ML21294A150
| ML21294A150 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 10/12/2021 |
| From: | Talen Energy, Susquehanna |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML21294A245 | List:
|
| References | |
| PLA-7935 | |
| Download: ML21294A150 (62) | |
Text
{{#Wiki_filter:SSES-FSAR Text Rev. 55 1.8 SYMBOLS AND TERMS USED IN ENGINEERING AND TEXT 1.8.1 TEXT DEFINITIONS AND ABBREVIATIONS Definitions used throughout the FSAR are listed in the Glossary of Terms, Table 1.8-1. Acronyms and technical abbreviations are listed in Tables 1.8-2 and 1.8-3, respectively . 1.8.2 DAAWING INDEX AND SYMBOLS Abbreviations used on drawings are listed in Table 1.8-5. Symbols used on GE supplied Pfping and Instrument Diagrams (P&ID's) are shown on Figure 1.8-1. Symbols for other P&ID's are shown on-Dwgs. M-100, Sh. 1, M-100, Sh. 2, M-100, Sh. 3, and M-100, Sh. 4. Logic Symbols and Instrument Symbols are shown on Figures 1.8-3 and 1.8-4, respectively. 1.8.3 PIPING IDENTIFICATION Piping is identified on the Piping and Instrument Diagrams {P&ID's) by a three-group identifier. The first group is the nominal pipe size in inches; the second is a three-letter group for the pipe class; and the third is a three-digit grou*p sequentially assigned Vvithin a pipe class. Example: 6"-HBD-117 Size Class Sequence The three-letter group for the pipe class is described in detail in Table 1.8-6. The three-digit sequence number is assigned consecutively to rdentify specific lines tn a pipe class as foflows: Piping common to both units 0-99 and 3001-3999 Piping for Unit 1 100-199 and 1000-1999 Piping for Unit 2 200-299 and 2000-2999 1.8.4 VALVE IDENTIFJCATION All manual and remotely operated vatves wifl have unique identification numbers for tracking purposes and 'Nill be shown on the P&ID's. Listed below ?re the numbering systems used for each group of vatves. AU manua, valves, except those which have a GE Master Parts List (MPL) number, and those valves supplied by vendors as part of the equipment package and not installed by Bechtel/PPL 'Nill be identified by the following method: FSAR Rev. 59 1.8~1
SSES-FSAR Text Rev. 55 1 006 t I r I I I Unit No. ---------------------------------- : 0-Cornmon 1-Unit 1 2-Unit 2 I System Identification I I (last 2 digits of P&IDs ----------------------------------- Sequence No. (3 digit numbers) Remote operated valves which do not have a GE MPL number! are identified by the operator number, e.g.: I I I I I I I Valve type _______________________________ J I I
'II I I
t I I I I I I Unit No.-------------------------------------------------J ' I I I I , I I I P&ID No. I I (last 2 digits) ---------------------------------------------------------. I I Sequence No. ------------------------- -------------- -- ---- -- - - --- -- --- --------- - - - - - : Those valves in GE's MPL are identified by the GE numbering systemt e.g.: E11 HV 1 F031 I I I l r I MPL System No. ------------------------- : (Referenced on figure notes) Valve Type ------- ------- ------------- -------- -- - ------ I I I 1 I I I Unit No. -------------------------------------------------------------. I I I I GE Valve No. ------ ------------------------------------~- ____________________________! FSAR Rev. 59 1.8-2
SSES-FSAR Text Rev. 55 Valves that are not numbered but are supplied as part of vendor mounted equipment will be identified in the vendor's operation and maintenance manuals. This is to avoid duplication of numbering these valves. 1.8.5 INSTRUMENT IDENTIFICATION 1.8.5.1 Instrument Components ldentffication of instruments and control devices is made by the use of one of the following numbering systems:
- 1. Instruments and devices within GE scope of design are numbered in accordance with the
~GE MPL system. Associated devices shown on P&ID's but without any numerical identity are numbered as in 2 betow.
- 2. Except as in 1 above, instrument identifications are based on Instrument Society of America (ISA) Standard S5.1~1973, as modified by Dwgs. M-100. Sh. 1, M-100, Sh. 2.
M-100, Sh. 3, and M-100, Sh. 4. In general, each instrument or device in a measurement loop is assigned the same number. however, loops containing instruments and devices identified in the GE MPL system are an exception to this rule. When a loop contains more than one instrument component of the same functional type. a suffix letter will be added and used to establish a unique identity for those components. Redundant measurement loops will be identified by the addition of a suffix letter to each instrument component or device in the loop. rn the case of redundant loops containing more than one instrument of the same functional type, the suffix letter will be foflowed by a number. Instrument and device numbers are constructed as follows: 1 .11 A1 I I I I I I I I I I 'I I t Functional Identification : I Per 8856-M-100 . ______________________________ _; I I I I I I Unit Number ______________________ ----------------------- ~ Last 2 Digits of P&IDNumber Loop Number Suffix ---------------------------- ---------------------------- ---------------------------- FSAR Rev. 59 1.8-3
SSES-FSAR Text Rev. 55 A zero in the unit number position indicates that the instrument or device is common to both units. 1.8.5.2 Instrument Location Instrument components and devices are mounted on racks and panels which are identified by a 5 character, alpha-numeric code. This code is marked adjacent to the instrument component identifier, as shown on Dwg. M-100, Sh. 1. The code numbers identify the unit number and the general location of the rack or panel by the following block-number assignment: C001 - C099 NSSS Local Panels and Racks C101 - C199 Turbine Building C201 - C299 Reactor Building C301 - C399 Radwaste Building C401 -C499 Primary Containment C501 - C599 Miscellaneous Locations C601 - C699 Control Structure C701 - C799 Administration Building A prefix digit is used to identify the unit or common plant assignment. With each block-number assignment above, the series from 076 through 099 are reserved for local racks and panels in heating and ventilation service. The following examples illustrate typical rack or panel assignments: 1 50 I I I I I I I I I I I I Unit Number -----------------------------..! I I I I I I I I
l
= Rack or Panet -------- -----------..-----------~
= Component Box------------------------------~
. . I I
I
= Plant Computer -------------------,------ ____ 1
'I I
I I I I I Control Structure ------------------------------------------------------ Panei Number FSAR Rev. 59 . 1,8-4
SSES-FSAR Text Rev. 55' 1.8.6 ELECTRICAL COMPONENT IDENTIFICATION This section describes the methods used to identify electrical equipment locations and to number electrical schemes, cables, and raceways. Additional information is contained in Section 8.3. 1.8.6.1 Equipment Location Numbers Each piece of electrical equipment is identified by an equipment number. To facilitate cable routing from one equipment loc.ation to another. a location number is also assjgned to each piece of electrical equipment. Generally, the equipment number and equipment location number for a specific piece of electrical equipment are identical. For large pieces 0f electrical equipment, such as switchgear, load centers, and motor control centers, which are compartmentalized , the - equipment location number consists of the basic equipment number plus additional suffixed information to identify a location within the equipment itsetf. The following two examples illustrate equipment location numbers: 1 1 01 I Unit Number - (1) *-------------------------- :I I I Equipment Ctassification - (Transformer)------------------~ I I Plant Area - (Turbine Building) __________________________________________ _1 I I I I < Sequential Number - (Transfonner No. 1) ---------------------------****------------- I 1 1 11 1 I I I I I Unit Number - (1) I t I I I I Equipment Classification - (Motor : *I I Control Center (MCC) )-- ---- - - ------ - - - ----- -- - - - --' . ,I I Plant Area - (Turbine Building) _______________________________ J I I I I Sequential number - (MCC No. 11) _________________________ ., ___________ _.I I I I I I I I t I l I Stack Number - (12. left to right)--------------------------------------------------- -' I I I I Cubicle Number - (1, top to bottom) ------ ---------- -- - --- --- -------- - - - - -- --- --- ----- - -- - - - -' In the first example, the equipment number and equipment location number for transfonners 1X101 are identical. !n the second example, the basic MCC equipment number 18111 is suffixed to establish an equipment location number, 1B 111121, which identifies a specific compartment within the MCC. FSAR Rev. 59 1.8-5
SSES-FSAR Text Rev. 55 To distinguish one piece of electrical equipment from other duplicate equipment used in the same service, a suffix fetter is added to the basic equipment number to establish individual equipment location numbers. For example, if main transfonner 1X101 is composed of a bank of three single phase transformers, the transformers for phases A, Band Care identified with equipment location numbers 1X101A. 1X101B and 1X101C, respectively. Equipment location numbers are generally assigned to items listed rn the circuit and raceway schedufes. Accordingly, most electrical equipment related to systems such as lighting, communications, and cathodic protection is not included. Electrical equipment which is an integral part of mechanical equipment is assigned the same number as the mechanical equipment. All major pieces of electrical equipment are listed in an equipment index. The equipment index provides a description of the equipment and identifies pertinent drawings such as applicable electrical layout drawings and P&ID's. 1.8.6.2 Scheme Numbers Each electrical scheme is identified by a six character number. The f;rst character is numeric and refers to the plant unit number for which the scheme is applicable. The second character is alphabetic and classifies the scheme by major plant system. The last four characters are numeric, with the exception of GE supplied cables, and provide a sequential, but arbitrary, identity for each scheme. Given below is an example of a typical scheme number. 1 Q I I I Unit Number - (1} *--------------------------------------~ I I I I I I I Plant System - (Nuclear Steam Supply System) __________________________ J I Scheme Sequential Number - (Arbitrary No. for RHR Pump 1A) _________________________: A log of all schemes is maintained in the scheme number index which contains pertinent information such as scheme description, scheme drawing number and source drawing number. 1.8.6.3 Scheme Cable Numbers Except for cabling associated with the plant lighting, communications, and cathodic protection systems, each cable in the plant is identified by a scheme cable number composed of nine characters. The first character is alphabetic and indicates the separation group to which the cable belongs. The second character is afso alphabetrc and denotes the system voltage !eve,. Characters three through eight identify the six character scheme number to which the cable is assigned. The ninth and finaf character is alphabetic, except for GE supplied cables, and provides a distinctive identity to each cable in the block diagram shown on the scheme drawings. The following two examples rllustrate typical sche111e cable numbers: FSAR Rev. 59 1.8-6
SSES-FSAR Text Rev. 55 100501 .B I* I I I I I I I I I ______________ .JI I I I I Separation Group - (Safeguard Channel A) I I I t I I I i Voltage Level -(120 VAC to 250 VDC Control) _______________________ J !
- 1 t
I Scheme Number - ( 1Q0501 for RHR Pump 1A) . ______________________
- _. ______ J Cable Identity- (Cable R in Block Diagram) -----------------------------------------------*
N M 1R3007 D I I I t I I I I
- I I Separation Group - (Non-Safety Related) ---------------- I I I
I I I I I I Voltage Level - (Low Level Instrumentation) _________________________ J I I I I I Scheme Number - (Radwaste Bldg. Sump Pump A) _____________________________ ; I 1 Cable Identity (Cable Din Block Diagram) -- ----------------------- - ------------- __ ----- - - - An alpha-numeric listing of all scheme cable numbers is maintained in the electrical circuit schedule. The circuit schedule also identifies the cable type, quantity of conductors, from and to equipment location numbers, and the cable routing. The circuit schedule uses the first t\vo characters of the scheme cable number as a facility code, ensures that separation and voltage criteria are not violated. A cable marker is affixed to each end of the cable for peITTtanent identmcation. Cable marl<ers for Class IE cables have distinguishing colors for each separation group. Additionally, all Class 1E cables are marked at regular intervals along their length with colors corresponding to the cable marker colors. 1.8.6.4 Raceway Numbers All scheduled electrical cable trays, conduits, conduit sleeves and junction boxes are identified by six character raceway numbers; PGCC ducts by seven character numbers and manholes by five character numbers. The two examples given below illustrate typical raceway numbers for engineered safety feature and non-safety feature cable trays, respectively. FSAR Rev. 59 1.8-7
. SSES-FSAR Text Rev. 55 6 !S B 99 I I I I Separatron Group - (Safeguard Channel A) ______ J I
I I I I I I t Unit Number - (1) ________________________________________ J I r I I I I Voltage Level - (120 VAC to 250 voe Control) ________________________ J
'I I
Main or Branch Run - (B) ______________________________________________________ __: I Section Number- (Tray Section 99) ------------------------------------------------------- -: 1 I
.e C I )
I I I I t t I r Unit Number - (1) ______________________________ J ( I I I I I I I I I I Voltage Level - (250 VDC to 480 VAC Power) ---*-----------~ I I I I I Main Run - (Main Tray B) ----------------------------------------:----.J l Branch Run - (Branch Tray C) __________________________________________________ J I Section Number - (Tray Section 85) *--- ______________________________
- _____________________:
The first character of each engineered safety feature cable tray is an alphabetic letter that relates to the. first character of each engineered safety feature scheme cable eligibfe for routing, therein. Non-safety feature cable tray, 'Nhose first character is numeric representing the unit number, may only contain scheme cable numbers prefixed by the letter N. This same practice was followed for conduit numbers as shown below. 8 1 E I r I Separation Group - : I I (Safeguard Channel A) ____________________ .J' I I I I r Unit Numb~r - (1) --------------- ----------- - * ---------- 2 Voltage Level - I (250 VDC to 480 VAC Power) __________________________________________ J I I I I I I I I I Conduit Sequential Number - (Arbitrary No.) ------------------------------------------ FsA*R Rev. 59 1.8-8
SSES-FSAR Text Rev. 55 1 .ti 005 Unit Number - (1) I I I I . Voltage Level - (13.SKV Power) __________________________ _: I I I I I Plant Area - (Turbine Bldg. Elev. 656') ------------------------------------' I Conduit Sequential Number - (Arbitrary No.) ________________________________________ ----~ An afphanumeric listing of all raceway numbers is maintained in the electrical Raceway Schedule, which also contains the raceway type, length from end point locations, percent fill, and list of included cables. Raceway markers are affixed to each raceway for permanent identification. Identification mali<ers for Class IE raceways are marked at regurar intervals along the length of the raceway with unique and distinguishing colors for each separation group corresponding to the cable marker colors. FSAR Rev. 59 1.8-9
SSES-FSAR NIMS Rev. 52 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 components 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. FSAR Rev. 64 Page 1 of 23
SSES-FSAR NIMS Rev. 52 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 facilitys 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 FSAR Rev. 64 Page 2 of 23
SSES-FSAR NIMS Rev. 52 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 components 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. FSAR Rev. 64 Page 3 of 23
SSES-FSAR NIMS Rev. 52 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. FSAR Rev. 64 Page 4 of 23
SSES-FSAR NIMS Rev. 52 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 Those deviations from Normal Conditions which require shutdown for correction of the conditions (Infrequent Incidents) 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). FSAR Rev. 64 Page 5 of 23
SSES-FSAR NIMS Rev. 52 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 operation, 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 events 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. FSAR Rev. 64 Page 6 of 23
SSES-FSAR NIMS Rev. 52 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 operation of the facility. FSAR Rev. 64 Page 7 of 23
SSES-FSAR NIMS Rev. 52 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. FSAR Rev. 64 Page 8 of 23
SSES-FSAR NIMS Rev. 52 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. FSAR Rev. 64 Page 9 of 23
SSES-FSAR NIMS Rev. 52 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 individuals 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. FSAR Rev. 64 Page 10 of 23
SSES-FSAR NIMS Rev. 52 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 movement 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 components 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. FSAR Rev. 64 Page 11 of 23
SSES-FSAR NIMS Rev. 52 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. FSAR Rev. 64 Page 12 of 23
SSES-FSAR NIMS Rev. 52 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 systems 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. FSAR Rev. 64 Page 13 of 23
SSES-FSAR NIMS Rev. 52 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 produce 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. FSAR Rev. 64 Page 14 of 23
SSES-FSAR NIMS Rev. 52 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. FSAR Rev. 64 Page 15 of 23
SSES-FSAR NIMS Rev. 52 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 valves, 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 testing 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. FSAR Rev. 64 Page 16 of 23
SSES-FSAR NIMS Rev. 52 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% Uk/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% Uk/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.
FSAR Rev. 64 Page 17 of 23
SSES-FSAR NIMS Rev. 52 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 floors floor response spectrum) when the surface is subjected to a given earthquakes 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 BWRs Safety Design Basis The safety design basis for a safety system states in functional 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. FSAR Rev. 64 Page 18 of 23
SSES-FSAR NIMS Rev. 52 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. FSAR Rev. 64 Page 19 of 23
SSES-FSAR NIMS Rev. 52 Table 1.8-1 SSES PROJECT GLOSSARY TERMS TERM DEFINITION REFERENCE Seismic Category II Plant features not required to assure 1) the integrity of the reactor coolant pressure boundary, 2) (Non-Seismic Category I) 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 hours 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. FSAR Rev. 64 Page 20 of 23
SSES-FSAR NIMS Rev. 52 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 Spring loaded piston actuated check valves operate as follows for the following modes: Check Valve (SLPACK) 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. FSAR Rev. 64 Page 21 of 23
SSES-FSAR NIMS Rev. 52 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 LOCAs, 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 generated 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. FSAR Rev. 64 Page 22 of 23
SSES-FSAR NIMS Rev. 52 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. FSAR Rev. 64 Page 23 of 23
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SSP.5-FSAF Sheet 1 of 9
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A AMBRR~ AK~ETER, AKPERE AC ALTEPNA1ING CURPENT AC~ ~IR CIRCUIT BREAKER ~IR Cf')MD AIR CONDITIONING UNIT AL ~L~RM OP ALUMINU~ A"'P A"PLIFIER OP A~PERE AN1stY ANALYZER J.WNU~ ANNUHCIA1'0B AD AIR OP ER AT ED APPROX APPFOX UIA'T E AR ALHHI RELAY ARPII ARPl~TU RE ~FP ARRFSTER OR ARR~NGF ,ssr l\5SE~DLY A/T A~PF.RE TRANSOUC~R AUTO AUTOf'IATIC AUTO '!'FAirs AUTO TR;\NSFORMER ATIX AUJI LT ARY AVG AVERAGING OR AVERAGE AWG A~E~TCAN VIRE GAUGE A~R OARRIFR BAT BATTERY BCT BUSHING C0PRE~T !RANSPOR~ER 8D BOA~D BOD BACK DRAFT DA"PRR lHL BASIC I~PULSE INSULATION LEVEL BKJ? BR EA ff P. R BLDG BUILDING BL'l BLOWER 9LF BOILER B/Pf BTLL OF ~ATF.RIAL BOD BOTTOfll OF DTJC'T snp BOTTOl'I OF PIPE BRG BEARING BSTP BOOSTER BYP BYP1a.SS CAB CABINET CAP . CAPACITOR OR CAPACITY
SSES-l'SAF Sheet 2 of 9 CAS CASING CAUS CA0STIC CAV CAVITY CB CONTROL BOABD cc CONTROL CABINET OR COOLING COIL CR CONDUCTIVITY CELL CH CONTROL HOUSE (SWITCHJARDt CHE11 CHE"ICAL CRG CHANGE OR CHARGF.R, CHARGING CHLOP CHLORINA!OR 0~ CHLORINE CHV CHrLLED WATER Clt~C CIRCULATING CKT CI~CtJIT CL CP.N'!'ER L!N! CLR COOLER CND CllNDf.JIT CNDS COND F.N SAT R COL COLUMN COfllM CO~NUNICATlON OR co~~ON CO?!P CO~POTEF, COMPONENT CON C9NSOLE CONO CONOEHSER OR CONOOC108 CON1 CONTROL CONTC CONTR"CTOR CONTD CONTiliUED COOL COOLING CR CONTROL ROOPI cs CONTROL swr~cH OR CONT~OL STATION CTR CE'1TEP. CrJB CU BICL e cw COOLIN.G WATER, CIRCllLATING VA TER, OB CLOC'(lfl SE CENTF!F Cf!~TRIFflGAL CONTAIN CONTA!Nt1F.NT COl'I PF COMPRESSOR CCW CO~PONENT COOLING WATER CNCT CONC ENT RATE DB DRY BULB DC DIRECT CURRENT DEIHN DEPfINERAtIZEFl Rev. JS, 07/A4
SSES-PSAR Shf!e-t 3 of 9 DE'! DETEC,-OR DIAG DI AGRA Pl DU DI Al'l!TER DH'! DIFF~F ENTI AL DISC DISCONNECT DISCH DISCHARGE DTST DISTRIBUTION DP OIFFF.REN1IAL PRESSUF! OR DEV POIN! DF DRIVE DRN DRATM OWG D1UWING !/P F.LECTFO TO PNEO~ATIC TqANSDOC!R ~r. ELEVATION !I.EC ELECTRIC OR ELECTPIClL !"ER Ef1 F.'RGE NCY EP'IT ELECTRICAL "!TAlLIC TUBING ENG ENGINE EtfT *ENTRRING !OUIP 'F.QUI PPIEN'!' !XC EXCITER OR EXCIT~TION F.X H EXH UJST EXHTR EXffAUSTEFl !XT EXTRACTOR t:XT N !XT?ACTION !-!NCL ENC LOS OR!. P.SP ENGINEERED SAFETY FEATURES F PLOW FC FAIL CLOSED FO PI~F. DAKPER 'PDP F!F.DEF FTG FIGURE FIL FILTER FL FLU'{! P'L~J PLEIIBL~ FLD PIELO PO F1'IL OP~N P/T FREOUP.NCY TRANSDUCER Pi FE!DVATEF 35, 07/81.i
SSES-FSAR I!~LIL.1a.§:~ ShP.et 4 of 9 'PS FLOW SUTCH G G8EEN GEN GENERA TOR GOV GOYERNOF GR GEAR GRD GROOND c;s* GLAND SEAt. GSS GUND S~AL STEA."' GfJI GTJIDE H HYDROGEN H&V REATING ~ND VEMTILATION~DNIT HI HIGH HP HIGH PRESSORE OF HOPSEPOVRR OR HIGH ~OINT HSE HOUSE - RT HEAT TRAC I NG fl'!'R HEATF.9 KV HIGH VOt~AGE P.WC HOT WATER CIRCULATION HYO HYO RAU LIC HYDZ HYDRAZINE H't'G HEATHIG flVAC HE~TING, VENTILATING & AI~ CONDITIONfNr, IN INLF.T, INPUT, OR INTAKE !NC IMCN'IING IND INDICATOP INST TNSTFU ~ENT INV INVERTER 1P INTP.R"EDIATE PRFSSURE !~() ISOLATlON OR rsot~TEO TTL INT ~RLOCJ( JB JUNCTION BOX KCIHL THOUSAND CIRCULAR "ILLS KU. KILOVOLT-APIPERE KV KILOVOLT ~w KILOWATT Pev. 35, 07/Bfl
I' SSES-PSAR
!l~LJLl&§:~
She~t 5 of 9 L Lt=:PT OR LF.V!L L" LIGHTNING ARRESTER LAB t,Br:>RATORY LU LEVEL AURP!, LOi LC LO A D C ENT F. R LCS toc,L CONTROL STATION LJJII I,HIIT t.IO LIOUID LO LUBRJC*TION OIL, LOW, OR LOCKOUT LP LOW PBESSURE LS LF.VF.L SWITCH OR tI8IT SWITCH L'!'G LIGHTING LVG LEAVING t1 ~A!N P!A~H "ACRINE
~AN JIIANOAL P!ISC fUSCEL U NEOUS "CC ~OTOR CONTROL CENT!P.
tn.:CH "ECHANIC OR NECHANICAL "FR flf ANU FA C'r lJ R E OP. "AN U F ACT UR E P PfG O!l ~-G MOTOR- GENEPATOP PO.N Hnt~ tHSC Pl 0 f1ISCEl. UN EOUS
!"O'rOR OPFP ATED t'lflV ~OTOR OPFRATED VANE OR VALVE P! s flfATN STEA!'!
f!T ~AIN TRANSf0Bf1ER
~TR r'!O'rOR OR "!FTEF NC NOPf'IALLY CLOSED NEC NATIO~~L ELECTRICAL COPE NESF NON-ENGINEFREP SAFETY FEATURES NP.UT NEUTRAL NO wn~BER OR ROR"ALLt OPEN NOR~ NORl'\AL .. p Kfd'IP PLATE NF.G NEGATIVE NP5 NEGATIVE PRASE SEQUENCE
SSES-PSAP Shei'3it 6 of 9 0 OXYGEN oc OVFRCUBPENT f"CB OIL CIRCUIT SPEAKER OL OVERLOAD OUT OUTLET OR OOTPUT p PRESSO RF. PAL PRESSURE ALAR" LOW PR PULL BOX OR PUSHBUTTON PH PRASP. PMG PERNANENT "AGNET GFNEBATOR PNL PANF.L PO POSITION OB POSITtON!B POS POSITION SWITCH OR POSITIVE pp PUPIP PnESS PRESSURE PS PRES SU PE SWITCH PT POTENTIAL TPANSFOR~FP OR P~PSSOPE TRANS~ITT~R PWF POWER POT POTEN"'I AL PW PILOT VIRE R FIGHT og RF.D FAD PAflJ AT ION ~ P. ~OOF rxH,USTER OR ~~DIATION ELE~ENT PF.l\CT PF. ACTOR REC FF.CE IV ER P:i:'CTPC PFCIRCOLATION OR PECIRCllt~T!~G FECP P.ECEPTACLE PF.CTIPI!P . Ff'.G RF.GJJLATOF P. F.F REFERENCP. Pfi REHEAT OR RELATIVE HOftIDITY 'RHEO RHf.OSTAT RHTF REH EAT E~ RI.Y RF.LAY IHI ROOK RR RAILBOAD FS REVERSING STA?TRR
SSP.S-FSAR Sh~et 7 of 9 RTD PBSISTAHC! T!"P!RlTORE DETECTOR "ECJ P RECIPROCATOP. OR RECIPROCATING
'P.HR RESIDUAL BF.AT RE"OVAL SAl'f P SA 1'PLT NG sc~v SCAV!NGIMG SCH F°,. SCREPIATIC SCRN SCREEN ~D .. STP.AN LOAD DRAIN OR DO~P VALVE SEC SECORD UY SECT ~ECTIOM SEL SELECTOR SFi~ ~EPP.RATOR SERV SERVICE SEW SF.WAGE SFT SHAFT SH SOFERH ! AT !R SHLD SHIELD S! SFAL TN R!LAY OR CONTACT SIG SIGNn SL . SEAL so SEAL OIL SP S PEEO OR SPARE S'rP! STEAPI STA STATION STFY ~TANDDY STD ~TA'fDARD STG STORAGE STR ST'-RTRF STfJP STA RTTJP SOBS TA SUB~TATION ~UCT SUCTION SUPHTR SCJPERH !AT EP St'JPT SUPPORT sv SOLP.NOID VAUR SIi 5WITCR OR SE~VICR IAT!R SWBl' SWITCH BOARD SWGP SWITCKGE~P .
SYM SYNCRRONOUS, SYCRRORIZI~G OR SYNCR~ONISft SWYD SVI~CHYABD (ie. SIITCHIMG STATION)
,~v . ~s. 01/84
.I
ss~s-rsu Sheet 8 of 9 SYS SYSTEM ss STA RTO P SOURCES SHED SHF.r.DI WG T TE~PEPATURE. TRANSPUC!R OR TRANSMITTER TAL TEMPE~kTUR! lLA~M TACH TACHOMETER TO TERl1INAL BOX TC THER"OCOOPLE OR TPIP COIL TD THIE DELAY TDC '1'.'IP,E DELAY (THIE DELAY TO CLOSE) TDD TIME DELAY (ON DEEHERGIZIMG) TDF. TI"E DELAY (ON EMERGIZIMG) TDO TI"E DELAY (TiftE DELAY TO OPEN) TE THFRPIAL F-LEPIENT TfL TELEPHONE TEf\ P TE~PEPATOP.E OR TE"PORARY T'ER"' TERIHN'AL TRE £PIO TRER~OSTAT THROT Tff ROTTL! TG OP T-G TORNING GEAF OR TUPBIME-GENERATOR 'f'K TANI!: TOS TOFOUE SWITCH TRAV ~IUV!LING 1'RANS TR~NSPEF OR TRANSPOR"ER TS TF.~PEP.ATUP.E SWITCH OR TEST SWITCH TIJRB TURBINE TfP TYPICAL TV TELEVISION (RECEIVER) TVC TELEVISION CA!U~RA (TRARS~ITTER)_ UP UPSTREA~ OR UPPER UV O'HDERVOLTAGE Uf' UN DFi Rf RR OU ! RCY V VOLT, VOLTAGE, VOLTIIIETER OR YALV! ,~c VACUUPI YAP *VAPOR 98 VIBROP! ETER V!NT V!MTILATIOM R.ev. 35, 07 /84
I SSES-FSAR SbP.et 9 of 9
'f!RT VERTICAL V/T YO LT TRAMS DOC!R II WEST, WHITE, OR VATT!!T!R WB VET BULB WHSfi! VAREROlJS!
VP WEATHRRPROOP 1/T VATT TRAMS DOC!~ WTR WATER
SS!S-FSU Sheet 1 of 2 Pipe and valve classes are desiqnated by a tbree-lettet code. - The first letter in~icates the pri ar, *al,e and flange pressure ratinq; th~ second letter, the type of aterial; and the third letter. the coae to which the pipinq is designed.
!!B~l-L!1.l~E_:_fBift!BJ_flB~~YBJ_JA%1Jg 1,
Unless othetvlse noted, all ratings are in accordance vith 11s1 816. 5 (NOTE A - ~pecific pressure a specific te peratnte B - 25001 C - 1500t D - 9001 E - 6001 F - 4 OOt G - 3001 ff - 1501 J - 1251 ANSI B16.1
- K- 1751 VOG Underwriters' Laboratories~ Inc.
L - 2S01 ANSI 816.1 X - Gravit? ratinq Y - General use as desiqnated on p!pinq class sheets. Ai1dit._ional pressure ratinqs for valves: M - 200t ( anufacturer*s ratinq) R - 1501 VOG P - 1001 (manufacturer's rating} R - 751 (manufacturer's ratinq) S - SOt WOG T - 2 St Aw wA (or *an u fa ct ur er
- s r a~ 1n g) z - Gen~ral use as desiqnated on p~pinq class sheets.
A - Alloy Steel B - C11rbon Steel c Austenitic Stainless Steel D - CoppP.r, Brass or Bronze ! - Copp~r-Mick4l P - carbon Steel - Copper Bearinq G - carbon steel - Linea R - cast Iron Rev. 35, 07/84
SSES-PSAR Sheet 2 of 2 J - Con*crete J - Vitrified Clay t - carbon Ste~l - Impact Tested ~ - cast Iron - Hiqh Silicon N - Carbon StP.P.1
- Galvanized P - Cast Iron - Ceaent Lined Q - Asbestos-Ce ent R - carbon steel-(RRR, Ser*ice later g !S Water svsteas)
A .. Nuclear Power Plant Co ponents, AS!'IE B&P, Code,, Sect. III. Class 1 8 - Nuclear Pov~r Plant components, ISM! B&PV Code, Sect. IIt, Class 2 C - Nuclear Pov~r Plant Co* pon en ts, ASME B&PY Code, Sect. III, Class 3 D - Pover Piplnq co~e, l MSl B31. 1. 0 F
- National Fire Protection Association coae G - Uniform Pluebinq Code H .. Pover Boilers, ASPIE B&PY Code,, Sect. I ~
J - A*erican water Works Stanaards Pev. 35, 07/8Q
INSTRUMEN T CONTROLLING (S ECON D AND THIRD LETTER) FUNCT ION MISCEL LANEOUS ABBREVIATIONS MISC ELL ANEOU S ABBREVIATIONS:
~
;1*, ADS AIS AUT0~1ATIC CEPRESSUR ZATION AIR SUPPLY ACID W A.STE oCORR CSIV:::,CAU STIC)
REACTOR BUIL Dlr-. C CLOSED CO ::JLIN G WATER REACTOR BU ILDING EQUIPMENT DRAIN TANK III ~ CONDUCTIVITY IND:CATOR OR TRANSMITTER RMC
*~
REMOTE MANUAL CONTROL
~~ CR OHS CONTROL ROil DR IVE HY[')RA ULIC SYSTT"M RMS RE~ OH" '-' Ar<UAL E;YE:TE'A REAC IUI( PFiOTEC I ION SYSTE M MEASURED VARIABLE
_ _\~ETTER) _ 0 " I~ !~ I CRS IC DRW C::JNUY.:fl VITl' RECOl <DIN GSWIICI I CYCLE T MER DIRTY RADWAST:0 REACTOR PR ESSURE VESSEL SELECTIV:C: SWITCH AUTO MATIC
,,,------------- --- DI FFERENTI AL TE MPERATUR E SWITCH OR .,r SQU,'.RE RO'.)T co,*wER-OR
~::=---- 1
- - CONV :CRTER ('CLTAGE / PNEUMATIC) TBCCW TU RBINE BU ILD NG CLOSED COOLING WATER CON DIJCTIV TY =1S SPECIAL EL:CCTRIC P:JWER SUPPLY 'l:EQU RED CID TER S TIJ RBI OTY RECOR DC:R SWITCH I FILTER DEMlt<. ERAL E R TIME LlLLA\' SWITC I I l'lDICATES CLOSES ON All / QR ELECl l{ l.;;/\,L FAILURL TORC UE OVERLOAD SWITCH
~LCW INDICAIOR TR/\NSM II IE R TOR C::IIF RECC RDF"R S'MTCH INDICATE S OPENS ON AIR OR ELECTRICAL FAILURE
-- -*- ----*--*- FLOW RECORDlr,G CONT'/OLLE R SWITCH TOT TO~OUE TR/\NSll/'ITTER I
HYDROGEN TRS - EMPERATL. RE RECORDER SWITC"1 HYDRAULIC CONTROL UNIT HYDROGEN ION CCNC CONV ERTER iCURRENT/PNEUM/\TIC)
.-, s I
--- ~---~1111----
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. L~ EL & DENSITY RECORDER SYSTEM NS, M-'-"'O SW :1(H SH ~or e il Mo,srnee OXY3E\J_ _ _ _ _ _ _
\J "C
02 URS LEVEL INDICATOR RECORDING SYSTEM PR ESSUR E _ __ P PRC _Q I LEVEL RECORDING SWITCH POSITION Po MA IN STEA'A ISO LATION VALV E
- - -- --- - ------- - ,- - - MILLIVOLT TO CURRENT SYSTEM f"
I
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~
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M/\ NUAL TR IP & RESET I~ PROC ESS Fl 0 \1'/ S MEAS,JRING POSl70N s*-~l - A.*11 -l r r r T 1__ 1 L PR ESS-TLMP INDEX
~
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---0--- SAlJl-j CE RS TYPE (:NE R) 8 --cl<i-- accrn~"'"' ~~ + CRW OPEN CRAIN FUNNE'- DRAIN SYSTE~ FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION
)- ->I<... FURNISHED WITH UNITS 1 & 2 a
'I" ASSOC AE EQUIPMFNT
~
--,,--, FINAL SAFETY ANALYSIS REPORT
--,f-- DIAP clRA~1 0P ERATEC CONTROL AIR OPERATEDWl7-l 50!..'c.NOln
'"' °' ,CM Off ~ " '"
--1'-'\-- OPENS ON FAILURE """ I _J_ PIPING SPEC CHANGF El I
~
(N OTTO RE US CCO ON FQ.J IP ~ :C:~T) MAIN CONHOL ROOM AIR OP ERATED WITH SOLENO IC RMS ON CONTROL ROCM PANEL
--ii>-- TE'APORARY STRAINER (S EE NOTE 2) PIPING AND INSTRUMENT SYMBOLS I
I FIGURE 1.8-1, Rev 54 I AutoCAD: Figure Fsar 1_8_1.dwg I I
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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£LCCT OUTPUTS. THIS O(VIC.C IS NORMALLY. VALVC nus I S ALSO o/~~' ~CO(i BfffifsBfQ~~RM1ss1vc r~c.T tCN 1oiMICH Jo<<JST BE su1s.-1to TO P(RMll lH( SIGNAL fLOw TO PAS.5 TO THC "'().T BLOCK . TH1S !!LOCK usco roR OH-i(A INPUT / OUTPUT PO-,.,:R SOUJK[S SUCH AS *111 OR l't'IOIUULI C,
- SOLENOID PILOT VALVC fOA .t,H AIR OPCR.ATCO VAL.'IC IS I.N CXAMPLC or THIS TTPC OEVICC, LINES ASSOCIATED WITH LOGIC SYMBOLS
~:~1~rs1~Rlo(f~~~~~CM;t.~~Cl:~XILIARY SIGNALS. THC OUTPUT ELECT . FLOW SIGNAL AUXILIARY SIGNAL SOURCE SUCH AS AIR FSAR REV.65 LEGEND OR HYDRAULIC SUSQUEHANNA STEAM ELECTRIC STATION
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M ECHANICAL LINKAGE START UNITS 1 & 2 START OF PRIMARY INITIATING SIGNAL FINAL SAFETY ANALYSIS REPORT
~ ETIER DESIGNATION SH. NO ZCfE OR REF . DV-/G LOGIC SYMBOLS MATCH CIRCLE- LE TT ER DESIGNATION ON DWG MUST MATCH LETTER DESIGNATION ON THE INTERFACING DWG FIGURE 1.8-3, Rev 54 AutoCAD: Figure Fsar 1_8-3.dwg
~ LARGE CASE ROUND CHART RECORDER I
BASIC CONTROLLER 1: 2-INPUT PROPORTIONAL AMPLIFIER (RACK-MOUNTED) 2 CONTAINS 2 OR 3-MODE CONTROL. SET POINT, MANUAL- ll'JPlfTS CONTAINS TWO INPUT CIRCUITS, INPUT BIAS ADJUSTMENT, ~ DIRECTLY OPERATED AUTOMATIC SELECTION, REVERSE OR DIRECT ACTING, OUTPUT BIAS ADJUSTMENT AND GAIN ADJUSTMENT r ,::<<c, GAIN CASCADE SWITCH, AND VALVE POSITION INDICATOR PNEUMATIC POSITION ER AND VALVE fTu-'3 PIPE MOUNTED
,NrnCATONC,CONH<OCUNG eYSOMme F
BASIC M/A CONTROLLER
- E 5-INPUT SUMMATION AMPLIFIER (RACK-MOUNTED) MILLIVOLTMETER TYPE CURRENT OUTPUT, CONTAINS 2 OR 3-MODE CONTROL, MANUAL AUTOMATIC C POSITION MODULATION OR TIME DURATlml SELECTION, REVERSE OR DIRECT ACTING, AND INDICATORS l'Jf--JUTS SAME AS 2 INPUT SUMMATION AMPLIFIER BUT PNEUMATICALLY OPERATED CONTROL VALVE CONTAINS 5 INPUT CIRCUITS CONTROL FORMS.
FOR PRO ESS VALUE AND VALVE POSITION. THE CONTROLLER Gh N MAY OR MAY NOT HAVE INTEGRAL I/P
+
SET POINT MUST COME FROM SOME EXTERNAL DEVICE, SYMBOL BIAS TRANSDUCER WITH ACTUATOR WILL ALSO BE USED TO INDICATE ONLY A MANUAL TO AUTO* s MATIC TRANSFER STATION.
~
ELECTRO-MECHA~JICAL cmJTROLLER f t ~ 8-INPUT SUMMATION AMPLIFIER (RACK-MOUNTED) RATIO SET STATION CONTAINS A 3-MODE CONTROLLER. EITHER CONTAINS A RATIO ADJUSTMENT KNOB 1.3 TO 3.01, INPUT "'P Ts SAME AS 2 INPUT SUMMA.TIDrJ AMPLIFIER BUT TIME, POSITION, OR CURREt.JT MODULATION. PNEUMATICALLY ACTUATED CONTROL VALVE l AND OUTPUT INDICATORS, AND THE RATIO AMPLIFIER. c;,N CONTAINS 8 INPUT CIRCUITS SET POINT COMES FROM A CONTROL SLIDE-MAY OR MAY NOT HAVE INTEGRAL I/P BIAS WIRE MOUNTED IN A SERVO-RECORDER. TRANSDUCER-WITHOUT ACTUATOR lm BIAS-MANUAL AUTOMATIC STATION CONTAINS A BIAS ADJUSTMENT KNOB -20 TO +20%, INPUT AND OUTPUT INDICATIORS, BIAS AMPLIFIER, AND MANUAL AUTOMATIC SELECTION. B SQUARE ROOT EXTRACTOR (PANEL OR RACK MOUNTED, CONTAINS CIRCUITRY TO EXTRACT THE SQUARE ROOT OF THE INPUT SIGNAL AND TRANSMIT A LINEAR SIGNAL. ITI RECORDER CONTROLLER IN COMMON CASE TWO OR THREE MODE CONTROLLER, SINGLE PEN RECORDER.
~
3-\/\IAY SOLENOID OPERATED VALVE
. ~
BASIC CASCADE COMBINATION FUNCTION GENERATOR (RACK MOUNTED) MINIATURE STRIP-CHART RECORDER lfflf1 f G t
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UTILIZES THE STANDARD CONTROLLER AS THE PRIMARY ~ CONTAINS CIRCUITRY TO PROVIDE AN OUTPUT AS A SINGLE PEN POTEMTIOMETRIC TYPE, 4" CHART EDGEWISE INDICATOR rm, AND A CONTROLLER WITHOUT INTEGRAL SET POINT ON THE SECONDARY. MUST BE OPERATED IN CASCADE OR IN MANUAL "
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i 0 VARIABLE FUNCTION OF THE INPUT SIGNAL. 1/2% ACCURACY. CASCADE (2 STANDARD CONTROLLERS) s SIGNAL SELECTOR (RACK MOUNTED) MINIATURE STRIP-CHART RECORDER 8 F TEMPERATURE SENSOR i ANY TWO STANDARD CONTROLLERS MAY BE OPERATED l~J L CONTAINS CIRCUITRY TO SELECT THE HIGHER OR LOWER [ ] TWO- PB*I POTENTIOMETRIC TYPE 4" CHART m u E CASCADE. THIS COMBINATION MAY BE OPERATED OUT N c OF A GROUP OF INPUT SIGNALS 1/2% ACCURACY. THERMOCOUPLE OR RESISTANCE C T TEMPERATURE DETECTOR OF CASCADE WITH THE "SECONDARY" CONTROLLER ON L O AUTOMATIC. MANUAL LOADING STATION
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MUL Tl PLIER-DIVIDER (RACK MOUtffED) COGCWOCC "'O~AWCS, GANG MO"'ff0 [:_] f CONTAINS A MANUAL KNOB, POWER SUPPLY, AND VALVE POSITION LIQUID-LEVEL TRANSMITTER GENERALLY USED FOR COMPENSATION OF FLOW SIGNALS INDICATOR SIMPLY PROVIDES 10-50mA INTO A 600 OHM LOAD POSITIVE DISPLACEMENT TYPE FOR REMOTE POSITIONING OF VALVES AND DAMPERS. MAY BE USED FOR A SET POINT STATION \/\/HEN POSITION IS l~OT CRITICAL. [jJ SET POINT STATION
~ ~
MILLVOLT CONVERTER LARGE-CASE, STRIP-CHART RECORDER CONTAINS THE PRECISE SET POINT UNIT ONLY. TO BE USED WHEM SET POINT MUST BE REMOTE FROM THE CONTROLLER. CONVERTS MILLIVOLT INPUT SIGNALS TO ELECTRONIC OUTPUT SIGNALS 10-50 mA. POTEMTIOMETRIC. 12" CHART, 1/4% OF FULL SCALE ACCURACY. [QJ ORIFICE PLATE INSTALLI\TION PLATE AND FLI\NGES INDUSTRIAL MASS FLOWMETER HIGH - LO\/\/ LIMIT STATION 5-UNIT POWER SUPPLY ~ WITH SELF-CONTAINED INTEGRATOR, DIGITAL READOUT
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CONTAINS ELECTRONIC CIRCUITRY TO LIMIT CONTROL SIGl~ALS LARGE-CASE. ROUIJD-CHART RECORDER TO A PRESET VALUE
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POTEIJTIOMETRIC 12" CHART, 1/4% ACCURACY ll*IDICATOR ELECTRICALLY OPERATED 250 SCALE, PANEL MOUNTING
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INTEGRATOR DIFFERENTIAL PRESSURE TRANSMITTER
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PANEL MOUNTED INTEGRATOR FOR TOTALIZING FLOW CONVERTS DIFFERENTIAL PRESSURE TO AN ELECTRONIC MANUALLY OPERATED ROTARY SWITCH SIGNALS. 6 DIGIT COUNTER SIGNAL. 10-50 rnA. [J
. (J) WITH PISTOL GRIP HANDLE.
RACK MOUNTED 3-MODE CONTROLLER S PROCESS PRESSURE TRAl'JSMITTER CONVERTS PROCESS PRESSURE TO AN ELECTRONIC INDICATING GAGE. DIRECTLY OPERATED p PROPORTIONAL CONTROLLER (]) 1*1101 SIGNAL. 10-50 mA. PANEL MOUNTING P+R PROPORTINAL PLUS RESET CONTROLLER PEScT PROPORTIONAL RESET PLUS RATE P+R+R PROPORTIONAL PLUS RESET PLUS RATE CONTROLLER RAT[ M/A MAN UAL-AUTO MATIC SELECTION r::7 Sll*JGLE UNIT POWER SUPPLY S. P. SET POINT
~ FSAR REV.65
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RACK-MOUNTED 2-MODE CONTROLLER PNEUMATICOPERATOR rROI' PROPORTIONAL PLUS RESET ADJUSTABLE HIGH OR LOW ALARM UNIT WITH COVER AND CURRENT-TO-PrJEUMATIC RESET B
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- TRANSDUCER SUSQUEHANNA STEAM ELECTRIC STATION ii UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT
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ELECTR-PfJEUMATIC CONVERTER PNEUMATIC COVER E] RACK-MOUNTED RATE ACTION DEVICE WITHOUT COVER BUT WITH CURRENT-TO-CONTAINS ADJUSTABLE RATE ACTION UtJIT PNEUMATIC TRANSDUCER RACK-MOUNTED SET POINT r7rl INSTRUMENT SYMBOLS
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THIS IS A PRECISE SET POlfJT UNIT FOR RACK MANUALLY OPERATED ROTARY SWITCH MOUNTING ONLY. ~ ~ HEAVY-DUTY PNEUMATIC OPERATOR [Q] WITH ROUND KNOB OPERATOR FIGURE 1.8-4, Rev 54 AutoCAD: Figure Fsar 1_8-4.dwg}}