ML081330049

Updated Final Safety Analysis Report (Ufsar), Revision 17, Chapter 1.0 - Introduction and General Description of Plant

ML081330049
Person / Time
Site:	LaSalle
Issue date:	04/14/2008
From:	Exelon Generation Co, Exelon Nuclear
To:	Office of Nuclear Reactor Regulation
References
RS-08-045
Download: ML081330049 (50)

Text

LSCS-UFSAR 1.0-i REV. 14, APRIL 2002 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT TABLE OF CONTENTS PAGE

1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF PLANT 1.1-1

1.1 INTRODUCTION

1.1-1 1.

1.1 REFERENCES

1.1-3

1.2 GENERAL

PLANT DESCRIPTION 1.2-1 1.2.1 Principal Design Criteria 1.2-1 1.2.1.1 Safety Design Criteria 1.2-2 1.2.1.2 Power-Generation Criteria 1.2-6 1.2.2 Station Description of Plant Features Important to Safety 1.2-7 1.2.2.1 Site Characteristics 1.2-8 1.2.2.1.1 Site Location and Size 1.2-8 1.2.2.1.2 Description of Site Environs 1.2-8 1.2.2.1.3 Meteorology 1.2-8 1.2.2.1.4 Hydrology 1.2-8 1.2.2.1.5 Geology and Seismology 1.2-9 1.2.2.1.6 Design Bases Dependent on Site Environs 1.2-9 1.2.2.1.6.1 Liquid Waste Effluents 1.2-9 1.2.2.1.6.2 Wind Loading and Seismic Design 1.2-10 1.2.2.1.6.3 Flooding 1.2-10 1.2.2.2 General Arrangement of Structures and Equipment 1.2-10 1.2.2.3 Nuclear System 1.2-11 1.2.2.3.1 Reactor Core and Control Rods 1.2-11 1.2.2.3.2 Reactor Vessel and Internals 1.2-12 1.2.2.3.3 Reactor Recirculation System 1.2-12 1.2.2.3.4 Residual Heat Removal System 1.2-13 1.2.2.3.5 Primary Reactor Water Cleanup System 1.2-13 1.2.2.3.6 Reactor Protection System 1.2-13 1.2.2.3.7 Main Steamline Flow Restrictors 1.2-14 1.2.2.3.8 Refueling Interlocks 1.2-14 1.2.2.3.9 Nuclear System Pressure Relief System 1.2-14 1.2.2.3.10 Reactor Core Isolation Cooling System 1.2-14 1.2.2.4 Containment 1.2-15 LSCS-UFSAR TABLE OF CONTENTS (Cont'd)

PAGE 1.0-ii REV. 15, APRIL 2004 1.2.2.4.1 Primary Containment 1.2-15 1.2.2.4.2 Secondary Containment 1.2-16 1.2.2.4.3 Standby Gas Treatment System 1.2-16 1.2.2.4.4 Containment and Reactor Vessel Isolation Control System 1.2-17 1.2.2.4.5 Main Steamline Isol ation Valve Leakage Control System (U2 deleted, U1 abandoned-in-place) 1.2-17 1.2.2.4.6 Reactor Building Isolation Dampers 1.2-17 1.2.2.4.7 Containment Vent and Purge System 1.2-17 1.2.2.5 Emergency Core Cooling Systems 1.2-18 1.2.2.5.1 High-Pressure Core Spray System 1.2-18 1.2.2.5.2 Automatic Depressurization System 1.2-18 1.2.2.5.3 Low-Pressure Core Spray System 1.2-18 1.2.2.5.4 Low-Pressure Coolant Injection 1.2-19 1.2.2.6 Auxiliary Systems 1.2-19 1.2.2.6.1 Reactor Building Closed Cooling Water System 1.2-19 1.2.2.6.2 CSCS Equipment Cooling 1.2-19 1.2.2.6.3 Shielding Building 1.2-20 1.2.2.6.4 Reactor Building Ventilation Radiation Monitoring System 1.2-20 1.2.2.6.5 Main Steamline Radiation Monitoring System 1.2-20 1.2.2.6.6 Nuclear Leak-Detection System 1.2-20 1.2.2.6.7 Standby A-C Power Supply 1.2-21 1.2.2.6.8 D-C Power Supply 1.2-21 1.2.2.6.9 Standby Liquid Control (SLC) System 1.2-21 1.2.2.6.10 Station Equipment and Safe Shutdown from Outside the Control Room 1.2-22 1.2.2.6.11 Combustible Gas Control 1.2-22 1.2.3 Station Description of Features Important to Power Generation 1.2-22 1.2.3.1 Power Conversion System 1.2-22 1.2.3.1.1 Turbine-Generator 1.2-22 1.2.3.1.2 Main Steamlines 1.2-23 1.2.3.1.3 Main Condenser 1.2-23 1.2.3.1.4 Circulating Water System 1.2-24 1.2.3.1.5 Condensate and Feedwater System 1.2-24 1.2.3.2 Electrical Systems and Instrumentation Control 1.2-24 1.2.3.2.1 Electrical Power System 1.2-24 1.2.3.2.2 Electrical Power System Process Control and Instrumentation 1.2-25 1.2.3.2.3 Nuclear System Process Control and Instrumentation 1.2-25 1.2.3.2.3.1 Reactor Manual Control System 1.2-25 1.2.3.2.3.2 Rod Sequence Control 1.2-26 1.2.3.2.3.3 Recirculation Flow Control System 1.2-26 1.2.3.2.3.4 Neutron Monitoring System 1.2-26 1.2.3.2.3.5 Reactor Vessel Instrumentation 1.2-26 LSCS-UFSAR TABLE OF CONTENTS (Cont'd)

PAGE 1.0-iii REV. 15, APRIL 2004 1.2.3.2.3.6 Process Computer System 1.2-26 1.2.3.3 Power Conversion Systems Process Control and Instrumentation 1.2-26 1.2.3.3.1 Pressure Regulator and Turbine-Generator Control 1.2-27 1.2.3.3.2 Feedwater System Control 1.2-27 1.2.3.4 Radioactive Waste Systems 1.2-27 1.2.3.4.1 Gaseous Radwaste System 1.2-27 1.2.3.4.2 Liquid Radwaste System 1.2-28 1.2.3.4.3 Solid Radwaste System 1.2-29 1.2.3.5 Radiation Monitoring and Control 1.2-29 1.2.3.6 Miscellaneous Power Generation 1.2-29 1.2.3.6.1 New and Spent Fuel Storage 1.2-29 1.2.3.6.2 Fuel Pool Cleanup System 1.2-29 1.2.3.6.3 Service Water System 1.2-30 1.2.3.6.4 Demineralized Water Makeup System 1.2-30 1.2.3.6.5 Station HVAC 1.2-30 1.2.3.6.6 Heating, Ventilating, and Air-Conditioning (HVAC) Systems 1.2-30 1.2.4 Glossary 1.2-31 1.2.4.1 Definitions 1.2-31 1.2.4.2 Acronyms used in LSCS-UFSAR 1.2-36 1.2.5 References 1.2-38

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 Information 1.3-1

1.4 IDENTIFICATION

OF AGENTS AND CONTRACTORS 1.4-1 1.5 REQUIREMENTS FOR FURT HER TECHNICAL INFORMATION

- CURRENT CONCERNS FROM LSCS ACRS LETTER 1.5-1 1.6 MATERIAL INCORPORATED BY REFERENCE 1.6-1 LSCS-UFSAR 1.0-iv REV. 15, APRIL 2004 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT LIST OF FIGURES AND DRAWINGS FIGURES NUMBER TITLE 1.2-1 Reactor System - Rated Power Heat Balance

DRAWINGS CITED IN THIS CHAPTER*

DRAWING* SUBJECT M-1 Property Plat M-2 General Site Plan M-3 Development Plan M-4 Gene ral Arrangement - Roof Plan M-5 General Arra ngement - Reactor Building Floor Plans M-6 General Arrang ement - Reactor Building Floor Plans M-7 General Arrangement - Main Floor Plan M-8 General A rrangement - Mezzanine Floor Plan M-9 General Arrangement - Ground Floor Plan M-10 General Arrang ement - Upper Basement Floor Plan M-11 General A rrangement - Basement Floor Plan M-12 General Arrangement - Miscellaneous Floor Plans M-13 General Arrangement - Section "A-A" M-14 General Arrangement - Section "B-B" M-15 General Arrangement - Section "C-C" M-16 General Arrangement - Section "D-D" M-17 General Arrangement - Section "E-E" and "F-F" M-18 General Arrangement - Section "G-G" and "H-H" M-19 General A rrangement - Lake Screen House M-20 General Arrangement - River Screen House M-21 General Arra ngement - Off-Gas Filter Building M-22 Service Building General Plans M-54 P&ID Symbols M-89 Standby Gas Treatment P&ID, Units 1 & 2 M-1455 Reactor Building Ventilation System P&ID, Unit 1 M-1456 Reactor Building Ventilation System P&ID, Unit 2 M-5054 Logic Block Diagram, Notes and Symbols

• The listed drawings are included as "General References" only; i.e., refer to the drawings to obtain additional detail or to obtain background information. These drawings are not part of the UFSAR. They are controlled by the Controlled Documents Program.

LSCS-UFSAR 1.1-1 REV. 14, APRIL 2002 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

1.1 INTRODUCTION

This Updated Final Safety Analysis Repo rt (UFSAR) is submitted for the nuclear power station designated as the LaSalle County Station (LSCS) Unit 1, in accordance with the requirements of 10 CF R 50 Section 50.71(e) as published in the Federal Register on May 9, 1980.

Written as if LaSalle is a single unit pl ant, but applying to both units unless expressly written for Unit 1 or Unit 2, the original LSCS Final Safety Analysis Report (FSAR) was submitted in April 1976.

The FSAR was written in accordance with Regulatory Guide 1.70, "Standard Fo rmat and Content of Safety Analysis Reports for Nuclear Power Plants," Re vision 2, September 1975. The last amendment to the original FSAR prior to the 10CFR50.71(e) update for Unit 1 is number 64. The original FSAR up through and including amendment 64 will be referred to herein as the "FSAR." The responses to the NRC questions comprise three volumes of the FSAR. The informatio n in the response(s) was current at the time the LaSalle Operating License was granted, and has not since been revised. However, the responses have been reviewed and applicable updated information has been incorporated into the UFSAR text as required. Therefore, the three FSAR volumes of responses to NRC questions are now "historical information" pursuant to the guidance provided in NEI 98-03, Revision 1.

This UFSAR is the updated version of the FSAR and follows the same format as the FSAR (with allowed content criteria as spec ified in NEI 98-03 Revision 1, and as endorsed by Regulatory Guide 1.181 09/99). The UFSAR contains a description of LSCS Unit 1 which is up-to-date as of not more than 6 months prior to the latest revision date. The latest UFSAR revision date is specified in the document control section at the beginning of Volume 1.

The UFSAR is revised in accordance with 10CFR50.71(e).

The Nuclear Regulatory Commission approved the transfer of the facility licenses from Commonwealth Edison (ComEd) Company to Exelon Generation Company, LLC (EGC) on August 3, 2000 (Reference 1). References in the UFSAR to ComEd, CECo, and Commonwealth Edison have been retained, as appropriate, instead of being changed to EGC to properly preserve the historical content.

LSCS-UFSAR 1.1-2 REV. 14, APRIL 2002 The LSCS Preliminary Safety Analysis Report (PSAR) was submitted on November 3, 1970 (Docket Nos. 50-373 and 50-374). The station was constructed under construction permits CPPR-99 and CPPR-100 which were issued on September 10, 1973. Unit 1 was authoriz ed to commence power operation under license No. NPF-11 which was granted on Apr il 17, 1982. Unit 2 was authorized to commence power operation under license No. NPF-18.

This power generating station is located in the agricultural area of Brookfield Township, LaSalle County, Illinois. It is approximately 55 direct-line miles southwest of Chicago and 20 miles west of Dresden Nuclear Power Station. The plant is on flat terrain about 220 feet above the Illinois River channel which traverses north central Illinois some 3-1/2 miles to the north of the site.

The station utilizes two single-cycle forced-circulation boiling water reactors, each rated at 3489 MWt and designed for 3559 MWt.

The gross electric output of each unit is 1183 MWe; the net output is 1154 M We from each General Electric (GE) turbine-generator. The NSSS supplier was GE (Nuclear Energy Division). The plant, except for the NSSS, was designed by Sargent & Lundy (S&L) Engineers.

LSCS-UFSAR 1.1-3 REV. 14, APRIL 2002 The containment design employs the BWR Ma rk II concept of over-under pressure suppression with multiple downcomers conn ecting the reactor drywell to the water-filled pressure suppression chamber. The primary containment is a steel-lined, post-tensioned, concrete enclosure, housing the reactor and the suppression pool. This primary containment is entirely enclosed in the reinforced concrete reactor building which is the second ary containment structure.

The power generation complex includes several contiguous buildings, two reactor buildings, an auxiliary building (housing the control room), the turbine building, diesel-generator buildings, the radwaste building, the service building, and the off-gas building. Other buildings such as th e gatehouse, warehous es, etc., are also located in the general plant area. A lake screen house on the intake flume is located about 800 feet east of the main building complex. A small river screen house, located on the Illinois River, provides makeup water to the cooling lake for the LaSalle County Station.

Condenser cooling for the station is prov ided from a perched cooling lake of 2058 acres. The ultimate heat sink for emergency core cooling is a submerged pond and intake flume that underlies the cooling lake and the natural grade of the site.

The station utilizes a single vent stack for elevated release of all gaseous waste. Liquid radwaste is stored for decay or concentrated to solid waste for controlled disposal at regulated storage sites. The shielding design and plant layout incorporate 16 years of reactor operating expe rience at CECo to restrict radiological exposures to as low as reasonably achievable levels. Estimated radiological doses for normal operations and classical postulat ed accidents are all fractional parts of the federal radiological guidelines for siting and operation of nuclear power plants.

1.1.1 References

1. Letter from D. M. Skay (NRR) to O. D.

Kingsley (ComEd), dated August 3, 2000, and the associated NRC safety evaluation report.

LSCS-UFSAR 1.2-1 REV. 14, APRIL 2002 1.2 GENERAL PLANT DESCRIPTION For the purposes of this UFSAR, the LaSalle County Station is described in terms of its safety functions via safety criteria, and in terms of its power generation functions via nonsafety or power generation criteria. These general criteria define the design approach for safety and for power generation objectives of the nuclear power plant. Although the distinctions between safety design criteria and power generation design criteria are not always clear-cut, this arbitrary division of criteria facilitates the safety analyses while also enabling a portrayal of the plant equipment in sufficient detail to assist in the understanding of its functional purpose. As a secondary categorization technique in this report, functionally related equipment is further grouped into "systems" which are discussed primarily for their importance to safety and secondly as they relate to power generation

objectives.

The summary overview in this chapter provides brief descriptio ns of the site, its environs, and the arrangement of the station building complex. This is followed by the specific safety features of the NSSS, the power conversion system, the electrical and instrumentation equipment, and the radioactive waste and auxiliary support systems. A glossary of terms is provided in Section 1.2.4.

1.2.1 Principal

Design Criteria LaSalle County Station was designed, fabricated, erected, and is operated in such a manner that the release of radioactivity to the environment does not exceed the limits and guideline values of applicable government regulations pertaining to the release of radioactive materials for normal operations and abnormal transients and accidents.

The station is designed in conformance with applicable government regulations, ASME Codes, IEEE Codes, and other appropriate standards as noted herein. Compliance with NRC Regulatory Guides is discussed specifically in Appendix B.

The electrical design for essential safety equipment is of such redundancy and independence that no single failure of active or passive components can prevent the required safety actions.

The mechanical design for the equipment which makes up the primary pressure boundary conforms with Sections II, III, VIII, IX, and XI of the ASME Boiler and Pressure Vessel Code. Mechanical separation criteria were also incorporated in the plant design.

The classification of structures, components, and systems is discussed in Section 3.2. Specific conformance to th e 56 general design criteria of 10 CFR 50, LSCS-UFSAR 1.2-2 REV. 13 Appendix A, is discussed in Section 3.1. Other criteria are disc ussed in Chapter 3.0 and throughout this UFSAR.

Single failures are considered fo r applicable safety situations.

The plant is designed to produce steam for direct use in a turbine-generator unit that feeds CECo's electrical network.

1.2.1.1 Safety Design Criteria

a. The fuel cladding is designed to retain integrity as a radioactive material barrier throughout the design power range. The fuel cladding is designed to accommodate, without loss of integrity, the pressures generated by the fission gases released from the fuel material throughout the design life of the fuel.

b. The reactor is designed so that there is no tendency for divergent oscillation of any operating characteristic, considering the interaction of the reactor with other appropriate station systems. c. The reactor core is so designed that its nuclear characteristics do not contribute to a divergent power transient.

d. The reactor core and reactivity control system are designed so that control rod action is capable of bringing the core subcritical and maintaining it so, even with the rod of highest reactivity worth fully withdrawn and unavailable for insertion.

e. Sufficient indications are provided to allow determination that the reactor is operating within the envelope of conditions considered in this safety analysis.

f. Those portions of the nuclear system that form part of the reactor coolant pressure boundary are designed to retain integrity as a radioactive material barrier following abnormal operational transients and credible accidents.

g. A primary containment is prov ided that completely encloses the reactor system. The containment employs the pressure-suppression concept.

h. It is possible to test primary containment integrity and leaktightness at periodic intervals.

LSCS-UFSAR 1.2-3 REV. 13 i. A secondary containment completely encloses the primary containment. This secondary containment includes the capability to control the release of radioactive materials from the primary containment.

j. Provisions are made to remove long-term energy from the primary containment as necessary to maintain the integrity of the containment following accidents which release energy into the containment.

k. Piping that penetrates the primary containment and could serve as a path for the uncontrolled release of any radioactive leakage to the environs is automatically isolated whenever such an

uncontrolled release of radioactive material is threatened. Such isolation is effected in time to limit radiological effects to significantly less than the prescribed radiation limits.

l. The primary and secondary containments, in conjunction with other engineered safety features, limit the radiological effects of accidents resulting from the release of radioactive material within these containment volumes to less than prescribed radiation limits.

m. The control room is shielded against radiation so that continued occupancy is possible under accident conditions.

n. In the event that the control room becomes uninhabitable, it is possible to bring the reactor from power range operation to cold shutdown conditions by a remote shutdown system located outside the control room.

o. A backup reactor shutdown system, independent of normal reactivity control provisions, has the capability to shut down the reactor from any normal operating condition and subsequently

to maintain the shutdown condition.

p. Interlocks or other automatic equipment are provided as backup to procedural controls to avoid conditions requiring needless functioning of nuclear safety systems or engineered safety features.

q. Faulted equipment is detected and isolated from the electrical systems with a minimum of di sturbance via activation of protective relaying in the event of equipment failure.

LSCS-UFSAR 1.2-4 REV. 13 r. The Class 1E power systems are designed as triple-bus systems, with any two buses being adequate to safely shut down the unit.

s. Standby electrical power sou rces are provided to allow prompt reactor shutdown and removal of decay heat under circumstances where normal auxiliary power is not available.

t. Where positive, precise action is immediately required in response to abnormal operational transients and accidents, such action is automatic and requires no decision or manipulation of controls by station operations personnel.

u. Voltage relays are used on the emergency equipment buses to isolate these buses from the normal electrical system in the event of loss of offsite power and concurrently to initiate starting of the standby emergency power system generators.

v. Standby electrical power sources have sufficient capacity to power all nuclear safety systems and engineered safety features requiring electrical power.

w. The design of nuclear safe ty systems and engineered safety features includes design a llowances for unusual natural phenomena such as earthquakes, floods, and storms on the site.

x. Nuclear safety systems and engineered safety features act to ensure that no violation of the reactor coolant pressure

boundary results from internal pressures caused by abnormal operational transients or accidents.

y. Provisions are made for control of active components of nuclear safety systems and engineered safety features from the control room during normal operations. During a remote shutdown condition, this control is purposely removed from the control

room. z. Engineered safety feat ures are designed to permit demonstration of their performance.

aa. Heat-removal systems are provided to remove decay heat generated in the core under circumstances wherein the normally

operational heat removal systems become inoperative. The capacity of such systems is adequate to prevent fuel cladding damage. The reactor is ca pable of being shut down automatically sufficiently fast to permit decay-heat-removal LSCS-UFSAR 1.2-5 REV. 13 systems to become effective following loss of operation of normal heat-removal systems.

bb. Emergency core cooling systems are provided to limit fuel cladding temperatures to le ss than the fragmentation temperature in the event of a loss-of-coolant accident.

cc. The emergency core cooling systems (ECCS) provide for continuity of core cooling over the complete range of postulated break sizes in the reactor coolant pressure boundary.

dd. Operation of the ECCS is initiated automatically when required, regardless of the availability of offsite power supplies and the

normal generating system of the station.

ee. Auxiliary systems, such as emergency cooling water, required heating and ventilating, communications, and lighting, are designed to function during normal and accident conditions.

ff. The fuel cladding, in conjunction with other plant systems, is designed to retain integrity throughout any abnormal operational transient.

gg. Gaseous, liquid, and solid waste disposal facilities are designed so that the discharge and offsite shipment of radioactive effluents can be made in accordance with applicable regulations.

hh. The radwaste systems are designed to minimize the release of radioactive materials from the st ation to the environs. Such releases as may be necessary during normal operations are limited to values that meet th e requirements of 10 CFR 20 and 10 CFR 50.

ii. The design of the systems provides means by which station operations personnel can be info rmed whenever specified limits on the release of radioactive material may be approached.

jj. The control room is shielded against radiation so that occupancy is possible under accident condit ions and so that radiation doses are less than those set by Criterion 19 of 10 CFR 50, Appendix A.

kk. Fuel handling and storage fa cilities are designed to prevent inadvertent criticality of new and spent fuel and to maintain shielding and cooling of spent fuel.

LSCS-UFSAR 1.2-6 REV. 13 1.2.1.2 Power-Generation Criteria

a. Reactor power level is manually controllable.

b. Control of the reactor is possible from a single location.

c. Reactor controls, including alarms, are arranged to allow rapid operator assessment of reactor conditions and the location of reactor system malfunctions.

d. Control equipment is provided to allow the reactor to respond automatically to both minor and major load changes including

abnormal operational transients.

e. Reactor controls, including alarms, are arranged to allow rapid operator assessment of reactor conditions and to locate reactor system malfunctions.

f. Backup heat-removal systems are provided to remove decay heat generated in the core under circumstances wherein the normal operational heat removal systems become inoperative. The capacity of such systems is adequate to prevent fuel cladding damage.

g. A means is provided by which station operators can be informed when limits on the release of radioactivity are approached.

h. The power conversion system is designed to ensure that any fission products or radioactivity associated with the steam and condensate during normal operation are contained safely inside the system or are released under controlled conditions in accordance with appropriate regulations and waste disposal procedures.

i. Sufficient normal and standby auxiliary sources of electrical power are provided to attain prompt shutdown and continued maintenance of the station in a safe condition under all credible circumstances.

j. Control of the nuclear system and the power-conversion equipment is possible from a central location.

k. Control equipment is provided to control the reactor pressure throughout its operating range.

LSCS-UFSAR 1.2-7 REV. 14, APRIL 2002

l. Control equipment in the feedwater system maintains the water level in the reactor vessel at the optimum level required by steam separators.

m. Metering for essential generators, transformers, and circuits is monitored in the control room.

n. Components of the power-conversion systems are designed to produce electrical power from the steam coming from the reactor, condense the steam into water, and return the water to the reactor as heated feedwater with a major portion of its gases and particulate impurities removed.

o. Gaseous, liquid, and solid radioactive waste disposal systems are designed so that in-plant pr ocessing, discharge of effluents, and offsite shipments are in accordance with all applicable federal regulations.

p. Auxiliary systems that are not required to effect safe shutdown of the reactor or maintain it in a safe condition are designed so that a failure of these systems does not prevent the essential auxiliary systems from performing their design functions.

q. Radiation shielding is designed and access control provisions are made to minimize radiation levels and provide the means to control radiation doses within the limits of published

regulations.

r. Radiation shielding is provided and access control patterns are established to allow a properly tr ained operating staff to control radiation doses within the limits of applicable regulations in any mode of normal station operations.

1.2.2 Station

Description of Plant Features Important to Safety This section provides an overview of those plant features of the LaSalle County Station that are important to safety considerations. The following subsections describe:

a. site characteristics - acreag e, location, environs, meteorology, hydrology, seismology, and site dependent design bases;

b. general arrangement of structures and equipment.

LSCS-UFSAR 1.2-8 REV. 14, APRIL 2002 Descriptive symbols appearing in P&ID's referenced by the UFSAR, and in UFSAR Figures which are based upon P&ID's, are defined on Drawing M-54 and M-5054.

The two side-by-side power generating units are essentially independent, although certain components are shared, such as the common control room, common radwaste facility, the st ation vent stack, etc.

1.2.2.1 Site Characteristics 1.2.2.1.1 Site Location and Size LSCS is located on an irregular pentagon ally shaped site (see Drawing No. M-1).

Approximately 3060 acres lie within the site boundaries, with 2058 acres being used for a cooling lake. A pipeline corridor, consisting of 815 acres, extends north from the site to the Illinois River, which is ap proximately 5.0 miles north of the reactor (see Drawing No. M-2).

1.2.2.1.2 Description of Site Environs Human population near the site is sp arse. Isolated farm homes and small groupings of houses typify the inhabited areas. The site environs are further described in Subsection 2.1.3 and Section 2.2.

1.2.2.1.3 Meteorology The site is subject to typical continental meteorology characterized by high variability and a wide range of temper ature extremes. The average annual precipitation at Ottawa based upon 89 years of record is 34 inches. This includes an annual average of 27 inches of snow. Thunderstorms occur on an average of 49 days per year.

The prevailing winds of this area are primarily south by southwest at an average of 10 mph. The probability of tornado occurrence at the site is 0.0016 for any given year, which converts to a recurrence interval of 625 years.

Dispersion of normal releases from the elevated station vent stack is further discussed in Section 2.3.

1.2.2.1.4 Hydrology The LSCS site is located in the Illinois Ri ver basin. The Illinois River is a perennial stream with a drainage area of approximately 7640 mi 2 surrounding the plant site. The normal pool elevation of the Marseilles pool is 483.25 feet MSL (USGS datum 1912 adjustment, which is 0.462 foot lowe r than USGS datum 1929 adjustment).

LSCS-UFSAR 1.2-9 REV. 13 The plant grade is 710 feet MSL (1929 datum). Therefore, the station site may be described as "floodproof" or "dry" with regard to floods in the Illinois River. Flood effects on the river screen hous e are discussed in Section 2.4.

1.2.2.1.5 Geology and Seismology The site is located in the Central Lowland Physiographic Province and in one of the most stable tectonic areas of the North Am erican Craton. The regional structure consists of a system of sedimentary basins, arches, and domes of Paleozoic age.

The depth to Precambrian rock is approximately 4200 feet at the site. Cambrian and Ordovician sandstones and dolomites form most of the sedimentary column overlying the Precambrian basement.

Pennsylvanian cyclothems, mainly shale, form a cap about 120 feet in thickness over the Ordovician strata in the site area.

Approximately 170 feet of predominantly Wisconsinan glacial drift overlies the bedrock surface in the site area. The ne arest major fault zone is the Sandwich Fault Zone, which is located approximately 26 miles northeast of this site and is noncapable. There are no geologic features at or near the site which would preclude its use for the construction and oper ation of the nuclear power station.

For seismic design of Seismic Category I structures, the maximum horizontal acceleration caused by the safe shutdown earthquake (SSE) is 20% of gravity at the free field foundation level. The operating-basis earthquake (OBE) is a horizontal acceleration of 10% of gravity at the found ation level. For additional information concerning geology and seismology consult Section 2.5.

1.2.2.1.6 Design Bases Dependent on Site Environs An elevated, 370-foot, station vent stack common to both units is provided for the continuous release of all gaseous effluents. In addition, a recombiner and charcoal bed adsorber system are employed to limit gaseous effluent releases from normal operations. This subject is furthe r discussed in Subsection 11.3.2.

1.2.2.1.6.1 Liquid Waste Effluents Liquid waste releases are controlled to ensure that concentrations at the point of discharge do not exceed 10 CFR 20 limits.

This subject is further discussed in Section 11.2.

LSCS-UFSAR 1.2-10 REV. 13 1.2.2.1.6.2 Wind Loading and Seismic Design The structures and components whose failure might conceivably contribute to an uncontrolled release of fission products are designed to resist tornado loads possessing a maximum wind velocity of 360 mph and an internal differential pressure of 3 psi in 3 seconds. This subject is further discussed in Section 3.3.

1.2.2.1.6.3 Flooding The plant design accounts for safety static water head pressures on plant structures. Consult Sections 2.4 and 3.4 and Subsection 1.2.2.1.4 for additional information.

1.2.2.2 General Arrangement of Structures and Equipment Station equipment is housed in the following principle structures:

a. reactor building - the nuclear steam supply system, the drywell, suppression pool, primary containment, new and spent fuel pools, refueling equipment, and emergency core cooling

equipment;

b. auxiliary building - the control room, the HVAC equipment, the station vent stack, and much of the station electrical switchgear;

c. turbine building - the power conversion equipment and feedwater cleanup equipment;

d. off-gas filter building - off-gas filters and associated equipment;

e. diesel-generator buildings - th e standby diesel generators, diesel oil storage tanks, CSCS cooling water pumps and strainers, and associated controls and instrumentation;

f. service building - the mach ine shop, offices, warehouses, and training rooms;

g. lake screen house - the service and circulating water pumps with their accompanying equipment and instrumentation;

h. river screen house - the lake makeup equipment and control instrumentation;

i. solid radwaste building - all solid radwaste disposal equipment;

LSCS-UFSAR 1.2-11 REV. 14, APRIL 2002 j. switchyard;

k. security gatehouse; and

l. interim radwaste storage facility

The arrangement of these structures on the station site is shown in Drawing No.

M-3. The arrangement of the equipment inside the main buildings is shown in Drawing Nos. M-4 through M-22.

1.2.2.3 Nuclear System The nuclear system includes a direct-cycle, forced-circulation, General Electric boiling water reactor that produces steam for direct use in the steam turbine. A heat balance showing the major parameters of the nuclear system for the rated power conditions is shown in Figure 1.2-1. This system is discussed in Chapter 4.0.

1.2.2.3.1 Reactor Core and Control Rods Fuel for the reactor core consists of slightly enriched uranium dioxide pellets sealed in Zircaloy tubes. These fuel rods are a ssembled into individual fuel assemblies.

Gross control of the core is achieved by movable, bottom-entry control rods which are positioned by individual control rod drives.

When a scram is signaled by the reactor protection system, the high-pressure water stored in an accumulator in the hydraulic control unit forces its control rod into the core. This system is disc ussed in Subsection 4.2.3.

A control rod velocity limiter is attached to each control rod to limit the velocity at which a control rod can fall out of the core should it become detached from its control rod drive. This action limits the rate of reactivity insertion resulting from a rod drop accident. The limiters contain no moving parts.

Control rod drive housing supports, located underneath the reactor vessel near the control rod housings, limit the travel of a control rod in the event that a control rod housing is ruptured. These supports prevent a nuclear excursion as a result of a housing failure and thus protect the fuel barrier.

Each fuel assembly has several fuel ro ds with a burnable poison, gadolina (Gd 2 O 3) mixed in solid solution with UO

2.

The initial core and reload fuel were prov ided by General Electric (GE). Beginning in 1999, reload fuel was provided by Siem ens Power Corporation, Nuclear Division (SPC). While the SPC fuel differs slightly from the GE fuel, the basic design requirements and description remain the same. Where design features and LSCS-UFSAR 1.2-12 REV. 13 analytical methods differ substantially between the two fuel vendors, the UFSAR test has been revised to describe or reference either the appropriate method, or both methods. 1.2.2.3.2 Reactor Vessel and Internals The reactor vessel contains: the core and supporting structures; the steam separators and dryers; the jet pumps; the control rod guide tubes; the distribution lines for the feedwater, core sprays, and liquid level control; the incore instrumentation, and other components. The main connections to the vessel include: steamlines, coolant recirculation lines, feedwater lines, control rod drive and incore nuclear instrument housings, high- and low- pressure core spray lines, low-pressure core injection lines, standby liquid control line, jet pump pressure-sensing lines, water level instrumentation, and control rod drive system return lines. The SS-clad low alloy steel reactor vessel is designed and fabricated in accordance with applicable codes for a pressure of 1250 psig. The nominal operating pressure in the steam space above the separators is 1020 psia.

The reactor core is cooled by demineralized water that enters the lower portion of the core and boils as it flows upward around the fuel rods. The steam leaving the core is dried by steam separators and dryers located in the upper portion of the reactor vessel. The steam is then directed to the turbine through the main steamlines. Each steamline is provided with two isolation valves in series, one on each side of the primary containment barrier. This system is described further in Subsection 5.2.2.

1.2.2.3.3 Reactor Recirculation System The reactor recirculation system pumps reactor coolant through the core. This is accomplished by two recirculation loops external to the reactor vessel but inside the primary containment. Each external loop contains one high capacity, motor-driven recirculation pump, a flow control valve, and two motor-operated gate valves for suction shutoff and discharge shutoff purpos es. Each pump suction line contains a flow measuring system. The variable-position flow control valve in the main recirculation pipe allows control of reacto r power level through the effects of coolant flow rate on moderator void content. The pumps can be operated at either high speed or low speed.

Low speed operation of the pumps provides capability for reduced recirculation flow during startup, shutdown, or other times of reduced power operation.

Jet pumps provide a continuous internal circulation path for the major portion of the core coolant flow. The jet pumps are located in the annular region between the core shroud and the vessel's inner wall; thus any recirculation line break would still LSCS-UFSAR 1.2-13 REV. 13 allow core flooding to approximately two-thirds of the core height--the level of the inlet of the jet pumps.

A detailed, comprehensive description of the reactor recirculation system is provided in Appendix G of the UFSAR.

1.2.2.3.4 Residual Heat Removal System The residual heat removal (RHR) system is a set of pumps, heat exchangers, and piping that fulfills the cooling functions under various configurations and conditions as follows:

a. Shutdown cooling and reactor vessel head spray - to remove residual heat (decay heat and sensible heat) from the nuclear boiler system after a normal shutdown and cooldown.

b.

• Steam condensing mode deleted per AIR 373-160-92-00108. (E01-2-9500158)

c. Low-pressure coolant injection mode - This capability is discussed in Subsection 1.2.2.5.3.

d. The primary containment cooling mode limits temperature, hence the pressure, by water sp ray action inside the primary containment when activated during an isolation event.

e. The suppression pool cooling mode limits the water temperature of the suppression pool follo wing a design-basis LOCA or following testing of the safety/relief valves and the RCIC system which discharge to the suppression pool.

This system is discussed further in Subsection 5.4.7.

1.2.2.3.5 Primary Reactor Water Cleanup System The reactor water cleanup system recirculates a portion of reactor coolant through a filter-demineralizer to remove particulate and dissolved impurities from the reactor system under controlled conditions (see Subsection 5.4.8).

1.2.2.3.6 Reactor Protection System The reactor protection system (RPS) is an electric logic network which initiates a rapid, automatic shutdown of the reactor. It acts in time to prevent fuel clad damage and any nuclear system process barrier damage associated with abnormal operational transients. The reactor protecti on system overrides all operator actions LSCS-UFSAR 1.2-14 REV. 13 and process controls. It uses a logic of one-out-of-two taken twice for protective actions. The design is based on a fail-safe philosophy that allows appropriate protective action even when a single failu re occurs. Some of the neutron monitors function uniquely as part of this nuclear safety system. The high neutron flux signals are used for this scram protection.

The source range monitors (SRM's) and the intermediate range monitors (IRM's) provide flux level indications during reactor startup and low-power operation.

This system is further di scussed in Section 7.2.

1.2.2.3.7 Main Steamline Flow Restrictors

A venturi-type flow restrictor is installed in each steamline. These devices limit the loss of coolant from the reactor vessel before the main steamline isolation valves are closed in case of a main steamline break outside the primary containment.

This system is further disc ussed in Subsection 5.4.4.

1.2.2.3.8 Refueling Interlocks A system of interlocks that restricts movement of refueling equipment and control rods when the reactor is in the refueling and startup modes is provided to prevent inadvertent criticality during refueling operations. The interlocks back up procedural controls that have the same objective. The interlocks affect the refueling platform, refueling platform hoists, fuel grapple, and control rods. This system is discussed in Section 7.7.

1.2.2.3.9 Nuclear System Pressure Relief System A pressure relief system consisting of sa fety/relief valves mounted on the main steamlines prevents excessive nuclear boiler pressure following either abnormal operational transients or accidents. This system is discussed in Subsection 5.2.2.

1.2.2.3.10 Reactor Core Isolation Cooling System Although not a safety system, the reactor core isolation cooling (RCIC) system provides makeup water to the reactor vessel when the vessel is isolated. It uses a steam-driven turbine-pump unit and operates automatically to maintain adequate water level in the reactor vessel. The RCIC pump takes water from the condensate storage tank or directly from the suppression pool or from the suppression pool via the RHR heat exchangers, depending on reactor conditions, and discharges it through the head spray nozzle of the reactor vessel to maintain reactor water level. This system is also discussed in Subsection 5.4.6.

LSCS-UFSAR 1.2-15 REV. 13 1.2.2.4 Containment The containment is a set of leaktight barriers which prohibit the release of fission products to the environs. Although these barriers include the fuel cladding and the reactor pressure vessel, the word "containment" connotes the structures in which the reactor pressure vessel and the nuclear process equipment operate. The

primary containment, utilizing the pressure suppression concept, and the secondary containment, including the reactor buildings with their atmospheric ventilation systems and the standby gas treatment system (SGTS), have the capability of minimizing rapid pressure transients. The containment provides an isolation function for the lines penetrating the prim ary containment. Ventilation dampers on secondary containment are provided to inhibit leakage. Both primary and secondary containments are designed to Class I seismic standards.

1.2.2.4.1 Primary Containment The primary containment is designed to limit the release of radioactivity to the environs subsequent to the postulated loss-of-coolant accident. The vapor suppression concept for the reduction of internal pressure is utilized in the LSCS design. The drywell is constructed above the wetwell in a single concrete vessel shaped like the frustrum of a cone on top of a right circular cyclinder. Unique features of the LSCS primary containment are as follows:

a. The drywell is lined with carbon steel.

b. The wetwell is lined in its entirety with stainless steel; this includes the central pedestal, the supporting columns, and the

ceiling. c. There are no projections, equipment, or galleries inside the wetwell; the wetwell is "structurally clean" internally.

d. Four vacuum breaker lines externally connect the wetwell and the drywell to provide internal pressure relief from the initial air pressurization of the wetwell. These vacuum breakers are serviced from within the secondary containment.

e. During normal operations, the suppression pool water volume is level-controlled; however, during servicing when the reactor head is removed, the suppression pool water is used to fill the reactor cavity and pool. (Suppression pool water is demineralized, filtered, and pumped for this dual usage.)

The drywell and wetwell are separated by a reinforced concrete floor which is penetrated by 98 stainless steel downco mers. The primary containment is a LSCS-UFSAR 1.2-16 REV. 13 posttensioned concrete and steel structure which houses the reactor vessel, the reactor coolant recirculation loops, and other principal connections of the reactor fluid loops making up the primary pressure boundary.

Cooling systems are provided to remove heat from the reactor core, the drywell, and the water in the suppression chamber, and thus provide continuous cooling of the

primary containment under postulated accident conditions. Isolation valves are used to ensure that radioactive materials which otherwise might be released from the reactor during the course of an accident are contained within the primary containment.

This subject is addressed in Section 6.2 and Reference 1.

1.2.2.4.2 Secondary Containment The reactor building completely surrounds the primary containment and functions as a secondary containment when the primary containment is closed and in service. The reactor building also houses refueling and reactor servicing equipment, new and spent fuel storage facilities, and othe r reactor safety and auxiliary systems.

The design of the reactor building includes provisions for seismic load resistance and low infiltration and exfiltration rates. The building consists of poured-in-place, reinforced concrete exterior walls up to the refueling floor. Above this level, the building structure is steel frame with insulated metal siding with sealed joints. Access to the secondary containment is through interlocked double doors.

This subject is addressed in Section 6.2.

1.2.2.4.3 Standby Gas Treatment System The standby gas treatment system (SGTS) consists of two identica l filter trains and interconnecting piping and ductwork. The individual trains are available to each reactor building (M-89).

Either train by itself is capable of exchanging both reactor building volumes once in a 24-hour period.

The system maintains a slightly negative internal building pressure and processes all gaseous effluent prior to its discharge via the station vent stack.

All SGTS equipment is powered from the essential buses and is started either automatically or manually from the control room. This system is further discussed in Subsection 6.5.1.

LSCS-UFSAR 1.2-17 REV. 15, APRIL 2004 1.2.2.4.4 Containment and Reactor Vessel Isolation Control System The primary containment and reactor vessel isolation control system automatically initiates closure of isolation valves to close off all potential leakage paths for radioactive material to the environs. This action is taken upon indication of a potential breach in the nuclear system pr ocess barrier. A containment and isolation status panel is provided in the control room to display the status and operations of the isolation control system. This system is further discussed in Subsection 6.2.4.

Although all pipelines that both penetrate the containment and offer a potential release path for radioactive material are provided with redundant isolation capabilities, the main steamlines, because of their large size and large mass flow rates, are given special isolation consideration. Automatic isolation valves are provided in each main steamline. Each is powered by both air pressure and spring force. 1.2.2.4.5 Main Steamline Isolat ion Valve Leakage Control System (U2 deleted, U1 abandoned-in-place)

The main steamline isolation valve leakag e control system (MSIV-LCS) provided originally has been deleted. The valve leakages are processed through the main steam lines, main steamline drains, and the main condenser. The system is discussed in Section 6.8.

1.2.2.4.6 Reactor Building Isolation Dampers The reactor building heating, ventilation, and air-conditioning system supply and discharge ducts are each supplied with two isolation dampers in series. These dampers are designed to maintain secondary containment isolation and are automatically closed whenever the standb y gas treatment system is initiated. These isolation dampers may also be manually closed from the local control panel (see Drawing Nos. M-1455 and M-1456). This system is further discussed in Subsections 6.2.4 and 7.3.7.

1.2.2.4.7 Containment Vent and Purge System Although not a safety system, a separate dual-train containment vent and purge system is connected, via isolation valving, in parallel with the SGTS filter trains.

This vent and purge equipment includes charcoal and HEPA filters with exhaust fans and ducting to the station vent stack. This equipment is to be used to clean up the primary and secondary containment atmospheres when low-level airborne

contamination exists, thereby attaining as low as reasonably achievable worker exposures. The SGTS is therefore reserved for the accident case and need not be operated for routine atmospheric cleanup.

LSCS-UFSAR 1.2-18 REV. 13 1.2.2.5 Emergency Core Cooling Systems Four emergency core cooling systems are pr ovided to maintain fuel cladding below fragmentation temperature in the event of a breach in the reactor coolant pressure boundary that results in a loss of reactor coolant. The systems are:

a. high-pressure core spray (HPCS) system;

b. automatic depressurization system (ADS);

c. low-pressure core spray (LPCS) system; and

d. low-pressure coolant injection (LPCI), an operating mode of the residual heat removal system.

These systems are further discussed in Section 6.3 1.2.2.5.1 High-Pressure Core Spray System The HPCS system provides and maintains an adequate coolant inventory inside the reactor vessel to maintain fuel cladding temperatures below fragmentation temperature in the event of breaks in the reactor coolant pressure boundary. The system is initiated by either high pressure in the drywell or low water level in the vessel. It operates independently of all other systems over the entire range of pressure differences from greater-than-normal operating pressure to zero. The HPCS system pump motor is powered by a diesel generator if auxiliary power is not available, and the system may also be us ed as a backup for the RCIC system. This system is further discussed in Subsection 6.3.2.

1.2.2.5.2 Automatic Depressurization System The automatic depressurization system rapidly reduces reactor vessel pressure in a

LOCA situation in which the HPCS system fails to maintain the reactor vessel water level. The depressurization provided by the system enables the low-pressure emergency core cooling systems to deliver cooling water to the reactor vessel. The ADS will not be activated unless either the LPCS or LPCI pumps are operating. This is to ensure that adequate coolant wi ll be available to maintain reactor water level after the depressurization. This system is further discussed in Subsection 6.3.2.

1.2.2.5.3 Low-Pressure Core Spray System The LPCS system consists of one independent pump and the valves and piping to deliver cooling water to a spray sparger over the core. The syst em is actuated by conditions indicating that a breach exists in the reactor coolant pressure boundary, LSCS-UFSAR 1.2-19 REV. 14, APRIL 2002 but water is delivered to the core only after reactor vessel pressure is reduced. This system provides the capability to cool the fuel by spraying water into the fuel channels. In conjunction with the HPCS , ADS, AND LPCI mode of RHR, the LPCS can maintain the fuel cladding below final acceptance criteria limits for the entire spectrum of breaks. This system is further discussed in Subsection 6.3.2.

1.2.2.5.4 Low-Pressure Coolant Injection Low-pressure coolant injection is an operating mode of the residual heat removal (RHR) system, but it is discussed here because the LPCI mode acts as an engineered safety feature in conjunction with the other emergency core cooling systems. LPCI uses the pump loops of the RHR system to inject cooling water directly into the pressure vessel. LPCI is actuated by conditions indicating a breach in the reactor coolant pressure boundary, but water is delivered to the core only after reactor vessel pressure is reduced.

LPCI operation provid es the capability of core reflooding, following a loss-of-coolant accident, in time to maintain the fuel cladding below final acceptance criteria limits. This system is further discussed in Subsection 6.3.2.

1.2.2.6 Auxiliary Systems Certain supportive equipment have functions which relate indirectly to the safety performance of those systems previously described in Subsections 1.2.2.1 through 1.2.2.5. Some supportive equipment regulates the internal environments in which the engineered safety systems normally operate, hence they contribute to the assurance of a "ready status" for these ESF systems. In case of accident, they provide standby power, added heat sink ca pacity for thermal control, and an added assurance of reactor shutdown capability. For completeness, this auxiliary equipment is briefly noted here because it indirectly supports safety objectives.

1.2.2.6.1 Reactor Building Closed Cooling Water System The reactor building closed cooling water system consists of five pumps, five heat exchangers, and control and instrumentation to provide adequate cooling for the reactor auxiliary systems. Spare equipment is provided to ensure adequate cooling capacity during normal conditions. This system is further discussed in Subsection 9.2.3.

1.2.2.6.2 CSCS Equipment Cooling The CSCS equipment cooling water system supplies cooling water to the RHR heat exchangers, diesel-generator coolers, core standby cooling system (CSCS) area coolers, and the LPCS and RHR pumps.

Each unit's CSCS consists of three separate electrical and physical divisions, one of which is sh ared between units.

LSCS-UFSAR 1.2-20 REV. 13 Each division is provided with separate pumps and draws cooling water from the CSCS cooling pond through separate intake pipes. Cooling water is returned to the station from the CSCS cooling pond through three discharge pipes corresponding to the three divisions of each unit.

1.2.2.6.3 Shielding Building The Mark II containment concept does not require a shielding building.

1.2.2.6.4 Reactor Building Ventilation Radiation Monitoring System The reactor building ventilation radiation monitoring system consists of radiation detectors which monitor the activity level of the normal exhaust from the reactor building en route to the station vent stack.

Upon detection of high radiation due to an accidental release, the reactor building is automatically isolated and the standby gas treatment system is started. For small, nonaccident releases of radioactivity, the drywell purge unit is utilized to exhaust to the station vent stack. This monitoring system is further di scussed in Subsection 7.6.1.

1.2.2.6.5 Main Steamline Radiation Monitoring System The main steamline radiation monitoring sy stem consists of fo ur gamma radiation monitors located external to the main steamlines just outside the primary

containment. The monitors are designed to detect a gross release of fission products from the fuel.

Upon detection of high radiation, the sign als generated by these monitors are used to provide an alarm in the control room. This system is further discussed in Subsection 7.6.1.

1.2.2.6.6 Nuclear Leak-Detection System The nuclear leak-detection system consists of temperature, pressure, flow, and radioactivity sensors and associated instrumentation with alarms used to detect and annunciate leakage from the following system:

a. nuclear boiler system, b. reactor water cleanup (RWCU) system, c. residual heat removal (RHR) system,

d. reactor core isolation cooling (RCIC) system, e. fuel pool cooling system, LSCS-UFSAR 1.2-21 REV. 13

f. feedwater system

g. fuel pool cooling system, and

h. instrument lines associated with the above systems.

Small leaks are detected by temperature and pressure changes, fill-up rates of drain sumps, and fission-product concentration inside the primary containment. Large leaks are also detected by changes in reactor water level and changes in flow rates in process lines. This sy stem is further discussed in Subsection 5.2.5.

1.2.2.6.7 Standby A-C Power Supply Standby a-c power is supplied from five di esel generators. Two diesel-generators are provided for each Unit 1 and 2. The other diesel-generator is arranged to serve essential auxiliaries for either Unit 1 or Unit 2.

This subject is further disc ussed in Subsection 8.3.1.

1.2.2.6.8 D-C Power Supply D-c power supplies consist of storage batteri es of ample capacity for all essential emergency loads for a minimum period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. D-c power supplies of three different voltage levels are provided for each of the two units.

Two independent 24-volt batteries are provided for each unit for neutron-monitoring instrumentation.

A three battery 125-volt system is provided for each unit for circuit breaker controls and other essential control systems. Each independent battery feeds its respective ESF division.

A separate 250-volt system is also prov ided for each unit for essential power required for valve operators and emergency pump motors.

These battery systems are desc ribed in Subsection 8.3.2.

1.2.2.6.9 Standby Liquid Control (SLC) System Although not intended to provide prompt reactor shutdown, as the control rods do, the standby liquid control system provides a redundant, independent, and different way to bring the nuclear fission reaction to subcriticality and to maintain subcriticality as the reactor cools. The system makes possible an orderly and safe shutdown in the event that not enough control rods can be inserted into the reactor LSCS-UFSAR 1.2-22 REV. 17, APRIL 2008 core to accomplish shutdown in the normal manner. The system is sized to counteract the positive reactivity effect from rated power to the cold shutdown condition. This system is discussed in Subsection 9.3.5.

1.2.2.6.10 Station Equipment and Safe Shutdown from Outside the Control Room

A separate remote shutdown control panel is provided in the auxiliary-electric equipment room, with sufficient indication to knowledgeably shut down the reactor from outside the control room. This it em is discussed in Subsection 7.4.4.

1.2.2.6.11 Combustible Gas Control The combustible gas control system consists of a hydrogen recombiner for each unit, with a crosstie between units for redundancy. In the event of a LOCA, the recombiner system can be actuated to pr event the hydrogen-oxygen level within the primary containment from reaching the flammability limit. The hydrogen recombining function of the hydrogen recombiners is abandoned in place. This system is further discussed in Subsection 6.2.5.

1.2.3 Station

Description of Features Important to Power Generation This section provides an overview of thos e plant features which are important to the power generation objective for LaSalle County Station. The power conversion equipment has the primary function of co nverting the internal steam energy to electricity. The power conversion system (PCS), with its associated process control and instrumentation, and the connected radwaste treatment systems are all important to power generation. Additionally, certain auxiliary systems which support the power equipment are briefly discussed in the following subsections. Examples of such auxiliaries include: new and spent fuel storage facilities, the fuel pool cleanup system, the reactor make up water demineralizer, and the HVAC systems that condition the facilities in which these PCS functions take place.

1.2.3.1 Power Conversion System

The power conversion system actually includes five interrelated systems: the turbine-generator, the main steamlines with control valving, the main condenser, the circulating water system, and the cond ensate and feedwater system. A brief description follows for each constituent system.

1.2.3.1.1 Turbine-Generator

The turbine is an 1800-rpm, tandem-com pound, six-flow, reheat unit with an electrohydraulic governor for normal operation. The turbine-generator is provided with an emergency trip system for turbin e overspeed. The approximate rating of the turbine-generator is 1,183,300 kW at 3.5 in. Hg abs exhaust pressure.

LSCS-UFSAR 1.2-23 REV. 14, April 2002 The generator is a direct-driven, 3-ph ase, 60-Hz, 25,000-Volt, 1800-rpm, hydrogen inner-cooled, synchronous generator rate d at 1,300,300 kVA at 0.90 power factor, 0.58 short circuit ratio at a maximu m hydrogen pressure of 75 psig.

The turbine-generator is discussed in Section 10.2.

A turbine gland seal subsystem is provided to minimize air in-leakage or radioactive steam out-leakage. The subsystem consis ts of a steam evaporator, steam seal pressure regulator, steam seal header, two full-capacity gland seal steam condensers with the associated piping, valves, and instrumentation. This subsystem is further discusse d in Subsection 10.4.3.

A steam bypass subsystem is provided which passes steam directly to the main condenser under the control of the pressure regulator. Steam is bypassed to the condenser whenever the reactor steaming rate exceeds the turbine-generator load (such as during generator synchronization or following a large electrical load rejection). The capacity of the turbine steam bypass subsystem is 23.6% of the reactor rated steam flow. This subsystem is further discussed in Subsection 10.4.4.

1.2.3.1.2 Main Steamlines In the context of the power conversion sytem, the main steamlines consist of four 26-inch-diameter lines from the outermost main steamline isolation valves to the main turbine stop valves. The use of four main steamlines permits testing of the turbine stop valves and main steamline is olation valves during station operation without load reduction. The design pressure and temperature of the main steamlines from the outermost MSIV to the turbine valve is 1250 psig at 575°F. This component is further discussed in Subsection 5.4.9.

1.2.3.1.3 Main Condenser The main condenser is a single-shell, single-pass, deaerating-type condenser with a divided water box. The condenser includes provisions for accepting up to 23.6% of the main steam flow at design conditions from the turbine bypass system and serves as a heat sink for se veral other flows, such as exhaust steam from the feed pump turbines, cascading heater drains, and feedwater heater shell operating vents. This item is disc ussed in Subsection 10.4.1.

A main condenser evacuation subsystem is provided to remove noncondensable gases from the condenser, including air inleakage and radiolytic dissociation products originating in the reactor, and to exhaust them to the gaseous radwaste system. The subsystem consists of two 100%-capacity, twin-element, two-stage, steam jet air ejector (SJAE) units complete with intercondensers for normal plant operation and a mechanical vacuum pump for use during startup and shutdown. This subsystem is discusse d in Subsection 10.4.2.

LSCS-UFSAR 1.2-24 REV. 13 1.2.3.1.4 Circulating Water System The circulating water system provides the condenser with a continuous supply of cooling water. The circulating water system takes water from a man-made perched cooling lake. Makeup water to the lake is provided from the Illinois River.

1.2.3.1.5 Condensate and Feedwater System The condensate and feedwater system delivers condensate from the condenser hotwell to the reactor pressure vessel. Condensate is pumped by four condensate pumps (one spare) through the intercondense r of the steam jet air ejector, the off-gas condenser, and the gland steam condenser. After leaving the gland steam condenser, the condensate is pumped thro ugh a full-flow condensate demineralizer system. The demineralizer effluent is then pumped by four condensate booster pumps (one spare) through the low-pressure heaters. The heaters are split into three one-third capacity parallel streams each stream consisting of five low pressure heaters in series. The last low-pressure heater discharges to the suction of the reactor feedwater pumps. The discharg e from the two turbine-driven reactor feedwater pumps and/or the motor-driven feedwater pump passes through the sixth stage of feedwater heating and then to the reactor pressure vessel. Feedwater flow is controlled by varying the speed of the turbine-driven feedwater pump or the position of the regulating valves on the motor-driven reactor feedwater pump. This system is further discussed in Subsection 10.4.7.

The condensate demineralizer subsystem is discussed in Subsection 10.4.6.

1.2.3.2 Electrical Systems and Instrumentation Control This subsection provides a general overview of the electrical subsystems and of instrumentation and control. All safety systems are supplied with redundant power supplies. This subject is further discussed in Chapters 7.0 and 8.0.

1.2.3.2.1 Electrical Power System The plant consists of two main generator units designated as Unit 1 and Unit 2. Each main generator is directly connected to a main power transformer through an isolated phase electrical bus duct. The main power transformers transform the output of each generator from the generator voltage to a nominal 345-kV for transmission.

The output of each main power transformer is connected to a 345-kV switchyard consisting of circuit breakers, disconnect switches, buses, and associated equipment.

LSCS-UFSAR 1.2-25 REV. 13 Overhead 345-kV transmission lines distri bute power to vari ous points on the transmission network. This system is further discussed in Chapter 8.0.

1.2.3.2.2 Electrical Power System Process Control and Instrumentation

Main generator electrical controls are lo cated in the station control room. These include the main generator circuit breake r controls, the synchronizing equipment, the generator excitation and voltage control equipment, and the circuit breaker controls for all main supply circuits to the auxiliary power system.

High-speed protective relaying equipment is provided for the main generators, main and auxiliary transformers, main buses, transmission lines, and interconnecting cables and bus ducts so as to provide proper clearing of this equipment in the event of electrical faults. The protective relay system includes breaker failure protection and backup relaying to ensure proper clearing of electrical faults in the event of a failure of the primary protective relaying.

Instrumentation is provided in the main control room for the main generator equipment. This includes indicating instruments for voltage, current, Megawatt (MW), megavolt ampere reactive (MVAR), and frequency. Recording instruments are provided for generator-MW output. KWh meters are provided for main generator outputs and for auxiliary power system loads.

Instrumentation is also provided for monitoring the generator and transformer performance.

Control of transmission line circuit breakers is by remote action from the station control room.

Electrical instrumentation is discussed in Chapter 7.0.

1.2.3.2.3 Nuclear System Process Control and Instrumentation

1.2.3.2.3.1 Reactor Manual Control System The reactor manual control system provides the means by which control rods are positioned from the control room to regula te reactor power. The system operates valves in each hydraulic control unit to change control rod position. Only one control rod can be manipulated at a time. The reactor manual control system includes these hydromechanical blocks that restrict control rod movement under

certain conditions as a backup to procedural controls. This system is discussed in Subsection 7.7.2.

LSCS-UFSAR 1.2-26 REV. 13 1.2.3.2.3.2 Deleted 1.2.3.2.3.3 Recirculation Flow Control System The recirculation flow control system ad justs the variable-position flow control discharge valve. This changes the coolant flow rate through the core and thereby changes the core power level. The system automatically matches the reactor power output to the load demand. This system is discussed in Subsection 7.7.3.

1.2.3.2.3.4 Neutron Monitoring System The neutron monitoring system is a system of incore neutron detectors and out-of-core electronic monitoring equipment. Th e system provides indication of neutron flux, which can be correlated to thermal power level for the entire range of flux conditions that can exist in the core. The local power range monitors (LPRM's) and average power range monitors (APRM's) allo w assessment of local and overall flux conditions during power range operation. Automatic control rod blocks, based on input signals from the neutron monitoring system, prevent rod withdrawal beyond the point of limited local reactor power for the existing reactor coolant flow rate. The traversing incore probe (TIP) syst em provides a means to calibrate the individual LPRM sensors.

1.2.3.2.3.5 Reactor Vessel Instrumentation In addition to instrumentation for the nuclear safety systems and engineered safety features, instrumentation is provided to monitor and transmit information that can be used to assess conditions existing inside the reactor vessel and the physical condition of the vessel itself. This instrumentation monitors reactor vessel pressure, water level, coolant temperature, reactor core differential pressure, coolant flow rates, and reactor vessel head inner seal ring leakage. This topic is further discussed in Subsection 7.7.1.

1.2.3.2.3.6 Proces s Computer System An on-line process computer is provided for each unit to monitor and log process variables and to make certain analytical computations. This system is further discussed in Subsection 7.7.7.

1.2.3.3 Power Conversion Systems Process Control and Instrumentation

The power conversion systems are controll ed by the equipment described in the following. Instrumentation is provided to sense a need for a controlling action.

LSCS-UFSAR 1.2-27 REV. 14, APRIL 2002 1.2.3.3.1 Pressure Regulator and Turbine-Generator Control The pressure regulator and turbine-generator instrumentation is classified as non-safety-related. It includes the remote turbine-generator controls, a redundant electrical supply, computer-operated automatic controls, and bypass valves and lines to relieve reactor vessel pressure.

The pressure regulator maintains control of the turbine control valves and turbine bypass valves to enable proper generator and reactor response to system load demand changes while maintaining the nuclear system pressure essentially constant.

The turbine-generator speed-load controls act to maintain constant turbine speed (generator frequency) and to respond to load changes by adjusting the reactor recirculation flow controller and the pressure regulator operating points.

The turbine-generator speed-load controls ca n initiate rapid closure of the turbine control valves (rapid opening of the turbine bypass valves) to prevent turbine overspeed upon loss of generator electric load. This is necessary to compensate for the delay of the nuclear boiler to respond to turbine-generator load fluctuations.

This item is discussed furt her in Subsections 7.7.5.

1.2.3.3.2 Feedwater System Control

A three-element controller is used to regulate the feedwater system so that proper water level is maintained in the reactor vessel. The controller us es main steam flow rate, feedwater flow rate, and reactor water level error signals. The feedwater control signal maintains a programmed level by varying the speed of the turbine-driven feedwater pumps and/or by varying the flow control valve position on the discharge of the constant speed motor-driven feedwater pump. Alternatively, operation in single element control is available.

1.2.3.4 Radioactive Waste Systems The radioactive waste systems provide a means to monitor, remove, treat, and dispose of radioactive wastes in a manner consistent with the applicable sections of 10 CFR 20 and 10 CFR 50, Appendix I.

1.2.3.4.1 Gaseous Radwaste System

Each unit has a completely independent gaseous radwaste system discharging to the station vent stack. All radioactive ga seous effluents are controlled and released via the station vent stack.

LSCS-UFSAR 1.2-28 REV. 13 The diffusion and dispersal characteristics of the station vent stack enable release without processing of low-level effluents.

Steam for the turbine gland sealing system is provided by an auxiliary steam seal evaporator. Because clean water is used as feedwater to the evaporator, the expected release of radionuclides from the gland seal condenser is expected to be minimal. The system is fully described in Subsection 10.4.3.

The main condenser is the largest volumetric source of gaseous radioactive effluent. Treatment of these gases includes high-temperature catalytic recombining, holdup for decay, high-efficiency particulate filtra tion, and charcoal adsorption. Effluent monitoring is provided to ensure that the released activity is well within federal limits. This system is discussed in Section 11.3.

1.2.3.4.2 Liquid Radwaste System

This system collects, treats, stores, and disposes of or recycles all radioactive liquid wastes. Liquid wastes are accumulated in sumps and drain tanks at various locations throughout the plant and are then transferred to collection tanks in the radwaste facility for subsequent treatmen t, storage, and transport to the solid radwaste system for ultimate disposal. Wa stes are processed on a batch basis, with each batch being processed by methods appropriate for the particulate type and quantity of isotopic materials present.

Processed liquid wastes are routed to the cycled condensate system or by the waste discharge piping to the river. The liquid wastes in the discharge piping are sufficiently diluted with cooling lake water to achieve a concentration for discharge into the Illinois River well within state and federal concentration limits. A design dilu tion factor of approximately 670 prior to discharge to the river is typical.

Radwaste equipment is selected, arranged , and shielded to permit operation, inspection, and maintenance with minimum personnel exposure. Processing equipment is selected and designed to require a minimum of maintenance.

Protection against accidental discharge of liquid radioactive waste is provided by instrument redundancy, for detection and alarm of abnormal conditions, and by procedural controls.

This system is discussed in Section 11.2.

LSCS-UFSAR 1.2-29 REV. 13 1.2.3.4.3 Solid Radwaste System Solid radioactive wastes are collected, proc essed, and packaged for storage. These wastes are generally stored on the site until the isotopes with short half-lives have decayed. Ultimately, the waste is lo aded and shipped to a burial site.

The solid radwaste system is designed to maintain radiation exposures to personnel "as low as reasonably achievable" during system operation and maintenance.

This system is discussed in Section 11.4.

1.2.3.5 Radiation Monitoring and Control Radiation monitoring systems are provided to monitor and control radioactivity in process and effluent streams and to activate appropriate alarms and controls.

A process radiation monitoring system is provided for indication and recording radiation levels associated with plant process streams and effluent paths leading to the environment. All effluents from the plant which are potentially radioactive are monitored.

Process radiation monitoring is also discussed in Sections 9.3 and 11.5.

1.2.3.6 Miscellaneous Power Generation

1.2.3.6.1 New and Spent Fuel Storage New and spent fuel storage racks are designed to prevent inadvertent criticality and load buckling. Sufficient coolant an d shielding are maintained to prevent overheating and excessive personnel exposure, respectively. The design of the fuel pool provides for corrosion resistance, adherence to Seismic Category I requirements, and prevention of keff from reaching 0.95 under flooded conditions.

The new fuel vault design prevents keff from reaching 0.90 under dry conditions, and 0.95 under flooded conditions. This subject is further discussed in Section 9.1.

1.2.3.6.2 Fuel Pool Cleanup System The fuel pool cooling and cleanup subsystem provides the removal of decay heat from stored spent fuel and maintains specified water temperature, purity, clarity, and level. This prevents spent fuel overheat and the buildup of excessive radioactive materials in the cooling water, thereby minimizing possible exposures to plant personnel.

LSCS-UFSAR 1.2-30 REV. 15, APRIL 2004 1.2.3.6.3 Service Water System The normal service water system supplies cooling water for turbine-generator and miscellaneous HVAC loads, fuel pool cooling, and the heat exchangers in the turbine building and reactor building closed cooling water systems. Service water for traveling screen wash and fire protection is also provided by this system. Gland water to the circulating water pumps is also provided by this system. This system is further discussed in Subsection 9.2.2.

1.2.3.6.4 Demineralized Water Makeup System

The demineralized water makeup system is abandoned-in-place and has been replaced with a vendor trailer. The demineralized water makeup system for LSCS, Units 1 and 2, provides demineralized water for plant usage. The system consists of a vendor trailer, which is capable of producing 72,000 gallons of demineralized water per day. A detailed discussion of the demineralized water makeup system is in Subsection 9.2.4.

1.2.3.6.5 Station HVAC The ventilation for the radwaste building is provided by a once-through system which uses evaporative coolers. Evaporative coolers (abandoned-in-place) are shutdown and controlled administratively.

Exhaust air is filtered through HEPA filters en route to the station vent stack. The station vent stack has a full-time stack monitoring system for radioactivity.

The HVAC systems are described in Section 9.4.

1.2.3.6.6 Heating, Ventilating, an d Air-Conditioning (HVAC) Systems Separate HVAC systems exist for the contro l room, the auxiliary electric equipment room and the rooms standby diesel gene rators. These HVAC systems, the CSCS equipment area coolers, and the switchgear heat-removal systems are designed to operate under all station conditions.

The CSCS equipment area cooling system co nsists of four water cooled air blowers for each primary containment that supplies cool air to respective CSCS pump cubicles.

All air distribution systems are designed so that airflow is directed from areas of lower contamination to areas of progressively higher potential contamination.

LSCS-UFSAR 1.2-31 REV. 13

1.2.4 Glossary

1.2.4.1 Definitions

The following definitions apply to the terms used in the LaSalle County Station, Units 1 and 2, Updated Final Safety Analysis Report:

Accident -- A single event, not reasonably expected during the course of station operation, that has been hypothesized fo r analysis purposes or postulated from unlikely but possible situations, and that causes or threatens a rupture of a radioactive material barrier.

Active Component

-- A safety related component ch aracterized by an automatically initiated change of state or discernible mech anical action in response to an imposed demand. Active Failure -- The failure of an active component to perform its function when called upon to do so by an initiating signal.

Administrative Controls -- The provisions relating to organization and management, personnel function procedures, recordkeeping, review and audit, and reporting necessary to ensure responsible operation of the facility.

Anticipated Operational Occurrences -- Those abnormal conditions of operation that are expected to occur one or more times during the life of the nuclear power unit, whose consequences do not affect safety.

Auxiliary Building -- A Seismic Category I building adjacent to the secondary containment (reactor building).

Availability -- The probability that a component will be operable when called upon to perform its specified function.

Available Reactor Power -- The steam power available for the turbine and other heat cycle equipment.

Boiling Length -- In a heated fuel bundle, the length that is producing net steam generation.

Channel -- An arrangement of one or more sensors and associated components used to evaluate station variables and produce discrete outputs used in logic. A channel terminates and loses its identity where in dividual channel outputs are combined in logic.

LSCS-UFSAR 1.2-32 REV. 14, APRIL 2002 Cold Shutdown -- The condition of the reactor when the reactor is shut down; the reactor coolant is maintained at less than 212° F, and the reactor vessel is near atmospheric pressure.

Components -- Items from which a functional system is assembled.

Design Basis -- That information which identifies the specific functions to be performed by a structure, syst em, or component, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design.

Design-Basis Accident -- A hypothesized accident, the characteristics and radiological consequences of which are utilized in the design of those systems and components pertinent to the preservation of radioactive material barriers.

Design Power -- Refers to the power level at wh ich the reactor is producing 102% of reactor vessel rated steam flow.

Diesel-Generator Building -- That Seismic Category I building which houses the standby diesel generator systems.

Drywell --A pressure and radioactive material barrier, surrounding the reactor vessel and its recirculation loops, that conveys steam resulting from a postulated LOCA to the suppression pool for condensation.

Emergency Core Cooling Systems -- The systems which furnish cooling water to the core to compensate for a loss of normal co oling capability during the postulated loss-of-coolant accidents.

Engineered Safety Features -- Systems provided to mi tigate the consequences of postulated accidents.

Excursion -- A sudden, very rapid rise in the reactor power level.

Functional Test -- The intentional operation or init iation of a system, subsystem, or component to verify that it operates within design tolerances.

Hot Shutdown -- The reactor condition when the mode switch is in the shutdown position and the reactor coolant temperature is greater than 212° F.

Hot Standby Mode -- The condition of the reactor when it is operating with the coolant temperature greater than 212° F, the system pressure less than 1060 psig, and the mode switch in the startup position.

Logic -- That array of components which combines individual bistable output signals to produce decision outputs.

LSCS-UFSAR 1.2-33 REV. 13 Loss-of-Coolant Accidents -- Those postulated accidents that result from the loss of reactor coolant at a rate in excess of the capability of the reactor coolant makeup system, and from breaks in the reactor coolant pressure boundary, up to and including a break equivalent in size to the double-ended rupture of the largest pipe of the reactor coolant system.

Minimum Critical Power Ratio (MCPR) -- The lowest ratio of that power which results in onset of transition boiling to the actual bundle power at the same location.

Module -- Any assembly of interconnected components that constitutes an identifiable device, instrument, or piece of equipment.

Nuclear Power Unit -- A nuclear power reactor and associated equipment necessary for electric power generation, including those structures, systems, and components required to provide reasonable assurance th at the facility can be operated without undue risk to the health and safety of the public.

Nuclear Steam Supply System (NSSS) -- A contractual term which designates those components of the nuclear power system and their related engineered safety features and instrumentation furnished by the nuclear steam supply system supplier (GE).

Operator Error -- An active deviation from writ ten operating procedures or nuclear station standard operating practices.

Passive Component -- A safety related component characterized by no change of state nor mechanical motion.

Passive Failure -- Loss of function of a passive component.

Power Generation Design Basis -- The unique design requirements that establish the power generation objective.

Power Generation Evaluation -- A comparison to show how the system satisfies the power generation design bases.

Power Generation System -- Any system not essential to safety, but essential to power generation.

Power Operation -- A time reference wh ich begins where "heatup" ends and includes continued operation of the station at power levels in excess of heatup power.

LSCS-UFSAR 1.2-34 REV. 13 Primary Containment -- The drywell in which the reactor vessel is located, the pressure suppression chamber, and the process lines out to the second isolation valve.

Rated Reactor Power -- Refers to the power level at which the reactor is producing 100% steam flow.

Reactor Building -- The Seismic Category I structure comprising the secondary containment.

Reactor Isolated -- A condition wherein the reactor is isolated from the condenser.

Reactor Mode Switch Positions -- Four modes of reactor operation for which switch positions are available as follows:

a. Shutdown Mode -- Condition of the reacto r when it is shut down, the reactor mode switch is in the shutdown mode position, and all operable control rods are fully inserted.

b. Startup Mode -- Condition of the reactor when the reactor mode switch is in the startup mode position.

c. Run Mode -- Condition of the reactor when the reactor mode switch is in the run mode position.

d. Refuel Mode -- Condition of the reactor when the reactor mode switch is in the refuel mode position.

Safety Design Basis -- The unique design requirements that establish the safety objective.

Safety Evaluation -- A comparison to show how the system satisfies the safety design basis.

Safety Related -- Those structures, and equipment necessary to maintain the integrity of the reactor coolant pressure boundary, to shut down the reactor and maintain it in a safe shutdown conditio n, and/or to prevent or mitigate the consequences of accidents.

Scram -- The simultaneous rapid insertion of all control rods into the core.

Secondary Containment or Reactor Building -- A Seismic Category I building that completely encloses the primary containment.

Sensor -- That part of a channel used to monitor a measurable power plant variable.

LSCS-UFSAR 1.2-35 REV. 13 Setpoint -- That value of a monitored plant va riable that results in a channel trip when the monitored variable reaches or exceeds this value.

Shutdown -- The reactor condition when the effective multiplication factor is sufficiently less than 1.0 such that the wi thdrawal of any one control rod could not produce criticality under the most restrictiv e potential conditions of temperature, pressure, burnup, and fission-product concentration.

Single Failure -- An occurrence that results in the loss of capab ility of a safety related component to perform its intended safety fuctions.

Source Material -- Uranium or thorium or any combination thereof, in any physical or chemical form; or ores which contain by weight one-twenti eth of one percent (0.05%) or more of uranium, thorium, or any combination thereof. Source material does not include special nuclear material.

Special Nuclear Material -- Plutonium, uranium-233, uranium enriched in the isotope 235, and any other material that the NRC, pursuant to the provisions of Section 51 of the Atomic Energy Act of 1954, as amended, determines to be special nuclear material, or any material artificia lly enriched by any of the foregoing. Special nuclear material does not include source material.

Standby Gas Treatment System (SGTS) -- An engineered safety system in the reactor building that processes leakage from the primary containment and discharges it after treatment to the atmosphere.

Station Vent Stack -- The common exhaust providin g elevated release (370 feet above grade) for all gaseous effluents and plant ventilation air.

Suppression Pool -- A pool of water, located in the suppression chamber under the drywell, which normally provides th e water seal between the drywell and containment.

Technical Specifications -- A set of detailed technical requirements and limits which establish the operational envelope for the plant, based on safety considerations.

Test Interval -- The elapsed time between the initia tion of sequential identical tests.

Trip -- The change of state of a bistable device from a normal condition.

Turbine Cycle Rated Power -- Rated power available for the turbine.

Unit -- A nuclear steam supply system, turbin e-generator, and supporting facilities.

LSCS-UFSAR 1.2-36 REV. 15, APRIL 2004 1.2.4.2 ACRONYMS USED IN LSCS-UFSAR ADS Automatic Depressurization System AEER Auxiliary Electric Equipment Room APRM Average Power Range Monitor ARI Alternate Rod Insertion ARM Area Radiation Monitor ATWS Anticipated Transients Without Scram BWR Boiling Water Reactor CCW Closed Cooling Water ComEd Commonwealth Edison Company CECo Commonwealth Edison Company CHF Critical Heat Flux CRD Control Rod Drive CRPI Control Rod Position Indication CSCS-ECWS Core Standby Cooling System - Equipment Cooling Water System DBA Design-Basis Accident DG Diesel Engine-Generator DIB Digital Isolation Block ECCS Emergency Core Cooling Systems EFCV Excess Flow Check Valve EGC Exelon Generation Company, LLC EHC Electrohydraulic Control ESF Engineered Safety Feature FA Full Arc (mode of TCV operation) FLECHT Full-Length Emergency Cooling Heat Transfer FPCC Fuel Pool Cooling and Cleanup FSAR Final Safety Analysis Report GE General Electric Company HCU Hydraulic Control Unit HEPA High-Efficiency Particulate Air/

Absolute (referring to filters) HPCS High-Pressure Core Spray HX Heat Exchanger H&V Heating and Ventilating HVAC Heating, Ventilating, and Air-Conditioning IGSCC Intergranular Stress Corrosion Cracking HWC Hydrogen Water Chemistry IAC Interim Acceptance Criteria (NRC)

IRM Intermediate Range Monitor IRSF Interim Radwaste Storage Facility LCO Limiting Condition of Operation LDS Leak-Detection System LOCA Loss-of-Coolant Accident LPCS Low-Pressure Core Spray LPRM Local Power Range Monitor LRCP Liquid Radwaste Control Panel LSCS LaSalle County Station LSSS Limiting Safety System Setting LSCS-UFSAR 1.2-37 REV. 15, APRIL 2004 LPZ Low Population Zone M/A Manual/Auto MCC Motor Control Center MCPR Minimum Critical Power Ratio MDRFP Motor Driven Reactor Feed Pump MG Motor-Generator Set MLD Mean Low Water Datum MSL Mean Sea Level MSIV Main Steam Isolation Valve MSIV-ICLTM Main Steam Isolation Valve Isolated Condenser Leakage Treatment Method MSIV-LCS Main Steam Isolation Valve Leakage Control System NB Nuclear Boiler NBR Nuclear Boiler Rated (power)

NED Nuclear Energy Division (GE) NMS Neutron-Monitoring System NSSS Nuclear Steam Supply System NSSSS Nuclear Steam Supply System Shutoff NSOA Nuclear Safety Operational Analysis OBE Operating Basis Earthquake OPRM Oscillation Power Range Monitor PA Public Address (System) PMF Probable Maximum Flood PMP Probable Maximum Precipitation P&ID Piping and Instrumentation Diagram PRM Power Range Monitor PSAR Preliminary Safety Analysis Report PCS Process Computer System RBM Rod Block Monitor RCPB Reactor Coolant Pressure Boundary RCIC Reactor Core Isolation Cooling RHR Residual Heat Removal RMC Reactor Manual Control RPS Reactor Protection System RPV Reactor Pressure Vessel RWCU Reactor Water Cleanup RWM Rod Worth Minimizer SAR Safety Analysis Report SGTS Standby Gas Treatment System SJAE Steam Jet Air Ejector S&L Sargent & Lundy SLC Standby Liquid Control SPC Siemens Power Corporation, Nuclear Division SPF Standard Project Flood SPS Standard Project Storm SRM Source Range Monitor SRV Safety/Relief Valve SSE Safe Shutdown Earthquake LSCS-UFSAR 1.2-38 REV. 15, APRIL 2004 SW Service Water TBCCW Turbine Building Closed Cooling Water TCV Turbine Control Valve TDRFP Turbine Driven Reactor Feed Pump TG Turbine-Generator TIP Traversing Incore Probe URC Ultrasonic Resin Cleaner

1.2.5 References

1. LaSalle County Station, "Mark II - Design Assessment Report," Commonwealth Edison Company, February 1976.

LSCS-UFSAR 1.3-1 REV. 13 1.3 COMPARISON TABLES

1.3.1 Comparison

with Similar Facility Designs

A comparison of the principal design features of the LaSalle County Station (LSCS) with those of other boiling water reactor facilities was included in Section 1.3 of the FSAR which compared LSCS with Zimmer 1, Washington

Public Power Supply System (WPPSS) 2, and Hatch 1, listing the design

characteristics of the following:

a. nuclear steam supply,

b. power conversion, c. engineered safety features,

d. containment,

e. radioactive waste management,

f. structural,

g. instrumentation and electrical, and

h. standby gas treatment.

This information was current at the time the LSCS Unit 1 operating license (OL) was granted and has not since been revised.

1.3.2 Comparison

of Final and Preliminary Information Table 1.3-9 of the FSAR provided a list of significant differences between the final and preliminary designs of the LaSalle County Station. This information

was current at the time the LSCS Unit 1 OL was granted and has not since

been revised.

LSCS-UFSAR 1.4-1 REV. 13 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS The identification of the principal agents and contractors involved in the design

and construction of the LaSalle County Station is included in Section 1.4 of the FSAR. This information was current at the time the LSCS-1 operating license was granted and has not since been revised.

LSCS-UFSAR 1.5-1 REV. 13 1.5 REQUIREMENTS FOR OTHER TECHNICAL INFORMATION - CURRENT CONCERNS FROM LSCS ACRS LETTER

The concerns of the Advisory Committee on Reactor Safeguards pertaining to

LSCS at the time the LSCS-1 OL was granted were addressed in Section 1.5 of the FSAR. Modifications made in response to those concerns have been identified and were documented in amendments to the FSAR.

LSCS-UFSAR 1.6-1 REV. 14, APRIL 2002 1.6 MATERIAL INCORPORATED BY REFERENCE Table 1.6-1 of the FSAR provided a list of all GE topical reports and any other

report or document which was incorporated in whole or in part by reference in the FSAR and had been previously filed with the NRC. Topical reports and other documents incorporated by reference in the FSAR and in annual UFSAR revisions are included in the reference sections of the applicable chapters in

the UFSAR.

Additional documents were incorporated into the UFSAR by reference in the appropriate sections when nuclear reload fuel fabricated by SPC was

introduced into the reactor cores with technical support shared by SPC and Commonwealth Edison (ComEd). Technical Requirements Manual also contains pertinent SPC licensing topical reports.

Additional documents were incorporated into the UFSAR by reference in the appropriate sections when the UFSAR was revised due to NRC approval of the Power Uprate License Amendment.