ML18220A905
| ML18220A905 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 05/11/2018 |
| From: | Public Service Enterprise Group |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18220A885 | List: |
| References | |
| LR-N18-0053 | |
| Download: ML18220A905 (39) | |
Text
SECTION 1 TABLE OF CONTENTS INTRODUCTION AND GENERAL DESCRIPTION OF PLANT Section Title Page 1.1 PROJECT IDENTIFICATION 1.1-1 1.2 PLANT SITE
SUMMARY
1.2-1
- 1. 2.1 Site Description 1.2-1
- 1. 2.2 Meteorology 1.2-1 1.2.3 Geology and Hydrology 1.2-1 1.2.4 Seismology 1.2-2 1.2.5 Marine Ecology 1.2-2 1.2.6 Environmental Radiation Monitoring 1.2-2 1.2. 7 Facility Safety Conclusions 1.2-2 1.3
SUMMARY
PLANT DESCRIPTION 1.3-1
- 1. 3.1 Structures 1.3-2 1.3.2 Nuclear Steam Supply System 1.3-3
- 1. 3.3 Reactor and Plant Control 1.3-4 1.3.4 Waste Disposal System 1.3-4
- 1. 3.5 Fuel Handling System 1.3-5 1.3.6 Turbine and Auxiliaries 1.3-5
- 1. 3. 7 Electrical System 1.3-6
- 1. 3.8 Engineered Safety Features 1.3-7 1.4 IDENTIFICATION OF CONTRACTORS 1.4-1 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5-1 1.5.1 17 x 17 Fuel Assembly 1.5-1 1.5.1.1 Rod Cluster Control Spider Tests 1.5-1 1.5.1.2 Grid Tests 1.5-1 1.5.1.3 Fuel Assembly Structural Tests 1.5-1 1-i SGS-UFSAR Revision 6 February 15, 1987
Section 1.5.1.4 1.5.1.5 1.5.1.6 1.5.1.7 1.5.2 1.5.2.1 1.5.2.2 1.5.3 1.6 SGS-UFSAR TABLE OF CONTENTS (Cont)
Title Guide Tube Tests Prototype Assembly Tests Departure from Nucleate Boiling Tests Incore Flow Mixing Other Programs Generic Programs of Westinghouse LOCA Heat Transfer Tests References for Section 1.5 LIST OF ACRONYMS 1-ii Page 1.5-2 1.5-2 1.5-2 1.5-2 1.5-2 1.5-2 1.5-3 1.5-3 1.6-1 Revision 6 February 15, 1987
Figure 1.2-1 General Site Plan SGS-UFSAR LIST OF FIGURES 1-iii Revision 7 July 22, 1987
SECTION 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 1.1 PROJECT IDENTIFICATION This Updated Final Safety Analysis Report is submitted pursuant to the requirements of 10CFR50.71 by Public Service Electric and Gas Company (PSE&G) for the two nuclear power units at its Salem Generating Station.
PSE&G and Westinghouse Electric Corporation have jointly participated in the design and construction of each unit.
On August 21,
Each unit employs a pressurized water reactor nuclear steam supply system furnished by Westinghouse which is similar in design concept to several other projects licensed by the Nuclear Regulatory Commission.
The only systems shared by the two units are Compressed Air, the Control Room Area intake air radiation monitors, parts of the Control Room Area Ventilation System, bulk Nitrogen Supply, Demineralized Water, and the Solid Radwaste Handling System.
There are a minimum of shared components; chemical drain and laundry hot shower tanks and pumps and the 20,000 barrel Bulk Fuel Oil Storage Tank are the only components in common.
The licensed core power for both units is 3459 MWt.
The approximate values for gross and net electrical outputs are 1178 MWe and 1135 MWe respectively for Unit 1 and 1182 MWe and 1139 MWe respectively for Unit 2.
The reactors are expected to be capable of outputs of approximately 3570 MWt, which corresponds to the valves-wide-open rating of the turbine generators of 1209 MWe gross and 1163 MWe net for Unit 1 and 1220 MWe gross and 1174 MWe net for Unit 2.
The containment and engineered safety features for both units have been designed and evaluated at the maximum power rating of 3570 MWt.
Loss-of-coolant accidents and those postulated accidents havin~ offsite dose consequences have been analyzed at the power rating of 3570 MWt.
1.1-1 SGS-UFSAR Revision 19 November 19, 2001 I
I
The remainder of Section 1 of this summarizes the features and criteria of the nuclear
- units, design out the similarities and differences with nY'C.CH:,,-ri zed water nuclear to other and basic power the same features as the Salem Station.
Sec::icn 2 contains a description and evaluation of the site and environs, supporting the suitability of the site for a nuclear plant of the size and type described.
Section 3 discusses the identification, description, and discussion of t:Ie archjt.ec::Jral and engineering design of structures, and to safety.
The reactor is described in Section 4.
Section 5 discusses the Reactor Coolant and related systems, and Sections 6 1 the emergency and systems.
Section 13 describes the Company's program for organization and training of plan:: personnel.
Section 14 contains an outline and description of the initial tests and operations associated with plant startup.
Section 15 is a safety evaluation summarizing the analyses which demonstrate the features.
SGS-IJFSAR of the Reactor Protection The consequences of various set forth in the N~clear and 10CFR50.67, Accident Source Term.
1.1-2 and the accidents are within the Commission's Rules and Revision 25 October 26, 2010
1.2 PLANT SITE
SUMMARY
1.2.1 Site Description The approximately 700 acre Salem site is located along the eastern shore of the Delaware River in Lower Alloways Creek Township, Salem County, New Jersey about 8 miles southwest of Salem, New Jersey.
The population density of the area surrounding the site is low.
Distance to the site boundary is about 4200 feet.
The nearest residence is approximately 3.4 miles west of the site in Bay View Beach, Delaware.
Other nearby residences are located 3.5 miles east-northeast and 3. 5 miles northwest of the site.
The population center distance is 15.5 miles.
The area is primarily utilized for agricultural pursuits, with heavy industry located generally 15 miles and beyond to the north of the site.
1.2.2 Meteorology The meteorological data pertinent to the Salem site has been
- reviewed, and there is no reason to anticipate unusual meteorological problems. The terrain is open and extremely flat, and the land-sea interaction favors a vigorous wind flow.
A meteorological tower facility was established northwest of the reactor area on the site to provide actual site meteorological data.
This data collection program has been terminated as sufficient data has been collected and analyzed to describe the dispersion parameters.
The tower has been relocated east of the site.
1.2.3 Geology and Hydrology An investigation of Salem site geology and hydrology was completed in 1967.
The nearest known faulting is approximately 25 miles from the site.
Test borings at the site indicate that subsurface conditions are adequate to support the structures.
The regional direction of ground water movement is toward the Delaware River, 1.2-1 SGS-UFSAR Revision 6 February 15, 1987
and all surface drainage at the Salem site flows directly into the river.
1.2.4 Seismology The site is located in a region which has experienced only infrequent minor earthquake activity.
No known faults exist in the basement rock or sedimentary deposits in the immediate vicinity of the site.
Significant earthquake motion is not expected at the site during the life of the facility.
The plant was conservatively designed to respond elastically, with no loss of function, to horizontal ground accelerations as high as 10 percent of gravity, and the design was checked for a hypothetical acceleration of 20 percent of gravity.
1.2.5 Marine Ecology A thorough study of the biological makeup of the Delaware Estuary is being conducted.
The study has continued since the plant went into operation to determine the effects (if any) of plant operation on the ecology of the Estuary.
1.2.6 Environmental Radiation Monitoring An environmental radiation monitoring program for the site and surrounding area is being conducted.
This program has continued since the plant went into operation to determine the effects (if any) of plant operation on radiation levels in the environment.
1.2.7 Facility Safety Conclusions The safety of the public and plant operating personnel and reliability of plant equipment and systems have been the primary considerations in the plant design.
The approach taken in fulfilling the safety consideration is three-fold. First, careful attention has been given to design so as to prevent the release of 1.2-2 SGS-UFSAR Revision 6 February 15, 1987
radioactivity to the environment under conditions which could be hazardous to the health and safety of the public.
Second, the plant has been designed so as to provide adequate protection for plant personnel wherever a potential radiation hazard exists.
- Third, Engineered Safety Features have been designed with redundancy and diversity, and to stringent quality standards.
Based on the over-all design of the plant including its safety features and the analyses of possible incidents including the design basis accident, it is concluded that the Salem Generating Station can be operated without undue risk to the health and safety of the public.
1.2-3 SGS-UFSAR Revision 6 February 15, 1987
I l
I ! I ll*
_JJ 15 ~
0 O[JCD Hope Creek Generating StatiOn
.c I
c LEGEND
- 1. Unit 1 Colainment
- 2. Unit 2 Contlilunent
- 3. Auxiliary laildlng
- 4. Unit 1 Full HMding
- 5. Unit 2 Ft.w1 H*dltO
- e. Service ***no
- 7. Turbine Gellll1ltOr Aree
'!.1 *.:., * !
I::J
'T
- l 11 II c:
'. GJ L
J o:-:c PUBLIC SERVICE ELECJRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Updated FSAR
- a. Admlnilnnrlian luldlng
- 9. Circulllllnl Waw ln'allul
- 10. s.mc. W.ur Intake
- 11. 500 ltV Swkctryard
. 12. Fuel Oil Ston~gt Tn
- 13. Clean fKilitiiM Buiklng
- 14. Conttolad Facl'lfiesluilding
- 15. Unit 3 Gu Tt.Wbine
- 16. SERVICE WATER ACCUMULATOR ENCLOSURE
~
- 1
~
~
~
13 12 pt£f..
~~
fJf-lfi REVISION 17 OCTOBER 16 1998 General Site Plan FIG. 1.2-1
SGS-UFSAR Figure Fl.2-2 intentionally deleted.
Refer to plant drawing 204803 in DCRMS Revision 27 November 25, 2013
1.3
SUMMARY
PLANT DESCRIPTION The inherent design of the pressurized water, closed-cycle reactor minimizes the quantities of fission products released to the atmosphere.
Four barriers exist between the fission product accumulation and the environment.
These are the uranium dioxide fuel matrix, the fuel cladding, the reactor vessel and coolant loops, and the reactor containment.
The consequences of a breach of the fuel cladding are greatly reduced by the ability of the uranium dioxide lattice to retain fission products.
Escape of fission products through a fuel cladding defect would be contained within the pressure vessel, loops, and auxiliary systems.
Breach of these systems or equipment would release the fission products to the reactor containment where they would be retained.
The reactor containment is designed to adequately retain these fission products under the most severe accident conditions, as analyzed in Section 15.
Several engineered safety features have been incorporated into the plant design to reduce the consequences of a loss-of-coolant accident (LOCA).
These safety features include an Emergency Core Cooling System (ECCS).
This system automatically delivers borated water to the reactor vessel for cooling the core under high and low reactor coolant pressure conditions.
The ECCS also serves to insert negative reactivity into the core in the form of borated water during plant cooldown following a steam line break or an accidental steam release.
Other safety features which have been included in the reactor containment design are a Containment Fan Cooler System which acts to effect depressurization of the containment following a LOCA and to remove particulate matter from the containment atmosphere, and a Containment Spray System which acts to depressurize the containment and remove elemental iodine from the atmosphere by washing action.
The Containment Spray System provides redundant backup by an alternate principle for the Containment Fan Cooler System for heat removal.
1.3-1 SGS-UFSAR Revision 6 February 15, 1987
1.3.1 Structures The major structures include a separate and independent Containment and Fuel Handling Building for each reactor, a
common Auxiliary Building with holdup tank vault, a common Turbine Building and a common Administration and Service Building.
General layouts of the Reactor Containment, Auxiliary Building, and interior component arrangements are shown on Figures 1. 2-1, 5.1-12 and Plant Drawings 204803, 204804, 204805, 204806, 204807 and 204808.
Seismic Criteria For Category I (seismic) equipment, dynamic methods or conservative static equivalents were used to determine that components and structures will operate or maintain their integrity, as required.
For Category II (seismic) equipment, static methods were used and non-seismic equipment meets applicable codes.
Definition of Seismic Categories Particular structures and equipment are classified according to seismic design.
The seismic definitions are:
- 1.
Category I (seismic)
SGS-UFSAR Those structures, mechanical components, the Reactor Protection
- System, and Engineered Safety Features Actuation System whose failure might cause or increase the severity of a LOCA.
Also, those structures and components vital to safe shutdown and isolation.
- 1. 3-2 Revision 27 November 25, 2013
- 2.
Category II (seismic)
Those structures and mechanical components that are not Category I (seismic), but which function in direct support of reactor operation.
1.3.2 Nuclear Steam Supply System The Nuclear Steam Supply System for each unit consists of a pressurized water reactor, Reactor Coolant System (RCS), and associated auxiliary fluid systems.
The RCS is arranged as four closed reactor coolant loops connected in parallel to the reactor vessel, each containing a reactor coolant pump and a steam generator.
An electrically heated pressurizer is connected to one of the loops.
The reactor core is composed of uranium dioxide pellets enclosed in Zircaloy-based tubing with welded end plugs.
The tubes are supported in assemblies by spring clip grid structures.
The control rods consist of clusters of stainless steel clad silver-indium-cadmium absorber rods located within the fuel assemblies.
The nuclear fuel is typically loaded in three regions, with the new fuel being introduced into the core interior and by its third cycle of operation being discharged from the core's outermost region to spent fuel storage.
The reactor vessel and reactor internals contain and support the fuel and control rods.
The reactor vessel is cylindrical with hemispherical heads and is clad with stainless steel.
The pressurizer is a cylindrical pressure vessel with hemispherical heads and is equipped with electrical heaters and spray nozzles for system pressure control.
1.3-3 SGS-UFSAR Revision 19 November 19, 2001
The st~am generators are vertical U-tube units utilizing Inconel tubes.
Integral moisture separating equipment reduces the moisture content of the steam at the turbine throttle to S 0.25 percent for Unit 1 and S 0.1 percent for Unit 2.
The reactor coolant pumps are vertical with controlled leakage shaft seals.
centri pumps equipped Auxiliary systems are provided to charge the RCS, add makeup water, purify reactor.coolant water, provide chemicals for corrosion inhibition and reactor
- control, cool system components, remove residual heat when the reactor is
- shutdown, cool the spent fuel storage pool, sample reactor coolant water, provide for emergency safety ection, and vent and drain the RCS.
- 1. 3. 3 Reactor and Plant Control The reactor is controlled by a coordinated combination of soluble neutron absorbers and mechanical control rods.
The control system allows the plant to accept step load changes of 10 percent and ramp load changes of 5 percent per minute over the load range of 15 to 95 percent power under normal operating conditions.
of each reactor and turbine is from each unit's control room.
- 1. 3. 4 Waste Disposal System The Waste Disposal Systems provide all the equipment necessary to collect, process, and prepare for disposal, all radioactive liquid, gaseous, and solid wastes produced as a result of reactor operation.
After liquid wastes are evaporated and/or demineralized if necessary to reduce activity levels.
The treated water from the demineralizers or the evaporator distillate may be recycled for use in the plant or may be discharged via the
- 1. 3-4 SGS-UFSAR Revision 24 May 11, 2009
condenser discharge ar concentrations well within the limits set forth in 10CFR20.
The evaporator concentrates and spent demineralizer resins are and from the site for ultimate in an authorized location.
Gaseous wastes are collected and held up for radioactive after which they nay be reused for blanketing tanks.
Decayed gases are discharged to the environment in a controlled manner which maintains the offsite dose well below the limits set forth in 10CFR20.
1.3.5 Fuel Handling E:ach reactor is refueled with to handle spent fuel llnder water from the time it leaves the reactor vessel until it is in a cask for dry storage at the Fuel Installation ( ISFSI) or for shipment from the site.
Underwater transfer of spent fuel provides an optically transparent radiation shield as well as a reliable source of coolant for removal of decay heat.
This system also provides capability for receiving, handling, and storing new fuel. The Spent Fuel Pool Cooling System has been redesigned to include a second 1.3.6 Turbine and Auxiliaries The turbine is a four with 44-inch last stage blades.
fuel pool pump.
six flow exhaust, 1800 rpm unit There are six combination moisL1re separator-steam reheater assemblies.
The turbine generators are rated as described in Section 1, with saturated inlet steam conditions of 765 psia, exhausting at 1.5 inches of mercury absolute, at zero percent makeup.
feedwater heating.
There are six stages of The turbine with an Control which uses an electronic controller and a
control valve movement.
- 1. 3-5 fire resistant fluid Revision 25 October 26, 2010 to
The condenser is of the single pass deaerating type. There are three strings of feedwater heaters, three one-third size condensate and heater drain pumps and two one-half size feedwater pumps. Drains from the two highest feedwater heaters are pumped into the Condensate System and drains from the four lowest feedwater heaters are cascaded to the condenser.
1.3.7 Electrical System Each main generator is a 1300 MVA, 25 kV, 3 phase, 60 cycle, 0.9 pf, 1800 rpm, 75 psig hydrogen inner-cooled unit with water cooled stator windings.
Field excitation is provided by a direct shaft driven brushless excitation system.
Each generator is connected to the primary side of three single phase main stepup transformers through isolated phase buses.
The secondary side of each main transformer delivers power to the 500 kV switchyard.
The station service systems consist of a 13.8 kV north ring bus, and 13.8 kV south bus sections, auxiliary and station power transformers, 4160 v, 460 v, 230 V, and 115 V ac and 250 V, 125 V, and 28 V de buses and equipment.
A third 500 kV system tie, the 13.8 kV north ring bus and 13.8 kV south bus sections, arrangement replaces the 69 kV single source described in the Preliminary Safety Analysis Report.
This provides a superior power supply system to the station.
Three diesel-generators per unit are provided as onsite sources of power in the event of complete loss of normal ac power.
These generators power the post-accident containment cooling equipment as well as the safety injection, centrifugal charging, and residual heat removal pumps to assure an acceptable post loss-of-coolant containment pressure transient and adequate core cooling.
Two-out-of-the three diesel-generators can handle the electrical load required for a unit in the event of a LOCA.
1.3-6 SGS-UFSAR Revision 14 December 29, 1995
1.3.8 Engineered Safety Features The LOCA features provided for each unit have sufficient of components and power sources such that under the conditions of a can maintain the of the containment and maintain the exposure operating of the public below the limits with incorporated in summarized below.
partial effectiveness.
the design of each unit set forth in 10CFR50. 67 1 even when The engineered features and the functions they serve are
- 1.
The ECCS injects borated water into the RCS.
This system limits damage to the core and limits the energy and fission products released into the containment a LOCA.
The system has been by The basic changes in the redesigned system are the use of t1..;o charging pumps from the Chemical and Volume Control System for high head injection in addition to their normal charging function and the relocation of the boron injection tank to the discharge side of these pumps.
The design of these pumps was changed from reciprocating to centrifugal.
Piping 1 val ving 1 and instrumentati.on were also revised as a result of the system redesign.
- 2.
A steel-lined concrete containment vessel of reinforced concrete wall, a with testable high integrity and a reinforced concrete base
- 3.
Reactor containment fan coolers and filters to reduce containment press;Jre and filter particulate matter following a LOCA.
- 1. 3-7 SGS-UFSAR Revision 25 October 26 1 2010
- 4.
A Containment Spray System to reduce containment pressure and remove iodine from the containment atmosphere.
- 5.
The Containment Isolation System incorporates valves and controls on piping systems penetrating the containment structure. These valves are arranged to provide two barriers between the RCS or containment atmosphere and the environment.
System design is such that failure of one valve to close will not prevent isolation, and no manual operation is required for immediate isolation.
Automatic isolation is initiated by a containment isolation signal, derived for Phase A isolation by the safety injection signal and high-high containment pressure signal for Phase B isolation.
- 6.
Power sources for the engineered safety features for each unit are provided by two 4 kV power circuits fed from the 500 kV system through the south 13 kV substation in the 500 kV switchyard.
The 500 kV switchyard arrangement consists of three 500 kV transmission lines connected to a breaker-and-a-half design with four 500-13 kV transformers.
Two of them are connected to the 500 kV main bus section 1, the other two are connected to Section 2.
- 1. 3-8 SGS-UFSAR Revision 14 December 29, 1995
.~
Two 500-13 kV transformers provide power to the south 13kV bus sections (one transformer per section) while the other two transformers feed the north 13kV ring bus.
Each south 13kV bus section feeds two 13-4 kV transformers, one for each unit, to provide off-site power for the engineered safety features and new Circulating Water Switchgear. The north 13 kV ring bus is normally operated split to allow one 500-13 kV transformer to feed two (one for each unit) 13-4 kV transformers for Group buses.
should one out of two 500-llkV transformers feeding the north 13kV ring bus be out of service, the ring bus will be realigned to provide power to all four 13-4kV transformers for both unit group buses from the remaining transformer.
If one out of two 500-13kV transformers feeding the south 13kV bus is out of service, transformers connected to the ring bus will be realigned in such a way that one transformer replaces the lost one while the other provides power to all four 13-4kV transformers for the group buses.
During this 500-13kV transformer swap over period, the double ended 4kV vital buses receive power from the second off-site power source.
Reliable diesel-generator power is provided for the engineered safeguards loads in the event of failure of station auxiliary power.
In addition, if external auxiliary power to the station is lost concurrent with an accident, power is available for the engineered safeguards from the diesel-generators, which are capable of supplying the engineered safeguards load to assure protection of the health and safety of the public in the event of a LOCA.
- 7.
All components necessary for the proper operation of the engineered safety features are operable from the control room
- 1.3-9 SGS-UFSAR Revision 15 June 12, 1996
1.4 IDENTIFICATION OF CONTRACTORS The Salem Generating Station was designed and constructed by Public Service Electric & Gas (PSE&G).
Westinghouse Electric Corporation designed and furnished the nuclear steam supply equipment and systems including the fuel assemblies.
PSE&G contracted United Engineers and constructors Inc. of Philadelphia, Pennsylvania, to supervise field erection.
PSE&G also engaged several consultants to provide technical assistance in various areas.
consultants are listed below.
These Consultant Southwest Research Institute, San Antonio, Texas
(Now Meteorological Evaluation systems, Inc. )
Amityville, Long Island, New York Dames and Moore, Cranford, New Jersey Pritchard - carpenter, Consultants, Coral Gables, Florida Radiation Management Corporation, Philadelphia, Pennsylvania Ichthyological Associates, Middletown, Delaware SGS-UFSAR 1.4-1 Program Quality Control Reactor Core and Nuclear Fuel Cycle Meteorology Geology, Hydrology, Seismology Hydrology Radiation Monitoring, Emergency Planning Marine Ecology Revision 6 February 15, 1987
I Porter-Gertz, Consultants, Inc.,
(Now Porter Consultants)
Ardmore, Pennsylvania Framatome Technologies, Inc., (FTI) of Lynchburg, Va. and Raytheon Corp.
Radiation Monitoring, Emergency Planning Unit 1 Steam Generator changeout During the operational phase, various consultants and contractors have been employed to support station operation.
These organizations are selected and perform the applicable service in accordance with the Salem QA Manual.
1.4-2 SGS-UFSAR Revision 18 April 26, :woo
1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION One of the design bases for the Salem Generating Station has been to utilize well-developed and proven design concepts, systems, and equipment, in order to minimize the potential for cost and schedule overruns and to enhance the reliability of operation.
As a consequence, there have been few requirements for research and development programs to confirm the adequacy of the design.
Those programs identified for Salem have been satisfactorily completed, as described in Section 1.5.1. Other programs were identified as valuable to define margins of conservatism or possible design improvements.
Relevant programs in this latter category are described in Section 1.5.2 1.5.1 17 x 17 Fuel Assembly A comprehensive test program for the 17 x 17 assembly has been successfully completed by Westinghouse.
Reference 1 contains a summary discussion of the program.
The following sections present specific references documenting individual portions of the program.
1.5.1.1 Rod Cluster Control Spider Tests Rod cluster control spider tests have been completed.
For a further discussion of these tests, refer to Section 4.2.3.4.
1.5.1.2 Grid Tests Verification tests of the structural adequacy of the grid design have been completed.
Refer to Section 4.2.3.4 and Reference 2 for a discussion of these tests.
1.5.1.3 Fuel Assembly Structural Tests Fuel assembly structural tests have been completed.
Refer to References 2 and 3 for a discussion of these tests.
1.5-1 SGS-UFSAR Revision 6 February 15, 1987
1.5.1.4 Guide Tube Tests Verification tests of the structural adequacy of the guide tubes have been completed.
Refer to References 3 and 4 for a discussion of these tests.
1.5.1.5 Prototxpe Assembly Tests Verification tests of the integrated fuel assembly and rod cluster control performance have been completed.
Refer to References 3 and 4 for a discussion of these tests.
1.5.1.6 Departure from Nucleate Boiling Tests The test program for experimentally determining the effect of the fuel assembly geometry on the departure from nucleate boiling (DNB) heat flux has been completed.
Refer to Reference 5 for a discussion of these tests.
1.5.1.7 Incore Flow Mixing The experimental test program to determine the effects of the fuel assembly geometry on mixing has been completed.
Refer to Reference 6 for a discussion of these tests.
1.5.2 Other Programs 1.5.2.1 Generic Programs of Westinghouse Reference 7 summarizes ongoing safety-related research and development programs that are being carried out for, or by, or in conjunction with the Westinghouse Nuclear Energy System Division and that are applicable to Westinghouse pressurized water reactors.
1.5-2 SGS-UFSAR Revision 6 February 15, 1987
1 5.2.2 LOCA Heat Transfer Tests E.xperimental test programs to determine the thermal-hydraulic characteristics of 17 x 17 fuel assemblies and to obtain experirnenta l reflooding transfer data under simulated loss-of-coolant accident (LOCA) condi tions have been completed.
Refer to Re.fer'ence 8 for a di scussion of these test.s.
A single rod burst test program to quantify the maximum assembly flow blockage whi,ch is assumed in the LOCA ana yses has been completed.
Refer to Reference 9 for a discussion of these tests.
The results of these two test programs have been used in the Emergency Core Cooling System analyses i n Chapter 15.
1.5.3 References for Section 1.5
- 1.
Eggleston, F. T.,
11Safety-Related Research and Development for Westinghouse Pressurized Water
- Reactors, Program Summaries - Spring 1976,." June 1976.
2.
Gesinski, L. and Chiang, D., "Safety Analysi s O*f the 17 x 17 Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident,"
WCAP-8236 (Proprietary) and WCAP-8288 (Non.. Proprietary), December 1973.
- 3.
DeMari o, E. E., "Hydraulic Flow Test of the 17 x 17 Fuel
- 4.
Cooper, F. W., Jr.,
u17 x 17 Dr ~ veline Component Tests -
Phase IB, II, III, D-Loop Drop and Deflection,.. WCAP-8446 (Propri etary) and WCAP-8449 (Non-Proprietary), December 1974p
- 5.
Hill, K.
W., et al.,
t'Effects of 17 x 17 Fuel Assembly Geom.etry on DNB,u WCAP-8296-P-A (Proprietary) and WCAP-8297-A (Non-Proprietary), February 1975 SGS-UFSAR 1.5-3 Revision 6 February 15, 1987
- 6.
Cadek, F. F.; Motley, F. E.; and Dominicis, D.P., "Effect of Axial Spacing on Interchannel Thermal Mixing with the R Mixing Vane Grid,"
WCAP-7941-P-A (Proprietary) and WCAP-7959-A (Non-Proprietary), January 1975.
- 7.
Eggleston, F. T., "Safety-Related Research and Development for Westinghouse Pressurized Water
- Reactors, Program Summaries -
Winter 1977 - Summer 1978," WCAP-8768, Revision 2, October 1978.
- 8.
"Westinghouse ECCS Evaluation Hodel - October 1975 Version,"
WCAP-8622 (Proprietary) and WCAP-8623 (Non-Proprietary),
November 1975.
- 9.
Kuchirka, P. J., "17 x 17 Design Fuel Rod Behavior During Simulated Loss-of-Coolant Accident Cqnditions," WCAP-8289 (Proprietary) and WCAP-8290 (Non-Proprietary), November 1974.
1.5-4 SGS-UFSAR Revision 6 February 15, 1987
1.6 LIST OF ACRONYMS The following is an alphabetical listing of the most frequently used acronyms in this report.
ALP ALS AMSAC -
ANS ANSI AO ASTM ATWS BIT BNWL BOL SGS-UFSAR Atomic Energy Commission Auxiliary Feedwater Storage Tank Auxiliary Feedwater Atomic Industrial Forum As Low as is Reasonably Achievable Actuation Logic Processor (AMSAC)
Actuation Logic System (AMSAC)
Actuation Mitigation System Actuation Circuitry American Nuclear Society American National Standards Institute Axial Offset American Society for Testing and Materials Anticipated Transient Without SCRAM Boron Injection Tank Battelle Northwest Laboratory Beginning-of-Life 1.6-1 Revision 10 July 22, 1990
BOP Balance-of-Plant BTP Branch Technical Position
- 1. 6-la SGS-UFSAR Revision 10 July 22, 1990
SGS-UFSAR THIS PAGE INTENTIONALLY BLANK 1.6-lb Revision 10 July 22, 1990
BWR CAACS CAP CASP CCP CERC CFR CIS CPS CRDM CRS CSAS css eves CVTR cws DAS DBA DBE DCRDR SGS-UFSAR Boiling Water Reactor Control Area Air Conditioning System Chemical Analysis Panel Containment Air Sampling Panel Centrifugal Charging Pump coastal Engineering Research Center Code of Federal Regulations Containment Isolation System condensate Polishing System Control Rod Drive Mechanism control Room supervisor Containment Spray Actuation System Containment Spray System Chemical and Volume Control System Carolina-Virginia Tube Reactor Circulating Water System Data Acquisition System Design Basis Accident Design Basis Earthquake Detailed Control Room Design Review 1.6-2 Revision 16 January 31, 1998
OTT DVRPC E&CD EACS ECCS EOL EPD EPRI EPZ ESF FPS SGS-UFSAR Double-Ended Pump Suction Decontamination Factor Departure from Nucleate Boiling Departure from Nucleate Boiling Ratio Department of Transportation differential pressure Dose Reduction Factor Ductility Transition Temperature Delaware Valley Regional Planning COmmission Engineering and construction Department Emergency Air COnditioning system Emergency core Cooling System End-of-Life Electric Production Department Electrical Power Research Institute Emergency Planning Area Engineered Safety Features Fire Protection system 1.6-3 Revision 16 January 31, 1998
FSAR GDC GM GPM GWD GWS HED HEPA HFP HHW HP I&C ICE IEEE I/0 LCR LDP LP LPG SGS-UFSAR Final Safety Analysis Report General Design Criteria Geiger-Mueller Gallons Per Minute Gigawatt Day Gaseous Waste System Human Engineering Deficiency High-Efficiency Particulate Air Hot Full Power High-High Water High Pressure Instrumentation and Control Instrumentation Controls and Electrical Institute of Electrical and Electronics Engineers Input/Output License Change Request Lighting Distribution Panel Low Pressure Liquified Petroleum Gas 1.6-4 Revision 10 July 22, 1990
LPM SGS~UFSAR Loose Parts Monitoring 1.6~4a Revision 10 July 22, 1990
THIS PAGE INTENTIONALLY BLANK
- 1. 6-4b SGS*UFSAR.
Revision 10 July 22, 1990
LNG LOCA LOFT LSP LWS MCD MEL MEL MI MIG MMA MMI MOL MSIV MSL MSR MWD/
MTU NBS NBU NCO SGS-UFSAR Liquified Natural Gas Loss-of-Coolant Accident Loss of Fluid Test Liquid sampling Panel Liquid Waste System Minor Civil Division Master Equipment List (Section 17.2)
Moderate Energy Lines Mechanical and Integrated Manual Inert Gas Manual Metal Arc Modified Mercalli Intensity Middle-of-Life Main Steam Isolation Valve Mean Sea Level Moisture Separator-Reheater Megawatt Days per Metric Ton of Uranium National Bureau of Standards Nuclear Business Unit Nuclear control Operators 1.6-5 Revision 16 January 31, 1998
NOS NRB NRC NSR NSSS NWS OBE OD O&M ORNL OS OSHA OTG PASS SGS-UFSAR Nil Ductility Transition National Electric Manufacturers' Association National Fire Protection Association Nuclear Instrumentation System Nuclear Mutual Limited Net Positive Suction Head Nuclear Oversight Department Nuclear Review Board Nuclear Regulatory Commission Nuclear Safety Review Department Nuclear Steam Supplier System National Weather Service Operating Basis Earthquake Outside Diameter Operations and Maintenance Oak Ridge National Laboratories Operations Superintendent Occupational and Safety Health Act Operational Test Group Post Accident Sampling System 1.6-6 Revision 22 May 5, 2006
PLUS PMH POPS PORC PORV PRT PSAR PSD PSE&G PSSUG PVRC PWR QA RA RAMPS RCC RCS RCCA
- SGS-UFSAR Parcel Land Use System Probable Maximum Hurricane Pressurizer OVerpressure Protection System Preoperational Testing Review committee Power-Operated Relief Valve Pressurizer Relief Tank Preliminary Safety Analysis Report Public Service Datum Public service Electric & Gas Public Service Electric & Gas Startup Group*
Pressure Vessel Research Committee Pressurized Water Reactor Quality Assurance Quality control Reduction Area Repair and Maintenance Procedure System Rod Cluster control Reactor Coolant syste~
Rod Cluster Control Assembly 1.6-7 Revision 6 February 15, 1987
RCFC RCS RCL RCP RCPB RCS REMP REP RG RBR RMS RMS RPS RSE RTD RVED RWST SBO SEC SER SGS-UFSAR Reactor Containment Fan Cooler Reactor coolant System Reactor Coolant Loop Reactor Coolant Pump Reactor Coolant Pressure Boundary Reactor COolant System Radiological Environmental Monitoring Program Radiation Exposure Permit Regulatory Guide Residual Heat Removal Radiation Monitoring System Root-Mean-square Reactor Protection System Reload Safety Evaluation Resistance Temperature Detector Reactor Vessel Examination Device Refueling Water Storage Tank Station Blackout Safeguards Equipment Control Safety Evaluation Report 1.6-8 Revision 15 June 12, 1996
SGB Steam Generator Blowdown SGS Salem Generating Station SIS Safety Injection System SMSA Standard Metropolitan Statistical Area SORC Station Operations Review Committee SPDS Safety Parameter Display System SRP Standard Review Plan SRSS Square-Root-of-the-Sum-of-the-Square SSE Safe Shutdown Earthquake SSG Salem Startup Group SSPS Solid State Protection System STA Shift Technical Advisor STGD Steam Turbine - Generator Division SWRI Southwest Research Institute sws Service Water System TDC Thermal Diffusion Coefficient TDH Total Dynamic Head TDS Total Dissolved Solid TLD Thermoluminescent Dosimeter 1.6-9 SGS-UFSAR Revision 6 February 15, 1987
TMI Three Mile Island T/MS Test/Maintenance System (AMSAC)
TSC Technical Support Center UE&C United Engineers & Constructors UFSAR Updated Fi~al Safety Analysis Report UPS Uninterruptible Power System USE Upper Shelf Energy UTG United Engineers and Constructors Test Group UTS Ultimate Tensile Stress VCT Volume Control Tank WDS Waste Disposal System w.g.
water gage WILMAPCO
- Wilmington Metropolitan Area Planning Council WMID WOL SGS-UFSAR Wisconsin-Michigan Inspection Device Wedge Opening Loading
- 1. 6-10 Revision 10 July 22, 1990