View source for L-2010-212, Turkey Point, Units 3 & 4, Updated Final Safety Analysis Report - Unit 4 Cycle 24 Update, Chapter 1, Introduction and Summary

{{Adams
| number = ML102870302
| issue date = 09/21/2010
| title = Turkey Point, Units 3 & 4, Updated Final Safety Analysis Report - Unit 4 Cycle 24 Update, Chapter 1, Introduction and Summary
| author name = 
| author affiliation = Florida Power &amp; Light Co
| addressee name = 
| addressee affiliation = NRC/NRR
| docket = 05000250, 05000251
| license number = 
| contact person = 
| case reference number = L-2010-212
| document type = Updated Final Safety Analysis Report (UFSAR)
| page count = 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&#4 - INDIAN POINT #2 - GINNA 

Line Item Notes  
: 1. Nominal reactor power level intermediate between Indian Point #2 and Ginna plants. Power level related to safety only in the ability to produce and remove the power in the core as designed. 
: 2. Directly related to Item 1 by conversion 
: 3. No change in the fraction of the total heat generated in the core.  
: 4. This limitation applies only to prevention of temperatures of the fuel rods and coolant corresponding to power in excess of this 

overpower limit. As demonstrated by the detailed examination    

of the rod withdrawal accident at power, presented in Section 

14, nuclear overpower can be 18 percent without exceeding this 

limit.
The primary consideration in overpower protection is not the actual value of the trip set point but rather the error 

allowances that make up the margin to trip. The set point is selected so that a minimum DNB ratio of 1.30 is maintained at the 

condition of the maximum overpower when all errors are taken in 

the adverse direction and with the most adverse pressure and 

temperature allowed by the high T trip. The combination of these protection channels (variable high T and overpower) limit the range of allowable conditions to a region of temperature, pressure and power which preclude DNB or core damage for credible 

accidents.     

TABLE 1.4-1 Sheet 8 of 15 Line Item Notes    The error allowances are subject to verification by performance tests of the installed system. The errors due to drift and set 

point reproducibility are errors quoted by many instrumentation  

manufacturers and are demonstrated in actual performance tests on 

the equipment before shipment. The improved performance is 

attributable to the use of a solid state system.  

The errors due to rod motion result from variations in axial flux distribution with rod motion. Because of this variation, ion 

chamber reading at a given axial location may differ for the same 

core average power level. These errors are reduced by the use of 

long ion chambers with top and bottom detectors, each equal in 

length to about one-half the core height. The detectors yield an 

average reading over one-half the axial length. 

5,6  The reactor coolant system design pressure for the four plants is 2500 psia. For all conditions the system pressure is limited by 

code safety valves set to open at design pressure and sized to 

prevent system pressure from exceeding code limitations.   

Equipment capabilities for overpressure protection are established 

by the complete loss of load without an immediate reactor trip. 

The maximum over-pressure for this transient is therefore a 

function of the  safety valve capacity and the maximum pressurizer 

surge rate and is not dependent on the value of the nominal 

operating pressure.
TABLE 1.4-1 Sheet 9 of 15 Line Item Notes The operating pressure is selected to ensure that desired thermal conditions are maintained in the core. The operating pressure is 

established and maintained between the upper and lower reactor 

trip limits to permit transient variations in either direction 

with the assistance of the Pressure Control System. 

7,8  There are no significant differences among the hot channel factors. For a detailed discussion, see Section 3.2.2. 
: 9. The differences in the DNBR at nominal conditions is  due to the slight differences in operation conditions. 
: 10. Same for all plants. 
: 11. The flow varies from plant to plant due to pump design and number of loops.
TABLE 1.4-1 Sheet 10 of 15 Line Item Notes  
: 12. The effective flow rate for heat transfer is essentially proportional to the total flow rate as determined by the core 

geometry. 
: 13. Effective flow area for heat transfer is determined by the mechanical design of fuel assemblies and core. 

14-24  There are no significant changes for these parameters from the previous plants. 
: 25. The active surface area is determined by the mechanical design of the core. 

26-29  The heat transfer parameters are determined by the required heat output, the heat transfer surface area and the design peaking 

factors for the core. They are related to clad integrity in that 

these conditions must be within the capability of the fuel and 

must also meet the thermal- hydraulic design criteria of DNB and 

fuel temperature. Extensive experience indicates that no problem 

exists at these thermal outputs. 
: 30. Same for all plants. 

31-32  The fuel central temperatures are not significantly different than those for the other plants. The temperatures are well below the UO 2 melting temperature of 4800 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 !!"!".
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* 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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_ *fnMt __ _ ScNrce _______ _ """ 111-, '" Q': ,.,,"
PROJECT . LOCA TI<?N OF POUR CLASS OF CONCKETE SLUMP OF CONCRETE CONTRACTOR ,ADMIXTURE SUPPLIER REMARKS: 
--MA. PTL Bv Client', Repr:. -(Cross. Out ,:\ ' *. _l" ,-. -I' * * " ... , , i .f :-'----:-" ."j LAB. NO. _______ _ NAME ADDRESS OF CLIENT COMPRESSION TEST MA. NO __ _ Mark Sect. Arfa Sq. Inch , 28,27 28.27 28*27 28.27 28.27 28.27 Crushing Load Lbs. Crushing Strength Lbs. pn S9. Inch 18-5 r Avu. (}t: ',: 
.* /  
" .. !-;. ';" [, l ... .. *'PIITSB(jRGHl LABORATORY*'* 
' .1','" 1'" 3gen N.vt 29th 33142 "': .... , .. :-...., .. . 
.. ., *. *-<:-.'-'Ct.*'
ft*;:' .. . ':' ,',.'"." MUnI"" PIIOTeCTlOM
_ C&.IDITe. 
'nt_ PUaLlC "1010 AU. II_11ft ',i' . '''' "11_ .u .... na AI 'tit_ CONPIODITIAL NoraTV 0" CLI...,... "NO AU'ntOIlIZATION . . ,,:;: . .;, .. Oil PUaLlCATION 0 .. "",...acn.
CO_U.ION.
011 DTIIAc:TI PItON 011 1t_."IIOIN
* DUll It_11ft ** 1I .. IItV .. raNOIN. Olfll wltlnaN _OVAL. ORDER No. .,,2955 . LABORATORY No., CLIENT'. No. ClIEtlT, S\lOJECTt Fint)\'\Css 11& 3tt'Jftt 1/2" 3/&' Noo ,. No.) 8 HoduluSl jU 3.fJwI c 3lt t No o .. Moo 8 Hodulust '(
0 RT ON PHYSICAL T&#xa3;S'TS OF CO:<<:RETE AGGREGATE
__ a ,.,. fJECKTEL CORPORATION P 0 B 321S" -, .. ,. " o I) ox -< ': 
," Florfda Cit,.,
* J.-'.' .* ,. .* * , Florfda Power $ Ltght Turke, Pofnt;' unfts 3 and It .. .:-. 
...... 
.' . ':: -.. , , ; , .sGmptes of coarse aggregate sampled by Pfttsburgh Tos,h,ng Laboretoc-, rapreS81,taHve at . . !J S T ... _ R E.' .. '!. L T S PERCENT C
95 a4'lO ss '-25 .... 60 0.:, liO 00 5 95 c:o JOO o 0 10 00 S PITTSBlRtJt TESTING LABORAToRy eel 1 AUn, Mro Warry 1 Attn. ;oro Wf) 3 C 1 A Un I " ... o* Wucott 1 Attn, Hro Woods 3 ClIent, Attn: Hro Stade-at Md o 2 i"G 1B-6 1,10 Har1te. < Vfu President  
'. , <N?" \;,: '. .. .. , eUENT'S No. ::ROJECT. 
-..... :-' .. ': :" .. ' ,." : . .. .* ' i' ..." .. . '. i ," ,.'; ' ..
TORY , , , Sa j q .. ,. *1 NoW. 2ttIt AVENUE, MIAMI, FLORIDA Ule! . (. * ' ., .. 0" * '.: ............... , ... : ****** ' :": ,"; ** ," .. ,'. ', r r. RT -*.
OF CONCRETE 
,; .. BECHTEL CORPORATION , P. O. Box 3218"" F10rfda Citl, Florida -, " '. , *, ..... i***: ,. "'."' *. .... : .':: ** ": .. .. ':': Turkey Pofrtt , ct, lktfts 2 and 3'\: 1 Florfda Power and Lfght . 
.. fine aggregate sampled b'y the -PfttsbL"'4Jh Testing Laboratory representatfve at c IE.! .... T .. R,..ES;.;:U 
... LT ... S ... ." ,. ," ."." :\',' ASTM 1:33...66 SP&#xa3;CIFJCATlOt' FOR F IHE AGGREGATE 
!,,", fii1er.esa 
;'iroeness Hodulu5I SIEVE SIZE c:,.....,.-
-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&#xa3;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 CE'''ENT, ____ 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&#xa3;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 &#xa3;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}}