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{{Adams
#REDIRECT [[L-2016-198, Turkey Point, Units 3 & 4, Updated Final Safety Analysis Report, Chapter 1, Introduction and Summary]]
| number = ML16330A180
| issue date = 10/29/2016
| title = Turkey Point, Units 3 & 4, Updated Final Safety Analysis Report, Chapter 1, Introduction and Summary
| author name =
| author affiliation = Florida Power & Light Co
| addressee name =
| addressee affiliation = NRC/NRR
| docket = 05000250, 05000251
| license number =
| contact person =
| case reference number = L-2016-198
| package number = ML16330A191
| document type = Updated Final Safety Analysis Report (UFSAR)
| page count = 127
}}
 
=Text=
{{#Wiki_filter:TABLE OF CONTENTS
 
Section Title Page
 
==1.0 INTRODUCTION==
AND SUMMARY 1-1
 
1.1  Site and Environment 1.1-1
 
1.2  Summary Description 1.2-1 1.2.1  Structures 1.2-2    Seismic Classification of Particular Structures and    Equipment 1.2-2 1.2.2  Nuclear Steam Supply System 1.2-2 1.2.3  Control System 1.2-3 1.2.4  Waste Disposal System 1.2-4 1.2.5  Fuel Handling System 1.2-4 1.2.6  Turbine and Auxiliaries 1.2-5 1.2.7  Electrical System 1.2-5 1.2.8  Engineered Safety Features 1.2-6 1.2.9  Fire Protection Program 1.2-6 1.2.10  Independent Spent Fuel Storage Installation (ISFSI) 1.2-6 1.2.11  References for Section 1.2.10 1.2-7 1.3 General Design Criteria 1.3-1 1.3.1  Overall Requirements 1.3-1 1.3.2  Protection by Multiple Fission Product Barriers 1.3-3 1.3.3  Nuclear and Radiation Controls 1.3-5 1.3.4  Reliability and Testability of Protection Systems 1.3-8 1.3.5 Reactivity Control 1.3-10 1.3.6 Reactor Coolant Pressure Boundary 1.3-12 1.3.7 Engineered Safety Features 1.3-14 1.3.8 Fuel and Waste Storage Systems 1.3-21 1.3.9 Effluents    1.3-23
 
1.4 Design Parameters and Unit Comparison 1.4-1 1.4.1  Design Developments Since Receipt of Construction Permit 1.4-1 Burnable Poison Rods 1.4-1 Safety Injection System 1.4-1 Containment Sumps 1.4-2 Safety Injection System Trip Signal 1.4-2    Containment Spray System Signal 1.4-3    Rod Stop and Reactor Trip on Startup 1.4-3 Isolation of the Control and Protection Systems 1.4-3    Electrical System Design 1.4-4    Auxiliary Coolant System 1.4-5 Waste Disposal System 1.4-5    Thermal Power Uprate 1.4-5    Extended Power Uprate 1.4-5    Low Level Waste Storage 1.4-6 1-i Revised 09/20/2016 C28 TABLE OF CONTENTS (Continued)
Section Title Page 1.5 Design Highlights 1.5-1 1.5.1  Power Level 1.5-1 1.5.2  Reactor Coolant Loops 1.5-1 1.5.3  Peak Specific Power 1.5-1 1.5.4  Fuel Assembly Design 1.5-2 1.5.5  Engineered Safety Features 1.5-2 1.5.6  Emergency Power 1.5-2 1.5.7  Emergency Containment Cooling and Filtering Systems 1.5-3 1.5.8  References 1.5-3 1.6 Research and Development Items 1.6-1 1.6.1  Initial Core Design 1.6-1 1.6.2  Development of Analytical Methods for Reactivity Transients from Rod Ejection Accidents 1.6-1 1.6.3  Safety Injection System Design 1.6-3 1.6.4  Systems for Reactor Control During Xenon Instabilities 1.6-4 1.6.5  Blowdown Capability of Reactor Internals 1.6-5  References 1.6-6 1.7 Identification of Contractors 1.7-1 1.8 Safety Conclusions 1.8-1 1.9 Quality Assurance Program 1.9-1 1.9.1  Purpose 1.9-1 1.9.2  Applicability 1.9-1 1.9.3  Organization 1.9-4 1.9.4  Scope 1.9-7 1.9.5  Design and Procurement 1.9-8 1.9.6  Shop Fabrication Quality Assurance 1.9-9 1.9.7  On-Site Construction, Erection, and Installation 1.9-10 
 
1-ii Revised 04/17/2013 APPENDICES
 
Appendix 1A Westinghouse Power Systems Division Quality Assurance Plan
 
Appendix 1B Quality Assurance Package for Concrete 
 
1-iii
 
LIST OF TABLES
 
Table Title 1.4-1 Comparison of Design Parameters
 
1-iv
 
LIST OF FIGURES
 
Figure                              Title
 
1.2-1 General Building Arrangement Plan
 
1.2-2 General Arrangement Plan EL 10'- 0" and Below
 
1.2-3 General Arrangement Ground Floor Plan EL. 18'- 0"
 
1.2-4 General Arrangement Operating Floor Plan EL. 42'- 0" and EL. 58'- 0"
 
1.2-5 General Arrangement Mezzanine Floor Plan and Section "A-A"
 
1.2-6 General Arrangement Sections "B-B" and "C-C"
 
1.2-7 General Arrangement Sections "D-D" and "E-E"
 
1.2-8 General Arrangement Unit 4 EDG Building Plan and Sections
 
1.9-1 Quality Assurance Organization - Florida Power & Light
 
1.9-2 Quality Assurance Organization - Bechtel
 
1.9-3 Quality Assurance Organization - Westinghouse
 
1-v Rev. 11  11/93 
 
==1.0 INTRODUCTION==
AND SUMMARY This Final Safety Analysis Report is submitted in support of an application by Florida Power & Light Company for a license to operate two nuclear power units designated as Turkey Point Units 3 and 4, located adjacent to oil and gas fired Unit 1 and a dual-convertible synchronous condenser/generator Unit 2 at the Turkey Point Plant, a steam electric generating facility situated on the shore of Biscayne Bay about 25 miles south of Miami, Florida.
 
The Turkey Point Units 3 and 4 reactors are pressurized light water moderated and cooled systems. Each reactor was originally designed to produce 2200 MWt. The corresponding steam and power conversion system, including turbine generator, was designed to permit the generation of 728 MW of gross electrical power. Subsequent Thermal Uprate increased the capacity of the reactors to produce 2300 MWt with a corresponding gross electrical power output of 775 MW. The Extended Power Uprate increases the capacity of the reactors to produce 2644 MWt with a corresponding gross electrical power output of 888 MW.
 
The nuclear power units incorporate a closed-cycle pressurized water Nuclear Steam Supply System and a Turbine-Generator System utilizing dry saturated steam. Equipment includes the Radioactive Waste Disposal System, Fuel Handling System, main transformers, main condenser and all auxiliaries, structures, and other on-site facilities required to provide complete and operable nuclear power units.
 
The nuclear safety systems, including containment and engineered safety features, are designed and evaluated for operation at the higher power level, which is used in the analysis of postulated loss-of-coolant accidents in this report.
 
The balance of this section summarizes the principal design features and safety criteria of the nuclear units, and compares them with some other pressurized water nuclear power plants employing the same technology and basic engineering features.
 
1-1 Revised 12/15/2015 C28 Section 2 contains a description and evaluation of the Turkey Point site and environs, supporting the suitability of the site for reactors of the size and
 
type described. Sections 3 and 4 describe the reactors and the reactor
 
coolant systems, Section 5 the structures and related systems, and Section 6
 
through 11 the emergency and other auxiliary systems.
 
Section 12 makes reference to the Company's program for organization and
 
training of personnel. Section 13 contains an outline and reference to the
 
initial tests and operations associated with startup.
 
Section 14 is a safety evaluation summarizing the analyses which demonstrate
 
the adequacy of the reactor protection systems, and the engineered safety
 
features. The consequences of various postulated accidents are within the
 
guidelines set forth in 10 CFR 50.67.
 
Section 15 makes reference to the Technical Specifications under which the
 
units are operated.
 
1-2 Revised 04/17/2013 C26 1.1 SITE AND ENVIRONMENT The site is on the shore of Biscayne Bay, about 25 miles south of Miami,
 
Florida. The area immediately surrounding the site is low and swampy and is
 
very sparsely populated, with much of it unsuited for development without
 
raising the elevation with fill. The nearest farming area lies in the
 
northwest quarter of a 5-mile arc from the site.
 
The area surrounding the site is flat and slopes very gently to the west from
 
sea level at the shoreline of Biscayne Bay to an elevation of about 10 ft
 
above MSL at a point some 8 to 10 miles inland. To the east across Biscayne
 
Bay from 5 to 8 miles, is a series of offshore islands running in a
 
northeast-southwest direction between the Bay and the Atlantic Ocean, the
 
largest of which is Elliott Key.
 
The site is well ventilated with air movement prevailing almost 100 per cent
 
of the time. The atmosphere in the area is generally unstable with diurnal
 
inversions of short duration.
 
The Miami area has experienced winds of hurricane force periodically. During
 
storms the plant may be subjected to flood tides of varying heights. 
 
Hurricane "Betsy" in 1965 produced the maximum flooding recorded, which was
 
about 10 feet above MSL. External flood protection is described in Appendix
 
5G.
 
The normal direction of natural drainage of surface and ground water in the
 
area of the site is to the east and south toward Biscayne Bay and will not
 
affect off-site wells. A radiological background study of the Turkey Point
 
area will be initiated approximately one year prior to initial startup of the
 
Unit 3. This will involve the collection of samples of air, soil, water,
 
marine life, biota and vegetation in the area. The bed rock beneath the
 
limerock fill is competent with respect to foundation conditions for the
 
nuclear units. The area is in a seismologically quiet region, all of Florida
 
being classified Zone 0 (the zone of least probability of damage) by the
 
Uniform Building Code, as published by International Conference of Building
 
Officials. 
 
1.1-1 Revised 04/17/2013 1.2  SUMMARY DESCRIPTION The inherent design of the pressurized water, closed-cycle reactor
 
significantly reduces the quantities of fission products which must release
 
to the atmosphere. Four barriers exist between the fission product
 
accumulation and the environment. These are the uranium dioxide fuel matrix,
 
the fuel cladding, the reactor  vessel and coolant loops, and the
 
containment. The consequences of a breach of the fuel cladding are greatly
 
reduced by the ability of the uranium dioxide lattice to retain fission
 
products. Escape of fission products through a fuel cladding defect would be
 
contained within the pressure vessel, loops and auxiliary systems. Breach of
 
these systems or equipment would release the fission products to the
 
containment where they would be retained. The containment is designed to
 
retain adequately these fission products under the accident conditions
 
analyzed in Section 14. 
 
Several engineered safety features have been incorporated into the design to 
 
reduce the consequences of a loss of coolant incident. These safety features
 
include a Safety Injection System. This system automatically delivers
 
borated water to the reactor vessel for core cooling under high and low
 
reactor coolant pressure conditions. The Safety Injection System also serves
 
to insert negative reactivity into the core in the form of borated water
 
during an uncontrolled cooldown following a steam line break or an accidental
 
steam release. Other safety features which have been included in the
 
containment design are an Emergency Containment Cooling System which would
 
effect a rapid depressurization of the containment following a loss of
 
coolant and a Containment Spray System which would also depressurize the
 
containment. The Emergency Containment Cooling System provides backup
 
cooling for the Containment Spray System. 
 
1.2-1 Revised 02/14/2013 C26 1.2.1  STRUCTURES 
 
The major structures are two Containments, one Auxiliary Building, two Turbine Buildings and one Control Building. A general plan of the building arrangements is shown on Figure 1.2-1. Figures 1.2-2 through 1.2-7 show the general internal layout and equipment locations within the buildings. 
 
Each containment is a right vertical, post-tensioned reinforced concrete cylinder with pre-stressed tendons in the vertical wall, a reinforced and post-tensioned concrete hemispherical domed roof and a substantial base slab of reinforced concrete. The containment is designed to withstand environmental effects and the internal pressure and temperature accompanying a loss-of-coolant accident. It also provides adequate radiation shielding for both normal operation and accident conditions 
 
Seismic Classification of Particular Structures and Equipment
 
Particular structures and equipment are classified according to seismic design. The definition of seismic classifications is given in Appendix 5A.
 
1.2.2  NUCLEAR STEAM SUPPLY SYSTEM 
 
Each Nuclear Steam Supply System consists of a pressurized water reactor, Reactor Coolant System, and associated auxiliary fluid systems. The Reactor Coolant System is arranged as three closed reactor coolant loops connected in parallel to the reactor vessel, each loop containing a reactor coolant pump and a steam generator. An electrically heated pressurizer is connected to one of the loops.
 
The reactor core is composed of uranium dioxide pellets enclosed in Zircaloy-4, ZIRLO, Optimized ZIRLOŽ High Performance Fuel Cladding Material tubes with welded end plugs. The tubes are supported in assemblies by a spring clip grid structure. The mechanical control rods consist of clusters of stainless steel clad absorber rods and guide tubes located within the fuel assembly.
 
1.2-2 Revised 03/11/2016 C28 The core fuel is loaded in three regions. New fuel is introduced into the outer region, and partially spent fuel is moved inward into  a checkerboard
 
pattern at successive refuelings when the inner region is discharged to spent
 
fuel storage.
 
The steam generators are vertical U-tube units containing Inconel tubes.
 
Integral separating equipment reduces the moisture content of the steam at
 
the steam generator outlet to 1/4 percent or less.
 
The reactor coolant pumps are vertical, single stage, centrifugal pumps
 
equipped with controlled leakage shaft seals.
 
Auxiliary systems are provided to charge the Reactor Coolant System and to
 
add makeup water, purify reactor coolant water, provide chemicals for
 
corrosion inhibition and reactor control, cool system components, remove
 
residual heat when the reactor is shutdown, cool the spent fuel storage pool,
 
sample reactor coolant water, provide for emergency safety injection, and
 
vent and drain the Reactor Coolant System.
 
1.2.3  CONTROL SYSTEM 
 
Each reactor is controlled by a coordinated combination of chemical shim and
 
mechanical control rods. The control system allows the units to accept step
 
load changes of 10% and ramp load changes of 5% per minute over the load
 
range of 15 to 100% power under nominal operating conditions.
 
Supervision of both the steam supply and turbine generator systems is
 
accomplished from the control room shared by Units 3 and 4. The control room
 
layout including location of control boards for each unit is shown in Figure
 
7.7-1. 
 
1.2-3 Revised 04/17/2013 The control room is approximately 40' x 70' and has control boards arranged to give adequate distance between operator areas to preclude interference.
The annunciators and alarms for the two units are separated and have distinguishable audible tones.
 
The waste disposal control boards are located in the Auxiliary Building, and the radwaste facility building, and permit the operator to control and monitor the processing of wastes from locations adjacent to the equipment.
 
1.2.4  WASTE DISPOSAL SYSTEM 
 
The Waste Disposal System provides equipment necessary to collect, process, and prepare for disposal of potentially radioactive liquid, gaseous, and solid wastes produced as a result of reactor operation. 
 
Contaminated liquid wastes are collected and processed by plant filters and demineralizers. The effluents are sampled to determine residual activity and monitored during discharge to the cooling canal system via the condenser discharge to assure concentrations below 10CFR20 guidelines. The filters and spent resins from demineralizers are processed, temporarily stored and disposed of in accordance with applicable regulations currently in force.
Packaged low level waste may be stored on-site in the Low Level Waste Storage Facility while awaiting transport to off-site disposal area.
 
Gaseous wastes are collected and stored until their radioactivity level is low enough to permit discharge to the environment at concentrations below 10 CFR 20 guidelines. 
 
1.2.5  FUEL HANDING SYSTEM 
 
The reactor is refueled with equipment designed to handle spent fuel under water from the time it leaves the reactor vessel until it is placed in a cask for transport to the on-site Independent Spent Fuel Storage Installation (ISFSI) Facility, or shipment off-site. Underwater transfer of spent fuel provides an optically transparent radiation shield, as well as a reliable source of coolant for removal of decay heat. This system also provides capability for receiving, handling and storage of new fuel.
 
1.2-4 Revised 12/05/2014 C28 1.2.6  TURBINE AND AUXILIARIES
 
The turbine is a tandem-compound, 3-element, 1,800 rpm unit having 45-inch
 
exhaust blading in the low pressure elements. Four combination moisture
 
separator-reheater units are employed to dry and superheat the steam between
 
the high and low pressure turbine cylinders.
 
A twin-shell deaerating type condenser with semi-cylindrical water boxes
 
bolted to both ends, steam jet air ejectors, three condensate pumps, two 60%
 
capacity motor-driven boiler feed pumps, and six stages of feedwater heaters
 
are provided. Three auxiliary steam-driven feedwater pumps and two standby
 
steam generator feedwater pumps are available in case of a complete loss of
 
normal feedwater.
 
1.2.7  ELECTRICAL SYSTEM
 
The main generator is an 1,800 rpm, 3 phase, 60 Hz, hydrogen-cooled unit. 
 
The main step-up transformer is a conventional two-winding forced oil-air
 
cooled unit.
 
The Station Service System consists of startup, auxiliary and C Bus
 
transformers, 4160V switchgear, 480V load centers, 480V motor control
 
centers, 120V AC distribution panels and 125V DC equipment.
 
Emergency power is supplied by alternate sources including four emergency
 
diesel generators. The emergency diesel generators are capable of operating
 
equipment required for the normal shutdown of one unit plus the equipment
 
required for a postulated loss-of-coolant accident in the second unit
 
assuming a single failure.
 
1.2-5 Revised 04/17/2013  C26 1.2.8  ENGINEERED SAFETY FEATURES
 
The Engineered Safety Features provided have redundancy of component and power sources such that under the conditions of a hypothetical loss-of-coolant accident, the systems can, even when operating with partial effectiveness, maintain the integrity of the containment and keep the off site activity levels below the guidelines of 10 CFR 50.67.
 
The systems provided are summarized below:
 
a) The Containment System provides a highly reliable leak-tight barrier against the escape of fission products. The containment penetrations are provided with a leak-test system utilized to check the integrity of those locations which are the most likely sources of containment leakage. 
 
b) The Safety Injection System provides borated water to cool the core by injection into both cold and hot legs of the reactor coolant system.
 
c) The Containment Spray System provides a spray of borated water to cool and thus depressurize the containment after a loss-of-coolant or main steam line break accident. 
 
d) The Emergency Containment Cooling System provides a heat sink to cool and thus depressurize the containment after a loss-of-coolant or main steam line break accident. 
 
1.2.9  FIRE PROTECTION PROGRAM 
 
The Fire Protection program is described in Section 9.6.1 1.2.10 INDEPENDENT SPENT FUEL INSTALLATION (ISFSI)
An Independent Spent Fuel Storage Installation (ISFSI) has been constructed on the Turkey Point site to provide Unit 3 and Unit 4 spent fuel 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 Turkey Point spent nuclear fuel, high-level radioactive waste, and reactor-related Greater than Class C (GTCC) waste into dry storage casks. 
 
1.2-6 Revised 09/20/2016 C28 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 plant safe 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.
1.2.11 REFERENCES for SECTION 1.2.10
: 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.
 
1.2-7 Revised 09/20/2016 C28
 
FINAL SAFETY ANALYSIS REPORT FIGURE 1.2-1 REFER TO ENGINEERING DRAWING 5610-C-2 
 
REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 GENERAL BUILDING ARRANGEMENT PLAN FIGURE 1.2-1
 
FINAL SAFETY ANALYSIS REPORT FIGURE 1.2-2 REFER TO ENGINEERING DRAWING 5610-M-55
 
REV. 13  (10/96
)  FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 GENERAL ARRANGEMENT PLAN  EL. 10'-0" FIGURE 1.2-2
 
FINAL SAFETY ANALYSIS REPORT FIGURE 1.2-3 REFER TO ENGINEERING DRAWING 5610-M-56
 
REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 GENERAL ARRANGEMENT GROUND FLOOR PLAN EL. 18' -0" FIGURE 1.2-3
 
FINAL SAFETY ANALYSIS REPORT FIGURE 1.2-4 REFER TO ENGINEERING DRAWING 5610-M-57, SHEET 1
 
REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 GENERAL ARRANGEMENT OPERATING FLOOR PLAN EL. 42'-0" & EL 58'-0" FIGURE 1.2-4
 
FINAL SAFETY ANALYSIS REPORT FIGURE 1.2-5 REFER TO ENGINEERING DRAWING 5610-M-58
 
REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 GENERAL ARRANGEMENT MEZZANINE FLOOR PLAN AND SECTION "A - A" FIGURE 1.2-5
 
FINAL SAFETY ANALYSIS REPORT FIGURE 1.2-6 REFER TO ENGINEERING DRAWING 5610-M-59
 
REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 GENERAL ARRANGEMENT SECTIONS "B - B" AND "C - C" FIGURE 1.2-6
 
FINAL SAFETY ANALYSIS REPORT FIGURE 1.2-7 REFER TO ENGINEERING DRAWING 5610-M-60
 
REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 GENERAL ARRANGEMENT SECTIONS "D - D" & "E - E" FIGURE 1.2-7
 
FINAL SAFETY ANALYSIS REPORT FIGURE 1.2-8 REFER TO ENGINEERING DRAWING 5614-M-724
 
REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNIT 4 GENERAL ARRANGEMENT UNIT 4 EDG BUILDING PLAN AND SECTIONS FIGURE 1.2-8
 
1.3 GENERAL DESIGN CRITERIA The general design criteria define or describe safety objectives and
 
approaches incorporated in the design. These general design criteria are
 
addressed explicitly in the pertinent sections in this report. The remainder
 
of this section, 1.3, presents a brief description of related features which
 
are provided to meet the design objectives reflected in the criteria. The
 
description is developed more fully in those succeeding sections of the
 
report indicated by the references.
 
The parenthetical numbers following the section headings indicate the numbers
 
of the 1967 proposed draft General Design Criteria (GDC).
 
1.3.1 OVERALL REQUIREMENTS (GDC 1-GDC 5)
 
All systems and components of the facility are classified according to their
 
importance. Those items vital to safe shutdown and isolation of the reactor
 
or whose failure might cause or increase the severity of an accident or
 
result in an uncontrolled release of excessive amounts of radioactivity are
 
designated Class I. Those items important to operation but not essential to
 
safe shutdown and isolation of the reactor or control of the release of
 
substantial amounts of radioactivity are designated Class III.
 
Class I systems and components are essential to the protection of the health
 
and safety of the public. Quality standards of material selection, design,
 
fabrication and inspection conform to the applicable provisions of recognized
 
codes, and good nuclear practice.
 
All systems and components designated Class I are designed so that there is
 
no loss of capability to perform their safety function in the event of the
 
maximum hypothetical seismic ground acceleration acting in the horizontal and
 
vertical directions simultaneously. The working stress for Class I item is
 
kept within code allowable values for the design seismic ground acceleration.
 
Similarly, measures are taken in the design to protect against high winds,
 
sudden barometric pressure changes, flooding, and other natural phenomena. 
 
The Containment and Auxiliary Building are designed to withstand the effects
 
of a tornado.
 
1.3-1 Revised 04/17/2013 Reference sections:
 
Section Title Section Site and Environment; Meteorology, Seismology 2.6, 2.11 Reactor Coolant System; Design Bases 4.1 Containment Structure; Design Bases 5.1 Electrical System; Design Bases 8.1 Unit 4 Emergency Diesel Generator Building 5.3.4 Structures, Systems and Equipment Appendix 5A 
 
The fire protection program for the nuclear units is described in the below referenced section:
 
Reference section:
 
Section Title Section Fire Protection Program 9.6.1 Certain components of the Auxiliary, Emergency and Waste Disposal Systems are shared by Units 3 and 4. Certain components of shared equipment may be called upon to fulfill either an emergency, or emergency and shutdown function. The design and its evaluation supports the capability to deal with the affected unit, while maintaining safe control of the second unit.
 
A complete set of as-built drawings is maintained throughout the life of the units. A set of all the quality assurance data generated during fabrication and erection of the essential components is retained.
 
Reference sections:
 
Section Title Section Records 12.4 Initial Tests and Operation 13 Functional Evaluation of the Components of the Appendix A Systems which are shared by the two units
 
1.3-2 Revised 09/20/2016 C28 1.3.2  PROTECTION BY MULTIPLE FISSION PRODUCT BARRIERS (GDC 6-GDC 10)
 
The reactor core with its related control and protection system is designed
 
to function throughout its design lifetime without exceeding acceptable fuel
 
limits specified to preclude damage. The core design, together with reliable
 
process and decay heat removal systems, provides for this capability under
 
all expected conditions of normal operation with appropriate margins for
 
uncertainties and anticipated transient situations.
 
The Reactor Control and Protection System is designed to actuate a reactor
 
trip for any anticipated combination of plant conditions, when necessary, to
 
ensure a minimum Departure from Nucleate Boiling (DNB) ratio equal to or
 
greater than the safety analysis limit value.
 
Reference sections:
 
Section Title Section Reactor, Design Basis, Reactor Design 3.1, 3.2 Instrumentation and Control, Protective Systems 7.2 Safety Analysis 14 The design of the reactor core and related protection systems ensures that
 
power oscillations which could cause fuel damage in excess of acceptable
 
limits are not possible.
 
The potential for possible spatial oscillations of power distribution
 
for the first core has been reviewed. It was concluded that low frequency
 
xenon oscillations could have occurred in the axial dimension and part length
 
control rods were provided to suppress these oscillations. The core was
 
determined to be stable to xenon oscillations in the X-Y dimension. Excore
 
instrumentation is provided to obtain necessary information concerning
 
power distribution. This instrumentation is adequate to enable the operator
 
to monitor xenon induced oscillations. The part length control rods were
 
removed from the core after the first few cycles of operation. Their removal
 
was based on a determination that their presence was not required, since the
 
control banks provide adequate means for controlling the xenon oscillations
 
1.3-3 Revised 04/17/2013 Reference section:
 
Section Title Section Reactor Design, Nuclear Design and Evaluation 3.2.1 Reactor Coolant System Pipe Rupture 14.3 The Reactor Coolant System in conjunction with its control and protective
 
provisions is designed to accommodate the system pressures and temperatures
 
attained under all expected modes of operation or anticipated system
 
interactions, and maintain the stresses within applicable code stress limits. 
 
The materials of construction of the pressure boundary of the Reactor Coolant
 
System are protected by control of coolant chemistry from corrosion phenomena
 
which might otherwise reduce the system structural integrity during its
 
service lifetime.
 
System conditions resulting from anticipated transients or malfunctions are
 
monitored, and appropriate action is automatically initiated to maintain the
 
required cooling capability and to limit system conditions to a safe level.
 
The system is protected from overpressure by means of pressure relieving
 
devices, as required by Section III of the ASME Boiler and Pressure Vessel
 
Code.
 
Isolatable sections of the system are provided with overpressure relieving
 
devices to closed systems such that the system code allowable relief pressure
 
within the protected section is not exceeded. 
 
Reference sections:
 
Section Title Section Reactor Coolant System, Design Basis 4.1 The design pressure and temperature of the containment exceeds the peak
 
pressure and temperature occurring as the result of the complete blowdown of
 
the reactor coolant through any pipe rupture of the Reactor Coolant System up
 
to and including the hypothetical severance of a reactor coolant pipe.
 
1.3-4 Revised 04/17/2013 Piping systems which penetrate the vapor barrier are anchored at the
 
containment liner. The main steam, feedwater, blow down and sample line
 
penetrations are designed stronger than the piping system so that the
 
containment will not be breached due to a hypothesized pipe rupture. Lines
 
connected to the Reactor Coolant System that penetrate the containment are
 
provided with whip restraints and supports. These restraints and supports
 
are designed to withstand the thrust moment and torque resulting from a
 
hypothesized rupture of the attached pipe or the loads induced by the maximum
 
hypothetical earthquake.
 
Isolation valves are supported to withstand, without impairment of valve 
 
operability, the loading of the design basis accident or maximum hypothetical
 
seismic conditions.
 
Reference section:
 
Section Title Section Containment Structure 5.1 1.3.3  NUCLEAR AND RADIATION CONTROLS (GDC 11 - GDC 18)
 
The units are equipped with a control room which contains the controls and
 
instrumentation necessary for operation of the reactor and turbine generator
 
under normal and accident conditions.
 
Sufficient shielding, distance, and containment integrity are provided to
 
assure that control room personnel shall not be subjected to doses for the
 
duration of the hypothetical accident conditions during occupancy of, ingress
 
to and egress from the control room which exceed a small fraction of 10 CFR
 
50.67 guidelines.
 
Instrumentation and controls essential to avoid undue risk to the health and
 
safety of the public are provided to monitor and maintain within prescribed
 
operating ranges the neutron flux temperatures, pressure, flow, and levels in
 
the Reactor Coolant System, Steam Systems, Containment and Auxiliary Systems.
 
The quantity and types of instrumentation provided are adequate for safe and
 
orderly operation of all systems and processes over the full operating range
 
of the units.
 
1.3-5 Revised 04/17/2013 C26 The operational status of the reactor is monitored from the control room. 
 
When the reactor is subcritical the spontaneous neutrons from the irradiated
 
fuel are continuously monitored and indicated by proportional counters
 
located in the instrument wells in the primary shield adjacent to the reactor
 
vessel. The source detector channels are checked prior to operations in which
 
criticality may be approached. Any appreciable increase in the neutron
 
source multiplication, including that caused by the maximum physical boron
 
dilution rate, is slow enough to give ample time  to start corrective action
 
(boron dilution stop and/or emergency boron injection) to prevent the core
 
from becoming critical.
 
When the reactor is critical, means for showing the relative reactivity
 
status of the reactor is provided by control bank positions displayed in the
 
control room. Periodic samples of the coolant boron concentration are taken.
 
The variation in concentration during core life provides a further check on
 
the reactivity status of the reactor including core depletion.
 
Instrumentation and controls provided for the protective systems are designed
 
to trip the reactor, when necessary, to prevent or limit fission product
 
release from the core and to limit energy release; to signal containment
 
isolation; and to control the operation of engineered safety features
 
equipment.
 
During reactor operation in the startup and power modes, redundant safety
 
limit signals will automatically actuate two reactor trip breakers which are
 
in series with the control rod drive mechanism coils. This action would
 
interrupt power and initiate reactor trip.
 
Reference section:
 
Section Title Section Instrumentation and Controls 7.1, 7.2, 7.4, 7.7
 
If the reactor protection system receives signals which are indicative of an
 
approach to an unsafe operating condition, the system actuates alarms,
 
prevents control rod motion, initiates load cutback, and/or opens the reactor
 
trip breakers.
 
1.3-6 Revised 04/17/2013 The basic reactor operating philosophy is to define an allowable region of power and coolant temperature conditions. This allowable range is defined by
 
the primary tripping functions, the overpower T trip, overtemperature T trip, and the nuclear overpower trip. The operating region below these trip
 
settings is designed so that no combination of power, temperatures and
 
pressure could result in DNBR less than the safety analysis limit value with
 
all reactor coolant pumps in operation. Additional tripping functions such
 
as a high pressurizer pressure trip, low pressurizer pressure trip, high
 
pressurizer water level trip, loss of flow trip, steam and feedwater flow
 
mismatch trip, steam generator low-low level trip, turbine trip, safety
 
injection trip, nuclear source and intermediate range level trips, and manual
 
trip are provided to back up the primary tripping functions for specific
 
accident conditions and mechanical failures.
 
Rod stops from nuclear overpower, overpower T and overtemperature T deviation are provided to prevent abnormal power conditions which could
 
result from excessive control rod withdrawal initiated by a malfunction of
 
the reactor control system or by operator error. The overpower T and overtemperature T rod stop setpoints are the same as the reactor trip setpoints effectively negating these functions.
 
Reference sections:
 
Section Title Section Safety Injection Systems 6.2 Reactor Protection System 7.2 Positive indications in the control room of leakage of coolant from the
 
Reactor Coolant System to the containment are provided by equipment which
 
permits continuous monitoring of the containment air activity. Deviations
 
from normal containment environmental conditions including air particulate
 
activity, radiogas activity, and, in the case of gross leakage, the liquid
 
inventory in the process systems and containment sump, will be detected.
 
For the case of leakage from the containment under accident conditions the
 
area radiation monitoring system supplemented by portable survey equipment
 
provides adequate monitoring of releases during an accident.
 
1.3-7 Revised 04/17/2013 Monitoring and alarm instrumentation are provided for waste storage and fuel
 
handling areas to detect inadequate cooling and to detect excessive radiation
 
levels. Radiation monitors are provided to maintain surveillance over the
 
release of radioactive gases and liquids. 
 
A controlled ventilation system removes gaseous radioactivity from the
 
atmosphere of the fuel storage and waste treating areas of the auxiliary
 
building and discharges it to the atmosphere via the plant vent or the Unit 3
 
Spent Fuel Pool stack vent. Radiation monitors are in continuous service in
 
these areas to actuate high-activity alarms on the control board annunciator,
 
as described in Section 11.2.3.
 
Reference sections:
 
Section Title Section Leakage Detection 6.5 Auxiliary Coolant System 9.3 Radiation Protection 11.2 1.3.4  RELIABILITY AND TESTABILITY OF PROTECTION SYSTEMS (GDC 19-GDC 26) 
 
Upon a loss of power to the control rod drive mechanism coils, the full
 
length rod cluster control assemblies (RCCAs) are released and free fall into
 
the core. The reactor internals, fuel assemblies, RCCAs and pressure
 
retaining drive system components are designed as Class I equipment. The
 
RCCAs are fully guided through the fuel assembly and for the maximum travel
 
of the control rod into the guide tube. Furthermore, the RCCAs are never
 
fully withdrawn from their guide thimbles in the fuel assembly. As a result
 
of these design safeguards and the flexibility designed into the RCCAs,
 
abnormal loadings and misalignments can be sustained without impairing
 
operation of the RCCAs.
 
Protection channels are designed with sufficient redundancy for individual
 
channel calibration and test to be made during operation without degrading
 
the reactor protection system. Removal of one trip circuit for test is
 
accomplished by placing that channel in a tripped mode. For example, a
 
two-out-of-three logic becomes a one-out-of-two logic. Testing will not
 
cause a trip unless a trip condition exists in a concurrent channel. The
 
trip signal furnished by the two remaining channels would be unimpaired in
 
this event.
 
1.3-8 Revised 04/17/2013 In the Reactor Protection System, two reactor trip breakers are provided to
 
interrupt power to the RCCA drive mechanisms. The breaker main contacts are
 
connected in series (with the power supply) so that opening either breaker
 
interrupts power to all full length RCCAs permitting the RCCAs to free fall
 
into the core. Each breaker is opened through an undervoltage trip coil or a
 
shunt trip coil. Each protection channel actuates two separate trip logic
 
trains, one for each reactor trip breaker undervoltage trip coil. The
 
protection system is thus inherently safe in the event of a loss of rod
 
control power.
 
Channel independence is carried throughout the system extending from the
 
sensor to the relay actuating the protective function. The protective and
 
control functions when combined are combined only at the sensor. A failure
 
in the control circuit does not affect the protection channel.
 
The power supplied to the channels are fed from four instrument buses. All
 
four buses are supplied by inverters.
 
The initiation of the engineered safety features provided for loss-of-coolant
 
accidents is accomplished from redundant signals derived from reactor coolant
 
system and containment instrumentation. The initiation signal for
 
containment spray comes from the coincidence of two sets of two-out-of-three
 
high containment pressure signals. Upon loss of voltage on a 4160 volt bus,
 
the associated emergency diesel generator will be automatically started and
 
connected to the bus.
 
The components of the protection system are designed and arranged so that the
 
mechanical and thermal environment accompanying any emergency situation in
 
which the components are required to function does not interfere with that
 
function.
 
The signal conditioning equipment of each protection channel in service at
 
power is capable of being calibrated and tripped independently by simulated
 
analog input signals to verify its operation without tripping the reactor.
 
Each reactor trip channel is designed so that trip occurs when the circuit is
 
de-energized; an open circuit or loss of channel power causes the system to
 
go into its trip mode. In two-out-of-three logic, the three channels are
 
equipped with separate primary sensors and each channel is energized from an
 
independent electrical power supply. 
 
1.3-9 Revised 04/17/2013 The signal for containment isolation is developed from two-out-of-three logic
 
in which each channel is separated and independent. The failure of any
 
channel does not interfere with the proper functioning of the isolation
 
circuit.
 
Redundancy in emergency power is provided by four emergency diesel-generator
 
sets, each capable of supplying a separate 4160 volt bus. Each unit's A and
 
B train of engineered safety features is powered by a separate emergency
 
diesel generator. Manual swing train D can be powered by either emergency
 
diesel generator of the associated unit. This swing train powers redundant
 
engineered safety features.
 
Diesel engine starting is accomplished by compressed air supplied solely for
 
the associated emergency diesel generator. The undervoltage relay scheme is
 
designed so that loss of 4160 volt power does not prevent the relay scheme
 
from functioning properly.
 
The ability of the emergency diesel generator sets to start within the
 
prescribed time and to carry load can periodically be checked. The emergency
 
diesel generator breaker is not closed automatically after starting during
 
this testing. The generator may be manually synchronized to its associated
 
4160 volt bus for loading.
 
Reference sections:
 
Section Title Section Instrumentation and Control;  Protection Systems 7.2 1.3.5 REACTIVITY CONTROL (GDC 27-GDC 32)
 
In addition to the reactivity control achieved by the rod cluster control
 
assemblies (RCCAs) as detailed in Section 7, reactivity control is provided
 
by the Chemical and Volume Control System which regulates the concentration
 
of boric acid solution neutron absorber in the Reactor Coolant System. The
 
system is designed to limit the rate of uncontrolled or inadvertent
 
reactivity changes to a value which provides the operators sufficient time to
 
correct the situation prior to system parameters exceeding design limits. 
 
The reactivity control systems provided are capable of making and holding the
 
core subcritical from any hot standby or hot operating condition, including
 
those resulting from power changes.
 
1.3-10 Revised 04/17/2013 The RCCAs are divided into two categories comprising control and shutdown rod
 
groups. One control group of RCCAs is used to compensate for short term
 
reactivity changes at power such as those produced due to variations in
 
reactor power requirements or in coolant temperature. The chemical shim
 
control is used to compensate for the more slowly occurring changes in
 
reactivity throughout core life such as those due to fuel depletion and
 
fission product buildup and decay.
 
The shutdown groups are provided to supplement the control groups of RCCAs to
 
make the reactor at least one percent subcritical (k eff = 0.99) following a trip from any credible operating condition to the hot, zero power condition,
 
assuming the most reactive RCCA remains in the fully withdrawn position.
 
Any time that the reactor is at power, the quantity of boric acid retained in
 
the boric acid tanks and ready for injection will always exceed that quantity
 
required to support a cooldown to cold shutdown conditions without letdown. 
 
Under these conditions, adequate boration can be achieved simply by providing
 
makeup for coolant contraction from a boric acid storage tank and the
 
refueling water storage tank. The minimum volume maintained in the boric
 
acid storage tanks, therefore, is that volume necessary to increase the RCS
 
boron concentration during the early phase of the cooldown of each unit such
 
that subsequent use of the refueling water storage tank for contraction
 
makeup will maintain the required shutdown margin throughout the remaining
 
cooldown. In addition, the boric acid storage tanks have sufficient boric
 
acid solution to achieve cold shutdown for each unit if the most reactive
 
RCCA is not inserted.
 
Boric acid is pumped from the boric acid storage tanks by one of two boric
 
acid transfer pumps to the suction of one of three charging pumps which
 
inject boric acid into the reactor coolant. Any charging pump and either
 
boric acid transfer pump can be operated from diesel generator power on loss
 
of offsite power. Boric acid can be injected by one pump at a rate which
 
takes the reactor to hot standby with no rods inserted in less than forty
 
minutes when a feed and bleed process is utilized (less than 30 minutes when
 
the available pressurizer volume is utilized). In forty additional minutes,
 
enough boric acid can be injected to compensate for xenon decay although
 
xenon decay below the equilibrium operating level does not begin until
 
approximately 15 hours after shutdown. If two boric acid pumps and two
 
charging pumps are available, these time periods are reduced. Additional
 
boric acid injection is employed if it is desired to bring the reactor to
 
cold shutdown conditions.
 
1.3-11 Revised 04/17/2013 The Reactor Protection System is capable of protecting against any single anticipated malfunction of the reactivity control system and is designed to 
 
limit reactivity transients to DNBR equal to or greater than the safety
 
analysis value due to any single malfunction in the  deboration controls.
 
Limits, which include considerable margin, are placed on the maximum
 
reactivity worth of control rods and on rates at which reactivity can be
 
increased, to ensure that the potential effects of a sudden or large change
 
of reactivity cannot: (a) rupture the reactor coolant pressure boundary; or
 
(b) disrupt the core, its support structures, or other vessel internals so as
 
to lose capability to cool the core.
 
The control rod cluster drive mechanisms are wired into preselected groups,
 
and are therefore prevented from being withdrawn in other than their
 
respective groups. The control rod drive mechanism is of the magnetic latch
 
type and the coil actuation is sequenced to provide variable speed rod
 
travel. The maximum insertion rate is analyzed in the detailed plant
 
analysis assuming two of the highest worth groups to be accidentally
 
withdrawn at maximum speed, yielding reactivity insertion rates of the order
 
of 11 x 10
-4 k/sec which is well within the capability of the overpower-overtemperature protection circuits to prevent core damage.
 
Reference sections:
 
Section Title Section Reactor Design Bases 3.1 Protection Systems 7.2 Regulating Systems 7.3 Chemical and Volume Control System 9.2 1.3.6 REACTOR COOLANT PRESSURE BOUNDARY (GDC 33-GDC 36)
 
The reactor coolant boundary is shown to be capable of accommodating without
 
rupture, the static and dynamic loads imposed as a result of a sudden
 
reactivity insertion such as a rod ejection.
 
1.3-12 Revised 04/17/2013 The operation of the reactor is such that the severity of an ejection accident is inherently limited. Since RCCAs are used to control load
 
variations only and boron dilution is used to compensate for core depletion,
 
only the RCCAs in the controlling groups are inserted in the core at power,
 
and at full power these rods are only partially inserted. A rod insertion
 
limit monitor is provided as an administrative aid to the operator to ensure
 
that this condition is met.
 
By using the flexibility in the selection of control rod groupings, radial
 
locations and position as a function of load, the design limits the maximum
 
fuel temperature for the highest worth ejected rod to a value which precludes
 
any resultant damage to the system pressure boundary from possible excessive
 
pressure surges.
 
The failure of a rod mechanism housing causing a rod cluster to be rapidly
 
ejected from the core is evaluated as a hypothetical, though not a credible
 
accident. While limited fuel damage could result from this hypothetical
 
event, the fission products are confined to the Reactor Coolant System and
 
the containment.
 
The reactor coolant pressure boundary is designed to reduce to an acceptable
 
level the probability of a rapidly propagating type failure.
 
In the core region of the reactor vessel it is expected that the ductility of
 
the material will change as a result of exposure to fast neutrons. This
 
change is evidenced as a shift in the Reference Nil Ductility Temperature RT
 
(ndt) which is factored into the operating procedures in such a manner that
 
full operating pressure is not applied until the vessel material is well
 
above the RT(ndt).
 
The value of the RT(ndt) is increased during the life of the unit as required
 
by the expected shift in the RT(ndt), and as confirmed by the experimental
 
data obtained  from irradiated specimens of reactor vessel materials.
 
The design of the reactor vessel and its arrangement in the system permits
 
accessibility during the service life to the entire internal surfaces of the
 
vessel and to the following external zones of the vessel: the flange seal
 
surface, the flange O.D. down to the cavity seal ring, the closure head
 
except around the drive mechanism adapters and the nozzle to reactor coolant
 
piping welds. The reactor arrangement within the containment provides
 
sufficient space for inspection of the external surfaces of the reactor
 
coolant piping, except for the area of pipe within the primary shielding
 
concrete.
 
1.3-13 Revised 04/17/2013 Monitoring of the RT(ndt) properties of the core region plates, forgings, weldments and associated heat treated zones are performed in accordance with
 
the version of ASTM E185,"Recommended Practice for Surveillance Tests on
 
Structural Materials in Nuclear Reactors," required by 10 CFR 50, Appendix H.
 
Samples of reactor vessel plate materials are retained and catalogued in case
 
future engineering development shows the need for further testing.
 
The material properties surveillance program includes not only the
 
conventional tensile and impact tests, but also fracture mechanics tests. 
 
The observed shifts in RT(ndt) of the core region materials with irradiation
 
will be used to confirm the calculated limits of startup and shutdown
 
transients.*
 
To define permissible operating conditions below RT(ndt), a pressure range is
 
established which is bounded by a lower limit for pump operation and an upper
 
limit which satisfies reactor vessel stress criteria. Since the normal
 
operating temperature of the reactor vessel is well above the maximum
 
expected RT(ndt), brittle fracture during normal operation is not considered
 
to be credible.
* As of 2010, the program also includes revised initial weld material properties from Framatome ANP Topical Report BAW-2308, Revisions 1A and 2A
 
(References 2 and 3). The NRC letter dated March 11, 2009 (Reference 4)
 
approved these Topical Reports for use at Turkey Point.
 
Reference sections:
 
Section Title Section Reactor Coolant System
 
System Design and Operation 4.2 Tests and Inspections 4.4 Vessel RT(ndt)
Appendix 4A
 
1.3.7 ENGINEERED SAFETY FEATURES (GDC 37-GDC 65)
 
The design, fabrication, testing and inspection of the core, reactor coolant
 
pressure boundary and their protection systems give assurance of safe and
 
reliable operation under all anticipated normal, transient, and accident
 
conditions. However, engineered safety features are provided in the facility
 
to back up the safety provided by these components. These engineered safety
 
features have been designed to cope with any size reactor coolant pipe break
 
up to and including the circumferential rupture of any pipe assuming
 
unobstructed discharge from both ends.
 
1.3-14 Revised 04/17/2013 The release of fission products from the reactor fuel is limited by the
 
Safety Injection System which, by cooling the core and limiting the fuel clad
 
temperature, keeps the fuel in place and substantially intact and limits the
 
metal-water reaction to an insignificant amount.
 
For any rupture of a steam pipe and the associated uncontrolled heat removal
 
from the core, the Safety Injection System adds shutdown reactivity so that
 
with  a stuck rod, no off-site power and minimum engineered safety features,
 
there is  no consequential damage to the fuel or the primary system and the
 
core remains in place and intact.
 
The Safety Injection System consists of high and low head centrifugal pumps
 
driven by electric motors, and passive accumulator tanks which are self
 
energized and which act independently of any actuation signal or power
 
source. The release of fission products from the containment is limited in
 
three ways:
: 1. Blocking the potential leakage paths from the containment. This is accomplished by:
: a. A steel-lined, concrete containment with testable penetrations.
: b. Isolation of process lines by the Containment Isolation System which imposes double barriers in each line which penetrates the
 
containment.
: 2. Reducing the containment pressure and thereby limiting the driving potential for fission product leakage by cooling the containment
 
atmosphere using the following independent systems.
: a. Containment Spray System
: b. Emergency Containment Cooling System
 
A comprehensive program of testing is formulated for all equipment systems
 
and system control vital to the functioning of engineered safety features and
 
associated secondary components such as the main steam isolation valves and
 
the Auxiliary Feedwater System. The program consists of performance tests of
 
individual pieces of equipment in the manufacturer's shop, integrated tests
 
of the system as a whole, and periodic tests of the actuation circuitry and
 
mechanical components to assure reliable performance. In the event that one
 
of the components should require maintenance as a result of failure to
 
perform during the test according to prescribed limits, the necessary
 
corrections will be made and the unit retested.
 
1.3-15 Revised 04/17/2013 C26 The units are supplied with normal, standby and emergency power sources as follows:
: 1. The normal source of auxiliary power during operation is the generator and switchyard via the C Bus transformer. Power is supplied via the
 
unit auxiliary transformer which is connected to the isolated phase bus
 
of the generator and the C Bus transformer which is connected to the
 
switchyard. 2. Power required during startup, shutdown and after reactor trip is supplied from the plant switchyard via the startup and C-Bus
 
transformers which has multiple lines running to the transmission
 
system. 
: 3. One emergency diesel generator is connected to each of the safety related 4160V busses to supply emergency power in the event of loss of
 
offsite power. The emergency diesel generators are capable of
 
automatically supplying the engineered safety features load required for
 
any loss-of-coolant accident assuming any credible single failure.
: 4. Emergency power supply for vital instruments, for control and for emergency lighting is supplied from 125V DC batteries.
 
The 4160V bus arrangement and logic network provides the capability for
 
certain loads to be powered by either emergency diesel generator of the
 
associated unit following the failure of one diesel generator unit to start.
 
For engineered safety features as are required to ensure safety in the event
 
of an accident or equipment failure, protection is provided primarily by the
 
provisions which are taken in the design to prevent the generation of
 
missiles. In addition, protection is also provided by the layout of
 
equipment or by missile barriers in certain cases. 
 
Layout and structural design specifically protect safety injection piping
 
leading to unbroken reactor coolant loops against damage as a result of the
 
maximum hypothetical accident.  (However, dynamic effects of postulated
 
primary loop pipe ruptures have been eliminated from the Turkey Point design
 
basis based on the resolution of Generic Letter 84-04, "Asymmetric LOCA
 
Loads," in NRC letter dated November 28, 1988.)  Injection lines penetrate
 
the missile barrier, and the injection headers are located in the missile
 
protected area between the missile barrier and the containment wall. 
 
Individual injection lines, connected to the injection headers, pass through
 
the barrier and then connect to the loops.
 
1.3-16 Revised 04/17/2013 Movement of the injection line, associated with rupture of a reactor coolant
 
loop, is accommodated by line flexibility and by the design of the pipe
 
supports such that no damage outside the missile barrier is possible.
 
Each engineered safety feature provides sufficient performance capability to
 
accommodate any single failure of an active component and still function in a
 
manner to avoid undue risk to the health and safety of the public.
 
All active components of the Safety Injection System (with the exception of
 
some injection line isolation valves) and the Containment Spray System are
 
located outside the containment and not subjected to containment accident
 
conditions.
 
Instrumentation, motors, cables and penetrations located inside the
 
containment are selected to meet the most adverse accident conditions to
 
which they may be subjected. These items are either protected from
 
containment accident conditions or are designed to withstand, without
 
failure, exposure to the combination of temperature, pressure, radiation and
 
humidity expected during the required operational period.
 
The reactor is maintained subcritical following a reactor coolant system pipe
 
rupture accident. Introduction of borated cooling water into the core
 
results in a net negative reactivity addition. No credit is taken for control
 
rod insertion.
 
The delivery of cold safety injection water to the reactor vessel following
 
accidental expulsion of reactor coolant does not cause further loss of
 
integrity of the Reactor Coolant System boundary.
 
Design provisions are made to facilitate access to the critical parts of the
 
reactor vessel internals, injection nozzles, pipes, valves and safety
 
injection pumps for visual or boroscopic inspection for erosion, corrosion
 
and vibration wear evidence, and for non-destructive inspection where such
 
techniques are desirable and appropriate.
 
The design provides for periodic testing of active components of the Safety
 
Injection System for operability and functional performance. The safety
 
injection pumps can be tested periodically during operation using the full
 
flow recirculation lines provided. The residual heat removal pumps are used
 
every time the residual heat removal loop is put into operation, and can be
 
tested periodically on recirculation alignments.
 
1.3-17 Revised 04/17/2013 An integrated safeguards test can be performed during refueling outages prior
 
to heatup. This test would not introduce flow into the Reactor Coolant
 
System but would demonstrate the operation of the valves, pump circuit
 
breakers, and automatic circuitry upon initiation of safety injection. A
 
test is performed during refueling outages to demonstrate the ability to
 
introduce flow into the reactor coolant system.
 
The accumulator tank pressure and level are continuously monitored during
 
reactor operation.
 
The accumulators and the safety injection piping up to the final isolation
 
valve is maintained full of borated water at refueling water concentration
 
while the reactor is in operation. Flow in each of the hot and cold leg
 
injection headers lines and in the main flow line for the residual heat
 
removal pumps is monitored by a flow indicator.
 
The design provides for capability to test initially, to the extent
 
practical, the full operational sequence up to the design conditions for the
 
Safety Injection System to demonstrate the state of readiness and capability
 
of the system.
 
Tests are performed to provide information to confirm valve operating times,
 
pump motor starting times, the proper automatic sequencing of load addition
 
to the diesel-generators, and delivery rates of injection water to the
 
Reactor Coolant System.
 
The following general criteria are followed to assure conservatism in
 
computing the required containment structural load capacity:
 
a) In calculating the containment pressure, rupture sizes up to and including a double-ended break of reactor coolant pipe are considered.
 
b) In considering post-accident pressure effects, various malfunctions of the emergency systems are evaluated including failures of a
 
diesel-generator, an emergency containment cooler and a containment
 
spray pump.
 
c) The pressure and temperature loadings obtained by analyzing various loss-of-coolant accidents, when combined with operating loads and design
 
wind or seismic forces, do not exceed the load-carrying capacity of the
 
structure, its access openings or penetrations.
 
1.3-18 Revised 04/17/2013 The reinforced concrete containment is not susceptible to a low temperature
 
brittle fracture. The containment liner is enclosed within the containment
 
and thus is not exposed to the temperature extremes of the environment.
 
Typically, the containment bulk ambient temperature during operation is between 50F and 120F. Operation with elevated normal bulk containment temperatures up to 125F for short periods of time during the summer months has been evaluated (See Section 14.0). The material for the containment penetrations, which are designed to Subsection B of Section III ASME B&PV Code has a RT(ndt) of 0F. The reactor coolant pressure boundary does not extend outside of the
 
containment. Isolation valves for all fluid system lines penetrating the
 
containment provide at least two barriers against leakage of radioactive
 
fluids to the environment in the event of a loss-of-coolant accident. These
 
barriers, in the form of isolation valves or closed systems, are defined on
 
an individual line basis. In addition to satisfying containment isolation
 
criteria, the valving is designed to facilitate normal operation and
 
maintenance of the systems and to ensure reliable operation of other
 
engineered safety features.
 
After completion of the containment structure an initial integrated leak rate
 
test was conducted at the calculated peak accident pressure, to verify that
 
the leakage rate is not greater than 0.25 per cent by weight of containment
 
air per day.
 
Leak rate tests are performed during unit shutdowns periodically on a
 
frequency determined by the Containment Leakage Rate Testing Program in
 
accordance with the Technical Specifications.
 
Following the reactor vessel closure head replacement containment opening
 
closure, a Type A Integrated Leakage Rate Test, ILRT, was performed in
 
accordance with the requirements of 10 CFR 50 Appendix J, Technical
 
Specifications and station procedures. Containment measurements were made
 
before, during, and following the ILRT to demonstrate structural integrity. 
 
Containment structural inspections were performed in accordance with ASME
 
Boiler and Pressure Vessel Code, Section XI, Subsection IWE & IWL, 1992
 
Edition with 1992 Addenda.
 
1.3-19 Revised 04/17/2013 Capability is provided to the extent practical for testing the functional operability of valves and associated apparatus during periods of reactor
 
shutdown.
 
Initiation of containment isolation employs coincidence circuits which allow
 
checking of the operability and calibration of one channel at a time.
 
Design provisions are made to the extent practical to facilitate access for
 
periodic visual inspection of important components of the Emergency
 
Containment Cooling and Containment Spray Systems.
 
The containment pressure reducing systems are designed to the extent
 
practical so that the spray pumps, spray valves and spray nozzles can be
 
tested periodically and after any component maintenance for operability.
 
Test lines (2-inch permanent for mini-flow, a permanent 6-inch for full-flow
 
for Unit 3 and Unit 4) for all the containment spray loops are located so
 
that all components up to the containment isolation valves may be tested. 
 
The manual isolation valves are checked for leakage during local leak rate
 
testing.
 
The containment spray nozzles in containment are periodically verified to be
 
unobstructed by verification of air flow by use of thermography or other
 
appropriate means. 
 
Capability is provided to test initially, to the extent practical, the
 
operational startup sequence beginning with transfer to alternate power
 
sources and ending with near design conditions for the Containment Spray and
 
the Emergency Containment Cooling Systems, including the transfer to the
 
alternate emergency diesel-generator power source.
 
Reference sections:
Section Title Section  Containment 5.1 Engineered Safety Features 6 Electrical System 8.1, 8.2
 
1.3-20 Revised 04/17/2013 C26 1.3.8 FUEL AND WASTE STORAGE SYSTEMS (GDC 66-GDC 69)
The new and spent fuel storage racks are designed so that it is impossible to insert assemblies in other than prescribed locations. The cask area rack for
 
each unit is designed with a missing cell to provide a space for storage of
 
the long handling tool. On Unit 3 the missing cell is located on the
 
southeast corner of the rack and on Unit 4 the missing cell is located on the
 
northeast corner of the rack. Proper installation of the racks places the
 
missing cell on the east side of the pool wall. Administrative controls
 
ensure proper installation. Borated water is used to fill the spent fuel
 
storage pit at a concentration to match that used in the refueling cavity and
 
refueling canal during refueling operations. The fuel is stored vertically
 
in an array with sufficient center-to-center distance between assemblies to
 
assure keff <0.95 with a sufficient soluble boron concentration present.
Criticality of the fuel assemblies in the cask area rack is prevented by the
 
inherent design of the rack which limits fuel assembly interaction. This is
 
done by fixing the minimum separation between assemblies and inserting
 
neutron poison between the assemblies. Criticality is prevented in the
 
Region I and Region II Spent Fuel Racks by loading patterns which include
 
specific fuel categories based on enrichment, burnup, and cooling times and
 
by use of a combination of Rod Cluster Control Assemblies (RCCAs), water gaps
 
and Metamic inserts.
During reactor vessel head removal and while loading and unloading fuel from
 
the reactor, the boron concentration is maintained at not less than that
 
required to shutdown the core to a k eff = 0.95. This shutdown margin maintains the core subcritical, even if all control rods are withdrawn from
 
the core. Periodic checks of refueling water boron concentration ensure the
 
proper shutdown margin.
The design of the fuel handling equipment incorporates built-in interlocks
 
and safety features, the use of detailed refueling instructions and
 
observance of minimum operating conditions provide assurance that no incident
 
could occur during the refueling operations that would result in a risk to
 
public health and safety.
The refueling water provides a reliable and adequate cooling medium for spent
 
fuel transfer. Heat removal is accomplished with a Residual Heat Removal
 
Heat Exchanger. 
 
1.3-21 Revised 04/17/2013 Adequate shielding for radiation protection is provided during reactor refueling by conducting all spent fuel transfer and storage operations under water. This permits visual control of the operation at all times while maintaining low radiation levels, less than 15 mr/hr, for periodic occupancy of the area by operating personnel. Pit water level is alarmed in the control room and water to be removed from the pit must be pumped out as there are no gravity drains. Shielding is provided for waste handling and storage facilities to permit operation within guidelines of 10CFR20.
 
Gamma radiation is continuously monitored at various locations in the Auxiliary Building. A high level signal is alarmed locally and is annunciated in the control room.
 
Auxiliary shielding for the Waste Disposal System and its storage components was designed to limit the dose rate to levels not exceeding 0.5 mr/hr in normally  occupied areas, to levels not exceeding 2.5 mr/hr in periodically occupied areas  and to levels not exceeding 15 mr/hr in short specific occupancy areas. Actual dose rates may exceed these design values over time due to accumulation of hot particles, debris, other operational factors, etc.
 
All waste handling and storage facilities are contained and equipment designed so that accidental releases directly to the atmosphere are monitored and will not exceed the guidelines of 10CFR50.67; refer also to Section 11.1.2, 14.2.2 and 14.2.3.
 
The refueling cavity, refueling canal and spent fuel storage pit are reinforced concrete structures with a seam-welded stainless steel plate liner. These structures are designed to withstand the anticipated earthquake loadings as Class I structures.
 
Reference sections:
 
Section Title  Section Fuel Storage and Handling 9.5 Waste Disposal System  11.1 Low Level Waste Storage  11.1 Radiation Protection  11.2 
 
1.3-22 Revised 12/05/2014 C27 1.3.9 EFFLUENTS (10 CFR Part 50, Appendix A, Criterion 60)
 
Liquid, gaseous, and solid waste disposal facilities are designed so that discharge of effluents and off-site shipments are in accordance with applicable governmental regulations.
 
Radioactive fluids entering the Waste Disposal System are collected in sumps and tanks until determination of subsequent treatment can be made. They are sampled and analyzed to determine the quantity of radioactivity, with an isotopic identification if necessary. Before discharge, radioactive fluids are processed as required and then released under controlled conditions. The system design and operation are characteristically directed toward minimizing releases to unrestricted areas. Discharge streams are appropriately monitored and safety features are incorporated to preclude releases in excess of 10 CFR 20 guidelines.
 
Radioactive gases are pumped by compressors through a manifold to one of the gas decay tanks where they are held a suitable period of time for decay.
Cover gases in the nitrogen blanketing system are re-used to minimize gaseous wastes. During normal operation, gases are discharged intermittently at a controlled rate from these tanks through the monitored plant vent.
 
Filter cartridges and the spent resins from the demineralizers are packaged and stored on-site until shipment off-site for disposal. Low level waste may be stored in the Low Level Waste Storage Facility while awaiting shipment off-site for disposal.
 
Reference sections:
 
Section Title  Section Waste Disposal System  11.1 Low Level Waste Storage      11.1
 
1.3-23 Revised 12/05/2014 C27C27 1.4 DESIGN PARAMETERS AND UNIT COMPARISON The original design parameters of the Turkey Point Units 3 and 4 are
 
presented in tabular form along with the comparisons of the major parameters
 
from the final designs of the H. B. Robinson Unit 2, Indian Point Unit 2 and
 
Ginna plants. The purpose and evaluation of the parameter differences from
 
the plant safety point of view among these plants are appended by reference
 
line number. Refer to Table 1.4-1. The design parameters in this table are
 
historical in nature and are not intended to describe the current design.
 
1.4.1 DESIGN DEVELOPMENTS SINCE RECEIPT OF CONSTRUCTION PERMIT 
 
Burnable Poison Rods
 
In order to reduce the dissolved poison requirement for control of excess
 
reactivity, burnable poison rods or integral burnable poisons are
 
incorporated in the core design so that changes in coolant density have less
 
effect on density of poison and the moderator temperature coefficient of
 
reactivity becomes less positive (See Section 3.2.1). 
 
Safety Injection System
 
A second high head safety injection system line and header has been added. 
 
This arrangement provides a redundant flow path for high head safety
 
injection water to the reactor coolant loops through the hot legs. To avoid
 
the possibility of steam binding due to injection into the hot legs early in
 
any LOCA transient when steam generators are still relatively hot, the valves
 
which control the flow paths to the hot legs are maintained closed by keeping
 
the motor circuit breakers locked open at the motor control centers. This
 
administrative control ensures that automatic or inadvertent manual actions
 
do not result in hot leg injection.
 
A valved cross-over in the residual heat removal pump discharge has been
 
added, with a valved by-pass around the residual heat exchangers. This is
 
used to maintain a constant flow through the residual heat removal loop and
 
to control cooldown.
 
1.4-1 Rev. 16  10/99 
 
An alternative path to the normal low head safety injection path is provided by MOV-872 using the RHR pumps. This alternative flow path is provided for
 
use in the long term post-LOCA operating mode after switchover to the cold
 
leg recirculation mode in the event a beyond-design basis passive failure
 
occurs in the normal low head flow path.
 
A fourth high head safety injection pump has been added to provide greater
 
flexibility for the system.
 
The power sources for the safety injection pumps were modified. Following
 
the modifications, each SI pump is powered by a separate emergency diesel
 
generator, therefore, the failure of an emergency diesel generator will only
 
result in the loss of one SI pump. Following this change, the operating unit
 
is required to have the two SI pumps associated with the unit and one SI pump
 
associated with the other unit operable to assure two SI pumps are operating
 
following a single failure.
 
Containment Sumps
 
The single post-MHA containment sump at the bottom of the reactor cavity with
 
two suction lines to the two residual heat removal pumps has been relocated
 
and increased to two individual 100% capacity sump suction inlets at
 
elevation 14'-0". Each of the two sump suction inlets provides suction to
 
its individual residual heat removal pump through a 14" diameter pipe.
 
Strainer assemblies are installed on elevation 14'-0". The water from the
 
strainer modules is piped to the 14" suction inlets. See Chapter 6,
 
Section 2.
 
Safety Injection System Trip Signal
 
The actuating signal for the Safety Injection System is any of the following
 
signals:
: a. Two out of three high containment pressure (approximately 10% design pressure).
: b. Two out of three low pressurizer pressure.
 
1.4-2 Revised 04/17/2013 C26C26
: c. Two out of three steam line differential pressure (between steam  generator header and main header) for any loop. 
: d. Two out of three high steam line flow in any steam line coincident with low T avg (2/3) or low steam line pressure (2/3). 
: e. Manually 
 
These signals Increase the initiation reliability and increase protection in
 
the case of a steam line rupture.  (See Sections 7 and 6.) 
 
Containment Spray System Signal
 
The actuating signal for the Containment Spray System is revised to operate
 
from two-out-of-three high and two-out-of-three containment high-high
 
pressure signal channels.  (See Sections 6 and 7.) 
 
Rod Stop and Reactor Trip on Startup
 
The automatic rod stop signal is actuated by an overpower or overtemperature
 
T, and by an intermediate range flux level setting as well as by a power range flux level, and the reactor trip signal on start-up is supplied by 
 
a high flux level setting.  (See Section 7.) 
 
Isolation of the Control and Protection Systems
 
Isolation of the entire control and protection systems is increased to 
 
include all channels except those for the pressurizer level and steam 
 
generator level.  (See Section 7.2) 
 
1.4-3 Rev. 3  7/85 Electrical System Design The voltage class of many safeguards motors was changed from 460 volt to 4000
 
volt, and they were connected to 4160 volt buses. The emergency diesel
 
generator power and voltage ratings for the original emergency diesel
 
generators were 2500 kW (continuous rating) and 4160 volt, respectively.
 
Two emergency diesel generators were originally connected to an emergency
 
power bus. This bus has been eliminated because a single failure on this bus
 
would prevent the emergency power from supplying the engineered safety
 
features equipment.
 
The reliability and capacity of the plant engineered safety features was
 
improved by: 
 
a) placing the engineered safety features equipment electrically closer to the offsite power supplies - namely to the startup transformers;
 
b) directly connecting the emergency diesel generators to the 4160 volt bus rather than by way of an emergency power bus;
 
c) providing emergency power cross-connections from the startup
 
transformers;
 
d) providing backup power connection from the C-Bus; and
 
e) adding two additional emergency diesel generators and dedicating two emergency diesel generators to each unit (one per emergency power
 
train). The existing emergency diesel generators were dedicated to
 
Unit 3 and the new emergency diesel generators were dedicated to  Unit 4. The new emergency diesel generators' power rating is 2874 kW (continuous rating).
 
1.4-4 Rev. 10  7/92 Auxiliary Coolant System Two component cooling headers provide a means to isolate certain passive failures (defined as a 50 gpm leak). A partition has been added to the component cooling surge tank. Each compartment is connected to one component cooling header. Following isolation of the headers, leakage in one header will not communicate through the tank to the intact header. Following addition of the CCW Head Tank, a leak in either header can reduce CCW system volume to the elevation of the CCW Surge Tank partition. Reduction of system inventory to that level will not affect the normal function of the CCW system as adequate inventory is retained to ensure that CCW pump NPSH requirements are satisfied in the non-leaking header.
 
Waste Disposal System
 
The waste disposal system has been designed as purely a waste process system, which includes demineralizers, monitor tanks, condensate tank and associated pumps. The system also includes equipment to prepare the waste for disposal.
(See Section 11.1.). The system also includes a Low Level Waste Storage Facility where low level waste may be stored while awaiting shipment to an off-site disposal facility.
 
Thermal Power Uprate
 
Appropriate sections of the UFSAR have been revised to reflect thermal power uprate. The thermal power uprate increased the original rating of 2200 Mwt to 2300 Mwt.
 
Extend Power Uprate
 
An extend power uprate (EPU) has been performed to increase the thermal power rating form 2300 MWt to 2644 MWt. The appropriate UFSAR sections have been revised to reflect changes due to this change.
 
1.4-5 Revised 12/05/2014 C27 Low Level Waste Storage A Low Level Waste Storage Facility (LLWSF) is to be utilized to provide interim low level waste storage capabilities for both Units 3 & 4. Conservatively, both units could produce up to a combined total of 840 cu. ft. of Class B/C low level radioactive waste (LLW) per year. This amount would fill approximately seven (7) type 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 radioactive waste is licensed under the General License provided to power reactor licensees under 10 CFR Part 50.
 
1.4-6 Revised 12/05/2014 C27
 
TABLE 1.4-1      Sheet 1 of 15
 
COMPARISON OF DESIGN PARAMETERS
 
TURKEY POINT 
#3 OR #4 ROBINSON #2 INDIAN POINT #2 GINNA REFERENCE FINAL REPORT FINAL REPORT FINAL  REPORT  FINAL REPORT LINE NO.
 
THERMAL AND HYDRAULIC DESIGN PARAMETERS
 
Total Primary Heat Output, MWt 2200 2200 2758 1300  1
 
Total Core Heat Output, Btu/hr 7479 x 10 6 7479 x 10 6 9413 x 10 6 4437 x 10 6  2 Heat Generated in Fuel, % 97.4 97.4 97.4 97.4  3
 
Maximum Thermal Overpower 12% 12% 12% 12%  4
 
System Pressure, Nominal, psia 2250 2250 2250 2250  5
 
System Pressure, Minimum Steady Stats, psia 2220 2220 2220 2220  6
 
Hot Channel Factors Heat Flux, F 3.23 3.23 3.23 3.38  7
 
Enthalpy Rise, Fll 1.77 1.77 1.77 1.77  8
 
DNB Ration at Nominal Conditions 1.81 1.81 2.00 2.15  9
 
Minimum DNBR for Design Transients 1.30 1.30 1.30 1.30  10
 
Coolant Flow Total Flow Rate, 1b/hr 101.5 x 10 6 101.5 x 10 6 136.3 x 106 67.3 x 10 6  11  Effective Flow Rate for Ht Transfer,1b/hr 97.0 x 10 6 97.0 x 10 6 130. x 10 6 64.3 x 10 6  12  Effective Flow Area for Ht transfer,ft 2 41.8 41.8 51.4 27.0 13  Average Velocity Along Fuel Rods, ft.sec 14.3 14.3 15.4 14.7  14  Average Mass Velocity, lb/hr-ft 2 2.32 x 10 6 2.32 x 10 6 2.53 x 10 6 2.38 x 10 6  15 Coolant Temperatures, oF  Nominal Inlet 546.2 546.2 543 551.9  16
 
Maximum Inlet Due to Instrumentation
 
Error and Deadband, oF 550.2 550.2 547 555.9 17  Average Rise in Vessel, oF 55.9 55.9 53.0 49.5 18  Average Rise in Core 58.3 58.3 55.5 52  19  Average in Core 575.4 575.4 571.0 578.0  20  Average in Vessel 574.2 574.2 569.5 577.0  21  Nominal Outlet of Hot Channel 642 642 633.5 634.0  22 Average Film Coefficient, Btu/hr-ft 2-F 5400 5400 5790 5590 23 Average Film Temperature Difference, oF 31.8 31.8 30.3 26.9 24 Heat Transfer at 100% Power
 
Active Heat Transfer Surface Area, ft 2 42,460 42,460 52,200 28,715 25  Average Heat Flux, Btu/hr-ft 2 171,600 171,600 175,600 150,500 26  Maximum Heat Flux, Btu/hr-ft 2 554,200 554,200 567,300 508,700 27  Average Thermal Output, kw/ft 5.5 5.5 5.7 4.88  28  Maximum Thermal Output, kw/ft 17.9 17.9 18.4 16.5  29 TABLE 1.4-1 (Continued)  Sheet 2 of 15
 
TURKEY POINT 
#3 OR #4 ROBINSON #2 INDIAN POINT #2 GINNA REFERENCE FINAL REPORT FINAL REPORT FINAL  REPORT  FINAL REPORT LINE NO.
 
Maximum Clad Surface Temperature at
 
Nominal Pressure, oF 657 657 657 657 30 Fuel Central Temperature, oF  Maximum at 100% Power 4150 4030 4090 3880  31 Maximum at Overpower 4400 4300 4380 4100  32
 
Thermal Output, kw/ft at Maximum Overpower 20.0 20.0 20.6 18.5  33
 
CORE MECHANICAL DESIGN PARAMETERS
 
Fuel Assemblies Design RCC Canless RCC Canless 15x15 RCC Canless 15 x 15 RCC Canless 14 x 14 34 Rod Pitch, in. 0.563 0.563 0.563 0.556  35 Overall Dimensions, In. 8.426 x 8.426 8.426 x 8.426 8.426 x 8.426 7.763 x 7.763  36
 
Fuel Weight (as UO 2), pounds 176,000 175,400 216,000 118,727 37  Total Weight, pounds 225,000 225,400 276,000 150,750  38  Number of Grids per Assembly 7      7 9      9        39 Fuel Rods
 
Number 32,028 32,028 39,372 21,659 40  Outside Diameter, In.
32,028 32,028 39,372 21,659  41  Diametral Gap, mils 7.5,7.5,8.5 6.5,7.5,8.5 6.5  6.5    42  Clad Thickness, in.
0.0243 0.0243 0.0243 0.0243  43  Clad Material Zircaloy Zircaloy Zircaloy Zircaloy  44 Fuel Pellets
 
Material UO2 Sintered UO2 Sintered UO2 Sintered UO Sintered    45 Density (% of Theoretical) 94,93,92 94-92-91 94-92-91 92-90  46  Diameter, in. 0.3659,0.3659, 0.3659 0.3669 0.3669  47  0.3649 Length, in. 0.6000 0.6000 0.6000 0.6000 48 Rod Cluster Control Assemblies
 
Neutron Absorber 5% Cd-15% In-80% Ag. 5% Cd-15% In-80% Ag. 5% Cd-15% In-80% Ag 5% d-5% In-80% Ag  49 Cladding Material Type 304 SS-Cold Type 304 SS-Cold Type 304 Ss-Cold Type 304 SS-Cold Worked Worked Worked Worked 50  Clad Thickness, in.
0.019 0.019 0.019 0.019  51  Number of Clusters 45 53 61 29  52  Number of Control Rods per Cluster 20 20 20 16  53 Core Structure
 
Core Barrel I.D./O.D., in. 133.875/137.875 133.875/137.875 148.0/152.5 109.0/112.5  54 Thermal Shield I.D./O.D., in.
142.625/148.0 142.625/148.0 158.5/164.0 115.3/122.5  55
 
FINAL NUCLEAR DESIGN DATA
 
Structural Characteristics
 
Fuel Weight (As UO 2), lbs. 176,000 175,400 216,000 118,727 56  Clad Weight, lbs 34,900 36,300 44,600 22,440  57  Core Diameter, in. (Equivalent) 119.5 119.5 132.5 96.5  58 TABLE 1.4-1 (Continued)  Sheet 3 of 15
 
TURKEY POINT 
#3 OR #4 ROBINSON #2 INDIAN POINT #2 GINNA REFERENCE FINAL REPORT FINAL REPORT FINAL  REPORT  FINAL REPORT LINE NO.
 
Core Height, in. (Active Fuel) 144 144 144 144  59
 
Reflector Thickness and Composition Top - Water plus Steel, in. 10 10 10 10  60 Bottom - Water plus Steel, in 10 10 10 10  61 Side - Water plus Steel, in. 15 15 15 15  62
 
H2O/U, (Cold Volume Ratio) 4.18 4.18 4.18 4.08  63 Number of Fuel Assemblies 157 157 193 121  64 UO2 Rods per Assembly 204 204 204 179  65 Performance Characteristics
 
Loading Technique 3 region, 3 region, 3 region, 3 region,  66  non-uniform non-uniform non-uniform non-uniform Fuel Discharge Burnup, MWD/MTU
 
Average First Cycle 13,000 13,000 14,200 14,126  67  Equilibrium Core Average 24,500 24,500 24,700 24,400  68 Feed Enrichments, w/o Region 1 1.85 1.85 2.2 2.44  69  Region 2 2.55 2.55 2.7 2.78  70  Region 3 3.10 3.10 3.2 3.48  71  Equilibrium 3.10 3.10 Control Characteristics
 
Effective Multiplication (Beginning of life)
 
Cold, No Power, Clean 1.180 1.180 1.257 1.188  72  Hot, No Power, Clean 1.138 1.138 1.199 1.137  73  Hot, Full Power, Xe and Sm Equilibrium 1.077 1.077 1.152 1.080  74 Rod Cluster Control Assemblies
 
Material 5% Cd-15% In-80% Ag 5% Cd-15% In-80% Ag 5% Cd-15% In-80% Ag 5% Cd-15% In-80% Ag 75 Number of RCC Assemblies 45 53 61 33  76  Number of Absorber per RCC Assembly 20 20 20 16  77  Total Rod Worth See Table See Table See Table 6.8%  78  3.2.1-3 3.2.1-3 3.2.1-3 Boron Concentrations
 
To shut reactor down with no rods
 
Inserted, clean(k eff=.99)  Cold/hot 1250 ppm/1210 ppm 1250 ppm/1210 ppm 1480 ppm/1370 ppm 1630 ppm/1580 ppm  79
 
To control at power with no rods inserted, clean/equilibrium xenon and samarium 1000 ppm/670 ppm 1000 ppm/920 ppm 1200 ppm/780 ppm 1470 ppm/1100 ppm  80 Boron worth, Hot 7.3 k/k 7.3 k/k 1% k/k / 89 ppm 1% k/k / 120 ppm  81 Boron worth, Cold 5.6 k/k 5.6 k/k 1% k/k / 72 ppm 1% k/k /  90 ppm  82 
 
TABLE 1.4-1 (Continued)  Sheet 4 of 15
 
TURKEY POINT 
#3 OR #4 ROBINSON #2 INDIAN POINT #2 GINNA REFERENCE FINAL REPORT FINAL REPORT FINAL  REPORT  FINAL REPORT LINE NO.
 
Kinetic Characteristics
 
Moderator Temperature Coefficient +0.3x10
-4to-3.5x10
-4 +0.3x10-4to-3.5x10 0.3x10-4 to +0.3x10-4to-3.5x10
-4 83  k/k/oF k/k -3.0x10
-4 k/k/oF    k/k/oF Moderator Pressure Coefficient -0.3x10
-6to3.4x10-6 -0.3x10-6to3.5x10-6 +0.3x10-6 to -0.3x10-6to3.5x10-6 84  k/k/psi k/k/psi +0.3x10
-6 k/k/psi    k/k/psi Moderator Void Coefficient +0.5x10
-3to-2.5x10
-3 +0.5x10-3to-2.5x10
-3 +0.03to-0.30 -0.10 to+0.30  85 k/k/ %void k/k/ % void k/g/cm k/g/cm Doppler Coefficient -1x10
-5to -1.6x10 1x10-5to-1.6x10 1.1x10-5to -1.0x10
-5to-1.6x10
-5 86  k/k/oF k/k/oF +1.8x10
-5 k/k/oF    k/k/oF REACTOR COOLANT SYSTEM - CODE REQUIREMENTS
 
Component Codes Reactor Vessel ASME III Class A ASME III Class A ASME III Class A ASME Class A  87
 
Steam Generator Tube Side ASME II Class A ASME III Class A ASME III Class A ASME III Class A  88 Shell Side AMSE III Class C ASME III Class C ASME III Class C ASME III Class A  89
 
Pressurizer ASME III Class A ASME III Class A ASME III Class A ASME III Class A  90
 
Pressurizer Relief Tank ASME III Class C ASME III Class C ASME III Class C ASME III Class C  91
 
Pressurizer Safety Valves ASME III ASME III ASME III ASME III  92
 
Reactor Coolant Piping USAS B31.1 USAS B31.1 USAS B31.1 USAS B31.1  93
 
PRINCIPAL DESIGN PARAMETERS OF THE
 
REACTOR COOLANT SYSTEM
 
Reactor Primary Heat Output, MWt 2200 2200 2758 1300  94 Reactor Primary Heat Output, Btu/hr 7508 x  10 6 7508 x  10 6 9413 x  10 6 4437 x 10 6  95 Operating Pressure, psig 2235 2235 2235 2235  96 Reactor Inlet Temperature 546.2 546.2 543 551.9  97 Reactor Outlet Temperature 602.1 602.1 596 601.4  98 Number of Loops 3 3 4 2  99 Design Pressure, psig 2485 2485 2485 2485  100
 
Design Temperature, oF 650 650 650 650 101 Hydrostatic Test Pressure (Cold), psig 3107 3107 3100 3110  102 Coolant Volume,including pressurizer,cu.ft. 9088 9088 12,600 6245  103 Total Reactor Flow, gpm 268,500 268,500 358,800 180,000  104 TABLE 1.4-1 (Continued)  Sheet 5 of 15
 
TURKEY POINT 
#3 OR #4 ROBINSON #2 INDIAN POINT #2 GINNA REFERENCE FINAL REPORT FINAL REPORT FINAL  REPORT  FINAL REPORT LINE NO.
 
PRINCIPAL DESIGN PARAMETER OF THE
 
REACTOR VESSEL
 
Material SA-302 Grade B, low SA-302 Grade B, low SA-302 Grade B, low SA-302 Grade B, low 105 alloy steel, inter- alloy steel, inter- alloy steel, inter- alloy steel, inter-nally clad with aus- nally clad with aus- ternally clad with  nally clad with aus-tenitic stainless tenitic stainless austenitic stainless tenitic stainless steel steel steel steel 
 
Design Pressure, psig 2485 2485 2485 2485  106
 
Design Temperature, oF 650 650 650 650 107 Operating Pressure, psig 2235 2235 2235 2235  108 Inside Diameter of Shell, in.
155.5 155.5 173 132  109 Outside Diameter Across Nozzles, in.
236 236 262 -7/16" 219 5/lo  110 Overall Height of Vessel & Enclosure
 
Heat, ft-in.
41-6 41-6 43' 9-11/16" 39' 1-5/16"  111 Minimum Clad Thickness, in.
5/32 5/32 7/32 5/32  112 PRINCIPLE DESIGN PARAMETERS OF THE 
 
STEAM GENERATORS
 
Number of Units 3 3 4 2  113 Type Vertical U-Tube with  Vertical U-Tube with Vertical U-Tube Vertical U-Tube  114 integral-moisture integral-moisture integral-moisture with integral-separator separator separator moisture separator Tube Material Inconel Inconel Inconel Inconel 115 Shell Material Carbon Steel Carbon Steel Carbon Steel Carbon Steel  116 Tube Side Design Pressure, psig 2485 2485 2485 2485  117 Tube Side Design Temperature, oF 650 650 650 650 118 Tube Side Design Flow, lb/hr 33.93 x 10 6 33.93 x 10 6 34.07 x 106 33.63 x 10 6  119 Shell Side Design Pressure, psig 1085 1085 1085 1085  120
 
Shell Side Design Temperature, oF 556 556 556 556 121 Operating Pressure, Tube Side,Nominal psig 2235 2235 2235 2235  122 Operating Pressure,Shell Side,Maximum,psig 1020 1020 1005 989  123 Maximum Moisture at Outlet at Full Load,% 1/4 1/4 1/4 1/4  124 Hydrostatic Test Pressure, Tube Side 3107 3110 3110 3110  125 (cold), psig
 
PRINCIPAL DESIGN PARAMETERS OF THE
 
REACTOR COOLANT PUMPS
 
Number of Units 3 3 4 2  126 Type Vertical, single Vertical, single Vertical, single Vertical, single  127 stage radial flow stage radial flow stage radial flow stage radial flow with bottom suction with bottom suction with bottom suction with bottom suction and horizontal and horizontal  and horizontal and horizontal discharge discharge discharge discharge Design Pressure, psig 2485 2485 2485 2485  128 Design Temperature, oF 650 650 650 650 129 Operating Pressure, Nominal, psig 2235 2235 2235 2235  130 
 
TABLE 1.4-1 (Continued)  Sheet 6 of 15
 
TURKEY POINT 
#3 OR #4 ROBINSON #2 INDIAN POINT #2 GINNA REFERENCE FINAL REPORT FINAL REPORT FINAL  REPORT  FINAL REPORT LINE NO.
 
Suction temperature, oF 546.5 546.5 556 551.9 131 Design Capacity, gpm 89,500 89,500 90,000 90,000  132 Design Head, ft 260 260 252 252  133 Hydrostatic Test Pressure (cold), psig 3107 3107 3110 3110  134 Motor Type A-C Induction A-C Induction A-C Induction A-C Induction single speed, single speed, single speed, single speed,  135 air cooled air cooled Motor Rating (nameplate) 6000 HP 6000 HP 6000 HP 6000 HP  136 PRINCIPAL DESIGN PARAMETERS OF
 
REACTOR COOLANT PIPING
 
Material Austenitic SS Austenitic SS Austenitic SS Austenitic SS  137 Hot Leg - I.D., in.
29 29 29 29  138 Cold Leg - I.D., in.
27-1/2 27-1/2 27-1/2 27-1/2  139 Between Pump and Steam Generator-I. D. in. 31 31 31 31  140 Design Pressure, psig 2485 2485 2486 2486  141 TABLE 1.4-1 Sheet 7 of 15
 
LINE ITEM COMPARISON
 
H. B. ROBINSON #2 - TURKEY POINT #3&#4 - INDIAN POINT #2 - GINNA
 
Line Item Notes 
: 1. Nominal reactor power level intermediate between Indian Point #2 and Ginna plants. Power level related to safety only in the ability to produce and remove the power in the core as designed.
: 2. Directly related to Item 1 by conversion
: 3. No change in the fraction of the total heat generated in the core. 
: 4. This limitation applies only to prevention of temperatures of the fuel rods and coolant corresponding to power in excess of this
 
overpower limit. As demonstrated by the detailed examination   
 
of the rod withdrawal accident at power, presented in Section
 
14, nuclear overpower can be 18 percent without exceeding this
 
limit.
The primary consideration in overpower protection is not the actual value of the trip set point but rather the error
 
allowances that make up the margin to trip. The set point is selected so that a minimum DNB ratio of 1.30 is maintained at the
 
condition of the maximum overpower when all errors are taken in
 
the adverse direction and with the most adverse pressure and
 
temperature allowed by the high T trip. The combination of these protection channels (variable high T and overpower) limit the range of allowable conditions to a region of temperature,
 
pressure and power which preclude DNB or core damage for credible
 
accidents.   
 
TABLE 1.4-1 Sheet 8 of 15 Line Item Notes    The error allowances are subject to verification by performance tests of the installed system. The errors due to drift and set
 
point reproducibility are errors quoted by many instrumentation 
 
manufacturers and are demonstrated in actual performance tests on
 
the equipment before shipment. The improved performance is
 
attributable to the use of a solid state system. 
 
The errors due to rod motion result from variations in axial flux distribution with rod motion. Because of this variation, ion
 
chamber reading at a given axial location may differ for the same
 
core average power level. These errors are reduced by the use of
 
long ion chambers with top and bottom detectors, each equal in
 
length to about one-half the core height. The detectors yield an
 
average reading over one-half the axial length.
 
5,6  The reactor coolant system design pressure for the four plants is 2500 psia. For all conditions the system pressure is limited by
 
code safety valves set to open at design pressure and sized to
 
prevent system pressure from exceeding code limitations. 
 
Equipment capabilities for overpressure protection are established
 
by the complete loss of load without an immediate reactor trip.
 
The maximum over-pressure for this transient is therefore a
 
function of the  safety valve capacity and the maximum pressurizer
 
surge rate and is not dependent on the value of the nominal
 
operating pressure.
TABLE 1.4-1 Sheet 9 of 15 Line Item Notes The operating pressure is selected to ensure that desired thermal conditions are maintained in the core. The operating pressure is
 
established and maintained between the upper and lower reactor
 
trip limits to permit transient variations in either direction
 
with the assistance of the Pressure Control System.
 
7,8  There are no significant differences among the hot channel factors. For a detailed discussion, see Section 3.2.2.
: 9. The differences in the DNBR at nominal conditions is  due to the slight differences in operation conditions.
: 10. Same for all plants.
: 11. The flow varies from plant to plant due to pump design and number of loops.
TABLE 1.4-1 Sheet 10 of 15 Line Item Notes 
: 12. The effective flow rate for heat transfer is essentially proportional to the total flow rate as determined by the core
 
geometry.
: 13. Effective flow area for heat transfer is determined by the mechanical design of fuel assemblies and core.
 
14-24  There are no significant changes for these parameters from the previous plants.
: 25. The active surface area is determined by the mechanical design of the core.
 
26-29  The heat transfer parameters are determined by the required heat output, the heat transfer surface area and the design peaking
 
factors for the core. They are related to clad integrity in that
 
these conditions must be within the capability of the fuel and
 
must also meet the thermal- hydraulic design criteria of DNB and
 
fuel temperature. Extensive experience indicates that no problem
 
exists at these thermal outputs.
: 30. Same for all plants.
 
31-32  The fuel central temperatures are not significantly different than those for the other plants. The temperatures are well below the UO 2 melting temperature of 4800 oF. 
: 33. The overpower linear power density is similar to that of Indian Point No. 2 and is still well within the fuel capability.
TABLE 1.4-1 Sheet 11 of 15 Line Item Notes 34-36  The fuel assembly design is not significantly changed with respect to type, rod pitch and overall dimensions.
: 37. The total amount of fuel utilized is primarily a function of the nominal power rating.
: 38. The total weight of each fuel assembly includes the  weight of the fuel, clad, grids, RCC guide tubes, and top and bottom nozzles.
: 39. The number of grids per assembly is primarily a function of the core length and the average coolant velocity along the fuel rods.
: 40. The total number of fuel rods is consistent with the fuel assembly design and number of fuel assemblies.
 
41-44  Same for all plants.
 
45-48  The design of the fuel pellets is not substantially different except that the pellets are reduced as a result of the use of
 
pressurized helium in the gap between pellets and cladding.
 
49-53  The rod cluster control design is the same for all four plants.
The number of RCC assemblies for each plant is determined based
 
upon the control requirements.
 
54-55  The core barrel and thermal shield diameters are consistent with the core diameter.
 
56-57  The same comments as for line items 37 and 38 apply here.
TABLE 1.4-1 Sheet 12 of 15 Line Item Notes
: 58. The core equivalent diameter is primarily a function of the nominal power rating.
 
59-62  Same for all plants.
: 63. The water to uranium ratio is equivalent to that of  Indian Point
#2 and Turkey Point #3 and #4. The Ginna ratio is slightly lower
 
because of the different fuel element geometry.
: 64. The number of fuel assemblies required is primarily a function of nominal power rating.
: 65. The number of fuel rods per assembly is primarily a function of core diameter and determined by use of 15 x 15 rather than 14 x 14
 
lattices. Any fuel assembly can be placed over an in-core   
 
instrumentation penetration and can accept a neutron flux probe.
: 66. The core loading procedures are the same.
 
67-68  The average first cycle and first burnups are not significantly different but are affected by the burnable poison.
 
69-71  The core enrichment requirements do not vary significantly among all plants.
 
72-74  The beginning-of-life effective multiplications are not significantly different.
 
75-77  The same comments as for Line Items 49, 50 and 51 apply here.
: 78. The total control rod worth is not significantly different.
 
TABLE 1.4-1 Sheet 13 of 15
 
Line Item Notes 79-82  The boron requirements for reactor shutdown and control are primarily a function of core life and temperature.
 
83-86  With the use of burnable poison, the moderator temperature coefficient is always negative throughout core life at power
 
operating conditions. The pressure coefficient, the moderator
 
void (density) coefficient and the Doppler coefficient are not
 
significantly different.
 
87-92  Section III of the ASME Boiler and Pressure Vessel Code is considered to be the better design guide because it has
 
significantly upgraded Section VIII and its associated Nuclear
 
Code Cases. It presents the latest skills in the analytical
 
techniques of pressure vessel design and improved knowledge of 
 
pressure vessel failure patterns.
: 93. The code requirements for piping design are the same for all plants.
94-95  Comments are the same as for Line Items 1 and 2
: 96. Comments are the same as for Line Items 5 and 6.
 
97-98  There is no significant change.
: 99. The number of coolant loops used are a function of the capability of the primary and secondary hardware.
 
100-101 The reactor coolant system design pressure and temperature are the same.
TABLE 1.4-1 Sheet 14 of 15 Line Item Notes
 
102. The hydro test pressure is essentially the same.
 
103. The reactor coolant system volume is primarily a function of the number of loops and the component arrangement, practically
 
proportional to the nominal power rating.
 
104. The same comment as for Line Item 11.
 
105-107 Same for all plants.
 
108. Same comment as for Line Items 5 and 6.
 
109-111 The physical dimensions of the reactor vessel are consistent with the core size.
 
112. The reactor vessel clad thickness is the same for all plants.
 
113-118 The steam generator design bases are the same. The number of generators is consistent with the number of coolant loops.
 
119. This is the value of reference line 11 divided by the number of loops for each plant.
 
120-122 Same for all plants.
 
123. The shell side maximum operating pressure corresponds to the steam pressure at no load.
 
124-125 Same for all plants.
TABLE 1.4-1 Sheet 15 of 15 Line Item Notes
 
126-130 The type of reactor coolant pump (shaft seal) and the design conditions are the same. The number of pumps is consistent with
 
the number of reactor coolant loops. Original seals have been
 
replaced with low-leakage seals per EC 280399 for Unit 3 and EC
 
280401 for Unit 4.
 
131. There is not significant change in the suction temperature to the pumps.
132. The pump capacities are practically the same.
 
133. The design head of the pumps meets the requirements of the component and piping pressure losses of each plant.
 
134. Same for all plants.
 
135-136 The type and design of the pump motors is the same.
 
137-140 The reactor coolant piping is essentially the same. The hot leg pipe diameter is designed to maintain the same flow velocity
 
limitation (<50 ft/sec) as used in the other plants. The pipe
 
between the steam generator and the pump is designed to meet the
 
allowable velocity limits at the pump inlet.
 
141. The piping design pressure is the same as that for other components of the Reactor Coolant System.
 
Revised 07/15/2016 C28 1.5  DESIGN HIGHLIGHTS The design of Turkey Point Units 3 and 4 is based upon proven concepts which
 
have been developed and successfully applied in the construction of
 
pressurized water reactor system. In subsequent paragraphs, a few of the
 
design features are listed which represent slight variation or extrapolations
 
from other units, such as San Onofre and Connecticut-Yankee, which were
 
operating at the time of the original license application.
 
1.5.1 POWER LEVEL
 
The license application power level of 2200 MWt was larger than the
 
capability of the Connecticut Yankee plant and represented a reasonable
 
increase over power levels of pressurized water reactors operating at the
 
time of the original Turkey Point license application. The capability of the
 
nuclear steam supply system (NSSS) to operate at an Extended Power uprate
 
core power level of 2644 MWt was verified in accordance with guidelines
 
contained in the NRC review Standard for Extended Power Uprates - RS-001
 
(Reference 1).
 
1.5.2 REACTOR COOLANT LOOPS
 
The Reactor Coolant System for the Turkey Point Units 3 and 4 consists of
 
three loops as compared with four loops for Connecticut-Yankee. The use of
 
three loops for the production of 2644 MWt requires an attendant increase in
 
the size and capacity of the Reactor Coolant System components such as the
 
reactor coolant pumps, piping and steam generators. These increases
 
represent reasonable engineering extrapolations of existing proven designs.
 
1.5.3 PEAK SPECIFIC POWER
 
The design rating (15 kw/ft)is slightly lower than that licensed in CVTR (17 kw/ft) and that of Saxton (19.1 kw/ft). The maximum overpower condition for EPU is 22.7 kw/ft (120%) compared to 20 kw/ft (118%) for CVTR.
 
1.5-1 Revised 04/17/2013 C26C26C26 1.5.4 FUEL ASSEMBLY DESIGN
 
The fuel assembly design incorporates the rod cluster control concept in a
 
canless assembly utilizing a grid spring to provide support for the 15 x 15
 
array of fuel rods. This concept incorporates the advantages of the Yankee
 
canless fuel assembly and the Saxton grid spring with the rod cluster control
 
scheme. Extensive out-of-pile tests have been performed on this concept and
 
operating experience is available from the San Onofre and Connecticut-Yankee
 
plants.
 
1.5.5 ENGINEERED SAFETY FEATURES
 
The engineered safety features provided are of the same types provided for
 
the Connecticut-Yankee plant augmented by borated water injection
 
accumulators. A Safety Injection System is provided which can be operated
 
from emergency on-site diesel power. An Emergency Cooling System is provided
 
for post-loss-of-coolant conditions. A Containment Spray System provides
 
cool, borated water spray into the containment atmosphere for additional
 
cooling capacity.
 
1.5.6 EMERGENCY POWER
 
In addition to the multiple ties to offsite power sources, four emergency
 
diesel generators are provided as emergency power supplies for the case of
 
loss of offsite power. The emergency diesel generators are capable of
 
operating sufficient safety injection and containment cooling equipment to
 
ensure an acceptable post-loss-of-coolant pressure transient for any credible
 
single failure.
 
1.5-2 Revised 02/14/2013 C26 1.5.7 EMERGENCY CONTAINMENT COOLING SYSTEM A cooling system is provided to reduce containment atmospheric pressure following a loss-of-coolant accident. The three cooling units (2 of 3 are required) can be operated from emergency on-site diesel power. 
 
1.
 
==5.8 REFERENCES==
: 1. RS-001, Revision 1, " Review Standard for Extended Power Uprates,"
December 2003.
: 2. Framatome ANP Topical Report BAW-2308, Revision 1A, "Initial RTndt of Linde 80 Weld Materials", Approved August 2005.
: 3. Framatome ANP Topical Report BAW-2308, Revision 2A, "Initial RTndt of Linde 80 Weld Materials", Approved March 2008.
: 4. Letter from Jason Paige, NRC, to Mano Nazar, FPL, "Turkey Point Units 3 and 4 - Exemption from the Requirements of 10 CFR part 50, Appendix G
 
and 10 CFR Part 50, Section 50.61 (TAC Nos. ME 1007 and ME 1008)",
 
March 11, 2010.
 
1.5-3 Revised 04/17/2013 C26C26C26C26C26 1.6  RESEARCH AND DEVELOPMENT ITEMS Research and development (as defined in Section 50.2 of the Code of Federal 
 
Regulations) was conducted regarding first cycle final core design details
 
and parameters, analytical methods for kinetics calculations, safety
 
injection (emergency core cooling) system, xenon stability, control systems
 
and capability of reactor internals to resist blowdown forces.
 
1.6.1  INITIAL CORE DESIGN
 
The detailed core design and thermal-hydraulics and physics parameters have
 
been finalized. The cycle one nuclear design, including fuel configuration
 
and enrichments, control rod pattern and worths, reactivity coefficients and
 
boron requirements are presented in Section 3.2.1 and the final thermal-
 
hydraulics design parameters are in Section 3.2.2. Section 3.2.3 presents
 
the fuel, fuel rod, fuel assembly and control rod mechanical design. The
 
core design incorporates fixed burnable poison rods (1) in the initial loading to ensure a negative moderator reactivity temperature coefficient at
 
operating temperature. This improves reactor stability and lessens the
 
consequences of a rod ejection or loss of coolant accident. The mechanical
 
design is presented in Section 3.2.3. Subsequent cycle specific values are
 
calculated and reviewed prior to each cycle and are presented in Appendices
 
14A and 14B.
 
1.6.2    DEVELOPMENT OF ANALYTICAL METHODS FOR REACTIVITY TRANSIENTS FROM  ROD EJECTION ACCIDENTS 
 
A control rod ejection accident is not considered credible, since it would 
 
require the failure of a control rod mechanism housing. Nevertheless, the 
 
reactivity, and associated pressure and temperature transients for this
 
accident have been analyzed. 
 
1.6-1 Rev. 16  10/99 Rod ejection analyses for this plant were originally performed using the CHIC-KIN code (2), which uses a point reactor kinetics model and a single channel fuel and coolant description. The CHIC-KIN code has been superseded
 
by the TWINKLE computer code (Reference 7) which solves multi-dimensional two
 
group transient diffusion equation using a finite differences technique. The
 
rod ejection analysis results are given in Section 14.2 of this report,
 
together with a brief description of the TWINKLE code. 
 
Results for ejection of the highest worth rod at both beginning and end of
 
core life and zero and full power are given in Section 14.2. These analyses
 
show that the temperature and pressure transients associated with a rod
 
ejection accident do not cause any consequential damage to the reactor
 
coolant system. 
 
The reactor core now contains fixed burnable poison rods or integral burnable
 
poisons. These, by allowing a reduction in the chemical shim concentration,
 
ensure that the moderator temperature coefficient of reactivity is always
 
negative at 100 percent power operating conditions. 
 
A positive moderator coefficient was expected at operating temperatures early
 
in the first fuel cycle in the original core design. The burnable poison
 
rods were borosilicate glass. Critical experiments have been conducted at
 
the Westinghouse Reactor Evaluation Center using rods containing 12.8 w/o
 
boron and Zircaloy clad UO 2 fuel rods, 2.27% enriched. These values are typical of this reactor also. These experiments showed that standard
 
analytical methods can be used to calculate the reactivity worth of the
 
burnable poison rods. The design basis and critical experiments are
 
described in reference (1). In-core testing completed in the Saxton reactor
 
has shown satisfactory performance of these rods. 
 
The consequences of a rod ejection accident are now lessened because the
 
moderator temperature coefficient of reactivity is mostly negative at
 
operating conditions. In addition, the effects of rod ejection are
 
inherently limited in this reactor in which boric acid chemical shim is
 
employed since the control rods need only to be inserted sufficiently to
 
handle load changes. 
 
1.6-2 Rev. 16  10/99 1.6.3 SAFETY INJECTION SYSTEM DESIGN 
 
The design of the safety injection system is essentially that proposed at the
 
time the construction permit was issued; that is, it includes nitrogen-
 
pressurized accumulators to inject borated water into the reactor coolant
 
system to rapidly and reliably reflood the core following a loss-of-coolant
 
accident. Additional analyses have been performed to demonstrate that the
 
accumulators in conjunction with other components of the emergency core
 
cooling system can adequately cool the core for any pipe rupture. These
 
analyses are presented in Section 14.3. The computer codes used for the
 
blowdown phase of the loss-of-coolant accident take into account the
 
accumulator injection. 
 
Research and development work has also been performed on the integrity of 
 
Zircaloy-clad fuel under conditions simulating those during a loss-of-coolant
 
accident. Under the conservatively evaluated temperatures predicted for the
 
fuel rods during loss-of-coolant accident, the clad may burst due to a 
 
combination of fuel rod internal gas pressure and the reduction of clad
 
strength  with temperature. Burst cladding could block flow channels in the
 
core, so that core cooling by the safety injection system would be
 
insufficient to prevent fuel rod melting. 
 
Rod burst experiments have therefore been conducted on Zircaloy rods. The 
 
results have been presented to the AEC in the Zion Station PSAR, Volume III.
 
Analytical studies with the amounts of flow blockage obtained from the clad 
 
rupture geometry observed to date show that rod bursting during a loss-
 
of-coolant accident does not preclude effective cooling of the core by the
 
Safety Injection System. 
 
1.6-3 Rev. 3-7/85 1.6.4  SYSTEMS FOR REACTOR CONTROL DURING XENON INSTABILITIES 
 
In the transition to large Zircaloy-clad-fuel cores, the potential of power 
 
spatial redistribution caused by instabilities in local xenon concentration
 
was created. 
 
Extensive analytical work has been performed on reactor core stability (3,4,5,6)
. These indicate that a core of this size may be unstable against axial power redistribution, but is stable against transverse power
 
oscillations. The reactor was therefore provided with instrumentation and
 
control equipment which would allow the operator to detect and suppress the
 
axial power oscillations. 
 
Part-length control rods were provided to control axial oscillations and to 
 
shape the axial power distribution. These were found to be not needed, used
 
nor assumed to be available to achieve reactor shutdown. Also, plant
 
operation at power was not allowed with part-length control rods. Their
 
removal does not cause any changes in the required reactor characteristics,
 
nor safety margins at full power, low power nor shutdown. Therefore, the
 
part-length control rods were removed and the manual control feature deleted.
 
In the event of axial power imbalance exceeding operating limits, various
 
levels of protection are invoked automatically. These include generation of
 
alarms, turbine power cutback and blocking of control rod withdrawal (Section
 
7.2). 
 
1.6-4 Rev. 3-7/85 1.6.5 BLOWDOWN CAPABILITY OF REACTOR INTERNALS 
 
The forces exerted on reactor internals and the core, following a loss-
 
of-coolant accident, were originally computed by employing the BLODWN-2
 
digital computer program developed for the space-time-dependent analysis of
 
multiloop PWR plants. The BLODWN-2 code has been superseded by the MULTIFLEX
 
code. This newer program, the models used and the results are discussed in 
 
Section 14.3.3. 
 
1.6-5 Rev 15 4/98 REFERENCES, Section 1.6 
: 1) Wood, P.M., Baller, E.A., et al, "Use of Burnable Poison Rods in Westinghouse Pressurized Water Reactors," WCAP 7113 (October 1967),   
 
NON-PROPRIETARY. 
: 2) Redfield, V.A., "CHIC-KIN...A Fortran Program for Intermediate and Fast Transients in a Water Moderated Reactor," WAPD-TM-479, (January 1,
 
1965). 
: 3) Poncelet, C.G. and Christie, A.M., "Xenon Induced Spatial Instabilities in Large Pressurized Water Reactors," WCAP-368O-20, (March 1968),
 
NON-PROPRIETARY. 
: 4) McGaugh, J.D., "The Effect of Xenon Spatial Variations and the Moderator Coefficient on Core Stability," WCAP-2983, (August 1966),
 
PROPRIETARY. 
: 5) Westinghouse Report, "Power Distribution Control in Westinghouse PWR's," WCAP-7208, (October 1968), PROPRIETARY. The NON-PROPRIETARY
 
version of this document is WCAP-7811. 
: 6) Westinghouse Report, "Power Maldistribution Investigations",
 
WCAP-7407-L, (January 1970), PROPRIETARY. 
: 7)  Risher, D.H. And Barry, R.F., "TWINKLE - A Multi-/dimensional Neutrons  Kinetics Computer Code," WCAP-7979-P-A (Proprietary), January 1975 and  WCAP-8028-A (non-Proprietary) January 1975. 
 
1.6-6    Rev. 16  10/99 1.7 IDENTIFICATION OF CONTRACTORS The information contained in this section pertains to the contractors who participated in the construction of Turkey Point Units 3 and 4. This information is for historical purposes only.
Turkey Point Units 3 and 4 are being supplied and constructed under two basic
 
agreements. The first is between the Westinghouse Electric Corporation and 
 
Florida Power & Light Company in which Westinghouse has agreed to furnish the
 
Nuclear Steam Supply Systems and associated auxiliary equipment, and the
 
turbine generators with accessories, and technical services. The second is
 
between the Bechtel Corporation and Florida Power & Light Company in which
 
the Bechtel Corporation agreed to perform all phases of construction in
 
accordance with the plans and engineering of Bechtel Associates. Bechtel
 
procures all materials to  complete the units.
 
Florida Power & Light Company reviews specifications, plans and engineering,
 
and inspects and approves the construction. 
 
Operation will be solely by Florida Power & Light Company using Westinghouse
 
and Bechtel advisory and consulting service. 
 
Florida Power & Light Company has engaged many consultants to conduct 
 
investigations and studies relative to the natural sciences and they are
 
listed in Section 2.1. Further, Southern Nuclear Engineering, Inc, has been
 
retained as a consultant on safety matters. 
 
1.7-1    Rev. 16  10/99 1.8  SAFETY CONCLUSIONS The safety of the public and operation personnel and reliability of equipment and systems have been the primary considerations in the design. The approach
 
taken in fulfilling the safety consideration is three-fold. First, careful 
 
attention has been given to the design so as to prevent the release of radioactivity to the environment under conditions which could be hazardous to the health and safety of the public. Second, the units have been designed so as to provide adequate radiation protection for personnel. Third, reactor systems and controls have been designed with a great degree of the redundancy
 
and with fail-safe characteristics. 
 
Based on the over-all design of the units including their safety features and the analyses of the possible incidents and of the hypothetical accident, it is concluded that Turkey Point Units 3 and 4 can be operated without undue risk
 
to the health and safety of the public. 
 
1.8-1 1.9  QUALITY ASSURANCE PROGRAM The following Section 1.9 of this updated FSAR is reflective of the Quality
 
Assurance Program applicable to the design, procurement, and construction of
 
systems, components, and structures of Turkey Point Units 3 and 4 and is
 
maintained here for completeness. Subsequent to the operating license,
 
Florida Power & Light has established and implemented a Quality Assurance
 
Program as described in the FPL Topical Quality Assurance Report which is in
 
compliance with the requirements of Appendix B to 10 CFR 50 and approved by
 
the NRC. Sections 1.9.3 through 1.9.7 represent historical descriptions of
 
the QA program in place during the construction of the Turkey Point Units.
 
The system, components, and structures to which the Topical QA Report program
 
is applicable were set forth in the Turkey Point Units 3 and 4 Q-List which
 
was approved by Florida Power & Light Nuclear Engineering Department. FPL
 
developed the Total Equipment Data Base (TEDB) in 1986 to expand the fields
 
in the Plant Q-List. The Plant Q-List and the TEDB have been concurrently
 
updated to reflect the latest as-built configuration. Both documents have
 
been used in parallel since the development of the TEDB in 1986. The TEDB
 
was not used as a sole source for design information until the Plant Q-List
 
was replaced with the TEDB in 1990. The TEDB contains as-built and approved
 
alternate information on a component level.
 
====1.9.1 PURPOSE====
 
The purpose of this program is to establish quality assurance requirements
 
for those systems, components, and structures, herein identified, which by
 
reason of their association with the safety requirement of the nuclear units
 
have had criteria and design bases established for them in the A.E.C. license
 
application. The program describes the organization, procedures, and actions
 
taken by Florida Power & Light Company and its consultants, contractors and
 
suppliers to assure that all applicable criteria and design bases have been
 
correctly translated into specifications, plans, and drawings, and that the
 
systems, components, and structures have been fabricated, erected, installed,
 
and constructed in accordance with the design requirements.
 
====1.9.2 APPLICABILITY====
 
The systems and structures to which this program is applicable are set forth
 
below. It is understood that such systems and structures include associated
 
1.9-1 Rev. 16  10/99
 
tanks, pumps, valves, piping, controls, instruments, supports, enclosures, wiring, and power supplies. In general these systems, components, and
 
structures have a vital role in the prevention or mitigation of the
 
consequences of accidents which could cause risk to the health and safety of
 
the public.
: 1. Reactor Coolant System Reactor vessel  Reactor vessel internals RCC assemblies and drive mechanisms  Steam generators Reactor coolant pumps Pressurizer and relief tank All reactor coolant piping, plus any other lines carrying reactor coolant under pressure
: 2. Containment System  Containment structure including polar crane Containment penetrations and cooling systems including personnel and equipment access penetrations All lines penetrating the containment, up to and in-cluding the first isolation valves
: 3. Main Steam and Feedwater Lines within the Containment
: 4. Main Steam Safety, Isolation and Atmospheric Dump Valves
: 5. New Fuel Storage Facilities
: 6. Auxiliary Feedwater System Auxiliary feedwater pumps and turbine drivers Condensate storage tank Steam, condensate and feedwater lines of auxiliary  feedwater system 
: 7. Emergency Diesel Generators, Day Tanks and Storage Tanks and Associated Starting Equipment
 
1.9-2 
: 8. Containment Polar Crane and Rail Support (Unloaded)
: 9. Refueling Water Storage Tanks
: 10. Emergency Containment Cooling
: 11. Intake Cooling Water Systems Intake structure and crane supports Intake cooling water pumps and motors Intake cooling water piping, from pumps to component cooling water heat exchanger inlets
: 12. Component Cooling System Component cooling heat exchangers Component cooling pumps and motors Residual heat removal pumps and motors (low-head safety  injection pumps)  Residual heat removal heat exchanges Component cooling surge tanks Component cooling head tank 
: 13. Spent Fuel Storage Facilities Spent fuel pit and racks Spent fuel pit cooling water pumps and motors Spent fuel pit heat exchangers Spent fuel pit demineralizer
: 14. Safety Injection System Containment spray pumps and motors Low-head safety injection pumps and motors (residual heat removal pumps)
High-head safety injection pumps and motors Containment spray headers  Accumulator system  Containment recirculation sumps
 
1.9-3 Revised 04/17/2013 C26C26
: 15. Chemical and Volume Control System Charging pumps  Volume control tank Boric acid blender Boric acid tanks Boric acid transfer pumps Boric acid filters  Heat exchangers Primary water storage tank
: 16. Fuel Transfer Tube
: 17. Motor-Driven Fire Pumps
: 18. Instrument Air System Dryers Receivers
: 19. Auxiliary Building Exhaust System
: 20. Control Building Ventilating System
: 21. Fuel Handling System
: 22. Vessel and Internals Lifting Devices
: 23. Electrical System 1.9.3  ORGANIZATION
 
Charts of the Turkey Point Quality Assurance organization are attached hereto
 
as Figures 1.9-1, 1.9-2 and 1.9-3. Responsibility for quality assurance
 
rests with Florida Power & Light Company's Vice President of Power Plant
 
Engineering and Construction. Reporting to him is the Manager of Power Plant
 
Engineering who is responsible for administration of all Florida Power &
 
Light Company power plant engineering functions. A project Manager has been
 
assigned to the Turkey Point Units No. 3 and 4 project. The 
 
1.9-4 Project Manager administers all detailed activities of the project and he is
 
responsible for review of all design documents such as drawings,
 
specifications, procedures, and the Final Safety Analysis Report. He is
 
assisted in his review by a full-time project engineer and by on-site quality
 
assurance engineers. He can call upon specialized engineering services such
 
as electrical, control, relay, cathodic protection, production, water
 
chemistry, and environmental as necessary. Quality assurance problems
 
referred to him from the field by Florida Power & Light quality assurance
 
engineers are handled directly with Bechtel's Project Engineer or with
 
Westinghouse's Project Manager. Should any matters not be resolved to his
 
satisfaction, the matter is taken up with the Manager of Power Plant
 
Engineering in the case of an engineering subject or with the Construction
 
Superintendent in case of a construction subject. The ultimate authority for
 
quality rests with Florida Power & Light Company's Vice President who can
 
implement any quality assurance measures required.
 
1.9-5 The effectiveness of the quality assurance program is continuously reviewed
 
by Florida Power & Light Company's Vice President through his executive
 
assistant.
 
Westinghouse Electric Corporation is responsible for performance of quality
 
control and quality assurance functions on components within its scope of
 
supply, the nuclear steam supply with its associated auxiliary systems. The
 
Westinghouse Quality Assurance Plan is given in Appendix 1A.
 
Bechtel Corporation as agent for Florida Power & Light Company is responsible
 
for quality control and quality assurance for those systems components and
 
structures within its scope of supply. Bechtel is responsible for assuring
 
that its system and structures are compatible with the nuclear steam supply
 
system. Bechtel is responsible for quality control and quality assurance in
 
the erection of the nuclear steam supply system, and these actions are
 
monitored by both Westinghouse and Florida Power & Light Company. As agent
 
for Florida Power & Light Company, Bechtel monitors the quality assurance
 
program of Westinghouse during shop fabrication of components.
 
Shop inspection by Bechtel is performed by the Inspection Department which is
 
an organization independent of engineering, project, and construction
 
departments. Inspection requirements are established by Bechtel design group
 
supervisors assigned to the projects.
 
Site quality control is the responsibility of the construction group through
 
the Job Engineer. Quality control is monitored by the Quality Assurance
 
Engineer who reports to the Project Engineer and is thus independent of
 
construction forces. Independent testing laboratories (such as Pittsburgh
 
Testing Laboratories in the case of concrete and rebar) perform testing and
 
inspection functions and report to the Quality Assurance Engineer.
 
The Welding Engineer reports to construction supervisors for work
 
assignments, but the technical requirements of his work are established by
 
Bechtel's Metallurgy and Quality Control Department, an organization which is
 
independent of all Bechtel construction divisions. The welding procedures,
 
for 
 
1.9-6
 
example, are established and qualified by this department. The Welding
 
Engineer's work is monitored by periodic visits from Bechtel's Chief Welding
 
Engineer, as well as by the Quality Assurance Engineer and Florida Power &
 
Light Company.
 
Florida Power & Light Company quality assurance engineers monitor all quality
 
control and quality assurance activities taking place on site. The quality
 
assurance engineers utilize checklists to guide the nature and extent of
 
monitoring requirements. Florida Power & Light Company monitors Bechtel
 
inspection staffing, documents, quality control procedures, tests, test
 
equipment calibrations, inspection and testing frequency, personnel
 
qualifications, material control, and storage and protection. All design
 
documents are received by them and are maintained for their use.
 
1.9.4  SCOPE
 
The quality assurance program includes procedures and activities in the
 
following areas:
: 1. Design and procurement
: a. Correct translation of regulations, criteria, and basis into detailed design
: b. Review of design  c. Design changes  d. Design interfaces  e. Procurement documents  f. Design documents  g. Document control 
: 2. Shop fabrication of purchased material  a. Inspection requirements  b. Inspection procedures  c. Acceptance criteria  d. Inspection reports 
: 3. On-site construction, erection, and installation    a. Materials control  b. Materials storage and protection
 
1.9-7 
: c. Inspection
: d. Testing
: e. Welding
: f. Calibration
: g. Special procedures and instructions  h. Records
: i. Inspection, test and operating status  j. Non-conforming materials
 
Procedures governing preoperational checkout and startup testing are
 
described in Section 13, as are procedures for operation including fueling.
 
1.9.5  DESIGN AND PROCUREMENT
 
Criteria, regulations, and design bases are translated into detailed designs
 
by engineers in various disciplines assigned to the project. Assignments of
 
engineers are made by Bechtel's Chief Mechanical, Civil, and Electrical
 
Engineers, who remain responsible for the technical content of the job. 
 
Designs are reviewed by design group supervisors assigned to the project and
 
by the Project Engineer. All design documents are transmitted to Florida
 
Power & Light Company for review and comment and by contract must have
 
Florida Power & Light Company approval before release for construction.
 
Design documents dealing with the interface between Bechtel-designed systems
 
and structures and Westinghouse-supplied systems and components are
 
transmitted by Bechtel to Westinghouse for review, comment, and approval. 
 
All Westinghouse design documents are sent to Bechtel for review and
 
transmittal to Florida Power & Light Company. All Florida Power & Light
 
Company comments on Westinghouse design documents are transmitted to
 
Westinghouse through Bechtel.
 
Bechtel design documents are also reviewed by specialty groups serving all
 
Bechtel projects, such as geology and soils engineering, and metallurgical
 
and materials engineering and the Scientific and Technical group. 
 
Containment design was reviewed by a Task Force in Bechtel's home office
 
responsible for conceptual design of containments on several Bechtel
 
projects.
 
1.9-8 
 
All nuclear project groups within the Bechtel Power and Industrial Division
 
review design problems together by means of meetings and information
 
exchanges. Nuclear project coordination is the responsibility of the
 
Division Manager of Engineering, Mr. Harvey Brush.
 
All construction is accomplish through the use of the above-mentioned design
 
documents. No deviations from the documents are permitted without
 
documentation of the change and submission for review and approval by the
 
Bechtel Project Engineer, Florida Power & Light Company, and Westinghouse. 
 
Document control is obtained through the periodic issuance of current print
 
registers, status of purchase orders-specifications, and equipment lists.
 
All specifications are reviewed by the inspection department for inspection
 
requirements. Procurement is made only from suppliers on Florida Power &
 
Light Company approved bidders list. In addition, suppliers must be approved
 
by the Bechtel Purchasing Department for current quality performance.
 
Design review meetings are held periodically between Bechtel, Westinghouse,
 
and Florida Power & Light Company to discuss design problems and resolve
 
interface responsibilities. Minutes of all meetings are published and
 
submitted for approval and review of all parties.
 
A procedure manual has been established for the job which prescribes all
 
documents, and transmitting, reviewing, and approving procedure manual also
 
exists, for Bechtel-Westinghouse documentation, transmittal, review and
 
approval.
 
1.9.6  SHOP FABRICATION QUALITY ASSURANCE
 
Bechtel as Florida Power & Light Company's agent performs periodic shop
 
inspection during fabrication of components within Bechtel's scope and
 
monitors Westinghouse shop quality assurance on components within their
 
scope. Bechtel's activities are conducted in accordance with "Inspection
 
Procedures - Bechtel Corporation Procured Items" and the Bechtel Shop
 
Inspection Manual. Reports of all inspections are made to the Bechtel
 
Project Engineer who transmits them for review to Florida Power & Light
 
1.9-9 
 
Company. Florida Power & Light Company maintains files of inspection reports
 
both in the General Office and in the field. All purchase orders require
 
Bechtel inspection release prior to shipment. Inspection requirements are
 
established by the project engineers and are required to include specific
 
acceptance standards. Bechtel inspectors are full-time Bechtel employees
 
operating on a regional basis and they do not perform expediting or other
 
work.
 
Shop inspection reports are forwarded to the Bechtel field Quality Assurance
 
Engineer who notes any unusual requirements such as work to be done at the
 
site to make the component comply with requirements.
 
1.9.7  ON-SITE CONSTRUCTION, ERECTION, AND INSTALLATION 
 
Quality assurance activities on-site are performed in accordance with the
 
Bechtel Quality Assurance Manual, and in accordance with many specifications,
 
procedures, and special instructions.
 
Material control is performed in accordance with the Bechtel Standard
 
Purchasing Procedure Manual. All receiving, checking, warehousing, records
 
and materials issuance is monitored by the Home Office purchasing department.
 
Incoming material is identified in accordance with coding instructions
 
contained in the procurement documents. The material is checked against the
 
specifications and a "Material Received Report" is prepared and forwarded to
 
Field Engineering. Field Engineering checks the material against the
 
specifications and verifies the prior receipt of all supporting certificates
 
and documentation. Non-complying material is specially marked. A QC-101,
 
102, or 103 form is prepared by Field Engineering and forwarded to the
 
Quality Assurance Engineer for checking and permanent filing.
 
Inspection and testing is performed in accordance with codes, manuals, or
 
special instructions depending on the subject matter. Inspector
 
qualifications and requirements are monitored by Florida Power & Light
 
Company. Inspection reports are checked by Florida Power & Light Company,
 
and the Quality Assurance Engineer, and permanently filed. Test equipment
 
calibration frequency is specified in procedures, calibration records are
 
filed, and calibration is monitored by Florida Power & Light Company. As 
 
1.9-10 as example of inspection, testing and documentation requirements on the job,
 
a quality assurance package for concrete is given in Appendix 1B listing
 
specifications, procedures, special instructions, and documentations.
 
Welding is performed in accordance with Bechtel Welding Standards, a document
 
issued by the Metallurgical Department in Bechtel's Home Office. This
 
document contains welding procedures, heat tracing procedures, and
 
qualification certificates. Welder qualification is performed in accordance
 
with the ASME code under the supervision of the Welding Engineer. Weld
 
inspection requirements are spelled out in a Welding Inspection Procedure
 
issued by Florida Power & Light Company, Bechtel, and Westinghouse. 
 
Documentation is maintained in the form of isometric drawings checked off,
 
inspection reports, radiograph reports, and radiographs.
 
Non-destructive testing is performed in accordance with applicable codes. 
 
Radiography is performed in accordance with Bechtel specifications prepared
 
specifically for the job.
 
For the Reactor Vessel Closure Head Replacement containment opening, welding was performed in accordance with welding standards developed by The Steam Generating Team specifically for the job. This includes welding procedures, qualification certificates and nondestructive testing. Welder qualification was performed in accordance with the ASME code under the supervision of the Project Welding Engineer.
 
Non-conforming material procedures include paint coding of rebar and tagging
 
of equipment found unsatisfactory at receiving inspection.
 
1.9-11  Revised 09/15/2005 
 
FLORIDA POWER & LIGHT COMPANY TURKEY POINT QUALITY ASSURANCE ORGANIZATION FIGURE 1.9-1
 
FLORIDA POWER & LIGHT COMPANY TURKEY POINT UNITS 3 & 4 QUALITY ASSURANCE ORGANIZATION FIGURE 1.9-2
 
FLORIDA POWER & LIGHT COMPANY TURKEY POINT UNITS 3 & 4 WESTINGHOUSE QUALITY ASSURANCE ORGANIZATION FIGURE 1.9-3
 
APPENDIX 
 
1A 
 
Westinghouse Power Systems Division 
 
Quality Assurance Plan 
 
1A-i WESTINGHOUSE PWR SYSTEMS DIVISION QUALITY ASSURANCE PLAN 
 
QUALITY ASSURANCE PLANNING
 
Purpose The Quality Assurance Plan of Westinghouse PWR Systems Division for the Nuclear Steam Supply System is set forth in this document. Its purpose is to describe
 
the procedures and actions used by Westinghouse to assure that the design, 
 
materials and workmanship employed in the fabrication and construction of 
 
systems, components and installations within the Westinghouse scope of responsibility in a nuclear power plant are controlled and meet all applicable
 
requirements of safety, reliability, operation and maintenance. 
 
This plan is a requirement for, but is not necessarily limited to, those 
 
components and systems of the plant having a vital role in the prevention or mitigation of the consequences of accidents which can cause undue risk to the health and safety of the public. These include Class I items as described by 
 
the Safety Analysis Report. 
 
Procedural Documents and Work Instructions
 
Written administrative and technical policies, procedures and instructions are
 
in use in Westinghouse to implement the Quality Assurance Plan. They are in 
 
formats appropriate to their applications, such as: 
 
Management Responsibility Statements Position Descriptions of Management and Professional Personnel Engineering Instructions Quality Assurance and Reliability Procedures Quality Control Notices Quality Control Plans Projects Procedures Purchasing Manual Procedures Construction Site Procedures 
 
1A-1 Technical and contractual information to assure effective implementation of  these policies and procedure is developed, documented and controlled through a
 
standard Westinghouse system which consists in part of: 
 
System Design Parameters Equipment Specifications Corporate Process Specifications Corporate Material Test Specifications Corporate Purchasing Department Specifications (including specifications for materials) 
 
Drawings Purchase orders Procedures are reviewed and revised on a continuing basis by the issuing 
 
authorities so that the procedures meet the needs for which they are intended.
 
Management reviews performance in accordance with these procedures to assure compliance. Independent audits, as described later, provide objective
 
assurance of both the adequacy of the procedures and compliance with them. 
 
ORGANIZATION
 
Organization Chart
 
Figure 1 shows the functional organization as related to quality assurance of 
 
the Westinghouse Nuclear Energy Systems Divisions, including the staff review and surveillance function of the Reliability Controls Group at the Westinghouse Corporate level. The Westinghouse organization provides the checks and
 
balances needed to foster an effective overall quality assurance program. 
 
The authority and responsibility of the manager of each activity on this 
 
organization chart is set forth in writing in an approved state of management 
 
responsibility. 
 
1A-2 The PWR Systems Quality Assurance department consists of four sections:
Mechanical Equipment, Pressure Vessels, Electrical, and Plant Quality
 
Assurance. The Quality Assurance department has responsibility for supplier
 
surveillance, audits of the nuclear steam supply scope at construction sites,
 
and quality assurance data feedback and analysis, as described elsewhere in
 
this Plan. Other Westinghouse divisions are organized for independence of a
 
quality assurance function, as shown in Figure 1. 
 
The corporate Director of Reliability Control, who reports to Westinghouse top
 
management through an organizational path independent of the Executive Vice 
 
President of Nuclear Energy Systems, is responsible for the surveillance and 
 
auditing of the quality assurance effort carried out by all the divisions in 
 
Westinghouse Nuclear Energy Systems. The Director of Reliability Control 
 
utilizes the services of the PWR Systems Quality Assurance department in 
 
carrying out audits of other activities in Nuclear Energy Systems. 
 
Functional Relationships, PWR Systems
 
PWR Systems Division id divided into a number of functional groups having both
 
direct and indirect responsibility for aspects of the design, fabrication and 
 
construction phases of the project. Close association and interchange of 
 
information at all levels exists among the functional groups. 
 
The table of Figure 2 illustrates the relationships among these groups. Figure
 
3 shows this information in flow chart form. For example, contractual 
 
requirements originate in Projects, and are distributed to Licensing and Reliability, system functional requirement groups, system design groups and the uipment design and procurement groups. It can be seen that all aspects of the roject are considered at each stage in the overall program, with the respective
 
lead functional group coordinating the efforts of the associated functional 
 
groups. 
 
1A-3 Figures 2 and 3 are intended to show graphically the overall quality assurance program. For the sake of clarity, variations among functional groups have not
 
been shown. Specifics of the functions are contained in the detailed 
 
documentation of the program. 
 
ASSURANCE OF DESIGN ADEQUACY
 
Specification of Technical Requirements
 
Engineering is responsible for designing or specifying equipment that conforms
 
to the requirements of the application for which it is intended. This responsibility includes the specification of quality control requirements that
 
will assure that the equipment will function as required in the system and 
 
plant. 
 
Systems Engineering designs the plant to meet functional, safety and regulatory
 
requirements. The component design engineers work closely with systems 
 
engineering to identify equipment limitations and to resolve functional requirements with equipment capabilities. The design of equipment also provides  for access to components for in-service inspection and maintenance as
 
required to assure continued integrity throughout the life of the plant. 
 
Written parameters are forwarded to component design engineers by systems engineering detailing the design requirements for the specific plant.
Equipment  Specifications or drawings are prepared by the component design engineers to cover these requirements. The term "Equipment Specification" as used in this  Quality Assurance Plan includes drawings when they are used instead of Equipment  Specifications. Detailed quality control requirements are specified in the Equipment Specification, or its references. Examples of these are nondestructive tests, acceptance standards, functional tests, and recording the  measured values of key characteristics. In the few cases when Equipment Specifications or design drawings are not used, the specific quality control requirements, tests and acceptance standards are identified in the
 
purchase order. 
 
1A-4 
 
Design Review for Compliance with Technical Requirements
 
Preliminary Equipment Specifications are reviewed within Westinghouse by
 
systems engineers, materials and process engineers, licensing engineers,
 
Quality Assurance, Projects, and others as required. These independent reviews
 
assure that Equipment Specifications meet systems requirements, conform to
 
established engineering standards, are adequate from a metallurgical and
 
welding point of view, meet all code requirements, satisfy all safety
 
requirements including those specified in safety analysis reports, contain
 
necessary quality control requirements, and conform with the customer's
 
contractual provisions. Written Engineering Instructions describe the
 
requirements of the review. 
 
Aspects of the equipment design that have an effect on that part of the plant 
 
design performed by the customer or architect-engineer are forwarded to them
 
for  their review. Customer or architect-engineer drawings which have an
 
effect on  the Westinghouse scope of supply are likewise sent to Westinghouse
 
engineers for  their review. 
 
Technical requirements are provided in the bid package to qualified suppliers
 
of components within the Westinghouse scope of responsibility. Suppliers' 
 
proposals responding to these bids are sent to engineering for review. The 
 
component design engineer evaluates the supplier's proposal for technical 
 
adequacy. He insists on sufficient functional design data to make an independent review of the supplier's design to assure that the equipment will 
 
meet all requirements. Consultants from the Westinghouse Research and 
 
Development Laboratory and outside experts are also used to review specific 
 
design features, as required. The component design engineer reviews how the 
 
supplier intends to meet the specified quality requirements. He reviews the 
 
proposed equipment for its capability to perform its function for the design 
 
life of the plant. 
 
Westinghouse does not permit exceptions in the proposal specifications that 
 
adversely affect the safety or reliability of the equipment. 
 
1A-5 Purchase requisitions prepared by the component engineer are the basis for 
 
purchase orders issued by Purchasing to suppliers. The purchase order is the
 
official contract document that covers the technical requirements in the form
 
of  the equipment specification. 
 
Purchase requisitions are reviewed by Component Engineering, System
 
Engineering, Projects and other functions, as necessary, to assure that
 
technical requirements have been transmitted correctly to suppliers of the
 
components. 
 
Purchase orders require suppliers to submit detail drawings, and manufacturing, inspection and test procedures as the work under the purchase order progresses.
 
This phase of the design is reviewed independently by Westinghouse component 
 
engineers. The written instructions for this phase are contained in an Administrative Specification and the Equipment Specification, which form part
 
of the purchase order. 
 
Formal Design Reviews
 
In addition to the routine reviews of technical requirements discussed above,
 
formal design reviews are conducted by the Reliability section on critical 
 
systems, subsystems and components to improve their reliability and to reduce 
 
fabrication, installation and maintenance costs. The design reviews are 
 
comprehensive, systematic studies by personnel representing a variety of 
 
disciplines who are not directly associated with the development of the
 
product. Specialists from other Westinghouse divisions and outside consultants
 
are used in the reviews as necessary. Information developed by the reviews is
 
recorded for evaluation and action by the cognizant design engineer. 
 
Not all equipment receives this formal design review. The design review
 
program is projected over a substantial period of time because of the
 
comprehensive nature of each review. Selection of equipment to be reviewed is
 
based on many considerations:  relation to safety, effect on plant performance
 
and availability, stage of design development, and others. 
 
1A-6 SUPPLIER QUALITY ASSURANCE 
 
Preaward Evaluation of Prospective Suppliers
 
Prior to considering a new supplier for placement of a purchase order, a 
 
supplier evaluation is conducted. This is done in accordance with a written 
 
check list. The results are documented in a report issued to management 
 
personnel of Purchasing, Engineering, Quality Assurance, and Projects. The 
 
evaluation is conducted by a team consisting of Purchasing, Engineering and 
 
Quality Assurance. Other personnel such as material and process engineers 
 
and manufacturing engineers participate as required. 
 
Considerations of the evaluation include: 
 
Previous experience with the supplier Physical plant facilities Quality control program and system Number and experience of design personnel Material control and raw material inspection In-process inspection Assembly and test capability Tool and gage control Special processes required Nondestructive testing Inspection and test equipment Records function Deficiencies in the supplier's organization or systems are resolved with the 
 
supplier's management prior to placing a purchase order. 
 
If an existing supplier does not maintain the quality level on Westinghouse 
 
orders, a similar team will review the supplier's problems and make 
 
recommendations to his management to correct the situation immediately.
 
When problems arise, Westinghouse specialists aid the supplier in specific 
 
1A-7 areas such as welding, manufacturing and nondestructive testing to resolve the problem. In this manner, Westinghouse assures the continued high level of 
 
supplier performance necessary to obtain the quality level required by the 
 
contract. 
 
Supplier Quality Control Requirements
 
Quality requirements that apply specifically to a component are contained in
 
the Equipment Specification. Requirements of a quality systems nature, not
 
peculiar  to a component, are contained in two standard documents. 
 
The first is entitled, "Administrative Specification for the Procurement of 
 
Nuclear Steam Supply System Components."  This document is applied in all 
 
component purchase orders. The Administrative Specification requires the 
 
supplier not only to manufacture equipment that conforms to purchase order 
 
requirements, but to assure himself and Westinghouse by means of appropriate 
 
inspections and tests that the equipment conforms to these requirements. The
 
quality control section of this specification contains specific requirements in
 
areas such as: 
 
Calibration of measurement and test equipment Control of drawings, specifications, procedures and other documents used in design or manufacture, and revisions to these documents 
 
Control and identification of material 
 
Maintenance of quality control records 
 
Test control through written test procedures and test records 
 
Nonconforming supplies, including identification and control to preclude further use 
 
The second document that specifies quality requirements is QCS-1, "Manufacturer's Quality Control Systems Requirements."  This document is
 
applied  to orders for more critical equipment such as components related to
 
safety. This document requires the supplier to maintain an adequate quality
 
control system. This specification meets the intent of Appendix IX of Section
 
III of  the ASME Boiler and Pressure Vessel Code in the area of 
 
1A-8 quality control system requirements. QCS-1 requires the following, among other
 
things:
Establishment and maintenance of a system for the control of quality that assures that all supplies and services meet all specification, drawing, and contract requirements. 
 
Application of the system to subcontracted items. 
 
Written procedures that implement the system. 
 
Qualification of personnel. 
 
Qualification and control of processes including welding, heat treating, nondestructive testing, quality audits and inspection techniques. 
 
Operation under a controlled manufacturing system such as process sheets, travelers, etc.
Written inspection plans for in-process and final inspection. 
 
Submittal of Inspection Check Lists for approval by Westinghouse; these check lists show inspection and test status.
Recording of results of each inspection operation. 
 
Repair procedures, with provision for Westinghouse approval of all procedures utilizing operations not performed in the normal manufacturing sequence. 
 
Written work and inspection instructions for handling, storage, shipping, preservation and packaging.
As required, inspection hold points are specified by Westinghouse in the Equipment Specification or elsewhere in the purchase order. These are points
 
of witness or inspection by Westinghouse beyond which work may not proceed
 
without approval by Westinghouse. 
 
Planning of Supplier Surveillance
 
Westinghouse PWR surveillance of suppliers during fabrication, inspection, 
 
testing and shipment of components is planned in advance and performed in 
 
accordance with written Quality Control Plans. These plans are prepared by 
 
Quality Assurance engineers and are based on the technical requirements of the
 
purchase order. The plans are reviewed and approved by engineering. 
 
The purpose of a Quality Control Plan is to provide planned guidance to the 
 
Quality Assurance field representative by (1) focusing attention on those 
 
1A-9 items which contribute most to quality and reliability, and (2) providing specific instructions for the witnessing, documentation, and acceptance of the
 
equipment, and for auditing to assure the supplier's compliance with all
 
quality  control requirements. The plan identifies the points during
 
manufacturing and  test that Quality Assurance intends to witness. 
 
The plan covers (1) the auditing of the supplier's quality control system and
 
operation procedures; (2) surveillance of key operations such as welding, 
 
nondestructive testing, production and nonoperating electrical testing; and (3)
 
inspection verification (for example, sampling review of radiographs, material
 
test reports, key dimensions, and operating electric tests). Special emphasis
 
is placed on the aspects of manufacture and inspection that most directly
 
affect performance of the equipment. Lead units of a new design get particular
 
attention in the supplier's shop by both Quality Assurance and Engineering
 
representatives. 
 
When surveillance is indicated, Quality Assurance develops a visit schedule 
 
depending on the supplier's performance. Visits are more frequent during the 
 
initial stages of manufacture, particularly to a new supplier, with frequency 
 
diminishing as the supplier demonstrates his capability. 
 
Surveillance of Suppliers
 
The purpose of Westinghouse surveillance of suppliers is to provide
 
Westinghouse management and customers first-hand objective assurance of
 
compliance with specified requirements. The principle followed is that the
 
supplier is responsible for inspecting and testing his product. The
 
Westinghouse field representative assures that the supplier has done this,
 
rather than attempting to perform the supplier's inspection for him or
 
duplicate the work he has done. The frequency and scope of Westinghouse
 
surveillance varies with criticality of equipment, supplier performance,
 
complexity of the component, and other factors. This determination is made by
 
Quality Assurance in conjunction with engineering. Quality Assurance Residents
 
are established as necessary. 
 
1A-10 surveillance is accomplished in accordance with Quality Control Plans. In 
 
addition, the field representative confirms on a continuing basis that the 
 
supplier's system is adequate to ensure that a quality product will be built. 
 
He sees that written instructions and procedures are kept current, that 
 
application of drawings and specifications is controlled, that corrective
 
action  is implemented, and that other necessary controls are effective. 
 
The Quality Assurance representative informs the supplier directly of problems
 
he discovers and obtains commitments to correct them. He brings these problems
 
to the attention of the supplier's management as required to obtain resolution.
 
Release of Equipment for Shipment
 
The Purchasing Administrative Specification requires the supplier to write a 
 
formal shipping release when he is satisfied that purchase order requirements 
 
have been met. When the Westinghouse Quality Assurance representative is 
 
satisfied that the equipment can be released for shipment, and after receipt of
 
the supplier's release, he prepares a Quality Control release form, and 
 
distributes copies to the supplier, buyer and engineer. The equipment can then
 
be released through normal engineering-purchasing channels for shipment. 
 
CONSTRUCTION SITE QUALITY ASSURANCE
 
Control of Site Work
 
Work on nuclear steam supply equipment, as performed by the construction 
 
contractor and subcontractors, is monitored for conformance to written 
 
procedures and specifications which cover areas such as receiving inspection, 
 
storage, cleanliness, erection, in-process and final inspection and quality 
 
control, and testing. Special processes such as welding, cleaning and 
 
nondestructive testing are performed in accordance with written procedures by 
 
qualified personnel. 
 
1A-11 During component installation, Westinghouse Nuclear Power Service monitors work on nuclear steam supply and engineered safeguards equipment, and on critical 
 
structures. Qualified personnel provide technical advice on various
 
disciplines  of construction such as welding, mechanical and electrical system,
 
instrumentation and control equipment, and start-up. 
 
Each man is responsible for overseeing that the Westinghouse nuclear steam 
 
supply equipment assigned to him is in good condition when received and that it
 
is stored, handled and installed properly according to applicable 
 
specifications, procedures, and manufacturers' instructions. Further, he 
 
verifies that the proper documents which record the critical actions and 
 
inspections associated with this work are prepared and filed. 
 
The headquarters Quality Assurance group consists of a staff organizationally 
 
separate from Nuclear Power Service. This group provides independent assurance
 
that quality-related activities are done in accordance with specifications and
 
procedures. Nuclear Power Service provides technical advise to the constructor
 
during critical operations. Personnel from headquarters audit site activities
 
and monitor records for adequacy. 
 
A procedure describes the system for identifying, reporting and obtaining 
 
disposition of nonconforming material, equipment or practices discovered at the
 
site. Nuclear Power Service personnel fill out a Field Deficiency Report to 
 
provide the cognizant engineering group with the information necessary for 
 
making proper and timely disposition of each problem. After the cognizant 
 
personnel make a disposition, it is noted on the Field Deficiency Report and 
 
returned to the field for action. Files of these reports are maintained to 
 
record all field deficiencies and to provide for long-term corrective action. 
 
Site personnel must discontinue work on the nonconforming equipment until 
 
disposition is made. 
 
1A-12 Qualification of Westinghouse Personnel Nuclear Power Service welding engineers are qualified to Level II as required
 
by the ASME Boiler and Pressure Vessel Code, Section III, Appendix IX. 
 
Nuclear Power Service personnel who advise and consult during the 
 
pre-operational and functional testing are graduates of the Westinghouse
 
Nuclear  Operator training program. 
 
QUALITY CONTROL RECORDS
 
The Administrative Specification described above requires suppliers to maintain
 
records for each test (nondestructive, electrical, performance) specified in
 
the purchase order. The record must show the test procedure, equipment and 
 
materials used, the acceptance standards applied, and the test results
 
obtained. The part or assembly tested, date of test, and test operator
 
identity is shown. 
 
The administrative specification and equipment specification also require 
 
maintenance of other records as required, such as material test reports, welder
 
qualifications, inspection records, etc. Records such as trip reports, 
 
deviation notices, and other quality-related documents form a part of the 
 
Quality Assurance records maintained by Westinghouse. 
 
Suppliers are required to maintain these records for specified periods, after
 
which they notify Westinghouse so a record file for the life of the plant can
 
be arranged. Suppliers are also required to transmit records to Westinghouse
 
as work is completed for added assurance of record availability. 
 
Records generated at the construction site are filed and maintained there. 
 
1A-13 NONCONFORMING MATERIAL, TREND ANALYSIS AND CORRECTIVE ACTION Deficiencies at Suppliers' Plants
 
The Administrative Specification and QCS-1, described above, contain specific
 
contractual requirements for controlling nonconforming material or workmanship.
 
The supplier must physically identify all material that does not conform to 
 
purchase order requirements and take necessary actions to preclude its further
 
use. All deviations are documented in writing and reviewed by engineering, 
 
quality control and other appropriate groups. First, consideration is given to
 
restoring the material to its specified condition or scrapping it. If that is
 
impractical, the deviation is considered from both an engineering and a quality
 
control point of view. If acceptable, the deviation is formally approved in 
 
writing by the cognizant engineer. A permanent file of these records is 
 
maintained. 
 
QCS-1 requires that the supplier's quality system provides for the Wi-cation
 
and evaluation of significant or recurring discrepancies and for alerting the
 
supplier's cognizant management to the need for corrective action. The
 
supplier must review corrective action for effectiveness and the need for
 
further action.
Deficiencies at the Construction Site
 
A written procedure provides for documented reporting of deficiencies found 
 
during plant construction. These reports are submitted by site engineering 
 
personnel to the cognizant engineering department. Like reports from
 
suppliers'  plants, these reports are reviewed for necessary action, formally
 
approved by  the cognizant engineer and permanently filed. 
 
1A-14 Trend Analysis and Corrective Action Plant Quality Assurance analyzes all deficiency data on Westinghouse-supplied 
 
equipment received from suppliers and from construction sites to determine 
 
patterns of occurrence by supplier, by component, or by process. With this as
 
a guide, Quality Assurance and cognizant engineers determine corrective actions
 
that are needed to prevent recurrence. This action is in addition to assuring
 
that the supplier or site personnel take corrective action of the individual 
 
deficiencies reported. 
 
AUDITS Suppliers' Plants
 
The Westinghouse audit function of suppliers is described in the section, 
 
"Supplier Quality Assurance", above. 
 
Construction Site
 
Plant Quality Assurance is responsible for conducting independent audits of 
 
Nuclear Steam Supply System work at the construction site to assure that proper
 
procedures and instructions are available and in use, and that adequate
 
controls  exist and are effective. Reports of audits are sent to top
 
management of the 
 
PWR Systems Division. 
 
Westinghouse Divisions
 
The Westinghouse Corporation has a formal audit procedure which applies to the
 
PWR System Division and all other divisions furnishing equipment or 
 
1A-15 services to the nuclear industry as well as other areas. The audit program is under the direction of the corporate Director of Reliability Control who is 
 
organizationally independent from the operating divisions. 
 
The purpose of the headquarters reliability control function is to provide an 
 
independent verification that the quality assurance programs of the
 
Westinghouse  divisions are effectively assuring that the product quality
 
complies with the  requirements of their customers and that the programs are
 
using the most 
 
effective approaches to prevent the manufacture of defective products. In 
 
addition, this group assists divisions in continually improving their quality 
 
control programs and provides help that may be required to institute the 
 
recommended improvements identified in the audits. 
 
Audits are performed of each division's quality assurance effort. An audit is
 
usually performed by a two or three man team, consisting of a member of the 
 
headquarters reliability control staff and the quality control manager of 
 
another division in the same product group as the division to be audited. 
 
The audit normally takes five days. The quality assurance systems and 
 
procedures that have been established by the division are reviewed to determine
 
if these systems and procedures are sufficient to provide an effective program.
 
Observations are then made to assure that the established systems and
 
procedures are being correctly followed. 
 
An oral presentation of the findings and conclusions of the audit is made to
 
the Division General Manager, Quality Assurance Manager, and other personnel 
 
affected by the audit findings. The items recommended for improvement in the
 
quality assurance program are presented as well as recommendations of
 
approaches for accomplishing these improvements. 
 
Following the audit, a written report containing the findings and
 
recommendations reviewed in the oral report is prepared and sent to the
 
attendees of 
 
1A-16 the meeting. In addition, a copy of the report is sent to the Vice President to whom the division reports and to the Corporate Director of Manufacturing.
 
This procedure assures that the attention of a high level of management is
 
directed  to actions needed to carry out the recommendations of the audit. 
 
The Division Manager is responsible for reviewing the audit report and for 
 
taking action to improve the Quality Assurance Program in those areas
 
identified  in the report as requiring improvement. In addition, the Vice
 
President sends a letter to each of his division managers after completion of
 
the audits for his group asking for the status of implementation of
 
corrective action for each item identified in the audit report. The answer
 
must be sent to the Vice President as well as the Corporate Reliability
 
Control Staff. This reply provides a basis for further follow by the
 
Corporate Reliability Control Staff to assure that the audit findings are
 
acted upon. 
 
1A-17 QUALITY ASSURANCE FUNCTIONAL ORAGANIZATION FOR WESTINGHOUSE NUCLEAR ENERGY SYSTEM FIGURE 1A-1
 
NUCLEAR ENERGY SYSTEM FUNCTIONAL GROUPS QUALITY ASSURANCE FLOW CHART FIGURE 1A-2 (Sheet 1 of 3)
 
NUCLEAR ENERGY SYSTEM FUNCTIONAL GROUPS QUALITY ASSURANCE FLOW CHART FIGURE 1A-2 (Sheet 2 of 3)
 
NUCLEAR ENERGY SYSTEM FUNCTIONAL GROUPS QUALITY ASSURANCE FLOW CHART FIGURE 1A-2 (Sheet 3 of 3)
 
NUCLEAR ENERGY SYSTEMS FUNCTIONAL GROUPS QUALITY ASSURANCE SCEMATIC FLOW DIAGRAM FIGURE 1A-3
 
APPENDIX 1B   
 
Quality Assurance Package for Concrete
 
1B-i QUALITY ASSURANCE PACKAGE FOR CONCRETE CONCRETE List of procedures (specifications, manuals, and forms) used for QC/QA Control.
 
Specs & Manuals  Description C-19 Forming, Placing, Curing and Finishing Concrete
 
C-20 Specifications for Concrete FC-2 Specification for performing Quality Control Materials testing and allied services SP-2 ACI Manual of Concrete Inspection
 
Quality Assurance for Concrete Placed for the Reactor Vessel Closure Head Containment Opening Closure:
All concrete placed for containment closure was in accordance with SGT procedures and specifications for concrete developed specifically for the job.
List of procedures (specifications, manuals, and forms) used for QC/QA Control.
Specs & Manuals  Description
 
QEP-11.03 SGT procedure for forming, placing, curing and    finishing concrete
 
7012-SPEC-C-003 SGT specification for concrete QC Forms  1. Concrete placement check card (attached).
: 2. Concrete placement Inspection Report (attached).
: 3. Concrete Placement and test Report (attached)
: 4. PTL concrete placement card (attached).
: 5. PTL report on physical tests of concrete coarse aggregates (attached).
: 6. PTL report on physical tests of concrete fine aggregates (attached).
: 7. PTL concrete temperature report (in field preplacement check-attached).
: 8. PTL daily concrete batch plant report (attached).
: 9. PTL report of test of 6"x 12" concrete cylinders (attached).
 
1B-1  Revised 09/15/2005 Special Instructions
: 1. Civil responsibilities for Unit 4 Containment Mat placement.
: 2. Concrete Batch Plant Technician's duties for Unit 4 Containment Mat  placement.
: 3. Quality Assurance guidelines for concrete placement inspectors for Unit 4 Containment Mat placement.
: 4. Turkey Point ready-mix driver instruction sheet for Unit #4 Containment Mat. 5. PTL supervision instructions to PTL technicians on records required for Unit 4 Containment Mat placement
: 6. PTL Field Technician guidelines for turbine Pedestal Unit #4 placement from QAE.
 
1B-1a  Revised 09/15/2005}}

Revision as of 03:12, 8 July 2018

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