ML17139C013
| ML17139C013 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 04/20/2017 |
| From: | Southern Nuclear Operating Co |
| To: | Office of Nuclear Reactor Regulation |
| Williams S | |
| Shared Package | |
| ML17136A052 | List: |
| References | |
| NL-17-0534 | |
| Download: ML17139C013 (185) | |
Text
1-i REV 21 5/08
1.0 INTRODUCTION
AND GENERAL DESCRIPTION OF THE PLANT TABLE OF CONTENTS Page
1.1 INTRODUCTION
............................................................................................................1.
1-1 1.1.1 License Requested................................................................................1.1-1
1.1.2 Plant
Units..............................................................................................1.1-1
1.1.3 Plant
Location.........................................................................................1.1-1
1.1.4 Containment
Type..................................................................................1.1-1
1.1.5 Nuclear
Steam Supply System..............................................................1.1-2
1.1.5.1 Reactor Type and Supplier....................................................................1.1-2 1.1.5.2 Power Output.........................................................................................1.1-2
1.1.6 Schedule
for Completion and Commercial Operation..........................1.1-2
1.1.7 Organization
of Contents.......................................................................1.1-2
1.1.7.1 Subdivisions...........................................................................................1.1-2 1.1.7.2 Standard Format....................................................................................1.1-3 1.1.7.3 References.............................................................................................1.1-3 1.1.7.4 Tables and Figures................................................................................1.1-3 1.1.7.5 Numbering of Pages..............................................................................1.1-3 1.1.7.6 Revising the FSAR.................................................................................1.1-3
1.2 GENERAL
PLANT DESCRIPTION...............................................................................1.2-1 1.2.1 Site Characteristics................................................................................1.2-1
1.2.1.1 Location .................................................................................................1.2-1 1.2.1.2 Site Ownership.......................................................................................1.2-1 1.2.1.3 Access to the Site...................................................................................1.2-1 1.2.1.4 Site Environs..........................................................................................1.2-1 1.2.1.5 Geology .................................................................................................1.2-1 1.2.1.6 Seismology.............................................................................................1.2-2 1.2.1.7 Hydrology...............................................................................................1.2-2 1.2.1.8 Meteorology............................................................................................1.2-3 1.2.1.9 Radiological Environmental Monitoring Program..................................1.2-3
1.2.2 General
Arrangement............................................................................1.2-4
1.2.3 Nuclear
Steam Supply System..............................................................1.2-6
1.2.3.1 Reactor Core..........................................................................................1.2-6 1.2.3.2 Reactor Coolant System........................................................................1.2-6 1-ii REV 21 5/08 TABLE OF CONTENTS Page
1.2.4 Steam
and Power Conversion System..................................................1.2-7 1.2.5 Containment...........................................................................................1.2-8
1.2.6 Safety
Features......................................................................................1.2-8 1.2.7 Unit Control............................................................................................1.2-9
1.2.8 Plant
Electrical Power............................................................................1.2-9
1.2.9 Plant
Instrumentation and Control System .................................................................................................1.2-10 1.2.10 Auxiliary Systems...................................................................................1.2-10
1.2.10.1 Chemical and Volume Control System..................................................1.2-11 1.2.10.2 Residual Heat Removal System............................................................1.2-11 1.2.10.3 Component Cooling System..................................................................1.2-12 1.2.10.4 Fuel Handling and Storage System.......................................................1.2-12 1.2.10.5 Sampling Systems.................................................................................1.2-13 1.2.10.6 Cooling Water Systems.........................................................................1.2-13 1.2.10.7 Plant Ventilation System........................................................................1.2-13 1.2.10.8 Plant Fire Protection System.................................................................1.2-13 1.2.10.9 Compressed Air Systems......................................................................1.2-14
1.2.11 Waste Disposal System.........................................................................1.2-14
1.3 COMPARISON
TABLES................................................................................................1.3-1
1.3.1 Comparison
with Similar Facility Designs..............................................1.3-1
1.3.2 Comparison
of Final and Preliminary Designs......................................1.3-1
1.4 IDENTIFICATION
OF AGENTS AND CONTRACTORS..............................................1.4-1 1.4.1 Applicant-Owner and Operator..............................................................1.4-1
1.4.1.1 Description of Business.........................................................................1.4-1 1.4.1.2 Description of Corporate Organization..................................................1.4-1 1.4.1.3 Technical Qualifications.........................................................................1.4-2
1.4.2 Nuclear
Operations Organization..........................................................1.4-2
1.4.2.1 Southern Nuclear Operating Company, Inc..........................................1.4-2
1.4.3 Architect-Engineer..................................................................................1.4-2
1.4.3.1 Bechtel Power Corporation....................................................................1.4-2 1-iii REV 21 5/08 TABLE OF CONTENTS Page 1.4.4 Nuclear Steam Supply System (NSSS) Supplier....................................1.4-3
1.4.4.1 Westinghouse's Qualifications and Experience................................................................................................1.4-3 1.4.4.2 Plants in Operation...................................................................................1.4-4 1.4.4.3 Westinghouse Facilities............................................................................1.4-8
1.4.5 Plant
Construction....................................................................................1.4-10
1.4.6 Division
of Responsibility..........................................................................1.4-10
1.4.6.1 Alabama Power Company (APC)............................................................1.4-10 1.4.6.2 Southern Nuclear Operating Company, Inc.............................................1.4-10
1.4.6.3 Bechtel Power Corporation......................................................................1.4-11 1.4.6.4 Westinghouse Electric Corporation..........................................................1.4-11
1.5 REQUIREMENTS
FOR FURTHER TECHNICAL INFORMATION..............................1.5-1
1.5.1 Programs
Required for Plant Operation...................................................1.5-1
1.5.2 Other
Programs Required for Plant Operation........................................1.5-2 1.5.3 17 x 17 Fuel Assembly Verification Tests................................................1.5-5
1.5.3.1 Rod Cluster Control (RCC) Spider Tests.................................................1.5-6 1.5.3.2 Grid Tests .................................................................................................1.5-6 1.5.3.3 Fuel Assembly Structural Tests................................................................1.5-7 1.5.3.4 Guide Tube Tests.....................................................................................1.5-7 1.5.3.5 Prototype Assembly Tests........................................................................1.5-7 1.5.3.6 Departure From Nucleate Boiling (DNB)..................................................1.5-7 1.5.3.7 Incore Flow Mixing....................................................................................1.5-8
1.5.4 LOCA Heat Transfer Tests (17 x 17).......................................................1.5-8
1.5.4.1 Blowdown Heat Transfer Testing.............................................................1.5-9 1.5.4.2 Single Rod Burst Test (SRBT).................................................................1.5-10
1.6 MATERIAL
INCORPORATED BY REFERENCE.........................................................1.6-1
1.6.1 Westinghouse
Topical Reports................................................................1.6-1
1.6.1.1 Reports Referenced in FSAR...................................................................1.6-1
1.6.2 Bechtel
Power Corporation Topical Reports............................................1.6-9
1.6.3 General
Reports.......................................................................................1.6-10
1.6.4 Cross
Reference of Engineering Drawings..............................................1.6-14
1-iv REV 21 5/08 TABLE OF CONTENTS Page 1.7 GLOSSARY OF TERMS................................................................................................1.7-1
1.7.1 Abbreviations............................................................................................1.7-1
1.7.2 Drawing
Index and Symbols.....................................................................1.7-1
1.7.3 Historical
Descriptions..............................................................................1.7-1
1-v REV 21 5/08 LIST OF TABLES
1.1-1 FSAR Revision Dates
1.3-1 Design Comparison
1.3-2 Comparison of Plant Characteristics
1.3-3 Significant Design Changes
1.3-4 Significant Design Changes Since PSAR
1.4-1 Westinghouse Pressurized Water Reactor Nuclear Power Plants
1.6-1 Cross Reference of Engineering Drawings
1.7-1 Technical Abbreviations
1-vi REV 21 5/08 LIST OF FIGURES
1.2-1 Units 1 and 2 Plant General Arrangement Plan at el 155 ft and 189 ft
1.2-2 Units 1 and 2 Plant General Arrangement Plan at el 139 ft
1.2-3 Units 1 and 2 Plant General Arrangement Plan at el 121 ft and 129 ft
1.2-4 Units 1 and 2 Plant General Arrangement Plan at el 105 ft 6 in. and Below
1.2-5 Unit 1 Plant General Arrangement Section "A-A"
1.2-6 Unit 1 Plant General Arrangement Section "B-B"
1.2-7 Plant General Arrangement Section "C-C"
1.2-8 Plant General Arrangement Section "D-D"
1.2-9 Units 1 and 2 Plant General Arrangement Roof Plan at el 175 ft
1.7-1 Fluid System Symbols
1.7-2 Piping Nomenclature
FNP-FSAR-1 1.1-0 REV 21 5/08
FNP-FSAR-1 1.1-1 REV 25 4/14
1.0 INTRODUCTION
AND GENERAL DESCRIPTION OF PLANT
1.1 INTRODUCTION
This Final Safety Analysis Report (FSAR) complies with the Standard Format and Content of
Safety Analysis Reports for Nuclear Power Plants (Revision 1) issued by the Nuclear
Regulatory Commission in October 1972. For a discussion of the format of this report refer to
subsection 1.1.7, Organization of Contents.
1.1.1 LICENSE
REQUESTED The Final Safety Analysis Report (FSAR) was submitted in support of the application of the
Alabama Power Company (APC) for the original operating license for a nuclear power plant, designated as Joseph M. Farley Nuclear Plant (FNP) Units 1 and 2, located on a site near
Dothan, Alabama. The updated Final Safety Analysis Report is maintained by Southern
Nuclear Operating Company, which became a jo int licensee and exclusive operator of Plant Farley in 1991.
1.1.2 PLANT
UNITS The original application was for two reactor units, each at a rated power level of 2652 MWt
under Section 103 of the Atomic Energy Act of 1954, as amended, and the regulations of the
Nuclear Regulatory Commission set forth in Part 50 of Title 10 of the Code of Federal
Regulations (10 CFR 50).
A license amendment request was submitted to the Nuclear Regulatory Commission in
February, 1997, to increase the rated power level of each unit. The amendment request was
approved by the Nuclear Regulatory Commission on April 29, 1998. The current rated power
level for each unit is 2775 MWt.
The two units are essentially the same, and the descriptions of one unit are interpreted as
applying to both units. Differences between the two units, and particularly structures, systems, and components which are shared between the two units, are specifically pointed out.
1.1.3 PLANT
LOCATION This plant site is located in Houston County, about 16.5 miles east of Dothan, Alabama. The
nearest public road is State Highway 95, which forms the western boundary of the site property
as shown on figure 2.1-3.
1.1.4 CONTAINMENT
TYPE The containment for each of the FNP units (designed by Bechtel Power Corporation) consists of
a steel lined prestressed concrete structure.
FNP-FSAR-1 1.1-2 REV 25 4/14 1.1.5 NUCLEAR STEAM SUPPLY SYSTEM 1.1.5.1 Reactor Type and Supplier The nuclear steam supply system (NSSS) for each of the two FNP units is a pressurized water
reactor (PWR). The Westinghouse Electric Corporation has designed and supplied units for the
FNP.
1.1.5.2 Power Output Each NSSS was originally designed for a warranted power output of 2660 MWt, including 8
MWt of heat from nonreactor sources (primarily pump heat), with a corresponding gross
electrical output of approximately 861 MWe. Each NSSS was originally expected to be capable
of an output of approximately 2774 MWt, and all steam and power conversion equipment, including the turbine generator, was expected to have the capability of generating a maximum
calculated gross output of approximately 898 MWe. Although the original license application
was for 2652 MWt, all safety systems, including the containment and engineered safeguards, were designed and evaluated for operation at the higher power level. The power rating of 2774
MWt was originally used in the analysis of all postulated accidents bearing significantly on the
acceptability of the site. The thermal hydraulic and nuclear aspects of the core were originally
evaluated on the basis of a core thermal output of 2652 MWt.
The rated thermal power level (i.e., core thermal power) of each NSSS was subsequently
increased to 2775 MWt. The corresponding NSSS thermal power level is 2785 MWt, including
10 MWt of heat from nonreactor sources (primarily pump heat). Analyses and evaluations at
these increased thermal power levels and a conservative electrical power level of 943.4 MWe.
The uprated gross electrical output is approximately 933.4 MWe.
1.1.6 SCHEDULE
FOR COMPLETION AND COMMERCIAL OPERATION The two FNP units were completed and began commercial operation as tabulated below:
Unit Completion of Construction (Fuel Loading)
Commercial Power Operation1 7-4-77 12-1-77 2 3-8-81 7-30-81
1.1.7 ORGANIZATION
OF CONTENTS 1.1.7.1 Subdivisions This FSAR is organized into 17 chapters, each of which consists of a number of sections that
are numerically identified by two numerals separated by a decimal (e.g., 3.4 is the fourth section of chapter 3). Further subdivisions are also referred to as subsections.
FNP-FSAR-1 1.1-3 REV 25 4/14 1.1.7.2 Standard Format This FSAR has been written to comply with the Standard Format and Content Safety Analysis
Reports for Nuclear Power Plants (Revision 1) as issued by the Nuclear Regulatory
Commission in October 1972. This FSAR uses the same chapter, section, subsection, and
paragraph headings as those used in the standard format. Where appropriate, the FSAR is
subdivided beyond the extent of the standard format to isolate all information specifically
requested in that document. Where information has been presented that is not specifically
requested by the standard format and this informati on is identified numerically (chapter, section, subsection, or paragraph), this information is presented under the appropriate general heading
as a subdivision following all subdivisions contai ning information specifically requested by the standard format. (For example, subsection 1.1.7 is not requested in the standard format. Since
it apparently belonged in section 1.1, it was placed after the six subsections containing
information requested by the standard format.)
1.1.7.3 References References to another location in the FSAR are made by chapter section, or subsection
number.
1.1.7.4 Tables and Figures Tabulations of data are designated "tables." They are identified by the section number, followed
by a number according to its order of mention in the section (e.g., table 3.3-5 is the fifth table of
section 3.3). Tables are located at the end of the applicable section. Drawings, sketches, curves, graphs, and engineering diagrams are all identified as "figures" and are numbered
according to the order of mention in the section (e.g., figure 3.4-2 is the second figure of section
3.4). Figures are located at the end of the applicable section. Some plant project drawings are
included in the FSAR by reference to the drawing identification number (e.g., D-177024) in lieu
of inclusion in the FSAR as a figure.
1.1.7.5 Numbering of Pages Pages are numbered sequentially within each section. For example, 1.1-4 is the fourth page of
section 1.1. When it becomes necessary during revision of this FSAR to insert a page(s)
between two existing pages within a section, letters will be used. (For example, to insert two pages between 3.2-4 and 3.2-5, the following page order would appear: 3.2-4, 3.2-4a, 3.2-4b, 3.2-5.)
1.1.7.6 Revising the FSAR When it becomes necessary to submit additional information or revise information presently contained in the FSAR as required by 10 CFR 50.71(e), the following procedures will be
followed:
FNP-FSAR-1 1.1-4 REV 25 4/14 1. When a change is made to the FSAR text, those pages affected will be marked with the date of change or revision number or both in the lower righthand corner and a vertical line in the righthand margin next to the material affected. Where it is
necessary to insert additional pages between existing pages within a section, letters will be used. (To insert a page between pages 3.2-4 and 3.2-5, the insert
will be numbered 3.2-4a.)
2 Figures will be revised by indicating the date of change or revision number or both in the lower righthand corner.
Such revisions shall reflect all changes up to a maximum of 6 months prior to the date of filing.
FNP-FSAR-1 REV 22 8/09 TABLE 1.1-1 FSAR REVISION DATES
FSAR REVISION DATE 0 7/82 1 7/83 2 7/84 3 7/85 4 7/86 5 7/87 6 7/88 7 7/89 8 7/90 9 7/91 10 6/92 11 6/93 12 10/94 12A 6/95 13 4/96 14 12/97 15 6/99 16 11/00 17 5/02 18 10/03 19 5/05 20 11/06 21 5/08
FNP-FSAR-1
1.2-1 REV 25 4/14
1.2 GENERAL
PLANT DESCRIPTION
1.2.1 SITE CHARACTERISTICS 1.2.1.1 Location The site is located in southeast Alabama on the west side of the Chattahoochee River, about 6
miles north of the intersection of U. S. Highway 84 and State Highway 95, as shown on figures
2.1-1 and 2.1-2. It is in the northeastern section of Houston County, Alabama, just across the
river from Early County, Georgia. The site is about 100 miles southeast of Montgomery, Alabama, and about 180 miles south-southwest of Atlanta, Georgia. The Universal Transverse
Mercator Grid Coordinates for the center of the Unit 1 containment are Zone 16-R, central
meridian 87 degrees, east 678,872.5 meters, north 3,455,620.1 meters; for the center of the
Unit 2 containment they are Zone 16-R, central meridian 87 degrees, east 679,871.6 meters, north 3,455,705.8 meters.
1.2.1.2 Site Ownership Alabama Power Company (APC) owns the 1850-acre site. Boundaries are shown on figure 2.1-
- 3.
1.2.1.3 Access to the Site All activities on the site are under the control of Southern Nuclear Operating Company. Access
to the plant proper is controlled by a securi ty fence and Security Force Member (SFM).
1.2.1.4 Site Environs There are no people living on the site. Approximately 45 percent of the land area around the
site is wooded, with the remainder being used for various agricultural purposes. The nearest
industry is located about 4 miles to the south in Early County, Georgia. The nearest developed
community is Columbia, Alabama, 5 miles to the north. The nearest major city is Dothan, Alabama, about 16.5 miles to the west, having an estimated 1980 population of 48,700.
1.2.1.5 Geology The site is located in the extreme southeastern portion of the East Gulf Coastal Plain
physiographic province, which covers about 65 percent of the State of Alabama. This province
is underlain by Mesozoic and Cenozoic sedimentary rocks which dip southward at 10 to 40 ft
per mile. These deposits consist of marine and nonmarine gravels, sands, silts, clays, marls, and their consolidated equivalents such as sandstone and limestone. In southeastern Alabama
and southwestern Georgia, the gentle south-dipping Paleocene through Oligocene sequence is
influenced by only minor structural features. These structures have been inactive since FNP-FSAR-1
1.2-2 REV 25 4/14 Miocene time, and do not affect materials underlying the site. The structure nearest the site is
the east-west trending Gordon anticline, 10 miles south of the site.
No major or active faults were found nor are believed to exist within the 50-mile radius studied.
No evidence of surface displacement was observed during the field investigation.
The site is characterized by two topographic features:
- 1. The gently undulating upland, which ranges from about el 130 to 240 ft mean sea level (msl).
- 2. The Chattahoochee River valley, which includes the floodplain and the river channel itself, all lying between el 130 and 70 ft msl.
The Lisbon formation (Eocene), which is a soft to moderately hard sedimentary rock and dense
sand formation approximately 130 ft thick, is the principal load-bearing material for plant
structures. The upper formational contact of the Lisbon is slightly undulating and ranges from el
90 to 103 ft msl.
1.2.1.6 Seismology The southern Chattahoochee River Valley is located within a broad region of infrequent seismic
activity encompassing southern Alabama, southern Georgia, and adjacent Florida. This region
is one of the least seismically active regions in the United States and is characterized by a few low-magnitude and low-intensity shocks. Historic records show that earthquakes have never
been felt at the site with an intensity greater than Modified Mercalli V and that no earthquake of
epicentral intensity greater than Modified Mercalli IV has occurred within 150 miles of the plant
site. In addition, the seismicity of the area was evaluated on the basis of the interim revised
Environmental Science Survey Administration (ESSA) Seismic Risk Map of the United States (Algermissen, 1969). This map shows that the site is located on the Zone 0 - Zone 1 border.
The closest distance from the site to a Zone 2 boundary is 85 miles and to a Zone 3 boundary is
255 miles.
The maximum intensity postulated as having been experienced at the site, as a result of any
historical earthquake, is low to moderate V. This intensity corresponds to a surface
acceleration of 0.03g on Hershberger's (1956) curve. Therefore, it is considered conservative
to select 0.10g surface acceleration as the safe shutdown earthquake (SSE). The SSE is that
earthquake which produces the vibratory ground motion for which Category I structures, systems, and components are designed to remain functional. A surface acceleration of 0.05g is
referred to as the 1/2 SSE. Category I structures, systems, and components will also be
designed to withstand the effects of vibratory motion at the 1/2 SSE in combination with other
appropriate loads. Vertical acceleration is taken as two-thirds of the above horizontal
accelerations. For detailed discussions, refer to sections 2.5 and 3.7.
FNP-FSAR-1
1.2-3 REV 25 4/14 1.2.1.7 Hydrology
The dominant hydrological features of the site region are the Chattahoochee River and some
small tributary streams. The Chattahoochee River joins the Flint River about 44 river miles
downstream of the plant site and forms the Apalachicola River, which empties into the Gulf of
Mexico 144 river miles downstream.
The area of the drainage basin affecting the Chattahoochee River at the site is about 8246
square miles. The average flow in the river during 33 years of record at Columbia, Alabama (about 6 river miles upstream) has been 10,600 ft 3/s, and the minimum flow was 1210 ft 3/s. The maximum historical flow, based on 60 years of record, was 207,000 ft 3/s during the flood of 1929. This corresponds to an estimated maximum flood stage at the site of about 124 ft msl.
The probable maximum flood for the Chattahoochee River at the Farley site has been
calculated to be 642,000 ft 3/s, which corresponds to an estimated maximum flood stage of 144.2 ft msl. This compares to the plant site grade of 154.5 ft msl.
1.2.1.8 Meteorology The climate at the site is typical of that in the Southern Gulf Coastal Plain, being hot and humid
in the summer and mild in the winter. Maximum rainfall in the 30-year period of record at
Blakely, Georgia, 15 miles northeast of the site, was about 11 in. in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and the maximum
average monthly rainfall was about 7 in. Napier Field, near Dothan, Alabama, about 22 miles
west-northwest of the site, is the closest offsite point for which wind direction and velocity data
are available. Wind directions are well distributed, with a slight predominance from the south-
southwest.
During the period from 1900 to 1966, there were 44 hurricanes or post-hurricane paths which
passed within 100 miles of the site. Since the site is approximately 80 miles from the southern
coast of Florida, the hurricane wind speeds are, in general, lower than those farther to the
south.
During the 50-year period from 1918 to 1967, for which 44 years of record are available, there
have been 39 tornadoes within a 25-mile radius of the site. During the period from 1958 to
1967, there were 13 tornadoes reported within this radius. The high frequency in the latter
period probably is due to improved reporting. The effects of severe weather were taken into
account in the design of plant buildings.
[HISTORICAL] In March 1971, an onsite meteorological program began operation. Data for the program are gathered from instruments placed at selected positions on the 240-ft microwave tower located north of the plant site. The program included measurements of wind speed and direction at 50 and 150 ft above the tower base and temperature measu rements at 35 and 200 ft above the tower base.
The instruments are installed on a 60-m (197-ft) tower located near the microwave tower at various heights. FSAR subsection 2.3.3 and table 2.3.10 reflect the current environmental monitoring equipment and locations.
FNP-FSAR-1
1.2-4 REV 25 4/14 1.2.1.9 Radiological Environmental Monitoring Program
The purpose of the Radiological Environmental Monitoring Program (REMP) is to measure
radioactive material in the environment which may be released from the plant. The type of
environmental measurement to be made is select ed so as to evaluate possible significant modes of human exposure. Background radiation reference measurements to assess radiation sources in the environment not attributable to the plant are made concurrently with
measurements of radiation in the environment due to plant operation. To accomplish this, two types of sampling stations are established--one type (indicator stations) is placed where
maximum doses attributable to the plant are expected, the other (control stations) is placed
where radioactive levels are not expected to be significantly influenced by radioactive releases from the plant. Appropriate samples are collected and analyzed to ensure that the significant
pathways of radioactive material to man are evaluated.
1.2.2 GENERAL
ARRANGEMENT The site plot plan, which indicates the arrangement and orientation of the plant buildings, storage pond, river water intake, and plant service water intake structures is shown on drawing
D-170084.
Each containment houses a nuclear steam supply system (NSSS) consisting of a reactor, steam generators, reactor coolant pumps, pressurizer, and some of the reactor auxiliaries.
The auxiliary buildings house the waste treatm ent facilities, engineered safeguards system components, heating and ventilation system com ponents, switchgear, laboratories, offices, laundry, and the control room. The spent-fuel pool and the new fuel storage facilities are
located in the auxiliary building of the respective units. These facilities are under controlled
ventilation whenever spent fuel is being moved or stored in that section. The transfer of fuel to
the spent-fuel pool of each unit from its associated containment is through a fuel transfer tube.
The emergency diesel generators are housed in the diesel generator building, which is located
south of the Unit 1 auxiliary building. The turbine building houses the turbine generator, condensers, feedwater heater, condensate and feedwater pumps, turbine auxiliaries, and
certain nonsafety-related switchgear for both units.
The solidification and dewatering facility located east of the auxiliary buildings is provided to
solidify or dewater spent radioactive resins and to solidify chemical drains and evaporator
concentrates. The facility will also be used for decontamination activities.
The service building on the south end of the turbine building provides office, shop, and
warehouse space for the plant. The computer/office building west of the service building
provides office, shop, and computer facilities. The outage support building, located northeast of
the Unit 2 containment building, provides office and storage space, and a craft break area.
The following structures, systems, and co mponents are shared by the two units.
FNP-FSAR-1
1.2-5 REV 25 4/14 Structures
Control room
Access control area
Diesel generator building
Auxiliary building HVAC penthouse Switchyards
Hot machine shop and decontamination room
Hot instrument shop
Ultimate heat sink storage pond, river intake structure, and service water intake structure
Water treatment plant
Drum storage area
Training center
Respirator storage
Solidification and dewatering facility
Low-level radwaste storage buildings
Secondary chemistry laboratory
Outage support building
Old steam generator storage facility Systems Control room HVAC system
Diesel generator fuel storage/transfer system
Diesel generator HVAC systems for shared diesel generators and switchgear rooms
Carbon dioxide supply system
Nitrogen supply system
Hydrogen supply system
Oxygen supply system
Demineralized water system
River water system to storage pond
Potable and sanitary water system Well water system
Auxiliary steam supply system
Reactor makeup water system
Service water intake structure compressed air system and backup nitrogen system
Service water intake structure HVAC
Cathodic protection system
Communications system
Components
One pressurizer heater station service transformer
Three of five diesel generators and auxiliary support equipment FNP-FSAR-1
1.2-6 REV 25 4/14 Diesel generator 600-volt switchgear and MCCs
Service water intake structure 600-volt switchgear, batteries and battery chargers
Fire protection system water supply
Fire pumps
Spent-fuel cask crane
Load centers H, J, N, K, L, R, and S
Service water discharge pipes to UHS pond and common discharge pipe to river
Spent-resin storage tanks
Unit 1 and Unit 2 plant layout drawings are shown in figures 1.2-1 through 1.2-9. The Unit 2
plant layout is generally a mirror image of the Un it 1 layout except for the shared spaces and equipment listed above. Figure 12.1-1 shows a layout of the combined control room for Units 1
and 2.
The criterion that is followed in the design of Units 1 and 2 is that each unit operates
independently of the other.
1.2.3 NUCLEAR
STEAM SUPPLY SYSTEM The nuclear steam supply system (NSSS) for each uni t consists of a pressurized water reactor and a three-loop reactor coolant system. The mechanical, thermal hydraulic, and nuclear
design of the reactor core is similar to the design of other Westinghouse units under
construction. The maximum expected thermal output of each unit is 2774 MWt. A heat
balance, showing the major parameters of the plant for the rated power condition, is shown in
figure 10.1-1 and for the maximum calculated power condition is shown in figure 10.1-2.
1.2.3.1 Reactor Core The reactor core is a three-region core. The fuel rods are pressurized, cold worked zircaloy
tubes containing slightly enriched uranium dioxide fuel. For the initial core, fuel assemblies with
the highest enrichments are placed in the core periphery, or outer region, and the two groups of
lower enrichment fuel assemblies are arranged in a selected pattern in the central region. In
reload cores, approximately 72 assemblies are discharged and the fresh fuel is loaded in the
center of the core with once burned fuel predominately on the periphery.
The fuel assembly is a canless type with the basic assembly consisting of the control rod guide
thimbles attached to the grids and the top and bottom nozzles. The fuel rods are held by the
spring clip grids in this assembly which provide a very stiff support for the fuel rods.
Full-length control rod assemblies and burnable absorber rods are inserted into the guide
thimbles of the fuel assemblies. The absorber sections of the control rods are fabricated of
silver-indium-cadmium sealed in stainless steel tubes. The absorber material in the fixed
burnable absorber rods is either in the form of borosilicate glass sealed in stainless steel tubes
or in the form of B 4 C in an alumina matrix sealed in zircaloy tubes.
Above the core, each cluster of control rods is attached to a spider connector and drive shaft, which is raised and lowered by a drive mechanism mounted on the reactor vessel head.
FNP-FSAR-1
1.2-7 REV 25 4/14 The control rod drive mechanisms for the full-length control rod assemblies are of the magnetic-
latch type. The latches are controlled by three magnetic coils. They are so designed that, upon
loss of power to the coils, the rod cluster control assembly is released and falls by gravity to
shut down the reactor.
1.2.3.2 Reactor Coolant System The reactor coolant system consists of three similar heat transfer loops connected in parallel to
the reactor vessel. Each loop contains a circulating pump and steam generator. The system
also includes a pressurizer, pressurizer relief tank, connecting piping, and instrumentation
necessary for operational control and protection.
The reactor coolant system transfers the heat generated in the core to the steam generators
where steam is generated to drive the turbine generator. Borated demineralized light water is
circulated at a flow rate and temperature consistent with achieving the desired reactor core
thermal hydraulic performance. The water also acts as a neutron moderator, reflector, and
solvent for the neutron absorber.
The reactor coolant pumps are Westinghouse vertical, single-stage, centrifugal pumps of the
shaft-seal type. The power supply systems for the pumps are designed so that adequate
coolant flow is maintained to cool the reactor core under all required conditions.
The steam generators are Westinghouse vertical U-tube units which contain inconel tubes.
Integral moisture separators reduce the moisture content of the steam to 0.10% or less.
The reactor coolant piping and all of the pressure containing and heat transfer surfaces in
contact with the reactor coolant are stainless steel, stainless steel clad. The steam generator
tubes and fuel tubes are inconel and zircaloy, respectively. The reactor core internals, including
the control rod drive shafts, are stainless steel.
An electrically heated pressurizer connected to one reactor coolant loop maintains reactor
coolant system pressure during normal operation, limits pressure variations during plant load
transients, and keeps system pressure within design limits during abnormal conditions.
1.2.4 STEAM
AND POWER CONVERSION SYSTEM The turbine generator is furnished by the Westinghouse Electric Corporation (modified by
Siemens) and is an 1800-rpm tandem-compound, four-flow exhaust, indoor unit designed for
saturated steam conditions. The unit utilizes two parallel strings of six feedwater heaters each.
The main steam generator feedwater pumps are driven by steam turbines supplied normally
with reheated steam and exhausting to the condenser.
The auxiliary feedwater system is designed to provide emergency heat removal, and two motor-driven pumps and one turbine-driven pump are provided.
FNP-FSAR-1
1.2-8 REV 25 4/14
1.2.5 CONTAINMENT
The containment uses a prestressed concrete design and is a vertical right cylindrical structure
with a dome and flat base. The containment interior is lined with carbon steel plate for
leaktightness.
Inside the containment, the reactor and other NSSS components are shielded with concrete. A
vent stack is attached to the outside of the containment and extends to an elevation 10 ft above
the top of the containment dome. Access to portions of the containment during power operation
is permissible.
The containment, in conjunction with engineered safety features, is designed to withstand the
internal pressure and coincident temperature resulting from the energy release of the loss-of-
coolant accident (LOCA) associated with 2774 MWt. The containment design conditions are for
an internal pressure of 54 psig and coincident temperature of 280°F.
1.2.6 SAFETY
FEATURES The engineered safety features system, especially the containment and cooling water systems, provides protection for the public and plant personnel against the accidental release of
radioactivity from the reactor system, particula rly as the result of a LOCA. These safety features function to localize, control, mitigate, and terminate such accidents so as to maintain
the exposure to the public within the requirements of 10 CFR 100.
The engineered safety features include, but are not limited to, the following systems:
- The containment spray system.
- The containment cooling system.
- The penetration room filtration system.
Subsection 6.1.1 provides a detailed list of engineering safety features systems.
The ECCS injects borated water into the reactor coolant system following a LOCA. This
provides cooling to limit core damage and fission product release and assures an adequate
shutdown margin regardless of temperature. The ECCS also provides continuous long-term
post-accident cooling of the core by recirculating borated water between the containment sump
and the reactor core.
The containment for each unit is equipped, as follows, with two independent, full-capacity
systems for cooling the containment atmosphere after the postulated LOCA:
A. The containment spray system supplies borated water to cool the containment atmosphere. The spray system, in combination with at least one of the FNP-FSAR-1
1.2-9 REV 25 4/14 containment air coolers, is sized to provide adequate cooling with one of the two
containment spray pumps in service on emergency power. These pumps take
suction from the refueling water storage tank. When this supply is depleted, suction of these pumps is aligned to pump water from the containment sump
directly into the containment during the recirculation mode of operation. Trisodium
phosphate is added to the sump to enhance iodine removal from the containment
atmosphere in the postaccident condition.
B. The containment cooling system is designed to provide containment atmosphere mixing and cooling. The system design basis is to provide adequate containment
cooling from the operation of one train of containment spray and one containment
cooler.
The penetration room filtration system for each unit collects and processes potential ECCS recirculation leakage to limit environmental activity levels following a LOCA. This system is also
used to mitigate the consequences of a fuel handling accident in the spent-fuel pool.
1.2.7 UNIT CONTROL The reactor is controlled by control rod movement and regulation of the boric acid concentration
in the reactor coolant. During steady-state operation, the reactor control system maintains a
programmed average reactor coolant temperature that rises in proportion to the load.
The solid-state protection logic system automatic ally initiates appropriate action whenever the parameters monitored by this system reach pr eestablished setpoints. This protection system acts to trip the reactor, actuate emergency core cooling, close containment isolation valves, and initiate the operation of other safety features systems.
1.2.8 PLANT
ELECTRICAL POWER The electrical systems are designed to provide re liable power sources for electrical equipment for startup, normal operation, safe shutdown, and emergency situations in both Units 1 and 2.
The Unit 1 output is fed through a step-up transformer to the 230-kV switchyard, and Unit 2
output is fed through a step-up transformer to the 500-kV switchyard. The 230-kV switchyard of
Unit 1 is connected to the high voltage transmission system through three 230-kV transmission
lines which approach the site from different directions. The 500-kV switchyard is connected to
the high-voltage transmission system by two 500-kV lines, and the 230-kV and 500-kV
substations are connected through two auto transformers. A switchable 230-kV shunt reactor is
provided to help control switchyard voltages and improve stability during transmission system light load conditions. In addition to carrying the electrical output of the plant, the high voltage
transmission system provides a means for power to be supplied to the plant from external sources. Startup power for normal loads and aux iliary power for emergency loads is taken from the 230-kV switchyard through startup auxiliary transformers. Since the emergency buses are
normally connected to and fed from offsite source s through the startup auxiliary transformer, in
case of emergency, power transfer to offsite sources will not be necessary. Normal plant
auxiliary power is taken from the generator main leads through the unit auxiliary transformer(s)
FNP-FSAR-1
1.2-10 REV 25 4/14 or may be taken directly from the startup aux iliary transformers. Offsite power for the engineered safeguards for each unit is provided from a startup auxiliary transformer. Onsite emergency power is provided by four diesel generators. Each unit has one diesel generator
dedicated to it specifically and two of the diesel generators are shared between the two units.
An additional diesel generator is dedicated as the alternate power source during station
blackout (SBO) events. Auxiliary power for each unit is utilized at 4160, 600, 480, and 208/120
volts. Direct current systems are also provided for emergency power, engineered safety features control, essential nuclear instrumentation, and switchyard control and relaying.
The normal and emergency sources of electrical power are each adequate to permit prompt
shutdown and maintain safe conditions under all credible circumstances. The capacity of the
power sources is sufficient for the required safety function under postulated abnormal
conditions.
1.2.9 PLANT
INSTRUMENTATION AND CONTROL SYSTEM Instrumentation and controls essential to avoid undue risk to the health and safety of the public
are provided to monitor and maintain nuclear power, primary coolant pressure, temperature, and control rod positions within prescribed operating ranges.
The nonnuclear regulating process and containment instrumentation measure temperatures, pressures, flows, and levels in the steam syst ems, containment, and auxilia ry systems. Process variables required on a continuous basis for the startup, operation, and shutdown of the unit are
indicated, recorded, and controlled from the control room. The quality and types of process
instrumentation provided ensure safe and orderly operation of all systems and processes over the full operating range of the plant.
The reactor control system provides for startup and shutdown of the reactor and for adjustment
of the reactor power in response to turbine load demand. The reactor is controlled by control
rod cluster motion, which is required for load follow transients and for startup and shutdown;
and by a soluble neutron absorber (boron in the form of boric acid) which is inserted during cold
shutdown, partially removed at startup, and adjusted in concentration during core lifetime to
compensate for such effects as fuel consumption and accumulation of fission products which
tend to slow the nuclear chain reaction. The control system permits the unit to accept step load
increases or reductions of 10 percent and ramp load increases of 5 percent per minute over the
load range from 5 percent to, but not exceeding, 100-percent power under normal operating
conditions, subject to xenon limitations.
Control of both the reactor and turbine generator is accomplished from the control room by NRC
licensed personnel.
1.2.10 AUXILIARY SYSTEMS Nuclear auxiliary systems are provided to perform the following functions:
A. Supply reactor coolant system requirements.
FNP-FSAR-1
1.2-11 REV 25 4/14 B. Purify reactor coolant.
C. Introduce chemicals for corrosion inhibition.
D. Introduce and remove chemicals for reactivity control.
E. Cool system components.
F. Remove residual heat during a portion of the reactor cooldown period and during shutdown of the reactor.
G. Cool the spent-fuel pool water.
H. Permit sampling of reactor coolant water.
I. Provide for safety.
J. Vent and drain the reactor cool ant system and the auxiliary systems.
K. Provide containment ventilation and cooling.
L. Dispose of liquid and gaseous wastes and provide for disposal of solid wastes.
These functions are performed by the following systems.
1.2.10.1 Chemical and Volume Control System The purity level in the reactor coolant system is controlled by continuous purification of a bypass
stream of reactor coolant. Water removed from the reactor coolant system is cooled in the regenerative heat exchanger. From there, the coolant flows to a letdown heat exchanger and then through a demineralizer where corrosion and fission products are removed. It then passes
through a filter and is sprayed into the volume control tank.
The chemical and volume control system automatica lly adjusts the amount of reactor coolant to compensate for changes in specific volume, due to coolant temperature changes and reactor
coolant pump shaft seal leakage, in order to maintain a constant level in the pressurizer.
1.2.10.2 Residual Heat Removal System The residual heat removal system is used to reduce the temperature of the reactor coolant at a
controlled rate from 350°F to a refueling temperature of approximately 140°F and to maintain
the proper reactor coolant temperature during refueling.
The residual heat removal pumps are used to circulate the reactor coolant through two residual
heat removal heat exchangers, returning it to the reactor coolant system through the low
pressure injection header.
FNP-FSAR-1
1.2-12 REV 25 4/14 1.2.10.3 Component Cooling System
The component cooling system provides an intermediate barrier between the reactor coolant
system and the service water system.
The component cooling system consists of th ree pumps and three heat exchangers to remove heat from the various auxiliary systems handling the reactor coolant. Corrosion inhibited
demineralized water is circulated by the system through the letdown heat exchanger, the reactor coolant pump coolers, sample heat exchangers, spent-fuel pool heat exchangers, recycle and
waste evaporators, waste gas compressors, etc. Component cooling water provides cooling for
the residual heat exchangers.
1.2.10.4 Fuel Handling and Storage System The reactor is refueled by the use of 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.
Transfer of spent fuel under water permits the economic use of an optically transparent
radiation shield as well as providing a reliable source of coolant for removal of residual heat.
The fuel handling system provides for the safe handling of rod cluster control assemblies and
for the required assembly and disassembly of reactor internals. The system is divided into two
pool regions: the refueling cavity, which is flooded for refueling, and the spent-fuel pool, which
is external to the reactor containment and is always accessible to plant personnel. The two pools are connected by the fuel transfer system, which transports the fuel from the refueling
cavity to the transfer canal by the use of:
A. Manipulator crane, which removes spent fuel from the reactor, located inside the containment above the refueling pool.
B. Fuel transfer carriage.
C. Upending devices.
D. Fuel transfer tube.
E. Fuel handling crane in the spent-fuel pool area.
F. Various devices used for handling the reactor vessel head and internals.
New fuel is stored dry in vertical racks in a storage area in the auxiliary building. Space is
provided for over one-third of a core and fuel assembly, and spacing is such as to preclude
criticality.
Each unit has a stainless steel lined, reinforced concrete spent fuel pool containing borated
water which provides storage for approximately 9 cores. Spent-fuel assemblies are stored in
vertical racks so spaced as to preclude criticality with administrative controls on placement and
no credit taken for the borated pool water. Additional credit is taken for the presence of soluble
boron in the pool water to maintain K eff < 0.95.
FNP-FSAR-1
1.2-13 REV 25 4/14 A shared independent spent-fuel storage installation (ISFSI) is provided for temporary storage
of spent fuel pending offsite transportation. The ISFSI is located inside the protected area south
of the diesel generator building and consists of five storage pads, each designed to store 12 spent-fuel casks. Following a suitable decay period, the spent fuel will be transported offsite for
disposal. The ISFSI is operated in accordance with the general license provisions of 10 CFR
Part 72 as described in the Joseph M. Farley 10 CFR 72.212 Report.
Purification and redundant cooling equipment is provided for the spent-fuel pool water. This
equipment may also be used for cleanup of refueling water after each fuel change in the
reactor.
1.2.10.5 Sampling Systems Two sampling systems are provided--one for t he reactor coolant and its auxiliary systems and one for the turbine steam and feedwater system. These systems are used for determining both chemical and radiochemical conditions of the various fluids used in the plant.
1.2.10.6 Cooling Water Systems The turbine generator condenser is cooled by the circulating water system, which is a closed
system, rejecting heat to cooling towers.
The service water requirements for the nuclear components, turbine building equipment, and
diesel generators are supplied by vertical centrifugal pumps taking suction from the storage
pond. Water makeup to the storage pond is supplied by vertical centrifugal pumps taking
suction from the river.
1.2.10.7 Plant Ventilation Systems Separate ventilation systems are provided for t he containment, auxiliary building fuel handling facility, control room and TSC, turbine building, and emergency diesel generator building. In
addition, a purge system and containment preacce ss filtration system are provided for the containment atmosphere.
The auxiliary building penetration rooms are ventilat ed by the penetration room filtration system, which includes filters for control of any leakage through the containment penetrations during
accident conditions.
1.2.10.8 Plant Fire Protection System The major fire protection system contains both diesel and electrical powered fire pumps which
supply the various hydrants, hose stations, sprinklers, and deluge systems. Supplementary to
these facilities, chemical fire extinguishi ng equipment is provided to accommodate special requirements for various classes of hazards.
FNP-FSAR-1
1.2-14 REV 25 4/14 Consideration is given to the use of noncombustible and fire resistant materials throughout the
facility, particularly in areas containing critical portions of the plant such as the containment, control room, and components of the engineered safeguards system.
Hydrants and hose stations are manually operated, while the sprinkler and deluge systems can be both automatic and manually actuated systems. A sufficient number of portable
extinguishers are placed at key locations for use in extinguishing fires. A complete description
of the fire protection system is presented in appendix 9B.
1.2.10.9 Compressed Air Systems Nonlubricated air compressors, in combination with aftercoolers, discharge compressed air to
air receivers which supply compressed air to a common header. This header furnishes
compressed air for the service air system and inst rument air system. Instrument air is dried
and filtered downstream of this common supply header.
The plant air system provides compressed air for normal maintenance service at various
stations throughout the plant, while the instrument air system provides compressed air for the
operation of all air operated instruments and valves.
1.2.11 WASTE DISPOSAL SYSTEM The waste disposal system provides controlled handling and disposal of liquid, gaseous, and
solid wastes. The waste processing system pr ovides all equipment necessary for controlled treatment and preparation for retention or disposal of all liquid, gaseous, and solid wastes
produced as a result of reactor operation.
The liquid waste processing system collects, processes, and recycles reactor grade water, removes or concentrates radioactive constituents, and processes them until suitable for release
or shipment off site. Liquid wastes are sampled and activity levels verified and recorded prior to
release. Processed liquid effluent from the r eactor coolant system will have been subjected to
the chemical and volume control system purif ication ion exchanger and the components of the waste processing system and will be within the lim its established by technical specifications.
The gaseous waste processing system functions to remove fission product gases from the
reactor coolant and to contain these gases during normal plant operation. The system also
collects the gases generated from the boron recycle evaporator. Waste gases are collected in
the vent header. These gases are withdrawn from the vent header by one of two compressors and discharged to a waste gas decay tank. The tank contents will be released to the
environment in accordance with technical specifications. Three waste gas decay tanks are
provided; each tank has a 45-day storage capacity. Gaseous wastes are discharged through
an absolute particle filter to the vent stack.
Solid wastes, when required, are stored in suitable containers for offsite disposal.
FNP-FSAR-1 1.3-1 REV 21 5/08
1.3 COMPARISON
TABLES 1.3.1 [HISTORICAL] [COMPARISONS WITH SIMILAR FACILITY DESIGNS
Table 1.3-1 presents a design comparison of the nuclear steam supply system design for the
Joseph M. Farley plant with the North Anna Power Station and the Surry Power Station.
Table 1.3-2 presents a design comparison of the containment system parameters and
engineering safety features design for the Farley plant with the Calvert Cliffs Power Station, the
Oconee Power Station, the Palisades Power Station, and the Turkey Point Power Station.
] 1.3.2 COMPARISON OF FINAL AND PRELIMINARY DESIGNS
All of the significant changes that have been made in the facility design since submittal of the
Preliminary Safety Analysis Report (PSAR) are listed in table 1.3-3. Each item in table 1.3-3 is
cross referenced to the appropriate section in the FSAR which describes the changes and the
bases for them.
Significant changes in NSSS design since the PSAR are listed, with reasons for change, in
table 1.3-4.
FNP-FSAR-1 REV 21 5/08
[HISTORICAL] [TABLE 1.3-1 (SHEET 1 OF 3)
DESIGN COMPARISON Nuclear Steam Supply System - Comparison with North Anna Power Station (Units 1 and 2) and Surry Power Station (Units 1 and 2)
Chapter Chapter Title References Number System/Component (FSAR) Significant Similarities Significant Differences
4.0 Reactor
Fuel Subsection 4.2.1 Similar to North Anna and Surry Differences exist in design parameters based on nuclear design and thermal-hydraulic design parameters (Surry has 15-x-15 fuel assembly)
Reactor vessel Subsection 4.2.2 Similar to No rth Anna and Surry Surry has a diffuser plate.
internals except as noted North Anna and Surry have a thermal shield (Farley has neutron pads) Reactivity Subsection 4.2.3 Similar to Nort h Anna and Surry Surry has 12 glass rods in control systems except as noted a burnable poison assembly Nuclear design Section 4.3 Similar to North Anna and Surry Differences exist in fuel except as noted burnup rates, fuel enrich-ments, k eff , and core kinetic characteristics Thermal-hydraulic Section 4.4 Similar to North Anna and Surry Surry has a core thermal design except as noted output of 2441 MW. Differ-ences exist in thermal and hydraulic and heat transfer parameters
5.0 Reactor
Coolant System Secti on 5.1, Similar to North Anna and Surry Differences exist because
5.2 except
as noted North Anna and Surry employ loop stop valves Reactor vessel* Section 5.4 Similar to North Anna and Surry Farley's steam generator is similar to the initial configura-tion of the Surry steam generator.
Surry steam generators have since been modified Reactor coolant pumps* Subsection 5.5.
1 Similar to North Anna and Surry None Steam generators* Subsection 5.5.2 Similar to North Anna and Surry None Piping Subsection 5.5.3 Similar to North Anna and Surry None FNP-FSAR-1 REV 21 5/08 TABLE 1.3-1 (SHEET 2 OF 3)
Chapter Chapter Title References Number System/Component (FSAR) Significant Similarities Significant Differences Residual heat Subsection 5.5.7 Functionally similar to North The RHRS has no safety removal system Anna and Surry function for North Anna (RHRS) or Surry Pressurizer* Subsection 5.5.10 Similar to North Anna and Surry None Loop stop valves None - - 6.0 Engineered Safety Features Emergency core Section 6.3 Similar to North Anna and Surry Farley uses the RHR heat cooling system except as noted exchangers for long term cooling. North Anna and Surry each have a separate low head recirculation system
7.0 Instrumentation
and Controls Reactor trip system Section 7.2 System functions are similar None to North Anna and Surry Engineered safety Section 7.3 System functions are similar None features systems to North Anna and Surry Systems required Section 7.4 System functions are similar None for safety shutdown to North Anna and Surry Safety-related Section 7.5 Parametric display is similar Actual physical con-display instrumen-to that of North Anna and figuration may differ due tation Surry to customer design philosophy Other safety Section 7.6 Operational functions are Some valve interlock systems similar to North Anna and designs differ slightly Surry on Surry Control systems Section 7.7 System functions are similar None to North Anna and Surry
9.0 Auxiliary
Systems Chemical and volume Subsection 9.3.4 Similar to North Anna and Farley uses a 4 wt/% boric control system Surry except as noted acid solution for make up.
North Anna and Surry each have a separate boron recovery system. Farley has a boron recovery subsystem
FNP-FSAR-1 REV 21 5/08 TABLE 1.3-1 (SHEET 3 OF 3)
Chapter Chapter Title References Number System/Component (FSAR) Significant Similarities Significant Differences 11.0 Radioactive Waste Management Source terms Section 11.1 Similar to North Anna and Differences are based on Surry except as noted plant operational influences Liquid waste Section 11.2 Functionally similar to None processing North Anna and Surry Gaseous waste Section 11.3 Functionally similar to North Anna and Surry have processing North Anna and Surry provisions for periodic except as noted release of the gaseous wastes Process radiation Section 11.4 Functionally similar to Differences which exist monitoring North Anna and Surry are due to differences in customer requirements 14.0 Initial Tests and Chapter 14 Si milar to North Anna and None Inspection Surry 15.0 Accident Analysis C hapter 15 Similar to North Anna and Additional analysis Surry presented for North Anna and Farley
_______________________ *Designed and manufactured to code in effect
]
FNP-FSAR-1 REV 22 8/09
[HISTORICAL][TABLE 1.3-2 (SHEET 1 OF 6)
COMPARISON OF PLANT CHARACTERISTICS Farley Calvert Cliffs Oconee Palisades Containment Units 1 and 2 Units 1 and 2 Units 1, 2 and 3 Unit 1 System Parameters FSAR FSAR FSAR FSAR Type Steel-lined, prestressed Steel-lined, prestressed Steel-lined, prestressed Steel-lined prestressed posttensioned concrete posttensioned concrete posttensioned concrete posttensioned concrete cylinder, curved dome cylinder, curved dome cylinder, curved dome cylinder, curved dome roof roof roof roof Design parameters Inside diameter (ft) 130 130 116 116 Inside height (ft) 183 182 208 190 Free volume (ft) 2,024,900 2,000,000 1,900,000 1,600,000 Design pressure (psig) 54 50 59 55 Concrete thickness (ft) Vertical wall 3-3/4 3-3/4 3-3/4 3 Dome 3-1/4 3-1/4 3-1/4 2-1/2 Containment leak Leaktight penetration Leaktight penetration Leaktight penetration Leaktight penetration preventi on and and continuous steel and continuous steel and continuous steel and continuous steel mitigation systems liner. Automatic isol a- liner. Automatic isola- liner. Automatic isola- liner. Automatic isola-tion where required. tion where required. tion where required. tion where required.
The exhaust from pene- The exhaust from pene- The exhaust from pene- The exhaust from pene-tration rooms to vent. tration rooms to vent. tration rooms to vent. tration rooms to vent.
Gaseous effluent purge Discharge through stack Discharge through stack Discharge through stack Discharge through stack
ENGINEERED SAFETY FEATURES
Safety injection system No. of high head (pumps) 3 3 3 3 No. of low head (pumps) 2 2 3 2 Containment fan coolers No. of units 4 4 3 4 Air flow capacity, 40,000 40,000 58,000 60,000]
each at emergency conditions (ft /min)
FNP-FSAR-1 REV 22 8/09 TABLE 1.3-2 (SHEET 2 OF 6)
Farley Calvert Cliffs Oconee Palisades Containment Units 1 and 2 Units 1 and 2 Units 1, 2 and 3 Unit 1 System Parameters FSAR FSAR FSAR FSAR Postaccident filters No. of units None 3 None None (ft /min) None 20,000 None None Containment spray No. of pumps 2 2 2 3 Emergency power Diesel-generator units 5 total for both units 3 total for both units Hydro 1 Safety injection tanks, number 3 4 2 4
FNP-FSAR-1 REV 22 8/09 TABLE 1.3-2 (SHEET 3 OF 6)
Turkey Point Units 3 and 4 References FSAR Significant Similarities Significant Differences By Sections Steel-lined, prestressed Containment types are the same for all posttensioned concrete units cylinder, curved dome roof 116 Design parameters are the same for Containment design parameters 6.2 169 Farley and Calvert Cliffs for Farley and Calvert Cliffs, and the rest of the listed units, differ because of differences in 1,550,000 dome height 59 3-1/2 3 Leaktight penetr ation and Containment Leak Prevention and Containment leak prevention and
6.2.4 continuous
steel liner. Mitigation Systems are of the same mitigation systems for Palisades Automatic isolation where design bas es for Farley, Calvert Cliffs, and Turkey Point differ from those of required. The exhaust from and Oconee Farley, Calvert Cliffs, and Oconee.
penetration rooms to vent.
The former type has no exhaust from penetration rooms to vent 6.2.2 Through particulate filter All units discharge through the stack and monitors part of except Turkey Point main exhaust systems.
2 The Safety Injection System is the same Oconee has 3 low head injection 6.2.2 2 basic design for all the listed plants pumps except Oconee 3 The number of containment fan coolers Calvert Cliffs and Palisades have 6.2.2.2 is the same for Farley, Calvert Cliffs, emergency-condition flowrates of and Palisades. They have 4 each. 60,000 ft /min. Oconee, Farley, and Oconee and Turkey Point ha ve 3 each Turkey Point hav e emergency flowrates of 58,000, 40,000, and 25,000 25,000 ft /min, respectively 3 Calvert Cliffs and Turkey Point 37,500 both have postaccident filters.
The other units do not 2 Farley, Calvert Cliffs, Oconee, and Palisades has 3 containment spray 6.2.3.1.1 Turkey Point have 2 containment spray pumps. The others have 2 each pumps each FNP-FSAR-1 REV 22 8/09 TABLE 1.3-2 (SHEET 4 OF 6)
Turkey Point Units 3 and 4 References FSAR Significant Similarities Significant Differences by Sections 2 total for both units Oconee derives its emergency 8.3.1.1 3 power from hydropower. Palisades utilizes 1 diesel generator; Turkey Point has 2 diesel generators for both plants; and Farley has a total of 5 diesels for both its plants.
Oconee utilizes 2 safety injection tanks, while Farley has 3, Calvert 6.2 Cliffs has 4, Palisades has 4, and Turkey Point has 3
FNP-FSAR-1 REV 22 8/09 TABLE 1.3-2 (SHEET 5 OF 6)
Farley Calvert Cliffs Oconee Containment System Parameters Units 1 and 2 Units 1 and 2 Units 1, 2 and 3 Electrical Components FSAR FSAR FSAR Standby power system Total of 5 diesels of Three diesels connected Two hydro units.
which 3 are shared between to 4-kV buses and shared 230-kV network and Units 1 and 2. between Units 1 and 2 startup transformers Diesels are connected to 4160-V buses Engineered safety feature buses Six 4160-V buses/unit Two 4-kV buses/unit Three 4160-V redundant divided into two separate divided into separate and buses per unit and redundant systems. redundant systems dc system Separate and redundant Four batteries between 2 Separate and redundant 125-V dc systems for ESF units divided to give two 125-V dc systems for in-loads. Separate dc systems separate and redundant 125-V strumentation and control for loads in auxiliary build- dc systems. Separate dc power system. Separate dc ing, turbine building, cool- systems are provided for systems for large loads, ing tower area, diesel generator turbine building and the switching station control, building, and switchyard switchyard and control of Keowee hydro station Vital instrumentation systems System comprised of 4 inverters Four inverters provided Four inverters arranged to arranged to give 4 separate between two units to give give 4 separate and redun-and redundant channels 4 separate and redundant dant buses channels per unit Offsite power system Unit 1 - 230-kV switchyard. 500-kV switchyard. Two Units 1 and 2 connected to Unit 2 - 500-kV switchyard. startup transformers shared the 230-kV switchyard and Each unit is comprised of between two units Unit 3 to 500-kV switch-two startup transformers yard. Each unit is pro-and two unit auxiliary transformers vided with one unit auxil-with the ESF buses supplied from iary and one startup startup transformers transformer
FNP-FSAR-1 REV 22 8/09 TABLE 1.3-2 (SHEET 6 OF 6)
Palisades Turkey Point Unit 1 Units 3 and 4 Reference FSAR FSAR Significant Similarities Significant Differences by Sections Total of 2 diesels. Two diesels connected Farley is similar to Oconee has hydro units. 8.3.1 Diesels are connected to 4160-V buses and Calvert Cliffs, Farley has one diesel to 2400-V buses shared between the two Palisades, and Turkey permanently aligned to an units Point, except as noted ES F bus per unit, as against that for Calvert Cliffs and Turkey Point, which have shared diesels only Two 2400-V separate Two 4160-V buses/unit Farley is similar to all None 8.3.1 and redundant buses divided into separate plants it is being com-and redundant systems pared with Separate and redundant Separate, redundant Farley is similar to A separate dc system is 8.3.2 125-V dc systems 125-V dc system supply- Calvert Cliffs and provided for non-ESF loads supplying ESF loads ing ESF loads and non- Oconee for Farley, in contrast to and non-ESF loads ESF loads. Separate that for Palisades and dc system for switch-Turkey Point yard Four inverters arranged Four inverters arranged Farley is similar to None 8.3.1 to give 4 separate to give 4 separate all plants being and redundant channels redundant channels considered Switchyard - 345 kV. Switchyard - 240 kV. Farley is similar to Farley has two unit auxil- 8.2] Two unit auxiliary and Each unit is provided Palisades iary and two startup trans-two startup trans- with one unit auxiliary formers compared to one of formers provided. and one startup trans-each for Oconee and Turkey ESF buses normally former Point. The startup trans-supplied from unit formers for Farley are not auxiliary trans-shared between the two units, formers as is the case for Calvert Cliffs
FNP-FSAR-1 REV 21 5/08 TABLE 1.3-3 (SHEET 1 OF 3)
SIGNIFICANT DESIGN CHANGES Change Discussed Item in FSAR Subsection Reason for Change
- 1. Steam jet air
9.4.4 Charcoal
filters were added in the steam jet air ejector ejector filtration discharge. Change initiated to reduce normal plant releases and/or plant personnel exposures
- 2. Radwaste ventilation
9.4.3 Charcoal
filters were added in the radwaste ventilation area. area Change initiated to reduce normal plant releases and/or plant personnel exposures
- 3. Steam generator 10.4.8 The blowdown heat exchanger was installed in place of the blowdown system blowdown flash tank. Change initiated to reduce normal plant releases and/or personnel exposures
- 4. Containment volume 3.8.1.1 The containment volume was increased. Containment volume increased to hold peak containment post-LOCA pressures within acceptable limits when considering the effect of post-reflood energy addition to the containment
- 5. Cable routing 8.3.1.4.3 Deleted reference to cable tray as Category I equipment. Cable tray supports are designed to meet Category I seismic requirements. No safety significance
- 6. Emergency lighting 9.5.3.3 Emergency lighting supplied either from plant batteries or individual battery packs and not from 600-V emergency buses. Change results in increased reliability
- 7. Startup auxiliary 8.2.1-3 Load growth required an increase in capacity. Size of transformers transformers could not be increased because interrupting capacity of switchgear would not be adequate for larger Two startup auxiliary trans-transformers. Four transformers decreased the number of formers provided in each unit ties between units instead of three for both
- 8. Service water to diesel 8.5.1c (11) Adequate separation of water pipes from Train A and Train B engine heat exchanger was impossible to maintain and still supply each engine from both headers; alternate safeguards such as barriers or pipes Each diesel engine HX was within pipes were not considered feasible. Each engine will supplied by only one be supplied water from Unit 1 and Unit 2 service water header until Unit 2 went into service, instead of two service water headers
FNP-FSAR-1 REV 21 5/08 TABLE 1.3-3 (SHEET 2 OF 3)
Change Discussed Item in FSAR Subsection Reason for Change
- 9. Main steam isolation valv es 10.3.9 The development of refined steam line blowdown mass and energy flows following postulated steam line breaks dictated Each main steam line the deletion of the swing check valves. It was judged im- contains two redundant swing practical to design a swing check valve for the new loadings. disc trip valves instead of The planned swing check valves were changed to swing disc one swing disc trip valve and trip valves to provide redundancy in stopping forward flow. one check valve.
It was determined that stopping backflow from the intact steam generators and piping would be assured by the redundant forward flow trip valves, obviating the previous need for swing check valves
- 10. Fuel handling area 9.4.2.2.2 Charcoal filters were added in the fuel handling area ventilation system ventilation system. Change initiated to reduce normal plant releases and/or personnel exposure
- 11. Containment purge 6.2.3.2.3 Charcoal filters were added in the containment purge system. system Change initiated to r educe normal plant releases and/or personnel exposure
- 12. Containment pre-access 6.2.3.2.3 This system was added in the containment ventilation system. filtration system Change initiated to reduce normal plant releases and/or personnel exposure
- 13. Containment isolation 6.2.4(1) (1) Deletion of valves in the containment pressure instrument barrier lines; refer to table 6.2-19, sheet 6 of 6, Note 5. Also refer to table 1.3-4, item 20. No safety significance (2) (2) Deletion of the vents from the airlock doors to the penetra- tion room. The airlock doors contain a double seal arrange- ment. After a review of our PSAR design, it was determined that total plant releases following an assumed single failure of the innermost seal would be minimized by removing the vent line, since the contribution to total site boundary dose due to leakage past the second seal would be less than the con- tribution associated with free venting of gross leakage past the inner seal to the penetration room, even when the effects of the penetration room filtration system are considered
- 14. Features for mitigating Appendix 3K Added a venting penthouse to avoid overpressurizing the main high-energy line rupture steam and feedwater valve room. Relocated hot shutdown panel outside containment due to environmental considerations. Motor-driven auxiliary feedwater pump rooms are now watertight and the turbine-driven auxiliary feedwater pump room is no longer watertight
FNP-FSAR-1 REV 21 5/08 TABLE 1.3-3 (SHEET 3 OF 3)
Change Discussed Item in FSAR Subsection Reason for Change
- 15. Moved the emergency pond 2.4.8 To remove spillway from the dike so as to eliminate potential spillway structure from effects on the downstream dike slope due to discharge over the northeast abutment of the spillway. Either design would be safe, but there is less the main dam to the north risk of damage with the current design. (Reference AEC letter leg of the pond of 3/1/73 for acceptance of the relocated spillway.)
FNP-FSAR-1 REV 21 5/08 TABLE 1.3-4 (SHEET 1 OF 3)
SIGNIFICANT DESIGN CHANGES SINCE PSAR Item Change in Design FSAR Reference Reason for Change
- 1. Core internals change - redesign 4.2.2.2 In order to simplify the core support structure, a design study was of thermal shield pads to neutron conducted to investigate the possibility of replacing the thermal pads shield with locally positioned plate members called neutron shielding pads. The plot of the circum ferential fluence di stribution of the vessel, which is very localized, showed this to be a practical approach. The design study es tablished the feasibility of replacing the thermal shield with a series of plate members of equivalent thickness attached to the core barrel by pins and bolts. Both analytical and experimental methods were utilized to determine design adequacy. The above design study is described in WCAP-7870, which has been submitted to the NRC
- 2. Core internals change - redesign 4.2.2.2 A standard flat plate forging was developed to replace the original of lower core support plate casting design
- 3. Core internals change - redesign 4.2.2.2 Allows a flat upper support plate to be used without affecting the of upper internals support system hydraulic flow characteristics, and result s in short support columns and guide tubes providing increased margins from possible flow-induced vibrations
- 4. Reactor vessel top and bottom head 5.4.1.4, To meet requirements of Section XI of the ASME Boiler and Pressure penetration and CRDM were 5.4.4.4 Vessel Code redesigned to meet ISI requirements
- 5. Addition of removable insulation 5.4.2 To prov ide access to these areas for inspection purposes on the closure and lower reactor vessel heads
- 6. Rod withdrawal stop from rod drop 7.2 The positive neutron flux rate trip ensures that the criteria signal and automatic turbine load appropriate for an ANS Condition IV event are met even for rod cutback initiated by rod drop have ejections from partial power. been replaced by the power range neutron flux positive rate trip
- 7. Analog RPI change to digital RPI 7.7.1.3 Improved performance with reduced setup and calibration time. An advancement in the "state of the art" 8. Steam generator blowdown de-11.2 Added system flexibility to the blowdown system mineralizers added to blowdown system 9. Changed all clean radioactive 11.2 To reduce system leakage sources waste treatment system valves to diaphragm valves except where system function or fluid condi-tions dictate otherwise
FNP-FSAR-1 REV 21 5/08 TABLE 1.3-4 (SHEET 2 OF 3)
Item Change in Design FSAR Reference Reason for Change
- 10. Fuel pellet density 4.3 The pellet densities cite d in the PSAR show a variation by fuel region of 91-, 92-, and 94-percent TD; whereas this FSAR cites 94-percent TD for all regions. The PSAR reflects the former design philosophy which provided lower density pellets for higher burnup regions to accommodate fuel swelling; however, experience has shown that swelling is not a strong function of burnup as previously believed, so that a uniform core pellet density can be employed. As discussed in section 1.5 and paragraph 4.2.1.3.1, the pellet density specification is subject to con- siderations of irradiation-induced densification
- 11. Revised method and accuracy of de-3.8.2 The Hach procedure is simpler, faster, and offers the same degree of termining water-soluble nitrates accuracy in tendon sheathing filler material from ASTM to Hach Chemical Co. procedure
- 12. Two startup auxiliary transformers 8.1.2 Load growth requires an increase in capacity. Size of transformers could are provided for each unit not be increased because interrupting capacity of switchgear would not be instead of three for two units adequate for larger transformers. Four transformers decreased the number of ties between units
- 13. Each diesel engine heat exchanger 8.3.1.1.7 (11) Adequate separation of water pipes from Train A and Train B was was supplied by only one impossible to do and still supply each engine from both headers, and service water header until Unit 2 alternate safeguards such as barriers of pipes within pipes were not went into service, instead considered feasible. Each engine will be supplied water from Unit 1 of two service water headers and Unit 2
- 14. Changed the stainless steel in 3.8.1.1.3 Type 304L was originally specified for better welding characteristics. the spent fuel pool and refueling Subsequent experience has shown Type 304 to be welded as easily as cavity from Type 304L to Type 304 Type 304 L
- 15. The material of the main steam Figure 3.2-8 To assure satisfactory results in impact tests as per ASME Section III. lines up the check valves was changed to A-155 KCF70 Class I
- 16. Deleted the 12-in. minimum 8.3.1.4.3 The design meets NRC requirements on 5-ft vertical and 3-ft horizontal vertical spacing between cable separation between cable trays of redundant safeguard trains. In a trays containing different classes few instances, it was not possible to meet the 12-in. separation of circuits in the same safeguard between cable trays containing different classes of circuits of the train as specified in same safeguard train. These instances have been examined by the paragraphs 8.7.2.5.c and designer on a case-by-case basis to ensure that the spacings are 8.7.2.5.d adequate 17. Fuel assembly change-4.1, 4.4.1 The average linear power is reduced with the 17-x-17 fuel assembly 15-x-15 fuel assemblies to 17-x-17 fuel assemblies
- 18. RCS pipe design code change from 5.5.3 The ASME III code w ent into effect subsequent to the PSAR. The code ANSI B31.7 to ASME III effectivity date covered Farley FNP-FSAR-1 REV 21 5/08 TABLE 1.3-4 (SHEET 3 OF 3)
Item Change in Design FSAR Reference Reason for Change
- 19. Steam flow restrictors installed 5.5.4, 10.3.6 Use of the integral steam flow restrictor for steam flow measurements in the steam generator outlet precludes the need for an additional flow measurement device. See nozzle section 3.2 for design criteria
- 20. Deletion of valves in containment 6.2.4 The present sealed sensing lines without valve arrangement provides pressure instrument lines automatic double barrier isolation without operator action and without sacrificing any reliability with regard to its safeguards function
- 21. Method of LOCA analysis 15.3, 15.4 The results of LOCA analysis are in conformance with 10 CFR 50.46 and appendix K
- 22. Specification of fuel design 4.4, 15.2 Acceptable fuel design limits for the 17-x-17 fuel assembly for antici- limits pated operational occurrences are presented in the references sections
FNP-FSAR-1 1.4-1 REV 25 4/14 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS
1.4.1 APPLICANT
- OWNER AND OPERATOR Alabama Power Company (APC) is the owner of the Farley Nuclear Plant. Alabama Power Company and Southern Nuclear Operating Company (SNC) are the licensees. SNC has
exclusive responsibility for and control over the physical construction, operation, and
maintenance of the facility.
1.4.1.1 Description of Business Alabama Power Company is a public utility incorporated under the laws of the State of Alabama
and is engaged in the generation, distribution, and sale of electricity at retail in 639 cities and towns and at wholesale to 15 municipal electric systems and 11 rural electric cooperatives.
Alabama Power Company's electric generating facilities consist of 8 steam electric generating
plants, 14 hydroelectric generating plants, and 1 combustion turbine generating plant, with an
installed capacity of 10,285,475 kW as of December 31, 1991.
Alabama Power Company, acting as its own general contractor, has constructed 27 steam
electric units and at present has none under construction. Within the last 30 years, these
include the following:
In Operation Unit Mwe Greene County 1 250 Greene County 2 250 Barry 4 350 Barry 5 700 Gorgas 10 700 Gaston 5 880 Miller 1 660 Miller 2 660 Miller 3 660 Miller 4 660 Joseph M. Farley Nuclear Plant No. 1 829 Joseph M. Farley Nuclear Plant No. 2 829 1.4.1.2 Description of Corporate Organization Alabama Power Company (APC) is a public utility incorporated under the laws of the State of
Alabama with its principal offices located at 600 North 18th Street, Birmingham, Alabama.
FNP-FSAR-1 1.4-2 REV 25 4/14 Southern Nuclear Operating Company (SNC) is responsible for operational support functions for the six nuclear units within the Southern electric system. APC and SNC are wholly owned
subsidiaries of The Southern Company.
1.4.1.3 Technical Qualifications The licensees and their affiliated companies have participated in the development of nuclear
power for about 30 years as members of Atomic Power Development Associates, Inc., and
Power Reactor Development Company, the designers and operators of the Enrico Fermi Atomic
Power Plant. The participation has consisted of both financial contributions and assignment of
personnel.
A course, Introduction of Nuclear Power, conducted by the University of Alabama, was begun in
January 1970. The course requirements, consisting of 144 classroom hours, were completed
by a group of 24 engineers in the Engineering, Power Supply, and Construction Departments.
A second group of 11 personnel completed the course on an audit basis.
The technical qualifications of APC and SNC are further delineated in section 13.1, Organization Structure of APC.
1.4.2 NUCLEAR
OPERATIONS ORGANIZATION 1.4.2.1 Southern Nuclear Operating Company, Inc.
Southern Nuclear Operating Company, Inc. (S NC) is a wholly owned subsidiary of The Southern Company.
SNC was formed from support organizations of Southern Company Services, Inc. (SCS),
Georgia Power Company, and APC, whose indi viduals collectively have many years of experience in the area of nuclear operations. SNC is responsible for the operational and
maintenance support of FNP. SNC is a licensee, along with APC, of FNP, as well as the sole
operator of the plant.
1.4.3 ARCHITECT
- ENGINEER SNC is responsible for the engineering and design of the Farley Nuclear Plant. Bechtel Power
Corporation has been retained by SNC as a subcontractor for the major portion of the plant
design.
1.4.3.1 Bechtel Power Corporation Bechtel Power Corporation is retained by SNC to assist in the engineering and design of the
Farley Nuclear Plant. Bechtel has extensive experience in the design and construction of over FNP-FSAR-1 1.4-3 REV 25 4/14 182 thermal generating units representing more than 82,900,000 kW of new generating capacity, of which 54 units are nuclear with more than 43,700,000 kW.
Among the numerous nuclear projects for which Bechtel is presently acting or has acted as
engineer-constructor are the following pressurized water reactor (PWR) units:
- Robert E. Ginna Nuclear Station (450,000 kW) for Rochester Gas & Electric Corporation (construction only).
- Turkey Point Units 3 and 4 (760,000 kW each) for Florida Power & Light Company.
- Palisades Unit 1 (770,000 kW) for Consumers Power Company.
- Point Beach Units 1 and 2 (450,000 kW each) for Wisconsin-Michigan Power Company.
- Oconee Units 1, 2, and 3 (840,000 kW each) for Duke Power Company (engineering only).
- Calvert Cliffs Units 1 and 2 (800,000 kW each) for Baltimore Gas and Electric Company.
- Millstone Unit 2 (830,000 kW) for Northeast Utilities Company (The Millstone Point Company).
- Davis-Besse Nuclear Power Station (872,000 kW) for the Toledo Edison Company and the Cleveland Electric Illuminating Company.
- Trojan Unit 1 (1,100,000 kW) for Portland General Electric Company.
1.4.4 NUCLEAR
STEAM SUPPLY SYSTEM (NSSS) SUPPLIER 1.4.4.1 Westinghouse's Qualifications and Experience Westinghouse Electric Corporation (Westinghouse) is responsible for supplying the nuclear
steam supply system (NSSS) and fuel for Farley Units 1 & 2.
Westinghouse has designed, developed, and manufactured nuclear power facilities since the
1950s, beginning with the world's first large central station nuclear power plant (Shippingport),
which produced power from 1957 to 1982. Completed or presently contracted commercial
nuclear capacity totals in excess of 97,000 MW. Westinghouse pioneered new nuclear design
concepts, such as chemical shim control of reactivity and the rod cluster control concept, throughout the last two decades. Westinghouse manufacturing facilities include the largest
commercial nuclear fuel fabrication facility in the world, and the world's most modern heat
transfer equipment production facility, as well as other facilities producing NSSS components.
FNP-FSAR-1 1.4-4 REV 25 4/14 Table 1.4-1 lists all Westinghouse pressurized water reactor (PWR) plants to date, including those plants currently under construction.
The US Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute have
contracted with Westinghouse for research into NSSS related activities. Westinghouse
experience was also utilized by the NRC and Metropolitan Edison immediately following the
Three Mile Island Unit 2 accident and remains as a heavy participant with the Westinghouse
Owner's Group of utilities in addressing the NRC's action plan for corrective actions.
1.4.4.2 Plants in Operation Westinghouse PWR plants in operation are as follows:
A. Shippingport Shippingport was the world's first large central station nuclear power plant. The
reactor plant was designed by the Bettis Atomic Power Laboratory, which is
operated by Westinghouse under a NRC contract. Shippingport's PWR has
produced power for Duquesne Light Company since December 1957.
B. Yankee-Rowe Singled out by the NRC as a "Nuclear Success Story," Yankee-Rowe went online
in November 1960. Owned and operated by the Yankee Atomic Electric Company, Yankee-Rowe has progressed from an initial rating of 120 MWe to its present 175
MWe rating. Westinghouse supplied the NSSS and the turbine generator.
C. Trino Vercellese (Enrico Fermi)
The Trino Vercellese nuclear plant was one of the first Westinghouse designed
plants to incorporate chemical shim control of reactivity. Chemical shim has since
become a standard feature of Westinghouse PWR control. Trino Vercellese
achieved initial criticality in June 1964 and began power operation in October
1964. The plant is rated at 260 MWe.
D. Chooz (Ardennes)
The Chooz plant is unique in that the Westinghouse PWR and its auxiliaries are
housed in man-made caverns. Ardennes, a joint Franco-Belgian undertaking, owned and operated by the Societa d 'Energie Nucleaire Franco-Belge des
Ardennes (SENA), is located in France near the French-Belgian border. Chooz
achieved initial criticality in October 1966 and began power operation in 1967.
E. San Onofre No. 1 San Onofre No. 1 employs the Westinghouse developed rod cluster control
concept which has since become a standard feature on the Westinghouse PWR.
FNP-FSAR-1 1.4-5 REV 25 4/14 Owned by the Southern California Edison Company and the San Diego Gas and Electric Company, the 450 MWe plant is located near San Clemente, California.
Westinghouse supplied the NSSS and the turbine generator. Initial criticality was
achieved in June 1967, and power operating began in January 1968.
F. Haddam Neck (Connecticut Yankee)
Owned and operated by the Connecticut Yankee Atomic Power Company, this
plant went critical in July 1967 and attained full power operation in December
1967. Like San Onofre No. 1, the plant employs rod cluster control in conjunction
with chemical shim control. Westinghouse supplied the NSSS and the turbine
generator. The plant has been uprated to 575 MWe.
G. Jose Cabrera - (Zorita)
The Jose Cabrera station is located near Zorita, Spain. The 153 MWe plant
employs rod cluster control, chemical shim control and a zircaloy-clad core.
Construction began in mid-1965, and power operation began in 1968. Jose
Cabrera is owned and operated by the Union Electra, S.A., a Spanish utility.
H. Beznau No. 1 and No. 2
Beznau No. 1, Switzerland's first commercial nuclear power plant, achieved initial
criticality in June 1969 and supplied power to the system in July 1969. The 350
MWe plant was designed and constructed by the Westinghouse-Brown Boveri
Consortium for the owner/operator utility, Nordostschweizerische Kraftwerke A.G.
The plant started producing power less than 4 years after award of the plant
contract. Beznau No. 2 achieved criticality in October 1971 and began commercial
operation in early 1972.
I. Robert Emmett Ginna
The Robert Emmett Ginna Plant, owned and operated by Rochester Gas and
Electric Corporation, is located in New York on the south shore of Lake Ontario.
Westinghouse supplied the 490 MWe plant on a turnkey basis. Construction
began in April 1966 with initial criticality being achieved in November 1969 (just 42
months after start of construction). Power was supplied to the system in December
1969.
J. Mihama No. 1 and Takahama No. 1
These plants are owned by the Kansai Electric Power Company, Inc. Mihama No.
1 is a two-loop, 320 MWe unit and marks the beginning of a line of Westinghouse
PWRs supplying the generation needs of the Far East. Westinghouse
International Company was the prime contractor for the Mihama project, supplying
the NSSS engineering, nuclear fuel, and some major system components.
Mihama No. 1 required only 44 months from the start of site construction to first
power production in August 1970. Takahama No. 1 is a three-loop, 780 MWe unit.
Initial criticality was achieved in March 1974.
FNP-FSAR-1 1.4-6 REV 25 4/14 K. H. B. Robinson No. 2 This plant is a three-loop, 707 MWe unit which was built on a turnkey basis for the
Carolina Power and Light Company. The plant is located at a site near Hartsville, South Carolina, on a man-made cooling lake. The construction permit was granted
in April 1967. The plant achieved criticality in August 1970 and first power to
system in October 1970.
L. Point Beach No. 1 and No. 2
The Point Beach Project consists of two 497 MWe units, which were built on a
turnkey basis for the Wisconsin Michigan Power Company and the Wisconsin
Electric Power Company. The plants are located near Two Creeks, Wisconsin, 90
miles north of Milwaukee on Lake Michigan. This was the first two-unit station to
utilize many common facilities and shared auxiliary systems. The construction permit for Point Beach No. 1 was granted in July 1967 with initial criticality and first
power to the system in November 1970. Point Beach No. 2 went critical in May
1972 and was available for commercial operation in October 1972.
M. Surry No. 1 and No. 2
The Surry Power Station, two three-loop 822 MWe units, is owned by the Virginia
Electric and Power Company. The James River Station is about 30 miles from
Norfork, Virginia. First criticality on Surry No. 1 was achieved in July 1972.
Commercial operation began in September 1972. Initial criticality on Surry No. 2
was achieved in March 1973.
N. Turkey Point No. 3 and No. 4
Florida Power and Light Company is the owner of a four-unit station in Biscayne
Bay, Florida. Turkey Point No. 3 and 4 of the station are three-loop, 745 MWe
plants. Commercial status for Turkey Point No.3 was achieved in December 1972.
Initial criticality for Turkey Point No. 4 was achieved in June 1973.
O. Indian Point No. 2 and No. 3
Consolidated Edison Company of New York operates three nuclear units located in
Buchanan, New York. Units 1 and 2 are owned by the Company and Unit 3 is
owned by the Power Authority of the State of New York. Units 2 and 3 are
Westinghouse PWRs rated at 873 and 965 MWe, respectively. Indian Point No. 2
achieved initial criticality in May 1973 and Indian Point No. 3 achieved initial
criticality in April 1976. P. Prairie Island No. 1 and No. 2
Northern States Power Company is the owner of these two-loop, 530 MWe units
located in Welch, Minnesota. Initial criticality was achieved in December 1973 for
Prairie Island No. 1, and in December 1974 for Prairie Island No. 2.
FNP-FSAR-1 1.4-7 REV 25 4/14 Q. Zion No. 1 and No. 2 Commonwealth Edison Company is the owner of these two four-loop, 1050 MWe
units. The units are located on Lake Michigan near Zion, Illinois. Initial criticality
was achieved in June 1973 for Zion No. 1 and in December 1973 for Zion No. 2.
R. Kewaunee
Wisconsin Public Service Corporation, Wisconsin Power and Light Company, and
Madison Gas and Electric Company are the owners of this two-loop, 541 MWe
plant located in Kewaunee, Wisconsin. Initial criticality was achieved in March
1974.
S. Ringhals No. 2
Statens Vattenfallsverk (SSPB) is the owner of this three-loop, 822 MWe unit
located in Sweden. Initial criticality was achieved in June 1974.
T. Donald C. Cook No. 1 and No. 2
Indiana and Michigan Electric Company is the owner of these four-loop, 1090 MWe
plants located in Bridgman, Michigan. These plants are the first to use the
Westinghouse Ice Condenser Containment design. Initial criticality was achieved
in January 1975 for Unit 1 and March 1978 for Unit 2.
U. Trojan
This four-loop, 1130 MWe plant is jointly owned by Portland General Electric
Company, Eugene Water and Electric Board, and Pacific Power and Light
Company. In addition to being the first commercial nuclear plant to operate in the
Pacific Northwest (located on the Oregon shore of the Columbia River near
Rainier, Oregon), Trojan is the first 17 x 17 fuel-rod-per-assembly plant to achieve
criticality. Initial criticality was achieved in December 1975.
V. Beaver Valley No. 1
This three-loop, 852 MWe plant is jointly owned by Duquesne Light Company, Ohio Edison Company, and Pennsylvania Powe r Company. Beaver Valley No. 1 is located on the Ohio River, 22 miles northwest of Pittsburgh, Pennsylvania.
Commercial operation began in early 1976.
W. Salem No. 1 and No. 2
Salem No. 1 & 2, owned jointly by the Public Service Electric and Gas Company, Philadelphia Electric Company, Atlantic Electric Company, and Delmarva Power
and Light Company, is located on Artificial Island, a man-made peninsula in Salem
County, New Jersey. The 1090 MWe, four-loop plant achieved initial criticality for
Unit 1 in late 1976 and Unit 2 achieved criticality in August 1980.
FNP-FSAR-1 1.4-8 REV 25 4/14 X. North Anna No. 1 and No. 2 Virginia Electric and Power Company owns the two approximately 907 MWe (net)
plants located 40 miles north of Richmond, Virginia, on Lake Anna. Unit 1
achieved criticality in May 1978 and Unit 2 achieved criticality in June 1980.
Y. Joseph M. Farley No. 1 and No. 2
The two Alabama Power Company units are located at Dothan, Alabama (approximately 180 miles south - southwest of Atlanta, Georgia). Unit 1 achieved criticality in August 1977 and Unit 2 achieved criticality in February 1981. The two
units were originally designed with a gross electrical output of approximately 861
MWe. Modifications to the HP turbine in the 1980's improved the turbine efficiency
to approximately 885 MWe (gross). Subsequent modifications performed in
support of power uprate in the late 1990's increased predicted gross output to
approximately 910 MWe (approximately 883 MWe net). Replacement of the Unit 1
and Unit 2 low pressure turbines in 2010 and 2011, respectively, increased predicted gross output to approximately 933.4 MWe.
Z. Sequoyah No. 1 and No. 2
The two units approximately 1148 MWe (net) are located on the Tennessee River
near Chattanooga, Tennessee. These two units are owned by Tennessee Valley
Authority. Sequoyah No. 1 received a full power license in September 1980.
1.4.4.3 Westinghouse Facilities Westinghouse, in its effort to plan for the future, has developed a broad range of facilities to
satisfy the needs of the nuclear industry. The following paragraphs briefly describe these
facilities.
A. Columbia Plant, Nuclear Fuel Division
The Columbia Plant is capable of performing all operations necessary to
manufacture finished nuclear fuel assemblies. These operations include
conversion of uranium hexafluoride to urani um dioxide powder, fabrication of fuel assembly grids, complete pellet loading, and final fabrication of assemblies. The
plant, located at Columbia, South Carolina, began full production in early 1970.
The Columbia Plant is the largest commercial nuclear fuel fabrication facility in the
world.
B. Westinghouse Pensacola Plant
The Westinghouse Pensacola Plant, located on Escambia Bay on the northwest
coast of Florida, produces precision reactor vessel internals, steam generators, and pressurizers. Contributing to the precision manufacturing capability is an
environmental control system which mi nimizes year round temperature changes FNP-FSAR-1 1.4-9 REV 25 4/14 throughout the shop area. Transportation facilities of the plant include a railroad spur which permits loading and unloading inside the shop, and access to barge
loading facilities on Escambia Bay.
C. Cheswick Plant, Electro-Mechanical Division
The Electro-Mechanical Division was established in Cheswick, Pennsylvania, in
1953 to manufacture canned motor primary coolant pumps for nuclear reactors.
Today, the product line has expanded to include shaft seal pumps (reactor coolant
pumps), valves from 4 inches to 31 inches, and control rod drive mechanisms, essential components of the Westinghouse PWR. The facility occupies 250,000
square feet and now contains the most modern facilities available for the
production and testing of nuclear plant components.
D. Specialty Metals Division
The Specialty Metals Division located in Blairsville, Pennsylvania, was completed
in late 1967. Several essential PWR component processes are accomplished at
Blairsville, including: the precision manufacture of inconel tubing for steam
generators, and the complete processing of zircaloy seamless tubing for nuclear
fuel cladding. At Blairsville, complete quality control facilities are utilized for the
evaluation and analysis of all specialty metal products used in Westinghouse
nuclear systems.
E. Westinghouse Nuclear Center
The headquarters of Westinghouse Nuclear Energy Systems is located just east of
Pittsburgh, in Monroeville, Pennsylvania.
Operating primarily as a headquarters and engineering facility, the complex houses many of the divisions which
encompass Westinghouse's nuclear activities associated with the electric utility
industry.
F. Zion Nuclear Training Center
The Westinghouse Electric Corporation and the Commonwealth Edison Company
of Chicago have built and are operating a nuclear training center at Zion, Illinois.
The 28,000 square foot training center contains classrooms, a training reactor, training material center, video recording facilities, and multi-plant nuclear power
plant simulators .Westinghouse staffs and operates the center, supplies all the
equipment required, and is responsible for the development and presentation of all
training programs. Commonwealth Edison provided the building, access to the
Zion nuclear units for conducting in-plant observation training, and advises and
assists Westinghouse in developing training programs.
G. Instrumentation, Technology, and Training Center - Strategic Operations Division
The Strategic Operations Division was formed to address the man/machine
interface including instrumentation, control, and training. Man/machine interface
activities include the technical support center, plant computer development, FNP-FSAR-1 1.4-10 REV 25 4/14 instrumentation development, simulator, and software development related to all of these activities. The Strategic Operations Division also provides comprehensive
training for nuclear power plant personnel.
1.4.5 PLANT
CONSTRUCTION Southern Nuclear is responsible for operation of Farley Nuclear Plant and utilizes contractors as
appropriate to support operational, construction, and outage activities. Contractors are required
to conform to technical qualification and quality assurance requirements associated with
assigned work activities.
1.4.6 DIVISION
OF RESPONSIBILITY 1.4.6.1 Alabama Power Company (APC)
Alabama Power Company and Southern Nuclear O perating Company (SNC) are the holders of the facility license. In fulfilling its responsibility, APC has delegated certain activities to other
organizations as explained in the paragraphs covering SNC, Southern Company Services, Inc.
(SCS), Bechtel Power Corporation, and Westinghouse Electric Corporation, which follow.
1.4.6.2 Southern Nuclear Operating Company, Inc.
SNC is a licensee, along with APC, and is the sole operator of Farley Nuclear Plant. SNC
provides technical and general operations and maintenance services for nuclear operations, including (i) work relating to the licensing, operation, surveillance, testing, maintenance, quality
assurance, outage planning, health physics, plant production, security, retirement, decommissioning, training, emergency planning and responses, waste management, engineering and environmental studies, nuclear fuel supply studies, and core design
engineering; (ii) performance or procurement of engineering and design work associated with
SNC nuclear operations; and (iii) making available to APC the services of qualified technicians
or specialists, inspectors, and supervisory per sonnel for many phases of nuclear operations.
Administrative, technical, training, operations, maintenance, performance and planning, and
plant modifications personnel make up the plant operating staff of SNC.
In addition, SNC also provides administrative services in support of SNC nuclear operations.
These services include procurement services, accounting and statistical services, employee
relation services, and information management services.
Southern Company Services, Inc. had a traditional relationship with Alabama Power Company
based on many years of association together as member companies in The Southern Company.
During this association, SCS acted as the architect-engineer for APC in the design and
engineering of new fossil-fired steam generating units. For the Farley Nuclear Plant (FNP),
SCS had the original responsibility for developing and implementing the design for the turbine FNP-FSAR-1 1.4-11 REV 25 4/14 generator building and the balance of the plant not assigned to Bechtel, including the review or audit, approval, and documentation of the basic desi gn concepts, detail designs, specifications, and drawings.
As a result of the consolidation of SCS and SNC nuclear expertise and in addition to being the
licensee, SNC also serves as its own Architec t/Engineer and performs the functions previously performed by SCS.
1.4.6.3 Bechtel Power Corporation Bechtel Power Corporation has been retained by SNC to act as its consultant on the nuclear
project portion of the plant. In this capacity, Bechtel is responsible for the review or audit, approval, and documentation of the basic design concepts, detail designs, specifications, and
drawings for the reactor building, auxiliary building, and other structures and facilities of the
FNP. Bechtel also acts in an advisory capacity to SNC on other matters that may be assigned
to them from time to time.
1.4.6.4 Westinghouse Electric Corporation Westinghouse Electric Corporation has been awarded a contract by APC to design and
fabricate the nuclear steam supply system. The contract covers the standard 3-loop plant of
FNP-FSAR-1 REV 21 5/08 TABLE 1.4-1 (SHEET 1 OF 6)
WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTS Number Commercial MWe of Plant Owner Utility Location Operation Net Loops Yankee-Rowe Yankee Atomic Electric Company Massachusetts 1961(a) 175 4 Trino Versellese (Enrico Ente Nazionale per l'Energia Italy 1965 260 4 Fermi) Elettrica (ENEL)
Chooz (Ardennes) Societe d'Energie Nucleaire France 1967 305 4 Franco-Belge des Ardennes (SENA) San Onofre No. 1 Southern California Edison Co.; California 1968 450 3 San Diego Gas and Electric Co.
Haddam Neck (Connecticut Connecticut Yankee Atomic Power Connecticut 1968 575 4 Yankee) Company Jose Cabrera-Zorita Union Electrica, S. A. Spain 1969 153 1 Beznau No. 1 Nordostschweizerische Switzerland 1969 350 2 Kraftwerke AG (NOK)
Robert Emmett Ginna Rochester Gas and Electric New York 1970 490 2 Corporation Mihama No. 1 The Kansai Electric Power Japan 1970 320 2 Company, Inc.
Point Beach No. 1 Wisconsin Electric Power Co.; Wisconsin 1970 497 2 Wisconsin Michigan Power Co.
H. B. Robinson No. 2 Carolina Power and Light Co. South Carolina 1971 707 3 Beznau No. 2 Nordostschweizerische Switzerland 1971 350 2 Kraftwerke AG (NOK)
Point Beach No. 2 Wisconsin Electric Power Co.; Wisconsin 1972 497 2 Wisconsin Michigan Power Co.
Surry No. 1 Virginia Electric and Power Co. Virginia 1972 822 3 Turkey Point No. 3 Florida Power and Light Co. Florida 1972 745 3 FNP-FSAR-1 REV 21 5/08 TABLE 1.4-1 (SHEET 2 OF 6)
Number Commercial MWe of Plant Owner Utility Location Operation Net Loops Indian Point No. 2 Consolidated Edison Company of New York New York 1974 970 4 Prairie Island No.1 Northern States Power Company Minnesota 1973 530 2 Turkey Point No. 4 Florida Power and Light Co. Florida 1973 745 3 Surry No. 2 Virginia Electric and Power Co. Virginia 1973 822 3 Zion No. 1 Commonwealth Edison Company Illinois 1973 1,050 4 Kewaunee Wisconsin Public Service Corp.; Wisconsin 1974 503 2 Wisconsin Power and Light Co.;
Madison Gas and Electric Co.
Prairie Island No. 2 Northern States Power Company Minnesota 1974 530 2 Takahama No. 1 The Kansai Electric Power Japan 1974 781 3 Company, Inc.
Zion No. 2 Commonwealth Edison Company Illinois 1974 1,050 4 Beaver Valley No. 1 Duquesne Light Company; Ohio Pennsylvania 1976 852 3 Edison Company; Pennsylvania Power Company Doel No. 1 Indivision Doel Belgium 1975 390 2 Doel No. 2 Indivision Doel Belgium 1975 390 2 Donald C. Cook No. 1 Indiana and Michigan Electric Michigan 1975 1,090 4 Company (AEP)
Donald C. Cook No. 2 Indiana and Michigan Electric Michigan 1978 ,090 4 Company (AEP)
Indiana Point No. 3 Power Authority of the State New York 1976 965 4 of New York (PASNY)
Ko-Ri No. 1 Korea Electric Co. Korea 1978 564 2 Ringhals No. 2 Statens Vattenfallsverk (SSPB)
Sweden 1975 822 3
FNP-FSAR-1 REV 21 5/08 TABLE 1.4-1 (SHEET 3 OF 6)
Number Commercial MWe of Plant Owner Utility Location Operation Net Loops Trojan Portland General Electric Co.; Oregon 1976 1,130 4 Eugene Water and Electric Board; Pacific Power and Light Company Almaraz No. 1 Union Electrica, S. A.; Spain 1981 902 3 Compania Sevillana de Electricidad, S. A.;
Hidroelectrica Espanola, S.A.
Diablo Canyon No. 1 Pacific Gas and Electric Co. California 1985 1,084 4 Joseph M. Farley No. 1 Alabama Power Company Alabama 1977 829 3 Lemoniz No. 1 Iberduero, S.A. Spain --(c) 902 3 Salem No. 1 Pacific Service Electric and New Jersey 1977 1,090 4 Gas Company; Philadelphia Electric Co.; Atlantic Electric Co.; Delmarva Power and Light Co.
Sequoyah No. 1 Tennessee Valley Authority Tennessee 1981 1,148 4 Almaraz No. 2 Union Electrica, S. A.; Spain 1984 902 3 Compania Sevillana de Electricidad, S. A.;
Hidroelectrica Espanola, S. A.
Angra 1 dos Reis Furnas-Centrais Electricas, Brazil 1984 626 2 S. A.
Asco No. 1 Fuerzas Electricas de Cataluna, Spain 1983 902 3 S. A. (FECSA)
Diablo Canyon No. 2 Pacific Gas and Electric Co. California 1986 1,160 4 Joseph M. Farley No. 2 Alabama Power Company Alabama 1981 829 3 North Anna No. 1 Virginia Electric and Power Co. Virginia 1978 898 3 North Anna No. 2 Virginia Electric and Power Co. Virginia 1980 898 3 FNP-FSAR-1 REV 21 5/08 TABLE 1.4-1 (SHEET 4 OF 6)
Number Commercial MWe of Plant Owner Utility Location Operation Net Loops Ohi No. 1 The Kansai Electric Power Japan 1979 1,122 4 Company, Inc.
Ohi No. 2 The Kansai Electric Power Japan 1979 1,122 4 Company, Inc.
Ringhals No. 3 Statens Vattenfallsverk (SSPB) Sweden 1981 900 3 Sequoyah No. 2 Tennessee Valley Authority Tennessee 1982 1,148 4 Asco No. 2 Fuerzas Electricas de Cataluna, Spain 1986 902 3 S. A. (FESCA); Empresa Nacional Hidroelectrica del Ribagorzana, S. A. (ENHER);
Fuerzas Hidroelectricas de Cataluna, S.A.; Hidroelectrica del Segre, S. A.
Krsko Savske Elektrane Ljubljana; Yugoslavia 1983 615 2 Slovenia; Electroprivreda Zagreb, Croatia Lemoniz No. 2 Iberduero, S. A. Spain --(c) 902 3 Watts Bar No. 1 Tennessee Valley Authority Tennessee
--(b) 1,177 4 William B. McGuire No. 1 Duke Power Company North Carolina 1981 1,180 4 Millstone No. 3 Northeast Nuclear Energy Co.
Connecticut 1986 1,156 4 Ringhals No. 4 Statens Vattenfallsverk (SSPB) Sweden 1983 900 3 Salem No. 2 Public Service Electric and New Jersey 1981 1,115 4 Gas Company; Philadelphia Electric Co.; Atlantic Electric Co.; Delmarva Power and Light Co.
Virgil C. Summer South Carolina Electric and South Carolina 1984 900 3 Gas Company Watts Bar No. 2 Tennessee Valley Authority Tennessee
--(c) 1,177 4 FNP-FSAR-1 REV 21 5/08 TABLE 1.4-1 (SHEET 5 OF 6)
Number Commercial MWe of Plant Owner Utility Location Operation Net Loops William B. McGuire No. 2 Duke Power Company North Carolina 1984 1,180 4 Byron No. 1 Commonwealth Edison Co.
Illinois 1985 1,120 4 Catawba No. 1 Duke Power Company South Carolina 1985 1,153 4 Comanche Peak No. 1 Texas Utilities Generating Co. Texas --(b) 1,150 4 Ko-Ri No. 2 Korea Electric Co. Korea 1983 605 2 Seabrook New Hampshire Yankee New Hampshire 1990 1,200 4 South Texas Project Unit Houston Lighting and Power Co.; Texas 1988 1,250 4 No. 1 Central Power and Light Co.;
City Public Service of San Antonio; City of Austin, Texas Sayago No. 1 Iberduero, S. A. Spain --(c) 1,000 3 Vandellos No. 2 Asociacion Nuclear Vandellos Spain 1988 930 3 Beaver Valley No. 2 Duquesne Light Company; Ohio Pennsylvania 1987 852 3 Edison Company; Pennsylvania Power Co.; Cleveland Electric Illuminating Company; Toledo Edison Company Braidwood No. 1 Commonwealth Edison Company Illinois 1988 1,120 4 Callaway No. 1 SNUPPS - Union Electric Co. Missouri 1984 1,150 4 Braidwood No. 2 Commonwealth Edison Company Illinois 1988 1,120 4 Byron No. 2 Commonwealth Edison Company Illinois 1987 1,120 4 Catawba No. 2 Duke Power Company South Carolina 1986 1,153 4 Comanche Peak No. 2 Texas Utility Generating Co. Texas --(b) 1,150 4
FNP-FSAR-1 REV 21 5/08 TABLE 1.4-1 (SHEET 6 OF 6)
Number Commercial MWe of Plant Owner Utility Location Operation Net Loops Comanche Peak No. 2 Texas Utility Generating Co. Texas --(b) 1,150 4 South Texas Project Unit Houston Lighting and Power Co.; Texas 1989 1,250 4 No. 2 Central Power and Light Co.;
City Public Service of San Antonio; City of Austin, Texas Maanshan No. 1 Taiwan Power Company Taiwan 1984 890 3 Wolf Creek SNUPPS - Wolf Creek Nuclear Kansas 1983 1,188 4 Operating Corporation Alvin W. Vogtle No. 1 Georgia Power Company; Georgia 1987 1,113 4 Oglethorpe Electric Membership Corp., Municipal Authority of Georgia; City of Dalton, Georgia Alvin W. Vogtle No. 2 Georgia Power Company; Georgia 1989 1,113 4 Oglethorpe Electric Membership Corp., Municipal Authority of Georgia; City of Dalton, Georgia Maanshan No. 2 Taiwan Power Company Taiwan 1985 890 3 Shearon Harris No. 1 Carolina Power and Light Co. North Carolina 1987 900 3 Napot Point No. 1 National Power Corp. Phillippines 1985 620 2
_________
- a. In the decommissioning phase.
- b. Uncompleted.
- c. Indefinitely postponed.
FNP-FSAR-1 1.5-1 REV 21 5/08
1.5 REQUIREMENTS
FOR FURTHER TECHNICAL INFORMATION The design of the Farley Nuclear Plant Units 1 and 2 is based upon proven concepts which
have been developed and successfully applied to the design of numerous other pressurized water reactor systems.
The term "research and development" (R&D) as used in this section is the same as that used by
the Commission in Section 50.2 of its 10 CFR 50 as follows:
"(n) 'Research and development' means (1) theoretical analysis, exploration or
experimentation; or (2) the extension of investigative findings and theories of a scientific
nature into practical application for experimental and demonstration purposes including
the experimental production and testing of models, devices, equipment, materials and
processes."
The research and development discussed in the FSAR is to confirm the engineering and design
values normally used to complete equipment and system designs. It does not involve the
creation of new concepts or ideas.
The technical information generated is used either to demonstrate the safety of the design and
more sharply define margins of conservatism or to lead to design improvements.
Each research and development program is briefly summarized for identification and its
relationship to the Farley Nuclear Plant Units 1 and 2 is discussed. Detailed discussions of
each R&D program are available in a more expanded summary form in reference 1 and other
references as noted throughout this section.
1.5.1 PROGRAMS
REQUIRED FOR PLANT OPERATION In the Farley PSAR, three programs were identified as required for plant design and operation.
A. Core Stability Evaluation.
B. Fuel Rod Burst Program.
C. In-Pile Fuel Densification Program.
A. Core Stability Evaluation (Item 1 in reference 1)
The purpose of this program is to establish means for the detection and control of
potential xenon oscillations and for the shaping of the axial power distribution for
improved core performance. This program has been completed. Refer to
reference 1 for a further discussion of these tests.
B. Fuel Rod Burst Program (Item 2 in reference 1)
The original rod burst program, a study of the performance of zircaloy cladding
under simulated loss-of-coolant accident (LOCA) conditions, has been completed.
FNP-FSAR-1 1.5-2 REV 21 5/08 It has supplied empirical data from which the effect of geometry distortion on the ability of the emergency core cooling system (ECCS) to meet the LOCA design
criteria has been determined using present analytical design techniques.
The program included burst and quench tests on single rods and burst tests on rod
bundles. As a result of single rod tests, specific design limits have been
established on peak clad temperature and allowable maximum metal water
reaction to assure effective core cooling. The multirod burst tests demonstrated
that even when rod to rod contact does occur after burst, the remaining flow area is
always sufficient to ensure adequate core cooling.
The single rod burst test program for the 17 x 17 fuel pin array is discussed in
subsection 1.5.4.3.
C. In-Pile Fuel Densification (Item 22 in reference 1)
Operating experience with uranium dioxide fuel has indicated that the fuel may
densify under irradiation, to a density higher than that to which it was
manufactured. This densification can lead to shorter active fuel length stacks, increased initial pellet to clad radial gaps, and pellet to pellet axial gaps. The
shorter fuel stack length gives rise to a small increase in overall, average linear
power density (kW/ft). Increased radial gaps reduce gap conductance and lead to
higher pellet temperatures. Axial gaps give rise to local power peaking due to
decreased neutron absorption.
Westinghouse fuel densification research was directed toward producing fuel with
a structure that minimizes in-pile densification (hereafter called stable fuel). The
objective of the program was to define material characteristics and manufacturing
processes that lead to stable fuel. Stable fuel is defined as fuel with a small
densification. Residual effects of densification were evaluated on a model
developed by this program. A more detailed description of the program and results
is presented in reference 1.
1.5.2 OTHER
PROGRAMS REQUIRED FOR PLANT OPERATION Other areas of research and development, as outlined below, are those which give added
confirmation that the design is conservative.
A. Burnable Absorber Program (Item 7 in reference 1)
Burnable absorber rod development is complete. The burnable absorber rods for
the first core are borosilicate glass encased in stainless steel tubes. The fixed rods
are used to reduce the concentration of boric acid absorber in the moderator, thereby ensuring that the moderator temperature coefficient of reactivity is always
within its limit specified in the Core Operating Limits Report, as required by the
Technical Specifications. Refer to reference 1 for a further discussion of this
program.
FNP-FSAR-1 1.5-3 REV 21 5/08 B. Fuel Development Program for Operation at High Power Densities (Item 8 in reference 1)
To demonstrate satisfactory operation of fuel at high burnup and power densities, and to define design margins, a program was designed to test fuel in both the
Saxton and Zorita reactors. The Saxton loose-lattice irradiation program was
designed to demonstrate fuel performance at conditions significantly in excess of
PWR design limits, and to establish power burnup limits for the fuel. The Zorita
reactor is the first PWR with a zircaloy core to operate at similar core conditions as
the current design units. Because of the timely manner in which fuel can be
irradiated in Zorita, four fuel assemblies are being tested there to demonstrate
satisfactory operation of the fuel in a commercial PWR environment.
Sustained successful operation of special Zorita fuel rods at peak design power
levels, in excess of those planned for the Farley Nuclear Units, will increase
assurance that the fuel had adequate performance margins to accommodate
transient overpower operation.
The Saxton loose-lattice irradiation and Saxton parametric irradiation subprograms
have been completed. It is concluded that the loose lattice program has
satisfactorily completed the test objective. The work of the loose-lattice
assemblies was partly performed under USAEC Contract AT (11-1) - 3044 and has
been reported on a quarterly basis (reference 10); a fuel materials performance
report has been published. (See reference 11.)
C. FLECHT (Full Length ECCS Heat Transfer Test) (Item 12 in reference 1)
The objective of the FLECHT program was to obtain experimental reflooding heat
transfer data under simulated loss-of-coolant accident conditions for use in
evaluating the heat transfer capabilities of pressurized water reactor emergency core cooling systems.
The current test results verified the ability of a bottom flooding ECCS design to
terminate the temperature increase during a LOCA. The LOCA evaluation
presented in this application utilized the results of the FLECHT program for the
analysis of the reflooding phase of the accident.
D. Loss-of-Coolant Analysis Program (Item 14 in reference 1)
This program has been completed with the results of the Flashing Heat Transfer
Program (item 13 in reference 1) being incorporated in the core thermal design
code used in the LOCA analysis presented in this application.
The loss-of-coolant analysis program was established to integrate, as appropriate, the more realistic heat transfer models obtained from experimental and analytical
development programs into the core thermal design codes used to evaluate the
loss-of-coolant accident.
FNP-FSAR-1 1.5-4 REV 21 5/08 E. Reactor Vessel Thermal Shock (Item 16 in reference 1)
The effects of safety injection water on the integrity of the reactor vessel, following a postulated loss-of-coolant accident, have been analyzed using data on fracture
toughness of heavy section steel at the beginning of plant life and after irradiation
corresponding to approximately 40 years of equivalent plant life.(a) The results show that under the postulated accident conditions, the integrity of the reactor
vessel is maintained.
Fracture toughness data are obtained from a Westinghouse experimental program
which is associated with the Heavy Section Steel Technology (HSST) Program at
ORNL and EURATOM programs. Since results of the analyses are dependent on
the fracture toughness of irradiated steel, efforts are continuing to obtain additional
confirmatory data. Data on 2-in.-thi ck specimens became available in 1970 from the HSST program. These data indicated a strong temperature dependence with a
rapid increase in toughness at approximately NDT. For results obtained in the
HSST program, the HSST Semiannual Progress Report, issued by the Oak Ridge
National Laboratory (quarterly beginning in 1974), should be consulted.
F. Blowdown Forces Program (Item 15 in reference 1)
The objective of the Blowdown Forces program was to develop a digital computer
program for the calculation of pressure, velocity, and force transients in the reactor
coolant system during a loss-of-coolant accident, and to utilize this code in the
calculation of blowdown forces on the fuel assemblies and reactor internals to
ensure that the stress and deflection criteria used in the design of these
components were met.
Westinghouse has completed the development of BLODWN-2, an improved digital
computer program for the calculation of local fluid pressure, flow, and density
transients in the reactor coolant system during a LOCA.
Extensive comparisons have been made between BLODWN-2 and test data.
Agreement between code predictions and data has been good.
Analysis using the BLODWN-2 program to evaluate the effects of blowdown is
presented elsewhere in this application. It was concluded from the analysis that
the design of this reactor meets the established design criteria.
________________
- a. The operating licenses for both FNP units have been renewed and the original licensed
operating terms have been extended by 20 years.
Reactor vessel neutron embrittlement was evaluated as a time-limited aging analysis (TLAA) for license renewal in accordance with 10
CFR Part 54. The results of this analysis are provided in chapter 18, subsection 18.4.1.
FNP-FSAR-1 1.5-5 REV 21 5/08 G. ESADA DNB Program (Item 11 in reference 1)
The ESADA DNB program has been completed. The experimental program was
conducted with rod bundles to determine the effect on DNB from:
- 1. Axially uniform and nonuniform heat flux distributions.
- 2. Radially uniform and nonuniform heat flux distributions.
- 3. Mixing vane grids.
Data obtained covered ranges for such parameters as pressure, temperature, mass flow, and heat flux pertinent to present PWR design. Good data agreement
was obtained with the nonuniform heat flux W-3 DNB correlation which had been
developed on single channel data. The scatter of rod bundle data was decidedly
less than the scatter of single channel data used to develop the correlation.
The effects of the new 17 x 17 fuel assembly geometry on the DNB heat flux is
discussed in subsection 1.5.3.6.
1.5.3 17 x 17 FUEL ASSEMBLY VERIFICATION TESTS A comprehensive test program for the 17 x 17 assembly has successfully been completed by
Westinghouse. Reference 1 contains a summary discussion on the program.
Some of the verification work described herein was conducted using 17 x 17 assemblies of
seven-grid design, whereas the selected 17 x 17 assembly design has eight grids. Tabulated
below are those 17 x 17 tests which utilized a seven-grid geometry, and the effect of adding an
eighth grid.
Test Parameter Effect Fuel Assembly Axial Stiffness Negligible effect at Structural blowdown impact forces (9) test
Lateral Impact Additional grid shares impact load (9)
Prototype Pressure Drop The margin between 7-grid Assembly Test design P and D-loop results (10) is adequate to cover the additional P resulting from the additional grip (<5 percent increase in P)
FNP-FSAR-1 1.5-6 REV 21 5/08 Test Parameter Effect Lift Force The margin between 7-grid design lift force and D-loop results (10) is adequate to cover the additional lift force resulting from the additional grid Rod Vibration Decreased span length results in improved vibration characteristics and reduced rod wear Departure from DNB Correlation Addition on a grid increases Nucleate Boiling mixing which increases DNB margin Incore Flow TDC TDC increases as grid Mixing spacing decreases (4) The above tabulation shows that additional design changes are not required (e.g. no new fuel
assembly hold-down spring) due to the addition of a grid, and seven-grid test information can be
used to assess the adequacy of the eight-grid design. Additional testing to investigate the eight-
grid assembly specifically is not required.
1.5.3.1 Rod Cluster Control (RCC) Spider Tests The 17 x 17 rod cluster control (RCC) spider (subsection 4.2.3.2) is conceptually similar to, but geometrically different from, the 15 x 15 spider. The 17 x 17 spider supports 24 rodlets (the 15
x 15 design supports 20); with no vane supporting more than two rodlets (same as 15 x 15
design). The RCC spider tests verified the structural adequacy of the design.
Spider tests have been completed. A vertical static load test approximately seven times the
design dynamic load did not result in spider vane to hub joint failure. A spider was tested to 2.8
x 10 6 steps without failure. The spider loading was 110 percent of the design value for 1.8 x 10 6 cycles and 220 percent of the design loading for 1 x 10 6 cycles. Design load is 3600 lb compression and 1800 lb tension. The spring test resulted in negligible preload loss.
1.5.3.2 Grid Tests The 17 x 17 grid (subsection 4.2.1.2) is conceptually similar but geometrically different from the 15 x 15 "R" grid. The purpose of the grid tests is to verify the structural adequacy of the grid
design.
FNP-FSAR-1 1.5-7 REV 21 5/08 The grid tests have been completed. Test results are in agreement with pretest design values.
The test results, along with fuel assembly structural test results, were factored into the seismic
analysis. (See reference 9.)
1.5.3.3 Fuel Assembly Structural Tests The 17 x 17 fuel assembly (subsection 4.2.1.2) tests were performed to determine mechanical
strength and properties. The fuel assembly parameters obtained were as follows: lateral and
axial stiffness, impact and internal structural dam ping coefficients, vibrational characteristics, and the lateral and axial impact response for postulated accident loads. The parameters
obtained from the lateral dynamic tests are used for seismic analysis, while those obtained from
the axial tests are incorporated in the LOCA (blowdown) accident analysis.
The fuel assembly structural tests have been completed. The fuel assembly structural test
results are factored into the seismic and blowdown analyses.
(9) 1.5.3.4 Guide Tube Tests To verify the structural adequacy of the guide tubes, an extensive series of tests was conducted
to determine guide tube deflection with simulated blowdown forces comparable to those
expected during a LOCA and to determine the maximum acceptable deflection which ensures
insertion of a control rod by free fall. Additional tests were conducted to determine fatigue
strength, displacement as a function of strain, and the natural frequencies of the guide tubes for
use in dynamic analyses. Refer to references 4 and 15 for a discussion of these items.
1.5.3.5 Prototype Assembly Tests The purpose of these tests was to demonstrate that the 17 x 17 fuel assembly and control rod
hardware designs perform as predicted. Two pr ototype assemblies were sequentially tested in order to obtain the required experimental data. A single set of control rod hardware, including
driveline, was used in the tests. The fuel assemblies were subjected to flow and system
conditions covering those most likely to occur in a plant during normal operation, as well as
during a pump overspeed transient. Seismic testing was not included in the test sequence.
These tests were used to verify the integrated fuel assembly and RCC performance in several
areas. Data obtained includes pressures and pressure drops throughout the system, hydraulic
loadings on the fuel assembly and drive line, control rod drop time and stall velocity, fuel rod
vibration and control rod, driveline, guide tube, and guide thimble wear during a lifetime of
operation.
1.5.3.6 Departure from Nucleate Boiling (DNB)
The effect of the 17 x 17 fuel assembly geometry on the departure from nucleate boiling (DNB)
heat flux has been determined experimentally and has been incorporated in a modified spacer FNP-FSAR-1 1.5-8 REV 21 5/08 factor for use with the W-3 correlation. The effect of cold-wall thimble cells in the 17-x-17 geometry has also been quantified.
A similar program was conducted to quantify the DNB performance of the R-type mixing vane
grid as developed for the 15-x-15 fuel assembly design.
(2) (3) The results of that program were used to develop a modified spacer factor which quantifies the power capability associated with
the use of the R mixing vane grid as well as the change in power capability due to the axial
spacing of the grids. The modified spacer factor, along with the W-3 correlation with the cold
wall factor, was shown to be applicable to cold wall thimble cells in the 15-x-15 geometry.
(3) 1.5.3.7 Incore Flow Mixing In the thermal-hydraulic design of a reactor core, the effect of mixing or turbulent energy transfer within the hot assembly is evaluated using the THINC code. The rate of turbulent energy
transfer is formulated in the THINC analysis in terms of a thermal diffusion coefficient (TDC).
A program (4) to determine the proper value of TDC for the R grid vane, as used in the 15-x-15 fuel assembly design, has been completed and showed that a design value of 0.038 (for 26-in.
spacing) can be used for TDC. These results also showed that TDC was independent of
Reynold's number, mass velocity, pressure, and quality over the ranges tested.
A similar TDC experimental program employed a geom etry typical of the 17 x 17 fuel assembly to determine the effects of the geometry on mixing and an appropriate value for TDC. A uniform
axial heat flux was used. There is no analytical reason to expect that the mixing coefficient
would be affected by a nonuniform axial heat fl ux. The THINC computer code considers the mixing in each increment along the heated length; within that increment, the heat flux is
considered uniform. The tests reported by Cadek (8) indicate that there was no difference, within experimental accuracy, between a test section with a uniform flux (Pitt) and 1/2 a cosine flux (Columbia). The heat flux varied between the simulated fuel rods in the test section to create a
thermal gradient in the radial direction. Using different flow rates and inlet temperatures, the
TDC for the 17 x 17 geometry was determined.
Status The TDC tests are completed and the results are reported in reference 12 and summarized in
subsection 4.4.3.3.
1.5.4 LOCA HEAT TRANSFER TESTS (17 x 17)
Extensive experimental program s have been completed to determine the thermal hydraulic characteristics of 15 x 15 fuel assemblies, and to obtain experimental reflooding heat transfer
data under simulated loss-of-coolant accident conditions.
Complementary experimental programs were completed with a simulated 17 x 17 assembly to
determine its behavior under similar LOCA conditions. Results from the 17 x 17 programs were FNP-FSAR-1 1.5-9 REV 21 5/08 compared with data from the 15 x 15 assembly test programs and were used to confirm predictions made by correlations and codes based on the 15 x 15 test results.
1.5.4.1 Blowdown Heat Transfer Testing (Formerly Titled Delayed Departure From Nucleate Boiling) The NRC Acceptance Criteria for emergency core cooling systems (ECCS) for light-water power
reactors was issued in Section 50.46 of 10 CFR 50 on December 28, 1973. It defines the basis
and conservative assumptions to be used in the evaluation of the performance of emergency
core cooling systems. Westinghouse believes that some of the conservatism of the criteria is
associated with the manner in which transient DNB phenomena are treated in the evaluation
models. Transient critical heat flux data presented at the 1972 specialists meeting of the
Committee on Reactor Safety Technology (CREST) indicated that the time to DNB can be
delayed under transient conditions. To demonstrate the conservatism of the ECCS evaluation models, Westinghouse initiated a program to ex perimentally simulate the blowdown phase of a
LOCA. This testing is part of the Electric Power Research Institute (EPRI) sponsored Blowdown
Heat Transfer Program, which was started early in 1976. Testing was completed in 1979. A
DNB correlation will be developed by Westinghouse from these test results for use in the ECCS
analyses.
Program
The program was divided into two phases. The Phase I tests started from steady state
conditions, with sufficient power to maintain nucleate boiling throughout the bundle, controlled
ramps of decreasing test section pressure or flow initiated DNB. By applying a series of
controlled conditions, investigation of the DNB was studied over a range of qualities and flows, and at pressures relevant to a PWR blowdown.
Phase I provided separate-effects data for heat transfer correlation development.
Typical parameters used for Phase I testing are as shown:
Initial Steady State Conditions Nominal Valve Pressure 1250 to 2250 psia Test section mass velocity 1.12 to 2.5 x 10 6 lb/h-ft 2 Core inlet temperature 550 to 600°F Maximum heat flux 306,000 to 531,000 Btu/h-ft 2 Transient Ramp Conditions Pressure decrease 0 to 350 psi/s and subcooled depressurization from 2250 psia Flow decrease 0 to 100 percent/s Inlet enthalpy Constant FNP-FSAR-1 1.5-10 REV 21 5/08 Phase II simulated PWR behavior during a LOCA to permit definition of the time delay associated with onset of DNB tests in this phase covered the large double-ended guillotine cold
leg break. All tests in Phase II were also started after establishment of typical steady state
operating conditions. The fluid transient was then initiated, and the rod power decay was
programmed in such a manner as to simulate the actual heat input of fuel rods. The test was
terminated when the heater rod temperatures reach a predetermined limit.
Typical parameters used for Phase II testing are as shown:
Initial Steady State Conditions Nominal Valve Pressure 2250 psia Test section mass velocity 2.5 x 10 6 lb/h-ft 2 Inlet coolant temperature 545°F Maximum heat flux 531,000 Btu/h-ft 2 Transient Conditions Simulated break Double-ended cold leg guillotine breaks Test Description
The experimental program was conducted in the J-loop at the Westinghouse Forest Hills Facility
with a full length 5 x 5 rod bundle simulating a section of a 15 x 15 assembly to determine DNB
occurrence under LOCA conditions.
The heater rod bundles used in this program were internally-heated rods, capable of a
maximum power of 18.3 kW/ft, with a total power of 135 kW (for extended periods) over the 12-
foot heated length of the rod. Heat was gener ated internally by means of a varying cross-sectional resistor which approximates a chopped cosine power distribution. Each rod was
adequately instrumented with a total of 12 clad thermocouples.
1.5.4.2 Single Rod Burst Test (SRBT)
The single rod burst test (SRBT) results were used to quantify the maximum assembly flow
blockage which is assumed in LOCA analyses.
Previously, single rod and multi-rod burst test (MRBT) have been completed on 15 x 15 fuel
assembly rods under conditions which exist during the loss-of-coolant accident. The conclusion
of these tests were that fuel rods burst in a staggered manner, so that maximum average
assembly-wise flow area blockage is 55 percent during blowdown and 65 percent during
reflood, based on the characteristics of the pressurized PWR fuel rod and the conservative peak clad temperature predicted during the LOCA transient.
The single rod burst test program for the 17 x 17 fuel assembly rods consisted of testing
specimens at the two internal pressures and the three heating rates listed below in a steam
atmosphere.
FNP-FSAR-1 1.5-11 REV 21 5/08 Heating Rate Internal Pressure (725°F to 1940°F) psi 5°F/s 1200, 1800 25°F/s 1200, 1800
100°F/s 1200, 1800 All specimens are then heated 5°F/s from 1940°F to about 2300°F, held for a short time, and
then cooled 5°F/s to 1200°F.
Metallography is done on specimens to determine the degree of wall thinning and the extent of
oxygen embrittlement.
In addition, tests were run on 15 x 15 fuel assembly rods to insure reproducibility of the 1972
single rod burst test results.
Facility The SRB tests are conducted in the Westinghouse Engineering Mechanics Laboratory in an
electrically heated furnace.
Status The single rod burst tests are complete and results are reported in reference 13. Results of
initial tests showed that the LOCA behavior of 17 x 17 clad in comparison to that of 15 x 15 clad
exhibited no significant differences in failure ductility. Because of the result and the geometry
sealing, the flow blockage (90 percent) as determined by 15 x 15 MRBT simulation can be used
for 17 x 17 fuel geometry.
FNP-FSAR-1 1.5-12 REV 21 5/08 REFERENCES
- 1. "Safety Related Research and Development for Westinghouse Pressurized Water Reactors, Program Summaries, Spring 1974," WCAP 8353 , September 1974.
- 2. Motley, F. E. and Cadek, F. F., "DNB Results for New Mixing Vane Grid (R), "WCAP-7695-L , July 1972, (Westinghouse Proprietary), and WCAP-7958 , October 1972.
- 3. Motley, F. E. and Cadek, F. F., " DNB Test Results for R Grids with Thimble Cold Wall Cells," WCAP-7695-L , Addendum 1, October 1972, (Westinghouse Proprietary), and WCAP-7958 , Addendum 1, October 1972.
- 4. Cadek, F. F., Motley, F. E., and Dominicis, D. P., "Effect of Axial Spacing on Interchannel Thermal Mixing with R Mixing Vane Grid," WCAP-7941-L , June 1972, (Westinghouse Proprietary), and WCAP-7959 , October 1972.
- 5. Tong, L. S., "Prediction of Departure From Nucleate Boiling for an Axially Non-Uniform Heat Flux Distribution," J. Nucl. Energy, 21, pp. 241-248 (1967).
- 6. Wilson, R. H., Stanek, L. J., Gellerstedt, J. S., and Lee, R. A., "Critical Heat Flux in a Nonuniformly Heated Rod Bundle," in Two-Phas e Flow and Heat Transfer in Rod Bundles, pp, 56-6 2, ASME, New York, November 1969.
- 7. Rosal, E. R., et al., "Rod Bundle Axial Non-Uniform Heat Flux Tests and Data," WCAP-7411 , December 1969 (Westinghouse Proprietary), and WCAP-7813 , December 1971.
- 8. Cadek, F. F., "Interchannel Thermal Mixing with Mixing Vane Grids," WCAP-7667-L , May 1971 (Westinghouse Proprietary), and WCAP-7755 , September 1971.
- 9. Gesinski, L., Chiang, D., and Nakazato, S., "Safety Analysis of the 17 x 17 Fuel Assembly" for Combined Seismic and Loss of Coolant Accident," WCAP 8288 , December 1973.
- 10. De Mario, E. E. and Nakazato, S., "Hydraulic Flow Test of the 17 x 17 Fuel Assembly, WCAP 8279 , February 1974.
- 11. Hill, K. W., Motley, F. E., Cadek, F. F., and Wenzel, A. H., "Effect of 17 x 17 Fuel Assembly Geometry on DNB," WCAP 8297 , March 1974.
- 12. Motley, F. E., Wenzel, A. H., and Cadek, F. F., "The Effect of 17 x 17 Fuel Assembly Geometry on Interchannel Thermal Mixing," WCAP 8299 , March 1974.
- 13. "Irradiation of 17x17 Demonstration Assemblies in Surry Units No. 1 and 2, Cycle 2," WCAP-8362 , July 1974.
- 14. Kochirka, P., "17x17 Design Fuel Rod Behavior during Simulated Loss-of-Coolant Accident," WCAP-8289 , (W Proprietary), WCAP-8290 , November 1974.
FNP-FSAR-1 1.5-13 REV 21 5/08
- 15. Cooper, F.W., Jr., "17 x 17 Driveline Component Tests - Phase IB, II, III, D-Loop Drop and Deflection" WCAP-8446 (W Proprietary) and WCAP-8449 (Non-proprietary), December 1974.
FNP-FSAR-1 1.6-1 REV 22 8/09 1.6 MATERIAL INCORPORATED BY REFERENCE
1.6.1 WESTINGHOUSE
TOPICAL REPORTS
1.6.1.1 Reports Referenced in FSAR
This section lists Westinghouse Topical Reports referenced throughout the FSAR, which provide bases for the design information presented.
REPORT TITLE SECTION REFERENCE NRC SUBMITTAL
- 1. PWR Staff, "Westinghouse Technical Position on Discrete Break Locations and Types for LOCA Analysis of the Primary Coolant Loop," Westinghouse Topical Report WCAP-8082 , May 1973.
3.6 June 1973 2. L. T. Gesinski, "Fuel Assembly Safety Analysis for Combined Seismic and Loss-of-Coolant Accident," WCAP-7950 , July 1972.
3.7, 3.9 July 1972
- 3. E. L. Vogeding, "Seismic Testing of Electrical and Control Equipment," WCAP-7817 and Supplements, December 1971, and WCAP-7897-L (W Proprietary). 3.7, 3.10, 7.6 January 1982
- 4. B.E. Olson, G. J. Bohn, "Indian Point No. 2 Primary Loop Vibration Test Program," WCAP-7662 (W Proprietary) and WCAP-7920 , September 1972.
3.9 August
1973 5. G. J. Bohn, "Indian Point No. 2 Internals Mechanical Analysis for Blowdown Excitation," WCAP-7822, December 1971, and WCAP-7332-L (W Proprietary).
3.9 December
1971 6. S. Kraus, "Neutron Shielding Pads," WCAP-7870 , May 1972. 1.3, 3.9, 4.2 July 1972
- 7. A. J. Kuenzel, "Westinghouse PWR Internals Vibration Summary 3-Loop Internals Assurance,"
WCAP-7765, September 1971, and WCAP-7765-L (W Proprietary).
3.9 October
1971 FNP-FSAR-1 1.6-2 REV 22 8/09 REPORT TITLE SECTION REFERENCE NRC SUBMITTAL
- 8. J. M. Hellman (Ed.), "Fuel Densification Experimental Results and Model For Reactor
Operation," WCAP-8219 , October 1973, and WCAP-8218 , October 1973 (W Proprietary). 4.1, 4.2, 4.3, 4.4, 15.3 October 1973
- 9. L. Gesinski, K. Chiang, S. Nakazato, "Safety Analysis of the 17 x 17 Fuel Assembly For Combined Seismic and Loss-of-Coolant Accident," WCAP-8288 , December 1973, and WCAP-8238. 4.2, 1.5 March 1973
- 10. W. J. Dollard, "Nuclear Fuel Division Quality Assurance Program Plan," WCAP-7800 , Revision 3, November 1979. 4.2, 17C.1 October 1973
- 11. E. E. Demario and S. Nakazato, "Hydraulic Flow Test of the 17 x 17 Fuel Assembly," WCAP-8279 , February 1974, and WCAP-8278. 1.5, 4.2, 4.4 March 1973
- 12. F. B. Skogen and A. F. McFarlane, "Control Procedures for Xenon-Induced X-Y Instabilities in Large PWRs" WCAP 3680-21 , (EURACE-2111), February 1969.
4.3 September
1969 13. R. F. Barry, et al., "The PANDA Code," WCAP-7757, April 1967.
4.3 December
1973 14. W.C. Gangloff and W.D. Loftus, "An Evaluation of Solid State Reactor Protection in Anticipated
Transients," WCAP-7706 July 1971, and WCAP-7706-L (W Proprietary) September 1971. 4.3, 7.1, 7.2, 7.3 September 1971
- 15. J. S. Moore, "Nuclear Design of Westinghouse PWRs with Burnable Poison Rods," WCAP-7806 , December 1971.
4.3 December
1971 16. S. Altomare and R. F. Barry, "The TURTLE 24.0 Diffusion Depletion Code," WCAP-7758 , September 1971, and WCAP-7213 , June 1968 (W Proprietary). 4.3, 15.1, 15.2, 15.3 December 1973
- 17. F.E. Motley and F.F. Cakek, "DNB Test Results for New Mixing Vane Grids (R)," WCAP-7695-L , July 1972 (W Proprietary), and WCAP-7958 , October 1972. 4.4, 1.5 July 1972
FNP-FSAR-1 1.6-3 REV 22 8/09 REPORT TITLE SECTION REFERENCE NRC SUBMITTAL
- 18. F.F. Cadek, "Interchannel Thermal Mixing with Mixing Vane Grids," WCAP-7667-L , May 1971 (W Proprietary), and WCAP-7755 , September 1971. 4.4, 1.5 May 1971 19. F.F. Cadek, F.E. Motley, and D. P. Dominicis, "Effects of Axial Spacing on Interchannel Thermal Mixing with the (R) Mixing Vane Grid," WCAP-7941-L, June 1972 (W Proprietary) and WCAP-7755 , September 1971.
4.4, 1.5 July 1972
- 20. L.E. Hochreiter, "Application of the THINC-IV Program to PWR Design," WCAP-8054 , October 1973, (W Proprietary), and WCAP-8195 , October 1974 4.4 December 1973 21. F.D. Carter, "Inlet Orificing of Open PWR Cores," WCAP-9004 , January 1969 (W Proprietary), and WCAP 7836, January 1972.
4.4 March
1969 22. K.W. Hill, F.E. Motley, and F.F. Cadek, "Effect of Local Heat Flux Spikes on DNB in Non Uniform Heated Rod Bundles," WCAP-8174, August 1973 (W Proprietary), and WCAP 8202, August 1973.
4.4 December
1973 23. S. Nakazato and E.E. DeMario, "Hydraulic Flow Test of the Fuel Assembly," WCAP-8279 , February 1974. 4.4, 1.5 March 1974
- 24. F.E. Motley and F.F. Cadek, "DNB Test Results for R Grid Thimble Cold Cells," WCAP-7958 , and WCAP-7958 , Addendum J.
1.5, 4.4 March 1973
- 25. D.H. Risher, Jr., "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," WCAP-7588, Revision 1, December 1971.
15.4 January 1972 26. L.E. Hochreiter, H. Chelemer, and P.T. Chu, "THINC-IV, An Improved Program for Thermal-
Hydraulic Analysis of Rod Bundle Cores," WCAP-7956 , June 1973.
4.4 October
1973 FNP-FSAR-1 1.6-4 REV 22 8/09 REPORT TITLE SECTION REFERENCE NRC SUBMITTAL
- 27. F.F. Cadek, K.W. Hill, F.E. Motley, and A.H. Wenzel, "Effects of 17 x 17 Fuel Assembly Geometry on DNB," WCAP-8297 , March 1974.
4.4, 1.5 April 1974
- 28. F.E. Motley, A.H. Wenzel, and F.F. Cadek, "The Effect of 17 x 17 Fuel Assembly Geometry on Interchannel Thermal Mixing," WCAP-8299 , March 1974. 4.4, 1.5 March 1974
- 29. F.E. Motley and F.F. Cadek, "Application of Modified Spacer Factor to L Grid Typical and Cold
Wall Cell DNB," WCAP-8030 , December 1972.
4.4 January
1973 30. W.S. Hazelton, "Sensitized Stainless Steel in Westinghouse PWR Nuclear Steam Supply
Systems," WCAP-7735, August 1971, and WCAP-7477-L (W Proprietary)
5.4 August
1971 31. K. Cooper, R.M. Starek and V. Miselis, "Overpressure Protection for Westinghouse Pressurized Water Reactors," WCAP-7769 , Revision 1, June 1972.
5.2 June 1972 32. W.S. Hazelton, S.L. Anderson and E.E. Yanichko, "Basis for Heatup and Cooldown Limit Curves,"
WCAP-7924, August 1972.
5.2 August
1972 33. F. Bordelon and A. Nahavandi, "A Space Dependent Loss-of-Coolant Accident and Transient Analysis for PWR Systems (SATAN Digital Computer Code)," WCAP-7854 , January 1972.
5.2 - 34. J. Locante and E.G. Igne, "Environmental Testing of Engineered Safety Features Related Equipment (NSSS-Standard Scope)," WCAP-7744 , Volume I, August 1971 and WCAP-7410-L Volume I, (W Proprietary). 3.11, 6.3, 7.3 September 1971
- 35. M.J. Bell, J.E. Bulkowski. L.F. Picone, "Investigation of Chemical Additives for Reactor Containment Sprays," WCAP-7154, March 1968 (W Proprietary).
6A February 1968 FNP-FSAR-1 1.6-5 REV 22 8/09 REPORT TITLE SECTION REFERENCE NRC SUBMITTAL
- 36. R. Bartholomew and J. Lipchak, "Test Report, Nuclear Instrumentation System Isolation
Amplifier," WCAP-7819 , Revision 1, January 1972.
7.2 January
1972 37. I. Garber, "Isolation Tests Process Instrumentation Isolation Amplifier Westinghouse Computer and Instrumentation Division Nucana 7300 Series,"
WCAP-7862, August 1972.
7.2 September
1972 38. D.N. Katz, "Solid State Logic Protection System Description," WCAP-7672, June 1971. 7.1, 7.2, 7.3 May 1971
- 39. J.T. Haller, "Engineered Safeguards Final Device or Actuator Testing," WCAP-7705, May 1972.
7.3 March
1974 40. A.E. Blanchard and D.N. Katz, "Solid State Rod Control System, Full Length," WCAP-7778 , December 1971, and WCAP-9012-L (W Proprietary).
7.7 December
1971 41. J.B. Reid, "Process Instrumentation for Westinghouse Nuclear Steam Supply System," WCAP-7913, January 1973. 7.1, 7.2, 7.3 March 1973
- 42. A.E. Blanchard, "Rod Position Monitoring," WCAP-7571 , March 1971.
7.7 April
1971 43. A.E. Blanchard, "Part Length Rod Control System," WCAP-7406, March 1971.
7.7 April
1971 44. "Source Term Data for W PWRs," Westinghouse Electric Corporation, Pittsburgh, PA, WCAP-8253 , April 1970.
11A.3 June 1974 45. D.H. Risher Jr., and R.F. Barry, "TWINKLE-A Multi-Dimensional Neutron Kinetics Computer Code,"
WCAP-8028 January 1973 (W Proprietary), WCAP-7979 November 1972. 15.1, 15.4, 15.2, 15.3 January 1970
- 46. F.M. Bordelon, "Calculation of Flow Coastdown After Loss of Reactor Coolant Pump (PHOENIX
Code)," WCAP-7973 , January 1973.
15.1 January 1973 FNP-FSAR-1 1.6-6 REV 22 8/09 REPORT TITLE SECTION REFERENCE NRC SUBMITTAL
- 47. D.B. Fairbrother and H.G. Hargrove, "WIT-6 Reactor Transient Analysis Computer Program Description," WCAP-7980, November 1972. 15.1, 15.2 September 1973
WCAP-7898, June 1972 and WCAP-7501. 15.1, 15.2 September 1972
- 49. T.W.T. Burnett, C.J. McIntyre, J.C. Buker, R.P. Rose, "LOFTRAN Code Description," WCAP-7907 , June 1972 and WCAP-7877 , October 1972. 15.2, 15.3 October 1972
- 50. C. Hunin, "FACTRAN, a Fortran IV Code for Thermal Transients in a UO 2 Fuel Rod," WCAP-7908, June 1972 and WCAP-7337. 15.2, 15.3 September 1972
WCAP-7909, June 1972 and WCAP-7635. 15.1, 15.2, 15.4 October 1972
- 52. M.A. Mangan, "Overpressure Protection for Westinghouse Pressurized Water
Reactors," WCAP-7769 , October 1971.
15.2 June 1972 53. V.J. Esposito, K. Kesavan, B.A. Maul, "WFLASH-A FORTRAN-IV Computer Program for Simulation of Transients in a Multiloop PWR," WCAP-8261 Rev. 1, July 1974 and WCAP-8200, Rev. 2.
15.3 July 1974 54. F.M. Bordelon, et al., "LOCTA-IV Program: Loss of Coolant Transient Analysis," WCAP-8305 , June 1974. 15.3, 15.4 July 1974
- 55. R.D. Kelly, et al., "Calculational Model for Core Reflooding after a Loss-of-Coolant Accident (WREFLOOD Code)," WCAP-8171, June 1974 and WCAP-8170. 15.4 July 1974 56. F.M. Bordelon and E.T. Murphy, "Containment Pressure Analysis Code (COCO)," WCAP-8326 , June 1974 and WCAP-8327. 15.4 July 1974 FNP-FSAR-1 1.6-7 REV 22 8/09 REPORT TITLE SECTION REFERENCE NRC SUBMITTAL
- 57. T.L. Buterbaugh, W.J. Johnson and S.D. Kopelic, "Westinghouse ECCS - Plant Sensitivity Studies," WCAP-8356, July 1974 and WCAP-8340
.15.3, 15.4 August 1974
- 58. F.M. Bordelon, W. Massie and T.A. Zordan, "Westinghouse ECCS Evaluation Model -
Summary" WCAP-8339 , July 1974. 15.3, 15.4 July 1974
- 59. F.M. Bordelon, et al., "SATAN-VI Program: Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant," WCAP-8306 , June 1974.
15.4 July 1974 60. K.D. Shepard, S. Cerni, J.R. Reavis, "An Evaluation of Fuel Rod Bowing," WCAP-8346 , May 1974. 4.4 May 1974 61. K.W. Hill, F.E. Motley, F.F. Cadek, "Effect of Bowed Rods on DNB," WCAP-8176 (W Proprietary) and WCAP-8323. 4.4 May 1974 62. H. Chelemer, J. Weisman, and L.S. Tong, "Subchannel Thermal Analysis of Rod Bundle
Cores," WCAP-7015 Rev. 1, January 1969.
4.4 February
1969 63. Wilson, J.F., "Electric Hydrogen Recombiner IEEE 323-1974 Qualification," WCAP-7709-6 Supplement 6 (Proprietary) October 1976, WCAP-7820 Supplement 6 (Nonproprietary), October 1976. 3.11, 6.2 October 1976
- 64. Lee, H. "Prediction of the of the Flow Induced Vibration Reactor Internals by Scale Model Tests,"
WCAP-8303 (W Proprietary) and WCAP-8317 , May 1974.
3.9 May 1974 65. P. Kochirka "17 x 17 Design Fuel Rod Behavior During Simulated Loss-of-Coolant
Accident," WCAP-8289 (W Proprietary)
WCAP-8290 November 1974. 1.5 November 1974 66. R.A. George, Y.C. Lee, G.H. Eng, "Revised Clad Flattening Model," WCAP-8377 (Westinghouse Proprietary) and WCAP-8381 (Nonproprietary), July 1974
4.2 August
1974 FNP-FSAR-1 1.6-8 REV 22 8/09 REPORT TITLE SECTION REFERENCE NRC SUBMITTAL
- 67. T.M. Burke, C.E. Meyer, and J. Shefcheck, "Analysis of Data from the Zion (Unit 1) THINC Verification," WCAP-8453 (Westinghouse Proprietary) and WCAP-8454 (Nonproprietary), December 1974.
4.4 January
1975 68. K.M. Vashi, "Documentation of Selected Westinghouse Structural Analysis Computer Codes," WCAP-8252 , April 1974.
Appendix 3L, 3M.6 June 1975
- 69. "Westinghouse ECCS Evaluation Model - October 1975 Version," WCAP-8623 , November 1975 (Nonproprietary). 15.4 November 1975 70. Julian, H.V., Tabone, C.J., Thompson, C.M., "Westinghouse ECCS-Three Loop Plant (17x17)
Sensitivity Studies," WCAP-8853, October 1976, (Non-proprietary).
15.4 October 1976 71. Vogeding, E.L., "Seismic Testing of Electrical and Control Equipment for Low Seismic Plants,"
WCAP-7817 Supplement 7, September 1976. 3.10, 7.6 October 1976
- 72. Figenbaum, E., "Seismic Testing of Electrical and Control Equipment for High Seismic Plants,"
WCAP-7821 Supplement 2, Addendum 1, November 1975.
October 1976 73. Jarecki, S.J., Coslow, B.J., Croasdaile, T.R., Lipchak, J.B., "Seismic Operability Demonstration Testing of the Nuclear Instrumentation System Bistable Amplifier," WCAP-8830 (Proprietary)
October 1976.
Ch. 3 November 1976
- 74. Land, R.E., "Mass and Energy Release Following Main Steam Ruptures," WCAP-8822 (Proprietary) September 1976 and WCAP-8860 (Nonproprietary)
September 1976.
6.2 September
1976 75. F. F. Cadek, F. E. Motley, and D. P. Dominicis, "Effect of Axial Spacing on Inter-channel Thermal Mixing with R Mixing Vane Grid," WCAP-7941-L , June 1972, (Westing-house Proprietary), and WCAP-7959 , October 1972.
1.5, 4.4 July 1972
FNP-FSAR-1 1.6-9 REV 22 8/09 REPORT TITLE SECTION REFERENCE NRC SUBMITTAL
- 76. L. S. Tong, "Prediction of Departure from Nucleate Boiling for an Axially Nonuniform Heat Flux Distribution," J. Nucl. Energy , 21 pp. 241-248 (1967). 1.5 July 1972 77. R. H. Wilson, L. J. Stanek, J. S. Gellerstedt, and R. A. Lee, "Critical Heat Flux in a Nonuniformly Heated Rod Bundle," in Two-Phase Flow and Heat Transfer in Rod Bundles, pp. 56-7, ASME, New York, November 1969.
1.5 July 1972 78. E. R. Rosal, et al
., "Rod Bundle Flux Tests and Data," WCAP-7411 , December 1969 (W Proprietary), and WCAP-7813 , December 1971. 1.5 July 1970 79. F. F. Cadek, "Interchannel Thermal Mixing with Mixing Vane Grids," WCAP-7667-L , May 1971 (W Proprietary), and WCAP-7755 , September 1971. 1.5, 4.4 July 1970
- 80. L. Gesinski, D. Chiang, and S. Nakazato, "Safety Analysis of the 17 x 17 Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident," WCAP-8288 , December 1973.
1.5, 4.2 - 81. "Irradiation of 17 x 17 Demonstration Assemblies in Surry Units No. 1 and 2, Cycle 2," WCAP-8362 , July 1974.
1.5 - 82. F. W. Cooper, Jr, "17 x 17 Driveline Component Tests-Phase IB, II, III, D-Loop Drop and Deflection," WCAP-8446 (W Proprietary) and WCAP-8449 (Non-proprietary), December 1974.
1.5 January
1975 1.6.2 BECHTEL POWER CORPORATION TOPICAL REPORTS Report No.
Title
- 1. B-TOP-3 "Design Criteria for Nucl ear Power Plants Against Tornadoes," March 1970
- 2. BP-TOP-1 "Seismic Analysis of Piping System," Revision 1, February 1974
FNP-FSAR-1 1.6-10 REV 22 8/09 Report No.
Title
- 3. BC-TOP-4 "Seismic Analysis of Structures and Equipment for Nuclear Power Plants," Revision 1, September 1972
- 4. BC-TOP-5 "Prestressed Concrete Nuclear Reactor Containment Structures," Revision 1, December 1972
- 5. BC-TOP-7 "Full Scale Buttress Test for Prestressed Nuclear Containment Structures," Revision 0, September 1972
- 6. BC-TOP-8 "Tendon End Anchor Reinforcement Test," Revision 0, September 1972
- 7. BC-TOP-1 "Containment Building, Liner Plate Design Report," Revision 1, December 1972 8. BN-TOP-1 "Testing Criteria for Integrated Leak Rate Testing of Primary Containment Structures for Nuclear Power Plants," Revision 1, November 1972.
- 9. BC-TOP-9A "Design of Structures for Missile Protection," Revision 2, September 1974
1.6.3 GENERAL
REPORTS Material used for initial design and background information and not required for evaluation of the Farley Nuclear Plant are as follows:
REPORT TITLE SECTION REFERENCE NRC SUBMITTAL
- 1. J.J. Szyslowski and R. Salvatori, "Determination of Design Pipe Breaks for the Westinghouse Reactor Coolant System," WCAP-7503 , Revision 1, February 1972.
3.6, 5.2 February 1972
- 2. J.S. Moore, "Westinghouse PWR Core Behavior Following Loss-of-Coolant Accident,"
WCAP-7422-L , January 1970, (W Proprietary), and WCAP-7422 August 1971.
3.9 January
1970 3. S. Fabic, "Loss-of-Coolant Analysis: Comparison Between BLODWN-2 Code Results and Test
Data," WCAP-7401 , November 1969.
3.9 February
1970 4. V.J. Placido, R.E. Schreiber, J. Skaritka, "Operational Experience - Westinghouse
Cores," WCAP-8183, October 1973.
4.2 July 1974 FNP-FSAR-1 1.6-11 REV 22 8/09 REPORT TITLE SECTION REFERENCE NRC SUBMITTAL
- 5. J.E. Outzs, "Plant Startup Test Report, H.B.
Robinson Unit No.2," WCAP-7844, January 1972.
4.3 January
1972 6. J.S. Moore, "Evaluation of Nuclear Hot Channel Factor Uncertainties," WCAP-7810 , December 1971. 4.3, 15.4 December 1971
- 7. J.S. Moore, "Power Distribution Control of Westinghouse PWRs," WCAP-7208 , September 1968 (W Proprietary), and WCAP-7811 , December 1971. 4.3 October 1968 8. G.C. Poncelet, "LASER-A Depletion Program for Lattice Calculations Based on MUFT and THERMUS," WCAP-2048 , July 1962.
4.3 July 1972 9. J.E. Olhoeft, "The Doppler Effect for a Non-Uniform Temperature Distribution in Reactor Fuel
Elements," WCAP-2048 , July 1962.
4.3 July 1962 10. R.J. Nodvik, et al., "Supplementary Report on Evaluation of Mass Spectrometric and Radiochemical Analyses of Yankee Core I Spent Fuel, Including Isotopes of Elements Thorium through Curium," WCAP-6086 , August 1969.
4.3 August
1969 11. F.B. Skogen and A.F. McFarlane, "Xenon-Induced Spatial Instabilities in Three Dimensions," WCAP-3680-22 (EURAEC-1976), March 1968.
4.3 September
1969 12. R.F. Barry, "LEOPARD, a Spectrum-Dependent, Non-Spatial Depletion Code for the IBM-7094,"
WCAP-3269-26 , September 1963. 4.3, 15.1, 15.3 September 1963
- 13. A.F. McFarlane, "Core Power Capability in Westinghouse PWRs WCAP-7267-L , October 1964 (W Proprietary), and WCAP-7809 , December 1971. 4.3 October 1969 14. A.F. McFarlane, "Power Peaking Factors" WCAP-7912-L , March 1972 (W Proprietary) and WCAP-7912 , March 1972.
4.3, 4.4 March 1973
FNP-FSAR-1 1.6-12 REV 22 8/09 REPORT TITLE SECTION REFERENCE NRC SUBMITTAL
- 15. J.O. Cermak, "Pressurized Water Reactor pH-Reactivity Effect," Final Report, WCAP-3696-8 (EURAEC-2074), October 1968.
4.3 October
1968 16. G.C. Poncelet and A.M. Christie, "Xenon-Induced Spatial Instabilities in Large PWRs,"
WCAP-3680-20 , (EURAEC-1974), March 1968.
4.3 March
1968 17. J.C. Lee, "Axial Xenon Transient Tests at the Rochester Gas and Electric Reactor," WCAP-7964 , June 1971.
4.3 June 1971 18. C.J. Kubit, "Safety Related Research and Development for Westinghouse PWRs, Program Summaries, Fall 1972," WCAP-8004 , December 1972. 4.3, 4.2 January 1973
- 19. J.A. Christensen, R.J. Allio, and A. Biancheria, "Melting Point of Irradiated UO 2," WCAP-6065. 4.2, 4.4 February 1965
- 20. C.G. Poncelet, "Burnup Physics of Heterogeneous Reactor Lattices," WCAP-6069 , June 1965.
4.4 June 1965 21. J. Shefcheck, "Application of the THINC Program to PWR Design," WCAP-7359-L, August 1969 (W Proprietary), and WCAP-7838, January 1972.
4.4 September
1969 22. E.H. Novendstern and R.O. Sandberg, "Single Phase Local Boiling and Bulk Boiling Pressure
Drop Correlations," WCAP-2850 , April 1966 (W Proprietary), and WCAP-7916, June 1972.
4.4 April
1966 23. G. Hetsroni, "Hydraulic Tests of the San Onofre Reactor Model," WCAP-3269-8, June 1964.
4.4 June 1964 24. R.L. Rosenthal, "An Experimental Investigation of the Effect of Open Channel Flow on Thermal-Hydrodynamic Flow Instability," WCAP-7966 , December 1972.
4.4 December
1972 25. M.G. Balfour, J.A. Christensen, and H.M. Farrari, "In-Pile Measurement of UO 2 , Thermal Conductivity," WCAP-2923 , March 1966.
4.4 March
1966 FNP-FSAR-1 1.6-13 REV 22 8/09 REPORT TITLE SECTION REFERENCE NRC SUBMITTAL
- 26. J.A. Nay, "Process Instrumentation for Westinghouse Nuclear Steam Supply Systems,"
WCAP-7671, April 1971, and WCAP-7547-L , March 1971 (W Proprietary).
5.2 May 1971 27. W.O. Shabbits, "Dynamics Fracture Toughness Properties of Heavy Section A533 Grade B, Class 1, Steel Plate," WCAP-7623 , December 1970.
4.2 January
1971 28. W.S. Hazelton, "Sensitized Stainless Steel in Westinghouse Heavy Section A533 Grade B Class 1 Steel Plate," WCAP-7623 , December 1970.
5.2 December
1970 29. L.F. Picone, "Evaluation of Protective Coatings for Use in Reactor Containment," WCAP-7198 , April 1968 (W Proprietary). 6A.8, 6A.9 April 1969
- 30. T.W.T. Burnett, "Reactor Protection System Diversity in Westinghouse Pressurized Water Reactors," WCAP-7306, April 1969. 7.1, 7.2, 15.4 April 1969
- 31. J.B. Lipchak and R.A. Stokes, "Nuclear Instrumentation System," WCAP-7669, April 1971. (Replaced by WCAP-8255
). 7.1, 7.2, 7.7 May 1971
- 32. J.J. Loving, "In-Core Instrumentation (Flux-Mapping System and Thermocouples)," WCAP 7607 , July 1971.
7.7 July 1971 33. A.E. Blanchard and Calpin, J.E. "Digital Rod Position Indication," WCAP-8014 , December 1974.
7.7 March
1974 34. W.C. Gangloff, "An Evaluation of Anticipated Operational Transients in Westinghouse Pressurized Water Reactors," WCAP-7486 , May 1971, and WCAP-7486-L. 15.2 May 1971 35. D.D. Malinowski, et al., "Radiological Consequences of a Fuel Handling Accident,"
WCAP-7878, December 1971, and WCAP-7518-L. (W Proprietary).
15.4 December 1971 FNP-FSAR-1 1.6-14 REV 22 8/09 REPORT TITLE SECTION REFERENCE NRC SUBMITTAL
- 36. F.F. Cadek, et al., PWR FLECT (Full Length Emergency Cooling Heat Transfer), Final Report,"
WCAP-7665, April 1971. 15.4, 15A July 1971
- 37. R.J. French, "Indian Point Unit No. 2 Rod Ejection Analysis," WCAP-2940 May 1966.
15.4 June 1966 38. J.S. Moore, "Evaluation of Nuclear Hot Channel Factor Uncertainties," WCAP-7810 , December 1971. 15.4 December 1971 39. "Safety Related Research and Development for Westinghouse Pressurized Water Reactors," Program Summaries, Spring 1974, WCAP-8353 , September 1974.
1.5 September
1974 40. "Anticipated Transients Without Reactor Trip in Westinghouse Pressurized Water Reactors," WCAP-8096, April 1973. Appendix 3A April 1973
- 41. F.M. Bordelon, "Small Loss of Coolant Accident Analysis for PWR Systems (SLAP Digital
Computer Code)," WCAP-7983 , November 1971.
15.3 September 1974 1.6.4 CROSS REFERENCE OF ENGINEERING DRAWINGS This table lists FSAR project drawing numbers and their affiliated titles that were associated with FSAR figure numbers which were removed in Revision 13. The project drawings contained the same information, but typically provided much more detail than the original FSAR figures. Any project drawings added to the FSAR will be added to this table with their appropriate drawing title.
FNP-FSAR-1 TABLE 1.6-1 (SHEET 28 OF 29)
Drawing Number Drawing Title REV 25 4/14 U-166233 Nuclear Instrumentation and Manual Trip Signals U-166234 Nuclear Instrumentation Permissives and Blocks U-166235 Primary Coolant Systems Trip Signals U-166236 Pressurizer Trip Signals U-166237 Steam Generator Trip Signals U-166238 Safeguards Actuation Signals U-166239 Rod Controls and Rod Blocks U-166240 Steam Dump Control
U-166241 Pressurizer Pressure and Level Control U-166242 Pressurizer Heater Control U-166243 Feedwater Control and Isolation U-166244 Auxiliary Feedwater Pumps Startup U-166245 Turbine Trips, Runbacks, and Other Signals
U-167647 Radiation Monitoring System Functional Block Diagram U-167649 Radiation Monitoring System Area Range R-21 and R-22 U-167650 ALA/APR Radiation Monitoring System Functional Block Diagram U-167651 Radiation Monitoring System Area Range R-17A, R-17B, and R-18 U-167652 ALA/APR Radiation Monitoring System Functional Block Diagram
U-170148 Solid State Protection System Interconnection
Diagram FNP-FSAR-1 TABLE 1.6-1 (SHEET 29 OF 29)
Drawing Number Drawing Title REV 25 4/14 U-202225 RCS Equipment Support S.G. Inlet Restraint
U-261455 RCS Equipment Support S.G. Inlet Restraint U-264612 Flow Diagram of the Reactor Vessel Head Vent System U-419610 Cable Drive Installation Assembly Arrangement
U-419916 Ex-Vessel Neutron Dosimetry Housing, Clamp &
Chain Stop
U-419917 Ex-Vessel Neutron Dosimetry Support Bar
U-419918 Ex-Vessel Neutron Dosimetry Chain Assembly U-419920 Ex-Vessel Neutron Dosimetry Installation - Unit 1
U-419289 Replacement Reactor Vessel Closure Head Outline Drawing - Unit 1 U-611138 Replacement Reactor Vessel Closure Head Outline Drawing - Unit 2
U-611432 Ex-Vessel Neutron Dosimetry Support Frame Installation - Unit 2 U-611433 Ex-Vessel Neutron Dosimetry Support Frame Assembly - Unit 2
FNP-FSAR-1 1.7-1 REV 21 5/08
1.7 GLOSSARY
OF TERMS This section presents abbreviations, symbols, indices, legends, and other aids to facilitate
review of this Final Safety Analysis Report. Information was compiled from applicable
specifications, drawings, FSAR sections, and related publications developed throughout the
design, construction, and documentation of the Farley Nuclear Plant.
1.7.1 ABBREVIATIONS
The technical abbreviations in table 1.7-1 are used where appropriate throughout the FSAR.
1.7.2 DRAWING
INDEX AND SYMBOLS
The piping and instrumentation diagrams (P&IDs) are listed in table 3.2-3. Drawings D-175016, sheet 1, D-175016, sheet 2, D-175016, sheet 3, figures 1.7-1 and 1.7-2 are provided to facilitate
the understanding of the figures and referenced drawings throughout the FSAR.
1.7.3 HISTORICAL
DESCRIPTIONS
Historical descriptions in the FSAR fall into one of the following categories:
- 1. Initial condition data, that may have been provided as part of original licensing activities, but that is not required for continued plant operation.
- 2. Initial test or analyses provided to document equipment acceptability.
- 3. References to initial construction related procedures or activities.
Material that falls into one of the above categories is not necessary to support the current
operations of the plant but is being retained in order to provide additional information concerning
the licensing history of FNP. The designation of material as historical is not to be used to justify
the removal and/or abandonment of any structure, system or component.
Historical sentences, paragraphs, sections and/or tables have been annotated with the word
"HISTORICAL" set off by brackets and marked with dollar signs ($). The annotation is placed at
the beginning of the historical material and the material itself is also clearly marked to insure
clarity. The following is an illustration of the annotation method; the historical material is
represented by the words "historical material" in the section set of brackets:
[HISTORICAL] [If historical material continues from one FSAR page to subsequent pages, each subsequent page is annotated exactly like the first page.]
FNP-FSAR-1 REV 21 5/08 TABLE 1.7-1 (SHEET 1 OF 15)
TECHNICAL ABBREVIATIONS Word Abbreviation absolute abs
absolute ampere abamp actual cubic feet per minute aft 3/min alternating current ac altitude alt
ampere(s) A
ampere-hour(s) Ah
ampere per square centimeter A/cm 2 anno domini A.D.
angstroms Å
ante meridian a.m.
antilogarithm log
-1 , antilog
approximately or approx
asymmetrical asym
atmosphere (standard) atm atomic mass unit (unified) u atomic number at no.
atomic percent at %
atomic weight at wt atomic weight unit awu audio-frequency af
average avg
FNP-FSAR-1 REV 21 5/08 TABLE 1.7-1 (SHEET 2 OF 15)
Word Abbreviation bar(s) bar
barn(s) b
barrel(s) bbl
Baume Be
billion electronvolts GeV biot(s) Bi
body centered cubic bcc boiling point bp brake horsepower bhp Brinell hardness number Bhn British thermal unit Btu British thermal unit per hour Btu/h British thermal unit Btu/h-ft-°F per hour per degree
Fahrenheit per foot (thermal conductivity)
calculated calc
calorie(s) cal
candela(s) cd
candlepower cp
Celsius (centigrade)
°C cent(s) ¢
center line cl centigram cg
centimeter(s) cm
FNP-FSAR-1 REV 21 5/08 TABLE 1.7-1 (SHEET 3 OF 15)
Word Abbreviation centimeter-gram-second cgs
centimeters per second cm/s centipoise cP
chemically pure cp coefficient coef
cologarithm colog
concentrated conc
constant const
cosecant csc
cosine cos
cotangent cot
coulomb(s) C
counts per minute cpm cubic cu
cubic centimeter(s) cc or cm 3 cubic feet per minute ft 3/min cubic feet per second ft 3/s cubic foot (feet) ft 3 cubic inch(es) in.3 cubic meter m 3 cubic micron(s) cu µ or µ3 cubic millimeter(s) cu mm, mm 3 cubic yard yd 3 FNP-FSAR-1 REV 21 5/08 TABLE 1.7-1 (SHEET 4 OF 15)
Word Abbreviation curies Ci
curies per minute Ci/min curies per second Ci/s cycles per second (hertz electronics)
Hz cylinder cyl
day day
debye(s) D
decibel(s) dB
degree(s) deg
degree Baume
°B degree Celsius (centigrade)
°C degree Fahrenheit
°F degree Kelvin (absolute)
K decimeter(s) dm
diameter diam, dia.
diamond pyramid hardness DPH direct current dc disintegration(s) dis
disintegrations per minute dpm disintegrations per second dps dollar(s) $
dyne(s) dyn
east E
FNP-FSAR-1 REV 21 5/08 TABLE 1.7-1 (SHEET 5 OF 15)
Word Abbreviation electromagnetic force emf electromagnetic unit emu electron volt(s) eV electrostatic units esu entropy units eu equation(s) Eq, Eqs equivalent equiv
erg(s) erg
exponential exp
exponential integral Ei Fahrenheit °F
farad(s) F
feet (foot) ft feet per minute ft/min feet per second ft/s fermi (=10
-13cm) F figure fig.
footcandle fc
footlambert fL
foot-pound ft-lb
franklin(s) Fr
frequency modulation fm gallon(s) gal
FNP-FSAR-1 REV 21 5/08 TABLE 1.7-1 (SHEET 6 OF 15)
Word Abbreviation gallons per minute gal/min gallons per second gal/s gallons per hour gal/h gauss G
gilbert(s) Gb
gram(s) g
gram-calorie g-cal
gram-molecular volume gmv grams per liter g/liter henry(-ies) H
hertz (cycle per second)
Hz high frequency hf high voltage hv horsepower hp
hour(s) h
hydrogen ion concentration, negative
logarithm of pH hyperbolic cosecant csch hyperbolic cosine cosh hyperbolic cotangent coth hyperbolic sine sinh inch(es) in.
inches of mercury in. Hg inches of water in. H 2 O FNP-FSAR-1 REV 21 5/08 TABLE 1.7-1 (SHEET 7 OF 15)
Word Abbreviation inch-pound in.-lb
inside diameter id integrated neutron flux nvt intermediate frequency if intramuscular(ly) im
intraperitoneal(y) ip
intraveneous(ly) iv
international angstrom IA joule(s) J
kelvin K
kilocalorie(s) kcal
kilocurie kCi
kilocycles per second kHz kiloelectron volt(s) keV kilogauss kG
kilogram(s) kg
kilogram meter kg-M kilogram-weight kg-wt
kilohm(s) K kilojoule(s) kJ
kiloliter(s) kliter
kilometer(s) km
kilo-oersted kOe
FNP-FSAR-1 REV 21 5/08 TABLE 1.7-1 (SHEET 8 OF 15)
Word Abbreviation kilovolt(s) kV
kilovolt-ampere(s) kVA
kilowatt(s) kW
kilowatt-hour(s) kWh
kinetic energy KE or T Knopp Hardness Number (microhardness)
KHN laboratory lab
lambert L
limit lim
liter(s) liters
logarithm (common) log logarithm (natural) ln lumen lm
lumens per watt lm/W lux lx
magnetomotive force mmf magnified 50 times 50X maximum max
maxwell(s) Mx
megacycles(s) Mc
megacycles per second Mc/s megacycles per second (electronics)
MHz megacycles per second (mechanics)
Mc/s FNP-FSAR-1 REV 21 5/08 TABLE 1.7-1 (SHEET 9 OF 15)
Word Abbreviation megavolts MV
megawatts MW
megawatt-day(s) MWday
megawatt-electric MWe
megawatt-hour(s) MWh
megawatt-second(s) MWs
megawatt-year(s) MWyear
megawatt-thermal MWt
megohm(s) M melting point mp meter(s) m
meter-kilogram second mks metric ton MT(tonne)
mho mho
microampere(s)
µA microangstrom
µÅ microbar µbar microbarn(s)
µb microcoulomb(s)
µC microcuries
µCi microgram
µg microfarad
µF microhenry
µH FNP-FSAR-1 REV 21 5/08 TABLE 1.7-1 (SHEET 10 OF 15)
Word Abbreviation microinch
µin.
micromicrofarad
µµF micromicrons
µµ micromole
µM micron(s)
µ microsecond
µs microvolt(s)
µV microwatt(s)
µW mile mi
miles per hour mph milliampere(s) mA
millicurie(s) mCi
milligauss mG
milligram mg
milligrams per decimeter per day mdd millihenry mH
milliliter(s) mliter
milli-mass-unit mmu
millimeter mm
millimicron(s) m
µ millimicrosecond(s) [nanosecond(s)
perferred] m
µs millimole(s) mM
million electron volts MeV FNP-FSAR-1 REV 21 5/08 TABLE 1.7-1 (SHEET 11 OF 15)
Word Abbreviation million volts MV milliroentgen per hour mR/h millisecond(s) ms
millivolt(s) MV
minimum min.
minute(s) min
molal molal
molar M mole mole
mole percent mole percent
molecular weight mol wt month mo
nanocuries nCi
nanosecond(s) ns
neper(s) Np
neutron flux nv neutrons per volume time nvt neutrons per square centimeter per second n/cm 2-s newton(s) N
north N
normal N nuclear magnetron nm number No.
FNP-FSAR-1 REV 21 5/08 TABLE 1.7-1 (SHEET 12 OF 15)
Word Abbreviation oersted Oe
ohm(s) ounce(s) oz
outside diameter od page p
pages pp
parts per billion ppb parts per million ppm percent % (graphics)
percent milli-k pcm picofarad(s) pF
poise P
post meridian p.m.
potential difference PD potential energy PE or V pound lb
pounds per cubic foot lb/ft 3 pounds per square foot lb/ft 2 pounds per square inch psi pounds per square inch absolute psia pounds per square inch differential psid pounds per square inch gauge psig pressure (millimeter of mercury) mm Hg FNP-FSAR-1 REV 21 5/08 TABLE 1.7-1 (SHEET 13 OF 15)
Word Abbreviation probable error pe radian rad
radioactivity (measure of) rad rad equivalent man rem Radiation Protection Guide RPG Radioactivity Concentration Guide RCG Rankins (degree)
°R revolutions per minute rpm revolutions per second rps roentgen(s) R
root mean square rms secant sec
second(s) s
Section sec.
sine sin
south S
specific gravity sp gr, s.g.
square sq
square centimeters cm 2 square foot ft 2 square inch(es) in.2 square kilometer(s) km 2 square meter(s) m 2 FNP-FSAR-1 REV 21 5/08 TABLE 1.7-1 (SHEET 14 OF 15)
Word Abbreviation square micron(s)
µ2 square millimeter(s) mm 2 stainless steel ss standard Std.
standard temperature and pressure STP steradian sr
tangent tan
temperature temp
tensile yield strength tys tesla (Wb/m
- 2) T thousand circular mills kcmil thousand electron volts keV ton(s) ton
trace Tr
transpose tr
ultimate tensile strength uts ultraviolet uv
velocity v
versus vs
volt(s) v
volume vol
volume parts per million vpm water gauge wg FNP-FSAR-1 REV 21 5/08 TABLE 1.7-1 (SHEET 15 OF 15)
Word Abbreviation watt(s) W
weber Wb
weight wt
weight percent wt/%
west W
x units xu yard(s) yd
year(s) year
REV 21 5/08 FLUID SYSTEM SYMBOLS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 1.7-1 (SHEET 1 OF 3)
REV 21 5/08 FLUID SYSTEM SYMBOLS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 1.7-1 (SHEET 2 OF 3)
REV 21 5/08 FLUID SYSTEM SYMBOLS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 1.7-1 (SHEET 3 OF 3)
REV 21 5/08 PIPING NOMENCLATURE JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 1.7-2 (SHEET 2 OF 2)