ML18116A111

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Redacted - Susquehanna Steam Electric Station, Units 1 & 2, Revision 68 to Final Safety Analysis Report, Chapter 1, Introduction
ML18116A111
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Site: Susquehanna  Talen Energy icon.png
Issue date: 10/16/2017
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SSES-FSAR Text Rev. 60 FSAR Rev. 64 1.1-1

1.1 INTRODUCTION

1.1.1 TYPE OF LICENSE This FSAR is submitted by PPL Susquehanna, LLC in support of its application for an operating license for Susquehanna Steam Electric Station (Susquehanna SES) Units 1 and 2.

1.1.2 IDENTIFICATION

OF APPLICANT Application is made by PPL Susquehanna, LLC, Two North Ninth Street, Allentown, Pennsylvania, 18101.

1.1.3 NUMBER

OF PLANT UNITS

The plant consists of two units which have a common control room, diesel generators and refueling floor, turbine operating deck, radwaste system, and other auxiliary systems.

1.

1.4 DESCRIPTION

OF LOCATION

The 2,355 acre plant site is located in Salem Township, Luzerne County, Pennsylvania, approximately 20 miles southwest of Wilkes-Barre, 50 miles northwest of Allentown and 70 miles northeast of Harrisburg.

1.1.5 TYPE OF NUCLEAR STEAM SUPPLY The Nuclear Steam Supply System for each unit consists of a General Electric Boiling Water Reactor, BWR/4 product line with a 3952 MWt nominal rating.

1.1.6 TYPE OF CONTAINMENT The containment is a pressure suppression type designated as Mark II. The drywell is a steel-lined concrete cone located above the steel-lined concrete cylindrical pressure suppression chamber.

The drywell and suppression chamber are separated by a concrete diaphragm slab which also serves to strengthen the entire system.

1.1.7 CORE THERMAL POWER LEVELS

The rated core thermal power for each unit is 3952 MWt. The nominal turbine generator output at 3952 MWt is 1300 MWe for both Unit 1 and Unit 2.

SSES-FSAR Text Rev. 60 FSAR Rev. 64 1.1-2 1.1.8 SCHEDULED FUEL LOAD AND OPERATION DATA Unit 1 original fuel load was on July 27, 1982 with a commercial operation date of June 8, 1983. Unit 2 original fuel load was March 28, 1984 with a commercial operation date of February 12, 1985.

1.1.9 FSAR ORGANIZATION The Susquehanna SES Final Safety Analysis Report (FSAR) has been organized using Regulatory Guide 1.70 Revision 2 (September, 1975).

The FSAR is divided into 18 chapters, using the same chapter, section, subsection, and paragraph headings that appear in the standard format.

Where information has been presented that has not been specifically requested by the standard format, the information is presented in the appropriate chapter as a section or subsection, and follows the information specifically requested by the standard format.

Tabulations of data are designated "tables" and are identified by the section number, followed by a dash and number of table according to its order in the text; e.g., Table 3.4-5 is the fifth table of Section 3.4. Drawings, pictures, sketches, curves, graphs, and engineering diagrams are identified as "figures" and are numbered in the same manner as tables.

This FSAR has been organized so that all figures and tables are at the end of each major section (with the exception of Chapter 15 Sections 15.1 through 15.5). The results of reload specific calculations for Chapter 15 Sections 15.1 through 15.5 (i.e., tables and figures) are included in Appendices 15C and 15D for Units 1 and 2, respectively. Appendix 15E contains similar information and analytical results associated with these sections for non-limiting events for the initial Susquehanna cycles for both Unit 1 and Unit 2. Reference reports or other documents, where applicable, have been tabulated in a separate subsection at the end of each Section. Reference numbers are listed sequentially under the section heading to which they apply. As an additional guide to the reader, Section 1.8 has been incorporated to provide a glossary of definitions, abbreviations, symbols, indices, legends, and other similar aids.

Contents of the overall FSAR are tabulated at the front of each volume. The level of breakdown corresponds to the numbered tabs. Each chapter is provided with the detailed table of contents for that chapter.

Amendments are issued as replacement sheets. The location of the amended material is indicated by a vertical bar in the margin of the page adjacent to the information. The amendment number and date appear at the left bottom corner of each page. Any change from the preceding issue, whether it be only a correction to a typographical error or deletion of a word or repositioning of words, is considered an amendment. Where an amendment bar appears adjacent to blank portions of a page, a deletion is indicated. Where pages have been changed only to reposition material with no change in content, only the amendment number and date are given.

A "List of Effective Pages," contained in its own binder, shows the current revision number of each individual page.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-1 1.2 GENERAL PLANT DESCRIPTION

1.2.1 PRINCIPAL

DESIGN CRITERIA The principal criteria for design, construction, and testing of the Susquehanna SES are summarized below. Specific criteria, codes and standards are addressed in Section 3.0.

1.2.1.1 General Design Criteria The Susquehanna SES design conforms to the requirements given in 10CFR50, Appendix A.

Specific compliance is discussed in Section 3.1.

1. The plant is designed, fabricated, and erected to produce electrical power in accordance with the codes, standards, and regulations as described in Section 3.1.
2. Safety related systems are designed to permit safe plant operation and to accommodate postulated accidents without endangering the health and safety of the public.

1.2.1.2 System Design Criteria

1.2.1.2.1 Nuclear System Criteria

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

throughout the design life of the fuel.

2. The fuel cladding, in conjunction with other plant systems, is designed to retain integrity throughout any abnormal operational transient.
3. Those portions of the nuclear system which form part of the reactor coolant pressure boundary are designed to retain integrity as a radioactive material barrier during normal operation following abnormal operational transients and accidents.
4. Heat removal systems are provided in sufficient capacity and operational adequacy 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.
5. Heat removal systems are provided to remove decay heat generated in the core under circumstances wherein the normal operational heat removal systems become inoperative. The capacity of such systems is adequate to prevent fuel clad damage.

The reactor is capable of being shut down automatically in sufficient time to permit decay heat removal systems to become effective following loss of operation of normal heat removal systems.

6. 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.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-2

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

1.2.1.2.2 Safety Related Systems Criteria 1.2.1.2.2.1 General

1. Safety systems act in response to abnormal operational transients so that fuel cladding retains its integrity as a radioactive material barrier.
2. Safety systems and engineered safety features act to ensure that no damage to the nuclear system process barrier results from internal pressures caused by abnormal operational transients or accidents.
3. Where positive, precise actions are required in immediate response to accidents, such actions are automatic and require no decision or manipulation of controls by operations personnel.
4. Essential safety actions are carried out by equipment of sufficient redundance and independence that no single failure of active components can prevent the required actions. For systems or components to which IEEE 279-1971 is 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.
5. Features of the station that are essential to the mitigation of accident consequences are designed, fabricated, and erected to quality standards which reflect the importance of the safety function to be performed.
6. The design of safety systems and engineered safety features includes allowances for environmental phenomena at the site.
7. Provision is made for control of active components of safety systems and engineered safety features from the control room.
8. Safety systems and engineered safety features are designed to permit demonstration of their functional performance requirements.

1.2.1.2.2.2 Containment and Isolation Criteria

1. 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.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-3 2. A secondary containment that completely encloses both primary containment and fuel storage areas is provided and is designed to act as a radioactive material barrier.

3. The primary and secondary containments, in conjunction with other engineered safety features, act to prevent radioactive material released from the containment volumes from exceeding the guideline values of applicable regulations.
4. 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.
5. Piping that 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 there is a threat of uncontrolled radioactive material being released. Such isolation is effected in time to prevent radiological effects from exceeding the values of applicable regulations.

1.2.1.2.2.3 Emergency Core Cooling System (ECCS) Criteria

1. ECCS systems are provided to limit fuel cladding temperature to temperatures below the onset of fragmentation (2200F) in the event of a loss of coolant accident.
2. The ECCS provides for continuity of core cooling over the complete range of postulated break sizes in the nuclear system process barrier.
3. The ECCS is diverse, reliable, and redundant.
4. Operation of the ECCS is initiated automatically when required regardless of the availability of off-site power supplies and the normal generating system of the plant.

1.2.1.2.3 Process Control Systems Criteria 1.2.1.2.3.1 Nuclear System Process Control Criteria

1. Control equipment is provided to allow the reactor to respond to limited load changes, major load changes and abnormal operational transients.
2. It is possible to control the reactor power level manually.
3. Control of the nuclear system is possible from a single location.
4. 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.
5. Interlocks, or other automatic equipment, are provided as a backup to procedural controls to avoid conditions requiring the actuation of nuclear safety systems or engineered safety features.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-4 6. If the control room is inaccessible, it is possible to bring the reactor from power range operation to a hot shutdown condition by manipulation of controls and equipment which are available outside of the control room. Furthermore, station design does not preclude the ability, in this event, to bring the reactor to a cold shutdown condition from the hot shutdown condition.

1.2.1.2.3.2 Power Conversion Systems Process Control Criteria

1. Controls are provided to maintain temperature and pressure to below design limitations. This system will result in a stable operation and response for all allowable variations.
2. Controls are designed to provide indication of system trouble.
3. Control of the power conversion system is possible from a single location.
4. Controls are provided to ensure adequate cooling of power conversion system equipment.
5. Controls are provided to ensure adequate condensate purity.
6. Controls are provided to regulate the supply of water so that adequate reactor vessel water level is maintained.

1.2.1.2.3.3 Electrical Power System Process Control Criteria

1. Controls are provided to ensure that sufficient electrical power is provided for startup, normal operation, prompt shutdown and continued maintenance of the station in a safe condition.
2. Control of the electrical power system is possible from a single location.

1.2.1.2.4 Electrical Power System Criteria

1. The station electrical power systems are designed to deliver the electrical power generated.
2. Sufficient normal 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 engineered safety features under postulated design basis accident conditions.
3. Standby electrical power sources are provided to allow prompt reactor shutdown and removal of decay heat under circumstances where preferred power is not available.

They provide sufficient power to all engineered safety features requiring electrical power.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-5 1.2.1.2.5 Fuel Handling and Storage Facilities

1. Fuel handling and storage facilities are located in the reactor building and are designed to preclude criticality and to maintain adequate shielding and cooling for spent fuel.

Additional spent fuel storage facilities are provided at the Independent Spent Fuel Storage Installation (ISFSI) located north of the Low Level Radwaste Holding Facility (LLRWHF). The ISFSI is described in detail in Section 11.7. Handling of spent fuel stored at the ISFSI is in the Reactor Building and is designed to preclude criticality and to maintain adequate shielding and cooling for spent fuel.

1.2.1.2.6 Auxiliary Systems Criteria

1. 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.

2. 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.

1.2.1.2.7 Power Conversion Systems Criteria Components of the power conversion systems are designed to fulfill the following basic objectives:

a) Generate electricity with the turbine generator from steam produced in the reactor, condense the exhaust steam in the condenser and return the condensed water to the reactor as heated feedwater with most of the non-condensable gases and impurities removed.

b) Ensure that any fission products or radioactivity associated with the steam and condensate during normal operation are safely contained inside the system or are released under controlled conditions in accordance with waste disposal procedures.

1.2.1.2.8 Radioactive Waste Disposal Criteria

1. Gaseous, liquid, and solid waste disposal facilities are designed so that the discharge of radioactive effluents, storage, and off-site shipment of radioactive material are made in accordance with applicable regulations.
2. These facilities include means for informing station operating personnel whenever operational limits on the release of radioactive material are exceeded.
3. A separate facility for interim on-site storage of low level radioactive waste material as of April 30, 1988 was included under the 10CFR Part 50 facility operating licenses.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-6 1.2.1.2.9 Shielding and Access Control Criteria

1. 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. The design and establishment of the above include conditions that deal with fission product release from failed fuel elements and contamination of station areas from system leakage.
2. The control room is shielded against radiation and has suitable environmental control so that occupancy under design basis accident conditions is possible.

1.2.2 PLANT

DESCRIPTION 1.2.2.1 Site Characteristics 1.2.2.1.1 Location and Size

In terms of its relationship to metropolitan areas, the plant site lies approximately 20 miles southwest of Wilkes-Barre, approximately 50 miles northwest of Allentown, and approximately 70 miles northeast of Harrisburg. The plant is a two unit, Boiling Water Reactor. Each unit has a nominal rating of 1300 MWe. It is located on a 1,574 acre property owned by PPL in Salem Township, Luzerne County, Pennsylvania, along the west bank of the Susquehanna River approximately 5 miles northeast of the borough of Berwick, Pennsylvania. In addition, 717 acres of the site are located on the east side of the river in Conyngham and Hollenback Township. Total site acreage is approximately 2,355 acres. There are no structures or facilities on the east side of the river with the exception of transmission lines and facilities. A map of the site area including major structures and facilities is provided as Figure 2.1-12. PPL owns the entire 1800 foot plant exclusion area (except for Township Route T-419) and has complete authority to regulate any and all access and activity within that area.

The property in Salem Township is open deciduous woodland, interspersed with grassland and orchards and is bounded on its eastern flank by the Susquehanna River, which has a low water elevation of 484 feet MSL in this vicinity. Much of the northern property boundary runs along the slopes of an east-west trending ridge rising to a maximum elevation of 1060 feet MSL. This ridge abuts a rolling plateau to the south which in turn falls gradually in an easterly direction toward the floodplain of the Susquehanna River. The plant site is located on this plateau at an approximate grade of 675 feet MSL. Rainfall runoff leads into two main valleys that form intermittent waterways draining to the Susquehanna River, east of the property.

Also, in Salem Township a portion of the long abandoned North Branch Canal runs north-south across the floodplain between the Susquehanna River and U.S. Route 11. Within the property limits, the northern portion of the canal is generally dry and overgrown with trees and shrubs whereas the southern portion contains stagnant water. A permanent 30 acre body of water named Lake "Took-a-While" is located just west of the canal. An approximate 400 acre recreation area has been developed on the floodplain.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-7 Acreage in Conyngham and Hollenback Townships of Luzerne County on the east side of the Susquehanna River is open to the public for hunting, fishing and hiking. One of the trails leads to the scenic view from Council Cup, a 700-foot high bluff overlooking the Susquehanna River valley.

A multiple-use land management program coordinated through a 10-year forest stewartship

plan is aimed at providing a mix of woodlands, farming, recreation, a wildlife habitat, timber production, and historical protection.

1.2.2.1.2 Road and Rail Access

US Route 11 runs north-south through the property along the western edge of the floodplain. In this vicinity, the highway has a pavement width of 36 feet and is of bitumen-topped concrete slab construction. Township road T419, which follows the toe of the east-west trending ridge described above, leads off US 11 to traverse the property and link with Township road T438 which passes through the western portion of the property. Both of these Township roads are paved in this vicinity. A railroad line on the floodplain parallels US 11 in traversing the property.

The North Shore Railroad Co. operates this line owned by The Commonwealth of Pennsylvania. The railroad is a single track, non-electrified line of standard gage. The nearest railroad station is at Berwick, 5 miles to the south. Access to the various facilities is provided as follows:

a) a MAIN and a SECONDARY ACCESS ROAD leading from US 11 at separate locations to serve the main power block and surrounding structures b) an ACCESS RAILROAD SPUR leading from the Conrail (Erie-Lackawanna) Railroad to serve the main power block and cooling tower areas

c) PLANT ROADS providing access to all structures as well as connecting with the Main and Secondary access roads d) a RIVER FACILITIES ACCESS ROAD leading from US 11 to the intake structure on the river bank

e) a PERIMETER PATROL ROAD paralleling and within the plant security fence

1.2.2.1.3 Description of Plant Environs

1.2.2.1.3.1 Geology and Soils The property is underlain by a series of tightly folded strata of Paleozoic age, trending generally northeast-southwest. Pleistocene glacial outwash deposits mantle much of the area, particularly in topographic depressions. Underlying these glacial deposits are strata of Devonian and older ages. The plant site area is underlain by a series of siltstone, sandstone and shale beds of the Hamilton and Susquehanna groups of Devonian age.

Soils in the area are derived from parent material of glacial origin. These soils are acidic, well drained and generally not well suited for agricultural purposes.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-8 1.2.2.1.3.2 Groundwater The two principal aquifer systems in the region are the unconsolidated glacial and alluvial deposits and the underlying bedrock formations. The glacial deposits consist of drift, till, and outwash materials and vary in permeability from very low to high. The thickness of these deposits is highly variable in the site vicinity, ranging from 1 or 2 feet to over 100 feet. Wells penetrating the bedrock produce water from Onondaga limestone and strata of the Hamilton siltstone group.

The groundwater table in the area is a subdued replica of the surface topography. At the site, the water table is found generally within 35 feet of the ground surface, usually just below the bedrock surface but sometimes within the overburden soils.

Groundwater is the primary source of water supply in the region. The plant potable water supply is obtained from groundwater. There will be no impact to surrounding wells (groundwater level) due to Plant usage, as documented in Dames & Moore report titled "Environmental Feasibility for Groundwater Supply at SSES" dated 9/24/86.

1.2.2.1.3.3 Hydrology The Susquehanna River Basin comprises an area of about 27,500 square miles in the states of New York, Pennsylvania and Maryland. The plant site is located on the west side of the Main Branch of the Susquehanna River, approximately 42 miles upstream from the confluence of the Main and West Branches at Sunbury, Pennsylvania. The Main Branch has its source at Otsego Lake about 35 miles southeast of Utica, New York. From Otsego Lake, the river flows generally southwest. The Lackawanna River joins it near Pittston, Pennsylvania. From there it flows past the site.

The extreme and average daily flows recorded at the gaging station at Wilkes-Barre, about 20 miles upstream from the site, are:

Flow Date (cfs) (gpm)

Minimum 528 2.38x10 5 September 27,1964 Average 13,000 5.85x10 6 70 years of record Maximum 345,000 1.55x10 8 June 24, 1972 (Hurricane Agnes)

The River's path is controlled by the geologic structure in the site area, following the regional folding and jointing pattern. Just north of the site area, the river cuts across the regional fold axis along a major joint set before swinging west-southwest and paralleling the regional fold

axis.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-9 The river gradient is approximately 1.5 feet per mile near the site. The river has long pools, short riffles and shallow bedrock flats. Scouring during spring floods has removed most silt deposits except in quiet pools. The river formerly had large deposits of coal silt, most of which have been removed by dredging. Below Shickshinny, a large pool extends downstream to a point near Mocanaqua; below this point the water is shallower and faster. The bed is rock and gravel, and the river is interspersed with islands. Both upstream and downstream portions of these islands have clean gravel bars.

1.2.2.1.3.4 Meteorology A modified continental-type climate prevails in the general area of the site. Normally, the frost-free season extends from late April to mid October. Minimum temperatures during December, January, and February usually are below freezing, but rarely dip below 0F. Maximum temperatures above 100F have seldom been recorded. The yearly relative humidity averages about 70 percent. Mean annual snowfall in the region is about 52 inches. In winter the area has about 40 percent of sunshine; the summer percentage is about 60 percent. Heavy thunderstorms have occasionally caused damage over limited areas and tornado-force winds have been reported.

The annual precipitation is about 38 inches. July is normally the wettest month, with an average rainfall of about 5 inches, and February is the driest, with about 2 inches.

The dominant wind is from the West-Southwest Sector.

1.2.2.1.3.5 Seismicity Of the very few earthquakes which have occurred in Pennsylvania during historical times, most have been in the southeastern part of the State.

Only minor damage has ever been recorded from earth movement in Pennsylvania, with the exception of the two disturbances at Wilkes-Barre in 1954. It is doubtful whether the latter were the direct result of an earthquake. Since the affected area was only 2000 square feet and no record was made of the disturbances at any of the nearest seismic stations, it is likely they were associated with mining activities and the readjustment of alluvial deposits.

Because of the small correlation between seismic activity and known faults or tectonics, the area can be said to constitute an inactive seismotectonic province.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-10 1.2.2.2 General Arrangement of Structures and Equipment The principal structures and facilities located at the Plant Site are shown on Figure 2.1-11.

The equipment arrangements for these structures are shown in Dwgs. M-220, Sh. 1, M-221, Sh. 1, M-222, Sh. 1, M-223, Sh. 1, M-224, Sh. 1, M-225, Sh. 1, M-226, Sh. 1, M-227, Sh. 1, M-230, Sh. 1, M-231, Sh. 1, M-232, Sh. 1, M-233, Sh. 1, M-234, Sh. 1, M-235, Sh. 1, M-236, Sh. 1, M-237, Sh. 1, M-240, Sh. 1, M-241, Sh. 1, M-242, Sh. 1, M-243, Sh. 1, M-244, Sh. 1, M-245, Sh. 1, M-246, Sh. 1, M-247, Sh. 1, M-248, Sh. 1, M-249, Sh. 1, M-250, Sh. 1, M-251, Sh. 1, M-252, Sh. 1, M-253, Sh. 1, M-254, Sh. 1, M-255, Sh. 1, M-256, Sh. 1, M-257, Sh. 1, M-258, Sh. 1, M-259, Sh. 1, M-260, Sh. 1, M-261, Sh. 1, M-270, Sh. 1, M-271, Sh. 1, M-272, Sh. 1, M-273, Sh. 1, M-274, Sh. 1, M-276, Sh. 1, M-280, Sh. 1, M-281, Sh. 1, M-282, Sh. 1, M-284, Sh. 1, M-5200, Sh. 1, and M-5200, Sh. 2.

1.2.2.3 Nuclear System The nuclear system includes a single cycle, forced circulation, General Electric Boiling Water Reactor producing steam for direct use in the steam turbine. Heat balances showing the major parameters of the nuclear system for the rated power condition, at rated core flow and at 108 Mlb/hr increased core flow, are shown in Figures 1.2-49 and 1.2-49-1 for Unit 1 and Figures 1.2-49-2 and 1.2-49-3 for Unit 2. The reactor heat balances differ slightly from the turbine heat balances (Figures 10.1-1a and 10.1-1b). The reactor heat balances are based on the measured moisture fraction exiting the reactor, and the moisture fraction exiting the MISV's is determined by the expected pressure drop between the reactor vessel steam dome and the MSIV exit. The turbine heat balance is based on turbine inlet design conditions, which allow for a slightly greater moisture fraction in the steam.

1.2.2.3.1 Reactor Core and Control Rods

The fuel for the reactor core consists of depleted, natural, and/or slightly enriched uranium dioxide pellets contained in sealed Zircaloy-2 tubes. These fuel rods are assembled into individual fuel assemblies. The number of fuel assemblies in the complete core is 764. Gross control of the core is achieved by movable, bottom entry control rods. The control rods are of cruciform shape and are distributed evenly throughout the core. The control rods are positioned by individual control rod drives.

Each fuel assembly has several fuel rods with gadolinia (Gd 2 O 3) mixed in solid solution with the UO 2. The gadolinia is burnable poison which diminishes the reactivity of the fresh fuel. It is depleted as the fuel reaches the end of its first cycle.

A conservative limit of plastic strain is the design criterion used for fuel rod cladding failure. The peak linear heat generation for steady-state operation is well below the fuel damage limit even late in life. Experience has shown that the control rods are not susceptible to distortion and have an average life expectancy many times the residence time of a fuel loading.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-11 1.2.2.3.2 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, core spray, and standby liquid control; the incore instrumentation; and other components. The main connections to the vessel include the steam lines, the coolant recirculation lines, the feedwater lines, the control rod drive 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 1050 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 core is cooled by reactor water that enters the lower portion of the core and boils as it flows upward around the fuel rods. The chemistry of reactor water is controlled to minimize corrosion of the fuel cladding, reactor vessel internals and reactor coolant system pressure boundary and to control the transport and deposition of corrosion product activity. 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 line(s). Each steam line is provided with two isolation valves in series, one on each side of the primary containment barrier.

1.2.2.3.3 Reactor Recirculation System

The Reactor Recirculation System pumps reactor coolant through the core to remove the energy generated in the fuel. This is accomplished by two recirculation loops external to the reactor vessel but inside the primary containment. 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.

1.2.2.3.4 Residual Heat Removal System

The Residual Heat Removal System (RHRS) consists of pumps, heat exchangers and piping that fulfill the following functions:

a. Removal of decay heat during and after plant shutdown.
b. Rapid injection of water into the reactor vessel following a loss of coolant accident, at a rate sufficient to reflood the core maintain fuel cladding below the limits contained in 10 CFR 50.46. This is discussed in Subsection 1.2.2.4.
c. Removal of heat from the primary containment following a loss-of-coolant accident (LOCA) to limit the increase in primary containment pressure. This is accomplished by cooling and recirculating the water inside the primary containment. The redundancy of

the equipment provided for the containment is further extended by a separate part of the RHRS which sprays cooling water into the drywell. This latter capability is discussed in Subsection 1.2.2.4.12.

d. Provide for cooling of the spent fuel pool(s) following a seismic event which results in a loss of normal spent fuel pool cooling, in conjunction with normal shutdown of both units.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-12

1.2.2.3.5 Reactor Water Cleanup System (RWCU)

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

1.2.2.4 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 allowable dose limits. These systems may be components, groups of components, systems, or groups of systems. Engineered Safety Feature (ESF) systems are included in this category. ESF systems have a sole function of mitigating the consequences of design basis accidents.

1.2.2.4.1 Reactor Protection System

The Reactor Protection System 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 Reactor Protection System overrides all operator actions and process controls.

1.2.2.4.2 Neutron-Monitoring System

Not all of the Neutron Monitoring System qualifies as a nuclear safety system; only those portions that provide high neutron flux signals and neutron flux oscillation signals to the Reactor Protection System are safety related. The intermediate range monitors (IRM), oscillation power range monitors (OPRM), and average power range monitors (APRM), which monitor neutron flux via in-core detectors, signal the Reactor Protection System in time to prevent excessive fuel clad temperatures as a result of abnormal operational transients.

1.2.2.4.3 Control Rod Drive System When a scram is initiated by the Reactor Protection System, the Control Rod Drive 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.

1.2.2.4.4 Nuclear System Pressure Relief System A Pressure Relief System, consisting of safety-relief valves mounted on the main steam lines, prevents excessive pressure inside the nuclear system following either abnormal operational transients or accidents.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-13 1.2.2.4.5 Reactor Core Isolation Cooling System The Reactor Core Isolation Cooling System (RCIC) provides makeup water to the reactor vessel whenever the vessel is isolated from the main condenser and feed water system. The RCICS uses a steam driven turbine-pump unit and operates automatically, in time and with sufficient

coolant flow, to maintain adequate reactor vessel water level.

1.2.2.4.6 Primary Containment A pressure-suppression primary containment houses the reactor vessel, the reactor coolant recirculating 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 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.

Hydrogen recombiners are provided in the drywell and wetwell to control combustible gases after a LOCA (not credited in the accident analysis).

1.2.2.4.7 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 which 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.

1.2.2.4.8 Secondary Containment

Any leakage from the primary containment system is to the secondary containment system.

This system includes the Standby Gas Treatment System and the Reactor Building Recirculation System. The secondary containment system is designed to minimize the release at ground level of airborne radioactive materials, and to provide for the controlled, filtered release of the reactor building atmosphere at roof level under accident conditions.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-14 1.2.2.4.9 Main Steam Line Isolation Valves Although process lines which 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 air 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.
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.

In addition the main steamline isolation valve leakage Isolated Condenser Treatment Method (Section 6.7) is provided to process the fission products after a LOCA. By directing the leakage from the closed main steamline isolation valves through the main steam drain line to the condenser, this leakage is processed prior to release to the atmosphere.

1.2.2.4.10 Main Steam Line Flow Restrictors

A venturi-type flow restrictor is installed in each steam line close to the reactor vessel. These devices limit the loss of coolant from the reactor vessel and prevent uncovering of the core before the main steam line isolation valves are closed in case of a main steam line break.

1.2.2.4.11 Emergency Core Cooling Systems (ECCS)

Four Core Standby Cooling Systems are provided to prevent excessive fuel clad temperatures if a breach in the nuclear system process barrier results in a loss of reactor coolant. The four Core Standby Cooling Systems are:

1. High Pressure Coolant Injection System (HPCIS)

The HPCIS provides and maintains an adequate coolant inventory inside the reactor vessel to prevent excessive fuel clad temperatures as a result of postulated small breaks in the Reactor Coolant Pressure Boundary (RCPB). A high pressure system is needed for such breaks because the reactor vessel depressurizes slowly, preventing low pressure systems from injecting coolant. The HPCIS 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.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-15 2. Automatic Depressurization System (ADS)

The ADS acts to rapidly reduce reactor vessel pressure in a LOCA situation in which the HPCIS fails to automatically maintain reactor vessel water level. The depressurization provided enables the low pressure standby cooling systems to deliver cooling water to the reactor vessel. The ADS uses some of the safety-relief valves which are part of the

Nuclear System Pressure Relief System. The automatic safety-relief valves are arranged to open when a break in the nuclear system process barrier has occurred and the HPCIS 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 Low Pressure Coolant Injection System Pumps are operating.

3. Core Spray System The Core Spray 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 breach exists in the nuclear system process barrier, 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 onto the core. Either core spray loop, together with another ECCS system, is capable of preventing excessive fuel clad temperatures following a LOCA.
4. Low Pressure Coolant Injection (LPCI)

Low Pressure Coolant Injection (LPCI) is an operating mode of the Residual Heat Removal System (RHRS) and is an engineered safety feature. LPCI uses the pump loops of the RHRS to inject cooling water at low pressure into a reactor recirculation loop. LPCI is actuated by conditions indicating a breach in the nuclear system process barrier, but water is delivered to the core only after reactor vessel pressure is reduced.

LPCI operation, together with the core shroud and jet pump arrangement, provides the capability of core reflooding following a LOCA in time to prevent excessive fuel clad temperatures.

1.2.2.4.12 Residual Heat Removal System (Containment Cooling)

The Residual Heat Removal System (RHRS) for containment cooling is placed in operation to limit the temperature of the water in the suppression pool following a design basis LOCA. In the containment cooling mode of operation, the RHRS pumps take suction from the suppression pool and deliver the water through the RHRS heat exchangers, where cooling takes place by transferring heat to the RHR service water. The fluid is then discharged back to the suppression pool.

As an alternative, RHRSW can be aligned to an RHR heat exchanger when RHR is aligned in the LPCI operating mode to support long term containment cooling.

Another portion of the RHRS is provided to spray water into the primary containment as a means of reducing containment pressure following a LOCA. This capability is in excess of the required energy removal capability and can be placed into service at the discretion of the

operator.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-16 1.2.2.4.13 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 control rod drive. The rate of reactivity insertion resulting from a rod drop accident is limited by this feature. The limiters contain no moving parts.

1.2.2.4.14 Control Rod Drive Housing Supports Control rod drive 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.

1.2.2.4.15 Reactor Building Recirculation and Standby Gas Treatment Systems The Reactor Building Recirculation System and the Standby Gas Treatment System (SGTS) are both a part of the secondary containment. The recirculation system has the capability of recirculating the reactor building air volume prior to its discharge via the SGTS, following a LOCA. Under normal wind conditions, the SGTS has the capability of maintaining a negative pressure within the reactor building with respect to the outside atmosphere. The air moving through the SGTS is filtered and discharged through the turbine building exhaust vent. 1.2.2.4.16 Standby ac Power Supply The Standby ac Power Supply System consists of four diesel-generator sets. The diesel-generators are sized so that three diesels can supply all the necessary power requirements for one unit in the design basis accident condition, plus the necessary required loads to effect the safe shutdown of the second unit. The diesel generators are specified to start up and attain rated voltage and frequency within 10 seconds. Four independent 4 kV engineered safety feature switchgear assemblies are provided for each reactor unit. Each diesel-generator feeds an independent 4 kV bus for each reactor unit.

Additionally, a spare diesel generator is provided which can be manually realigned as a replacement for any one of the other four diesel generators. This spare diesel generator has the emergency loading capability of any of the other four diesel generators.

Each diesel-generator starts automatically upon loss of off-site power or detection of a nuclear accident. The necessary engineered safety feature system loads are applied in a preset time sequence. Each generator operates independently and without paralleling during a loss of off-site power or LOCA signal.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-17 1.2.2.4.17 dc Power Supply Each reactor unit is provided with four independent 125 V and two independent 250 V dc systems. Each dc system is supplied from a separate battery bank and battery charger. The 125 V dc systems are provided to supply station dc control power and dc power to four diesel generators and their associated switchgears. The 250 V dc systems are provided to supply power required for the larger loads such as dc motor driven pumps and valves.

Additionally, a separate 125V dc system is provided for the spare diesel generator. This separate 125V dc system is provided to supply dc control power and dc power to the spare diesel generator auxiliaries and its associated switchgear.

The 125 and 250-V dc Systems are designed to supply power adequate to satisfy the engineered safety feature load requirements of the unit with the postulated loss of off-site power and any concurrent single failure in the dc system.

1.2.2.4.18 Residual Heat Removal Service Water System A Residual Heat Removal Service Water System is provided to remove the heat rejected by the Residual Heat Removal System during shutdown operation and accident conditions.

1.2.2.4.19 Emergency Service Water System The Emergency Service Water System supplies water to cool the standby diesel-generators and the ECCS and Engineered Safety Features equipment rooms, and other essential heat loads.

1.2.2.4.20 Main Steam Line Radiation Monitoring System

The Main Steam Line Radiation Monitoring System 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. A trip signal to the Mechanical Vacuum Pump (MVP) and its suction valves is generated by the monitors upon detection of a high high radiation signal.

1.2.2.4.21 Reactor Building Ventilation Radiation Monitoring System

The Reactor Building Ventilation Radiation Monitoring System consists of a number of radiation monitors arranged to monitor the activity level of the ventilation exhaust from the reactor building. Upon detection of high radiation, the reactor building is automatically isolated and the Standby Gas Treatment System is started.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-18 1.2.2.4.22 Nuclear Leak Detection System The Nuclear Leak Detection 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:

1) Main steam lines
2) Reactor water cleanup (RWCU) system
3) Residual heat removal (RHR) system
4) Reactor core isolation cooling (RCIC) system
5) High pressure coolant injection (HPCI) system
6) Instrument lines

Small leaks generally are detected by temperature and pressure changes, fillup rate of drain sumps, and fission product concentration inside the primary containment. Large leaks are also detected by changes in reactor water level and changes in flow rates in process lines.

1.2.2.5 Instrumentation and Control 1.2.2.5.1 Nuclear System Process Control and Instrumentation 1.2.2.5.1.1 Reactor Manual Control System

The Reactor Manual Control System provides the means by which control rods are manipulated from the control room for gross power control. The system controls valves in the Control Rod Drive Hydraulic System. Only one control rod can be manipulated at a time. The Reactor Manual Control System includes the controls that restrict control rod movement (rod block) under certain conditions as a backup to procedural controls.

1.2.2.5.1.2 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 in core power level. The system is arranged to automatically adjust reactor power output to the

load demand by adjusting the frequency of the electrical power supply for the reactor recirculation pumps.

1.2.2.5.1.3 Neutron Monitoring System

The Neutron Monitoring System is a system of in-core neutron detectors and out-of-core electronic monitoring equipment. The system provides indication of neutron flux, which can be correlated to thermal power level for the entire range of flux conditions that may exist in the core. The system also provides detection of neutron flux oscillations, which may indicate thermal-hydraulic instability. The source range monitors (SRM) and the intermediate range monitors (IRM) provide flux level indications during reactor startup and low power operation.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-19 The local power range monitors (LPRM), oscillation power range monitors (OPRM) and average power range monitors (APRM) allow assessment of local and overall flux conditions during power range operation. The average power range monitors also provide post-accident neutron flux information. A rod block monitor (RBM) is provided to prevent rod withdrawal when reactor power should not be increased at the existing reactor coolant flow rate. The Traversing In-core Probe System (TIPS) provides a means to calibrate the individual LPRM's.

1.2.2.5.1.4 Refueling Interlocks A system of interlocks, restricting the movements of refueling equipment and control rods when the reactor is in the refuel mode, 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 bridge, the refueling bridge hoists, the fuel grapple and control

rods.

1.2.2.5.1.5 Reactor Vessel Instrumentation

In addition to instrumentation provided for the nuclear safety systems and engineered safety features, instrumentation is provided to monitor and transmit information that can be used to assess conditions existing inside the reactor vessel and the physical condition of the vessel itself. The instrumentation provided monitors reactor vessel pressure, water level, temperature, internal differential pressures and coolant flow rates, and top head flange leakage.

1.2.2.5.1.6 Process Computer System An on-line process computer is provided to monitor and log process variables and to make certain analytical computations. In conjunction with approved operating procedure, the rod worth minimizer function prevents improper rod withdrawal under low power conditions. The effect of the rod worth minimizer function is to limit the reactivity worth of the control rods by enforcing adherence to the preplanned rod pattern.

1.2.2.5.1.7 Remote Shutdown System

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 main control room becomes uninhabitable.

1.2.2.5.2 Power Conversion Systems Process Control and Instrumentation

1.2.2.5.2.1 Pressure Regulator and Turbine Control

The pressure regulator maintains control of turbine control valves; it regulates pressure at the turbine inlet and, therefore, the pressure of the entire nuclear system. Pressure regulation is coordinated with the turbine speed system and the load control system so that rapid control valve closure can be initiated when necessary to provide turbine overspeed protection for large load rejection.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-20 1.2.2.5.2.2 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.

1.2.2.5.2.3 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.

1.2.2.6 Electrical Systems

1.2.2.6.1 Transmission and Generation Systems Redundant sources of off-site 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.

The two independent off-site sources provide auxiliary power for startup and for operating the engineered safety feature systems.

The main generator for each unit is an 1800-rpm, three-phase, 60-cycle synchronous machine rated at 1354 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 24 kV to 230 kV on Unit No. 1 and from 24 kV to 500 kV on Unit No. 2 by the unit main transformers and supplied by overhead lines to the 230 kV and 500 kV switchyards, respectively.

1.2.2.6.2 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 and some non-class 1E loads while the non-Class 1E system supplies the balance of plant equipment.

The Class 1E ac system for each unit consists of four independent load groups. Two independent off-site power systems provide the normal electric power to these groups. Each load group includes 4.16 kV switchgear, 480 V load centers, motor control centers and 120 V control and instrument power panel. The vital ac instrumentation and control power supply systems include battery systems, static inverters. Voltages listed are nominal values, and all electrical equipment essential to safety is designed to accept a range of +/-10 percent in voltage.

Four independent diesel generators are shared between the two units. Additionally, a spare diesel generator is provided which can be manually realigned as a replacement for any one of the other four diesel generators. This spare diesel generator has the emergency loading capability of any of the other four diesel generators. Each diesel generator is provided as a standby source of emergency power for one of the four Class 1E ac load groups in each unit.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-21 Assuming the total loss of off-site power 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 design basis accident on one unit and a forced shutdown of the second unit. The non-Class 1E ac system includes 13.8 kV switchgear, 4.16 kV switchgear, 480 V load centers and motor control centers. Four independent Class 1E 125V dc batteries and two independent Class 1E 250V 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 the Class 1E 125 and 250 V batteries.

An additional circuit breaker is provided for each non-class 1E load connected to the class 1E system for redundant fault protection. Additionally, a separate 125V dc system is provided for the spare diesel generator. This separate 125V dc system is provided to supply dc control power and dc power to spare diesel generator auxiliaries and its associated switchgear. These systems are discussed in Chapter 8. 1.2.2.7 Fuel Handling and Storage Systems 1.2.2.7.2 Fuel Pool Cooling and Cleanup System A Fuel Pool Cooling and Cleanup 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. 1.2.2.7.3 Fuel Handling Equipment The major fuel servicing and handling equipment includes the reactor building cranes, the refueling service platform, fuel and control rod servicing tools, fuel sipping and inspection devices, and other auxiliary servicing tools.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-22 1.2.2.8 Cooling Water and Auxiliary Systems 1.2.2.8.1 Service Water System The Service Water System is desig ned to a)Furnish cooling water to various heat exchangers located in the several plant buildings b)Furnish water for diluting the oxidizing and non-oxidizing biocides and for injecting th em into the circulating water systems. This is an intermittent service.The system consists of three 50 percent 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 that acts as the heat sink. Equipment that requires service water and is common to Units 1 and 2 is provided with inter-ties to both service water systems so that either can provide the water. 1.2.2.8.2 Residual Heat Removal Service Water System (RHRSWS)

The objective of the RHRSWS is to provide a reliable supply of cooling water for heat removal from the Residual Heat Removal System under post-accident conditions and supply a source of water if post-accident flooding of the core or primary containment is required. The system consists of two independent loops per unit, each of 100 percent capacity, and each loop consisting of two pumps, valves, piping and controls. Each loop uses the common spray pond with its spray distribution network as a heat sink. During operation the pumps take water from the spray pond and circulate it through the tube side of the RHR heat exchangers. The warm water is returned to the spray pond through a network of spray nozzles that produce the cooling effect by causing an enthalpy gradient as a result of the convective heat transfer and partial evaporative cooling of the spray droplets. A

radiation monitor is provided to check the radioactivity of the service water leaving each RHR heat exchanger. In the event of a high activity level (a tube leak in the RHR heat exchanger),

an alarm will sound and the operator will make the decision to isolate the heat exchanger and minimize the volume of contaminated water that flows to the spray pond.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-23 1.2.2.8.3 Emergency Service Water System (ESWS)

The objective of the ESWS is to supply cooling water to the RHR pumps and associated room coolers during normal and emergency conditions, as necessary, to safely shutdown the reactor or support normal and emergency conditions, as necessary, to safely shutdown the reactor or support "hot standby" conditions and in addition supply cooling water to the Diesel-Generator Units. The ESWS provides a reliable supply of cooling water to emergency equipment under a loss of off-site power condition or LOCA. The system consists of two independent loops supplying both units supplying both units (denoted "A" and "B") each of 100 percent capacity and containing two pumps, valves, piping and controls. Each loop uses the spray pond with its spray distribution system (common to both the Emergency Service Water and RHR Service Water Systems) as a heat sink. The ESWS is designed with sufficient redundancy so that no single active or passive system component failure can prevent it from achieving its safety objective. During operation, the ESWS pumps take water from the spray pond and circulate it through the various heat exchangers in the system. The warm water is returned to the spray pond through either a network of spray nozzles or directly through piping that bypass the spray arrays. The spray nozzles produce the cooling effect by causing an enthalpy gradient as a result of the convective heat transfer and partial evaporative cooling of the spray droplets.

1.2.2.8.4 Reactor Building Closed Cooling Water System

The Reactor Building Closed Cooling Water System is designed to accomplish the following objectives:

a) Provide cooling water to auxiliary plant equipment associated with the nuclear system and located in the reactor and radwaste buildings.

b) Provide cooling water to reactor building chilled water system in the event of unavailability of the chillers or loss of off-site power.

The Reactor Building Closed Cooling Water System consists of two 100 percent capacity pumps, two 100 percent capacity heat exchangers, a head tank, chemical addition tank, associated piping, valves and controls. The reactor building cooling water system is a closed loop cooling water system using inhibited demineralized water. The systems for Units 1 and 2

are separate from each other.

During normal plant operation one pump and heat exchanger will be in service, transferring heat to the service water system, with the other pump on automatic standby. Upon complete loss of off-site power, without occurrence of DBA, both cooling water pumps will start automatically when the buses are re-energized by Diesel Generators. The reactor building closed cooling water heat exchangers can be transferred from service water to emergency service water and one pump can be taken out of service, both by remote manual switching.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-24 1.2.2.8.5 Turbine Building Closed Cooling Water System (TBCCWS)

The TBCCWS is designed to provide cooling water to the auxiliary plant equipment associated with the nuclear and power conversion systems in the Turbine Building. The TBCCWS consists of two 100 percent capacity pumps, two 100 percent capacity heat exchangers, a head tank, chemical addition tank, associated piping and valves. The Turbine Building Cooling water System is a closed loop cooling water system using inhibited demineralized water. The systems for Units 1 and 2 are separate from each other. During normal plant operation, the turbine building closed cooling water heat exchanger transfers heat from the Turbine Building Closed Cooling Water System to the Service Water system. After a loss of off-site power, the pumps start automatically the pumps start automatically and the turbine building closed cooling water heat exchangers can be transferred by remote switching to Emergency Service Water System.

The heat load during this period will be rejected to the emergency service water. One turbine building closed cooling water pump and heat exchanger will be normally in service and the other pump will be on automatic standby.

1.2.2.8.6 Standby Liquid Control System Although not intended to provide rapid reactor shutdown, the Standby Liquid Control System provides a redundant, independent, and alternative method to the control rods to bring the reactor subcritical and to maintain it subcritical as the reactor cools. The system makes possible an orderly and safe shutdown if 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 system will also be used to buffer suppression pool pH to prevent iodine re-evolution following a postulated design basis loss of coolant accident.

1.2.2.8.7 Fire Protection System A Fire Protection System supplies fire fighting water to points throughout the plant. In addition, automatic Halon and carbon dioxide protection systems and portable fire extinguishers are also

provided.

1.2.2.8.8 Plant Heating, Ventilating, and Air-Conditioning Systems

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

1.2.2.8.9 Compressed Air System

The Compressed Air Systems (e.g., instrument air, service air and containment instrument air) supply air of suitable quality and pressure for various plant operations.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-25 1.2.2.8.10 Makeup Water Treatment System A Makeup Water Treatment System furnishes a supply of treated water suitable for use as makeup for the plant.

1.2.2.8.11 Domestic and Sanitary Water Systems

A water system for drinking and sanitary uses is provided for the plant.

1.2.2.8.12 Plant Equipment and Floor Drainage Systems

The Plant Equipment and Floor Drainage System handles both radioactive and non-radioactive drains. Drains which may contain radioactive materials are pumped to the radwaste system for cleanup, reuse, or discharge. Non-radioactive drains are discharged to the environs.

1.2.2.8.13 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.

1.2.2.8.14 Plant Communication System The Plant Communication System provides communication between various plant buildings and locations.

1.2.2.8.15 Process Valve Stem Leakoff System

The Process valve stem leak-off collection system is designed to reduce and control leakage to the atmosphere from valves greater than 2 1/2 in. that are used in the turbine building in systems containing radioactive steam or water and not connected to the main condenser.

Valves in the turbine building were originally provided with valve stem packing leakoff connections. Research and testing has shown that improved packing provides an effective seal to prevent leakage into the Turbine Building. As a result, these leakoff connections are in the process of being removed and package configurations changed, as appropriate, to conform with the new requirements. As part of this effort, leakoff isolation valves and piping will be removed (or abandoned in place) and the leakoff collection header piping will be removed or abandoned in place.

1.2.2.8.16 Diesel Auxiliary Systems

Diesel auxiliary systems are those systems which directly support operation of the emergency diesel generators. The following are diesel auxiliary systems:

a) Diesel Generator Fuel Oil Storage and Transfer System

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-26 b) Diesel Generator Cooling Water System

c) Diesel Generator Starting System d) Diesel Generator Lubrication System

e) Diesel Generator Combustion Air Intake and Exhaust System

1.2.2.8.17 Auxiliary Steam System The auxiliary steam system consists of two electrode steam boilers and auxiliary equipment.

The system is designed to provide flexibility for accommodating varying steam demands during all operating modes.

1.2.2.9 Power Conversion System

1.2.2.9.1 Turbine-Generator

The turbine-generator consists of the turbine, generator, exciter, controls, and required subsystems designed for a nominal plant rating output of 1300 MWe for both Unit 1 and Unit 2.

Each turbine is an 1800 rpm, tandem-compound, six-flow, non-reheat unit with an electrohydraulic control 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, three-phase, 60 Hz, 24,000 V, 1800 rpm, conductor-cooled, synchronous generator rated on the basis of guaranteed best turbine efficiency MW rating at 0.935 power factor, 75 psig hydrogen pressure. The generator-exciter system is shaft-driven, complete with static type voltage regulator and associated switchgear. The following are the turbine generator auxiliary systems:

a) Generator Hydrogen System b) Generator Seal Oil System c) Turbine Lube Oil System d) Steam Seal System e) Gland Exhaust System f) Generator Stator Cooling System

1.2.2.9.2 Main Steam System

The main steam system delivers steam from the nuclear boiler system via four 24 in. OD steam lines to the turbine-generator. This system also supplies steam to the steam jet air ejectors, the reactor feed pump turbines, the main condenser hotwell at startup and low loads, and the steam seal evaporator.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-27 1.2.2.9.3 Main Condenser 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, and each section is located below one of three low-pressure turbines. The condensers are supported on the turbine foundation mat, with rubber expansion joints provided between each turbine exhaust opening and the steam inlet connections in the condenser shells.

During normal operation, steam from the low pressure turbine is exhausted directly downward into the condenser shells through exhaust openings in the bottom of the turbine casings and is condensed. The condenser also serves as a heat sink for several other flows, such as exhaust steam from feed condenser drain, gland seal condenser drain, feedwater heater shell operating vents, and condensate pump suction vents.

During abnormal conditions the condenser is designed to receive (not simultaneously) turbine bypass steam, feedwater heater high level dump(s), and relief valve discharge (from crossover steam lines, feedwater heater shells, steam seal regulator, and various steam supply lines).

Other flows occur periodically; they originate from condensate pump and reactor feed pump startup vents, reactor feed pump and condensate pump minimum recirculation flows, feedwater line startup flushing, turbine equipment clean drains, low point drains, deaerating steam, makeup, condensate, etc.

1.2.2.9.4 Main Condenser Gas Removal System

The main condenser Gas Removal System removes the non-condensable gases from the main condenser and exhausts them to the Off-Gas System. One steam jet air ejector (100 percent capacity), is provided for the removal of air and radiolysis gases during normal operation. One motor-driven mechanical vacuum pump is to establish or maintain vacuum during startup and shutdown.

1.2.2.9.5 Steam Seal System

The steam seal system provides clean, non-radioactive steam to the seals of the turbine valve packings and the turbine shaft packings. The sealing steam is supplied by the seal steam evaporator. The auxiliary boiler provides an auxiliary steam supply for startup and when the seal steam evaporator is not operating.

1.2.2.9.6 Steam Bypass and Pressure Control System The turbine steam bypass and pressure control system control 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.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-28 c) During transient power operation when the reactor steam generation exceeds the turbine steam requirements.

d) When cooling down the reactor.

1.2.2.9.7 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 a natural draft cooling tower.

1.2.2.9.8 Condensate Cleanup System

The function of the Condensate Cleanup System is to maintain the required purity of the

feedwater flowing to the reactor.

The system consists of full flow deep bed demineralizers using ion exchange resins which remove dissolved and a portion of the suspended solids from the feedwater to maintain the purity necessary for the reactor. The demineralizers will also remove some of the radioactive material produced by corrosion as well as fission product carryover from the reactor. The radioactivity from these sources does not have a significant effect on the resins.

1.2.2.9.9 Condensate and Feedwater System The Condensate and Feedwater System is designed to deliver the required feedwater flow to the reactor vessels during stable and transient operating conditions throughout the entire operating range from startup to full load to shutdown. The system operates using four condensate pumps to pump deaerated condensate from the hotwell of the main condenser through the steam jet air ejector condenser, the gland steam condenser, the condensate filters, and thence to the condensate demineralizer. The demineralized feedwater then flows through three parallel strings of feedwater heaters, each string consisting of five heaters, to the suction of three reactor feed pumps which deliver the feedwater to the reactor.

1.2.2.9.10 Condensate and Refueling Water Storage and Transfer System

The function of the Condensate and Refueling Water Storage and Transfer System is to store condensate to be used as follows:

a) Supply water for the RCIC and HPCI systems.

b) Maintain the required condensate level in the hotwell either by receiving excess condensate rejected from the main condensate system or by supplying condensate to the main condensate system to makeup for a deficiency.

c) Fill up the reactor well of either reactor during refueling and receive this water back for storage after it has been cleaned up by the demineralizer.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-29 d) Provide condensate where required for miscellaneous equipment in the radwaste building and both reactor buildings.

The makeup to condensate storage tanks and the refueling storage tank is provided by the demineralized water storage tank.

1.2.2.10 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. Various methods are used to achieve this end, e.g. collection, filtration, holdup for decay, dilution and concentration.

1.2.2.10.1 Liquid Radwaste System The Liquid Radwaste System collects, treats, stores, and disposes of all 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 building 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 which 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.

1.2.2.10.2 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 off-site storage in approved shipping containers. Examples of these wastes are spent control rods, and in-core ion chambers.

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

If off-site shipment of solidified liners or dry active waste is not practicable, these items may be temporarily stored at the Low Level Radioactive Waste Holding Facility, as described in Section 11.6, provided they are packaged for off-site disposal.

1.2.2.10.3 Gaseous Radwaste System Radioactive gaseous wastes are discharged to the reactor building vent via the Gaseous Radwaste System. This system provides hydrogen-oxygen recombination, filtration, and holdup of the off-gases to ensure a low rate of release from the reactor building vent.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-30 The off-gases from the main condenser are the greatest source of gaseous radioactive waste. The treatment of these gases reduces the released activity to below permissible levels.

1.2.2.11 Radiation Monitoring and Control

1.2.2.11.1 Process Radiation Monitoring

Radiation monitors are provided on various lines to monitor for radioactive materials released to the environs via process liquids and gases or for detection of process system malfunctions. These monitors annunciate alarms and/or provide signals to initiate isolation and corrective actions.

1.2.2.11.2 Area Radiation Monitors Radiation monitors are provided to monitor for abnormal radiation at various locations in the reactor building, turbine building, and radwaste building. These monitors annunciate alarms when abnormal radiation levels are detected.

1.2.2.11.3 Site Environs Radiation Monitors

Radiation monitors are provided outside the plant buildings to monitor radiation levels. These data are used for determining the contribution of plant operations to on-site and off-site radiation levels.

1.2.2.11.4 Liquid Radwaste System Control Liquid wastes to be discharged are handled on a batch basis with protection against accidental discharge provided by procedural controls. Instrumentation, with alarms, to detect abnormal concentration of the radwastes, is provided.

1.2.2.11.5 Solid Radwaste Control

The Solid Radwaste System collects, treats, and prepares solid radioactive wastes for off-site shipment. Wastes are handled on a batch basis. Radiation levels of the various batches are

determined by the operator.

1.2.2.11.6 Gaseous Radwaste System Control

The Gaseous Radwaste System is continuously monitored by the turbine building vent radiation monitor and the off-gas pre-treatment radiation monitor. A high level signal will annunciate

alarms.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-31 1.2.2.12 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 and 10CFR50. It is also designed to protect certain plant components from radiation exposure resulting in unacceptable alterations of material properties or activation.

1.2.2.13 Steam Dryer Storage

A separate Steam Dryer Storage Facility(SDSF) is provided within the plant protected area for the storage, shielding, and radioactive decay of replaced Reactor steam dryers. The steam dryers are cut in half and packaged into steel containers for storage in the SDSF. They are not considered as radioactive waste but are treated as irradiated plant equipment. The SDSF is a reinforced concrete vault with removable roof slab access only, meeting 10CFR20 dose limits.

1.2.2.14 FLEX Equipment Storage Building

A separate FLEX Equipment Storage Building is provided within the plant protected area for the storage of portable equipment needed to respond to a Beyond Design Basis External Event (BDBEE). B.5.b equipment (i.e., pumper truck, etc.) is also stored in this building. This is strictly an emergency equipment storage facility (no personnel occupancy amenities) constructed to meet all plant extreme environmental conditions (i.e., seismic, tornado, missile).

THIS FIGURE HAS BEEN REPLACED BY DWG.

M-220, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-1 replaced by dwg. M-220, Sh. 1 FIGURE 1.2-1, Rev. 57 AutoCAD Figure 1_2_1.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-221, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-2 replaced by dwg. M-221, Sh. 1 FIGURE 1.2-2, Rev. 56 AutoCAD Figure 1_2_2.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-222, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-3 replaced by dwg. M-222, Sh. 1 FIGURE 1.2-3, Rev. 48 AutoCAD Figure 1_2_3.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-223, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-4 replaced by dwg. M-223, Sh. 1 FIGURE 1.2-4, Rev. 48 AutoCAD Figure 1_2_4.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-224, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-5 replaced by dwg. M-224, Sh. 1 FIGURE 1.2-5, Rev. 48 AutoCAD Figure 1_2_5.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-225, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-6 replaced by dwg. M-225, Sh. 1 FIGURE 1.2-6, Rev. 48 AutoCAD Figure 1_2_6.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-226, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-7 replaced by dwg. M-226, Sh. 1 FIGURE 1.2-7, Rev. 48 AutoCAD Figure 1_2_7.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-227, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-8 replaced by dwg. M-227, Sh. 1 FIGURE 1.2-8, Rev. 48 AutoCAD Figure 1_2_8.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-230, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-9 replaced by dwg. M-230, Sh. 1 FIGURE 1.2-9, Rev. 57 AutoCAD Figure 1_2_9.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-231, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-10 replaced by dwg. M-231, Sh. 1 FIGURE 1.2-10, Rev. 49 AutoCAD Figure 1_2_10.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-232, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-11 replaced by dwg. M-232, Sh. 1 FIGURE 1.2-11, Rev. 48 AutoCAD Figure 1_2_11.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-233, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-12 replaced by dwg. M-233, Sh. 1 FIGURE 1.2-12, Rev. 49 AutoCAD Figure 1_2_12.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-234, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-13 replaced by dwg. M-234, Sh. 1 FIGURE 1.2-13, Rev. 48 AutoCAD Figure 1_2_13.doc THiS FIGURE HAS BEEN REPLACED BY DWG.

M-235, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-14 replaced by dwg. M-235, Sh. 1 FIGURE 1.2-14, Rev. 48 AutoCAD Figure 1_2_14.doc THiS FIGURE HAS BEEN REPLACED BY DWG.

M-236, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-15 replaced by dwg. M-236, Sh. 1 FIGURE 1.2-15, Rev. 48 AutoCAD Figure 1_2_15.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-237, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-16 replaced by dwg. M-237, Sh. 1 FIGURE 1.2-16, Rev. 48 AutoCAD Figure 1_2_16.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-240, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-17 replaced by dwg. M-240, Sh. 1 FIGURE 1.2-17, Rev. 48 AutoCAD Figure 1_2_17.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-241, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-18 replaced by dwg. M-241, Sh. 1 FIGURE 1.2-18, Rev. 55 AutoCAD Figure 1_2_18.doc

THIS FIGURE HAS BEEN REPLACED BY DWG.

M-243, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-20 replaced by dwg. M-243, Sh. 1 FIGURE 1.2-20, Rev. 55 AutoCAD Figure 1_2_20.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-244, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-21 replaced by dwg. M-244, Sh. 1 FIGURE 1.2-21, Rev. 56 AutoCAD Figure 1_2_21.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-245, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-22 replaced by dwg. M-245, Sh. 1 FIGURE 1.2-22, Rev. 48 AutoCAD Figure 1_2_22.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-246, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-23 replaced by dwg. M-246, Sh. 1 FIGURE 1.2-23, Rev. 48 AutoCAD Figure 1_2_23.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-247, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-24 replaced by dwg. M-247, Sh. 1 FIGURE 1.2-24, Rev. 48 AutoCAD Figure 1_2_24.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-248, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-25 replaced by dwg. M-248, Sh. 1 FIGURE 1.2-25, Rev. 48 AutoCAD Figure 1_2_25.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-249, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-26 replaced by dwg. M-249, Sh. 1 FIGURE 1.2-26, Rev. 48 AutoCAD Figure 1_2_26.doc

THIS FIGURE HAS BEEN REPLACED BY DWG.

M-251, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-28 replaced by dwg. M-251, Sh. 1 FIGURE 1.2-28, Rev. 56 AutoCAD Figure 1_2_28.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-252, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-29 replaced by dwg. M-252, Sh. 1 FIGURE 1.2-29, Rev. 48 AutoCAD Figure 1_2_29.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-253, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-30 replaced by dwg. M-253, Sh. 1 FIGURE 1.2-30, Rev. 55 AutoCAD Figure 1_2_30.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-254, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-31 replaced by dwg. M-254, Sh. 1 FIGURE 1.2-31, Rev. 55 AutoCAD Figure 1_2_31.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-255, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-32 replaced by dwg. M-255, Sh. 1 FIGURE 1.2-32, Rev. 48 AutoCAD Figure 1_2_32.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-256, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-33 replaced by dwg. M-256, Sh. 1 FIGURE 1.2-33, Rev. 48 AutoCAD Figure 1_2_33.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-257, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-34 replaced by dwg. M-257, Sh. 1 FIGURE 1.2-34, Rev. 48 AutoCAD Figure 1_2_34.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-258, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-35 replaced by dwg. M-258, Sh. 1 FIGURE 1.2-35, Rev. 48 AutoCAD Figure 1_2_35.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-259, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-36 replaced by dwg. M-259, Sh. 1 FIGURE 1.2-36, Rev. 48 AutoCAD Figure 1_2_36.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-260, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-37 replaced by dwg. M-260, Sh. 1 FIGURE 1.2-37, Rev. 48 AutoCAD Figure 1_2_37.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-261, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-38 replaced by dwg. M-261, Sh. 1 FIGURE 1.2-38, Rev. 48 AutoCAD Figure 1_2_38.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-270, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-39 replaced by dwg. M-270, Sh. 1 FIGURE 1.2-39, Rev. 56 AutoCAD Figure 1_2_39.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-271, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-40 replaced by dwg. M-271, Sh. 1 FIGURE 1.2-40, Rev. 48 AutoCAD Figure 1_2_40.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-272, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-41 replaced by dwg. M-272, Sh. 1 FIGURE 1.2-41, Rev. 48 AutoCAD Figure 1_2_41.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-273, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-42 replaced by dwg. M-273, Sh. 1 FIGURE 1.2-42, Rev. 48 AutoCAD Figure 1_2_42.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-274, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-43 replaced by dwg. M-274, Sh. 1 FIGURE 1.2-43, Rev. 48 AutoCAD Figure 1_2_43.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-276, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-44 replaced by dwg. M-276, Sh. 1 FIGURE 1.2-44, Rev. 48 AutoCAD Figure 1_2_44.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-280, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-45 replaced by dwg. M-280, Sh. 1 FIGURE 1.2-45, Rev. 48 AutoCAD Figure 1_2_45.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-281, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-46 replaced by dwg. M-281, Sh. 1 FIGURE 1.2-46, Rev. 48 AutoCAD Figure 1_2_46.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-282, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-47 replaced by dwg. M-282, Sh. 1 FIGURE 1.2-47, Rev. 48 AutoCAD Figure 1_2_47.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-284, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-48 replaced by dwg. M-284, Sh. 1 FIGURE 1.2-48, Rev. 48 AutoCAD Figure 1_2_48.doc AutoCAD: Figure Fsar 1_2_49.dwg FSAR REV.67 FIGURE 1.2-49, Rev 60 UNIT 1 HEAT BALANCE AT RATED POWER WITH 100 X 10 LBm/hr CORE FLOW SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT THIS FIGURE HAS BEEN REPLACED BY DWG.

M-5200, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-50 replaced by dwg. M-5200, Sh. 1 FIGURE 1.2-50, Rev. 52 AutoCAD Figure 1_2_50.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-5200, Sh. 2 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-51 replaced by dwg. M-5200, Sh. 2 FIGURE 1.2-51, Rev. 52 AutoCAD Figure 1_2_51.doc THIS FIGURE HAS BEEN INTENTIONALLY LEFT BLANK FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure intentionally left blank FIGURE 1.2-52, Rev. 51 AutoCAD Figure 1_2_52.doc AutoCAD: Figure Fsar 1_2_49_1.dwg FSAR REV.67 FIGURE 1.2-49-1, Rev 60 UNIT 1 HEAT BALANCE AT RATED POWER WITH 108 X 10 LBm/hr INCREASED CORE FLOW SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT 6

AutoCAD: Figure Fsar 1_2_49_2.dwg FSAR REV.67 FIGURE 1.2-49-2, Rev 3 UNIT2 HEAT BALANCE AT RATED POWER WITH 100 X 10 LBm/hr CORE FLOW SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT 6

AutoCAD: Figure Fsar 1_2_49_3.dwg FSAR REV.67 FIGURE 1.2-49-3, Rev 3 UNIT 2 HEAT BALANCE AT RATED POWER WITH 108 X 10 LBm/hr INCREASED CORE FLOW SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT 6

SSES-FSAR Rev. 49, 04/96 1.3-1

1.3 COMPARISON

TABLES

1.3.1 COMPARISONS

WITH SIMILAR FACILITY DESIGNS

This subsection highlights the historical principal design features of the plant and compares its major features with other boiling water reactor facilities. The design of this facility was based on proven technology obtained during the development, design, construction, and operation of boiling water reactors of similar types.

The data, performance, characteristics, and other information presented here represent the then current Susquehan na Steam Electric Station desi gn as it compared to similar designs available at that time. To maintain this original design comparison, these tables will not be revised to reflect current plant des ign, other than the previous addition of the fifth ("E") emergency diesel genera tor to Tables 1.3-6 and 1.3-7.

1.3.1.1 Nuclear Steam Supply System Design Characteristics

Table 1.3-1 summarizes the design and operati ng characteristics for the nuclear steam supply systems. Parameters are related to rated power output for a single plant unless otherwise noted.

1.3.1.2 Power Conversion System Design Characteristics Table 1.3-2 compares the power conv ersion system design characteristics.

1.3.1.3 Engineered Safety Features Design Characteristics

Table 1.3-3 compares the engineered safety features design characteristics.

1.3.1.4 Containment De sign Characteristics

Table 1.3-4 compares the contai nment design characteristics.

1.3.1.5 Radioactive Waste Managem ent Systems Design Characteristics

Table 1.3-5 compares the radioactive waste management design characteristics.

SSES-FSAR Rev. 49, 04/96 1.3-2 1.3.1.6 Structural Design Characteristics Table 1.3-6 compares the struct ural design characteristics.

1.3.1.7 Instrumentation and Electrical Systems Design Characteristics

Table 1.3-7 compares the inst rumentation and electrical systems design characteristics.

1.3.2 COMPARISON

OF FINAL AND PRELIMINARY INFORMATION

All of the significant changes that have been made in the facility design between submission of the last PSAR revision and Revision 0 of the FSAR are listed in Table 1.3-8. Each item in Table 1.3-8 is cross-referenced to the appropriate portion of the FSAR which describes the changes and the bases for them.

SSES-FSARTABLE1.3-3COMPARISON OF ENGINEERED SAFETY FEATURES DESIGN CHARACTERISTICS Page 1 of 2 Rev. 49, 04/96 SSES-FSARTABLE1.3-3COMPARISON OF ENGINEERED SAFETY FEATURES DESIGN CHARACTERISTICS Page 2 of 2 Rev. 49, 04/96

SSES-FSAR Rev. 49, 04/96 Page 1 of 3 TABLE 1.3-8 SIGNIFICANT DESIGN CHANG ES FROM PSAR TO FSAR*

ITEM CHANGE REASON FO R CHANGE FSAR PORTION IN WHICH CHANGE IS DISCUSSED Recirculation flow measurement The recirculation flow measurement design was changed from a flow element to an elbow-tap type. To improve flow measurement accuracy. 7.3.1, 7.6.1 Recirculation system The pressure interlock for RHR shutdown mode was changed. NRC Requirement for diversity. 7.3.1, 7.6.1 Nuclear fuel The number of fuel pins in each fuel bundle has been changed from 7 x 7 to 8 x 8. Improved fuel performance by increasing safety margins.

4.2 Nuclear

boiler An additional test mode was added for closing MSIV's one at a time to 90% of full open in the fast mode (close in slow mode already

existed). Verifies that the spring force on the

valves will cause them to close under loss-of-air conditions.

5.4 Main steam line isolation A main condenser low vacuum initiation of the main steam line isolation was added. NRC requirement 7.3.1 Main steam line isolation Reactor isolation was deleted for high water level initiation actuation. To provide improved plant

availability.

5.4 Main steam line drain system A main steam line drain system was improved. Prevent accumulation of condensate in an idle line outboard of MSLIV.

5.4 Feedwater

sparger The thermal sleeve was changed to provide improved design of sparger to nozzle.

To eliminate vibration, failure, and leakage. 5.3 Standby liquid control (SLC) system Interlocks on the SLC system were revised. To prevent inadvertent boron injection during system testing. 9.3.5 and 7.4.1 SSES-FSAR

Rev. 49, 04/96 Page 2 of 3 TABLE 1.3-8 SIGNIFICANT DESIGN CHANGES FROM PSAR TO FSAR*

ITEM CHANGE REASON FOR CHANGE FSAR PORTION IN WHICH CHANGE IS DISCUSSED RCIC & HPCI steam supply A warmup bypass line and valve was added. Permits pressurizing and pre-warming of the steam supply line downstream to the turbine during reactor vessel heatup. 5.4 and 6.3 RCIC & HPCI vacuum breaker system A vacuum breaker system was added to the turbine exhaust line into the suppression pool. To prevent backup of water in the pipe and consequential high dynamic pipe loads and reactions. 5.4 and 6.3 RCIC & HPCI system Each component has been made capable of functional testing. Improved testability 5.4 and 6.3 Automatic depressurization system (ADS) The interlocks on the automatic depressurization system were revised. To meet IEEE-279 requirements. 7.3.1 RPV code The RPV was partially updated to ASME 1971 code and Summer 1971 addenda. Update to applicable code as much as practical.

5.2 Level

instrumentation The RPV level instrumentation was revised to eliminate Yarway columns and replace them with a conventional condensing chamber type; also, separation and

redundancy features were added.

Improve ECCS separation per IEEE 279 and improve reliability.

7.3.1 Leak detection system The leak detection system was revised to upgrade the capability. To meet IEEE-279 requirements. 7.6.1 Reactor vibration monitoring A confirmatory vibration monitoring test was added. NRC requirement 14.2 SSES-FSAR

Rev. 49, 04/96 Page 3 of 3 TABLE 1.3-8 SIGNIFICANT DESIGN CHANGES FROM PSAR TO FSAR*

ITEM CHANGE REASON FOR CHANGE FSAR PORTION IN WHICH CHANGE IS DISCUSSED Primary Containment Concrete Delineation of compressive strengths for pozzolan vs. non-pozzlan Type II Portland cements. Update to reflect current engineering design requirements.

3.8B RPV Insulation Correct the RPV Insulation Description Revised support beams on as-build RPV Insulation Panels 5.3.3.1.4 Safety Related Conduits & Trays Correct separation statements for conduits and trays. Question 7.4 of Amend. #5 of PSAR (Revised per requirement of Reg.

Guide 1.75 - 1974).

3.12 Tornado Loading Revised Tornado Loading combinations.

To reflect latest NRC recommendations in the Standard Review Plan.

3.3

  • NOTE: Design changes listed are only those which have occurred between the last SSES PSAR Amendment and Revision 0 of the FSAR. The NRC has been notified of all other design changes prior to the last PSAR amendment by previous amendments to the PSAR.

SSES-FSAR NIMS Rev. 62 TABLE 1.6-4 OTHER TOPICAL REPORTS FSAR Rev. 64 Page 1 of 3 REPORT NUMBER TITLE REFERENCED IN FSAR SECTION AE-RTL-788 Void Measurements in the Region of Subcooled and Low Quality Boiling (April 1966) 4.4 ANL-5621 Boiling Density in Vertical Rectangular Multichannel Sections with Natural Circulation (November 1956) 4.4 ANL-6385 Power-to-Void Transfer Functions (July 1961) 4.4 BHR/DER 70-1 Radiological Surveillance Studies at a Boiling Water Nuclear Power Reactor (March 1970) 11.1 BMI-1163 Vapor Formation and Behavior in Boiling Heat Transfer (February 1957) 4.4 CF 59-6-47 (ORNL) Removal of Fission Product Gases From Reactor Off Gas Streams by Adsorption (June 11, 1959) 11.3 ST1-372-38 Kinetic Studies of Heterogeneous Water Reactors (April 1966) 4.4 TID-4500 Relap 3 - A Computer Program for Reactor Blowdown Analysis IN-1321 (June 3970) 3.6 UCRL-50451 Improving Availability and Readiness of Field Equipment Through Periodic Inspection, p. 10 (July 16, 1968) 18.3 WAPD-BT-19 A Method of Predicting Steady-Boiling Vapor Fractions in Reactor Coolant Channels (June 1960) 4.4 ANF-524(P)(A) Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors, Revision 2, Supplement 1 Revision 2 and Supplement 2, Advanced Nuclear Fuels Corporation, Richland WA 99352, November 1990 4.1, 4.4, 15.3 ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," August 1990 5.2, 15.0, 15.1, 15.2, 15.3, 15.5 NE-092-001A "SSES Power Uprate Licensing Topical Report," and NRC letter dated November 30, 1993, from Thomas E. Murley to Robert G. Byram (PP&L).

Subject:

Licensing Topical Report for Power Uprate with Increased Core Flow, Rev. 0, Susquehanna Steam Electric Station, Units 1 and 2 (PLA-3788) (TAC NOS. M83426 and M83427) with 10.2, 15.6 SSES-FSAR NIMS Rev. 62 TABLE 1.6-4 OTHER TOPICAL REPORTS FSAR Rev. 64 Page 2 of 3 REPORT NUMBER TITLE REFERENCED IN FSAR SECTION enclosed Safety Evaluation Report PL-NF-89-005-A "Qualification of Transient Analysis Methods for BWR Design and Analysis," Issue Date: July 1992.

15.1 XN-NF-80-19(P)(A) Volume 1, Supplement 3, "Advanced Nuclear Fuels Methodology for Boiling Water Reactors-Benchmark Results for the CASMO-3G/MICROBURN-B Calculation Methodology," EXXON Nuclear Company, Richland, WA 99352, November 1990 15.1 XN-NF-80-19(P)(A) "Exxon Nuclear Methodology for Boiling Water Reactors; Neutronic Methods for Design and Analysis," Volume 1, and Volume 1 Supplements 1 and 2, March 1983.

15.0, 15.4 XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," February 1987 15.0, 15.3 XN-NF-84-105(P)(A) Volume 1 Supplement 4, "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, Void Fraction Model Comparison to Experimental Data," June 1988 15.3 ANF-91-048(P)(A) "Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model,"

and Correspondence, January 1993.

6.3 XN-NF-80-19(P)(A) Volume 3, Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," Siemens Power Corporation, January 1987 4.1, 15.0 XN-NF-79-59(P)(A) "Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies," November 1979 4.1 EMF-CC-074(P)(A) Volume 1, 2 and 4 "STAIF - A Computer Program for BWR Stability Analysis in the Frequency Domain" 4.1, 4.4 ANF-89-98(P)(A) Rev. 1 and Rev. 1 Supplement 1, "Generic Mechanical Design Criteria for BWR Fuel Designs," Advanced Nuclear Fuels Corporation, May 1995 4.2 NFQM "Nuclear Fuel Business Group Quality Management Manual", NFQM, Rev. 1 Framatome - ANP, U.S. Version, July 2003.

4.2 CENPD-400-P-A "Generic Topical Report for the ABB Option III Oscillation Power Range Monitor (OPRM)" 3.1, 4.4, 7.6 SSES-FSAR NIMS Rev. 62 TABLE 1.6-4 OTHER TOPICAL REPORTS FSAR Rev. 64 Page 3 of 3 REPORT NUMBER TITLE REFERENCED IN FSAR SECTION EMF-93-177 (P)(A)

& SUPPLEMENT 1 "Mechanical Design for BWR Fuel Channels" Siemens Power Corporation, August 2005 4.2 EMF-2209 (P)(A) "SPCB Critical Power Correlation," September 2003 4.1, 4.4 EMF-2158(P)(A) "Siemens Power Corporation Methodology For Boiling Water Reactors - Evaluation and Validation of CASMO-4/Microburn-B2" Rev. 0, October 1999 4.1, 4.3, 4.4, 15.4, 15.5 XN-NF-80-19(P)(A) "Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads", Exxon Nuclear Company, June 1986 5.2, 15.0, 15.1, 15.2, 15.4, 15.5 EMF-2361(P)(A) EXEM BWR-2000 ECCS Evaluation Model", Framatome ANP, May 2001 6.3 ANF-1358(P)(A) "The Loss of Feedwater Heating Transient in Boiling Water Reactors", Advanced Nuclear Fuels Corporation, September 2005 15.1 XN-NF-85-74(P)(A) "RODEX2A(BWR) Fuel Thermal-Mechanical Evaluation Model", Exxon-Nuclear Company, Inc,. August 1986 4.1 MEF-93-177(P)(A)

Rev. 1 "Mechanical Design for BWR Fuel Channels,"

August 2005

SSES-FSAR Table Rev. 47 FSAR Rev. 64 1.7-1 START HISTORICAL 1.7 ELECTRICAL, INSTRUMENTATION, AND CONTROL DRAWINGS

Table 1.7-1 contains a list of non-proprietary electrical, instrumentation and control (EI&C) drawings. This table lists those drawings which were considered to be necessary to evaluate the safety-related features in Chapters 7 and 8 of the Susquehanna Unit 1 and 2 FSAR. All the drawings listed in Table 1.7-1 are considered historical.

END HISTORICAL

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 1 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Abnormal occurrence Any reportable occurrence that is determined by the Commission to be significant from the standpoint of public health or safety Abnormal Operational Transients Infrequent design events that may be reasonably expected during the course of planned operations, including events that are a result of (or follow)a single equipment malfunction or operator error. Power failures, pump trips, and rod withdrawal errors are typical of the single malfunctions or errors which may initiate the events in this category.

Acceptable Demonstrated to be adequate by the safety analysis of the Plant. Accident A single event, not reasonably expected to occur during the course of plant operations, that has been hypothesized for analyses purposes or postulated from unlikely but conceivable situations and that causes or threatens to cause a violation of one or more fission product barriers.

Achieving Criticality All actions which are normally accomplished in bringing the Plant from a condition in which all control rods are fully inserted to a condition in which nuclear criticality is achieved and maintained.

Achieving Shutdown Achieving shutdown begins where power operation ends and includes all actions normally accomplished in achieving nuclear shutdown (more than one rod subcritical) following power operation.

Activated Device A mechanical module in a system used to accomplish an action. An activated device is controlled by an actuation device.

Active components a. Those components whose operability is relied upon to perform a safety function such as safe shutdown of the reactor or mitigation of the consequences of a postulated pipe break in the reactor coolant pressure boundary. b. Active component is one in which mechanical motion must occur to complete the component's intended function.

Active failure The failure of an active component such as a piece of mechanical equipment, component of the electrical supply system or instrumentation and control equipment to act on command to perform its design function. Examples include the failure of a motor-operated valve to move to its correct position, the failure of an electrical breaker or relay to respond, the failure of a pump, fan, or diesel generator to start, etc.

Actuation Device An electrical or electromechanical module controlled by an electrical decision output used to produce mechanical operation of one or more activated devices, thus achieving necessary action.

Additional Plant Capability Event An event which neither qualifies as neither an abnormal operational transient nor an accident but which is postulated to demonstrate some special capability of the Plant.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 2 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Administrative Controls Measures to prevent the existence or development of an unsafe condition in connection with the operation of the reactor. They also define the administrative action to be taken in the event a safety limit or allowed condition for operation is exceeded. Requirements concerning the facility's organization and management, procedures, record keeping, review and audit, and reporting are specified.

Alteration of the Reactor Core The act of moving any component in the region above the core support plate, below the upper grid and within the shroud. Normal control rod movement with the control rod drive hydraulic system is not defined as a core alteration. Normal movement of in-core instrumentation and the traversing in-core probe is not defined as a core alteration.

Alternate Rod Injection An alternate means of inserting control rods. One of the features provided in order to mitigate a postulated anticipated transient without scram (ATWS) event.

Analog channel calibration Adjustment of channel output such that it responds, with acceptable range and accuracy, to know values of the parameter which the channel measures. Calibration shall encompass the entire channel, including alarm or trip, and shall be deemed to include the channel functional test.

Analog channel check A qualitative determination of acceptable operability by observation of channel behavior during operation. This determination may include comparison of the channel with other independent channels measuring the same variable.

Analog channel functional test Injection of a simulated signal into the channel to verify that it is operable, including alarm and/or trip initiating action.

Anticipated operational occurrences Anticipated operational occurrences mean those conditions of normal operation which are expected to occur one or more times during the life of the nuclear power unit and include but are not limited to loss of power to all recirculation pumps, tripping of the turbine-generator set, isolation of the main condensers, and loss of all off-Site power (10 CFR 50 Appendix A).

Anticipated transients (with Scram) This group of anticipated abnormal transients include events which present a demand for protection action by the Reactor Protection System and which have a probability of occurrence greater than 10-³ per year. The events which fall in this category of anticipated transients are listed below: a. Loss of load b. Excessive load increase c. Loss of one feedwater pump d. Loss of flow (one pump) e. Rod withdrawal f. Startup accident g. Accidental depressurization of Reactor Coolant System h. Plant blackout

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 3 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Anticipated Transients Without Scram (ATWS) Anticipated operational occurrences which require reactor shutdown followed by the failure to insert all control rods; ie. a failure to SCRAM.

Associated Areas The off-Site equipment, facilities and structures which are necessary for the operation of the Project. These include the makeup water pump facility, the makeup water pipeline, the discharge structure, the transmission line, the railroad spur, and the rights-of-way and access roads associated with the above.

Average linear power density Total thermal power produced in the fuel rods divided by the total active fuel length of all rods in the core.

Average rod power Total thermal power produced in the fuel rods divided by the number of fuel rods (assuming all rods have equal length).

Channel An arrangement of components and sensors as required to generate a single protective action signal when required by a generating station condition. A Channel loses its identity where single action signals are combined.

Class IE Electric Systems The safety classification of the electric equipments and systems that are essential to emergency reactor shutdown containment isolation, reactor core cooling and reactor heat removal or otherwise are essential in preventing significant release of radiation to the environment.

Closed System Piping system containing fluid, not freely accessible to the environment, penetrating containment but not communicating with either primary coolant pressure boundary or containment atmosphere.

Cold Shutdown When the reactor is in the shutdown mode; the reactor coolant is maintained at equal to or less than 200°F, and the reactor vessel is vented to containment atmosphere.

Common mode failure The failure of two or more components of the same or similar design by the same failure mechanism. Such failure mechanisms for compon ents may result from the adverse conditions from a design basis event for which the components were expected by design to remain functional. Such failures may result from a design deficiency or manufacturing deficiency. Redundant equipment can be made inoperable by this mechanism.

Components Items from which a system is assembled. Containment (primary and secondary) The structures that enclose components of the reactor coolant pressure boundary and which provides an essentially leaktight barrier against the uncontrolled release of fission products to the environment.

Containment Atmosphere Free volume enclosed by the primary containment.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 4 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Containment integrity Exists when all the following conditions exist: a. All nonautomatic containment isolation valves not required for normal operation are closed or under administrative control. b. Blind flanges are properly installed where required. c. The equipment door is properly closed and sealed. d. At least one door in each personnel air lock is properly closed and sealed. e. All automatic containment isolation trip valves are operable or closed. f. The containment leakage satisfies Technical Specification.

Containment isolation Establishment of mechanical barrier(s) in appropriate fluid systems penetrating the Containment which would otherwise represent open paths for the fission products in the event of a loss-of-coolant accident inside the Containment.

Controlled Access Area The area immediately surrounding the principal Project Structures, enclosed with a fence or other suitable physical barrier, such that entry into this area is controlled. This area will encompass the Reactor Buildings, the Turbine Buildings, the Auxiliary Buildings, Control Building, Diesel-Generator Buildings, Radwaste Building, and the Cooling Towers.

Controls Methods and devices by which actuation is used to affect the value of a variable.

When used with respect to nuclear reactors, means apparatus and mechanisms, the actuation of which directly affects the reactivity or power level of the reactor.

Cooldown Cooldown begins where achieving shutdown ends and includes all actions normally accomplished in the continued removal of decay heat and the reduction of nuclear system temperature and pressure.

Critical items Those structures, units (or components) and systems which require a degree of design review, verification, inspection and documentation over and above that applied in the course of normal engineering, procurement and construction. As a minimum, critical items include all structures and systems required to maintain the integrity of the reactor primary system pressure boundary, to provide Containment Engineered Safety Features, assure safe shutdown under all conditions and continued residual heat removal.

Degree of redundancy The difference between the number of sensors of a variable and the number of sensors which when tripped will cause an automatic system trip.

Design Basis "Design basis" means that information which identifies the specific functions to be performed by a structure, system, or component of a facility, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design. These values may be (1) restraints derived from generally accepted "state of art" practices for achieving functional goals, or (2) requirements derived from analysis (based on calculation and/or experiments) or the effects of a postulated accident for which a structure, system, or component must meet its functional goals.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 5 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Design Basis Accidents (DBA) The hypothesized accident whose characteristics and consequences are utilized in the design of those systems and components pertinent to the preservation of radioactive material barriers and the restriction of radioactive material release from the barriers upon occurrence of a loss-of-coolant accident. The potential radiation exposures resulting from a DBA are not exceeded by any similar accident postulated from the same general accident assumptions.

Design Basis Events (DBE) Postulated events used in a design to establish the performance requirements of structures, systems, and components.

Design features Those features of the facility such as materials of construction and geometric arrangements which, if altered or modified, would have a significant effect on safety.

Design Power The power level equal to 102% of the licensed or rated core thermal power level. The design power level is equivalent to 4031 MWt.

Diffuser The submerged section of the discharge pipeline which has multiple ports. Dilution Zone The boundary of the dilution zone is defined as that point where the Plant discharge is mixed with the Susquehanna River.

Discharge structure The diffuser section, connecting discharge pipeline, and anchors, both the shoreline anchors and river bed anchors.

Drywell A pressure-containing envelope surrounding the reactor and its recirculation loops which will channel steam resulting from the LOCA through the suppression pool for condensation. Part of primary containment.

Emergencies Unplanned events characterized by risks sufficient to require immediate action to avoid or mitigate an abrupt or rapidly deteriorating situation.

Emergency Conditions (Infrequent Incidents) Those deviations from Normal Conditions which require shutdown for correction of the conditions or repair of damage in the system. The conditions have a low probability of occurrence but are included to provide assurance that no gross loss of structural integrity will result as a concomitant effect of any damage developed in the system.

Engineered Safeguards (Same as Engineered Safety Features).

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 6 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Engineered Safety Features (ESF) a. Features of a unit or system other than reactor trip or those used only for normal operatio n, that are provided to prevent, limit or mitigate the release of radioactive material in excess of 10 CFR 50.67 limits. b. Engineered Safety Feature System (ESFS) consists of those systems, including essential support systems or components thereof the primary purpose of which during a design basis accident (DBA) will be to: (1) Retain fuel temperatures within design limits by maintaining fuel coolant inventory and temperatures within design limits. (2) Maintain fuel temperatures within design limits by inserting auxiliary negative reactivity. (3) Prevent the escape of radioactive materials to the environment in excess of 10 CFR 50.67 limits by isolation of the systems or structures. (4) Reduce the quantity of radioactivity available for leakage and its potential for leakage by purification, cleanup, containment heat removal and containment pressure reduction. (5) Control the concentration of combustible gases in the containment systems within established limits.

Exclusion area A circle within a radius of 1800 ft from the centerline of the reactors, as defined by 10CFR 100.3. Extended Load Line Limit Analysis Safety analyses performed to demonstrate adequate safety margins in support of a license amendment permitting operation with an elevated load line on the power-flow map; i.e. with increased thermal power at a given recirculation flow.

Failure The termination of the ability of an item to perform its required function. Failures may be unannounced and not detected until the next test (unannounced failure), or they may be announced and detected by any number of methods at the instant of occurrence (announced failure).

Faulted Condition (Limiting Faults) Those combinations of conditions associated with extremely-low-probability, postulated even ts whose consequences are such that the integrity and operability of the nuclear energy system may be impaired to the extent that considerations of public health and safety are involved. Such considerations require compliance with safety criteria as may be specified by jurisdictional authorities.

Functional Test The manual operation or initiation of a system, subsystem, or component to verify that it functions within design tolerances (eg, the manual start of a core spray pump to verify that it runs and that it pumps the required volume of water).

General Design Criteria (GDC) A set of design criteria for structures, systems, and components important to safety, which are given in Appendix A to 10 CFR 50, and provide reasonable assurance that the Plant can be operated without undue risk to the health and safety of the public.

Globe Stop Check Valve (GCK) These valves shall be designed to normally function as check valves, but in addition they shall be provided with means for positive shutoff using manual or mechanical actuators.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 7 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Heatup Heatup begins where achieving criticality ends and includes all actions which are normally accomplished in approaching nuclear system rated temperature and pressure by using nuclear power (reactor critical). Heatup extends through warmup and synchronization of the turbine generator.

High radiation area Any area, accessible to personnel, in which there exists radiation originating in whole or in part within licensed material at such levels that a major portion of body could receive in any one hour a dose in excess of 100 mrem.

Hot shutdown See Technical Specification Section 1.1. Startup/Hot standby See Technical Specification Section 1.1. Immediate Immediate means that the required action will be initiated as soon as practicable considering the safe operation of the unit and the importance of the required action.

Inactive components Those components whose operability (eg, valve opening or closing, pump operation or trip) are not relied upon to perform the system function during the transients or events considered in the respective operating condition categories.

Incident Any natural or accidental event of infrequent occurrence and its related consequences which affect the Plant operation and require the use of Engineered Safety Feature systems. Such events, which are analyzed independently and are not assumed to occur simultaneously, include the loss-of-coolant accident, steam line ruptures, steam generator tube ruptures, etc. A system blackout may be an insolated occurrence or may be concurrent with any event requiring Engineered Safety Feature systems use.

Incident Detection Circuitry Includes those trip systems which are used to sense the occurrence of an incident. Increased Core Flow Operation with core flow greater than 100% of original design. Used to provide additional reactivity at end of core life to permit a longer fuel cycle and more economic operation.

Instrument Calibration An instrument calibration means the adjustment of an instrument signal output so that it corresponds, within acceptable range and accuracy, to a known value(s) of the parameter which the instrument monitors. Calibration shall encompass the entire instrument including actuation, alarm, or trip.

Channel check See Technical Specification Section 1.1. Channel Functional Test See Technical Specification Section 1.1. Irradiated Fuel Fuel that has been in the reactor during reactor operation. Isolated Condition Condition in which the reactor is isolated from the main condenser. Limiting conditions for operation The lowest functional capability or performance levels of equipment required for safe operati on of the facility.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 8 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Limiting safety system settings Settings for automatic protective devices are related to those variables having significant safety functions. (Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting has been chosen so that automatic protective action will correct the most severe abnormal situation anticipated before a safety limit is exceeded).

Linear power density The thermal power produced per unit length of fuel rod (kW/ft). Since fuel assembly geometry is standardized, this is the unit of power density most commonly used. For all practical purposes it differs from kW/liter by a constant factor which includes geometry and the fraction of the total thermal power which is generated in the fuel rod.

Load Group An arrangement of buses, transfers, switching equipment, and loads fed from a common power supply. Local heat flux The heat flux at the outer surface of the cladding (Btu/ft²hr). For nominal rod parameters this differs from linear power density by a constant factor.

Logic That array of components which combines individual bistable output signals to produce decision outputs. Logic channel A logic channel is a group of logic matrices which operate in response to the digital single action signals from the analog channels to generate a protective action signal.

Logic System functional test See Technical Specification Section 1.1. Long Term The remainder of the recovery period following the short term. In comparison with the short term where the main concern is to remain within NRC specified site criteria, the long-term period of operation involves bringing the Plant to cold shutdown conditions where access to the Containment can be gained and repair effected.

Loss-of-Coolant Accident (LOCA) Those postulated accidents that result from the loss of reactor coolant, at a rate in excess of the capability of the Reactor Coolant Makeup System, from breaks of pipes containing reactor coolant, up to and including a break equivalent in size to the double-ended rupture of the largest pipe of the Reactor Coolant System.

Low Population Zone (LPZ) The area included in a three mile radius from the midpoint of the centerline between the two reactor buildings on on Plant Site, and in 10 CFR 100.3 as defined.

Low power physics tests Tests below a nominal five percent of rated power which measure fundamental characteristics of the reactor core and related instrumentation.

Manual Component A component, the operability of which is relied upon to perform a manual nuclear safety function such as providing manual action or operator information required for initiation of action for safe shutdown of the reactor of mitigation of the consequences of an accident.

Material surveillance program The provisions for the placement of reactor vessel material specimens in the reactor vessel, and the program of periodic withdrawal and testing of such specimens to monitor, over the service life of the vessel, changes in the fracture toughness properties of the vessel as a result of neutron irradiation.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 9 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Minimum Critical Heat Flux Ratio (MCHFR) The lowest value of the ratio of critical heat flux (that heat flux which results in transition boiling) to the actual heat flux at the same location.

Minimum degree of redundancy The degree of redundancy below which operation is prohibited or otherwise restricted by the Technical Specifications.

Missile Barrier A Physical barrier which protects essential components, systems or structures from potential missiles arising from consequences of a loss-of coolant accident.

Mode See Technical Specification Section 1.1. Module Any assembly of interconnected components which constitutes an identifiable device, instrument, or piece of equipment. A module can be disconnected, removed as a unit, and replaced with a spare. It has definable performance characteristics which permit it to be tested as a unit. A module could be a card or other subassembly of a larger device, provided it meets the requirements of this definition.

Normal conditions Normal conditions are any condition in the course of system startup, operation in the design power range, hot standby and system shutdown, other than Upset, Emergency, Faulted or Testing Conditions.

Normal operation Operation of the plant under planned, anticipated conditions including, but not limited to, the following: a. Reactor critical (any temperature)

b. Power operation c. Reactor startup d. Reactor shutdown e. Refueling f. Periodic testing g. Nuclear system cooldown h. Nuclear system heatup i. Standby (reactor shutdown, nuclear system maintained at constant temperature)

Nuclear-fueled electrical generating facility (the Plant) The reactor, turbine-generator, cooling tower, associated buildings (reactor building, turbine building, and administration building), and the switchyard.

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

Nuclear Safety Operational Analysis A systematic identification of the requirements for the limitations on plant operation necessary to satisfy nuclear safety operational criteria.

Nuclear Safety Operational Criteria A set of standards used to select nuclear safety operational requirements.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 10 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Nuclear System Generally includes those systems most closely associated with the reactor vessel which are designed to contain or be in communication with the water and steam coming from or going to the reactor core. The nuclear system includes the following: a. Reactor vessel b. Reactor assembly and internals c. Reactor core d. Main steam lines from reactor vessel out to and including the isolation valve outside the Containment e. Neutron monitoring system f. Reactor recirculation system g. Control rod drive system h. Residual heat removal system i. Reactor core isolation cooling system j. Emergency core cooling systems k. Reactor water cleanup system l. Feedwater system piping between the reactor vessel and the first valve outside the Containment m. Pressure relief system Nuclear System Process Barrier See Reactor Coolant Pressure Boundary Occupational dose Include exposure of an individual to radiation (i) in a restricted area; or (ii) in the course of employment in which the individual's duties involve exposure to radiation, provided that "occupational dose" shall not be deemed to include any exposures of an individual to radiation for the purpose of medical diagnosis or medical therapy of such individual.

Operable See Technical Specification Section 1.1. Operating A system or component is operating when it is performing its intended functions in the required manner. Operating Cycle Interval between the end of one refueling outage and the end of the next subsequent refueling outage. Operational The adjective "operational", along with its noun and verb forms, is used in reference to the working or functioning of the Plant, in contrast to the design of the Plant.

Operating reports These reports include the Startup Report, First Year Operation Report, and Semiannual Operating Records. Operating reports are submitted in writing to the Director, Division of Reactor Licensing, USNRC, Washington, D.C. 20545.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 11 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Operating Basis Earthquake (OBE) That earthquake which produces vibratory ground motion for which those structures, systems, and components necessary for power generation are designed to remain operable.

Operator Any individual who manipulates a control of a facility. An individual is deemed to manipulate a control if he directs another to manipulate a control Operator error An active deviation from written operating procedures or nuclear plant standard operating practices. A single operator error is the set of actions that is a direct consequence of a single reasonably expected erroneous decision. The set of actions is limited as follows: a. Those actions that could be performed by only one person. b. Those actions that would have constituted a correct procedure had the initial decision been correct. c. Those actions that are subsequent to the initial operator error and that affect the designed operation of the plant but are not necessarily directly related to operator error.

Passive Component A component in which mechanical mo vement does not occur in order for the component to perform its intended function.

Passive failure The structural failure of a static component which limits the component's effectiveness in carrying out its design function. When applied to a fluid system, this could mean a break in the pressure boundary.

Peaking Factor The ratio of the maximum fuel rod surface heat flux in an assembly to the average surface heat flux of the core.

Penetration Assembly, Elec. Provides the means to allow passage of electrical circuits through a single aperture (nozzle or other opening) in the containment pressure barrier, while maintaining the integrity of the pressure barrier. Pennsylvania Power & Light Company (PP&L) The owner-operator of the Project, having total controlling ownership. Period of recovery The time necessary to bring the Plant to a cold shutdown and regain access to faulted equipment. The recovery period is the sum of the short-term and long-term periods.

Place in Cold Shutdown Condition Conduct an uninterrupted normal Plant shutdown operation until the cold shutdown condition is attained.

Place in Isolated Condition Conduct an uninterrupted normal isolation of the reactor from the main (turbine) condenser including the closure of the main steam line isolation valves.

Place in Shutdown Condition Conduct an uninterrupted normal plant shutdown operation until shutdown is attained.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 12 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Planned Operation Normal plant operation under planned conditions in the absence of significant abnormalities. Operations subsequent to an incident (transient, accident, or special event) are not considered planned operations until the procedures being followed or equipment being used are identical to those used during any one of the defined planned operations. The established planned operations can be considered as a chronological sequence: refueling outage; achieving criticality; heatup; power operation; achieving shutdown; cooldown; refueling outage.

The following planned operations are identified: a. Refueling Outage b. Achieving Criticality c. Heatup d. Reactor Power Operation e. Achieving Shutdown f. Cooldown Plant Those structures, systems and components that make up the Susquehanna Steam Electric Station. Power density The thermal power produced per unit volume of the core (kW/liter). Power Generation When used to modify such words as design basis, action and system, this term indicates that the objective, design basis, action, or system is related to the mission of the Plant, to generate electrical power, as opposed to concerns considered to be of primary safety importance. Thus, the words "power generation" identify aspects of the Plant which are not considered to be of primary importance with respect to safety.

Power Generation Design Basis The power generation design basis for a power generation system states in functional terms the unique design requirements which establish the limits within which the power generation objective shall be met. A safety system may have a power generation design basis which states in functional terms the unique design requirements which establish the limits within which the power generation objective for the system shall be met.

Power Generation Evaluation Shows how the system satisfies some or all of the power generation design bases. Because power generation evaluations are not directly pertinent to public safety, generally they are not included. However, where a system or component has both safety and power generation objectives, a power generation evaluation can clarify the safety versus power generation capabilities.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 13 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Power Generation System Any system, the actions of which are not essential to a safety action, but which are essential to a power generation action. Power generation systems are provided for any of the following purposes: a. To carry out the mission of the Plant-generate electrical power through planned operation. b. To avoid conditions which would limit the ability of the Plant to generate electrical power. c. To facilitate and expedite the return to conditions permitting the use of the Plant to generate electrical power following an abnormal operational transient, accident, or special event.

Power operation condition When the reactor is critical and the neutron flux power range instrumentation indicates greater than two percent of rated power.

Power uprate Evaluations, tests, modifications, setpoint changes, and license amendments which permitted an increase in rated thermal power from the original 3293 Mwt to 3441 Mwt; allowing an increase in the nominal generator rating from approximately 1100 to approximately 1150 MWe, and in the net plant rating from approximately 1050 MWe to approximately 1100 MWe. Extended Power Uprate (EPU): The operating license for both units was further modified to permit operation at 3952 MWt with a nominal generator output of 1300 MWe.

Preferred power source That power supply which is preferred to furnish electrical energy under accident or post accident. It is obtained from start-up transformers. The switchgear is arranged to auto transfer from one preferred source to another preferred source in the event the preferred source fails.

Preferred power system The off-site external commercial power system. Preoperational Test Program The preoperational test program applicable to the nuclear steam supply system is the test program conducted prior to fuel loading. The test program applicable to other Plant systems is the test program conducted prior to that system's required operation.

Principal design criteria The criteria which establish the necessary design, fabrication, construction, testing and performance requirements for structures, systems and components important to safety, that is, structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public.

Principal Project structures The Reactor Buildings, Control Buildings, Diesel Generator Building, Radwaste Building, Turbine Buildings, and Cooling Towers.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 14 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Probable maximum flood (PMF) The hypothetical flood characteristics (peak discharge, volume and hydrograph shape) that are considered to be the most severe "reasonably possible": at a particular location, based on relative comprehensive hydrometeorological analyses of critical runoff producing precipitation (and snowmelt, if pertinent) and hydrological factors favorable for maximum flood runoff.

Refer to Chapter 2 of the SSES FSAR for specific values which apply to the Susquehanna Steam Electric Station.

Probable maximum precipitation (PMP) The theoretically greatest precipitation over the applicable drainage area that would prod uce flood flows that have virtually no risk of being exceeded.

Refer to Chapter 2 of the SSES FSAR for specific values which apply to the Susquehanna Steam Electric Station.

Probable maximum winds The hypothetical tornado or other cyclonictype windstorm that might result from the most severe combinations of meteorological parameters that are considered reasonably possible in the region involved, if the tornado or other type windstorm should approach the point under study along a critical path and at optimum rate of movement.

Protection System The aggregate of the protective signal system and the protective actuator system Protective action a. Protective action at the channel level is the generation of a signal by a single channel when the variable(s) sensed exceeds a limit. b. Protective action at the system level is the operation of sufficient actuated equipment to accomplish a protective function (for example: rapid insertion of control rod, closing of containment isolation valves, safety injection, core spray).

Protective Actuator System An arrangement of components that performs a protective action when it receives a signal from the protective signal system (for example: control rods, their drive mechanisms and their trip mechanisms; isolation valves, their operators and their contractors; core spray pumps, their motors and circuit breakers).

Protective function Any one of the functions necessary to limit the safety consequences of a design basis event (for example: rapid reduction of reactor power following a control rod ejection, isolation of the Containment following a steam line break, removal of heat from the core following a loss-of-coolant-accident.

Quality Assurance (QA) All those planned and systematic actions necessary to provide adequate confidence that a structure, system or component will perform satisfactorily in service. Quality assurance includes quality control, which comprises those quality assurance actions related to physical characteristics of a material, structure, component, or system which provide a means to control the quality of the material, structure, component, or system of predetermined requirements.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 15 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Quality Control (QC) Those quality assurance actions related to physical characteristics of a material, structure, component, or system which provide a means to control the quality of the material, structure, component or system to predetermined requirements.

Q-Listed system Q-Listed systems, structures and components are those which prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. They include materials, structures, and equipment whose failure could cause significant release of radioactivity to the environment comparable to 10 CFR 50.67 limits at the Site exclusion distance, or which are vital to the safe shutdown of the Plant, or which are necessary for the removal of decay and sensible heat from the reactor.

Quality Group A classification which identifies the importance of structures, systems, and components with respect to Plant safety functions in accordance with definitions given in NRC Regulatory Guide 1.26. Radiation area Any area, accessible to personnel, in which there exists radiation originating in whole or in part within licensed material, at such levels that a major portion of the body could receive in any one hour a dose in excess of 5 mrem, or in any five consecutive days a dose in excess of 100 mrem.

Radioactive Material Barrier Includes the systems, structures, or equipment that together physically prevent the uncontrolled release of radioactive materials. The barriers are the fuel cladding, the reactor coolant system and the Containment.

Radioactive waste Radioactive wastes are solids, liquids, and gaseous effluents from the radioactive waste systems that have concentration or radioactivity in excess of background.

Rated power The power level at which the reactor is producing 100 percent of reactor vessel rated steam flow. This is the maximum power that could be authorized by the operating license. Rated coolant flow, rated neutron flux and rated nuclear system pressure refer to values of these parameters when the reactor is at rated power.

Reactivity A state variable of neutron chain reactions which is indicative of a deviation in the chain reaction from criticality. It is measured in terms of where p=keff-1/keff. Positive valves correspond to a supercritical state and negative values to a subcritical state.

Usage has established "units" of delta k/k for reactivity change. This term (delta k/k) is used to represent a departure from criticality, and is referred to as reactivity worth. Reactivity worth is the reactivity attributable to the specified component material, portion of material, or void in the nuclear reactor.

Reactor Building Structural complex enclosing the primary containment, and forming secondary containment. Reactor Coolant System The vessels, pipes, pumps, tubes, valves and similar process equipment that contain the steam, water, gases, and radioactive materials coming from, going to, or in communication with the reactor vessel.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 16 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Reactor coolant pressure boundary (RCPB) All those pressure-containing components such as pressure vessels, piping, pumps and v alves, which are (1) part of the Reactor Coolant System, or (2) connected to the Reactor Coolant System, up to and including any and all of the following: a. The outermost containment isolation valve in system piping which penetrates primary reactor Containment. b. The second of two valves normally closed during normal reactor operation in system piping which does not penetrate primary reactor Containment. The Reactor Coolant System safety and relief valves.

Reactor critical When the neutron chain reaction is self-sustaining and keff = 1.0. Reactor Power Operations Reactor power operation begins after heatup is complete and includes any operation with the mode switch in the "Startup" or "Run" position with the reactor critical and above 1 percent rated power. Reactor Vessel Pressure Unless otherwise indicated, reactor vessel pressures are those measured by the reactor vessel steam space detectors.

Redundant equipment or system An equipment or system that duplicates the essential function of another equipment or system to the extent that either may perform the required function regardless of the state of operation or failure of the other.

Refueling Mode See Technical Specification Section 1.1. Refueling operation condition Any operation within the Containment involving movement of core components when the vessel head is completely unbolted or removed and there is fuel in the reactor.

Refueling Outage The period of time between the shutdown of the unit prior to a refueling and the startup of the unit after that refueling. For the purpose of designating frequency of test ing and surveillance, a refueling outage shall mean a regularly scheduled outage. However, where such outages occur within 8 months of the completion of the previous refueling outage, the required surveillance testing need not be performed until the next regularly scheduled outage.

Refueling shutdown condition When the reactor is subcritical by at least 10,000 pcm, Tavg is 140°F, and fuel or fuel inserts are scheduled to be moved to or from the reactor core.

Reliability The probability that a component will perform its specified function without failure for a specified time in a specified environment.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 17 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Reportable Occurrence They are as follows: 1) Failure of the reactor protection system or other systems settings to initiate the required protective function by the time a monitored parameter reaches the setpoint specified as the limiting safety-system setting in the technical Specifications or failure to complete the required protective function. 2) Operation of the unit or affected systems when any parameter or operation subject to a limiting condition for operation is less conservative than the least conservative aspect of the limiting condition for operation established in the technical specifications. 3) Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment.

4) Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady-state conditions during power operation greater than or equal to 1% k/k; a calculated reactivity balance indicating 5 shutdown margin less conservative than specified in the technical specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if subcritical, an unplanned reactivity insertion of more than 0.5% k/k; or occurrence of any unplanned criticality. 5) Failure or malfunction or one or more components which prevents or could prevent, by itself, the fulfillment of the functional requirements of system(s) used to cope with accidents analyzed in the SAR. 6) Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems(s) used to cope with accidents analyzed in the SAR. 7) Conditions arising from natural or manmade events that as a direct result of the event, require plant shutdown, operation of safety systems, or other protective measures required by the technical specifications.
8) Errors discovered in the transient or accident analyses or in the method used for such analysis as described in the safety analysis report or in the bases for the technical specifications but have or could have permitted reactor operation in a manner less conservative than assumed in the analyses 9) Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than that assumed in the accident analyses in the SAR or technical Specifications bases; or discovery during plant life of conditions not specifically considered in the SAR or technical Specifications that require remedial action or corrective measures to prevent the existence or development of an unsafe condition.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 18 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Response spectrum A plot of the maximum response of single degree of freedom bodies, at a damping value expressed as a percent of critical damping, of different natural frequencies, mounted on the surface of interest (that is, on the ground for the ground response spectrum or on the floor of a building for that floor's floor response spectrum) when the surface is subjected to a given earthquake's motion.

NOTE: The response spectrum is not the floor motion or the ground motion.

Restricted area Any area access which is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials. "Restricted Area" shall not include any areas used as residential quarters, although a separate room or rooms in a residential building may be set apart as a restricted area.

Rod power or rod integral power The length integrated linear power in one rod (kW). Run Mode See Technical Specification Section 1.1. Safe Shutdown Earthquake (SSE) The "Safe Shutdown Earthquake" is that maximum probable earthquake which produces the vibratory ground motion for which structures, systems and components designed to Seismic Category I requirements remain functional.

Safety When used to modify such words as objective, design basis, action, and system, the word indicates that, that objective, design basis, action, or system is related to concerns considered to be of safety significance, as opposed to the Plant mission - to generate electrical power. Thus, the word "safety" identifies aspects of the plant which are considered to be of importance with respect to safety. A safety objective or safety design basis does not necessarily indicate that the system is an engineered safety feature.

Safety Action An ultimate action in the Plant which is essential to the avoidance of specified conditions considered to be of safety significance. The specified conditions are those that are most directly related to the ultimate limits on the integrity of the radioactive material barriers and the release of radioactive material. There are safety actions associated with planned operation, abnormal operational transients, accidents, and special events. Safety actions include such actions as reactor scram, emergency core cooling, reactor shutdown from outside the control room and the indication to the operator of the values of certain process variables.

Safety Class A classification which identifies the importance of structures systems, and components with respect to Plant functions in accordance with definitions given in ANSI N212 for BWR's Safety Design Basis The safety design basis for a safety system states in functi onal terms the unique design requirements that establish the limits within which the safety objective shall be met. A power generation system may have a safety design basis which states in functional terms the unique design requirements that ensure that neither planned operation nor operational failure by the system results in conditions for which Plant safety actions would be inadequate.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 19 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Safety Evaluation Shows how the system satisfies the safety design bases. A safety evaluation is performed only for those systems that have safety design bases. Safety evaluations form the basis for the Technical Specifications and establish why specific safety limitations are imposed.

Safety limits Limits upon important process variables which are found to be necessary to reasonably protect the integrity of certain of the physical barriers which guard against the uncontrolled release of radioactivity.

Safety Related See Section 17.2.2 for this definition Safety System Any system, group of systems, components, or groups of components the actions of which are essential to accomplishing a safety action.

Scram Refers to the automatic rapid insertion of control rods into the reactor core in response to the detection of undesirable conditions.

Seiche An oscillation of the surface of a lake or landlocked sea that varies in period from a few minutes to several hours and is thought to be initiated chiefly by local variations in atmospheric pressure aided in some instances by winds and tidal currents and that continues for a time after the inequalities of atmospheric pressure have disappeared.

Seismic Category I Plant features required to assure 1) the integrity of the reactor coolant pressure boundary, 2) the capability to shut down the reactor and maintain it in a safe shutdown condition, or 3) the capability to prevent or mitigate the consequences of accidents which could result in potential off-Site exposures comparable to the guideline exposure of 10 CFR 50.67.

Plant features required to meet NRC GDC-1 of Appendix A to 10 CFR 50 and Appendix B of 10 CFR 50. Plant features required to meet NRC GDC-2 of Appendix A to 10 CFR 50 and Proposed Appendix A to 10 CFR 100. Plant features designed to withstand effects of the Safe Shutdown Earthquake.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 20 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Seismic Category II (Non-Seismic Category I) Plant features not required to assure 1) the integrity of the reactor coolant pressure boundary, 2) the capability to shut down the reactor and maintain it in a safe shutdown condition, or 3) the capability to prevent or mitigate the consequences of accidents which could result in exposures comparable to the guidelines of exposures of 10 CFR 50.67.

Plant features not required to meet NRC GDC-1 of Appendix A to 10 CFR 50 and Appendix B to 10 CFR 50.

Plant features not required to meet NRC GDC-2 of Appendix A to 10 CFR 50 and proposed Appendix A to 10 CFR 100. Plant features not designed to withstand the effects of the Safe Shutdown Earthquake.

Senior Operator Any individual designated by a facility licensee under 10 CFR 50 to direct the licensed activities of licensed operators.

Sensor That part of a channel used to detect variations in a measured variable. Service conditions Environmental, power, and signal conditions expected as a result of normal operating requirements, expected extremes in operating requirements, and postulated conditions appropriate for the design basis events of the station.

Short term The time immediately following the incident during which automatic actions are performed, system responses are checked, type of incident is identified and preparations for long-term recovery operations are made. The short term is the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following initiation of system operations after the incident.

Shut Down See Technical Specification Section 1.1. Shutdown Mode See Technical Specification Section 1.1. Simulated Automatic Actuation Simulated automatic actuation means applying a simulated signal to sensor to actuate the circuit in question.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 21 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Single failure a. An occurrence which results in the loss of capability of a component to perform its intended safety functions. Multiple failures resulting from a single occurrence are considered to be a single failure. Fluid and electrical systems are considered to be designed against and assumed single failure if neither (1) a single failure of any active component (assuming passive components function properly) nor (2) a single failure of any passive component (assuming active components function properly) results in a loss of the capability of the system to perform its safety functions. b. Single failures are spontaneous occurrences imposed upon safety systems that are required to respond to a design basis event. They are postulated in spite of the fact that they were designed to remain functional under the adverse condition imposed by the accident. No mechanism for the cause of the single failure need be postulated. Single failures of passive components in electrical systems should be assumed in designing against a single failure.

Site Features Those features that are important to safety by virtue of the physical setting of the Plant.

Spring Loaded Piston Actuated Check Valve (SLPACK) Spring loaded piston actuated check valves operate as follows for the following modes: a. During Normal Flow

A spring loaded piston operator is held open by air pressure. Meanwhile, the valve is fully open by action of force due to flow alone. b. During Accidental Loss of Operator Air
The valve shall remain in the fully open position when the flow rate is equal to or greater than the normal flow rate indicated in the Valve Data Sheets. With a flow rate less than normal, the valve may be partially open due to the force of spring against force due to flow. c. Upon Reversal of Flow
Valve shall tightly shut as a normal check valve. In addition, the Control room operator will assist in starting valve closure by sending a remote signal to open a fail-open solenoid valve, releasing air pressure from the operator cylinder. All signal wiring will be furnished by others.

Standby power source The power supply that is selected to furnish electrical energy when the preferred power supply is not available. It consists if an electrical generating unit and all necessary auxiliaries, usually a diesel generator set.

Standby power system Those on Site power sources and their distribution equipment provided to energize devices essential to safety and capable of operation independently of the preferred power system.

Startup Mode See Technical Specification Section 1.1. Startup testing After fuel has been loaded into the reactor, testing is conducted under conditions similar to those for Hot Functional Testing with the reactor subcritical to complete those tests which could not be completed during the initial hot functional testing and those which must be done with the core in position.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 22 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Suppression Pool A pool of water, located in the lower section of the Containment. During relief-valve discharge and postulated LOCA's, it serves as a heat sink and a pressure-suppression water pool comparable to the pool in the torus or suppression chamber of earlier BWR plants.

Surveillance Frequency See Technical Specification Section 1.4. Surveillance requirements Requirements relating to test, calibration or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within the safety limits and that the limiting conditions of operation will be met.

Swing bus A bus that is automatically transferred to one or the other of two redundant standby power sources. System Redundancy System that duplicates an essential function of another system to the extent that either may perform a required function regardless of the state of operation or failure of the other.

Technical Specifications (as used in the FSAR) Encompass the nuclear safety operational requirements and limits to be used by plant operations and management personnel. They are prepared in accordance with the requirements of 10 CFR 50.36 and are incorporated by reference into the operating license issued by the U.S. Nuclear Regulatory Commission.

Testing Conditions Testing conditions are those tests in addition to the ten (10) hydrostatic or pneumatic tests permitted by ASME Section III, paragraphs NB-6222 and NB-6322 including leak tests or subsequent hydrostatic tests.

Testable Check Valve These valves are designed to normally function as a check valve, but in addition, they shall be provided with a manual test lever to prove operability during shutdown.

Test Duration The elapsed time between test initiation and test termination. Test Interval The elapsed time between the initiation of identical tests. Thermal power The total core heat transfer rate from the fuel and the coolant. Tornado criteria The design parameters applicable to the design tornado, such as rotational and translational velocities, design pressure differential and associated time interval and the tornado-generated missile impact load with a statement of whether the imposed loads will be established simultaneously in establishing the tornado design.

Transition Boiling Transition boiling means the boiling regime between nucleate and film boiling. Transition boiling is the regime in which both nucleate and film boiling occur intermittently with neither type being completely stable.

Trip The change of state of a bistable device that represents the change from a normal condition. A trip signal, which results from a trip, is generat ed in the channels of a trip system and produces subsequent trips and trip signals throughout the system as directed by the logic.

Trip System That portion of a system encompassing one or more channels, logic and bistable devices used to produce signals to the actuation logic. A trip system terminates and loses its identity where outputs are combined in logic.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 23 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Type tests Tests made on one or more units to verify adequacy of design. Ultimate Heat Sink The spray pond and associated structures and components. Unrestricted area Any area access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials and any area used for residential quarters.

Upset Conditions (Incidents of Moderate Frequency) Any deviations from Normal Conditions anticipated to occur often enough that design should include a capability to withstand the conditions without operational impairment. The Upset Conditions include those transients which result from any single operator error or control malfunction, transients caused by a fault in a system component requiring its isolation from the system and transients due to loss of load or power. Upset Conditions include any abnormal incidents not resulting in a forced outage and also forced outages for which the corrective action does not include any repair of mechanical damage. The estimated duration of an Upset Condition shall be included in the Design Specifications.

THIS FIGURE HAS BEEN REPLACED BY DWG.

M-100, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.8-2A replaced by dwg. M-100, Sh. 1 FIGURE 1.8-2A, Rev. 55 AutoCAD Figure 1_8_2A.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-100, Sh. 2 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.8-2B replaced by dwg. M-100, Sh. 2 FIGURE 1.8-2B, Rev. 56 AutoCAD Figure 1_8_2B.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-100, Sh. 3 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.8-2C replaced by dwg. M-100, Sh. 3 FIGURE 1.8-2C, Rev. 48 AutoCAD Figure 1_8_2C.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-100, Sh. 4 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.8-2D replaced by dwg. M-100, Sh. 4 FIGURE 1.8-2D, Rev. 55 AutoCAD Figure 1_8_2D.doc SSES-FSAR Text Rev. 60 FSAR Rev. 64 1.1-1

1.1 INTRODUCTION

1.1.1 TYPE OF LICENSE This FSAR is submitted by PPL Susquehanna, LLC in support of its application for an operating license for Susquehanna Steam Electric Station (Susquehanna SES) Units 1 and 2.

1.1.2 IDENTIFICATION

OF APPLICANT Application is made by PPL Susquehanna, LLC, Two North Ninth Street, Allentown, Pennsylvania, 18101.

1.1.3 NUMBER

OF PLANT UNITS

The plant consists of two units which have a common control room, diesel generators and refueling floor, turbine operating deck, radwaste system, and other auxiliary systems.

1.

1.4 DESCRIPTION

OF LOCATION

The 2,355 acre plant site is located in Salem Township, Luzerne County, Pennsylvania, approximately 20 miles southwest of Wilkes-Barre, 50 miles northwest of Allentown and 70 miles northeast of Harrisburg.

1.1.5 TYPE OF NUCLEAR STEAM SUPPLY The Nuclear Steam Supply System for each unit consists of a General Electric Boiling Water Reactor, BWR/4 product line with a 3952 MWt nominal rating.

1.1.6 TYPE OF CONTAINMENT The containment is a pressure suppression type designated as Mark II. The drywell is a steel-lined concrete cone located above the steel-lined concrete cylindrical pressure suppression chamber.

The drywell and suppression chamber are separated by a concrete diaphragm slab which also serves to strengthen the entire system.

1.1.7 CORE THERMAL POWER LEVELS

The rated core thermal power for each unit is 3952 MWt. The nominal turbine generator output at 3952 MWt is 1300 MWe for both Unit 1 and Unit 2.

SSES-FSAR Text Rev. 60 FSAR Rev. 64 1.1-2 1.1.8 SCHEDULED FUEL LOAD AND OPERATION DATA Unit 1 original fuel load was on July 27, 1982 with a commercial operation date of June 8, 1983. Unit 2 original fuel load was March 28, 1984 with a commercial operation date of February 12, 1985.

1.1.9 FSAR ORGANIZATION The Susquehanna SES Final Safety Analysis Report (FSAR) has been organized using Regulatory Guide 1.70 Revision 2 (September, 1975).

The FSAR is divided into 18 chapters, using the same chapter, section, subsection, and paragraph headings that appear in the standard format.

Where information has been presented that has not been specifically requested by the standard format, the information is presented in the appropriate chapter as a section or subsection, and follows the information specifically requested by the standard format.

Tabulations of data are designated "tables" and are identified by the section number, followed by a dash and number of table according to its order in the text; e.g., Table 3.4-5 is the fifth table of Section 3.4. Drawings, pictures, sketches, curves, graphs, and engineering diagrams are identified as "figures" and are numbered in the same manner as tables.

This FSAR has been organized so that all figures and tables are at the end of each major section (with the exception of Chapter 15 Sections 15.1 through 15.5). The results of reload specific calculations for Chapter 15 Sections 15.1 through 15.5 (i.e., tables and figures) are included in Appendices 15C and 15D for Units 1 and 2, respectively. Appendix 15E contains similar information and analytical results associated with these sections for non-limiting events for the initial Susquehanna cycles for both Unit 1 and Unit 2. Reference reports or other documents, where applicable, have been tabulated in a separate subsection at the end of each Section. Reference numbers are listed sequentially under the section heading to which they apply. As an additional guide to the reader, Section 1.8 has been incorporated to provide a glossary of definitions, abbreviations, symbols, indices, legends, and other similar aids.

Contents of the overall FSAR are tabulated at the front of each volume. The level of breakdown corresponds to the numbered tabs. Each chapter is provided with the detailed table of contents for that chapter.

Amendments are issued as replacement sheets. The location of the amended material is indicated by a vertical bar in the margin of the page adjacent to the information. The amendment number and date appear at the left bottom corner of each page. Any change from the preceding issue, whether it be only a correction to a typographical error or deletion of a word or repositioning of words, is considered an amendment. Where an amendment bar appears adjacent to blank portions of a page, a deletion is indicated. Where pages have been changed only to reposition material with no change in content, only the amendment number and date are given.

A "List of Effective Pages," contained in its own binder, shows the current revision number of each individual page.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-1 1.2 GENERAL PLANT DESCRIPTION

1.2.1 PRINCIPAL

DESIGN CRITERIA The principal criteria for design, construction, and testing of the Susquehanna SES are summarized below. Specific criteria, codes and standards are addressed in Section 3.0.

1.2.1.1 General Design Criteria The Susquehanna SES design conforms to the requirements given in 10CFR50, Appendix A.

Specific compliance is discussed in Section 3.1.

1. The plant is designed, fabricated, and erected to produce electrical power in accordance with the codes, standards, and regulations as described in Section 3.1.
2. Safety related systems are designed to permit safe plant operation and to accommodate postulated accidents without endangering the health and safety of the public.

1.2.1.2 System Design Criteria

1.2.1.2.1 Nuclear System Criteria

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

throughout the design life of the fuel.

2. The fuel cladding, in conjunction with other plant systems, is designed to retain integrity throughout any abnormal operational transient.
3. Those portions of the nuclear system which form part of the reactor coolant pressure boundary are designed to retain integrity as a radioactive material barrier during normal operation following abnormal operational transients and accidents.
4. Heat removal systems are provided in sufficient capacity and operational adequacy 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.
5. Heat removal systems are provided to remove decay heat generated in the core under circumstances wherein the normal operational heat removal systems become inoperative. The capacity of such systems is adequate to prevent fuel clad damage.

The reactor is capable of being shut down automatically in sufficient time to permit decay heat removal systems to become effective following loss of operation of normal heat removal systems.

6. 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.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-2

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

1.2.1.2.2 Safety Related Systems Criteria 1.2.1.2.2.1 General

1. Safety systems act in response to abnormal operational transients so that fuel cladding retains its integrity as a radioactive material barrier.
2. Safety systems and engineered safety features act to ensure that no damage to the nuclear system process barrier results from internal pressures caused by abnormal operational transients or accidents.
3. Where positive, precise actions are required in immediate response to accidents, such actions are automatic and require no decision or manipulation of controls by operations personnel.
4. Essential safety actions are carried out by equipment of sufficient redundance and independence that no single failure of active components can prevent the required actions. For systems or components to which IEEE 279-1971 is 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.
5. Features of the station that are essential to the mitigation of accident consequences are designed, fabricated, and erected to quality standards which reflect the importance of the safety function to be performed.
6. The design of safety systems and engineered safety features includes allowances for environmental phenomena at the site.
7. Provision is made for control of active components of safety systems and engineered safety features from the control room.
8. Safety systems and engineered safety features are designed to permit demonstration of their functional performance requirements.

1.2.1.2.2.2 Containment and Isolation Criteria

1. 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.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-3 2. A secondary containment that completely encloses both primary containment and fuel storage areas is provided and is designed to act as a radioactive material barrier.

3. The primary and secondary containments, in conjunction with other engineered safety features, act to prevent radioactive material released from the containment volumes from exceeding the guideline values of applicable regulations.
4. 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.
5. Piping that 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 there is a threat of uncontrolled radioactive material being released. Such isolation is effected in time to prevent radiological effects from exceeding the values of applicable regulations.

1.2.1.2.2.3 Emergency Core Cooling System (ECCS) Criteria

1. ECCS systems are provided to limit fuel cladding temperature to temperatures below the onset of fragmentation (2200F) in the event of a loss of coolant accident.
2. The ECCS provides for continuity of core cooling over the complete range of postulated break sizes in the nuclear system process barrier.
3. The ECCS is diverse, reliable, and redundant.
4. Operation of the ECCS is initiated automatically when required regardless of the availability of off-site power supplies and the normal generating system of the plant.

1.2.1.2.3 Process Control Systems Criteria 1.2.1.2.3.1 Nuclear System Process Control Criteria

1. Control equipment is provided to allow the reactor to respond to limited load changes, major load changes and abnormal operational transients.
2. It is possible to control the reactor power level manually.
3. Control of the nuclear system is possible from a single location.
4. 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.
5. Interlocks, or other automatic equipment, are provided as a backup to procedural controls to avoid conditions requiring the actuation of nuclear safety systems or engineered safety features.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-4 6. If the control room is inaccessible, it is possible to bring the reactor from power range operation to a hot shutdown condition by manipulation of controls and equipment which are available outside of the control room. Furthermore, station design does not preclude the ability, in this event, to bring the reactor to a cold shutdown condition from the hot shutdown condition.

1.2.1.2.3.2 Power Conversion Systems Process Control Criteria

1. Controls are provided to maintain temperature and pressure to below design limitations. This system will result in a stable operation and response for all allowable variations.
2. Controls are designed to provide indication of system trouble.
3. Control of the power conversion system is possible from a single location.
4. Controls are provided to ensure adequate cooling of power conversion system equipment.
5. Controls are provided to ensure adequate condensate purity.
6. Controls are provided to regulate the supply of water so that adequate reactor vessel water level is maintained.

1.2.1.2.3.3 Electrical Power System Process Control Criteria

1. Controls are provided to ensure that sufficient electrical power is provided for startup, normal operation, prompt shutdown and continued maintenance of the station in a safe condition.
2. Control of the electrical power system is possible from a single location.

1.2.1.2.4 Electrical Power System Criteria

1. The station electrical power systems are designed to deliver the electrical power generated.
2. Sufficient normal 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 engineered safety features under postulated design basis accident conditions.
3. Standby electrical power sources are provided to allow prompt reactor shutdown and removal of decay heat under circumstances where preferred power is not available.

They provide sufficient power to all engineered safety features requiring electrical power.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-5 1.2.1.2.5 Fuel Handling and Storage Facilities

1. Fuel handling and storage facilities are located in the reactor building and are designed to preclude criticality and to maintain adequate shielding and cooling for spent fuel.

Additional spent fuel storage facilities are provided at the Independent Spent Fuel Storage Installation (ISFSI) located north of the Low Level Radwaste Holding Facility (LLRWHF). The ISFSI is described in detail in Section 11.7. Handling of spent fuel stored at the ISFSI is in the Reactor Building and is designed to preclude criticality and to maintain adequate shielding and cooling for spent fuel.

1.2.1.2.6 Auxiliary Systems Criteria

1. 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.

2. 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.

1.2.1.2.7 Power Conversion Systems Criteria Components of the power conversion systems are designed to fulfill the following basic objectives:

a) Generate electricity with the turbine generator from steam produced in the reactor, condense the exhaust steam in the condenser and return the condensed water to the reactor as heated feedwater with most of the non-condensable gases and impurities removed.

b) Ensure that any fission products or radioactivity associated with the steam and condensate during normal operation are safely contained inside the system or are released under controlled conditions in accordance with waste disposal procedures.

1.2.1.2.8 Radioactive Waste Disposal Criteria

1. Gaseous, liquid, and solid waste disposal facilities are designed so that the discharge of radioactive effluents, storage, and off-site shipment of radioactive material are made in accordance with applicable regulations.
2. These facilities include means for informing station operating personnel whenever operational limits on the release of radioactive material are exceeded.
3. A separate facility for interim on-site storage of low level radioactive waste material as of April 30, 1988 was included under the 10CFR Part 50 facility operating licenses.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-6 1.2.1.2.9 Shielding and Access Control Criteria

1. 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. The design and establishment of the above include conditions that deal with fission product release from failed fuel elements and contamination of station areas from system leakage.
2. The control room is shielded against radiation and has suitable environmental control so that occupancy under design basis accident conditions is possible.

1.2.2 PLANT

DESCRIPTION 1.2.2.1 Site Characteristics 1.2.2.1.1 Location and Size

In terms of its relationship to metropolitan areas, the plant site lies approximately 20 miles southwest of Wilkes-Barre, approximately 50 miles northwest of Allentown, and approximately 70 miles northeast of Harrisburg. The plant is a two unit, Boiling Water Reactor. Each unit has a nominal rating of 1300 MWe. It is located on a 1,574 acre property owned by PPL in Salem Township, Luzerne County, Pennsylvania, along the west bank of the Susquehanna River approximately 5 miles northeast of the borough of Berwick, Pennsylvania. In addition, 717 acres of the site are located on the east side of the river in Conyngham and Hollenback Township. Total site acreage is approximately 2,355 acres. There are no structures or facilities on the east side of the river with the exception of transmission lines and facilities. A map of the site area including major structures and facilities is provided as Figure 2.1-12. PPL owns the entire 1800 foot plant exclusion area (except for Township Route T-419) and has complete authority to regulate any and all access and activity within that area.

The property in Salem Township is open deciduous woodland, interspersed with grassland and orchards and is bounded on its eastern flank by the Susquehanna River, which has a low water elevation of 484 feet MSL in this vicinity. Much of the northern property boundary runs along the slopes of an east-west trending ridge rising to a maximum elevation of 1060 feet MSL. This ridge abuts a rolling plateau to the south which in turn falls gradually in an easterly direction toward the floodplain of the Susquehanna River. The plant site is located on this plateau at an approximate grade of 675 feet MSL. Rainfall runoff leads into two main valleys that form intermittent waterways draining to the Susquehanna River, east of the property.

Also, in Salem Township a portion of the long abandoned North Branch Canal runs north-south across the floodplain between the Susquehanna River and U.S. Route 11. Within the property limits, the northern portion of the canal is generally dry and overgrown with trees and shrubs whereas the southern portion contains stagnant water. A permanent 30 acre body of water named Lake "Took-a-While" is located just west of the canal. An approximate 400 acre recreation area has been developed on the floodplain.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-7 Acreage in Conyngham and Hollenback Townships of Luzerne County on the east side of the Susquehanna River is open to the public for hunting, fishing and hiking. One of the trails leads to the scenic view from Council Cup, a 700-foot high bluff overlooking the Susquehanna River valley.

A multiple-use land management program coordinated through a 10-year forest stewartship

plan is aimed at providing a mix of woodlands, farming, recreation, a wildlife habitat, timber production, and historical protection.

1.2.2.1.2 Road and Rail Access

US Route 11 runs north-south through the property along the western edge of the floodplain. In this vicinity, the highway has a pavement width of 36 feet and is of bitumen-topped concrete slab construction. Township road T419, which follows the toe of the east-west trending ridge described above, leads off US 11 to traverse the property and link with Township road T438 which passes through the western portion of the property. Both of these Township roads are paved in this vicinity. A railroad line on the floodplain parallels US 11 in traversing the property.

The North Shore Railroad Co. operates this line owned by The Commonwealth of Pennsylvania. The railroad is a single track, non-electrified line of standard gage. The nearest railroad station is at Berwick, 5 miles to the south. Access to the various facilities is provided as follows:

a) a MAIN and a SECONDARY ACCESS ROAD leading from US 11 at separate locations to serve the main power block and surrounding structures b) an ACCESS RAILROAD SPUR leading from the Conrail (Erie-Lackawanna) Railroad to serve the main power block and cooling tower areas

c) PLANT ROADS providing access to all structures as well as connecting with the Main and Secondary access roads d) a RIVER FACILITIES ACCESS ROAD leading from US 11 to the intake structure on the river bank

e) a PERIMETER PATROL ROAD paralleling and within the plant security fence

1.2.2.1.3 Description of Plant Environs

1.2.2.1.3.1 Geology and Soils The property is underlain by a series of tightly folded strata of Paleozoic age, trending generally northeast-southwest. Pleistocene glacial outwash deposits mantle much of the area, particularly in topographic depressions. Underlying these glacial deposits are strata of Devonian and older ages. The plant site area is underlain by a series of siltstone, sandstone and shale beds of the Hamilton and Susquehanna groups of Devonian age.

Soils in the area are derived from parent material of glacial origin. These soils are acidic, well drained and generally not well suited for agricultural purposes.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-8 1.2.2.1.3.2 Groundwater The two principal aquifer systems in the region are the unconsolidated glacial and alluvial deposits and the underlying bedrock formations. The glacial deposits consist of drift, till, and outwash materials and vary in permeability from very low to high. The thickness of these deposits is highly variable in the site vicinity, ranging from 1 or 2 feet to over 100 feet. Wells penetrating the bedrock produce water from Onondaga limestone and strata of the Hamilton siltstone group.

The groundwater table in the area is a subdued replica of the surface topography. At the site, the water table is found generally within 35 feet of the ground surface, usually just below the bedrock surface but sometimes within the overburden soils.

Groundwater is the primary source of water supply in the region. The plant potable water supply is obtained from groundwater. There will be no impact to surrounding wells (groundwater level) due to Plant usage, as documented in Dames & Moore report titled "Environmental Feasibility for Groundwater Supply at SSES" dated 9/24/86.

1.2.2.1.3.3 Hydrology The Susquehanna River Basin comprises an area of about 27,500 square miles in the states of New York, Pennsylvania and Maryland. The plant site is located on the west side of the Main Branch of the Susquehanna River, approximately 42 miles upstream from the confluence of the Main and West Branches at Sunbury, Pennsylvania. The Main Branch has its source at Otsego Lake about 35 miles southeast of Utica, New York. From Otsego Lake, the river flows generally southwest. The Lackawanna River joins it near Pittston, Pennsylvania. From there it flows past the site.

The extreme and average daily flows recorded at the gaging station at Wilkes-Barre, about 20 miles upstream from the site, are:

Flow Date (cfs) (gpm)

Minimum 528 2.38x10 5 September 27,1964 Average 13,000 5.85x10 6 70 years of record Maximum 345,000 1.55x10 8 June 24, 1972 (Hurricane Agnes)

The River's path is controlled by the geologic structure in the site area, following the regional folding and jointing pattern. Just north of the site area, the river cuts across the regional fold axis along a major joint set before swinging west-southwest and paralleling the regional fold

axis.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-9 The river gradient is approximately 1.5 feet per mile near the site. The river has long pools, short riffles and shallow bedrock flats. Scouring during spring floods has removed most silt deposits except in quiet pools. The river formerly had large deposits of coal silt, most of which have been removed by dredging. Below Shickshinny, a large pool extends downstream to a point near Mocanaqua; below this point the water is shallower and faster. The bed is rock and gravel, and the river is interspersed with islands. Both upstream and downstream portions of these islands have clean gravel bars.

1.2.2.1.3.4 Meteorology A modified continental-type climate prevails in the general area of the site. Normally, the frost-free season extends from late April to mid October. Minimum temperatures during December, January, and February usually are below freezing, but rarely dip below 0F. Maximum temperatures above 100F have seldom been recorded. The yearly relative humidity averages about 70 percent. Mean annual snowfall in the region is about 52 inches. In winter the area has about 40 percent of sunshine; the summer percentage is about 60 percent. Heavy thunderstorms have occasionally caused damage over limited areas and tornado-force winds have been reported.

The annual precipitation is about 38 inches. July is normally the wettest month, with an average rainfall of about 5 inches, and February is the driest, with about 2 inches.

The dominant wind is from the West-Southwest Sector.

1.2.2.1.3.5 Seismicity Of the very few earthquakes which have occurred in Pennsylvania during historical times, most have been in the southeastern part of the State.

Only minor damage has ever been recorded from earth movement in Pennsylvania, with the exception of the two disturbances at Wilkes-Barre in 1954. It is doubtful whether the latter were the direct result of an earthquake. Since the affected area was only 2000 square feet and no record was made of the disturbances at any of the nearest seismic stations, it is likely they were associated with mining activities and the readjustment of alluvial deposits.

Because of the small correlation between seismic activity and known faults or tectonics, the area can be said to constitute an inactive seismotectonic province.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-10 1.2.2.2 General Arrangement of Structures and Equipment The principal structures and facilities located at the Plant Site are shown on Figure 2.1-11.

The equipment arrangements for these structures are shown in Dwgs. M-220, Sh. 1, M-221, Sh. 1, M-222, Sh. 1, M-223, Sh. 1, M-224, Sh. 1, M-225, Sh. 1, M-226, Sh. 1, M-227, Sh. 1, M-230, Sh. 1, M-231, Sh. 1, M-232, Sh. 1, M-233, Sh. 1, M-234, Sh. 1, M-235, Sh. 1, M-236, Sh. 1, M-237, Sh. 1, M-240, Sh. 1, M-241, Sh. 1, M-242, Sh. 1, M-243, Sh. 1, M-244, Sh. 1, M-245, Sh. 1, M-246, Sh. 1, M-247, Sh. 1, M-248, Sh. 1, M-249, Sh. 1, M-250, Sh. 1, M-251, Sh. 1, M-252, Sh. 1, M-253, Sh. 1, M-254, Sh. 1, M-255, Sh. 1, M-256, Sh. 1, M-257, Sh. 1, M-258, Sh. 1, M-259, Sh. 1, M-260, Sh. 1, M-261, Sh. 1, M-270, Sh. 1, M-271, Sh. 1, M-272, Sh. 1, M-273, Sh. 1, M-274, Sh. 1, M-276, Sh. 1, M-280, Sh. 1, M-281, Sh. 1, M-282, Sh. 1, M-284, Sh. 1, M-5200, Sh. 1, and M-5200, Sh. 2.

1.2.2.3 Nuclear System The nuclear system includes a single cycle, forced circulation, General Electric Boiling Water Reactor producing steam for direct use in the steam turbine. Heat balances showing the major parameters of the nuclear system for the rated power condition, at rated core flow and at 108 Mlb/hr increased core flow, are shown in Figures 1.2-49 and 1.2-49-1 for Unit 1 and Figures 1.2-49-2 and 1.2-49-3 for Unit 2. The reactor heat balances differ slightly from the turbine heat balances (Figures 10.1-1a and 10.1-1b). The reactor heat balances are based on the measured moisture fraction exiting the reactor, and the moisture fraction exiting the MISV's is determined by the expected pressure drop between the reactor vessel steam dome and the MSIV exit. The turbine heat balance is based on turbine inlet design conditions, which allow for a slightly greater moisture fraction in the steam.

1.2.2.3.1 Reactor Core and Control Rods

The fuel for the reactor core consists of depleted, natural, and/or slightly enriched uranium dioxide pellets contained in sealed Zircaloy-2 tubes. These fuel rods are assembled into individual fuel assemblies. The number of fuel assemblies in the complete core is 764. Gross control of the core is achieved by movable, bottom entry control rods. The control rods are of cruciform shape and are distributed evenly throughout the core. The control rods are positioned by individual control rod drives.

Each fuel assembly has several fuel rods with gadolinia (Gd 2 O 3) mixed in solid solution with the UO 2. The gadolinia is burnable poison which diminishes the reactivity of the fresh fuel. It is depleted as the fuel reaches the end of its first cycle.

A conservative limit of plastic strain is the design criterion used for fuel rod cladding failure. The peak linear heat generation for steady-state operation is well below the fuel damage limit even late in life. Experience has shown that the control rods are not susceptible to distortion and have an average life expectancy many times the residence time of a fuel loading.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-11 1.2.2.3.2 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, core spray, and standby liquid control; the incore instrumentation; and other components. The main connections to the vessel include the steam lines, the coolant recirculation lines, the feedwater lines, the control rod drive 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 1050 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 core is cooled by reactor water that enters the lower portion of the core and boils as it flows upward around the fuel rods. The chemistry of reactor water is controlled to minimize corrosion of the fuel cladding, reactor vessel internals and reactor coolant system pressure boundary and to control the transport and deposition of corrosion product activity. 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 line(s). Each steam line is provided with two isolation valves in series, one on each side of the primary containment barrier.

1.2.2.3.3 Reactor Recirculation System

The Reactor Recirculation System pumps reactor coolant through the core to remove the energy generated in the fuel. This is accomplished by two recirculation loops external to the reactor vessel but inside the primary containment. 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.

1.2.2.3.4 Residual Heat Removal System

The Residual Heat Removal System (RHRS) consists of pumps, heat exchangers and piping that fulfill the following functions:

a. Removal of decay heat during and after plant shutdown.
b. Rapid injection of water into the reactor vessel following a loss of coolant accident, at a rate sufficient to reflood the core maintain fuel cladding below the limits contained in 10 CFR 50.46. This is discussed in Subsection 1.2.2.4.
c. Removal of heat from the primary containment following a loss-of-coolant accident (LOCA) to limit the increase in primary containment pressure. This is accomplished by cooling and recirculating the water inside the primary containment. The redundancy of

the equipment provided for the containment is further extended by a separate part of the RHRS which sprays cooling water into the drywell. This latter capability is discussed in Subsection 1.2.2.4.12.

d. Provide for cooling of the spent fuel pool(s) following a seismic event which results in a loss of normal spent fuel pool cooling, in conjunction with normal shutdown of both units.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-12

1.2.2.3.5 Reactor Water Cleanup System (RWCU)

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

1.2.2.4 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 allowable dose limits. These systems may be components, groups of components, systems, or groups of systems. Engineered Safety Feature (ESF) systems are included in this category. ESF systems have a sole function of mitigating the consequences of design basis accidents.

1.2.2.4.1 Reactor Protection System

The Reactor Protection System 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 Reactor Protection System overrides all operator actions and process controls.

1.2.2.4.2 Neutron-Monitoring System

Not all of the Neutron Monitoring System qualifies as a nuclear safety system; only those portions that provide high neutron flux signals and neutron flux oscillation signals to the Reactor Protection System are safety related. The intermediate range monitors (IRM), oscillation power range monitors (OPRM), and average power range monitors (APRM), which monitor neutron flux via in-core detectors, signal the Reactor Protection System in time to prevent excessive fuel clad temperatures as a result of abnormal operational transients.

1.2.2.4.3 Control Rod Drive System When a scram is initiated by the Reactor Protection System, the Control Rod Drive 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.

1.2.2.4.4 Nuclear System Pressure Relief System A Pressure Relief System, consisting of safety-relief valves mounted on the main steam lines, prevents excessive pressure inside the nuclear system following either abnormal operational transients or accidents.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-13 1.2.2.4.5 Reactor Core Isolation Cooling System The Reactor Core Isolation Cooling System (RCIC) provides makeup water to the reactor vessel whenever the vessel is isolated from the main condenser and feed water system. The RCICS uses a steam driven turbine-pump unit and operates automatically, in time and with sufficient

coolant flow, to maintain adequate reactor vessel water level.

1.2.2.4.6 Primary Containment A pressure-suppression primary containment houses the reactor vessel, the reactor coolant recirculating 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 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.

Hydrogen recombiners are provided in the drywell and wetwell to control combustible gases after a LOCA (not credited in the accident analysis).

1.2.2.4.7 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 which 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.

1.2.2.4.8 Secondary Containment

Any leakage from the primary containment system is to the secondary containment system.

This system includes the Standby Gas Treatment System and the Reactor Building Recirculation System. The secondary containment system is designed to minimize the release at ground level of airborne radioactive materials, and to provide for the controlled, filtered release of the reactor building atmosphere at roof level under accident conditions.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-14 1.2.2.4.9 Main Steam Line Isolation Valves Although process lines which 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 air 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.
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.

In addition the main steamline isolation valve leakage Isolated Condenser Treatment Method (Section 6.7) is provided to process the fission products after a LOCA. By directing the leakage from the closed main steamline isolation valves through the main steam drain line to the condenser, this leakage is processed prior to release to the atmosphere.

1.2.2.4.10 Main Steam Line Flow Restrictors

A venturi-type flow restrictor is installed in each steam line close to the reactor vessel. These devices limit the loss of coolant from the reactor vessel and prevent uncovering of the core before the main steam line isolation valves are closed in case of a main steam line break.

1.2.2.4.11 Emergency Core Cooling Systems (ECCS)

Four Core Standby Cooling Systems are provided to prevent excessive fuel clad temperatures if a breach in the nuclear system process barrier results in a loss of reactor coolant. The four Core Standby Cooling Systems are:

1. High Pressure Coolant Injection System (HPCIS)

The HPCIS provides and maintains an adequate coolant inventory inside the reactor vessel to prevent excessive fuel clad temperatures as a result of postulated small breaks in the Reactor Coolant Pressure Boundary (RCPB). A high pressure system is needed for such breaks because the reactor vessel depressurizes slowly, preventing low pressure systems from injecting coolant. The HPCIS 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.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-15 2. Automatic Depressurization System (ADS)

The ADS acts to rapidly reduce reactor vessel pressure in a LOCA situation in which the HPCIS fails to automatically maintain reactor vessel water level. The depressurization provided enables the low pressure standby cooling systems to deliver cooling water to the reactor vessel. The ADS uses some of the safety-relief valves which are part of the

Nuclear System Pressure Relief System. The automatic safety-relief valves are arranged to open when a break in the nuclear system process barrier has occurred and the HPCIS 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 Low Pressure Coolant Injection System Pumps are operating.

3. Core Spray System The Core Spray 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 breach exists in the nuclear system process barrier, 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 onto the core. Either core spray loop, together with another ECCS system, is capable of preventing excessive fuel clad temperatures following a LOCA.
4. Low Pressure Coolant Injection (LPCI)

Low Pressure Coolant Injection (LPCI) is an operating mode of the Residual Heat Removal System (RHRS) and is an engineered safety feature. LPCI uses the pump loops of the RHRS to inject cooling water at low pressure into a reactor recirculation loop. LPCI is actuated by conditions indicating a breach in the nuclear system process barrier, but water is delivered to the core only after reactor vessel pressure is reduced.

LPCI operation, together with the core shroud and jet pump arrangement, provides the capability of core reflooding following a LOCA in time to prevent excessive fuel clad temperatures.

1.2.2.4.12 Residual Heat Removal System (Containment Cooling)

The Residual Heat Removal System (RHRS) for containment cooling is placed in operation to limit the temperature of the water in the suppression pool following a design basis LOCA. In the containment cooling mode of operation, the RHRS pumps take suction from the suppression pool and deliver the water through the RHRS heat exchangers, where cooling takes place by transferring heat to the RHR service water. The fluid is then discharged back to the suppression pool.

As an alternative, RHRSW can be aligned to an RHR heat exchanger when RHR is aligned in the LPCI operating mode to support long term containment cooling.

Another portion of the RHRS is provided to spray water into the primary containment as a means of reducing containment pressure following a LOCA. This capability is in excess of the required energy removal capability and can be placed into service at the discretion of the

operator.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-16 1.2.2.4.13 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 control rod drive. The rate of reactivity insertion resulting from a rod drop accident is limited by this feature. The limiters contain no moving parts.

1.2.2.4.14 Control Rod Drive Housing Supports Control rod drive 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.

1.2.2.4.15 Reactor Building Recirculation and Standby Gas Treatment Systems The Reactor Building Recirculation System and the Standby Gas Treatment System (SGTS) are both a part of the secondary containment. The recirculation system has the capability of recirculating the reactor building air volume prior to its discharge via the SGTS, following a LOCA. Under normal wind conditions, the SGTS has the capability of maintaining a negative pressure within the reactor building with respect to the outside atmosphere. The air moving through the SGTS is filtered and discharged through the turbine building exhaust vent. 1.2.2.4.16 Standby ac Power Supply The Standby ac Power Supply System consists of four diesel-generator sets. The diesel-generators are sized so that three diesels can supply all the necessary power requirements for one unit in the design basis accident condition, plus the necessary required loads to effect the safe shutdown of the second unit. The diesel generators are specified to start up and attain rated voltage and frequency within 10 seconds. Four independent 4 kV engineered safety feature switchgear assemblies are provided for each reactor unit. Each diesel-generator feeds an independent 4 kV bus for each reactor unit.

Additionally, a spare diesel generator is provided which can be manually realigned as a replacement for any one of the other four diesel generators. This spare diesel generator has the emergency loading capability of any of the other four diesel generators.

Each diesel-generator starts automatically upon loss of off-site power or detection of a nuclear accident. The necessary engineered safety feature system loads are applied in a preset time sequence. Each generator operates independently and without paralleling during a loss of off-site power or LOCA signal.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-17 1.2.2.4.17 dc Power Supply Each reactor unit is provided with four independent 125 V and two independent 250 V dc systems. Each dc system is supplied from a separate battery bank and battery charger. The 125 V dc systems are provided to supply station dc control power and dc power to four diesel generators and their associated switchgears. The 250 V dc systems are provided to supply power required for the larger loads such as dc motor driven pumps and valves.

Additionally, a separate 125V dc system is provided for the spare diesel generator. This separate 125V dc system is provided to supply dc control power and dc power to the spare diesel generator auxiliaries and its associated switchgear.

The 125 and 250-V dc Systems are designed to supply power adequate to satisfy the engineered safety feature load requirements of the unit with the postulated loss of off-site power and any concurrent single failure in the dc system.

1.2.2.4.18 Residual Heat Removal Service Water System A Residual Heat Removal Service Water System is provided to remove the heat rejected by the Residual Heat Removal System during shutdown operation and accident conditions.

1.2.2.4.19 Emergency Service Water System The Emergency Service Water System supplies water to cool the standby diesel-generators and the ECCS and Engineered Safety Features equipment rooms, and other essential heat loads.

1.2.2.4.20 Main Steam Line Radiation Monitoring System

The Main Steam Line Radiation Monitoring System 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. A trip signal to the Mechanical Vacuum Pump (MVP) and its suction valves is generated by the monitors upon detection of a high high radiation signal.

1.2.2.4.21 Reactor Building Ventilation Radiation Monitoring System

The Reactor Building Ventilation Radiation Monitoring System consists of a number of radiation monitors arranged to monitor the activity level of the ventilation exhaust from the reactor building. Upon detection of high radiation, the reactor building is automatically isolated and the Standby Gas Treatment System is started.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-18 1.2.2.4.22 Nuclear Leak Detection System The Nuclear Leak Detection 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:

1) Main steam lines
2) Reactor water cleanup (RWCU) system
3) Residual heat removal (RHR) system
4) Reactor core isolation cooling (RCIC) system
5) High pressure coolant injection (HPCI) system
6) Instrument lines

Small leaks generally are detected by temperature and pressure changes, fillup rate of drain sumps, and fission product concentration inside the primary containment. Large leaks are also detected by changes in reactor water level and changes in flow rates in process lines.

1.2.2.5 Instrumentation and Control 1.2.2.5.1 Nuclear System Process Control and Instrumentation 1.2.2.5.1.1 Reactor Manual Control System

The Reactor Manual Control System provides the means by which control rods are manipulated from the control room for gross power control. The system controls valves in the Control Rod Drive Hydraulic System. Only one control rod can be manipulated at a time. The Reactor Manual Control System includes the controls that restrict control rod movement (rod block) under certain conditions as a backup to procedural controls.

1.2.2.5.1.2 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 in core power level. The system is arranged to automatically adjust reactor power output to the

load demand by adjusting the frequency of the electrical power supply for the reactor recirculation pumps.

1.2.2.5.1.3 Neutron Monitoring System

The Neutron Monitoring System is a system of in-core neutron detectors and out-of-core electronic monitoring equipment. The system provides indication of neutron flux, which can be correlated to thermal power level for the entire range of flux conditions that may exist in the core. The system also provides detection of neutron flux oscillations, which may indicate thermal-hydraulic instability. The source range monitors (SRM) and the intermediate range monitors (IRM) provide flux level indications during reactor startup and low power operation.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-19 The local power range monitors (LPRM), oscillation power range monitors (OPRM) and average power range monitors (APRM) allow assessment of local and overall flux conditions during power range operation. The average power range monitors also provide post-accident neutron flux information. A rod block monitor (RBM) is provided to prevent rod withdrawal when reactor power should not be increased at the existing reactor coolant flow rate. The Traversing In-core Probe System (TIPS) provides a means to calibrate the individual LPRM's.

1.2.2.5.1.4 Refueling Interlocks A system of interlocks, restricting the movements of refueling equipment and control rods when the reactor is in the refuel mode, 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 bridge, the refueling bridge hoists, the fuel grapple and control

rods.

1.2.2.5.1.5 Reactor Vessel Instrumentation

In addition to instrumentation provided for the nuclear safety systems and engineered safety features, instrumentation is provided to monitor and transmit information that can be used to assess conditions existing inside the reactor vessel and the physical condition of the vessel itself. The instrumentation provided monitors reactor vessel pressure, water level, temperature, internal differential pressures and coolant flow rates, and top head flange leakage.

1.2.2.5.1.6 Process Computer System An on-line process computer is provided to monitor and log process variables and to make certain analytical computations. In conjunction with approved operating procedure, the rod worth minimizer function prevents improper rod withdrawal under low power conditions. The effect of the rod worth minimizer function is to limit the reactivity worth of the control rods by enforcing adherence to the preplanned rod pattern.

1.2.2.5.1.7 Remote Shutdown System

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 main control room becomes uninhabitable.

1.2.2.5.2 Power Conversion Systems Process Control and Instrumentation

1.2.2.5.2.1 Pressure Regulator and Turbine Control

The pressure regulator maintains control of turbine control valves; it regulates pressure at the turbine inlet and, therefore, the pressure of the entire nuclear system. Pressure regulation is coordinated with the turbine speed system and the load control system so that rapid control valve closure can be initiated when necessary to provide turbine overspeed protection for large load rejection.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-20 1.2.2.5.2.2 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.

1.2.2.5.2.3 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.

1.2.2.6 Electrical Systems

1.2.2.6.1 Transmission and Generation Systems Redundant sources of off-site 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.

The two independent off-site sources provide auxiliary power for startup and for operating the engineered safety feature systems.

The main generator for each unit is an 1800-rpm, three-phase, 60-cycle synchronous machine rated at 1354 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 24 kV to 230 kV on Unit No. 1 and from 24 kV to 500 kV on Unit No. 2 by the unit main transformers and supplied by overhead lines to the 230 kV and 500 kV switchyards, respectively.

1.2.2.6.2 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 and some non-class 1E loads while the non-Class 1E system supplies the balance of plant equipment.

The Class 1E ac system for each unit consists of four independent load groups. Two independent off-site power systems provide the normal electric power to these groups. Each load group includes 4.16 kV switchgear, 480 V load centers, motor control centers and 120 V control and instrument power panel. The vital ac instrumentation and control power supply systems include battery systems, static inverters. Voltages listed are nominal values, and all electrical equipment essential to safety is designed to accept a range of +/-10 percent in voltage.

Four independent diesel generators are shared between the two units. Additionally, a spare diesel generator is provided which can be manually realigned as a replacement for any one of the other four diesel generators. This spare diesel generator has the emergency loading capability of any of the other four diesel generators. Each diesel generator is provided as a standby source of emergency power for one of the four Class 1E ac load groups in each unit.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-21 Assuming the total loss of off-site power 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 design basis accident on one unit and a forced shutdown of the second unit. The non-Class 1E ac system includes 13.8 kV switchgear, 4.16 kV switchgear, 480 V load centers and motor control centers. Four independent Class 1E 125V dc batteries and two independent Class 1E 250V 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 the Class 1E 125 and 250 V batteries.

An additional circuit breaker is provided for each non-class 1E load connected to the class 1E system for redundant fault protection. Additionally, a separate 125V dc system is provided for the spare diesel generator. This separate 125V dc system is provided to supply dc control power and dc power to spare diesel generator auxiliaries and its associated switchgear. These systems are discussed in Chapter 8. 1.2.2.7 Fuel Handling and Storage Systems 1.2.2.7.2 Fuel Pool Cooling and Cleanup System A Fuel Pool Cooling and Cleanup 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. 1.2.2.7.3 Fuel Handling Equipment The major fuel servicing and handling equipment includes the reactor building cranes, the refueling service platform, fuel and control rod servicing tools, fuel sipping and inspection devices, and other auxiliary servicing tools.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-22 1.2.2.8 Cooling Water and Auxiliary Systems 1.2.2.8.1 Service Water System The Service Water System is desig ned to a)Furnish cooling water to various heat exchangers located in the several plant buildings b)Furnish water for diluting the oxidizing and non-oxidizing biocides and for injecting th em into the circulating water systems. This is an intermittent service.The system consists of three 50 percent 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 that acts as the heat sink. Equipment that requires service water and is common to Units 1 and 2 is provided with inter-ties to both service water systems so that either can provide the water. 1.2.2.8.2 Residual Heat Removal Service Water System (RHRSWS)

The objective of the RHRSWS is to provide a reliable supply of cooling water for heat removal from the Residual Heat Removal System under post-accident conditions and supply a source of water if post-accident flooding of the core or primary containment is required. The system consists of two independent loops per unit, each of 100 percent capacity, and each loop consisting of two pumps, valves, piping and controls. Each loop uses the common spray pond with its spray distribution network as a heat sink. During operation the pumps take water from the spray pond and circulate it through the tube side of the RHR heat exchangers. The warm water is returned to the spray pond through a network of spray nozzles that produce the cooling effect by causing an enthalpy gradient as a result of the convective heat transfer and partial evaporative cooling of the spray droplets. A

radiation monitor is provided to check the radioactivity of the service water leaving each RHR heat exchanger. In the event of a high activity level (a tube leak in the RHR heat exchanger),

an alarm will sound and the operator will make the decision to isolate the heat exchanger and minimize the volume of contaminated water that flows to the spray pond.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-23 1.2.2.8.3 Emergency Service Water System (ESWS)

The objective of the ESWS is to supply cooling water to the RHR pumps and associated room coolers during normal and emergency conditions, as necessary, to safely shutdown the reactor or support normal and emergency conditions, as necessary, to safely shutdown the reactor or support "hot standby" conditions and in addition supply cooling water to the Diesel-Generator Units. The ESWS provides a reliable supply of cooling water to emergency equipment under a loss of off-site power condition or LOCA. The system consists of two independent loops supplying both units supplying both units (denoted "A" and "B") each of 100 percent capacity and containing two pumps, valves, piping and controls. Each loop uses the spray pond with its spray distribution system (common to both the Emergency Service Water and RHR Service Water Systems) as a heat sink. The ESWS is designed with sufficient redundancy so that no single active or passive system component failure can prevent it from achieving its safety objective. During operation, the ESWS pumps take water from the spray pond and circulate it through the various heat exchangers in the system. The warm water is returned to the spray pond through either a network of spray nozzles or directly through piping that bypass the spray arrays. The spray nozzles produce the cooling effect by causing an enthalpy gradient as a result of the convective heat transfer and partial evaporative cooling of the spray droplets.

1.2.2.8.4 Reactor Building Closed Cooling Water System

The Reactor Building Closed Cooling Water System is designed to accomplish the following objectives:

a) Provide cooling water to auxiliary plant equipment associated with the nuclear system and located in the reactor and radwaste buildings.

b) Provide cooling water to reactor building chilled water system in the event of unavailability of the chillers or loss of off-site power.

The Reactor Building Closed Cooling Water System consists of two 100 percent capacity pumps, two 100 percent capacity heat exchangers, a head tank, chemical addition tank, associated piping, valves and controls. The reactor building cooling water system is a closed loop cooling water system using inhibited demineralized water. The systems for Units 1 and 2

are separate from each other.

During normal plant operation one pump and heat exchanger will be in service, transferring heat to the service water system, with the other pump on automatic standby. Upon complete loss of off-site power, without occurrence of DBA, both cooling water pumps will start automatically when the buses are re-energized by Diesel Generators. The reactor building closed cooling water heat exchangers can be transferred from service water to emergency service water and one pump can be taken out of service, both by remote manual switching.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-24 1.2.2.8.5 Turbine Building Closed Cooling Water System (TBCCWS)

The TBCCWS is designed to provide cooling water to the auxiliary plant equipment associated with the nuclear and power conversion systems in the Turbine Building. The TBCCWS consists of two 100 percent capacity pumps, two 100 percent capacity heat exchangers, a head tank, chemical addition tank, associated piping and valves. The Turbine Building Cooling water System is a closed loop cooling water system using inhibited demineralized water. The systems for Units 1 and 2 are separate from each other. During normal plant operation, the turbine building closed cooling water heat exchanger transfers heat from the Turbine Building Closed Cooling Water System to the Service Water system. After a loss of off-site power, the pumps start automatically the pumps start automatically and the turbine building closed cooling water heat exchangers can be transferred by remote switching to Emergency Service Water System.

The heat load during this period will be rejected to the emergency service water. One turbine building closed cooling water pump and heat exchanger will be normally in service and the other pump will be on automatic standby.

1.2.2.8.6 Standby Liquid Control System Although not intended to provide rapid reactor shutdown, the Standby Liquid Control System provides a redundant, independent, and alternative method to the control rods to bring the reactor subcritical and to maintain it subcritical as the reactor cools. The system makes possible an orderly and safe shutdown if 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 system will also be used to buffer suppression pool pH to prevent iodine re-evolution following a postulated design basis loss of coolant accident.

1.2.2.8.7 Fire Protection System A Fire Protection System supplies fire fighting water to points throughout the plant. In addition, automatic Halon and carbon dioxide protection systems and portable fire extinguishers are also

provided.

1.2.2.8.8 Plant Heating, Ventilating, and Air-Conditioning Systems

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

1.2.2.8.9 Compressed Air System

The Compressed Air Systems (e.g., instrument air, service air and containment instrument air) supply air of suitable quality and pressure for various plant operations.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-25 1.2.2.8.10 Makeup Water Treatment System A Makeup Water Treatment System furnishes a supply of treated water suitable for use as makeup for the plant.

1.2.2.8.11 Domestic and Sanitary Water Systems

A water system for drinking and sanitary uses is provided for the plant.

1.2.2.8.12 Plant Equipment and Floor Drainage Systems

The Plant Equipment and Floor Drainage System handles both radioactive and non-radioactive drains. Drains which may contain radioactive materials are pumped to the radwaste system for cleanup, reuse, or discharge. Non-radioactive drains are discharged to the environs.

1.2.2.8.13 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.

1.2.2.8.14 Plant Communication System The Plant Communication System provides communication between various plant buildings and locations.

1.2.2.8.15 Process Valve Stem Leakoff System

The Process valve stem leak-off collection system is designed to reduce and control leakage to the atmosphere from valves greater than 2 1/2 in. that are used in the turbine building in systems containing radioactive steam or water and not connected to the main condenser.

Valves in the turbine building were originally provided with valve stem packing leakoff connections. Research and testing has shown that improved packing provides an effective seal to prevent leakage into the Turbine Building. As a result, these leakoff connections are in the process of being removed and package configurations changed, as appropriate, to conform with the new requirements. As part of this effort, leakoff isolation valves and piping will be removed (or abandoned in place) and the leakoff collection header piping will be removed or abandoned in place.

1.2.2.8.16 Diesel Auxiliary Systems

Diesel auxiliary systems are those systems which directly support operation of the emergency diesel generators. The following are diesel auxiliary systems:

a) Diesel Generator Fuel Oil Storage and Transfer System

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-26 b) Diesel Generator Cooling Water System

c) Diesel Generator Starting System d) Diesel Generator Lubrication System

e) Diesel Generator Combustion Air Intake and Exhaust System

1.2.2.8.17 Auxiliary Steam System The auxiliary steam system consists of two electrode steam boilers and auxiliary equipment.

The system is designed to provide flexibility for accommodating varying steam demands during all operating modes.

1.2.2.9 Power Conversion System

1.2.2.9.1 Turbine-Generator

The turbine-generator consists of the turbine, generator, exciter, controls, and required subsystems designed for a nominal plant rating output of 1300 MWe for both Unit 1 and Unit 2.

Each turbine is an 1800 rpm, tandem-compound, six-flow, non-reheat unit with an electrohydraulic control 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, three-phase, 60 Hz, 24,000 V, 1800 rpm, conductor-cooled, synchronous generator rated on the basis of guaranteed best turbine efficiency MW rating at 0.935 power factor, 75 psig hydrogen pressure. The generator-exciter system is shaft-driven, complete with static type voltage regulator and associated switchgear. The following are the turbine generator auxiliary systems:

a) Generator Hydrogen System b) Generator Seal Oil System c) Turbine Lube Oil System d) Steam Seal System e) Gland Exhaust System f) Generator Stator Cooling System

1.2.2.9.2 Main Steam System

The main steam system delivers steam from the nuclear boiler system via four 24 in. OD steam lines to the turbine-generator. This system also supplies steam to the steam jet air ejectors, the reactor feed pump turbines, the main condenser hotwell at startup and low loads, and the steam seal evaporator.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-27 1.2.2.9.3 Main Condenser 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, and each section is located below one of three low-pressure turbines. The condensers are supported on the turbine foundation mat, with rubber expansion joints provided between each turbine exhaust opening and the steam inlet connections in the condenser shells.

During normal operation, steam from the low pressure turbine is exhausted directly downward into the condenser shells through exhaust openings in the bottom of the turbine casings and is condensed. The condenser also serves as a heat sink for several other flows, such as exhaust steam from feed condenser drain, gland seal condenser drain, feedwater heater shell operating vents, and condensate pump suction vents.

During abnormal conditions the condenser is designed to receive (not simultaneously) turbine bypass steam, feedwater heater high level dump(s), and relief valve discharge (from crossover steam lines, feedwater heater shells, steam seal regulator, and various steam supply lines).

Other flows occur periodically; they originate from condensate pump and reactor feed pump startup vents, reactor feed pump and condensate pump minimum recirculation flows, feedwater line startup flushing, turbine equipment clean drains, low point drains, deaerating steam, makeup, condensate, etc.

1.2.2.9.4 Main Condenser Gas Removal System

The main condenser Gas Removal System removes the non-condensable gases from the main condenser and exhausts them to the Off-Gas System. One steam jet air ejector (100 percent capacity), is provided for the removal of air and radiolysis gases during normal operation. One motor-driven mechanical vacuum pump is to establish or maintain vacuum during startup and shutdown.

1.2.2.9.5 Steam Seal System

The steam seal system provides clean, non-radioactive steam to the seals of the turbine valve packings and the turbine shaft packings. The sealing steam is supplied by the seal steam evaporator. The auxiliary boiler provides an auxiliary steam supply for startup and when the seal steam evaporator is not operating.

1.2.2.9.6 Steam Bypass and Pressure Control System The turbine steam bypass and pressure control system control 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.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-28 c) During transient power operation when the reactor steam generation exceeds the turbine steam requirements.

d) When cooling down the reactor.

1.2.2.9.7 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 a natural draft cooling tower.

1.2.2.9.8 Condensate Cleanup System

The function of the Condensate Cleanup System is to maintain the required purity of the

feedwater flowing to the reactor.

The system consists of full flow deep bed demineralizers using ion exchange resins which remove dissolved and a portion of the suspended solids from the feedwater to maintain the purity necessary for the reactor. The demineralizers will also remove some of the radioactive material produced by corrosion as well as fission product carryover from the reactor. The radioactivity from these sources does not have a significant effect on the resins.

1.2.2.9.9 Condensate and Feedwater System The Condensate and Feedwater System is designed to deliver the required feedwater flow to the reactor vessels during stable and transient operating conditions throughout the entire operating range from startup to full load to shutdown. The system operates using four condensate pumps to pump deaerated condensate from the hotwell of the main condenser through the steam jet air ejector condenser, the gland steam condenser, the condensate filters, and thence to the condensate demineralizer. The demineralized feedwater then flows through three parallel strings of feedwater heaters, each string consisting of five heaters, to the suction of three reactor feed pumps which deliver the feedwater to the reactor.

1.2.2.9.10 Condensate and Refueling Water Storage and Transfer System

The function of the Condensate and Refueling Water Storage and Transfer System is to store condensate to be used as follows:

a) Supply water for the RCIC and HPCI systems.

b) Maintain the required condensate level in the hotwell either by receiving excess condensate rejected from the main condensate system or by supplying condensate to the main condensate system to makeup for a deficiency.

c) Fill up the reactor well of either reactor during refueling and receive this water back for storage after it has been cleaned up by the demineralizer.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-29 d) Provide condensate where required for miscellaneous equipment in the radwaste building and both reactor buildings.

The makeup to condensate storage tanks and the refueling storage tank is provided by the demineralized water storage tank.

1.2.2.10 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. Various methods are used to achieve this end, e.g. collection, filtration, holdup for decay, dilution and concentration.

1.2.2.10.1 Liquid Radwaste System The Liquid Radwaste System collects, treats, stores, and disposes of all 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 building 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 which 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.

1.2.2.10.2 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 off-site storage in approved shipping containers. Examples of these wastes are spent control rods, and in-core ion chambers.

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

If off-site shipment of solidified liners or dry active waste is not practicable, these items may be temporarily stored at the Low Level Radioactive Waste Holding Facility, as described in Section 11.6, provided they are packaged for off-site disposal.

1.2.2.10.3 Gaseous Radwaste System Radioactive gaseous wastes are discharged to the reactor building vent via the Gaseous Radwaste System. This system provides hydrogen-oxygen recombination, filtration, and holdup of the off-gases to ensure a low rate of release from the reactor building vent.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-30 The off-gases from the main condenser are the greatest source of gaseous radioactive waste. The treatment of these gases reduces the released activity to below permissible levels.

1.2.2.11 Radiation Monitoring and Control

1.2.2.11.1 Process Radiation Monitoring

Radiation monitors are provided on various lines to monitor for radioactive materials released to the environs via process liquids and gases or for detection of process system malfunctions. These monitors annunciate alarms and/or provide signals to initiate isolation and corrective actions.

1.2.2.11.2 Area Radiation Monitors Radiation monitors are provided to monitor for abnormal radiation at various locations in the reactor building, turbine building, and radwaste building. These monitors annunciate alarms when abnormal radiation levels are detected.

1.2.2.11.3 Site Environs Radiation Monitors

Radiation monitors are provided outside the plant buildings to monitor radiation levels. These data are used for determining the contribution of plant operations to on-site and off-site radiation levels.

1.2.2.11.4 Liquid Radwaste System Control Liquid wastes to be discharged are handled on a batch basis with protection against accidental discharge provided by procedural controls. Instrumentation, with alarms, to detect abnormal concentration of the radwastes, is provided.

1.2.2.11.5 Solid Radwaste Control

The Solid Radwaste System collects, treats, and prepares solid radioactive wastes for off-site shipment. Wastes are handled on a batch basis. Radiation levels of the various batches are

determined by the operator.

1.2.2.11.6 Gaseous Radwaste System Control

The Gaseous Radwaste System is continuously monitored by the turbine building vent radiation monitor and the off-gas pre-treatment radiation monitor. A high level signal will annunciate

alarms.

SSES-FSAR Text Rev. 74 FSAR Rev. 68 1.2-31 1.2.2.12 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 and 10CFR50. It is also designed to protect certain plant components from radiation exposure resulting in unacceptable alterations of material properties or activation.

1.2.2.13 Steam Dryer Storage

A separate Steam Dryer Storage Facility(SDSF) is provided within the plant protected area for the storage, shielding, and radioactive decay of replaced Reactor steam dryers. The steam dryers are cut in half and packaged into steel containers for storage in the SDSF. They are not considered as radioactive waste but are treated as irradiated plant equipment. The SDSF is a reinforced concrete vault with removable roof slab access only, meeting 10CFR20 dose limits.

1.2.2.14 FLEX Equipment Storage Building

A separate FLEX Equipment Storage Building is provided within the plant protected area for the storage of portable equipment needed to respond to a Beyond Design Basis External Event (BDBEE). B.5.b equipment (i.e., pumper truck, etc.) is also stored in this building. This is strictly an emergency equipment storage facility (no personnel occupancy amenities) constructed to meet all plant extreme environmental conditions (i.e., seismic, tornado, missile).

THIS FIGURE HAS BEEN REPLACED BY DWG.

M-220, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-1 replaced by dwg. M-220, Sh. 1 FIGURE 1.2-1, Rev. 57 AutoCAD Figure 1_2_1.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-221, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-2 replaced by dwg. M-221, Sh. 1 FIGURE 1.2-2, Rev. 56 AutoCAD Figure 1_2_2.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-222, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-3 replaced by dwg. M-222, Sh. 1 FIGURE 1.2-3, Rev. 48 AutoCAD Figure 1_2_3.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-223, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-4 replaced by dwg. M-223, Sh. 1 FIGURE 1.2-4, Rev. 48 AutoCAD Figure 1_2_4.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-224, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-5 replaced by dwg. M-224, Sh. 1 FIGURE 1.2-5, Rev. 48 AutoCAD Figure 1_2_5.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-225, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-6 replaced by dwg. M-225, Sh. 1 FIGURE 1.2-6, Rev. 48 AutoCAD Figure 1_2_6.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-226, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-7 replaced by dwg. M-226, Sh. 1 FIGURE 1.2-7, Rev. 48 AutoCAD Figure 1_2_7.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-227, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-8 replaced by dwg. M-227, Sh. 1 FIGURE 1.2-8, Rev. 48 AutoCAD Figure 1_2_8.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-230, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-9 replaced by dwg. M-230, Sh. 1 FIGURE 1.2-9, Rev. 57 AutoCAD Figure 1_2_9.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-231, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-10 replaced by dwg. M-231, Sh. 1 FIGURE 1.2-10, Rev. 49 AutoCAD Figure 1_2_10.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-232, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-11 replaced by dwg. M-232, Sh. 1 FIGURE 1.2-11, Rev. 48 AutoCAD Figure 1_2_11.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-233, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-12 replaced by dwg. M-233, Sh. 1 FIGURE 1.2-12, Rev. 49 AutoCAD Figure 1_2_12.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-234, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-13 replaced by dwg. M-234, Sh. 1 FIGURE 1.2-13, Rev. 48 AutoCAD Figure 1_2_13.doc THiS FIGURE HAS BEEN REPLACED BY DWG.

M-235, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-14 replaced by dwg. M-235, Sh. 1 FIGURE 1.2-14, Rev. 48 AutoCAD Figure 1_2_14.doc THiS FIGURE HAS BEEN REPLACED BY DWG.

M-236, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-15 replaced by dwg. M-236, Sh. 1 FIGURE 1.2-15, Rev. 48 AutoCAD Figure 1_2_15.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-237, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-16 replaced by dwg. M-237, Sh. 1 FIGURE 1.2-16, Rev. 48 AutoCAD Figure 1_2_16.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-240, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-17 replaced by dwg. M-240, Sh. 1 FIGURE 1.2-17, Rev. 48 AutoCAD Figure 1_2_17.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-241, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-18 replaced by dwg. M-241, Sh. 1 FIGURE 1.2-18, Rev. 55 AutoCAD Figure 1_2_18.doc

THIS FIGURE HAS BEEN REPLACED BY DWG.

M-243, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-20 replaced by dwg. M-243, Sh. 1 FIGURE 1.2-20, Rev. 55 AutoCAD Figure 1_2_20.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-244, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-21 replaced by dwg. M-244, Sh. 1 FIGURE 1.2-21, Rev. 56 AutoCAD Figure 1_2_21.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-245, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-22 replaced by dwg. M-245, Sh. 1 FIGURE 1.2-22, Rev. 48 AutoCAD Figure 1_2_22.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-246, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-23 replaced by dwg. M-246, Sh. 1 FIGURE 1.2-23, Rev. 48 AutoCAD Figure 1_2_23.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-247, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-24 replaced by dwg. M-247, Sh. 1 FIGURE 1.2-24, Rev. 48 AutoCAD Figure 1_2_24.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-248, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-25 replaced by dwg. M-248, Sh. 1 FIGURE 1.2-25, Rev. 48 AutoCAD Figure 1_2_25.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-249, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-26 replaced by dwg. M-249, Sh. 1 FIGURE 1.2-26, Rev. 48 AutoCAD Figure 1_2_26.doc

THIS FIGURE HAS BEEN REPLACED BY DWG.

M-251, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-28 replaced by dwg. M-251, Sh. 1 FIGURE 1.2-28, Rev. 56 AutoCAD Figure 1_2_28.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-252, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-29 replaced by dwg. M-252, Sh. 1 FIGURE 1.2-29, Rev. 48 AutoCAD Figure 1_2_29.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-253, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-30 replaced by dwg. M-253, Sh. 1 FIGURE 1.2-30, Rev. 55 AutoCAD Figure 1_2_30.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-254, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-31 replaced by dwg. M-254, Sh. 1 FIGURE 1.2-31, Rev. 55 AutoCAD Figure 1_2_31.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-255, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-32 replaced by dwg. M-255, Sh. 1 FIGURE 1.2-32, Rev. 48 AutoCAD Figure 1_2_32.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-256, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-33 replaced by dwg. M-256, Sh. 1 FIGURE 1.2-33, Rev. 48 AutoCAD Figure 1_2_33.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-257, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-34 replaced by dwg. M-257, Sh. 1 FIGURE 1.2-34, Rev. 48 AutoCAD Figure 1_2_34.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-258, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-35 replaced by dwg. M-258, Sh. 1 FIGURE 1.2-35, Rev. 48 AutoCAD Figure 1_2_35.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-259, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-36 replaced by dwg. M-259, Sh. 1 FIGURE 1.2-36, Rev. 48 AutoCAD Figure 1_2_36.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-260, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-37 replaced by dwg. M-260, Sh. 1 FIGURE 1.2-37, Rev. 48 AutoCAD Figure 1_2_37.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-261, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-38 replaced by dwg. M-261, Sh. 1 FIGURE 1.2-38, Rev. 48 AutoCAD Figure 1_2_38.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-270, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-39 replaced by dwg. M-270, Sh. 1 FIGURE 1.2-39, Rev. 56 AutoCAD Figure 1_2_39.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-271, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-40 replaced by dwg. M-271, Sh. 1 FIGURE 1.2-40, Rev. 48 AutoCAD Figure 1_2_40.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-272, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-41 replaced by dwg. M-272, Sh. 1 FIGURE 1.2-41, Rev. 48 AutoCAD Figure 1_2_41.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-273, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-42 replaced by dwg. M-273, Sh. 1 FIGURE 1.2-42, Rev. 48 AutoCAD Figure 1_2_42.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-274, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-43 replaced by dwg. M-274, Sh. 1 FIGURE 1.2-43, Rev. 48 AutoCAD Figure 1_2_43.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-276, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-44 replaced by dwg. M-276, Sh. 1 FIGURE 1.2-44, Rev. 48 AutoCAD Figure 1_2_44.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-280, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-45 replaced by dwg. M-280, Sh. 1 FIGURE 1.2-45, Rev. 48 AutoCAD Figure 1_2_45.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-281, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-46 replaced by dwg. M-281, Sh. 1 FIGURE 1.2-46, Rev. 48 AutoCAD Figure 1_2_46.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-282, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-47 replaced by dwg. M-282, Sh. 1 FIGURE 1.2-47, Rev. 48 AutoCAD Figure 1_2_47.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-284, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-48 replaced by dwg. M-284, Sh. 1 FIGURE 1.2-48, Rev. 48 AutoCAD Figure 1_2_48.doc AutoCAD: Figure Fsar 1_2_49.dwg FSAR REV.67 FIGURE 1.2-49, Rev 60 UNIT 1 HEAT BALANCE AT RATED POWER WITH 100 X 10 LBm/hr CORE FLOW SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT THIS FIGURE HAS BEEN REPLACED BY DWG.

M-5200, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-50 replaced by dwg. M-5200, Sh. 1 FIGURE 1.2-50, Rev. 52 AutoCAD Figure 1_2_50.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-5200, Sh. 2 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.2-51 replaced by dwg. M-5200, Sh. 2 FIGURE 1.2-51, Rev. 52 AutoCAD Figure 1_2_51.doc THIS FIGURE HAS BEEN INTENTIONALLY LEFT BLANK FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure intentionally left blank FIGURE 1.2-52, Rev. 51 AutoCAD Figure 1_2_52.doc AutoCAD: Figure Fsar 1_2_49_1.dwg FSAR REV.67 FIGURE 1.2-49-1, Rev 60 UNIT 1 HEAT BALANCE AT RATED POWER WITH 108 X 10 LBm/hr INCREASED CORE FLOW SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT 6

AutoCAD: Figure Fsar 1_2_49_2.dwg FSAR REV.67 FIGURE 1.2-49-2, Rev 3 UNIT2 HEAT BALANCE AT RATED POWER WITH 100 X 10 LBm/hr CORE FLOW SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT 6

AutoCAD: Figure Fsar 1_2_49_3.dwg FSAR REV.67 FIGURE 1.2-49-3, Rev 3 UNIT 2 HEAT BALANCE AT RATED POWER WITH 108 X 10 LBm/hr INCREASED CORE FLOW SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT 6

SSES-FSAR Rev. 49, 04/96 1.3-1

1.3 COMPARISON

TABLES

1.3.1 COMPARISONS

WITH SIMILAR FACILITY DESIGNS

This subsection highlights the historical principal design features of the plant and compares its major features with other boiling water reactor facilities. The design of this facility was based on proven technology obtained during the development, design, construction, and operation of boiling water reactors of similar types.

The data, performance, characteristics, and other information presented here represent the then current Susquehan na Steam Electric Station desi gn as it compared to similar designs available at that time. To maintain this original design comparison, these tables will not be revised to reflect current plant des ign, other than the previous addition of the fifth ("E") emergency diesel genera tor to Tables 1.3-6 and 1.3-7.

1.3.1.1 Nuclear Steam Supply System Design Characteristics

Table 1.3-1 summarizes the design and operati ng characteristics for the nuclear steam supply systems. Parameters are related to rated power output for a single plant unless otherwise noted.

1.3.1.2 Power Conversion System Design Characteristics Table 1.3-2 compares the power conv ersion system design characteristics.

1.3.1.3 Engineered Safety Features Design Characteristics

Table 1.3-3 compares the engineered safety features design characteristics.

1.3.1.4 Containment De sign Characteristics

Table 1.3-4 compares the contai nment design characteristics.

1.3.1.5 Radioactive Waste Managem ent Systems Design Characteristics

Table 1.3-5 compares the radioactive waste management design characteristics.

SSES-FSAR Rev. 49, 04/96 1.3-2 1.3.1.6 Structural Design Characteristics Table 1.3-6 compares the struct ural design characteristics.

1.3.1.7 Instrumentation and Electrical Systems Design Characteristics

Table 1.3-7 compares the inst rumentation and electrical systems design characteristics.

1.3.2 COMPARISON

OF FINAL AND PRELIMINARY INFORMATION

All of the significant changes that have been made in the facility design between submission of the last PSAR revision and Revision 0 of the FSAR are listed in Table 1.3-8. Each item in Table 1.3-8 is cross-referenced to the appropriate portion of the FSAR which describes the changes and the bases for them.

SSES-FSARTABLE1.3-3COMPARISON OF ENGINEERED SAFETY FEATURES DESIGN CHARACTERISTICS Page 1 of 2 Rev. 49, 04/96 SSES-FSARTABLE1.3-3COMPARISON OF ENGINEERED SAFETY FEATURES DESIGN CHARACTERISTICS Page 2 of 2 Rev. 49, 04/96

SSES-FSAR Rev. 49, 04/96 Page 1 of 3 TABLE 1.3-8 SIGNIFICANT DESIGN CHANG ES FROM PSAR TO FSAR*

ITEM CHANGE REASON FO R CHANGE FSAR PORTION IN WHICH CHANGE IS DISCUSSED Recirculation flow measurement The recirculation flow measurement design was changed from a flow element to an elbow-tap type. To improve flow measurement accuracy. 7.3.1, 7.6.1 Recirculation system The pressure interlock for RHR shutdown mode was changed. NRC Requirement for diversity. 7.3.1, 7.6.1 Nuclear fuel The number of fuel pins in each fuel bundle has been changed from 7 x 7 to 8 x 8. Improved fuel performance by increasing safety margins.

4.2 Nuclear

boiler An additional test mode was added for closing MSIV's one at a time to 90% of full open in the fast mode (close in slow mode already

existed). Verifies that the spring force on the

valves will cause them to close under loss-of-air conditions.

5.4 Main steam line isolation A main condenser low vacuum initiation of the main steam line isolation was added. NRC requirement 7.3.1 Main steam line isolation Reactor isolation was deleted for high water level initiation actuation. To provide improved plant

availability.

5.4 Main steam line drain system A main steam line drain system was improved. Prevent accumulation of condensate in an idle line outboard of MSLIV.

5.4 Feedwater

sparger The thermal sleeve was changed to provide improved design of sparger to nozzle.

To eliminate vibration, failure, and leakage. 5.3 Standby liquid control (SLC) system Interlocks on the SLC system were revised. To prevent inadvertent boron injection during system testing. 9.3.5 and 7.4.1 SSES-FSAR

Rev. 49, 04/96 Page 2 of 3 TABLE 1.3-8 SIGNIFICANT DESIGN CHANGES FROM PSAR TO FSAR*

ITEM CHANGE REASON FOR CHANGE FSAR PORTION IN WHICH CHANGE IS DISCUSSED RCIC & HPCI steam supply A warmup bypass line and valve was added. Permits pressurizing and pre-warming of the steam supply line downstream to the turbine during reactor vessel heatup. 5.4 and 6.3 RCIC & HPCI vacuum breaker system A vacuum breaker system was added to the turbine exhaust line into the suppression pool. To prevent backup of water in the pipe and consequential high dynamic pipe loads and reactions. 5.4 and 6.3 RCIC & HPCI system Each component has been made capable of functional testing. Improved testability 5.4 and 6.3 Automatic depressurization system (ADS) The interlocks on the automatic depressurization system were revised. To meet IEEE-279 requirements. 7.3.1 RPV code The RPV was partially updated to ASME 1971 code and Summer 1971 addenda. Update to applicable code as much as practical.

5.2 Level

instrumentation The RPV level instrumentation was revised to eliminate Yarway columns and replace them with a conventional condensing chamber type; also, separation and

redundancy features were added.

Improve ECCS separation per IEEE 279 and improve reliability.

7.3.1 Leak detection system The leak detection system was revised to upgrade the capability. To meet IEEE-279 requirements. 7.6.1 Reactor vibration monitoring A confirmatory vibration monitoring test was added. NRC requirement 14.2 SSES-FSAR

Rev. 49, 04/96 Page 3 of 3 TABLE 1.3-8 SIGNIFICANT DESIGN CHANGES FROM PSAR TO FSAR*

ITEM CHANGE REASON FOR CHANGE FSAR PORTION IN WHICH CHANGE IS DISCUSSED Primary Containment Concrete Delineation of compressive strengths for pozzolan vs. non-pozzlan Type II Portland cements. Update to reflect current engineering design requirements.

3.8B RPV Insulation Correct the RPV Insulation Description Revised support beams on as-build RPV Insulation Panels 5.3.3.1.4 Safety Related Conduits & Trays Correct separation statements for conduits and trays. Question 7.4 of Amend. #5 of PSAR (Revised per requirement of Reg.

Guide 1.75 - 1974).

3.12 Tornado Loading Revised Tornado Loading combinations.

To reflect latest NRC recommendations in the Standard Review Plan.

3.3

  • NOTE: Design changes listed are only those which have occurred between the last SSES PSAR Amendment and Revision 0 of the FSAR. The NRC has been notified of all other design changes prior to the last PSAR amendment by previous amendments to the PSAR.

SSES-FSAR NIMS Rev. 62 TABLE 1.6-4 OTHER TOPICAL REPORTS FSAR Rev. 64 Page 1 of 3 REPORT NUMBER TITLE REFERENCED IN FSAR SECTION AE-RTL-788 Void Measurements in the Region of Subcooled and Low Quality Boiling (April 1966) 4.4 ANL-5621 Boiling Density in Vertical Rectangular Multichannel Sections with Natural Circulation (November 1956) 4.4 ANL-6385 Power-to-Void Transfer Functions (July 1961) 4.4 BHR/DER 70-1 Radiological Surveillance Studies at a Boiling Water Nuclear Power Reactor (March 1970) 11.1 BMI-1163 Vapor Formation and Behavior in Boiling Heat Transfer (February 1957) 4.4 CF 59-6-47 (ORNL) Removal of Fission Product Gases From Reactor Off Gas Streams by Adsorption (June 11, 1959) 11.3 ST1-372-38 Kinetic Studies of Heterogeneous Water Reactors (April 1966) 4.4 TID-4500 Relap 3 - A Computer Program for Reactor Blowdown Analysis IN-1321 (June 3970) 3.6 UCRL-50451 Improving Availability and Readiness of Field Equipment Through Periodic Inspection, p. 10 (July 16, 1968) 18.3 WAPD-BT-19 A Method of Predicting Steady-Boiling Vapor Fractions in Reactor Coolant Channels (June 1960) 4.4 ANF-524(P)(A) Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors, Revision 2, Supplement 1 Revision 2 and Supplement 2, Advanced Nuclear Fuels Corporation, Richland WA 99352, November 1990 4.1, 4.4, 15.3 ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," August 1990 5.2, 15.0, 15.1, 15.2, 15.3, 15.5 NE-092-001A "SSES Power Uprate Licensing Topical Report," and NRC letter dated November 30, 1993, from Thomas E. Murley to Robert G. Byram (PP&L).

Subject:

Licensing Topical Report for Power Uprate with Increased Core Flow, Rev. 0, Susquehanna Steam Electric Station, Units 1 and 2 (PLA-3788) (TAC NOS. M83426 and M83427) with 10.2, 15.6 SSES-FSAR NIMS Rev. 62 TABLE 1.6-4 OTHER TOPICAL REPORTS FSAR Rev. 64 Page 2 of 3 REPORT NUMBER TITLE REFERENCED IN FSAR SECTION enclosed Safety Evaluation Report PL-NF-89-005-A "Qualification of Transient Analysis Methods for BWR Design and Analysis," Issue Date: July 1992.

15.1 XN-NF-80-19(P)(A) Volume 1, Supplement 3, "Advanced Nuclear Fuels Methodology for Boiling Water Reactors-Benchmark Results for the CASMO-3G/MICROBURN-B Calculation Methodology," EXXON Nuclear Company, Richland, WA 99352, November 1990 15.1 XN-NF-80-19(P)(A) "Exxon Nuclear Methodology for Boiling Water Reactors; Neutronic Methods for Design and Analysis," Volume 1, and Volume 1 Supplements 1 and 2, March 1983.

15.0, 15.4 XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," February 1987 15.0, 15.3 XN-NF-84-105(P)(A) Volume 1 Supplement 4, "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, Void Fraction Model Comparison to Experimental Data," June 1988 15.3 ANF-91-048(P)(A) "Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model,"

and Correspondence, January 1993.

6.3 XN-NF-80-19(P)(A) Volume 3, Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," Siemens Power Corporation, January 1987 4.1, 15.0 XN-NF-79-59(P)(A) "Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies," November 1979 4.1 EMF-CC-074(P)(A) Volume 1, 2 and 4 "STAIF - A Computer Program for BWR Stability Analysis in the Frequency Domain" 4.1, 4.4 ANF-89-98(P)(A) Rev. 1 and Rev. 1 Supplement 1, "Generic Mechanical Design Criteria for BWR Fuel Designs," Advanced Nuclear Fuels Corporation, May 1995 4.2 NFQM "Nuclear Fuel Business Group Quality Management Manual", NFQM, Rev. 1 Framatome - ANP, U.S. Version, July 2003.

4.2 CENPD-400-P-A "Generic Topical Report for the ABB Option III Oscillation Power Range Monitor (OPRM)" 3.1, 4.4, 7.6 SSES-FSAR NIMS Rev. 62 TABLE 1.6-4 OTHER TOPICAL REPORTS FSAR Rev. 64 Page 3 of 3 REPORT NUMBER TITLE REFERENCED IN FSAR SECTION EMF-93-177 (P)(A)

& SUPPLEMENT 1 "Mechanical Design for BWR Fuel Channels" Siemens Power Corporation, August 2005 4.2 EMF-2209 (P)(A) "SPCB Critical Power Correlation," September 2003 4.1, 4.4 EMF-2158(P)(A) "Siemens Power Corporation Methodology For Boiling Water Reactors - Evaluation and Validation of CASMO-4/Microburn-B2" Rev. 0, October 1999 4.1, 4.3, 4.4, 15.4, 15.5 XN-NF-80-19(P)(A) "Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads", Exxon Nuclear Company, June 1986 5.2, 15.0, 15.1, 15.2, 15.4, 15.5 EMF-2361(P)(A) EXEM BWR-2000 ECCS Evaluation Model", Framatome ANP, May 2001 6.3 ANF-1358(P)(A) "The Loss of Feedwater Heating Transient in Boiling Water Reactors", Advanced Nuclear Fuels Corporation, September 2005 15.1 XN-NF-85-74(P)(A) "RODEX2A(BWR) Fuel Thermal-Mechanical Evaluation Model", Exxon-Nuclear Company, Inc,. August 1986 4.1 MEF-93-177(P)(A)

Rev. 1 "Mechanical Design for BWR Fuel Channels,"

August 2005

SSES-FSAR Table Rev. 47 FSAR Rev. 64 1.7-1 START HISTORICAL 1.7 ELECTRICAL, INSTRUMENTATION, AND CONTROL DRAWINGS

Table 1.7-1 contains a list of non-proprietary electrical, instrumentation and control (EI&C) drawings. This table lists those drawings which were considered to be necessary to evaluate the safety-related features in Chapters 7 and 8 of the Susquehanna Unit 1 and 2 FSAR. All the drawings listed in Table 1.7-1 are considered historical.

END HISTORICAL

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 1 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Abnormal occurrence Any reportable occurrence that is determined by the Commission to be significant from the standpoint of public health or safety Abnormal Operational Transients Infrequent design events that may be reasonably expected during the course of planned operations, including events that are a result of (or follow)a single equipment malfunction or operator error. Power failures, pump trips, and rod withdrawal errors are typical of the single malfunctions or errors which may initiate the events in this category.

Acceptable Demonstrated to be adequate by the safety analysis of the Plant. Accident A single event, not reasonably expected to occur during the course of plant operations, that has been hypothesized for analyses purposes or postulated from unlikely but conceivable situations and that causes or threatens to cause a violation of one or more fission product barriers.

Achieving Criticality All actions which are normally accomplished in bringing the Plant from a condition in which all control rods are fully inserted to a condition in which nuclear criticality is achieved and maintained.

Achieving Shutdown Achieving shutdown begins where power operation ends and includes all actions normally accomplished in achieving nuclear shutdown (more than one rod subcritical) following power operation.

Activated Device A mechanical module in a system used to accomplish an action. An activated device is controlled by an actuation device.

Active components a. Those components whose operability is relied upon to perform a safety function such as safe shutdown of the reactor or mitigation of the consequences of a postulated pipe break in the reactor coolant pressure boundary. b. Active component is one in which mechanical motion must occur to complete the component's intended function.

Active failure The failure of an active component such as a piece of mechanical equipment, component of the electrical supply system or instrumentation and control equipment to act on command to perform its design function. Examples include the failure of a motor-operated valve to move to its correct position, the failure of an electrical breaker or relay to respond, the failure of a pump, fan, or diesel generator to start, etc.

Actuation Device An electrical or electromechanical module controlled by an electrical decision output used to produce mechanical operation of one or more activated devices, thus achieving necessary action.

Additional Plant Capability Event An event which neither qualifies as neither an abnormal operational transient nor an accident but which is postulated to demonstrate some special capability of the Plant.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 2 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Administrative Controls Measures to prevent the existence or development of an unsafe condition in connection with the operation of the reactor. They also define the administrative action to be taken in the event a safety limit or allowed condition for operation is exceeded. Requirements concerning the facility's organization and management, procedures, record keeping, review and audit, and reporting are specified.

Alteration of the Reactor Core The act of moving any component in the region above the core support plate, below the upper grid and within the shroud. Normal control rod movement with the control rod drive hydraulic system is not defined as a core alteration. Normal movement of in-core instrumentation and the traversing in-core probe is not defined as a core alteration.

Alternate Rod Injection An alternate means of inserting control rods. One of the features provided in order to mitigate a postulated anticipated transient without scram (ATWS) event.

Analog channel calibration Adjustment of channel output such that it responds, with acceptable range and accuracy, to know values of the parameter which the channel measures. Calibration shall encompass the entire channel, including alarm or trip, and shall be deemed to include the channel functional test.

Analog channel check A qualitative determination of acceptable operability by observation of channel behavior during operation. This determination may include comparison of the channel with other independent channels measuring the same variable.

Analog channel functional test Injection of a simulated signal into the channel to verify that it is operable, including alarm and/or trip initiating action.

Anticipated operational occurrences Anticipated operational occurrences mean those conditions of normal operation which are expected to occur one or more times during the life of the nuclear power unit and include but are not limited to loss of power to all recirculation pumps, tripping of the turbine-generator set, isolation of the main condensers, and loss of all off-Site power (10 CFR 50 Appendix A).

Anticipated transients (with Scram) This group of anticipated abnormal transients include events which present a demand for protection action by the Reactor Protection System and which have a probability of occurrence greater than 10-³ per year. The events which fall in this category of anticipated transients are listed below: a. Loss of load b. Excessive load increase c. Loss of one feedwater pump d. Loss of flow (one pump) e. Rod withdrawal f. Startup accident g. Accidental depressurization of Reactor Coolant System h. Plant blackout

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 3 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Anticipated Transients Without Scram (ATWS) Anticipated operational occurrences which require reactor shutdown followed by the failure to insert all control rods; ie. a failure to SCRAM.

Associated Areas The off-Site equipment, facilities and structures which are necessary for the operation of the Project. These include the makeup water pump facility, the makeup water pipeline, the discharge structure, the transmission line, the railroad spur, and the rights-of-way and access roads associated with the above.

Average linear power density Total thermal power produced in the fuel rods divided by the total active fuel length of all rods in the core.

Average rod power Total thermal power produced in the fuel rods divided by the number of fuel rods (assuming all rods have equal length).

Channel An arrangement of components and sensors as required to generate a single protective action signal when required by a generating station condition. A Channel loses its identity where single action signals are combined.

Class IE Electric Systems The safety classification of the electric equipments and systems that are essential to emergency reactor shutdown containment isolation, reactor core cooling and reactor heat removal or otherwise are essential in preventing significant release of radiation to the environment.

Closed System Piping system containing fluid, not freely accessible to the environment, penetrating containment but not communicating with either primary coolant pressure boundary or containment atmosphere.

Cold Shutdown When the reactor is in the shutdown mode; the reactor coolant is maintained at equal to or less than 200°F, and the reactor vessel is vented to containment atmosphere.

Common mode failure The failure of two or more components of the same or similar design by the same failure mechanism. Such failure mechanisms for compon ents may result from the adverse conditions from a design basis event for which the components were expected by design to remain functional. Such failures may result from a design deficiency or manufacturing deficiency. Redundant equipment can be made inoperable by this mechanism.

Components Items from which a system is assembled. Containment (primary and secondary) The structures that enclose components of the reactor coolant pressure boundary and which provides an essentially leaktight barrier against the uncontrolled release of fission products to the environment.

Containment Atmosphere Free volume enclosed by the primary containment.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 4 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Containment integrity Exists when all the following conditions exist: a. All nonautomatic containment isolation valves not required for normal operation are closed or under administrative control. b. Blind flanges are properly installed where required. c. The equipment door is properly closed and sealed. d. At least one door in each personnel air lock is properly closed and sealed. e. All automatic containment isolation trip valves are operable or closed. f. The containment leakage satisfies Technical Specification.

Containment isolation Establishment of mechanical barrier(s) in appropriate fluid systems penetrating the Containment which would otherwise represent open paths for the fission products in the event of a loss-of-coolant accident inside the Containment.

Controlled Access Area The area immediately surrounding the principal Project Structures, enclosed with a fence or other suitable physical barrier, such that entry into this area is controlled. This area will encompass the Reactor Buildings, the Turbine Buildings, the Auxiliary Buildings, Control Building, Diesel-Generator Buildings, Radwaste Building, and the Cooling Towers.

Controls Methods and devices by which actuation is used to affect the value of a variable.

When used with respect to nuclear reactors, means apparatus and mechanisms, the actuation of which directly affects the reactivity or power level of the reactor.

Cooldown Cooldown begins where achieving shutdown ends and includes all actions normally accomplished in the continued removal of decay heat and the reduction of nuclear system temperature and pressure.

Critical items Those structures, units (or components) and systems which require a degree of design review, verification, inspection and documentation over and above that applied in the course of normal engineering, procurement and construction. As a minimum, critical items include all structures and systems required to maintain the integrity of the reactor primary system pressure boundary, to provide Containment Engineered Safety Features, assure safe shutdown under all conditions and continued residual heat removal.

Degree of redundancy The difference between the number of sensors of a variable and the number of sensors which when tripped will cause an automatic system trip.

Design Basis "Design basis" means that information which identifies the specific functions to be performed by a structure, system, or component of a facility, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design. These values may be (1) restraints derived from generally accepted "state of art" practices for achieving functional goals, or (2) requirements derived from analysis (based on calculation and/or experiments) or the effects of a postulated accident for which a structure, system, or component must meet its functional goals.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 5 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Design Basis Accidents (DBA) The hypothesized accident whose characteristics and consequences are utilized in the design of those systems and components pertinent to the preservation of radioactive material barriers and the restriction of radioactive material release from the barriers upon occurrence of a loss-of-coolant accident. The potential radiation exposures resulting from a DBA are not exceeded by any similar accident postulated from the same general accident assumptions.

Design Basis Events (DBE) Postulated events used in a design to establish the performance requirements of structures, systems, and components.

Design features Those features of the facility such as materials of construction and geometric arrangements which, if altered or modified, would have a significant effect on safety.

Design Power The power level equal to 102% of the licensed or rated core thermal power level. The design power level is equivalent to 4031 MWt.

Diffuser The submerged section of the discharge pipeline which has multiple ports. Dilution Zone The boundary of the dilution zone is defined as that point where the Plant discharge is mixed with the Susquehanna River.

Discharge structure The diffuser section, connecting discharge pipeline, and anchors, both the shoreline anchors and river bed anchors.

Drywell A pressure-containing envelope surrounding the reactor and its recirculation loops which will channel steam resulting from the LOCA through the suppression pool for condensation. Part of primary containment.

Emergencies Unplanned events characterized by risks sufficient to require immediate action to avoid or mitigate an abrupt or rapidly deteriorating situation.

Emergency Conditions (Infrequent Incidents) Those deviations from Normal Conditions which require shutdown for correction of the conditions or repair of damage in the system. The conditions have a low probability of occurrence but are included to provide assurance that no gross loss of structural integrity will result as a concomitant effect of any damage developed in the system.

Engineered Safeguards (Same as Engineered Safety Features).

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 6 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Engineered Safety Features (ESF) a. Features of a unit or system other than reactor trip or those used only for normal operatio n, that are provided to prevent, limit or mitigate the release of radioactive material in excess of 10 CFR 50.67 limits. b. Engineered Safety Feature System (ESFS) consists of those systems, including essential support systems or components thereof the primary purpose of which during a design basis accident (DBA) will be to: (1) Retain fuel temperatures within design limits by maintaining fuel coolant inventory and temperatures within design limits. (2) Maintain fuel temperatures within design limits by inserting auxiliary negative reactivity. (3) Prevent the escape of radioactive materials to the environment in excess of 10 CFR 50.67 limits by isolation of the systems or structures. (4) Reduce the quantity of radioactivity available for leakage and its potential for leakage by purification, cleanup, containment heat removal and containment pressure reduction. (5) Control the concentration of combustible gases in the containment systems within established limits.

Exclusion area A circle within a radius of 1800 ft from the centerline of the reactors, as defined by 10CFR 100.3. Extended Load Line Limit Analysis Safety analyses performed to demonstrate adequate safety margins in support of a license amendment permitting operation with an elevated load line on the power-flow map; i.e. with increased thermal power at a given recirculation flow.

Failure The termination of the ability of an item to perform its required function. Failures may be unannounced and not detected until the next test (unannounced failure), or they may be announced and detected by any number of methods at the instant of occurrence (announced failure).

Faulted Condition (Limiting Faults) Those combinations of conditions associated with extremely-low-probability, postulated even ts whose consequences are such that the integrity and operability of the nuclear energy system may be impaired to the extent that considerations of public health and safety are involved. Such considerations require compliance with safety criteria as may be specified by jurisdictional authorities.

Functional Test The manual operation or initiation of a system, subsystem, or component to verify that it functions within design tolerances (eg, the manual start of a core spray pump to verify that it runs and that it pumps the required volume of water).

General Design Criteria (GDC) A set of design criteria for structures, systems, and components important to safety, which are given in Appendix A to 10 CFR 50, and provide reasonable assurance that the Plant can be operated without undue risk to the health and safety of the public.

Globe Stop Check Valve (GCK) These valves shall be designed to normally function as check valves, but in addition they shall be provided with means for positive shutoff using manual or mechanical actuators.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 7 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Heatup Heatup begins where achieving criticality ends and includes all actions which are normally accomplished in approaching nuclear system rated temperature and pressure by using nuclear power (reactor critical). Heatup extends through warmup and synchronization of the turbine generator.

High radiation area Any area, accessible to personnel, in which there exists radiation originating in whole or in part within licensed material at such levels that a major portion of body could receive in any one hour a dose in excess of 100 mrem.

Hot shutdown See Technical Specification Section 1.1. Startup/Hot standby See Technical Specification Section 1.1. Immediate Immediate means that the required action will be initiated as soon as practicable considering the safe operation of the unit and the importance of the required action.

Inactive components Those components whose operability (eg, valve opening or closing, pump operation or trip) are not relied upon to perform the system function during the transients or events considered in the respective operating condition categories.

Incident Any natural or accidental event of infrequent occurrence and its related consequences which affect the Plant operation and require the use of Engineered Safety Feature systems. Such events, which are analyzed independently and are not assumed to occur simultaneously, include the loss-of-coolant accident, steam line ruptures, steam generator tube ruptures, etc. A system blackout may be an insolated occurrence or may be concurrent with any event requiring Engineered Safety Feature systems use.

Incident Detection Circuitry Includes those trip systems which are used to sense the occurrence of an incident. Increased Core Flow Operation with core flow greater than 100% of original design. Used to provide additional reactivity at end of core life to permit a longer fuel cycle and more economic operation.

Instrument Calibration An instrument calibration means the adjustment of an instrument signal output so that it corresponds, within acceptable range and accuracy, to a known value(s) of the parameter which the instrument monitors. Calibration shall encompass the entire instrument including actuation, alarm, or trip.

Channel check See Technical Specification Section 1.1. Channel Functional Test See Technical Specification Section 1.1. Irradiated Fuel Fuel that has been in the reactor during reactor operation. Isolated Condition Condition in which the reactor is isolated from the main condenser. Limiting conditions for operation The lowest functional capability or performance levels of equipment required for safe operati on of the facility.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 8 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Limiting safety system settings Settings for automatic protective devices are related to those variables having significant safety functions. (Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting has been chosen so that automatic protective action will correct the most severe abnormal situation anticipated before a safety limit is exceeded).

Linear power density The thermal power produced per unit length of fuel rod (kW/ft). Since fuel assembly geometry is standardized, this is the unit of power density most commonly used. For all practical purposes it differs from kW/liter by a constant factor which includes geometry and the fraction of the total thermal power which is generated in the fuel rod.

Load Group An arrangement of buses, transfers, switching equipment, and loads fed from a common power supply. Local heat flux The heat flux at the outer surface of the cladding (Btu/ft²hr). For nominal rod parameters this differs from linear power density by a constant factor.

Logic That array of components which combines individual bistable output signals to produce decision outputs. Logic channel A logic channel is a group of logic matrices which operate in response to the digital single action signals from the analog channels to generate a protective action signal.

Logic System functional test See Technical Specification Section 1.1. Long Term The remainder of the recovery period following the short term. In comparison with the short term where the main concern is to remain within NRC specified site criteria, the long-term period of operation involves bringing the Plant to cold shutdown conditions where access to the Containment can be gained and repair effected.

Loss-of-Coolant Accident (LOCA) Those postulated accidents that result from the loss of reactor coolant, at a rate in excess of the capability of the Reactor Coolant Makeup System, from breaks of pipes containing reactor coolant, up to and including a break equivalent in size to the double-ended rupture of the largest pipe of the Reactor Coolant System.

Low Population Zone (LPZ) The area included in a three mile radius from the midpoint of the centerline between the two reactor buildings on on Plant Site, and in 10 CFR 100.3 as defined.

Low power physics tests Tests below a nominal five percent of rated power which measure fundamental characteristics of the reactor core and related instrumentation.

Manual Component A component, the operability of which is relied upon to perform a manual nuclear safety function such as providing manual action or operator information required for initiation of action for safe shutdown of the reactor of mitigation of the consequences of an accident.

Material surveillance program The provisions for the placement of reactor vessel material specimens in the reactor vessel, and the program of periodic withdrawal and testing of such specimens to monitor, over the service life of the vessel, changes in the fracture toughness properties of the vessel as a result of neutron irradiation.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 9 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Minimum Critical Heat Flux Ratio (MCHFR) The lowest value of the ratio of critical heat flux (that heat flux which results in transition boiling) to the actual heat flux at the same location.

Minimum degree of redundancy The degree of redundancy below which operation is prohibited or otherwise restricted by the Technical Specifications.

Missile Barrier A Physical barrier which protects essential components, systems or structures from potential missiles arising from consequences of a loss-of coolant accident.

Mode See Technical Specification Section 1.1. Module Any assembly of interconnected components which constitutes an identifiable device, instrument, or piece of equipment. A module can be disconnected, removed as a unit, and replaced with a spare. It has definable performance characteristics which permit it to be tested as a unit. A module could be a card or other subassembly of a larger device, provided it meets the requirements of this definition.

Normal conditions Normal conditions are any condition in the course of system startup, operation in the design power range, hot standby and system shutdown, other than Upset, Emergency, Faulted or Testing Conditions.

Normal operation Operation of the plant under planned, anticipated conditions including, but not limited to, the following: a. Reactor critical (any temperature)

b. Power operation c. Reactor startup d. Reactor shutdown e. Refueling f. Periodic testing g. Nuclear system cooldown h. Nuclear system heatup i. Standby (reactor shutdown, nuclear system maintained at constant temperature)

Nuclear-fueled electrical generating facility (the Plant) The reactor, turbine-generator, cooling tower, associated buildings (reactor building, turbine building, and administration building), and the switchyard.

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

Nuclear Safety Operational Analysis A systematic identification of the requirements for the limitations on plant operation necessary to satisfy nuclear safety operational criteria.

Nuclear Safety Operational Criteria A set of standards used to select nuclear safety operational requirements.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 10 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Nuclear System Generally includes those systems most closely associated with the reactor vessel which are designed to contain or be in communication with the water and steam coming from or going to the reactor core. The nuclear system includes the following: a. Reactor vessel b. Reactor assembly and internals c. Reactor core d. Main steam lines from reactor vessel out to and including the isolation valve outside the Containment e. Neutron monitoring system f. Reactor recirculation system g. Control rod drive system h. Residual heat removal system i. Reactor core isolation cooling system j. Emergency core cooling systems k. Reactor water cleanup system l. Feedwater system piping between the reactor vessel and the first valve outside the Containment m. Pressure relief system Nuclear System Process Barrier See Reactor Coolant Pressure Boundary Occupational dose Include exposure of an individual to radiation (i) in a restricted area; or (ii) in the course of employment in which the individual's duties involve exposure to radiation, provided that "occupational dose" shall not be deemed to include any exposures of an individual to radiation for the purpose of medical diagnosis or medical therapy of such individual.

Operable See Technical Specification Section 1.1. Operating A system or component is operating when it is performing its intended functions in the required manner. Operating Cycle Interval between the end of one refueling outage and the end of the next subsequent refueling outage. Operational The adjective "operational", along with its noun and verb forms, is used in reference to the working or functioning of the Plant, in contrast to the design of the Plant.

Operating reports These reports include the Startup Report, First Year Operation Report, and Semiannual Operating Records. Operating reports are submitted in writing to the Director, Division of Reactor Licensing, USNRC, Washington, D.C. 20545.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 11 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Operating Basis Earthquake (OBE) That earthquake which produces vibratory ground motion for which those structures, systems, and components necessary for power generation are designed to remain operable.

Operator Any individual who manipulates a control of a facility. An individual is deemed to manipulate a control if he directs another to manipulate a control Operator error An active deviation from written operating procedures or nuclear plant standard operating practices. A single operator error is the set of actions that is a direct consequence of a single reasonably expected erroneous decision. The set of actions is limited as follows: a. Those actions that could be performed by only one person. b. Those actions that would have constituted a correct procedure had the initial decision been correct. c. Those actions that are subsequent to the initial operator error and that affect the designed operation of the plant but are not necessarily directly related to operator error.

Passive Component A component in which mechanical mo vement does not occur in order for the component to perform its intended function.

Passive failure The structural failure of a static component which limits the component's effectiveness in carrying out its design function. When applied to a fluid system, this could mean a break in the pressure boundary.

Peaking Factor The ratio of the maximum fuel rod surface heat flux in an assembly to the average surface heat flux of the core.

Penetration Assembly, Elec. Provides the means to allow passage of electrical circuits through a single aperture (nozzle or other opening) in the containment pressure barrier, while maintaining the integrity of the pressure barrier. Pennsylvania Power & Light Company (PP&L) The owner-operator of the Project, having total controlling ownership. Period of recovery The time necessary to bring the Plant to a cold shutdown and regain access to faulted equipment. The recovery period is the sum of the short-term and long-term periods.

Place in Cold Shutdown Condition Conduct an uninterrupted normal Plant shutdown operation until the cold shutdown condition is attained.

Place in Isolated Condition Conduct an uninterrupted normal isolation of the reactor from the main (turbine) condenser including the closure of the main steam line isolation valves.

Place in Shutdown Condition Conduct an uninterrupted normal plant shutdown operation until shutdown is attained.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 12 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Planned Operation Normal plant operation under planned conditions in the absence of significant abnormalities. Operations subsequent to an incident (transient, accident, or special event) are not considered planned operations until the procedures being followed or equipment being used are identical to those used during any one of the defined planned operations. The established planned operations can be considered as a chronological sequence: refueling outage; achieving criticality; heatup; power operation; achieving shutdown; cooldown; refueling outage.

The following planned operations are identified: a. Refueling Outage b. Achieving Criticality c. Heatup d. Reactor Power Operation e. Achieving Shutdown f. Cooldown Plant Those structures, systems and components that make up the Susquehanna Steam Electric Station. Power density The thermal power produced per unit volume of the core (kW/liter). Power Generation When used to modify such words as design basis, action and system, this term indicates that the objective, design basis, action, or system is related to the mission of the Plant, to generate electrical power, as opposed to concerns considered to be of primary safety importance. Thus, the words "power generation" identify aspects of the Plant which are not considered to be of primary importance with respect to safety.

Power Generation Design Basis The power generation design basis for a power generation system states in functional terms the unique design requirements which establish the limits within which the power generation objective shall be met. A safety system may have a power generation design basis which states in functional terms the unique design requirements which establish the limits within which the power generation objective for the system shall be met.

Power Generation Evaluation Shows how the system satisfies some or all of the power generation design bases. Because power generation evaluations are not directly pertinent to public safety, generally they are not included. However, where a system or component has both safety and power generation objectives, a power generation evaluation can clarify the safety versus power generation capabilities.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 13 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Power Generation System Any system, the actions of which are not essential to a safety action, but which are essential to a power generation action. Power generation systems are provided for any of the following purposes: a. To carry out the mission of the Plant-generate electrical power through planned operation. b. To avoid conditions which would limit the ability of the Plant to generate electrical power. c. To facilitate and expedite the return to conditions permitting the use of the Plant to generate electrical power following an abnormal operational transient, accident, or special event.

Power operation condition When the reactor is critical and the neutron flux power range instrumentation indicates greater than two percent of rated power.

Power uprate Evaluations, tests, modifications, setpoint changes, and license amendments which permitted an increase in rated thermal power from the original 3293 Mwt to 3441 Mwt; allowing an increase in the nominal generator rating from approximately 1100 to approximately 1150 MWe, and in the net plant rating from approximately 1050 MWe to approximately 1100 MWe. Extended Power Uprate (EPU): The operating license for both units was further modified to permit operation at 3952 MWt with a nominal generator output of 1300 MWe.

Preferred power source That power supply which is preferred to furnish electrical energy under accident or post accident. It is obtained from start-up transformers. The switchgear is arranged to auto transfer from one preferred source to another preferred source in the event the preferred source fails.

Preferred power system The off-site external commercial power system. Preoperational Test Program The preoperational test program applicable to the nuclear steam supply system is the test program conducted prior to fuel loading. The test program applicable to other Plant systems is the test program conducted prior to that system's required operation.

Principal design criteria The criteria which establish the necessary design, fabrication, construction, testing and performance requirements for structures, systems and components important to safety, that is, structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public.

Principal Project structures The Reactor Buildings, Control Buildings, Diesel Generator Building, Radwaste Building, Turbine Buildings, and Cooling Towers.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 14 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Probable maximum flood (PMF) The hypothetical flood characteristics (peak discharge, volume and hydrograph shape) that are considered to be the most severe "reasonably possible": at a particular location, based on relative comprehensive hydrometeorological analyses of critical runoff producing precipitation (and snowmelt, if pertinent) and hydrological factors favorable for maximum flood runoff.

Refer to Chapter 2 of the SSES FSAR for specific values which apply to the Susquehanna Steam Electric Station.

Probable maximum precipitation (PMP) The theoretically greatest precipitation over the applicable drainage area that would prod uce flood flows that have virtually no risk of being exceeded.

Refer to Chapter 2 of the SSES FSAR for specific values which apply to the Susquehanna Steam Electric Station.

Probable maximum winds The hypothetical tornado or other cyclonictype windstorm that might result from the most severe combinations of meteorological parameters that are considered reasonably possible in the region involved, if the tornado or other type windstorm should approach the point under study along a critical path and at optimum rate of movement.

Protection System The aggregate of the protective signal system and the protective actuator system Protective action a. Protective action at the channel level is the generation of a signal by a single channel when the variable(s) sensed exceeds a limit. b. Protective action at the system level is the operation of sufficient actuated equipment to accomplish a protective function (for example: rapid insertion of control rod, closing of containment isolation valves, safety injection, core spray).

Protective Actuator System An arrangement of components that performs a protective action when it receives a signal from the protective signal system (for example: control rods, their drive mechanisms and their trip mechanisms; isolation valves, their operators and their contractors; core spray pumps, their motors and circuit breakers).

Protective function Any one of the functions necessary to limit the safety consequences of a design basis event (for example: rapid reduction of reactor power following a control rod ejection, isolation of the Containment following a steam line break, removal of heat from the core following a loss-of-coolant-accident.

Quality Assurance (QA) All those planned and systematic actions necessary to provide adequate confidence that a structure, system or component will perform satisfactorily in service. Quality assurance includes quality control, which comprises those quality assurance actions related to physical characteristics of a material, structure, component, or system which provide a means to control the quality of the material, structure, component, or system of predetermined requirements.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 15 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Quality Control (QC) Those quality assurance actions related to physical characteristics of a material, structure, component, or system which provide a means to control the quality of the material, structure, component or system to predetermined requirements.

Q-Listed system Q-Listed systems, structures and components are those which prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. They include materials, structures, and equipment whose failure could cause significant release of radioactivity to the environment comparable to 10 CFR 50.67 limits at the Site exclusion distance, or which are vital to the safe shutdown of the Plant, or which are necessary for the removal of decay and sensible heat from the reactor.

Quality Group A classification which identifies the importance of structures, systems, and components with respect to Plant safety functions in accordance with definitions given in NRC Regulatory Guide 1.26. Radiation area Any area, accessible to personnel, in which there exists radiation originating in whole or in part within licensed material, at such levels that a major portion of the body could receive in any one hour a dose in excess of 5 mrem, or in any five consecutive days a dose in excess of 100 mrem.

Radioactive Material Barrier Includes the systems, structures, or equipment that together physically prevent the uncontrolled release of radioactive materials. The barriers are the fuel cladding, the reactor coolant system and the Containment.

Radioactive waste Radioactive wastes are solids, liquids, and gaseous effluents from the radioactive waste systems that have concentration or radioactivity in excess of background.

Rated power The power level at which the reactor is producing 100 percent of reactor vessel rated steam flow. This is the maximum power that could be authorized by the operating license. Rated coolant flow, rated neutron flux and rated nuclear system pressure refer to values of these parameters when the reactor is at rated power.

Reactivity A state variable of neutron chain reactions which is indicative of a deviation in the chain reaction from criticality. It is measured in terms of where p=keff-1/keff. Positive valves correspond to a supercritical state and negative values to a subcritical state.

Usage has established "units" of delta k/k for reactivity change. This term (delta k/k) is used to represent a departure from criticality, and is referred to as reactivity worth. Reactivity worth is the reactivity attributable to the specified component material, portion of material, or void in the nuclear reactor.

Reactor Building Structural complex enclosing the primary containment, and forming secondary containment. Reactor Coolant System The vessels, pipes, pumps, tubes, valves and similar process equipment that contain the steam, water, gases, and radioactive materials coming from, going to, or in communication with the reactor vessel.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 16 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Reactor coolant pressure boundary (RCPB) All those pressure-containing components such as pressure vessels, piping, pumps and v alves, which are (1) part of the Reactor Coolant System, or (2) connected to the Reactor Coolant System, up to and including any and all of the following: a. The outermost containment isolation valve in system piping which penetrates primary reactor Containment. b. The second of two valves normally closed during normal reactor operation in system piping which does not penetrate primary reactor Containment. The Reactor Coolant System safety and relief valves.

Reactor critical When the neutron chain reaction is self-sustaining and keff = 1.0. Reactor Power Operations Reactor power operation begins after heatup is complete and includes any operation with the mode switch in the "Startup" or "Run" position with the reactor critical and above 1 percent rated power. Reactor Vessel Pressure Unless otherwise indicated, reactor vessel pressures are those measured by the reactor vessel steam space detectors.

Redundant equipment or system An equipment or system that duplicates the essential function of another equipment or system to the extent that either may perform the required function regardless of the state of operation or failure of the other.

Refueling Mode See Technical Specification Section 1.1. Refueling operation condition Any operation within the Containment involving movement of core components when the vessel head is completely unbolted or removed and there is fuel in the reactor.

Refueling Outage The period of time between the shutdown of the unit prior to a refueling and the startup of the unit after that refueling. For the purpose of designating frequency of test ing and surveillance, a refueling outage shall mean a regularly scheduled outage. However, where such outages occur within 8 months of the completion of the previous refueling outage, the required surveillance testing need not be performed until the next regularly scheduled outage.

Refueling shutdown condition When the reactor is subcritical by at least 10,000 pcm, Tavg is 140°F, and fuel or fuel inserts are scheduled to be moved to or from the reactor core.

Reliability The probability that a component will perform its specified function without failure for a specified time in a specified environment.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 17 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Reportable Occurrence They are as follows: 1) Failure of the reactor protection system or other systems settings to initiate the required protective function by the time a monitored parameter reaches the setpoint specified as the limiting safety-system setting in the technical Specifications or failure to complete the required protective function. 2) Operation of the unit or affected systems when any parameter or operation subject to a limiting condition for operation is less conservative than the least conservative aspect of the limiting condition for operation established in the technical specifications. 3) Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment.

4) Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady-state conditions during power operation greater than or equal to 1% k/k; a calculated reactivity balance indicating 5 shutdown margin less conservative than specified in the technical specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if subcritical, an unplanned reactivity insertion of more than 0.5% k/k; or occurrence of any unplanned criticality. 5) Failure or malfunction or one or more components which prevents or could prevent, by itself, the fulfillment of the functional requirements of system(s) used to cope with accidents analyzed in the SAR. 6) Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems(s) used to cope with accidents analyzed in the SAR. 7) Conditions arising from natural or manmade events that as a direct result of the event, require plant shutdown, operation of safety systems, or other protective measures required by the technical specifications.
8) Errors discovered in the transient or accident analyses or in the method used for such analysis as described in the safety analysis report or in the bases for the technical specifications but have or could have permitted reactor operation in a manner less conservative than assumed in the analyses 9) Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than that assumed in the accident analyses in the SAR or technical Specifications bases; or discovery during plant life of conditions not specifically considered in the SAR or technical Specifications that require remedial action or corrective measures to prevent the existence or development of an unsafe condition.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 18 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Response spectrum A plot of the maximum response of single degree of freedom bodies, at a damping value expressed as a percent of critical damping, of different natural frequencies, mounted on the surface of interest (that is, on the ground for the ground response spectrum or on the floor of a building for that floor's floor response spectrum) when the surface is subjected to a given earthquake's motion.

NOTE: The response spectrum is not the floor motion or the ground motion.

Restricted area Any area access which is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials. "Restricted Area" shall not include any areas used as residential quarters, although a separate room or rooms in a residential building may be set apart as a restricted area.

Rod power or rod integral power The length integrated linear power in one rod (kW). Run Mode See Technical Specification Section 1.1. Safe Shutdown Earthquake (SSE) The "Safe Shutdown Earthquake" is that maximum probable earthquake which produces the vibratory ground motion for which structures, systems and components designed to Seismic Category I requirements remain functional.

Safety When used to modify such words as objective, design basis, action, and system, the word indicates that, that objective, design basis, action, or system is related to concerns considered to be of safety significance, as opposed to the Plant mission - to generate electrical power. Thus, the word "safety" identifies aspects of the plant which are considered to be of importance with respect to safety. A safety objective or safety design basis does not necessarily indicate that the system is an engineered safety feature.

Safety Action An ultimate action in the Plant which is essential to the avoidance of specified conditions considered to be of safety significance. The specified conditions are those that are most directly related to the ultimate limits on the integrity of the radioactive material barriers and the release of radioactive material. There are safety actions associated with planned operation, abnormal operational transients, accidents, and special events. Safety actions include such actions as reactor scram, emergency core cooling, reactor shutdown from outside the control room and the indication to the operator of the values of certain process variables.

Safety Class A classification which identifies the importance of structures systems, and components with respect to Plant functions in accordance with definitions given in ANSI N212 for BWR's Safety Design Basis The safety design basis for a safety system states in functi onal terms the unique design requirements that establish the limits within which the safety objective shall be met. A power generation system may have a safety design basis which states in functional terms the unique design requirements that ensure that neither planned operation nor operational failure by the system results in conditions for which Plant safety actions would be inadequate.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 19 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Safety Evaluation Shows how the system satisfies the safety design bases. A safety evaluation is performed only for those systems that have safety design bases. Safety evaluations form the basis for the Technical Specifications and establish why specific safety limitations are imposed.

Safety limits Limits upon important process variables which are found to be necessary to reasonably protect the integrity of certain of the physical barriers which guard against the uncontrolled release of radioactivity.

Safety Related See Section 17.2.2 for this definition Safety System Any system, group of systems, components, or groups of components the actions of which are essential to accomplishing a safety action.

Scram Refers to the automatic rapid insertion of control rods into the reactor core in response to the detection of undesirable conditions.

Seiche An oscillation of the surface of a lake or landlocked sea that varies in period from a few minutes to several hours and is thought to be initiated chiefly by local variations in atmospheric pressure aided in some instances by winds and tidal currents and that continues for a time after the inequalities of atmospheric pressure have disappeared.

Seismic Category I Plant features required to assure 1) the integrity of the reactor coolant pressure boundary, 2) the capability to shut down the reactor and maintain it in a safe shutdown condition, or 3) the capability to prevent or mitigate the consequences of accidents which could result in potential off-Site exposures comparable to the guideline exposure of 10 CFR 50.67.

Plant features required to meet NRC GDC-1 of Appendix A to 10 CFR 50 and Appendix B of 10 CFR 50. Plant features required to meet NRC GDC-2 of Appendix A to 10 CFR 50 and Proposed Appendix A to 10 CFR 100. Plant features designed to withstand effects of the Safe Shutdown Earthquake.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 20 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Seismic Category II (Non-Seismic Category I) Plant features not required to assure 1) the integrity of the reactor coolant pressure boundary, 2) the capability to shut down the reactor and maintain it in a safe shutdown condition, or 3) the capability to prevent or mitigate the consequences of accidents which could result in exposures comparable to the guidelines of exposures of 10 CFR 50.67.

Plant features not required to meet NRC GDC-1 of Appendix A to 10 CFR 50 and Appendix B to 10 CFR 50.

Plant features not required to meet NRC GDC-2 of Appendix A to 10 CFR 50 and proposed Appendix A to 10 CFR 100. Plant features not designed to withstand the effects of the Safe Shutdown Earthquake.

Senior Operator Any individual designated by a facility licensee under 10 CFR 50 to direct the licensed activities of licensed operators.

Sensor That part of a channel used to detect variations in a measured variable. Service conditions Environmental, power, and signal conditions expected as a result of normal operating requirements, expected extremes in operating requirements, and postulated conditions appropriate for the design basis events of the station.

Short term The time immediately following the incident during which automatic actions are performed, system responses are checked, type of incident is identified and preparations for long-term recovery operations are made. The short term is the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following initiation of system operations after the incident.

Shut Down See Technical Specification Section 1.1. Shutdown Mode See Technical Specification Section 1.1. Simulated Automatic Actuation Simulated automatic actuation means applying a simulated signal to sensor to actuate the circuit in question.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 21 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Single failure a. An occurrence which results in the loss of capability of a component to perform its intended safety functions. Multiple failures resulting from a single occurrence are considered to be a single failure. Fluid and electrical systems are considered to be designed against and assumed single failure if neither (1) a single failure of any active component (assuming passive components function properly) nor (2) a single failure of any passive component (assuming active components function properly) results in a loss of the capability of the system to perform its safety functions. b. Single failures are spontaneous occurrences imposed upon safety systems that are required to respond to a design basis event. They are postulated in spite of the fact that they were designed to remain functional under the adverse condition imposed by the accident. No mechanism for the cause of the single failure need be postulated. Single failures of passive components in electrical systems should be assumed in designing against a single failure.

Site Features Those features that are important to safety by virtue of the physical setting of the Plant.

Spring Loaded Piston Actuated Check Valve (SLPACK) Spring loaded piston actuated check valves operate as follows for the following modes: a. During Normal Flow

A spring loaded piston operator is held open by air pressure. Meanwhile, the valve is fully open by action of force due to flow alone. b. During Accidental Loss of Operator Air
The valve shall remain in the fully open position when the flow rate is equal to or greater than the normal flow rate indicated in the Valve Data Sheets. With a flow rate less than normal, the valve may be partially open due to the force of spring against force due to flow. c. Upon Reversal of Flow
Valve shall tightly shut as a normal check valve. In addition, the Control room operator will assist in starting valve closure by sending a remote signal to open a fail-open solenoid valve, releasing air pressure from the operator cylinder. All signal wiring will be furnished by others.

Standby power source The power supply that is selected to furnish electrical energy when the preferred power supply is not available. It consists if an electrical generating unit and all necessary auxiliaries, usually a diesel generator set.

Standby power system Those on Site power sources and their distribution equipment provided to energize devices essential to safety and capable of operation independently of the preferred power system.

Startup Mode See Technical Specification Section 1.1. Startup testing After fuel has been loaded into the reactor, testing is conducted under conditions similar to those for Hot Functional Testing with the reactor subcritical to complete those tests which could not be completed during the initial hot functional testing and those which must be done with the core in position.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 22 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Suppression Pool A pool of water, located in the lower section of the Containment. During relief-valve discharge and postulated LOCA's, it serves as a heat sink and a pressure-suppression water pool comparable to the pool in the torus or suppression chamber of earlier BWR plants.

Surveillance Frequency See Technical Specification Section 1.4. Surveillance requirements Requirements relating to test, calibration or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within the safety limits and that the limiting conditions of operation will be met.

Swing bus A bus that is automatically transferred to one or the other of two redundant standby power sources. System Redundancy System that duplicates an essential function of another system to the extent that either may perform a required function regardless of the state of operation or failure of the other.

Technical Specifications (as used in the FSAR) Encompass the nuclear safety operational requirements and limits to be used by plant operations and management personnel. They are prepared in accordance with the requirements of 10 CFR 50.36 and are incorporated by reference into the operating license issued by the U.S. Nuclear Regulatory Commission.

Testing Conditions Testing conditions are those tests in addition to the ten (10) hydrostatic or pneumatic tests permitted by ASME Section III, paragraphs NB-6222 and NB-6322 including leak tests or subsequent hydrostatic tests.

Testable Check Valve These valves are designed to normally function as a check valve, but in addition, they shall be provided with a manual test lever to prove operability during shutdown.

Test Duration The elapsed time between test initiation and test termination. Test Interval The elapsed time between the initiation of identical tests. Thermal power The total core heat transfer rate from the fuel and the coolant. Tornado criteria The design parameters applicable to the design tornado, such as rotational and translational velocities, design pressure differential and associated time interval and the tornado-generated missile impact load with a statement of whether the imposed loads will be established simultaneously in establishing the tornado design.

Transition Boiling Transition boiling means the boiling regime between nucleate and film boiling. Transition boiling is the regime in which both nucleate and film boiling occur intermittently with neither type being completely stable.

Trip The change of state of a bistable device that represents the change from a normal condition. A trip signal, which results from a trip, is generat ed in the channels of a trip system and produces subsequent trips and trip signals throughout the system as directed by the logic.

Trip System That portion of a system encompassing one or more channels, logic and bistable devices used to produce signals to the actuation logic. A trip system terminates and loses its identity where outputs are combined in logic.

SSES-FSAR NIMS Rev. 52 FSAR Rev. 64 Page 23 of 23 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Type tests Tests made on one or more units to verify adequacy of design. Ultimate Heat Sink The spray pond and associated structures and components. Unrestricted area Any area access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials and any area used for residential quarters.

Upset Conditions (Incidents of Moderate Frequency) Any deviations from Normal Conditions anticipated to occur often enough that design should include a capability to withstand the conditions without operational impairment. The Upset Conditions include those transients which result from any single operator error or control malfunction, transients caused by a fault in a system component requiring its isolation from the system and transients due to loss of load or power. Upset Conditions include any abnormal incidents not resulting in a forced outage and also forced outages for which the corrective action does not include any repair of mechanical damage. The estimated duration of an Upset Condition shall be included in the Design Specifications.

THIS FIGURE HAS BEEN REPLACED BY DWG.

M-100, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.8-2A replaced by dwg. M-100, Sh. 1 FIGURE 1.8-2A, Rev. 55 AutoCAD Figure 1_8_2A.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-100, Sh. 2 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.8-2B replaced by dwg. M-100, Sh. 2 FIGURE 1.8-2B, Rev. 56 AutoCAD Figure 1_8_2B.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-100, Sh. 3 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.8-2C replaced by dwg. M-100, Sh. 3 FIGURE 1.8-2C, Rev. 48 AutoCAD Figure 1_8_2C.doc THIS FIGURE HAS BEEN REPLACED BY DWG.

M-100, Sh. 4 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 1.8-2D replaced by dwg. M-100, Sh. 4 FIGURE 1.8-2D, Rev. 55 AutoCAD Figure 1_8_2D.doc

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