. A plan for utilization of any data recorded should be available.
Safety Guide 13 - 1. The spent fuel facility should meet 9.1; Fuel Storage Facility Category I seismic requirements. 15.4 Design Basis - GDC 61
- 2. The facility should be designed to prevent tornado winds or missiles from causing significant pool water loss or fuel damage.
- 3. Interlocks should be provided to prevent cranes passing over or within striking distance of fuel when fuel handling is not in progress.
- 4. The fuel pool enclosure should be provided for radioactive particulate cleanup and controlled leakage. The ventilation and filtration capability is assessed on the basis that one fuel bundle clad-gap activity might be released.
1.7-7
TABLE 1.7-1 (Continued)
Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section
- 5. The fuel pool structure should be designed to withstand a cask drop without substantial water loss or fuel damage.
- or -
The cranes capable of carrying heavy loads 9.1-2.2.2 (including the spent fuel cask) should be designed to provide single-failure-proof handling of heavy loads.
- 6. Water cleanup, makeup and drain connections should be designed to preclude substantial water loss in case of maloperation or line rupture.
- 7. Low water and high radiation conditions should be alarmed in control room.
- 8. The water makeup system should be Category I and redundant backup water makeup arrangements should be available. The backup system need not be permanently installed.
Safety Guide 14 - 1. Flywheel material should meet the following test 5.5 Reactor Coolant Pump criteria:
Flywheel Integrity - GDC a. NDT 10 F 4 b. Three specimens exhibit Cv (WR) 50 ft-lb
- c. Minimum fracture toughness at operating temperature equivalent to KIc dynamic 100 ksi in.
- e. Flame cut flywheels have 1/2 in. of stock left for machining to size.
- 2. The flywheel should be designed to meet the following criteria:
- a. Normal operative speed stresses 1/3 minimum specified yield strength.
- b. Design overspeed 110% anticipated maximum overspeed. The basis for the assumed design overspeed should be stated.
- c. Design overspeed stresses 2/3 minimum specified yield strength.
1.7-8 Amendment No. 20 (4/04)
TABLE 1.7-1 (Continued)
Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section
- 3. The flywheels should be tested at the design overspeed.
- 4. In place in-service inspection should occur:
- b. About every 10 years - exposed surface inspection and complete UT.
Safety Guide 15 - 1. Yield and tensile strength of reinforcing 3.8.2 Testing of Reinforcing bars should be tested:
Bars fo Concrete a. One specimen/50 tons or fraction Structures - GDC 1 thereof from each heat,
- b. Tests conform to ASTM A-370-68,
- c. Acceptance conform to ASTM A-615-68
- 2. Reinforcing bars deformation should be inspected to assure conformance to ASTM A-615-68 Safety Guide 16 - 1. Routine, non-routine and special reports 13.6; Reporting of Operating should be prepared and filed according to 16.6.6 Information - 10 CFR tabular data in Section 13.6 and 16.6.6, 20, 50.36, 70 & 73 explicitly following Safety Guide 16.
Safety Guide 17 - 1. Control of access to vital areas of the 13.7 Protection Against nuclear power plant should include:
Industrial Sabotage, a. Exclusion fencing dated 10/20/71. b. Fence surveillance
- c. Access point guards
- d. Key and magnetic locks
- e. Security organization plan with line responsibility delegated 1.7-9
TABLE 1.7-1 (Continued)
Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section
- f. Individual personnel and visitor identification and inventory procedures
- g. Procedures for vehicle and personnel entry control and search authorization.
- h. Procedures for internal and potential external emergencies including sabotage.
- i. Procedures for security incident investigation and audit.
- 2. Means for selection and review of reliable plant personnel should include established employment standards, periodic employee performance review, detection procedure for unusual behavior patterns and plant security training of personnel.
- 3. Means should be provided for continual monitoring of vital equipment status.
- 4. The design and arrangement of equipment and facilities should include consideration of minimizing opportunity for industrial sabotage.
Safety Guide 18 1. Not applicable.
Safety Guide 19 - 1. Liner (Containment Vessel) seam welds Not applicable.
Non-destructive should have:
Examination of Primary a. Radiographs of first ten feet and at Containment Liners - least one foot of succeeding 50 foot or GDC 1 fraction for each welder and welding position, where accessible.
- c. Soap solution or equivalent sensitivity leak tests.
- 2. Penetrations, airlock and access opening welds-using techniques in Section V, ASME B & PV Code should 1.7-10
TABLE 1.7-1 (Continued)
Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section have:
- a. Welds between flued fittings and pipe lines are 100% radiographed.
- c. All other welds fully examined according to NE-5120 of Section III
- 3. Welders and welding procedures should be qualified according to Section IX.
- 4. Non-destructive examination personnel should be qualified according to Section V.
- 5. Each radiograph location should be greater than 10 feet apart and should be recorded.
- 6. Non-destructive examinations should be formed as soon as practical following each lineal weld increment.
- 7. Accepted welds should meet the applicable B & PV Code standards.
- 8. Where unacceptable welds are found, repair and reexamination by the original non-destructive methods should be preformed in accordance with the B & PV Code.
- 9. Records of both accepted and repaired defective welds should be retained by applicant.
Safety Guide 20 - For reactor internals similar to the prototype 4.2; 5.6 Vibration design; Measurements on
- 1. The preoperational functional tests should Reactor Internals -
be done in all significant flow modes of GDC 1 normal operation, under the same test conditions and for at least as 1.7-11
TABLE 1.7-1 (Continued)
Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section long a test duration as imposed on the prototype design.
- 2. Following 1, above, the internals should be removed from the reactor vessel. Visual and non-destructive examinations should be made of:
- a. major load bearing elements,
- b. lateral, vertical and torsional restraints within the vessel,
- c. locking and bolting devices,
- d. other locations examined in the prototype design,
- e. the reactor vessel interior for loose parts or foreign material.
- 3. A summary of the inspection should be submitted to the Commission indicating similarity and/or dissimilarity of vibrational characteristics to the prototype design, with corrective actions where applicable.
Safety Guide 21 - Normal and potential paths for release of 11.4; Measuring and radioactive material during normal plant 16.3.8 Reporting of Effluents operation should be monitored. The sensitivity from Nuclear Power and calibration of monitors, and the Plants - 10 CFR frequencies of sampling and analysis should 50.36(a)(2) be specified in the Technical Specifications.
Records of releases should be retained and reported in conformance with Appendix A of Guide 21.
- a. Quantity and isotopic analysis,
- b. Hourly meteorology during release,
- c. Tank concentrations of isotopes > 10-4 c/cc; concen. Of H-3.
- d. Continuous iodine sampler; weekly sample analyzed for I-131; 10-4 Ci/sec measurable; 10-10 Ci/cc in tank measurable,
- e. Continuous particulate sampler; weekly sample analyzed for gross 1.7-12
TABLE 1.7-1 (Continued)
Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section
, , Ba-140, La-140 and I-131; 10-4 Ci/sec --emitters measurable; 10-10 Ci/cc -emitters in tank measurable.
- f. Monthly composite -emitters analysis
- g. Quarterly composite of Sr-89, Sr-90, plus gross of .
- 2. Gross , ; Ba-La-140; I-131; dissolved fission gases; Sr-89-90; principal -
emitters and activation products and H-3.
- a. Sample makeup, compositing and aliquotting plus minimum concentration sensitivities in Guide 21, c.4. should be met.
Safety Guide 22 - The protection system should be designed to 7.3.2.6 Periodic Testing of permit periodic testing to include as closely as Protection System practical the required performance of the Actuation Functions - actuation devices and actuated equipment, GDC-20 except that only actuation devices can be tested during reactor operation.
- 1. The actuation devices are included by testing in judiciously selected groups and preventing operation of certain actuation equipment during actuation device tests.
- 2. Positive isolating means should be used, indicated in the control room, to prevent expansion of a by-pass condition to redundant or diverse channels.
- 3. Where actuated equipment is tested only during reactor shutdown.
- a. Adverse safety or operability implications are shown if actuation during reactor operation occurred.
- b. Protection channel reliability should be shown to be adequately high on low frequency testing.
1.7-13
TABLE 1.7-1 (Continued)
Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section Safety Guide 23 - 1. Instruments should be installed to measure 2.3.3 Onsite Meteorological wind direction, wind speed and air Programs - 10 CFR temperature at 10 meters above ground 50.34(a)(1) and (b)(1) and 30 meters above the lower sensors, on a tower isolated from structures effects.
- 2. Analog or digital records may be provided and strip chart recorders should be located in the control room.
- 3. Instrument accuracy:
- a. Wind direction +/- 5o.
- b. Wind speed, time averaged, +/- 0.5 mph; starting speed, < 1 mph.
- c. Temperature, time averaged, +/- 0.5o C; Temperature difference averaged +/-
0.1o C.
- 4. Instruments should be serviced to provide
> 90% data recovery; semi-annual calibration.
- 5. Data should be averaged over more than 15 minutes, once each hour.
- 6. Monthly or seasonal and annual joint frequency distributions should be reported, for each stability class. /Q values for accidental and annual average effluent releases diffusion effects should be calculated and the calculation assumptions annotated.
Safety Guide 24 - Care is taken to prevent air in-leakage to the 15.4 Assumptions Used for gas storage tanks to minimize possibility of Evaluating the Potential explosive H2 -O2 mixture. Tanks are isolated Radiological from each other during use. Over pressure Consequences of a relief system design minimizes a) likelihood of Pressurized Water premature release due to operator error or Reactor Radioactive valve malfunction, and b) radiation exposure to Gas Storage Tank on-site personnel during venting. Tank and Failure system components located and shielded to minimize likelihood of premature release due to internal or external missiles and 1.7-14
TABLE 1.7-1 (Continued)
Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section mechanical accidents. The assumptions used for the tank failure accident evaluation follow Guide 24.
Safety Guide 25 - 15.4 The assumptions used for the fuel handling Assumptions Used for accident evaluation are based on Evaluating the Potential
- 1. The highest power assembly peak linear Radiological power density = 20.5 kw/ft.;
Consequences of a
- 2. Maximum centerline temperature = 4500 F; Fuel Handling Accident
- 3. Average burnup 25,000 MWD/MT, (peak local burnup 45,000);
- 4. Technical specification minimum time to move fuel;
- 5. Fuel rod pressure 1200 psig
- 6. Minimum water depth = 23 feet;
- 8. Radial peaking factor 1.65 used for inventory calculation, full power end-of-life operation;
- 9. Iodines 99.75% inorganic and 0.25%
organic;
- 10. Effective decontamination factors in water for I = 100 and for noble gases = 1
- 11. Fuel pool building activity release period =
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />;
- 12. Filter efficiencies for inorganic I=90% (75%
air inventory) and for organic I=70% (25%
air inventory)
- 13. Atmospheric release and diffusion characteristics similar in Guides 24 and 25 for ground level releases.
Safety Guide 26 - Acceptable quality standards for water-and 3.2 Quality Group steam containing components (not reactor Classifications and coolant pressure boundary, turbine or Standards-GDC 1 condensers) in systems required for Quality Group B:
- 1. Emergency core cooling,
- 2. Post-accident containment heat removal,
- 3. Reactor shutdown,
- 5. Steam generator secondary to outermost containment isolation valves and
- 6. Connections to RCS boundary incapable of isolation during all normal operating modes by two valves; 1.7-15
TABLE 1.7-1 (Continued)
Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section this Quality Group B should meet ASME - B &
PV Section III, Components Class 2; Quality Group C: Cooling water, seal water and feedwater components in systems listed in Quality Group B, plus radioactive waste management system components whose failure might release activity exceeding a potential site boundary exposure > 170 mrem; this Quality Group C should meet ASME-B&PV Section III, Components Class 3; Quality Group D: Any remaining components as part of a system that might contain radioactive material, not Group B or C; this Quality Group D should meet ASME - B&PV,Section VIII, Division 1, or ANSI B31.1.0 (for piping and valves), or API-620, API-650, AWWA-D 100, or ANSI B 96.1 (for Atmospheric or 0-15 psig Storage Tanks).
Safety Guide 27 - 1. The circulating water system should be 2.4.9; Ultimate Heat Sink - capable of providing sufficient cooling for at 2.4.11; GDC-44 least 30 days under accident conditions, 9.2 and procedures for extended capability after 30 days are available.
- 2. The heat sink should be capable of withstanding the effects of the postulated maximum hurricane and failure of the intake conduit.
- 3. The heat sink should consist of 2 water sources (2 conduits) or one conduit of extremely low failure probability.
- 4. Procedures include corrective action for water supply if 1. or 2. above temporarily not satisfied.
Safety Guide 28 - ANSI N45.2-1971, Quality Assurance Program Quality Assurance Requirements for Nuclear Power Plants Program Requirements provide the guidelines for establishing and
- 10 CFR 50, Appendix executing a QA program during the design and B construction phases and provide an adequate basis for complying which the QA criteria, App.
B, 10 CFR 50.
1.7-16
TABLE 1.7-1 (Continued)
Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section Safety Guide 29 - The following items including their foundations 3.2, 3.7 Seismic Design and supports, are designated Category I, i.e., 3.8, 3.10 Classification - GDC 2; should be designed to withstand the effects of 10 CFR 50, Appendix A the safe Shutdown Earthquake and remain 10 CFR 100, Appendix A functional:
- a. reactor coolant pressure boundary
- b. reactor core and vessel internals
- c. subsystems for
- 1) emergency core cooling
- 2) post-accident containment a) heat removal b) atmosphere control
- 3) reactor shutdown
- 5) spent fuel pool cooling
- d. portions of steam and feedwater systems from the steam generators secondary sides through the outermost containment isolation valves, piping 2 1/2 inches and larger, and the first valve, either NC or capable of automatic closure during normal reactor operation.
- e. cooling water, component cooling, auxiliary feedwater and seal water subsystems required for:
- 1) emergency core cooling
- 2) post-acident containment a) heat removal b) atmosphere control
- 4) spent fuel pool cooling
- 5) reactor coolant components important to safety
- f. radioactive waste handling subsystems whose failure would result in potential off-site exposures approaching 10 CFR 100 guidelines
- g. subsystems required to
- 1) supply fuel for emergency equipment
- 2) monitor and actuate systems important to safety
- h. protection system
- i. spent fuel pool structure and racks
- j. reactivity control systems 1.7-17
TABLE 1.7-1 (Continued)
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GDC References Requirements FSAR Section
- k. control room and associated life support systems
- l. primary and secondary reactor containment
- m. on-site emergency electrical power system needed for items a through l Category I seismic design requirements extend to the first seismic restraint (including the interface component) beyond the defined boundaries.
Safety Guide 30 - Requirements for the installation, inspection, 8.1 Quality Assurance and testing of instrumentation and electric Requirements for the equipment included in ANSI N45.2.4 - 1972, Installation, Inspection, also designated IEEE 336 - 1971, are and Testing of generally acceptable for complying with the Instrumentation and pertinent QA requirements of 10 CFR 50, App.
Electric Equipment - B.
- 1. ANSI N45.2.4 - 1972 should be used in conjunction with ANSI N45.2-1971.
- 2. Requirements of Subdivision 1.1, ANSI N45.2.4 - 1972, apply to plant operation phase.
Safety Guide 31 - Weld fabrications for austenitic stainless steel 17.1 Control of Stainless core support structures and Class 1, 2 and 3 Steel Welding - GDC 1; components should comply with Section III and 10 CFR 50, Appendix B Section IX of the ASME B&PV Code, supplemented by;
- 1. The procedure qualification should require certain tests for delta-ferrite, welding heat input be specified and transverse side bend test specimen be examined for fissures.
- 2. The results of item 1. tests should be included in qualification test report.
- 3. Welding materials for production welds should meet ASME B&PV Code,Section III and delta-ferrite requirements in item 1.
- 4. Production welds should meet item 1.
requirements including 1.7-18
TABLE 1.7-1 (Continued)
Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section instruments calibrated to same standards used for procedure qualification.
- 5. If item 4. is not met, additional side bend examinations should be done.
- 6. Production welds heat input requirements in item 1 should be met.
Safety Guide 32 - IEEE STD 308-1971 may be used in 8.2 Use of IEEE STD-308- implementing GDC 17 except where conflict in 1971, Criteria for Class resolution occurs; 1E Electric Systems...
- 1. Availability of offsite Power - a preferred
- GDC 17; design includes two immediate access circuits from the transmission network; an acceptable design would substitute a delayed access circuit if it conforms to Criterion 17 and the other circuit is immediately accessible.
- 2. The capacity of the battery charger supply should be based on the largest combined demands of steady-state loads and the charging capacity to restore the battery to the fully charged state.
Safety Guide 33 - The requirements for administrative controls to 17.2 Quality Assurance safely operate nuclear power plants given in Program Requirements proposed standard ANS-3.2 (Nov. 2, 1972) and (operation) - 10 NSI N45.2 - 1971 are generally acceptable for CFR50, Appendix B complying with 10 CRF50, Appendix B.
Appendix A to Safety Guide 33 should be used to assure minimum plant operating and maintenance procedures coverage.
1.7-19 Amendment No. 17, (10/99)
TABLE 1.7-1 (Continued)
Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section Information Guide 1 - 1. Containment system boundaries 6.2 Primary Reactor delineated on figures.
Containment Systems (Steel Construction)
- 2. Max. internal and external pressures, 3.8.2; 3.9 temperatures and design bases; design loading combinations, as Normal, Upset Emergency conditions; stress intensity limits, means or tests to demonstrate operability under Emergency Conditions.
- 3. Fracture toughness criteria 3.8.2.1.5
- 4. Seismic design criteria 3.7
- 5. Design criteria for mechanical, thermal and 3.8.2 nonaxisymmetric loads.
- 6. Design criteria for those penetrations 3.8.2 whose pipes could transmit accident vibrations.
- 7. Design provisions for leak testing of 6.2.1.4 penetrations.
- 8. Design criteria and test considerations for 6.2.1.4 components to withstand completed containment integrated pressure tests.
Information Guide 2 - 1. Protection system reactor trip, ESF trip and 7.2; 7.3 Instrumentation and other safety actions:
Electrical Systems a. Lists of Systems designed and built by the NSSS similar and dissimilar to the in-service systems employed elsewhere; discussion of differences
- b. List of systems by non-NSSS
- c. Justification of any features not meeting IEEE Std 279-1971 and GDC criteria
- 2. Controls systems by NSSS: 7.5; 7.7
- a. List of systems similar to his in-service systems employed elsewhere.
1.7-20
TABLE 1.7-1 (Continued)
Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section
- b. List and discussion of systems dissimilar along with safety significance evaluation of differences
- 3. Seismic design criteria 3.10
- a. Protective action initiation during DBE.
- c. Documentation of equipment qualification tests
- 4. Quality assurance description as applied to 7.1; 17.1 reactor protection, ESF and emergency power equipment
- 5. Criteria and bases to establish minimum 7.1; 8.3.1.2 independence and redundance of subject systems and their provision quality assurance including electrical cable.
- a. Derating
- b. Routing in hostile areas
- c. Vital and non-vital sharing in trays
- d. Fire detection and protection
- e. Marking and tray marking, and
- f. Spacing of wires and components at terminations
- 6. Design criteria to protection from potential 3.11 effects of normal and accident radiation levels. Documentation of equipment radiation qualification tests.
- 7. Description of and documentation of 3.11 qualification tests on safety related equipment within the containment to withstand the DBA environment.
- 8. Identify limiting temperature on 3.11 instrumentation and control equipment that requires reactor shutdown, expected worst case temp.-humidity environment and criteria to loss of HVAC will 1.7-21
TABLE 1.7-1 (Continued)
Document/Title/ Criteria Compliance -
GDC References Requirements FSAR Section not adversely affect safety equipment operation. Documentation of qualification tests in worst case environment
- 9. On the spot identification of safety 7.1; 8.3.1.2.3 equipment, safety channels and cabling description
- 10. Description of method for periodic testing 7.3.2.6 of subject equipment including conformance to IEEE Std 279-1971
- 11. Information readouts of conditions in the 7.5 reactor, the reactor coolant, and the containment:
- a. Design criteria
- b. Type
- c. Number of channels
- d. Range
- e. Accuracy
- f. Location
- 12. Justify and non-conformance by the 8.3 emergency power system to Safety Guides 6 and 9, and to IEEE Std 308-1971
- 13. Reasons for and safety significance of any 1.3 changes in subject equipment from that presented in the PSAR 1.7-22