ML21133A080

0 to Updated Final Safety Analysis Report, Chapter 1, Sections 1.1 Though 1.6

ML21133A080
Person / Time
Site:	Limerick
Issue date:	04/29/2021
From:	Exelon Generation Co
To:	Office of Nuclear Reactor Regulation
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LGS UFSAR CHAPTER 1 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

1.1 INTRODUCTION

This UFSAR is submitted to fulfill the requirements of 10CFR50.71(e) for Exelons nuclear power station designated as the Limerick Generating Station, Units 1 and 2. These units hold full power utilization facility (Class 103) licenses, NPF-39 and NPF-85. These licenses were issued in August 1985 and August 1989 respectively.

LGS is located on the east bank of the Schuylkill River in Limerick Township of Montgomery County, Pennsylvania, approximately 4 river miles downriver from Pottstown, 35 river miles upriver from Philadelphia, and 49 river miles above the confluence of the Schuylkill with the Delaware River. The site contains 595 acres (423 acres in Montgomery County and 172 in Chester County).

Each of the LGS units employs a GE BWR originally designed and licensed to operate at a rated core thermal power of 3293 MWt (100% steam flow) with a corresponding gross electrical output of 1092 MWe. Approximately 37 MWe are used for auxiliary power. Subsequent to issuing the original operating licenses, LGS Units 1 and 2 were reevaluated with regard to rerating power to 3458 MWt (Rerate Power). The acceptability of the rerate evaluations stems from the fact that LGS Units 1 and 2 were originally designated for steam flow capabilities at least 5% above its original rating. In addition, improvements in the analytical techniques based on more realistic assumptions, plant performance feedback, and the latest fuel designs resulted in a significant increase in the calculated operational margins related to safety analyses. Subsequently, the core thermal power was uprated from 3458 MWt to 3515 MWt as part of the measurement uncertainty recapture power uprate (MUR PU). The existing analyses assume a 2% uncertainty in feedwater flow, which is reduced to 0.3% using leading edge ultrasonic flow meters. This is acceptable because the analyzed conditions and margins of the safety analyses are not affected by this change in FW flow measurement.

The reactor design power level used in various analyses is discussed in Section 6.3 and Chapter 15.

The containment system, designed by Bechtel Power Corporation, limits the release of radioactive materials to the environs subsequent to the occurrence of a postulated LOCA so that the offsite doses are below the values stated in 10CFR50.67. The design employs the drywell/pressure-suppression features of the BWR/Mark II containment concept. The containment consists of a dual barrier: the primary containment, and the secondary containment. The primary containment is a steel-lined reinforced concrete pressure-suppression system of the over-and-under configuration. The secondary containment is the enclosure that encloses the reactor, and its primary containment, and fuel storage areas.

Condenser cooling is provided by water circulated through natural draft cooling towers.

The LGS PSAR was submitted on February 26, 1970 (AEC Dockets 50-352 and 50-353). The construction permits, CPPR-106 and CPPR-107, were originally issued on June 19, 1974.

Environmental impact is discussed in the Applicant's Environmental Report - Construction Permit Stage (Revised) dated May 1972. The Atomic Energy Commission (now Nuclear Regulatory Commission) issued the LGS Final Environmental Statement Construction Stage in November 1973.

CHAPTER 01 1.1-1 REV. 16, SEPTEMBER 2012

LGS UFSAR The LGS FSAR was originally submitted on March 17, 1981 and docketed on July 27, 1981. The NRC issued its SER in August 1983 and nine supplements to NUREG-0991 through August 1989.

A discussion of environmental impact was also submitted as the EROL, and the NRC issued its LGS Final Environmental Statement Operating License Stage, NUREG-0974, in April 1984.

The LGS Unit 1 low power (5%) operating license (NPF-27) was issued on October 26, 1984 and the full power operating license (NPF-39) was issued on August 8, 1985. The LGS Unit 1 entered commercial operation on February 1, 1986.

The LGS Unit 2 was issued an operating license (NPF-83) to load fuel and conduct testing up to, but not including, initial criticality on June 22, 1989. The low power (5%) operating license (NPF-

84) was issued July 10, 1989 and the full power operating license (NPF-85) was issued August 25, 1989. The LGS Unit 2 entered commercial operation on January 8, 1990. The licensee received a license amendment for a 5% increase in rated power to 3458 MWt for LGS Unit 2 on February 16, 1995 and for Unit 1 on January 24, 1996. The licensee received license amendments for an additional 1.65% increase in rated power to 3515 MWt for Unit 1 and Unit 2 on April 8, 2011.

The FSAR was prepared and submitted in accordance with 10CFR50.34(b). Its format and content are in accordance with Regulatory Guide 1.70 (Rev 3), "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants", November, 1978. Following its initial update, the UFSAR will be updated in accordance with 10CFR50.71.

The relationships of other various reports submitted by the licensee are discussed below.

As discussed in Appendix 9A, the Fire Protection Evaluation Report was prepared and submitted in response to a September 30, 1976 NRC request. Since it contains the information suggested by Regulatory Guide 1.70 to be provided in Section 9.5.1.3, the FPER, in its entirety, was incorporated into the UFSAR as Appendix 9A in accordance with 10CFR50.32. The FPER will be updated in accordance with 10CFR50.71.

As discussed in Appendix 3A, the Design Assessment Report was prepared and submitted in response to a generic industry concern of thermohydrodynamic loads resulting from SRV operations and/or discharges during a LOCA. The DAR contains considerable design information and will be updated in accordance with 10CFR50.71. Proprietary DAR information is discussed in Appendix 3B.

The Quality Assurance Program was prepared and submitted in accordance with 10CFR50.34(b).

Its format and content were prepared in accordance with Regulatory Guide 1.70 (Rev 3). The Quality Assurance Program is incorporated into the UFSAR as Section 17.2 and will be updated in accordance with 10CFR50.54(a)(3) and 10CFR50.71.

In accordance with 10CFR50.34(b)(6)(v) and Regulatory Guide 1.70, the Emergency Plan, while a separate report, is referenced in Section 13.3 of the UFSAR. The Emergency Plan will be updated in accordance with 10CFR50.54(q) and 10CFR50, Appendix E.

The Security Plan was prepared and submitted in accordance with 10CFR50.34(c). It is identified as a separate report withheld from public disclosure in UFSAR Section 13.6, in accordance with Regulatory Guide 1.70 (Rev 3). The Security Plan will be updated in accordance with 10CFR50.54(p).

CHAPTER 01 1.1-2 REV. 16, SEPTEMBER 2012

LGS UFSAR The Safeguards Contingency Plan was prepared and submitted in accordance with 10CFR50.34(c), and withheld from public disclosure pursuant to 10CFR2.790(d). No discussion of this report is provided in the UFSAR, consistent with Regulatory Guide 1.70 (Rev 3). The Safeguards Contingency Plan will be updated in accordance with 10CFR50.54(p).

As discussed in Section 1.0 of the Environmental Qualification Report, the EQR was prepared and submitted in response to a May 1980 Commission Memorandum and Order (CLI 80-21). It has been revised to fulfill the requirements of 10CFR50.49. The EQR is referenced in Section 3.11 of the UFSAR but is considered to be a licensee working document and update submittals are not planned.

The EROL was prepared and submitted in accordance with 10CFR51.21. Its format and content are in accordance with Regulatory Guide 4.2 (Rev 2), "Preparation of Environmental Reports for Nuclear Power Plants", July 1976. Regulatory Guide 4.2, section 7.1, "Station Accidents Involving Radioactivity" was later superseded by the Commission's Interim Position on Accident Consideration Under NEPA (45FR40101).

As discussed in section 1.1 of the Probabilistic Risk Assessment (PRA) and the Severe Accident Risk Assessment (SARA), the PRA and SARA were prepared and submitted in response to an initial May 6, 1980 NRC request. Because they contain the information required by the NRC's Interim Position on Accident Considerations under NEPA (45FR40101), the PRA and SARA were used in the analysis performed for EROL Section 7.1, in accordance with 10CFR50.32. At the request of the NRC Limerick Project Manager, the PRA is acknowledged in UFSAR Section 15.11; it is not referenced therein. The PRA and SARA are considered to be licensee working documents and update submittals are not planned.

CHAPTER 01 1.1-3 REV. 16, SEPTEMBER 2012

LGS UFSAR 1.2 GENERAL PLANT DESCRIPTION 1.2.1 SITE CHARACTERISTICS A summary of the site characteristics for LGS is provided below. Detailed discussions on the site characteristics are provided in Chapter 2 of the UFSAR.

1.2.1.1 Location LGS is located in southeastern Pennsylvania on the Schuylkill River about 1.7 miles southeast of the limits of the Borough of Pottstown and about 20.7 miles northwest of the Philadelphia city limits. The Schuylkill River passes through the site and separates the western portion, which is located in East Coventry Township, Chester County, from the eastern portion, which is partly in Limerick Township and partly in Lower Pottsgrove Township, both in Montgomery County, Pennsylvania. All of the major plant structures are located in Limerick Township. The site location map is shown in Figure 2.1-1.

1.2.1.2 Site Environs and Access The site is located in gently rolling countryside, traversed by numerous valleys containing small creeks or streams that empty into the Schuylkill River. Two parallel streams, Possum Hollow Run and Brooke Evans Creek, cut through the site in wooded valleys, running southwest into the Schuylkill River.

The area surrounding the site can be generally classified as rural and open. A large portion of the land is used for agricultural purposes with the remainder of the area being either vacant or woodland with scattered residences.

The main access to the plant is from U.S. Highway 422 which runs east and west about one mile north of the site. Access to the site and all activities thereon are under the control of the licensee.

1.2.1.3 Geology The site is situated in the Triassic Lowland section of the Piedmont Physiographic Province. This section is characterized by a gently rolling land surface formed on an eroded low plateau.

The rocks in the region surrounding the site include Precambrian and Lower Paleozoic crystalline rocks and folded sedimentary strata, and essentially unfolded Triassic sedimentary rocks and igneous intrusions. The Triassic rocks belong to the Newark Group which is divided into the basal Stockton Formation and the Brunswick, Lockatong, and Hammer Creek Lithofacies.

Bedrock at the site underlies a thin cover of residual soil. The Brunswick red siltstone, sandstone, and shale is the predominant bedrock formation. Gray shale and argillite of the Lockatong Lithofacies, light gray sandstones and conglomerates of the Hammer Creek Lithofacies, and intruded diabase and associated hornfels are also found in the area. The strata exhibit gentle homoclinal dips to the north and northwest. The thickness of the Newark Group overlying the Paleozoic and Precambrian basement rocks at the site is on the order of 8000 feet.

The dominant structural feature of the region is the Regional Appalachian Orogenic Belt. This belt is marked by the northeast- southwest orientation of the axes and lineation of most of the structural features and stratigraphic contacts.

CHAPTER 01 1.2-1 REV. 20, SEPTEMBER 2020

LGS UFSAR 1.2.1.4 Seismology The seismicity of the site was evaluated on the basis of historical earthquakes, damage resulting from these shocks, and the regional and local geologic structure. The site lies in a region that has experienced a moderate amount of earthquake activity. Most of the reported earthquakes have occurred in the Piedmont province. Some minor shocks have occurred in a northeast-southwest trend along the Fall Zone, the physiographic boundary between the Piedmont and the Coastal Plain to the southeast. Some scattered activity has occurred in the Coastal Plain. No recent faulting has been mapped in the area of the site.

Based on the seismic history and the geologic structure of the region, no significant earthquake ground motion is expected at the site during the life of the proposed facility. Despite this demonstrative evidence to the contrary, the remote possibility that the geologic structure in the epicentral areas of the significant regional earthquakes could exist in the site area is considered.

Thus, a design basis earthquake is hypothesized equivalent to the 1871 Wilmington, Delaware earthquake (Intensity VII) near the site. Such an event is highly improbable and this hypothesis is very conservative.

The plant has the capability for safe shutdown if subjected to a peak horizontal ground acceleration at foundation level of 0.15 g. Seismic Category I structures are founded on the competent siltstone and sandstone bedrock of the Brunswick Lithofacies. Design spectra consistent with the safe shutdown earthquake are used for the dynamic analysis of the seismic Category I structures and equipment.

1.2.1.5 Hydrology In the site area, Triassic-age siltstone, sandstone, and shale are found at shallow depths beneath a thin cover of residual soils. The residual soils are relatively impermeable. Most groundwater in the area is found in joints, fractures, and other secondary openings in the rock. The groundwater table is found at relatively shallow depths, except in the vicinity of pumping wells. Because of the limited quantities of available groundwater, surface water is the primary source of supply in the region. Groundwater accounts for only about 3% of the total industrial and commercial use in the region. However, numerous domestic wells extract small quantities of water from the Triassic strata.

A number of wells are located in the general site area. However, the geologic, hydrologic, and topographic conditions are such that the possibility of any adverse effect on these wells by operation of LGS is extremely remote.

The flows of the Schuylkill vary widely at different points along the river. This is mainly due to the varying topography and climatological and geohydrologic conditions along its course. The extreme and average daily flows as recorded at Pottstown gauging station (about 5.5 river miles upstream from the site) are:

Flow (cfs) Date Minimum 87 August 13, 1930 (instantaneous)

Average 1793 October 1926-September 1969 (daily)

Maximum 95,900 June 1972 (instantaneous)

CHAPTER 01 1.2-2 REV. 20, SEPTEMBER 2020

LGS UFSAR The probable maximum flood peak and stage at the plant site are estimated to be 500,000 cfs and el 174' (MSL), respectively.

The principal uses of the Schuylkill River are for municipal and industrial water supply. The river is also used for recreational fishing and boating.

1.2.1.6 Meteorology The general climate of the site is best described as humid continental. The region is dominated by continental air masses in winter, and by alternating continental and maritime tropical air masses in the summer. The site is near the track of most eastwardly moving low pressure systems that are brought from the interior of the U.S. by the prevailing westerlies. Annual average wind speeds in the region are between 9 and 10 mph and temperatures rarely exceed 100?F or drop below 0?F. The region receives a moderate amount of precipitation which is well distributed over the year.

Five years of meteorological data collected on the site have verified that the general regional conditions do exist at the site and that no unusual meteorological conditions prevail.

1.2.2 PRINCIPAL DESIGN CRITERIA The principal criteria for design, construction, and testing of LGS are summarized below. Specific criteria, codes, and standards are addressed in Sections 3.1 and 3.2.

1.2.2.1 General Design Criteria The LGS design conforms to the requirements given in 10CFR50, Appendix A, "General Design Criteria for Nuclear Power Plants." Specific compliance is discussed in Section 3.1.

a. The plant is designed, fabricated, erected, and operated to produce electrical power in a safe and reliable manner.

b. The plant is designed, fabricated, erected, and operated in such a way that the release of radioactive materials 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 for abnormal transients and accidents. Safety-related systems are designed to permit safe plant operation and to accommodate postulated accidents without endangering the health and safety of the public.

1.2.2.2 System Design Criteria 1.2.2.2.1 Nuclear System Criteria

a. The fuel cladding is designed to retain integrity, so that any failures are within acceptable limits, as a radioactive material barrier for the design power range and for any abnormal transient.

CHAPTER 01 1.2-3 REV. 20, SEPTEMBER 2020

LGS UFSAR

b. Those portions of the nuclear system that form part of the nuclear system process barrier are designed to retain integrity as a radioactive material barrier during normal operation and following abnormal operational transients and accidents.

c. Heat removal systems are provided to remove heat generated in the reactor core for the full range of normal operational conditions from shutdown to design power, and for any abnormal operational transient. Heat removal systems are provided to remove decay heat generated in the core under remote circumstances where the normal operational heat removal systems become inoperative. The capacity of such systems is adequate to prevent fuel cladding damage.

d. The reactor core and reactivity control systems 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. Backup reactor shutdown capability is provided independent of normal reactivity control provisions. This backup system has the capability to shut down the reactor from any operating condition, and subsequently to maintain the shutdown condition.

f. The nuclear system is designed so there is no tendency for divergent oscillation of any operating characteristics through hardware and administrative controls, considering the interaction of the nuclear system with other appropriate plant systems.

g. The reactor core is designed so that its nuclear characteristics do not contribute to a divergent power transient through hardware and administrative controls .

1.2.2.2.2 Safety-Related Systems Criteria 1.2.2.2.2.1 General

a. Safety systems act in response to abnormal operational transients so that fuel cladding retains its integrity as a radioactive material barrier to keep any failures within acceptable limits.

b. Safety systems and ESF act to ensure that no damage to the nuclear system process barrier results from internal pressures caused by abnormal operational transients or accidents.

c. Where positive, precise actions are immediately required in response to accidents, these actions are automatic and require no decision or manipulation of controls by operations personnel.

d. Essential safety actions are carried out by equipment of sufficient redundancy and independence so that no single failure of active components prevents the required actions. For systems or components to which IEEE 279 and/or IEEE 308 are applicable, single failures of passive electrical components are considered as well as single failures of active components in recognition of the higher anticipated failure rates of passive electrical components relative to passive mechanical components.

CHAPTER 01 1.2-4 REV. 20, SEPTEMBER 2020

LGS UFSAR

e. Features of the station that are essential to the mitigation of accident consequences are designed, fabricated, and erected to quality standards that reflect the importance of the safety function to be performed.

f. The design of safety systems and ESF includes allowances for environmental phenomena at the site.

g. Provision is made for control of active components of nuclear safety systems and ESF from the control room.

h. Safety systems and ESF are designed to permit demonstration of their functional performance requirements.

1.2.2.2.2.2 Containment and Isolation Criteria

a. A primary containment is provided to completely enclose the reactor vessel. It is designed to act as a radioactive material barrier during accidents that release radioactive material into the primary containment. It is possible to test the primary containment integrity and leak-tightness at periodic intervals.

b. A secondary containment that completely encloses both the primary containment and fuel storage areas is provided and designed to act as a radioactive material barrier.

c. The primary and secondary containments, in conjunction with other ESF, act to prevent radioactive material released from the containment volumes from exceeding the guideline values of applicable regulations.

d. Provisions are made for the removal of energy from within the primary containment as necessary to maintain the integrity of the containment system following accidents that release energy to the primary containment.

e. Piping that both penetrates the primary containment structure and could serve as a path for the uncontrolled release of radioactive material to the environs is automatically isolated whenever such uncontrolled radioactive material release is threatened. Such isolation is effected in time to prevent radiological effects from exceeding the guideline values of applicable regulations.

1.2.2.2.2.3 Emergency Core Cooling Systems Criteria

a. The ECCS is provided to prevent excessive fuel clad temperature as a result of a LOCA.

b. The ECCS provides for continuity of core cooling over the complete range of postulated break sizes in the nuclear system process barrier.

c. The ECCS is diverse, reliable, and redundant.

d. Operation of the ECCS is initiated automatically when required, regardless of the availability of offsite power supplies and the normal generating system of the plant.

1.2.2.2.3 Process Control Systems Criteria CHAPTER 01 1.2-5 REV. 20, SEPTEMBER 2020

LGS UFSAR 1.2.2.2.3.1 Nuclear System Process Control Criteria

a. Control equipment is provided for recirculation flow control to allow the operator to respond manually to load changes.

b. It is possible to manually control the reactor power level.

c. Control of the nuclear system is possible from a single location.

d. Nuclear system process controls are arranged to allow the operator to rapidly assess the condition of the nuclear system and to locate process system malfunctions.

e. Interlocks or other automatic equipment are provided as a backup to procedural controls to avoid conditions requiring the actuation of nuclear safety systems or ESF.

f. If the control room is inaccessible, it is possible to bring the reactor from power range operation to a cold shutdown condition by manipulation of controls and equipment that are available outside of the control room.

1.2.2.2.3.2 Power Conversion Systems Process Control and Instrumentation Criteria

a. Controls are provided to maintain temperature and pressure to below design limitations. These systems result in a stable operation and response for all allowable variations.

b. Controls and instrumentation are designed to provide equipment protection, indication, and alarm in the event of power conversion system trouble.

c. Control of the power conversion system is possible from locations accessible during normal operation.

d. Controls are provided to ensure adequate cooling of power conversion system equipment.

e. Controls are provided to ensure adequate condensate purity.

f. Controls are provided to regulate the supply of water so that adequate reactor vessel water level is maintained.

1.2.2.2.3.3 Electrical Power Systems Process Control Criteria

a. Controls are provided to ensure that sufficient electrical power is provided for startup, normal operation, and to attain prompt shutdown and continued maintenance of the station in a safe condition.

b. Control of the electrical power system is possible from locations accessible during normal operation.

1.2.2.2.4 Power Conversion Systems Criteria CHAPTER 01 1.2-6 REV. 20, SEPTEMBER 2020

LGS UFSAR

a. The power conversion system components 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.

b. The power conversion system components are designed to ensure that fission products or radioactivity associated with the steam and condensate are safely contained inside the system or are released under controlled conditions in accordance with waste disposal procedures and the plant Technical Specifications.

1.2.2.2.5 Electrical Power Systems Criteria

a. The station electrical power systems are designed to efficiently deliver the electrical power generated.

b. 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. The capacity of the power sources is adequate to accomplish all required ESF under postulated DBA conditions.

1.2.2.2.6 Auxiliary Systems Criteria

a. Fuel handling and storage facilities are designed to prevent criticality and to maintain adequate shielding and cooling for spent fuel.

b. Multiple independent station auxiliary systems are provided for the purpose of cooling and servicing the station, the reactor, and the station containment systems under various normal and abnormal conditions.

c. 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 systems from performing their design functions.

1.2.2.2.7 Radioactive Waste Management Systems Criteria

a. Gaseous, liquid, and solid waste management facilities are designed so that the discharge of radioactive effluents and offsite shipment of radioactive materials are made in accordance with applicable regulations.

b. The waste management systems, design includes means of informing station operations personnel whenever operational limits on the release of radioactive material are exceeded.

1.2.2.2.8 Shielding and Access Control Criteria

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

b. The control room is shielded against radiation and has suitable environmental control so that occupancy under DBA conditions is possible.

CHAPTER 01 1.2-7 REV. 20, SEPTEMBER 2020

LGS UFSAR 1.2.2.2.9 Fuel Handling and Storage Facilities Fuel handling and storage facilities are located in the refueling area (Zone III) of the secondary containment and are designed to preclude criticality and to maintain adequate shielding and cooling for spent fuel.

1.2.3 GENERAL ARRANGEMENT OF STRUCTURES AND EQUIPMENT The principal structures are listed below:

a. Main power block comprised of the following:

1. Two reactor enclosures with common refueling area

2. Two turbine enclosures with common operating floor

3. One control structure with common control room

4. Two diesel generator enclosures

5. One radwaste enclosure

6. One chemistry laboratory expansion

7. One administration building

8. One auxiliary boiler enclosure

9. One warehouse and shop

b. One circulating water pump structure

c. Two cooling towers with two acid/chlorine enclosures and a common valve and meter pit

d. One spray pond and spray pond pump structure

e. One Schuylkill pump structure

f. One Perkiomen pump structure

g. One water treatment enclosure

h. One fuel oil pump structure

i. One sewage treatment plant The arrangement of structures on the site is shown in drawing C-2. The general arrangements for the major power block structures and the spray pond pump structure are shown in drawings M-110, M-111, M-112, M-113, M-114, M-115, M-116, M-117, M-118, M-119, M-120, M-121,M-122, M-123, M-124, M-125, M-126, M-127, M-128, M-129, M-130, M-131, M-132, M-133, M-134, M-135, M-136, M-137, M-138, M-140, M-141, M-142, M-143, M-144, M-145, M-146, M-388, M-389, and M-390. The layout of piping and equipment within the reactor enclosure and primary containment is shown in drawings M-206, M-207, M-208, M-209, M-210, M-211, M-213, M-215, M-217, M-218, M-219, M-220, M-221, M-222, M-223, M-225, M-226, M-227, M-228, M-229, M-230, M-231, M-232, M-234, M-235, M-236, M-237, M-238, M-239, M-240, M-241, M-242, M-246, M-288, M-289, M-290, M-291, M-292, M-293, M-295, M-296, M-297, M-298, M-299, M-CHAPTER 01 1.2-8 REV. 20, SEPTEMBER 2020

LGS UFSAR 300, M-301, M-302, M-303, M-305, M-306, M-307, M-308, M-309, M-310, M-311, M-312, M-313, M-316, M-317, M-318, M-319, M-320, M-321, M-322, M-323, and M-326.

1.2.4 SYSTEM DESCRIPTION A summary of the systems provided for LGS is provided below.

1.2.4.1 Nuclear System The nuclear system includes a single-cycle, forced circulation, GE BWR producing steam for direct use in the steam turbine. A heat balance showing the major parameters of the nuclear system for the rated power condition is shown in Figure 1.2-84.

1.2.4.1.1 Reactor Core and Control Rods The fuel for the reactor core consists of slightly enriched uranium dioxide pellets contained in sealed Zircaloy-2 fuel rods. These fuel rods are assembled into individual fuel assemblies. Gross control of the core is achieved by movable, bottom-entry control rods. The control rods are of cruciform shape and are dispersed throughout the lattice of fuel assemblies. The rods are controlled by individual hydraulic systems.

1.2.4.1.2 Reactor Vessel and Internals The reactor vessel contains the core and supporting structure; the steam separators and dryers; the jet pumps; the control rod guide tubes; distribution lines for the feedwater, and core spray; the incore instrumentation; and other components. The main connections to the vessel include the steam lines, the coolant recirculation lines, the feedwater lines, the CRD and nuclear instrumentation housings, and the ECCS lines.

The reactor vessel is designed and fabricated in accordance with applicable codes for a pressure of 1250 psig. The nominal operating pressure is 1060 psia in the steam space above the separators. The vessel is fabricated of carbon steel and is clad internally with stainless steel (except for the top head which is not clad).

The reactor core is cooled by demineralized feedwater that enters the lower portion of the core and is heated 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 four main steam lines. Each steam line is provided with two isolation valves in series, one on each side of the primary containment barrier.

1.2.4.1.3 Reactor Recirculation System The reactor recirculation system consists of two recirculation pump loops external to the reactor vessel. These loops provide the piping path for the driving flow of water to the reactor vessel jet pumps that provide a continuous internal circulation path for the major portion of the core coolant flow. Each loop has one motor-driven recirculation pump. Recirculation pump speed can be varied to allow some control of reactor power level through the effects of coolant flow rate on moderator void content.

1.2.4.1.4 Residual Heat Removal System CHAPTER 01 1.2-9 REV. 20, SEPTEMBER 2020

LGS UFSAR The RHR system consists of pumps, heat exchangers, and piping that fulfill the following functions:

a. Removal of decay and sensible heat during and after plant shutdown

b. Injection of water into the reactor vessel following a LOCA, to reflood the core independent of other core cooling systems. This is discussed Section 1.2.4.2.13.

c. Removal of heat from the primary containment following a LOCA to limit the increase in primary containment pressure. This is accomplished by cooling and recirculating the suppression pool water (containment cooling) and by spraying the drywell and suppression pool air spaces (containment spray) with suppression pool water.

1.2.4.1.5 Reactor Water Cleanup System A RWCU system is provided to clean up the reactor cooling water, to reduce the amounts of activated corrosion products in the water, and to remove excess reactor coolant from the nuclear system under controlled conditions.

1.2.4.1.6 Nuclear Leak Detection System The nuclear leak detection and monitoring system consists of temperature, pressure, flow, and fission product sensors with associated instrumentation and alarms. This system detects and annunciates leakage in the following systems:

a. Main steam lines

b. RWCU system

c. RHR system

d. RCIC system

e. Feedwater system

f. ECCS systems

g. Miscellaneous systems Small leaks generally are detected by monitoring the temperature, radiation levels, and drain sump fill-up and pump-out rates. Large leaks are also detected by changes in reactor water level and changes in flow rates in process lines.

1.2.4.2 Safety-Related Systems Safety-related systems provide actions necessary to assure safe shutdown, to protect the integrity of radioactive material barriers, and/or to prevent the release of radioactive material in excess of allowable dose limits. These systems can be components, groups of components, systems, or CHAPTER 01 1.2-10 REV. 20, SEPTEMBER 2020

LGS UFSAR groups of systems. ESF systems are included in this category. ESF systems function to mitigate the consequences of DBAs.

1.2.4.2.1 Reactor Protection System The RPS initiates a rapid, automatic shutdown (scram) of the reactor. This action is taken in time to prevent excessive fuel cladding temperatures and any nuclear system process barrier damage following abnormal operational transients. The RPS overrides all operator actions and process controls.

1.2.4.2.2 Neutron Monitoring System Those portions of the NMS that are part of the RPS are safety-related. The IRM and APRM monitor neutron flux via incore detectors and provide scram logic inputs to the RPS to initiate a scram in time to prevent excessive fuel clad damage as a result of overpower transients (Upscale neutron flux), upscale simulated thermal power, and OPRM upscale are conditions that provide scram logic signals.

1.2.4.2.3 Control Rod Drive System When a scram is initiated by the RPS, the CRD system inserts the negative reactivity necessary to shut down the reactor. Each control rod is controlled individually by a hydraulic control unit.

When a scram signal is received, high pressure water from an accumulator for each rod forces each control rod rapidly into the core.

1.2.4.2.4 Control Rod Velocity Limiter A control rod velocity limiter is a part of each control rod and limits the velocity at which a control rod can fall out of the core should it become detached from its CRD. The rate of reactivity insertion resulting from a rod-drop accident is limited by this feature. The limiters contain no moving parts.

1.2.4.2.5 Control Rod Drive Housing Supports CRD housing supports are located underneath the reactor vessel near the control rod housings.

The supports limit the travel of a control rod in the event that a control rod housing is ruptured.

The supports prevent a nuclear excursion as a result of a housing failure, thus protecting the fuel barrier.

1.2.4.2.6 Nuclear System Pressure Relief System A pressure relief system, consisting of SRVs mounted on the main steam lines, prevents excessive pressure inside the nuclear system during operational transients or accidents.

1.2.4.2.7 Reactor Core Isolation Cooling System The RCIC system provides makeup water to the reactor vessel whenever the vessel is isolated from the main condenser and feedwater system. The RCIC system uses a steam-driven turbine-pump unit and operates automatically in time, and with sufficient coolant flow, to maintain adequate reactor vessel water level for events defined in Section 5.4.6.1.

CHAPTER 01 1.2-11 REV. 20, SEPTEMBER 2020

LGS UFSAR 1.2.4.2.8 Primary Containment A pressure-suppression primary containment houses the reactor vessel, the reactor coolant recirculation loops, and other branch connections of the reactor primary system. The pressure-suppression system consists of a drywell, a pressure-suppression chamber storing a large volume of water, a connecting vent system between the drywell and the water pool, isolation valves, containment cooling systems, and other service equipment. In the event of a process system piping failure within the drywell, reactor water and steam would be released into the drywell air space. The resulting increased drywell pressure would then force a mixture of air, steam, and water through the vents into the pool of water stored in the suppression chamber. The steam would condense rapidly in the suppression pool, resulting in a rapid pressure reduction in the drywell. Air transferred from the drywell to the suppression chamber pressurizes the suppression chamber and is subsequently vented to the drywell to equalize the pressure between the two chambers. Cooling systems remove heat from the reactor core, the drywell, and from the water in the suppression chamber, thus providing continuous cooling of the primary containment under accident conditions. Appropriate isolation valves are actuated during this period to ensure containment of radioactive materials within the primary containment.

1.2.4.2.9 Primary 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 process lines that are potential leakage paths for radioactive material to the environs. This action is taken upon indication of a potential breach in the nuclear system process barrier.

1.2.4.2.10 Secondary Containment Any leakage from the primary containment system is contained within the secondary containment system. This system includes the SGTS and the RERS. The secondary containment system is designed to minimize the release of airborne radioactive materials, and to provide for the controlled, filtered release of the reactor enclosure atmosphere under accident conditions.

1.2.4.2.11 Main Steam Isolation Valves Although process lines that penetrate the primary containment, and offer a potential release path for radioactive material, are provided with redundant isolation capabilities, the main steam lines, because of their large size and large mass flow rates, are given special isolation consideration.

Two automatic isolation valves, each powered by both pneumatic pressure and spring force, are provided in each main steam line. These valves fulfill the following objectives:

a. To prevent excessive damage to the fuel barrier by limiting the loss of reactor coolant from the reactor vessel resulting either from a major leak from the steam piping outside the primary containment, or from a malfunction of the pressure control system, resulting in excessive steam flow from the reactor vessel

b. To limit the release of radioactive materials, by closing the nuclear system process barrier, in case of a gross release of radioactive materials from the fuel to the reactor coolant and steam CHAPTER 01 1.2-12 REV. 20, SEPTEMBER 2020

LGS UFSAR

c. To limit the release of radioactive materials by closing the primary containment barrier, in case of a major leak from the nuclear system inside the primary containment.

1.2.4.2.12 Main Steam Line Flow Restrictors A venturi-type flow restrictor is installed in each steam line. These devices limit the loss of coolant from the reactor vessel and prevent uncovering of the core before the MSIVs are closed in case of a main steam line break.

1.2.4.2.13 Emergency Core Cooling Systems Four independent core standby cooling systems are provided to maintain fuel clad temperatures below the limits of 10CFR50.46 in the event of a breach in the RCPB that results in a loss of reactor coolant. The four core standby cooling systems are as follows:

a. High Pressure Coolant Injection System The HPCI system provides and maintains an adequate coolant inventory inside the reactor vessel to limit fuel clad temperatures as a result of postulated small breaks in the RCPB. A high pressure system is needed for such breaks because the reactor vessel depressurizes slowly, preventing low pressure systems from injecting coolant. The HPCI system includes a turbine-driven pump powered by reactor steam. The system is designed to accomplish its function on a short-term basis without reliance on plant auxiliary power supplies other than the dc power supply.

b. Automatic Depressurization System The ADS acts to rapidly reduce reactor vessel pressure in a LOCA situation in which the HPCI system fails to maintain reactor vessel water level. The depressurization provided enables the low pressure ECCS to deliver cooling water to the reactor vessel. The ADS uses some of the relief valves that are part of the nuclear system pressure relief system. The automatic relief valves are arranged to open on conditions indicating that both a break in the nuclear system process barrier has occurred and the HPCI system is not delivering sufficient cooling water to the reactor vessel to maintain the water level above a preselected value. The ADS will not be activated unless either the core spray or the LPCI system is operating.

c. Core Spray System The CS system consists of two independent pump loops that deliver cooling water to spray spargers over the core. The system is actuated by conditions indicating that a breach exists in the RCPB, 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 on the core. Either loop functioning in conjunction with the ADS or HPCI can provide sufficient fuel cladding cooling following a LOCA.

d. Low Pressure Coolant Injection CHAPTER 01 1.2-13 REV. 20, SEPTEMBER 2020

LGS UFSAR LPCI is an operating mode of the RHR system. LPCI uses the pump loops of the RHR system to inject cooling water into the reactor system. LPCI is actuated by conditions indicating a breach in the RCPB, but water is delivered to the core only after reactor vessel pressure is reduced. LPCI operation provides the capability of core reflooding following a LOCA in time to maintain the fuel cladding below prescribed temperature limits.

1.2.4.2.14 Residual Heat Removal System (Containment Cooling)

The RHR system for containment cooling is placed in operation to limit the temperature of the water in the suppression pool and of the atmospheres in the drywell and suppression chamber following a design basis LOCA, to control the pool temperature during normal operation and to reduce the pool temperature following an isolation transient. In the containment cooling mode of operation, the RHR system pumps take suction from the suppression pool and deliver the water through the RHR system heat exchangers, where cooling takes place by transferring heat to the RHR service water system. The fluid is then discharged back to the suppression pool, the drywell or suppression chamber spray headers, or to the RPV.

1.2.4.2.15 Control Room Heating, Ventilating and Air Conditioning System The control room HVAC system provides ventilation, cooling, and control of environmental conditions in the control room areas for the safety and comfort of operating personnel during normal operations and during postulated accident conditions. The system includes air filter units used to remove contaminants that are potentially present in the air following a postulated accident before introducing the air into the control room HVAC system.

1.2.4.2.16 Reactor Enclosure Recirculation System and Standby Gas Treatment System Both the RERS and the SGTS service the secondary containment. The recirculation system has the capability of recirculating the reactor enclosure air volume prior to its discharge via the SGTS, following a LOCA. The SGTS has the capability of maintaining a negative pressure within the reactor enclosures and the refueling area zones of the secondary containment with respect to the outside atmosphere. The air moving through the SGTS is filtered and discharged through the north exhaust stack.

1.2.4.2.17 Standby AC Power Supply The standby ac power supply system consists of four diesel generator sets per unit. The diesel generators are sized so that any three diesels can supply all the necessary power requirements for one unit in the DBA condition. The diesel generators are designed to start and be able to accept load within 10 seconds. Four independent 4 kV ESF switchgear assemblies are provided for each reactor unit. Each diesel generator feeds an independent 4 kV bus for each reactor unit.

Each diesel generator starts automatically upon LOOP or detection of a LOCA. The necessary safety-related loads are applied in a preset time sequence. Each generator operates independently and without paralleling during a LOOP or LOCA signal.

1.2.4.2.18 DC Power Supply CHAPTER 01 1.2-14 REV. 20, SEPTEMBER 2020

LGS UFSAR Each unit is provided with two independent Class 1E 125/250V, two independent Class 1E 125V, one non-Class 1E 125/250V, and one non-Class 1E 250V DC systems. The systems are provided to supply station DC control power, DC power to diesel generators, their associated switchgear, ESF systems, and DC motor driven pumps and valves.

The Class 1E DC systems are designed to supply power adequate to satisfy the safety-related load requirements of the unit with the postulated LOOP and any concurrent signal failure in the DC system.

1.2.4.2.19 Residual Heat Removal Service Water System The purpose of the RHRSW system is to provide a reliable supply of cooling water for heat removal from the RHR system during normal shutdown operations and under postaccident conditions. It can also supply a source of water if postaccident flooding of the primary containment is required.

The system consists of two independent loops, each of 100% capacity, supplying one RHR heat exchanger in each unit. During postaccident conditions the system uses the common spray pond as the heat sink. Interconnections are provided to allow use of a cooling tower as a heat sink during normal operations and, if conditions permit, during postaccident operation.

1.2.4.2.20 Emergency Service Water System The purpose of the ESW system is to provide a reliable supply of cooling water to emergency equipment during LOOP and postaccident conditions.

The system consists of two independent loops. Each loop supplies corresponding safety-related equipment in each unit. The safety- related cooling loads consist of equipment room coolers, control room chillers, and the RHR pump coolers.

ESW for the diesel generators is supplied by either loop. The ESW system uses the common spray pond as a heat sink. Interconnections are provided to allow use of a cooling tower as a heat sink if conditions permit.

1.2.4.2.21 Main Steam Line Radiation Monitoring System The MSL-RMS consists of four gamma radiation monitors located external to the main steam lines just outside of the primary containment. The monitors are designed to detect a gross release of fission products from the fuel. Upon detection of high radiation, an alarm signal is initiated by the monitors.

1.2.4.2.22 Reactor Enclosure and Refueling Area Ventilation Radiation Monitoring System The REVE-RMS and RAVE-RMS consists of a number of radiation monitors arranged to monitor the activity level of the ventilation exhaust from the reactor enclosure and refueling area. Upon detection of high radiation, the affected area is automatically isolated and the SGTS and RERS (for reactor enclosure only) are started.

1.2.4.2.23 Remote Shutdown System CHAPTER 01 1.2-15 REV. 20, SEPTEMBER 2020

LGS UFSAR A remote shutdown panel and associated procedures are provided for each unit so that the plant can be maintained in a safe shutdown condition in the event that the control room becomes uninhabitable.

1.2.4.2.24 Standby Liquid Control System Although not intended to provide rapid reactor shutdown, the SLCS provides a redundant, independent, and alternate way to bring the reactor subcritical and to maintain it subcritical 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 core to accomplish normal shutdown. The system is sized to counteract the positive reactivity effect from rated power to the cold shutdown condition. The SLCS also provides pH control for the primary containment water inventory following a LOCA.

1.2.4.2.25 Deleted.

1.2.4.2.26 Redundant Reactivity Control System The RRCS is designed to provide a redundant and diverse method of shutting down the reactor, in the unlikely event that the RPS does not scram the reactor as a result of an anticipated operating transient. The RRCS logic is initiated when either the high reactor pressure or low reactor water level setpoints are reached. A signal is then sent to open the ARI valves that vent the CRD scram air header to insert the control rods into the reactor. A signal is also transmitted to the RPT breakers to trip the reactor recirculation pumps to reduce the reactor power. An initiation of the RRCS logic by high reactor pressure will cause the feedwater pumps to automatically runback. If reactor power has not decreased to a predetermined level, within a specified period of time, the RRCS logic will initiate a feedwater runback and the injection of a neutron poison solution into the reactor, via the SLCS, and shut down the reactor.

The system consists of control panels, their associated ATWS detection and actuation logic, and the necessary interface logic to the recirculation system, the feedwater system, the SLCS, the RWCU system and the ARI components of the CRD system required to perform specific functions in response to an ATWS event.

1.2.4.3 Instrumentation and Control Systems 1.2.4.3.1 Nuclear System Process Control and Instrumentation 1.2.4.3.1.1 Reactor Manual Control System The reactor manual control system provides the means by which control rods are positioned from the control room for power control. The system operates valves in the CRD hydraulic system to control rod position. Only one control rod can be manipulated at a time. The reactor manual control system includes the logic that restricts control rod movement (rod block) under certain conditions as a backup to procedural controls.

1.2.4.3.1.2 Recirculation Flow Control System The recirculation flow control system controls the speed of the reactor recirculation pumps.

Adjusting the pump speed changes the coolant flow rate through the core. This effects changes CHAPTER 01 1.2-16 REV. 20, SEPTEMBER 2020

LGS UFSAR in core power level. The system is operated manually to match reactor power output to the load demand by adjusting the frequency of the electrical power supply for the reactor recirculation pumps.

1.2.4.3.1.3 Neutron Monitoring System The NMS is a system of incore neutron detectors and electronic monitoring equipment. The system provides indication of neutron flux that can be correlated to thermal power level for the entire range of flux conditions that may exist in the core. The SRM and IRM provide flux level indications during reactor startup and low power operation. The LPRM and APRM allow assessment of local and overall flux conditions during power range operation. The TIP system provides a means to calibrate the individual LPRMs utilizing a gamma-measuring probe. The NMS provides inputs to the reactor manual control system to initiate rod blocks if preset alarm flux limits are exceeded, and inputs to the RPS to initiate a scram if trip limits are exceeded.

1.2.4.3.1.4 Refueling Interlocks A system of interlocks, restricting the movements of refueling equipment and control rods when the reactor is in the refueling and startup modes, is provided to prevent an inadvertent criticality during refueling operations. The interlocks back up procedural controls that have the same objective. The interlocks affect the refueling platform, the refueling platform hoists, the fuel grapple, and control rods.

1.2.4.3.1.5 Reactor Vessel Instrumentation In addition to instrumentation provided for the nuclear safety systems and ESF, 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. The instrumentation provided monitors reactor vessel pressure, water level, temperature, internal differential pressures and coolant flow rates, and top head flange leakage.

1.2.4.3.1.6 Deleted 1.2.4.3.1.7 Plant Monitoring System The PMS is a centralized, integrated system which performs the process monitoring and calculation that are necessary for the effective evaluation of normal and emergency power plant operation. The PMS acquires and records process data (e.g., temperatures, pressures, flows, status indicators) to produce displays, logs, and plots of current or historical plant performance which are presented to plant personnel in the plant main control room.

1.2.4.3.1.8 Rod Worth Minimizer The RWM monitors and enforces adherence to established low power level rod insert and withdraw sequences. This function prevents the operator from establishing control rod patterns that are not consistent with the prescribed sequence by initiating the appropriate rod insert and withdraw block. When RWM is inoperable both insert and withdraw blocks are enforced unless the RWM is bypassed. The RWM enforces control rod sequences designed to limit individual control rod worths to acceptable levels as determined by the rod-drop accident design basis.

CHAPTER 01 1.2-17 REV. 20, SEPTEMBER 2020

LGS UFSAR 1.2.4.3.2 Power Conversion Systems Process Control and Instrumentation 1.2.4.3.2.1 Pressure Regulator and Turbine-Generator Control The pressure regulator controls the turbine control and turbine bypass valves to maintain the nuclear system pressure essentially constant.

The turbine-generator speed-load control is set above the desired load and grid generator frequency. Load changes are made by the operator by adjusting the reactor recirculation flow control system and control rods.

The turbine-generator speed-load controls initiate rapid closure of the turbine control valves (rapid opening of the turbine bypass valves) to prevent turbine overspeed on loss of the generator electric load.

1.2.4.3.2.2 Feedwater System Control A three-element control system regulates the feedwater system so that proper water level is maintained in the reactor vessel. Signals used by the control system are main steam flow rate, reactor vessel water level, and feedwater flow rate. The feedwater control signal is used to control the speed of the steam turbine-driven feedwater pumps.

1.2.4.3.2.3 Electrical Power System Control Controls for the electrical power system are located in the control room to permit safe startup, operation, and shutdown of the plant.

1.2.4.4 Electrical Systems 1.2.4.4.1 Transmission and Generation Systems Redundant sources of offsite power are provided to each unit by separate transmission lines to ensure that no single failure of any active component can prevent a safe and orderly shutdown.

Each unit is provided with an independent substation, which is 220 kV for Unit 1 and 500 kV for Unit 2. The substations are connected by an autotransformer and ultimately feed into the PJM interconnection through the 220 kV and 500 kV transmission systems. Two independent offsite sources provide auxiliary power for startup and for operating the safety-related systems.

The main generator for each unit is an 1800 rpm, 3 phase, 60 Hz synchronous unit rated at 1264.97 MVA. Each generator is connected directly to the turbine shaft and is equipped with an excitation system coupled directly to the generator shaft. Power from the generators is stepped up from 22-220 kV on Unit 1 from 22-500 kV on Unit 2 by the unit main transformers and supplied by overhead lines to the 220 kV and 500 kV switchyards, respectively.

1.2.4.4.2 Electric Power Distribution Systems The electric power distribution system includes Class 1E and non- Class 1E ac and dc power systems. The Class 1E power system supplies all safety-related equipment, while the non-Class 1E system supplies the balance of plant equipment.

CHAPTER 01 1.2-18 REV. 20, SEPTEMBER 2020

LGS UFSAR The Class 1E ac system for each unit consists of four independent load groups. Two independent offsite power systems provide the normal electric power to these groups. Each load group includes a 4 kV switchgear, a 440 V load center, 440 V MCCs, and 120 V control and instrument power panels. The vital ac instrumentation and control power supply systems include battery systems and static inverters.

There are four independent diesel generator sets for each unit. Each diesel generator is provided as a standby source of power for one of the four Class 1E ac load groups in each unit. Assuming the total LOOP and failure of one diesel generator, the remaining diesel generators have sufficient capacity to operate all the equipment necessary to prevent undue risk to public health and safety in the event of a DBA on one unit and an emergency shutdown of the second unit.

The non-Class 1E ac system includes 13.2 kV switchgear, 2.3 kV switchgear, 440 V load centers, MCCs, and 120 V control and instrument power panels.

Two independent Class 1E 125 V dc batteries and two independent Class 1E 125/250 V dc batteries and associated battery chargers provide direct current power for the Class 1E dc loads of each unit. Power for non-Class 1E dc loads is supplied from 125/250 V and 250 V non-Class 1E batteries and associated battery chargers.

1.2.4.5 Fuel Handling and Storage Systems 1.2.4.5.1 New and Spent Fuel Storage The fuel storage racks are designed to prevent load buckling and inadvertent criticality under dry and flooded conditions. Sufficient coolant and shielding are maintained to prevent overheating and excessive personnel exposure, respectively. New and spent fuel will be stored in the spent fuel pool in addition to spent fuel stored in the Independent Spent Fuel Storage Installation.

1.2.4.5.2 Fuel Pool Cooling and Cleanup System The FPCC system is provided to remove decay heat from spent fuel stored in the fuel pool and to maintain specified water temperature, purity, clarity, and level.

1.2.4.5.3 Fuel Handling Equipment The major fuel servicing and handling equipment includes the reactor enclosure cranes, refueling service platform, fuel and control rod servicing tools, fuel sipping and inspection devices, and other auxiliary servicing tools.

1.2.4.6 Cooling Water and Auxiliary Systems 1.2.4.6.1 Service Water System The service water system supplies cooling water to equipment required for normal plant operation.

The system consists of three 50% capacity pumps with associated piping and valves. The cooling water supply to the pumps is taken from the cooling tower basin, while the water being returned from the system is discharged into the cooling tower.

1.2.4.6.2 Reactor Enclosure Cooling Water System CHAPTER 01 1.2-19 REV. 20, SEPTEMBER 2020

LGS UFSAR The RECW system is a closed-loop cooling water system that provides cooling water for miscellaneous reactor auxiliary plant equipment. The RECW system consists of two 100%

capacity pumps, two 100% capacity heat exchangers, a head tank, chemical addition tank, associated piping, valves, and controls. One RECW pump is normally in service and the other pump is on automatic standby. During normal plant operation, heat is transferred from the RECW system to the service water system.

1.2.4.6.3 Turbine Enclosure Cooling Water System The TECW system is a closed-loop cooling system that provides cooling water to the auxiliary plant equipment associated with the nuclear and power conversion systems in the turbine enclosure. The TECW system consists of two 100% capacity pumps, two 100% capacity heat exchangers, a head tank, chemical addition tank, and associated piping and valves. During normal plant operation, the TECW heat exchanger transfers heat from the TECW system to the service water system. One TECW pump is normally in service and the other pump is on automatic standby.

1.2.4.6.4 Fire Protection System A fire protection system supplies fire fighting water to points throughout the plant. An automatic carbon dioxide protection system, in addition to portable fire extinguishers, is provided. An automatic halon fire protection system provides protection in the auxiliary electrical equipment room.

1.2.4.6.5 Plant Heating, Ventilating, and Air Conditioning Systems The HVAC systems supply and circulate filtered fresh air for personnel comfort and equipment cooling.

1.2.4.6.6 Instrument Air System The instrument air system supplies compressed air of suitable quality and pressure for power plant operation.

1.2.4.6.7 Clarified and Domestic Water Systems The clarified and domestic water systems provide filtered, clarified water for plant and personnel use.

1.2.4.6.8 Demineralized Water Makeup System A demineralized water makeup system is provided to furnish a supply of treated water suitable for use as makeup for the plant.

1.2.4.6.9 Plant Equipment and Floor Drainage System The plant equipment and floor drainage system handles both radioactive and nonradioactive drains. Drains that can contain radioactive materials are pumped to the radwaste system for cleanup, reuse, or discharge. Nonradioactive drains are processed to remove oil and chemicals and then discharged to the environs.

CHAPTER 01 1.2-20 REV. 20, SEPTEMBER 2020

LGS UFSAR 1.2.4.6.10 Process Sampling System The process sampling system is provided to monitor the operation of plant equipment and to provide information needed to make operational decisions.

1.2.4.6.11 Plant Communication System The plant communication system provides communication between various plant structures and locations.

1.2.4.7 Power Conversion Systems 1.2.4.7.1 Turbine-Generator The turbine-generator consists of the turbine, generator, exciter, controls, and required subsystems designed for a nominally rated MWe 1231.1.

The turbine is an 1800 rpm, tandem-compound, nonreheat steam turbine and an EHC system.

The main turbine comprises one double-flow high pressure turbine and three double-flow low pressure turbines. Exhaust steam from the high pressure turbine passes through moisture separators before entering the three low pressure turbines.

The generator is a direct-driven, 3 phase, 60 Hz, 22,000 V, 1800 rpm, synchronous generator rated at 1265 MVA on the basis of guaranteed best turbine efficiency MW rating at an expected 0.973 power factor at 75 psig hydrogen pressure.

1.2.4.7.2 Main Steam System The main steam system delivers steam from the nuclear boiler system via four 26 inch OD steam lines to the turbine-generator. This system also supplies steam to the SJAE, the RFPT, the main condenser hotwell at startup and low loads, and the steam seal evaporator.

1.2.4.7.3 Main Condenser The main condenser system condenses and deaerates the exhaust steam from the main turbine and RFPT, and provides a heat sink for the turbine bypass system. The main condenser is a triple-pass, triple-pressure, deaerating-type with a reheating-deaerating hotwell and divided water boxes. The condenser consists of three sections with each section located below one of three low pressure turbines.

1.2.4.7.4 Main Condenser Evacuation System The main condenser evacuation system removes the noncondensable gases from the main condenser. Two redundant SJAE are provided for air removal during normal operation, and one motor-driven vacuum pump is provided for air removal during startup.

1.2.4.7.5 Steam Seal System CHAPTER 01 1.2-21 REV. 20, SEPTEMBER 2020

LGS UFSAR The steam seal system provides clean, nonradioactive steam to the seals of the turbine valve packings and the turbine shaft packings. The sealing steam is supplied by the steam seal evaporator. The auxiliary boiler provides an auxiliary steam supply for startup and can be used as a backup to the steam seal evaporator, when available, during power operation.

1.2.4.7.6 Turbine Bypass and Pressure Control System The turbine bypass and pressure control system provides control of the reactor pressure for the following operating modes:

a. During reactor heatup to rated pressure

b. While the turbine is being brought up to speed and synchronized

c. During normal power operation and transient power operation when the reactor steam generation exceeds the turbine steam requirements

d. When shutting down the reactor 1.2.4.7.7 Circulating Water System The circulating water system is a closed-loop system designed to circulate the flow of water required to remove the heat load from the main condenser and auxiliary heat exchanger equipment and discharge it to the atmosphere through natural draft cooling towers.

1.2.4.7.8 Condensate Cleanup System The function of the condensate cleanup system is to maintain the required purity of the feedwater to the reactor. The system consists of full flow condensate filter/demineralizers and deep bed condensate demineralizers. The condensate filter/demineralizers remove suspended solids from the condensate and feedwater streams. Downstream of and in series with the condensate filter/demineralizers, the deep bed condensate demineralizers use ion exchange resins that remove dissolved impurities from the feedwater. A bypass exists around one of the condensate filter/demineralizers to provide a means to control the iron/transition metal concentration in the reactor recirculation system. The deep bed condensate demineralizers also remove some of the radioactive material produced by corrosion as well as fission product carryover from the reactor.

1.2.4.7.9 Condensate and Feedwater Systems The condensate and feedwater systems are designed to deliver the required feedwater flow to the reactor vessel during stable and transient operating conditions throughout the entire operating range from startup to full load to shutdown. The system operates using three condensate pumps to pump deaerated condensate from the hotwell of the main condenser through the SJAE condenser, the gland steam packing exhauster condenser, and then to the condensate cleanup system. The demineralized feedwater then flows through three parallel strings of low pressure feedwater heaters to the suction of three reactor feed pumps that deliver the feedwater through one set of three parallel high pressure heaters to the reactor.

1.2.4.7.10 Condensate and Refueling Water Storage Facilities The condensate and refueling water storage facilities provide storage of condensate water for use in normal plant operations and refueling operations.

CHAPTER 01 1.2-22 REV. 20, SEPTEMBER 2020

LGS UFSAR 1.2.4.8 Radioactive Waste Systems The radioactive waste systems are designed to confine the release of plant-produced radioactive material to well within the limits specified in 10CFR20 and 10CFR50, Appendix I. Various methods are used to achieve this end (e.g., collection, filtration, holdup for decay, dilution, and concentration). The pre-1994 10CFR20, Appendix B limits were used for the original licensing basis of the plant. Current liquid effluent releases are limited to ten-times the Effluent Concentration Limit (ECL) for each isotope specified in post-1994 10CFR20, Appendix B, Table 2, Column 2. Current gaseous and liquid effluent releases are controlled by the Radioactive Effluent Controls Program defined by the Technical Specifications.

1.2.4.8.1 Liquid Radwaste System The liquid radwaste system collects, treats, stores, and disposes of radioactive liquid wastes.

These wastes are collected in sumps and drain tanks at various locations throughout the plant and then transferred to the appropriate collection tanks in the radwaste enclosure prior to treatment, storage, and disposal. Processed liquid wastes are returned to the condensate system, packaged for offsite shipment, or discharged from the plant.

Equipment is selected, arranged, and shielded to permit operation, inspection, and maintenance within radiation allowances for personnel exposure. For example, tanks and processing equipment that will contain significant radiation sources are shielded and sumps, pumps, instruments, and valves are located in controlled access rooms or spaces. Processing equipment is selected and designed to require a minimum of maintenance.

Valving redundancy, instrumentation for detection, alarms of abnormal conditions, and procedural controls protect against the accidental discharge of liquid radioactive waste.

1.2.4.8.2 Solid Radwaste System Solid wastes originating from nuclear system equipment are stored for radioactive decay in the fuel storage pool and prepared for reprocessing or offsite storage in approved shipping containers. Examples of these wastes are spent control rods, and incore ion chambers.

Process solid wastes are collected, dewatered, packaged, and stored in shielded compartments prior to offsite shipment. Examples of these solid wastes are filter residue, spent resins, paper, air filters, rags, and used clothing.

1.2.4.8.3 Gaseous Radwaste System Radioactive gaseous wastes are discharged to the turbine enclosure vent via the gaseous radwaste system. This system provides hydrogen-oxygen recombination, filtration, and holdup of the offgases to ensure a low rate of release from the turbine enclosure vent.

The offgases from the main condenser are the greatest source of gaseous radioactive waste.

The treatment of these gases reduces the released activity to well below permissible levels even with some defective fuel elements.

1.2.4.9 Radiation Monitoring and Control CHAPTER 01 1.2-23 REV. 20, SEPTEMBER 2020

LGS UFSAR 1.2.4.9.1 Process Radiation Monitoring Radiation monitors are provided on various lines to monitor and control radioactivity in process and effluent streams and to activate appropriate alarms and controls.

1.2.4.9.2 Area Radiation Monitors Radiation monitors are provided to monitor for abnormal radiation at various locations in the plant.

These monitors activate alarms when abnormal radiation levels are detected.

1.2.4.9.3 Site Environs Radiation Monitors Radiation monitors are provided outside the plant structures to monitor radiation levels. These data are used for determining the contribution of plant operations to onsite and offsite radiation levels.

1.2.4.10 Shielding Shielding is provided throughout the plant, as required, to reduce radiation levels to operating personnel and to the general public within the applicable limits set forth in 10CFR20, 10CFR50, and 10CFR50.67. It is also designed to protect certain plant components from radiation exposure resulting in unacceptable alterations of material properties or activation.

1.2.5 Control Of Unit 2 Construction Activities During Unit 1 Operation To meet the requirements of 10CFR50.34(b)(6)(vii), an evaluation was performed which verified that safety-related structures, systems, and components of Unit 1 were protected from potential hazards resulting from construction activities on Unit 2. This evaluation also identified the plant administrative and managerial controls required to provide assurance that technical specification limiting conditions for operation were not exceeded. Administrative and managerial controls existed in the form of design controls, procedures and training during the construction of Unit 2.

The plant Technical Specifications were taken into consideration throughout the development of these controls. The Plant Operations Review Committee was not required to review the Unit 2 construction activity controls or the revisions thereto.

LGS utilizes a unitized design for safety-related structures and systems with only the following common structures and systems as exceptions: the control enclosure and supporting systems, spray pond, spray pond pumphouse and supporting systems, ESW system, RHRSW system, and SGTS. All portions of the Unit 1 systems and those common systems required for the safe operation and shutdown of Unit 1 were completed prior to Unit 1 fuel load. Unit 2 systems and structures required to assure that Unit 1 limiting conditions for operation were not exceeded had also been completed. A PAB that was maintained in accordance with an established separation program served to define the Unit 1 and common operations area. The PAB considered personnel access for both security and ALARA radiation exposure. The LGS security plan covered all aspects of Unit 1 and common security.

Administrative and managerial controls governed the activities associated with Unit 2 construction to prevent the safe operation of Unit 1 from being affected. Design controls, procedures, safety programs, or training sessions existed to ensure that the following areas were addressed during Unit 2 construction:

CHAPTER 01 1.2-24 REV. 20, SEPTEMBER 2020

LGS UFSAR

- Separation of Unit 1 and common safety-related systems and equipment (including those Unit 2 systems, structures, and equipment required for operation of Unit 1) from Unit 2 construction activities.

- Yard activities, such as excavation, blasting, grading, use of heavy equipment, and erection of structures, such that consideration is given to site drainage, protection of underground utilities, and Unit 1 operation.

- Maintenance of construction personnel radiation exposures as low as reasonably achievable.

- Minimizing the probability of the LOOP to Unit 1 due to Unit 2 construction activities.

- The control of the use of hazardous and toxic chemicals which may have been used during Unit 2 construction.

- The control of the use and storage of compressed gas cylinders which may have been used during Unit 2 construction such that Unit 1 was not jeopardized.

- Construction cranes were prevented from damaging Unit 1 and common structures and systems important to safety by governing placement and operation.

- Rigging and the movement of equipment was governed to ensure potential heavy load drops did not affect Unit 1 or common safe shutdown equipment.

- The safe operation of Unit 1 was maintained by considering the fire barrier, missile protection, radiation protection, pressure boundary, flood control, and security aspects of walls, ceilings and floors when construction openings were made.

- Use of safe onsite transportation and traffic practices.

- Fire protection programs were enforced and not affected by construction activities associated with Unit 2.

CHAPTER 01 1.2-25 REV. 20, SEPTEMBER 2020

LGS UFSAR 1.3 COMPARISON TABLES 1.3.1 COMPARISONS WITH SIMILAR FACILITY DESIGNS This section highlights the principal design features of the plant and compares the major features with those of other BWR facilities. The design of this facility is based on proven technology attained during the development, design, construction, and operation of BWRs of similar types.

The data, performance, characteristics, and other information presented here represent the original design, and licensed operating conditions.

The following tables summarize the plant design characteristics for LGS, PBAPS, SSES, and Zimmer Unit 1:

Table No. System 1.3-1 Nuclear Steam Supply System 1.3-2 Engineered Safety Features and Auxiliary Systems 1.3-3 Power Conversion Systems 1.3-4 Containment 1.3-5 Structural Design 1.3-6 Radioactive Waste Management Systems 1.3-7 Electrical Power Systems 1.3.2 COMPARISON OF FINAL AND PRELIMINARY INFORMATION Significant changes made in the facility design between the PSAR stage and the FSAR stage are listed in Table 1.3-8. Design changes are only those that have occurred since the last PSAR amendment. Notice of all other design changes has been given through amendments to the PSAR. Each item in Table 1.3-8 is cross-referenced to the appropriate section of the UFSAR which describes the present design.

CHAPTER 01 1.3-1 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 1.3-1 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS LGS* SSES ZIMMER 1 PBAPS THERMAL AND HYDRAULIC DESIGN (Section 4.4)

Rated power, MWt 3293 3293 2436 3293 Design power, MWt 3435 3439 2550 3440 Steam flow rate, lb/hr 14.159x106 13.48x106 10.477x106 13.381x106 Core coolant flow rate, lb/hr 100x106 100.5x106 100.5x106 102.5x106 Feedwater flow rate, lb/hr 14.127x106 13.44x106 10.447x106 10.447x106 Feedwater temperature, F 420 383 420 376.1 System pressure, nominal in steam dome, psia 1020 1020 1020 1020 Average power density, kW/liter 48.7 48.7 50.51 50.8 Maximum linear heat generation rate, kW/ft 13.4 13.4 13.4 18.35 Average linear heat generation rate, kW/ft 5.3 5.3 5.4 7.049 Maximum heat flux, Btu/hr-ft2 361,600 361,600 354,255 425,060 Average heat flux, Btu/hr-ft2 144,100 143,700 144,032 163,230 Maximum uranium dioxide (U02) temperature, F 3435 3435 3435 4430 Average volumetric fuel temperature, F 2130 2130 2130 2780 Average fuel rod surface temperature, F 566 566 566 560 Minimum critical power ratio (MCPR) 1.22 1.25 1.24 -

Coolant enthalpy at core inlet, Btu/lb 526.1 521.8 527.4 521.2 Core maximum exit voids within assemblies, % 77.1 76.6 75 79 Core average exit quality, % steam 14.1 13.2 13.2 13.2 Design power peaking factor(1)

Maximum relative assembly power 1.4 1.4 1.4 1.4 Design power peaking factor(1)

Maximum relative assembly power 1.4 1.4 1.4 1.4 Axial peaking factor 1.4 1.4 1.4 1.5 NUCLEAR DESIGN (FIRST CORE)

(Section 4.3)

Water/UO2 volume ratio (cold, beginning of cycle 2.74 2.80 2.55 2.43

[BOC])

Reactivity with strongest control rod out, keff <0.99 <0.99 <0.99 <0.99 CHAPTER 01 1.3-2 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 1.3-1 (Contd)

COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS LGS* SSES ZIMMER 1 PBAPS Dynamic void coefficient (end of cycle 1 [EOC-1])

a. At core average voids, % 39.7 39.7 40.54 (2)

b. At rated output, ¢/% -7.48 -7.48 -8.57 (2)

Fuel temperature doppler coefficient (BOC) at -1.85x10-5 -1.85x10-5 -1.94x10-5 (2) rated output (1/k) (dk/dt) (1/C)

Initial average U-235 enrichment wt, % 1.88 1.88 1.90 2.19 Initial cycle exposure, Mwd/short ton 9600 9600 9200 (2)

CORE MECHANICAL DESIGN (Sections 4.2 and 4.6)

Fuel assembly Number of fuel assemblies 764 764 560 764 Fuel rod array 8x8 8x8 8x8 7x7 Overall length, in 176 176 176 176 Weight of UO2 per assembly, lb 456 458 466 490 (pellet type)

Weight of fuel assembly, lb 680 665 698 676 Fuel rods Number per fuel assembly 62 62 63 63 Outside diameter, in 0.483 0.483 0.493 0.563 Cladding thickness, in 0.032 0.032 0.034 0.032 Diametral gap, pellet to cladding, in 0.009 0.009 0.009 0.011 Length of gas plenum, in 9.48 10 14 16 Cladding material Zircaloy-2 Zircaloy-2 Zircaloy-2 Zircaloy-2 Cladding process Freestanding Freestanding Freestanding Freestanding loaded tubes loaded tubes loaded tubes loaded tubes Fuel pellets Material UO2 UO2 UO2 UO2 Density, % of theoretical 94 95 95 95 Diameter, in 0.410 0.410 0.416 0.487 Length, in 0.410 0.410 0.420 0.5 Fuel channel Overall length, in 166.9 166.9 166.9 166.9 CHAPTER 01 1.3-3 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 1.3-1 (Contd)

COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS LGS* SSES ZIMMER 1 PBAPS Thickness, in 0.100,0.080(3) 0.080 0.100 0.080 Cross section dimensions, in 5.48x5.48 5.48x5.48 5.48x5.48 5.44x5.44 Material Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Core assembly Fuel weight as UO2, lb 348,939 349,912 260,551 369,790 Core diameter (equivalent), in 187.1 187.1 160.2 187.1 Core height (active fuel), in 150 150 146 144 Reactor control system Movable control Movable control Movable control Movable control rods and variable rods and variable rods and variable rods and variable Method of variation of reactor power forced coolant forced coolant forced coolant forced coolant flow flow flow flow Number of movable control rods 185 185 137 185 Shape of movable control rods Cruciform Cruciform Cruciform Cruciform Pitch of movable control rods, in 12.0 12.0 12.0 12.0 Control material in movable rods Boron Carbide B4C granules B4C granules B4C granules (B4C) granules compacted in compacted in compacted in compacted in stainless steel stainless steel stainless steel stainless steel tubes tubes tubes tubes Type of control rod drives Bottom entry Bottom entry Bottom entry Bottom entry locking piston locking piston locking piston locking piston Type of temporary reactivity control Burnable poison; Burnable poison; Burnable poison; Burnable poison; gadolinia-urania gadolinia-urania gadolinia-urania gadolinia-urania fuel rods fuel rods fuel rods fuel rods Incore neutron instrumentation Total number of LPRM detectors 172 172 124 172 Number of incore LPRM penetrations 43 43 31 43 Number of LPRM detectors per penetration 4 4 4 4 Number of SRM penetrations 4 4 4 4 Number of IRM penetrations 8 8 8 8 Total nuclear instrument penetrations 55 55 43 55 CHAPTER 01 1.3-4 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 1.3-1 (Contd)

COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS LGS* SSES ZIMMER 1 PBAPS Range (and number) of detectors:

a. SRM Shutdown Shutdown Shutdown Shutdown through through through through criticality(4) criticality(4) criticality(4) criticality(4)

Prior to criticality Prior to criticality Prior to criticality Prior to criticality

b. IRM to low power (8) to low power (8) to low power (8) to low power (8)

Power range monitors 1% to 125% 5% to 125% 1% to 125% 1% to 125%

power power power power

- LPRM 172 172 124 172

- APRM 6 6 6 6 Number and type of incore neutron sources 7; Sb-Be 7; Sb-Be 5; Sb-Be 7; Sb-Be REACTOR VESSEL DESIGN (Section 5.3)

Material Carbon steel/ Carbon steel/ Carbon steel/ Carbon steel/

stainless clad stainless clad stainless clad stainless clad Design pressure, psig 1250 1250 1250 1250 Design temperature, F 575 575 575 575 Inside diameter, ft-in 20-11 20-11 18-2 20-11 Inside height, ft-in 72-1 72-11 69-10 72-11 Minimum base metal thickness (cylindrical 6.187 6.19 5.375 6.3125 section), in Minimum cladding thickness, in 1/8 1/8 1/8 1/8 REACTOR COOLANT RECIRCULATION SYSTEM DESIGN (Section 5.4)

Number of recirculation loops 2 2 2 2 Design pressure

a. Inlet leg, psig 1250 1250 1250 1148

b. Outlet leg, psig 1500 1500 1575 1326 Design temperature, F 575 575 575 562 Pipe diameter, in 28 28 20 28 Pipe material, ANSI 316 304/316 304/316 304/316 CHAPTER 01 1.3-5 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 1.3-1 (Contd)

COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS LGS* SSES ZIMMER 1 PBAPS Recirculation pump flow rate, gpm 45,200 45,200 32,500 45,200 Number of jet pumps in reactor 20 20 20 20 MAIN STEAM LINES (Section 10.3)

Number of steam lines 4 4 4 4 Design pressure, psig 1115 1250 1250 1115 Design temperature, F 582 575 575 583 Pipe diameter, in 26 26 24 26 Pipe material Carbon steel Carbon steel Carbon steel Carbon steel (1)

Local and total peaking factors are not used as input to the thermal/hydraulic codes and are therefore not included here.

(2)

Available values for PBAPS first core were developed on a different basis and are not directly comparable to LGS, SSES, or Zimmer 1.

(3)

LGS Unit 1 will use 80 mil thick fuel channels beginning with refueling cycle 2.

LGS Unit 2 will use 80 mil thick fuel channels beginning with the initial core.

• Based on the original licensed operating conditions.

CHAPTER 01 1.3-6 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 1.3-2 COMPARISON OF ENGINEERED SAFETY FEATURES AND AUXILIARY SYSTEMS DESIGN CHARACTERISTICS LGS* SSES ZIMMER 1 PBAPS EMERGENCY CORE COOLING SYSTEMS (Systems sized on design power)

(See UFSAR Section 6.3)

Core spray system Number of loops 2 2 1 2 Flow rate, gpm, per pump 6350 at 105 psid 6350 at 105 psid 4725 at 119 psid 6250 at 122 psid High pressure coolant injection system Number of loops 1 1 1 1 Flow rate, gpm 5600 minimum 5000 at 1172-165 1330 at 1110 psid, 5000 at 1120-150 psia, 4625 at 200 psid psid Automatic depressurization system Number of relief valves 5 6 6 5 Low pressure coolant injection Number of loops 4 2 3 2 Number of pumps 4 4 3 4 Flow rate, gpm/pump 10,000 at 20 psid 10,650 at 20 psid 5050 at 20 psid 10,000 at 20 psid AUXILIARY SYSTEMS (See UFSAR Sections 5.4 and 9.2)

Residual heat removal system Reactor shutdown cooling mode:

Number of pumps 2 2 3 4 Flow rate, gpm/pump 10,000 10,000 5050 10,000 Duty, Btu/hr/heat exchanger 41.6x106 44x106 30.8x106 70.0x106 Number of heat exchangers 2 2 3 4 Primary containment cooling mode:

Flow rate, gpm/heat exchanger 10,000 10,000 5050 10,000 Service water system Flow rate, gpm/heat exchanger 12,000 9000 5000 14,000 Number of pumps 3 2 4 3 Reactor core isolation cooling system Flow rate, gpm 600 at 1120 psid 600 at 1172-165 psia 400 at 1120 psid 616 at1120 psid Fuel pool cooling and cleanup system Capacity, Btu/hr 11.25x106 13.2x106 6.9x106 11.25x106

• Based on the original licensed operating conditions.

CHAPTER 01 1.3-7 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 1.3-3 COMPARISON OF POWER CONVERSION SYSTEM DESIGN CHARACTERISTICS LGS* SSES ZIMMER 1 PBAPS TURBINE-GENERATOR (See UFSAR Section 10.2)

Design power, MWe (gross) 1092 1085 830 1098 Generator speed, rpm 1800 1800 1800 1800 Design steam flow, lb/hr (maximum 14.14x106 13.46x106 11.00x106 13.36x106 guaranteed)

Inlet pressure, psig 965 965 950 965 STEAM BYPASS SYSTEM (See UFSAR section 10.4.4)

Capacity, % design steam flow 25 25 25 25 MAIN CONDENSER (See UFSAR Section 10.4.1)

Heat removal capacity, Btu/hr 7725x106 7890x106 7053x106 7600x106 CIRCULATING WATER SYSTEM (See UFSAR Section 10.4.5)

Number of pumps 4 4 3 3 Flow rate, gpm/pump 113,000 112,000 150,000 250,000 CONDENSATE AND FEEDWATER SYSTEMS (See UFSAR Section 10.4.7)

Design flow rate, lb/hr 14.12x106 13.44x106 10.971x106 13.33x106 Number of condensate pumps 3 4 3 3 Number of condensate booster pumps - - 3 -

Number of feedwater pumps 3 3 2 3 Condensate pump drive AC power AC power AC power AC power Booster pump drive - - AC power -

Feedwater pump drive Turbine Turbine Turbine Turbine

• Based on the original licensed operating conditions.

CHAPTER 01 1.3-8 REV. 16, SEPTEMBER 2012

LGS UFSAR Table 1.3-4 COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS LGS* SSES ZIMMER 1 PBAPS PRIMARY CONTAINMENT (See UFSAR Sections 3.8 and 6.2)

Type Pressure- Suppression Pressure-suppression Pressure-suppression Pressure-suppression Construction Concrete with steel liner Concrete with steel liner Concrete with steel liner Free-Standing Steel Drywell Frustum of cone upper Frustum of cone upper Frustum of cone upper Light bulb shape, steel portion portion portion vessel Suppression chamber Cylindrical lower portion Cylindrical lower portion Cylindrical lower portion Torus, steel vessel Suppression chamber internal design 55 53 45 56 pressure, psig Suppression chamber external design 5 5 2 2 pressure, psi Drywell internal design pressure, psig 55 53 45 56 Drywell external design pressure, psi 5 5 2 2 Drywell free volume, ft3 (low water) 243,580 239,600 180,000 159,000 Suppression chamber free air volume, ft3 147,670 148,590 93,000 19,000 (high water) (high water) 159,540 159,130 (low water) (low water)

Suppression pool water volume, ft3 134,600(max) 131,550(max) 95,762 135,000 122,120(min) 122,410(min)

Submergence of downcomers below 121/4 11 10 4 suppression pool surface, ft (high water) 10 (low water)

CHAPTER 01 1.3-9 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 1.3-4 (Cont'd)

LGS* SSES ZIMMER 1 PBAPS Design temperature of drywell, F 340 340 340 281 Design temperature of suppression 220 220 275 281 chamber, F Downcomer vent pressure loss factor 2.23 2.5 2.17 6.21 Break area/total vent area 0.0159 0.016 0.008 0.019 Calculated maximum pressure after 44.0 44 40.4 40 blowdown to drywell, psig Calculated maximum suppression 30.6 29 35.6 25 chamber pressure after LOCA blowdown, psig Initial suppression pool temperature rise 43 40 35 32 during LOCA blowdown, F Leakage rate, % free volume/day 0.5 0.5 0.635 at 45 psig and 0.5 340F SECONDARY CONTAINMENT (See UFSAR Section 3.8)

Type Controlled leakage, roof Controlled leakage, Controlled leakage, Controlled leakage, level release elevated release elevated release elevated release Construction Lower levels Reinforced concrete Reinforced concrete Reinforced concrete Reinforced concrete Upper levels Reinforced concrete Steel super-structure and Steel super-structure and Steel super-structure and super-structure and siding siding siding siding Roof Reinforced concrete Steel decking Steel decking Steel decking

• Based on the original licensed operating conditions.

CHAPTER 01 1.3-10 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 1.3-4 (Cont'd)

LGS* SSES ZIMMER 1 PBAPS Internal design pressure, psig below 0.25 0.25 0.25 0.25 atmosphere Design inleakage rate % free 200 100 100 100 volume/day at 0.25 in wg. Reactor enclosure Refueling area 50 100 100 100

• Based on the original licensed operating conditions.

CHAPTER 01 1.3-11 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 1.3-5 COMPARISON OF STRUCTURAL DESIGN CHARACTERISTICS LGS SSES ZIMMER 1 PBAPS SEISMIC DESIGN (See UFSAR Section 3.7)

Operating basis earthquake

- horizontal(g) 0.075 0.05 0.10 0.05

- vertical(g) 0.05 0.033 0.07 0.033 Safe shutdown earthquake

- horizontal(g) 0.15 0.10 0.20 0.12

- vertical(g) 0.10 0.067 0.14 0.08 WIND DESIGN (See UFSAR Section 3.3)

Maximum sustained wind speed (mph) 90 80 90 87 TORNADO DESIGN (See UFSAR Section 3.3)

Translational speed (mph) 60 60 60 -

Rotational speed (mph) 300 300 300 300 CHAPTER 01 1.3-12 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 1.3-6 RADIOACTIVE WASTE MANAGEMENT SYSTEMS DESIGN CHARACTERISTICS LGS SSES ZIMMER 1 PBAPS GASEOUS RADWASTE (See UFSAR Section 11.3 Design bases, noble gases, Ci/sec 100, 000 at 30 min 100, 000 at 30 min 100, 000 at 30 min 100, 000 at 30 min Process treatment Recombiner and ambient Recombiner and Chilled charcoal Recombiner charcoal delay ambient charcoal delay Design condenser inleakage, cfm 75 30 12.5 54 Release point-height above ground, ft 197 201 172 500 LIQUID RADWASTE (See UFSAR Section 11.2)

Treatment of:

Floor drains F, D, R F, D, R F, E, R F, O Equipment drains F, D, R F, D, R F, D, R F, D, R Chemical drains F, D, R E, D concentrates to E, D concentrates to Neutralized radwaste, F, O solid radwaste, distillate solid radwaste, distillate R R Laundry drains F, O Diluted and sent to Reverse osmosis F, O circulating water discharge discharge (1) Legend:

D = demineralized F = filtered E = evaporator/concentrator R = recycled, i.e., returned to condensate storage O = discharged CHAPTER 01 1.3-13 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 1.3-7 COMPARISON OF ELECTRICAL POWER SYSTEMS DESIGN CHARACTERISTICS LGS SSES ZIMMER 1 PBAPS TRANSMISSION SYSTEM (See UFSAR Section 8.2)

Outgoing lines, number-rating 3-500 kV 1-230 kV 3-345 kV 4.500 kV 2-230 kV (Unit 1) 1-500 kV (Unit 2)

NORMAL AUXILIARY AC POWER (See UFSAR Sections 8.2 and 8.3)

Incoming lines, number-rating 3-500 kV 2.230 kV 1-69 kV 1-230 kV 2-230 kV (common to both units) 1-345 kV 13.8 kV Auxiliary transformers, number 2 1 per unit 1(unit auxiliary) 2 Startup transformers, numbers 2 2 (common to both units) 2 2 Safeguard transformers, numbers 2 + 1 spare - - 2 STANDBY AC POWER SUPPLY (See UFSAR Section 8.3)

Number of diesel generators 8 4 (common to both units) 3 4 Number of 4160 V shutdown buses 8 4 per unit 3 8 Number of 480 V shutdown buses 8 4 load centers and 8 8 8 MCCS per unit; 8 MCCs common to both units CHAPTER 01 1.3-14 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 1.3-7 (Cont'd)

LGS* SSES ZIMMER 1 PBAPS DC POWER SUPPLY (See UFSAR Section 8.3)

Number of 125 V or 250 V batteries 4-125 V 4-125 V 3-125 V 4-125/250 V per unit 6-125/250 V 2-250 V 1-250 V 2-250 V 2-250 V per unit Number of 125 V buses 33 4 per unit 3 8 Number of 250 V buses 8 buses and 8 MCCs 2 load centers and 3 1 8 MCCs per unit CHAPTER 01 1.3-15 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 1.3-8 SIGNIFICANT DESIGN CHANGES FROM PSAR TO FSAR UFSAR SECTION IN WHICH ITEM CHANGE REASON FOR CHANGE SUBJECT IS DISCUSSED Nuclear fuel The arrangement of fuel rods in This change improves fuel 4.2 each fuel bundle was changed performance by increasing safety from 7x7 to 8x8. margins.

Nuclear pressure relief system The two safety valves were This change improves the 5.2.2 deleted and the number of relief pressure suppression capability of valves was increased from 11 to the nuclear relief system.

14.

Feedwater sparger The design of the thermal This change eliminated the 5.3 sleeve connecting the sparger to possibility of vibration and leakage.

the reactor vessel has been changed.

Reactor recirculation system The 4 inch bypass line around This changed improves the 5.4 the recirculation pump discharge integrity and reliability of the valve has been deleted. recirculation loop.

Process sensors The monitoring devices of This change improves testability. 7.2, 7.3 process parameters have been changed (e.g., pressure, level, and flow) for the RPS and the ESF systems from process switches to transmitters and trip units.

Main condenser low vacuum The MSIV and main steam drain This eliminates reliance on the 7.2, 7.3 line isolation valves, rather than turbine control system for isolation the turbine stop valves and of the main condenser from the turbine bypass valves, are main steam supply.

automatically closed if there is loss of vacuum in the main condenser.

CHAPTER 01 1.3-16 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 1.3-8 (Cont'd)

UFSAR SECTION IN WHICH ITEM CHANGE REASON FOR CHANGE SUBJECT IS DISCUSSED Standby ac power system Bypassing of the following diesel To conform with BTP ICSB 17.

generator trip signals, if there is a LOCA, has been added:

a. Phase overcurrent

b. Ground neutral overcurrent

c. Ant-motoring

d. Jacket coolant high temperature/low pressure

e. Lube oil high temperature/low pressure Control room HVAC system The capability for total isolation For conformance with Regulatory 6.4, 9.4.1 and recirculation of control room Guides 1.78 and 1.95.

atmosphere on toxic chemical detection signal has been added.

Reactor enclosure HVAC The requirement for complete The reactor enclosure normal 9.4.2 isolation closure of the isolation valves in exhaust system serves those the reactor enclosure normal areas of the reactor enclosure exhaust duct before any which have a low release potential.

contaminated air is released Releases of radioactivity during after its detection has been the short time period required for deleted. Isolation valves will trip isolation valve closure would be a in normal time on detection of very small fraction of 10CFR50.67 airborne contamination. guidelines.

Drywell unit coolers Changed from two-speed to This change was made to obtain 9.4.5 single fan motors of sufficient motors qualified for 340oF horsepower for LOCA operation operation.

and ILRT operation.

CHAPTER 01 1.3-17 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 1.3-8 (Cont'd)

UFSAR SECTION IN WHICH ITEM CHANGE REASON FOR CHANGE SUBJECT IS DISCUSSED Reactor enclosure and control Added steam flooding isolation This prevents damage to 9.4.2 structure HVAC systems dampers safeguard equipment and protects personnel from pipe breaks outside containment.

Drywell HVAC duct-work Added pressure relief valves This prevents damage to ducts 9.4.5 during LOCA pressure surge in drywell.

Mechanical vacuum pump The pump exhaust was rerouted The charcoal filters will capture the 9.4.4 to the turbine equipment iodine discharged from the compartment exhaust duct vacuum pumps, as identified by leading to the charcoal filters EPRI Report NP-495.

and vent stack.

Mechanical vacuum pump The room air exhaust was The charcoal filters will capture 9.444 compartment changed from nonfiltered to iodine released to the room per filtered exhaust. EPRI Report NP-495.

Liquid Waste Management The radwaste evaporator Effective processing of chemical 11.2 System system and associated waste can be provided by the floor equipment was not completely drain subsystem, thereby installed for plant operation. eliminating operational problems associated with evaporative processing.

Gaseous waste management Change the offgas treatment The charcoal adsorption method 11.3 system system from cryogenic provides advantages in reliability distillation type to charcoal and maintainability, while limiting adsorption type. offsite doses to levels as low as or lower than can be achieved with the cryogenic distillation method.

CHAPTER 01 1.3-18 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 1.3-8 (Cont'd)

UFSAR SECTION IN WHICH ITEM CHANGE REASON FOR CHANGE SUBJECT IS DISCUSSED Solid radwaste management Because the evaporator was not Due to the previously described system completely installed for plant change in the liquid waste operation, a system for solidification management system, the evaporator of evaporator concentrates was also concentrates will not be produced.

not installed. Dewatered resin wastes will be packaged in approved high integrity containers consistent with the requirements established by the South Carolina Department of Health and Environmental Control and will meet the intent of BTP ESTB 11-3.

The vibrating hoppers were eliminated from system design because of anticipated operational problems and because they are not required for effective functioning of the solid radwaste system.

Primary containment instrument A seismic Category I backup gas To allow for long-term operation of 9.3.1, 5.4.7, 1.13 gas system supply for operation of ADS valves ADS valves to provide an alternate has been added. shutdown cooling path for conformance with Regulatory Guide 1.139. Also satisfies NUREG-0694 item.

Bypass leakage barrier design Added vent lines to potential bypass To minimize leakage in order to 6.2.3 leakage paths to vent to the minimize offsite doses in the event of secondary containment and added a a LOCA.

feedwater fill system to maintain a wate seal on the feedwater lines.

Residual heat removal system Addes intertie piping and valves 'A' To allow use of RHR pumps 'B' and 'D' 5.4 and 'B' and between RHR 'C' and 'D' in conjunction with the RHR heat discharge loops. exchanger in the shutdown cooling mode for greater system flexibility during maintenance.

CHAPTER 01 1.3-19 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 1.3-8 (Cont'd)

UFSAR SECTION IN WHICH ITEM CHANGE REASON FOR CHANGE SUBJECT IS DISCUSSED Ultimate heat sink Added soil-bentonite lining to the To minimize construction delays in 2.5, 9.2.6 spray pond bottom and soil event that the potential for greater slopes. than expected seepage rates was found.

Safeguard piping fill system Provided a redundant safety- To provide increased assurance 6.3 grade piping fill system. that water hammer events will not occur during ECCS system startup.

Spent fuel pool Added makeup water To provide Seismic Category I, 9.1 connections from the ESW safety-grade makeup water supply System. to the spent fuel pools to provide cooling in the event of loss of the spent fuel pool cooling system in conformance with Regulatory Guide 1.13.

Reactor recirculation system Deleted recirculation loops To improve the integrity and 5.4 equalizing line. reliability of the recirculation loops.

Reactor coolant pressure Changed portions of the RHR To minimize potential for IGSCC. 5.2.3 boundary piping LPCI, head spray, and shutdown cooling piping to type 316L stainless steel with less than 0.02% carbon. Changed recirculation loop piping to type 316K stainless steel.

Battery System Added a non-Class IE battery to To provide separation of Class 1E 8.3 power non-Class IE dc loads. and non-Class 1E loads in conformance with Regulatory Guide 1.75 CHAPTER 01 1.3-20 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 1.3-8 (Cont'd)

UFSAR SECTION IN WHICH ITEM CHANGE REASON FOR CHANGE SUBJECT IS DISCUSSED Reactor enclosure recirculation RERS no longer recirculates air Separation prevents the 9.4.2.1 system from the refueling area which is temperature and humidity of the permanently separated from the refueling area from affecting the reactor enclosure at el 352'. reactor enclosure below through Refueling area now exhausts RERS mixing, reduces the time for directly to the SGTS upon drawdown of the reactor enclosure isolation. by eliminating the possibility of mixing airborne activity through reactor enclosure 1, the refueling area, and reactor enclosure 2 following a LOCA.

Redundant Reactivity Control Added RRCS To reduce the possibility of an 7.6.1.8 & 7.6.2.8 System ATWS event and the automatically mitigate the consequences of an ATWS event by independently monitoring the RPV dome pressure and water level and automatically initiating the ARI, RPT, SLCS flow and feedwater runback.

Alternate rod insertion Added ARI solenoid valves to To provide independent solenoid 4.6.1.2.5.4 the scram valve pilot air header. valves to bleed air from the scram valve pilot air header on low water level or high dome pressure in the RPV when detected by the RRCS to increase the reliability of control rod insertion.

Recirculation Pump Trip Added redundant trip breakers To trip the reactor recirculation 7.1 & 7.6 to the power supply to each pump motors in the event of high reactor recirculation pump dome pressure or low water level motor. in the RPV, and thereby reduce the reactor power level in the event of an ATWS.

CHAPTER 01 1.3-21 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 1.3-8 (Cont'd)

UFSAR SECTION IN WHICH ITEM CHANGE REASON FOR CHANGE SUBJECT IS DISCUSSED HPCI Flow Split Added a cross connection so To ensure adequate mixing of 6.3.2.2.1 that HPCI flow is injected to the SLCS flow injected into the RPV.

RPV through both the core spray and feedwater spargers.

Standby Liquid Control System Added a third SLCS pump & To permit increased SLCS 9.3.5 explosive valve. maintenance flexibility.

Reactor Building Recirculation Added REPS Vent. This vent is needed to allow System simultaneous SGTS operation to the reactor enclosure and refueling area. The SGTS duct is seal welded to ensure RERS prefiltration after the RERS fan starts.

Standby Gas Treatment System SGTS filters have 8 inch deep The 8 inch deep SGTS charcoal 6.5.1.1.2 charcoal adsorber with and adsorber provides a residence assigned efficiency of 99.0% for time 0.68 seconds after drawdown.

removal of inorganic iodines. This is sufficient to meet the filtering efficiency specified in table 2 of Regulatory Guide 1.52 (Rev 2).

Reactor Recirculation Pump Both of the Unit 1 The ASD is more reliable and Table 1.3-8, 1.10-1, Table 3.2-1, Motor Power Supply Recirculation Pump Motors efficient than the MG Set for Sect. 3.8.4.1.8, Sect. 4.4.3.3.2, Power Supplies have been controlling Reactor Recirculation 4.4.3.5, Table 5.2-3, Sect. 5.4.1.3, changed from a Motor- Pump speed. Table 5.4-1, Sect. 7.1.1.2, Generator (MG) Set to an Table 7.1-1, Sect. 7.7.1.3.1.1, Adjustable Speed Drive (ASD). 7.7.1.3.2, 7.7.1.3.3.2.1, 7.7.1.3.3.3, 7.7.1.3.3.4, 7.7.1.3.3.4.3, 7.7.1.3.3.4.4, 7.7.1.3.3.4.5, 7.7.1.3.3.4.6, 7.7.1.3.3.4.7, 7.7.1.3.3.4.8, 7.7.1.3.3.5, 7.7.1.3.5.1, 7.7.1.3.5.2, 7.7.2.3.1, 7.7.2.3.3, Fig. 7.7-7, Sect. 8.1.6.1.12, Fig. 8.1-4, Table 8.3-28, Table 8.3-29, CHAPTER 01 1.3-22 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 1.3-8 (Cont'd)

UFSAR SECTION IN WHICH ITEM CHANGE REASON FOR CHANGE SUBJECT IS DISCUSSED Sect. 9.4.4.1, 9.4.4.2.4, 9.5.1.2.3.2, 9.5.1.2.9, 9.5.1.2.12, Sect. 9A.2.3, 9A.2.12, 9A.5.8.12, Table 9A-1, Fig. 9A-8, Sect. 15.3.1.1.1, 15.3.2.1.1, 15.3.2.2.3, 15.3.2.3.3, 15.4.5.1.1, 15.4.5.2.1, 15.4.5.3.2, 15.9.6.3.3 Reactor Recirculation Pump Recirculation Pump Motor The ASD is more reliable and Table 1.3-8, 1.10-1, Table 3.2-1, Motor Power Supply Power Supply has been efficient to improve system Sect. 3.8.4.1.8, Sect. 4.4.3.3.2, changed from a Motor- response to perturbations and 4.4.3.5, Sect. 5.4.1.3, Table 5.4-1, Generator (MG) Set to an transients. In addition, the Sect. 7.1.1.2, Table 7.1-1, Adjustable Speed Drive (ASD) ASD units are designed to be Sect. 7.7.1.3.1.1, 7.7.1.3.2.1, for Unit 2. The Unit 1 MG sets tolerant to single failures 7.7.1.3.2.2, 7.7.1.3.3.2.1, replacement was performed 7.7.1.3.3.3.1, 7.7.1.3.3.3.2, previously. 7.7.1.3.3.4.1, 7.7.1.3.3.4.2, 7.7.1.3.3.4.3, 7.7.1.3.3.4.4, 7.7.1.3.3.4.5, 7.7.1.3.3.4.6, 7.7.1.3.3.4.7, 7.7.1.3.3.4.8, 7.7.1.3.3.5, 7.7.1.3.5.1, 7.7.1.3.5.2, 7.7.2.3.1, 7.7.2.3.3, Fig. 7.7-7.1. Fig. 7.7-2.2, Sect. 8.1.6.1.12, Fig. 8.1-4, Table 8.3-28, Table 8.3-29, Sect. 9.4.4.1, 9.4.4.2.4, 9.5.1.2.3.2, 9.5.1.2.9, 9.5.1.2.12, Sect. 9A.2.3, 9A.2.12, 9A.5.8.25, Table 9A-1 Fig. 9A-8, Sect. 15.3.1.1.1, 15.3.2.1.1, 15.3.2.2.3, 15.3.2.3.3, 15.4.5.1.1, 15.4.5.2.1, 15.4.5.3.2, 15.9.6.3.3 CHAPTER 01 1.3-23 REV. 17, SEPTEMBER 2014

LGS UFSAR 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4.1 APPLICANT Exelon Generation Company, LLC (EGC) is the sole applicant for the utilization facility licenses and will operate the plant upon completion. Prime contractors and principal consultants are identified in Sections 1.4.2, through 1.4.5.

The Applicant has been responsible for the design and construction of, and currently operates, eight multi-unit fossil fuel power plants and one 2-unit nuclear power plant. Additionally, two large multi-unit hydroelectric generating plants, one of which is a pumped-storage plant, and several diesel engine and combustion turbine-driven generator installations have also been designed, constructed, and are currently being operated by the Applicant.

The Applicant also has an ownership interest in two operating fossil fuel plants and one operating nuclear power plant. These plants provide capacity and energy to the Applicant's system but are not operated by the Applicant.

These facilities, which had a net capacity of 8,197,600 KWe at the end of 1977, constitute the Applicant's electric generating system.

The Applicant has been active in the development of atomic energy for electric generation for many years. In 1952 it became a charter member of the Dow Chemical-Detroit Edison Nuclear Power Development Project, which subsequently became Atomic Power Development Associates, Inc. This organization designed and developed a fast breeder power reactor for the Atomic Energy Commission's Power Demonstration Program. The Applicant also took part in the formation of Power Reactor Development Company, which was organized to finance, construct, own, and operate the fast breeder reactor designed by Atomic Power Development Associates, Inc., for the Enrico Fermi Atomic Power Station.

The Applicant's engineers have had experience in many phases of nuclear projects including:

assignments to Atomic Power Development Associates, Inc. for design and development of core and fuel elements, shielding design, coordination of research at various levels on the metallurgical and chemical aspects of fuel elements; shift supervisor duties and preoperational duties, including preparation of plant operation manuals, at the Enrico Fermi Atomic Power Station; assignment to Power Reactor Development Company for coordination of control instrumentation and electrical features; assignment to the Nautilus nuclear submarine project for field engineering and mechanical operations during startup and initial operation; assignment to the nuclear reactor at Shippingport, Pennsylvania, for training and operational duties; assignment to the Knolls Atomic Power Laboratory for participation in prototype design of a sodium boiler for a submarine reactor; construction, startup, operation, and decommissioning of PBAPS Unit 1, a 40,000 KWe capacity unit employing an HTGR; and construction, startup, operation, and maintenance of PBAPS Units 2 and 3, each unit consisting of a BWR having an original rated core thermal power of 3293 KWt and producing a corresponding net electrical output of 1065 MWe.

1.4.2 ARCHITECT-ENGINEER AND CONSTRUCTOR The Applicant has retained Bechtel Power Corporation and Bechtel Construction, Inc. to provide architectural, engineering, construction, and startup services for LGS. In addition, Bechtel is CHAPTER 01 1.4-1 REV. 16, SEPTEMBER 2012

LGS UFSAR responsible for procurement of equipment other than the NSSS, turbine-generators, and certain other major components that have been purchased by the Applicant. Bechtel has been continuously engaged in engineering and construction activities since 1898. A review of recent tabulations of nuclear units in the continental United States that are planned, under construction, or in operation, indicates that Bechtel is responsible for the engineering design of approximately 60 of these units and, in addition, is charged with responsibility for construction of over 40 units.

Bechtel is, therefore, eminently qualified to provide the required services for station design, equipment procurement, construction, and startup.

1.4.3 NUCLEAR STEAM SUPPLY SYSTEM SUPPLIER GE has the contract to design, fabricate, and deliver the boiling water-type NSSS and nuclear fuel for LGS, as well as to provide technical direction for installation and startup of these systems. GE has been engaged in the development, design, construction, and operation of boiling water reactors since 1955. A review of recent tabulations of nuclear units in the United States that are planned, under construction, or in operation, reveals that approximately 65 of these units employ GE BWRs. Thus, GE has substantial experience, knowledge, and capability to design, manufacture, and furnish technical assistance for the installation and startup of the LGS NSSS.

1.4.4 TURBINE-GENERATOR SUPPLIER GE was the original contractor to design, fabricate, and deliver the turbine-generators for LGS, as well as to provide technical assistance for installation and startup of this equipment. GE has a long history in the application of turbine-generators to nuclear power stations dating back to 1955.

Over 100 of the nuclear units planned, under construction, or in operation in the United States employ GE turbine-generators. GE is, therefore, well qualified to design, fabricate, and deliver the turbine-generators for LGS, and to provide technical assistance for the installation and startup of this equipment. Siemens Power Corporation has the contract to design, fabricate and install the retrofit high pressure and low pressure turbines. Like GE, Siemens is well qualified to perform this work.

1.4.5 CONSULTANTS The licensee has engaged consultants to provide information and recommendations in a number of specialized fields. Principal consultants include:

a. Gilbert Associates: siting

b. Dames & Moore: geology, seismology, groundwater hydrology

c. Meteorological Evaluation Services, Inc.: meteorology

d. Buchart-Horn: archeology

e. MPR Associates, Inc.: quality assurance

f. C. L. Hosler: cooling tower studies

g. Radiation Management Corporation: aquatic and terrestrial biology, and radioactive releases and their effects

h. Nuclear Associates International Corp: core analysis CHAPTER 01 1.4-2 REV. 16, SEPTEMBER 2012

LGS UFSAR

i. Kibbe & Associates: nuclear fuel supply

j. Nuclear Energy Services: inservice inspection

k. J. E. Edinger: limnology

l. Hydrocon: Schuylkill River hydrology, radioactive release dispersion analysis

m. Betz-Converse-Murdoch, Inc.: sewage treatment

n. Sanders & Thomas: Schuylkill River soundings, road relocation

o. E. H. Bourquard Associates, Inc: water supply

p. Tippetts-Abbett-McCarthy-Stratton: water supply, hydrology CHAPTER 01 1.4-3 REV. 16, SEPTEMBER 2012

LGS UFSAR 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5.1 CURRENT DEVELOPMENT PROGRAMS 1.5.1.1 Instrumentation for Vibration of Reactor Intervals Vibration testing for reactor internals has been performed on virtually all GE BWR plants. At the time of issue of Regulatory Guide 1.20, test programs for compliance were instituted. The first BWR/4 plant of this size, Browns Ferry 1, is considered a prototype design and was instrumented and subjected to both cold and hot, two-phase flow testing to demonstrate that flow-induced vibrations similar to those expected during operation do not cause damage. Subsequent plants that have internals similar to those of the prototypes are tested in compliance with the requirements of Regulatory Guide 1.20 to confirm the adequacy of the design with respect to vibration. Combined with the system for monitoring of vibration within the reactor vessel and its immediate piping, additional equipment is provided to monitor selected rotating machines within the plant in order to provide early warning of excessive machine vibration.

1.5.1.2 Core Spray Distribution GE has a program underway to study BWR/6 core spray distributions using a combination of single nozzle steam and air tests, single and multiple nozzle analytical models, and full-scale air tests. This methodology was confirmed by a full-scale 30° sector steam test conducted during 1979 and reported in Reference 1.5-1. The NRC has agreed "that the overall empirical/engineering method outlined by GE ... is an acceptable method for verification of the currently assumed core spray distributions which are used to justify conservatisms of the spray cooling heat transfer coefficients in ECCS-LOCA licensing calculations."

For other BWR plants, the NRC believes that "there is a sufficient technical basis to permit continued plant operation and licensing in the interim period while these additional tests and information are being developed. This interim conclusion is based on:

a. The existence of a considerable safety margin between available and required spray flow indicated by preliminary analyses and measurements provided for each size BWR/1 through BWR/5;

b. The relative ease with which ECCS re-analyses could be performed to establish an acceptable power limit in the unlikely event that test results do not support the spray flows currently assumed;

c. The possibility that plants under construction could modify their spray nozzles or aiming pattern to provide a better spray distribution, if future test results indicate the desirability of such changes (particularly applicable to the BWR/6, where the type of preliminary measurements referenced in a. above are not yet available);

d. The existence of counter-current-flow-limiting phenomena in many plants would provide a steam/water layer on top of the core which should force a more even distribution of the core spray; CHAPTER 01 1.5-1 REV. 13, SEPTEMBER 2006

LGS UFSAR

e. The aforementioned empirical/engineering method is expected to provide timely confirmation of the spray flow margin presently believed to exist."

1.5.1.3 Core Spray and Core Flooding Heat Transfer Effectiveness Due to the incorporation of an 8x8 fuel rod array with unheated "water rods," tests have been conducted to demonstrate the effectiveness of the ECCS in the new geometry.

These tests are regarded as confirmatory only, since the geometry change is very slight and the "water rods" provide an additional heat sink in the inside of the bundle that improves heat transfer effectiveness.

There are two distinct programs involving the core spray. Testing of the core spray distribution has been accomplished and submitted (Reference 1.5-2). The other program concerns the testing of core spray and core flooding heat transfer effectiveness. The results of testing with stainless steel cladding are reported in Reference 1.5-3. The results of testing using Zircaloy cladding are reported in Reference 1.5-4.

1.5.1.4 Verification of Pressure-Suppression Design The Mark II pressure-suppression test program was initiated in the fall of 1975 to investigate suppression pool dynamic phenomena. Phase I blowdown tests were completed late in 1975.

These tests utilized a single 24 inch diameter (590 mm ID) downcomer that vented into a 7 foot (2.13 meters) inside diameter tank, representative of a single downcomer/pool cell in a typical Mark II suppression pool. The objective of this phase of testing was to quantify pool dynamics phenomena, particularly the effect of wetwell pressurization on pool swell, and the load associated with the low mass flux steam condensation, or chugging. Primary variables were simulated break size, initial vent submergence, and wetwell air space configuration, i.e., vented or closed wetwell.

The Phase II tests were generally similar to the Phase I tests, except a 20 inch diameter (489 mm ID) downcomer was used. The Phase I and II tests thus bound the range of vent-to-pool area ratios of all Mark II containments. Although the test objectives were similar during Phases I and II, some changes were made in the Phase II test matrix after review of the Phase I data. For example, since the Phase I test had shown that wetwell configuration was the variable that had the most pronounced effect on pool dynamics, the decision was made to concentrate the testing effort on the closed wetwell configuration, which is characteristic of the Mark II containment.

In place of the open wetwell tests, additional blowdowns were included in the Phase II test matrix in order to investigate the effect of saturated liquid versus saturated steam breaks and the effect of downcomer bracing configuration.

As was the case for the Phase I tests, the primary Phase II variables were simulated break size and initial vent submergence. The Phase III tests investigated the pool temperature sensitivity of pool swell and of the load associated with the chugging phenomenon. Only a single break size and vent submergence were tested, with pool temperature alone being a variable. A significant number of blowdowns were performed to yield a statistically significant data set.

CHAPTER 01 1.5-2 REV. 13, SEPTEMBER 2006

LGS UFSAR 1.5.1.5 Critical Heat Flux Testing A program for critical heat flux testing was established and was to be similar to that described in Reference 1.5-5. Since that time, however, a new analysis has been performed and the General Electric BWR Thermal Analysis Basis (GETAB) program initiated. The results of that analysis and related testing is described in Reference 1.5-6.

1.5.1.6 Fuel Assembly Structural Testing Although tests are being conducted to determine the effects of vibration on fuel assembly spacers and to determine the forces to which the assemblies are subjected during shipment, there is no special program at present concentrating on structural testing, and no topical report is anticipated.

1.

5.2 REFERENCES

1.5-1 "Core Spray Design Methodology Confirmation Test," NEDO-24712, (August 1979).

1.5-2 "BWR Core Spray Distribution," Licensing Topical Report NEDO-10846, (April 1973).

1.5-3 "Modeling the BWR/6 Loss-of-Coolant Accident: Core Spray and Bottom Flooding Heat Transfer Effectiveness," Licensing Topical Report NEDO-10801, (March 1973).

1.5-4 "Emergency Core Cooling Test of an Internally Pressurized, Zircaloy Clad, 8x8 Simulated BWR Fuel Bundle," Licensing Topical Report NEDO-20231, (December 1973).

1.5-5 "Design Basis for Critical Heat Flux Conditions in Boiling Water Reactors," APED-5286, (September 1966).

1.5-6 "General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application," Licensing Topical Report NEDO-10958-A, (January 1977).

CHAPTER 01 1.5-3 REV. 13, SEPTEMBER 2006

LGS UFSAR 1.6 MATERIAL INCORPORATED BY REFERENCE/GENERAL REFERENCE Table 1.6-1 provides a tabulation of topical reports and other documents incorporated by reference in this UFSAR.

Note: For clarification of the terms Incorporation by Reference and General Reference, refer to Reg. Guide 1.181, Content Of The Updated Final Safety Analysis In Accordance With 10 CFR 50.71(e), September, 1999 (endorsement of NEI 98-01, Revision 1, June, 1999).

CHAPTER 01 1.6-1 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 1.6-1 REFERENCED REPORTS Report Referenced in Number UFSAR Section A. General Electric Company Reports APED-4827 Maximum Two-Phase Vessel 6.2 Blowdown from Pipes (1965)

APED-5286 Design Basis for Criteria Heat 1.5 Flux Condition in BWRs (Sept 1966)

APED-5458 Effectiveness of Core Standby 5.4 Cooling Systems for General Electric Boiling Water Reactors (March 1968)

APED-5460 Design and Performance of General 3.9 Electric BWR Jet Pumps (July 1968)

APED-5555 Impact Testing on Collet 4.6 Assembly for Control Rod Drive Mechanism (7RDB144A) (Nov 1967)

APED-5652 Stability and Dynamic Performance 4.1 of the General Electric Boiling Reactor APED-5706 In-Core Neutron Monitoring System 7.2 For General Electric Boiling Water Reactors (Nov 1968, Revised April 1969)

APED-5750 Design and Performance of General 3.9, 5.4 Electric Boiling Water Reactor Main Steam Line Isolation Valves (March 1969)

APED-5756 Analytical Methods for Evaluating 15.4, 15.7 the Radiological Aspects of the General Electric Boiling Water Reactor (March 1969)

GEAP-5620 Failure Behavior in ASTM A 106B 5.2 Pipes Containing Axial Through-Wall Flows (April 1968)

GE-NE-L12- GE14 Fuel Design Cycle-Independent 9.1.6 00884-00-01P Analyses for Limerick Generating Station, Units 1 and 2, March 2001 J11-03898-01 GE14 Spent Fuel Storage Rack 9.1.6

-SFP Criticality Analysis for Limerick Generating Station, Unit 2, March, 2001 J11-03932-00 GE14 Spent Fuel Storage Rack 9.1.6

-SFP Criticality Analysis for the Limerick Generating Station, Unit 1, May, 2001 NEDC-32601P-A Methodology and Uncertainties for 4.1 Safety Limit MCPR Evaluations (latest approved revision)

NEDC-32868-P GE14 Compliance with Amendment 22 9.1.6 of NEDE-24011-P-A (GESTAR II)

(latest approved revision)

CHAPTER 01 1.6-2 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 1.6-1 (Cont'd)

Report Referenced in Number UFSAR Section NEDE-10813 PDA-Pipe Dynamic Analysis Program 3.6 for Pipe Rupture Movement (Proprietary Filing)

NEDE-10958 General Electric BWR Thermal 15.0 Analysis Basis (GETAB):

Data, Correlation, and Design Application (November 1973)

NEDE-20371-02 User Guide and Engineering 4.1 Description of HEATER Computer Program (March 1974)

NEDE-20566 Analytical Model for Loss-of 3.9, 6.3 Coolant Accident (LOCA) in Accordance with 10CFR50, Appendix K (Nov. 1975)

NEDE-20944-P BWR/4 and BWR/5 Fuel Design 1.8, 4.2 (October 1976) Amendment 1 4.3 (January 1977)

NEDE-21175-3-P BWR Fuel Assembly Evaluation 3.9 of Combined Safe Shutdown Earthquake (SSE) and Loss-of-Coolant Accident (LOCA)

Loadings (July 1982)

NEDE-21354-P BWR Fuel Channel Mechanical 3.9 Design and Deflection (September 1976)

NEDE-21821 Boiling Water Reactor 3.9 Feedwater Nozzle/Sparger Final Report (March 1978)

NEDE-24011-P-A General Electric Standard 4.1, 4.2 NEDE-24011-P- Application for Reactor Fuel, A-US including the United States Supplement (latest approved revision)

CHAPTER 01 1.6-3 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 1.6-1 (Cont'd)

Report Referenced in Number UFSAR Section NEDE-24057-P Assessment of Reactor 3.9 Internal Vibration in BWR/4 and BWR/5 Plants (November 1977)

NEDE-24222 Assessment of BWR Mitigation of 15.8 ATWS (NUREG-0460 Alternate 3)

(Volume 1 May 1979, Volume 2 December 1979)

NEDO-10173 Current State of Knowledge of 11.1 High Performance BWR Zircaloy-Clad UO2 Fuel (May 1970)

NEDO-10320 The General Electric Pressure 6.2 Suppression Containment Analytical Model (April 1971) Supplement 1 (May 1971)

NEDO-10466 Power Generation Control Complex 7.7, 8.1 Design Criteria and Safety Evaluation (Revision 2, March 1978)

NEDO-10505 Experience with BWR Fuel 11.1 Through September 1971 (May 1972)

NEDO-10527 Rod-Drop Accident Analysis for 15.4A Large Boiling Water Reactors (March 1972) Supplement 1 (July 1972) Supplement 2 (January 1973)

NEDO-10585 Behavior of Iodine in Reactor 5.2, 15.6 Water During Plant Shutdown and Startup (August 1972)

NEDO-10602 Testing of Improved Jet Pumps 3.9 for BWR/6 Nuclear System (June 1972)

NEDO-10739 Methods for Calculating Safe 6.3 Test Intervals and Allowable Repair Times for Engineered Safeguard Systems (January 1973)

CHAPTER 01 1.6-4 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 1.6-1 (Cont'd)

Report Referenced in Number UFSAR Section NEDO-10801 Modeling the BWR/6 Loss-of- 1.5 Coolant Accident: Core Spray and Bottom Flooding Heat Transfer Effectiveness (March 1973)

NEDO-10802 Analytical methods of Plant 5.2, 15.0, Transient Evaluations for 15.1 General Electric Boiling Water Reactor (April 1973)

NEDO-10846 BWR Core Spray Distribution 1.5 (April 1973)

NEDO-10871 Technical Derivation of BWR 11.1 1971 Design Basis Radioactive Source Terms (March 1975)

NEDO-10899 Chloride Control in BWR Coolants 5.2 (June 1973)

NEDO-10958-A General Electric BWR Thermal 1.5 Analysis Basis (GETAB): Data, Correlation, and Design Application (January 1977)

NEDO-10959 General Electric BWR Thermal 15.0 Analysis Basis (GETAB): Data, Correlation, and Design Application (November 1973)

NEDO-11209- Nuclear Energy Business Group 1.8 04A Boiling Water Reactor Quality Assurance Program (February 1980)

NEDO-12037 Summary of X-Ray and Gamma Ray 12.3 Energy and Intensity Data (January 1970)

NEDO-20231 Emergency Core Cooling Tests 1.5 of an Internally Pressurized, Zircaloy-Clad, 8x8 Simulated BWR Fuel Bundle (December 1973)

NEDO-20533 The General Electric Mark III 6A Pressure-Suppression Containment System Analytical Model (June 1974) Supplement 1 (September 1975)

CHAPTER 01 1.6-5 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 1.6-1 (Cont'd)

Report Referenced in Number UFSAR Section NEDO-20566 General Electric Company Model 3.9, 6.3 for Loss-of-Coolant Accident Analysis in Accordance with 10CFR50, Appendix K (January 1976)

NEDO-20922 Experience with BWR Fuel 11.1 Through September 1974 (June 1975)

NEDO-20944-1 BWR/4 and BWR/5 Fuel Design 4.2, 4.3 (October 1976) Amendment 1 (January 1977)

NEDO-20953 Three-Dimensional Boiling 15.4 Water Reactor Core Simulator (May 1976)

NEDO-21061 Mark II Containment Dynamic 3.8, 6.2 Forcing Function Information Report NEDO-21142 Realistic Accident Analysis 15.4, 15.6 The RELAC Code (October 1977)

NEDO-21143 Conservative Radiological 15.4 Accident Evaluation - The CONAC01 Code (March 1976)

NEDO-21159 Airborne Release from BWR's for 11.1 Environmental Impact Evaluation (March 1976)

NEDO-21231 Banked Position Withdrawal 15.4 Sequence (January 1977)

NEDO-21617A Analog Transmitter/Trip Unit 7.0, 7.2 System for Engineered Safe-Guard Sensor Inputs NEDO-21660 Experience with BWR Fuel 11.1 through December 1976 (July 1977)

CHAPTER 01 1.6-6 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 1.6-1 (Cont'd)

Report Referenced in Number UFSAR Section NEDO-21708 Radiation Effects in BWR 5.3 NEDO-21778-A Transient Pressure Rises 5.3 Affecting Fracture Toughness Requirements for BWRs (December 1978)

NEDO-23538 Users Manual for CRPLS01 4.1 Program (December 1976)

NEDO-24057 Assessment of Reactor Internals 3.9 Vibration in BWR/4 and BWR/5 Plants (November 1977)

NEDO-24154 Qualification of the One- 5.2, 15.0 Dimensional Core Transient Model for BWRs (October 1978)

NEDO-24708A Additional Information Required 1.13, 6.3 for NRC Staff Generic Report on BWRs, Rev. 1 (Dec. 1980)

NEDO-24712 Core Spray Design Methodology 1.5 Confirmation Tests (August 1979)

NEDO-24951 BWR Owners' Group NUREG-0737 1.13 Implementation: Analyses and Positions Submitted to the NRC (June 1981)

NEDO-31897 GE Nuclear Energy, "Generic 6.2, 15.0 NEDC-31897P-A Guidelines For GE Boiling Water Reactor Power Uprate,"

Class I (Non-proprietary),

February 1992; and Class III (Proprietary), May 1992.

NEDO-33484 GE Hitachi Nuclear Energy, "Safety 1.1, 1.2, 1.3, 1.4, NEDC-33484P Analysis Report for Limerick Generating 1.10, 3.6, 3.8, 3.9, Station Units 1 and 2 Thermal Power 4.3, 4.4, 5.2, 5.3, Optimization," Class I (Non-proprietary), 7.2, 7.7, 8.2, 9.1, March 2010; and Class III (Proprietary), 9.2, 10.1, 10.2, March 2010. 10.4, 15.0, 15.2, 15.6 NEDC-31585P BWROG Report for increasing MSIV 6.7, 15.6 Leakage Rate Limits and Elimination of Leakage Control Systems NEDM-10320 The GE Pressure Suppression 6.2 Containment Analytical Model,"

March 1971 CHAPTER 01 1.6-7 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 1.6-1 (Cont'd)

Report Referenced in Number UFSAR Section NEDO-20533 The General Electric Mark III 6.2 Pressure Suppression Containment System Analytical Model," June 1974.

NEDO-20566A General Electric Company 6.2 Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K -

Volume II, January 1976 NUREG-0800 U.S. Nuclear Regulatory 6.2 Commission Standard Review Plan, Section 6.2.1.1.C, "Pressure -

Suppression Type BWT Containments,"

Revision 6, August 1984 NRC Letter Letter from Ashok Thadani, 6.2 Director Division of Systems Safety and Analysis, Office of Nuclear Reactor Regulation, U.S.

Nuclear Regulatory Commission, to Gary L. Sozzi, Manager Technical Services, General Electric Nuclear Energy, "Use of SHEX Computer Program and ANSI/ANS 5.1-1979 Decay Heat Source Term For Containment Long-Term Pressure and Temperature Analysis,"

July 13, 1993.

NEDO-21061 Mark II Containment Dynamic 3.8, 6.2 Forcing Functions Information Report, Rev. 4, November 1981 NUREG-0487 U.S. Nuclear Regulatory 6.2 Commission "Mark II Containment Lead Plant Program Load Evaluation and Acceptance Criteria," October 1978 NUREG-0808 U.S. Nuclear Regulatory 6.2 Commission, "Mark II Containment Program Load Evaluation and Acceptance Criteria," August 1981 CHAPTER 01 1.6-8 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 1.6-1 (Cont'd)

Report Referenced in Number UFSAR Section U.S. Nuclear Regulatory 6.2 Commission, "Safety Evaluation Report Related to the Operation of Limerick Generating Station, Units 1 and 2," NUREG-0991, August 1983, and Supplements (Docket Nos.

50-352 and 50-353).

NEDO-30832 Elimination of Limit On Local 6.2 Suppression Pool Temperature For SRV Discharge With Quenchers, Class I, December 1984 GE Nuclear Energy, "Generic 6.2 Evaluations of General Boiling Water Reactor Power Uprate,"

Licensing Topical Report NEDC-31984P, Class III (Proprietary),

July 1991; NEDO-31984, Class I (Non-proprietary), March 1992; and Supplements 1 & 2.

D. Gobel, "Thermo-Hydraulic 6.2 Quencher Design of the Safety Relief System," Revision 1, R14-25/1978, Kraftwerk Union, April 1978 NEDC-32170P Limerick Generating Station 6.2 Units 1 and 2, SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis, Class III (Proprietary), Revision 1, June 1993 NEDC-32225P Power Rerate Safety Analysis 6.2 Report for Limerick Generating Station Units 1 & 2 (September 1993)

HI-2104779 Criticality Safety Evaluation for GNF2 9.1 Fuel in the SFP at Limerick, Revision 1 NEDC-33270P GNF2 Advantage Generic Compliance 9.1 with NEDE 24011-P-A (GESTAR II)

(latest approved revision)

NEDC-33627P GNF2 Fuel Design Cycle Independent 15.7 Analyses for Limerick Generating Station Units 1 and 2, Global Nuclear Fuels Document, (latest approved revision)

B. Other Reference Reports AI-75-2 Thermal Hydrogen Recombiner 6.2 System for Water-Cooled Reactors, Rockwell International (Rev.2(P),

July 1975)

CHAPTER 01 1.6-9 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 1.6-1 (Cont'd)

Report Referenced in Number UFSAR Section AI-77-55 Thermal Hydrogen Recombiner 6.2 System for Mark I and II Boiling Water Reactors, Rockwell International (Sept 1977)

BHR/DER 70-1 Radiological Surveillance 11.1 Studies at a Boiling Water Nuclear Power Reactor (March 1970)

OCF-1 Nuclear Containment Isolation 6.2 System, Owens-Corning Fiberglass Corporation (Jan 1979)

TID-4500 Relap 3 - A Computer Program 3.6 for Reactor Blowdown Analysis IN-1321 (June 1970)

C. Bechtel Power Corporation Reports BC-TOP-1 Containment Building Liner 3.8 Plant Design (Rev. 1 Dec 1972)

BC-TOP-4-A Seismic Analyses of Structures 3.7 and Equipment for Nuclear Power Plants (Rev. 3, November 1974)

BC-TOP-9A Design of Structures for Missile 3.3, 3.5 Impact (Rev. 2, Sept 1974)

BN-TOP-2 Design for Pipe Rupture Effects 3.6 (Rev. 2, May 1974)

BN-TOP-4 Subcompartment Pressure Analysis 3.6, 6A (Nov 1972) Rev. 1 (July 1976)

D. Exelon Reports Limerick Environmental Qualification Report 1.13, 3.11, 7.1, 8.1 Limerick Probabilistic Risk Assessment 15.11 Limerick Severe Accident Risk Assessment Limerick Environment Report - Operating 5.2, 2.3.2 License Stage CHAPTER 01 1.6-10 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 1.6-1 (Cont'd)

Report Referenced in Number UFSAR Section E. Siemens Reports ER 9605 Missile Probability 3.5, 10.2 Methodology for Limerick Generating Station, Units 1 and 2, with Siemens Retrofit Turbines Siemens Power Corporation Engineering Report, Revision 2, June 18, 1997 - SIEMENS PROPIETARY CT-27496 "Missile Analysis Report" 3.5, 10.2 Limerick Units 1 and 2, March 1, 2013 (Includes ER-9605, Rev. 2 as Appendix A)

CHAPTER 01 1.6-11 REV. 17, SEPTEMBER 2014