ML102870302
| ML102870302 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 09/21/2010 |
| From: | Florida Power & Light Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| L-2010-212 | |
| Download: ML102870302 (133) | |
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 1.2-2 Structures and Equipment 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 System 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 1.3-6
1.3.4 Reliability
and Testability of 1.3-9 Protection Systems 1.3.5 Reactivity Control 1.3-12
1.3.6 Reactor
Coolant Pressure Boundary 1.3-14
1.3.7 Engineered
Safety Features 1.3-17 1.3.8 Fuel and Waste Storage Systems 1.3-26
1.3.9 Effluents
1.3-28
1.4 Design
Parameters and Unit Comparison 1.4-1
1.4.1 Design
Developments Since Receipt of 1.4-1 Construction Permit Burnable Poison Rods 1.4-1 Safety Injection System 1.4-1 Containment Sumps 1.4-2 Emergency Containment Filtering System 1.4-2 Safety Injection System Trip Signal 1.4-2
1-i Rev. 16 10/99
TABLE OF CONTENTS (Continued) Section Title Page
1.4.1 Containment
Spray System Signal 1.4-3 (Cont'd) Rod Stop and Reactor Trip on Startup 1.4-3 Isolation of the Control 1.4-3 and Protection Systems Electrical System Design 1.4.4 Auxiliary Coolant System 1.4-5 Waste Disposal System 1.4-5 Thermal Power Uprate 1.4-5
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
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 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 Rev. 16 10/99 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 Units 1 and 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 is designed to produce initially 2200 MWt, and is
capable of an uprated output of 2300 MWt. Each steam and power conversion
system, including its turbine generator, is designed to permit generation of
760 Mw of gross electrical power, with a corresponding uprated gross
electrical output of approximately 795Mw.
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 Rev. 14 2/97 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 100.
Section 15 makes reference to the Technical Specifications under which the
units are operated.
1-2 Rev. 3-7/85 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
1.1-1 Rev 8 7/90 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.
1.1-2 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 an Emergency Containment Filtering System which would remove
elemental iodine from the atmosphere by absorption 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
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
tubes with welded end plugs. The tubes are supported in assemblies by a
spring clip grid structure. The mechanical control rods consist of
1.2-2 clusters of stainless steel clad absorber rods and guide tubes located within the fuel assembly. 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 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 and disposed of in accordance
with applicable regulations currently in force.
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 shipment from the site. Underwater transfer of spent fuel provides an
optically transparent radiation shield, as well as a
1.2-4 Rev. 13 10/96 reliable source of coolant for removal of decay heat. This system also provides capability for receiving, handling and storage of new fuel.
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 60% capacity 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 Rev. 16 10/99
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 100.
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.
e) The Emergency Containment Filtering System provides a rapid cleanup of iodine from the containment atmosphere after a loss-of-coolant
accident.
1.2-6 Rev. 16 10/99 1.2.9 FIRE PROTECTION PROGRAM
The Fire Protection program is described in Appendix 9.6A.
1.2-7 Rev. 5 7/87
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,
1.3-1 Rev. 14 2/97
sudden barometric pressure changes, flooding, and other natural phenomena. The Containment and Auxiliary Building are designed to withstand the effects
of a tornado.
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 Appendix 9.6A
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.
1.3-2 Rev. 16 10/99 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 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
1.3-3 Rev. 16 10/99 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.
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.
1.3-4 Rev. 16 10/99 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.
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.
1.3-5 Rev. 16 10/99 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
100 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.
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
1.3-6 Rev. 16 10/99 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.
1.3-7 Rev. 16 10/99 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.
1.3-8 Revised 10/23/2006 C22C22C22 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.
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
1.3-9 Rev. 16 10/99 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.
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.
1.3-10 Revised 04/29/2005
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.
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.
1.3-11 Rev. 10 7/92
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.
1.3-12 Rev. 16 10/99
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.
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
1.3-13 Rev. 16 10/99 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.
1.3-13a Rev. 16 10/99 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.
1.3-14 Rev. 16 10/99 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).
1.3-15 Rev. 16 10/99 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.
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.
1.3-16 Rev. 16 10/99 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.
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.
1.3-17 Rev. 16 10/99
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.
1.3-18 Rev. 13 10/96
- 2. Reducing the fission product concentration in the containment atmosphere by filtration.
- 3. 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.
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.
1.3-19 Rev. 3 7/85
- 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. 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.
1.3-20 Rev. 16 10/99 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.
1.3-21 Rev. 16 10/99
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.
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.
1.3-22 Rev. 16 10/99
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.
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 50 F and 120 F. Operation with elevated normal bulk containment temperatures up to 125 F 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 0 F.
1.3-23 Rev. 16 10/99 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.
1.3-24 Revised 09/15/2005 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 Filtering 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 and temporary 6-inch for full-
flow) 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.
1.3-25 Revised 12/01/2006 C22 Reference sections:
Section Title Section
Containment 5.1
Engineered Safety Features 6
Electrical System 8.1, 8.2
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 k eff <0.95 with a sufficient soluble boron concentration present. Criticality of the fuel assemblies in the spent fuel 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.
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.
1.3-26 Revised 09/29/2005 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 10CFR100; 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
Radiation Protection 11.2
1.3-27 Revised 09/29/2005
1.3.9 EFFLUENTS
(GDC 70)
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.
Reference sections:
Section Title Section Waste Disposal System 11.1
1.3-28 Rev. 16 10/99
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
recirculation mode in the event a 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.
Emergency Containment Filtering System
Three emergency containment filtering units have been added. (See Section
6.3)
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 6/30/2008 C23
- 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.)
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.
1.4-5 Rev. 16 10/99
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 10 6 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, o F 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, o F 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, o F 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 32,0228 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 UO 2 Sintered UO 2 Sintered UO 2 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
H 2 O/U, (Cold Volume Ratio) 4.18 4.18 4.18 4.08 63 Number of Fuel Assemblies 157 157 193 121 64 UO 2 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 -4 to-3.5x10-4 +0.3x10-4 to-3.5x10-4 -0.3x10-4 to +0.3x10-4 to-3.5x10-4 83 k/k/o F k/k -3.0x10 -4 k/k/o F k/k/o F Moderator Pressure Coefficient -0.3x10 -6 to3.4x10-6 -0.3x10-6 to3.5x10-6 +0.3x10-6 to -0.3x10-6 to3.5x10-6 84 k/k/psi k/k/psi +0.3x10 -6 k/k/psi k/k/psi Moderator Void Coefficient +0.5x10 -3 to-2.5x10-3 +0.5x10-3 to-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 -5 to -1.6x10 1x10-5 to-1.6x10-5 -1.1x10-5 to -1.0x10 -5 to-1.6x10-5 86 k/k/o F k/k/o F +1.8x10-5 k/k/o F k/k/o F 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 10 6 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 - 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 o F.
- 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.
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.
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 a thermal uprate core power
level of 2300 MWt was verified in accordance with guidelines contained in the
Westinghouse Topical Report WCAP-10263 (Reference 1); this core power uprate
methodology was similarly followed by the North Anna, Salem, Indian Point 2, Callaway, and Vogtle plants for their core power upratings.
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 2300 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 is 22.0 kw/ft (118%) compared to 20 kw/ft (118%) for CVTR.
1.5-1 Rev. 16 10/99 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 and Filtering
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 Rev. 16 10/99
1.5.7 EMERGENCY
CONTAINMENT COOLING AND FILTERING SYSTEMS
Separate and independent cooling and filtering systems are provided to reduce
containment pressure and airborne fission products respectively in the
containment atmosphere following a loss-of-coolant accident. The three
cooling units and three filtering units (2 of 3 are required) can be operated
from emergency on-site diesel power.
1.
5.8 REFERENCES
- 1. Topical Report, WCAP-10263, "A Review Plan for Uprating the Licensed Power of a Pressurized Water Reactor," Westinghouse Electric Corporation.
1.5-3 Rev. 16 10/99
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 and Filtering Units
- 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 pump and motor Spent fuel pit heat exchanger 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 Rev. 16 10/99
- 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)
... "-: -':! ... ----:1 :(-. ---:t !!"!".
cC! o! JIlnt!: !. A
- c T c Sheet 2 of*) : :: :. 1 J J 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
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- 'iroeness Hodulu5I SIEVE SIZE c
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-319 1' . No. ,._ tto ** 8 10 .'00 30 SO No. 100 3/f1' No. ,. tlo. 8 No. 16 tto. 30 SO No. 100 3/8 11 NoC) It N041 8 Noo 16 Ho. 30 No. 50 No. 100 eel 1 FP'L, Attn. Werry .1 Attn. Hr. Wf11fbms 3 Client, Attn. Hr. Wescott 1 Attn. Hr. Battin :3 CH-:-r;t w f) H':I. Attn. Mr" 2 PG IB-7 cwa: ... _. " 100 95 .. 100 80Q100 SO... 8S 25"1' 60 1Qc...,30 2<;> 10 too 95",,;00 i SO.,. 8S 2S ... 60 . -10", 30 2;'" 10 100 95<>>100 80 .. 100 , 500 8S 25 ... 60 100 30 2 .... 10 Tf.STIIIG LABCiI. ATIII' John "'0 Harl1eeo-Vice Prelo '-PO"N ..... A BATCHERMAN
, SHIFT_. _____ _ PROJECT: -. Turkey Poi nt, Uni ts 3 & It, Job NO" S610' ---_.MIX SP£C.SLUMP
'.' -:. ---------- P LACEM ENT LO CATION/ELEVATl.Otl _______ -:-_________ -MAX. TEMP
- _____ _ JATCH No. , AIR TEMP. CEMENT TEt-tP. c.
TEMP. F.AGG. TEMP. WATER TEr'IP. ICE TEMP. #I ICE/C.V. #I CHILLED WATER/C.V. i T '. .' , -.......... <'" .. -* .E, * ,', ;-." :. ! -" -"</ .. -.... _--.-.--_ .. -.......... --_ .. _ .. --.----.. -----.-. .-"- ... -.. ---.- -MIXING TIJ1E/SATCH . -----------.....;...-----.f---.-.-..---+-----.-t--- ... -----_._--+-,--- TRUCK NO. CONC. TEHP PlAt-IT TEMP. @ DISCHARGE . SLUMP (KELlV BAll) ----________ __ .......... _+_--, I -----_._--..;.: __ ..... __ ._-1--. __ . ..::..-"-.,-----1-------+----- __ 1--------AIR CONTEllT I ----.. ----.. ---.. -.-...-.,....-.. . ............-- .... --+-----........ L,..".. ........ --------* .:.-.,----+-----.f------ WEIGHT' I APPEARANCE OF MIX 1 I I I I j I -.-----.---_-L- __ --J... __ --1 ___ .l....-__ _____ __L..;. ___ __ BatCh Plant TeChnician Page _of _ Respectfully submitted,' PITTSBURGH TESTING LABORATORY-John W. Her 11ee Vice President .; ... '. '.' . .;.' ," , ;.-. ;;.; .. .f . .: . '. . I --;.'. j J ** f" *.**. :."-. . . ". " , .... ':. ; ... \ ....... : ....... .,. .. :' CLIENT, PROJECT: '. ;Bechtel Corporation -, ,; .... , , '. . '0. Box "A"' . Florida City, Florfda . . I t r *..* .. .,.,: .... I!d f:*j*':.* Turkey Point, Units 3 & Job No. S6l0 .. . . --... ;" . 1'0"'" MA (j)' 0' . .. , ., . .. . .' .. PLANT ________________ __ : OATCHERMAN, ________ _ SHI FT _______ _ MI X CLASS, ____ '!"-_ PLACEI-'.HiT LOCATIOtl/ELEVATlOtl. _______ ....... ____________ .,'_-'_:. : '. _____ _ TICKET . " * .,j" .... -'.-..... 'DATCH' MOISTURE SLUMP" * ..,.. :*:'t\ !..!!1I".J!Q.!. NO.;' CU.YO. -RECORDATION + .. -HETER , "'ETER REMARKS'** .;:: ': ' . CEMENT_'_' ._ ... .. ----........ . ., ..... , '.-.. _,'. '.' .... ......... ' .. c. AGG. __ __ _ *f. AGG. (w l-/ATER -------;-.,.:\ ,: I.' .' RET'WL _________ _ AEA ___________ _ :,' ICE CEENT, ____ C. AGG. __ __ _ F. AGG. __ ..:(w:.. ___ _ WATER RET AEA: ____________ _ -----------------;l ... ..... '." c. AGG-. _-_-_-_-_-_- _-F. AGG. __ ... @--. __ _ WATER._------- RET'WL _________ _ AEA ___________ __ ICE CEfotEIIT __ ___ _ C. AGG. ___ _____ __ F. AGG. _____ ______ __ _____ _ RET'rlL AEA ------------- ICE DESIGN MIX: AGG. SAAP LED CEMENT F.M. C. AGG. FREE MOISTURE CA F. AGG. WATER FA RET'WL AEA CEMENT SAMPLED @ ICE FROt1 B8tch Plant Technician Page _of_ 1B-9 -------_ ..... , . *, r. ;'. WEATHER @ @ . Respectfully submitted, PITTSBURGH TESTING LABORATORY John W. Har llee . Vice President } .. i . . /' . '. * .. *t' .. ;-:-.... . (I.' '. , MARK 701' 701 781 781 7al 7al "', ". '.' " , :.. ',:. ... ;. ".,. '< '.'. '. ..
- JOB NO. .!;-\ ...
CUSTOMER"' ORDER . ',' DATI! 19G9 .* 7
- REPORT OF TEST .OF . '" Ii 12"* CONCRETE CYLINDERS leC:TlONAL AREA 10.INC:H 28.0 27 .
20.027 28 0 27 28:';27 FOR ;." .. Both'l:c 1 Co;"porCiti on p.. o. 3210 'Fle-rida City,'ln$) COMPRESSION TEST LOAD. C:RUIHING ITRENGTH Lal. rr.R 10. INCH 11(1',000 5090 159 0 000 fJS20 11f1:500 50iO AYER. SOlO ... ., .;1' ...... ., . .. (j AOIit DAYI 7 7" 7 28 28 20 ' .. :".': ',: .... MPLE G concreto cyU ndo.J"s cost on 3-20e>69 by representativo at 1: C5 PM et J 0 and recaivcdin lobor£tory on PROJeCT I !ul'key Poi nt Units 3 " If LOC ... TION OF POUR' I Unit 1i4 contoinment struc1;\:re \'1&11 '. CL ... SS OF CONCRETE I 2P5 .LUMP OF CONCRETE I 2" CONTRACTOR ... 01011 XT URE .UPPLIER '11 eel 3 1 3 I 1 1 2 I Poir.t Client Attn; Mr. Cli".!n";; P;ttn.: CHant Attn: Itt; Flu .. _ Po .... L i !j:-:t I.li*kai' PG @> Box Woods Ston" t::) Har>'hmd Attn: :*11-.
- lr *. Wi 1 Ji li':\S ,: Perrir.3 Jr . 18-10 panel 1149 0 7Uo to t3 z*o £1. 09 0 to 99 0 . DwSo Wator : 9 galGo QU ... NTITY I 9 TRUCK NO. I 695 TICKET NO. I li07582 MIXINO TIME Ito mino . BLOO. PERMIT NO. I}}