Glossary of Terms and Abbreviations used at Diablo Canyon Power Plant (DCPP)
- Aging Management
- is a program for monitoring and dispositioning materials and components whose characteristics change with time or use. PG&E defines aging management as “Engineering, operations, and maintenance activities to control age-related degradation and to mitigate failures of systems, structures, or components (SSC) that are due to aging mechanisms.
- As Low As reasonably Achievable (ALARA)
- refers to maintaining offsite radioactive releases and occupational radiation exposures as low as achievable in a reasonable, cost-effective manner.
- Capacity Factor
- is the fraction of power actually produced compared to the maximum which could be produced by operating at full power during a period of time (expressed in percent).
- Civil Penalty
- is a penalty in the form of a monetary fine levied by the Nuclear Regulatory Commission for a significant violation of its regulations.
- Control Rods
- are long slender metal-clad rods which move into or out-of nuclear fuel assemblies in the reactor core to control the rate of the nuclear fission process. The rods contain a neutron absorbing material which, when inserted into the fuel, absorb neutrons, slowing down the fission rate and thus the heat generation rate and reducing the power level of the reactor.
- Design Bases
- are the current features and criteria upon which the nuclear plant is designed and are also the bases for Nuclear Regulatory Commission review and approval.
- Diesel Generator (DG)
- is a standby source of emergency electrical power needed to power pumps and valves to provide cooling water to the fuel in the reactor to prevent its overheating and possible melting. The diesel generator is designed to start up and provide power automatically if normal power is lost.
- Emergency Operations Center (EOC)
- is the facility away from the immediate vicinity of the plant which is used to direct the operations for mitigation of and recovery from an accident.
- Emergency Preparedness (EP)
- is the assurance that the plant and its personnel are practiced and prepared for postulated emergencies to be able to mitigate them and recover with a minimum of damage and health effects.
- Engineered Safety Features (ESF)
- are the features (systems and equipment) engineered into the plant to mitigate the effects of anticipated and postulated accidents.
- Erosion/Corrosion
- is a phenomenon which takes place in carbon steel power plant water systems. The inside metal pipe will continually corrode due to galvanic action, forming a magnetite coating as erosion (due to high water velocity and/or changes in flow direction) continually wears away the magnetite layer, permitting the corrosion layer to reform, etc. The continual combination of effects wears away and thins the pipe wall.
- Escalated Enforcement Action
- is action taken by NRC beyond a notice of violation of its requirements for a single severe violation or recurring violations. Examples include a civil penalty, suspension of operations, and modification or revocation of a license to operate a nuclear plant.
- Final Safety Analysis Report (FSAR)
- is the document which describes the plant design, safety analysis, and operations for Nuclear Regulatory Commission review and approval for licensing for plant operation.
- Fitness for Duty (FFD)
- describes the state of an employee (cleared to access the nuclear plant) being in sound enough physical and mental condition to adequately and safely carry out his or her duties without adverse effects.
- High Impact Team (HIT)
- is a term denoting a multi-disciplinary or multi-functional team of people put together to focus on solving a particular problem or perform a particular task. The disciplines included are those necessary to effectively accomplish the task.
- High Level Waste (HLW)
- is highly radioactive waste, usually in the form of spent fuel (or fuel which has been discharged from the reactor as waste) containing a high level (as defined by NRC regulations) of radioactive fission products. HLW is handled remotely, using water or a thick container as a radiation shield.
- Individual Plant Examination (IPE)
- is a level 2 Probabilistic Risk Assessment (PRA) analysis of plant accident sequences. The analysis includes core damage progression through the release of radioactive material to the containment and the subsequent containment failure but stops short of determining potential impact on the public or property. The NRC requested all nuclear plants be analyzed in this way to get a better understanding of severe accident behavior. An IPEEEis an IPE which is initiated by External Events to the plant.
- INPO
- the Institute of Nuclear Power Operators is a nuclear industry group formed after the Three Mile Island accident to help improve nuclear plant operations through regular assessments of each nuclear plant, evaluations, best practices, and nuclear operator training accreditation.
- ISFSI
- or Independent Spent Fuel Storage Installation, is the term for DCPP’s on-site storage facility for the dry cask storage of spent nuclear fuel.
- Inservice Inspection (ISI) and Inservice Testing (IST)
- are the practices of inspecting and testing certain selected components periodically during their service lives to determine degradation patterns and to repair, if necessary, any degradation beyond acceptable limits.
- Licensee Event Reports (LERs)
- are reports from the plant operator to the Nuclear Regulatory Commission describing off-normal events or conditions outside established limits at a nuclear plant.
- Loss of Offsite Power (LOOP)
- is an occurrence whereby the normal supply of electrical power from offsite is interrupted. Nuclear reactors need power from offsite when shutdown for spent fuel cooling and residual heat removal. There are usually several sources of offsite power; however, loss of all sources would result in the automatic start-up of the diesel generators to supply power.
- Low Level Waste (LLW)
- is waste containing a low level of radioactivity as defined by NRC regulations. LLW is usually in the form of scrap paper, plastic, tape, tubing, filters, scrap parts, dewatered resins, etc. LLW requires packaging to prevent the spread of contamination but little radiation shielding.
- Maintenance Rule
- is the NRC proposed rule which requires that nuclear power plant licensees monitor the performance or condition, or provide effective preventative maintenance of certain structures, systems and components against licensee-established goals. The Rule becomes effective July 10, 1996.
- Microbiologically-Influenced (or Induced) Corrosion (MIC)
- is corrosion, usually in the form of pitting, on steel piping systems containing stagnant or low-flow water conditions. The corrosion is caused by surface-attached microbe-produced chemicals which attack the piping surface. Depending on severity, MIC is controlled by mechanical and chemical cleaning combined with biocides.
- Mid-Loop Operation
- is an infrequently-used Refueling Outage procedure in which, after shutdown and a cooling period, reactor coolant is lowered below the hot and cold legs, permitting work to be performed in a relatively dry environment. The operation is a relatively high-risk condition due to the potential for loss of cooling.
- Motor-Operated Valves
- are valves opened or closed by remotely-or locally-operated integral electric motors. The valves are used in power plant piping systems to divert, block or control the flow of steam or water.
- Nuclear Excellence Team (NET)
- is a organization of several well-qualified senior people whose mission is “To improve plant performance through the use of performance-based self-assessments within the NPG (Nuclear Power Generation) organization.” The Team is augmented by at least one other PG&E and one outside individual with expertise appropriate to the particular investigation.
- Nuclear Regulatory Commission (NRC)
- is the Federal agency which regulates and licenses the peaceful uses of domestic nuclear and radioactive applications such as nuclear power plants, experimental nuclear reactors, medical and industrial radioisotope applications, radioactive waste, etc.
- Nuclear Steam Supply System (NSSS)
- is the nuclear reactor and its closely associated heat removal systems which produce steam for the turbine. The NSSS usually includes the nuclear reactor, nuclear fuel, reactor coolant pumps, pressurizer, steam generators, and connected piping.
- Operational Capacity Factor
- is the capacity factor as measured between, but not including, refueling outages.
- Probabilistic Risk Assessment (PRA)
- is a formal process for quantifying the frequencies and consequences of accidents to predict public health risk.
- Protected Area
- is the outermost area of the nuclear plant which is protected by physical means, a security system, and security force to prevent unauthorized entry (see also Vital Area).
- Quality Assurance (QA)
- comprises all those planned and systematic actions necessary to provide confidence that a structure, system or component will perform satisfactorily in service.
- Reactor Coolant System (RCS)
- is the collection of piping, reactor vessel, steam generators, pumps, pressurizer, and associated valves which function to circulate water through the reactor to remove heat.
- Refueling Outage
- is a normal shutdown of a nuclear power unit to permit refueling of the reactor, along with maintenance, inspections and modifications. Typical DCPP refueling outages occur about every 18 months and last for about two months. The outages are numbered by unit number (1 or 2), “R”, and the consecutive outage number. For example, “1R5” is the fifth Refueling Outage for Unit 1 since start-up.
- Reliability Centered Maintenance (RCM)
- is the practice of maintaining equipment on the basis of the logical application of reliability data and expert knowledge of the equipment, i.e., a systems approach. Normal preventive maintenance (PM) is performed on the basis of time, i.e., maintenance operations are performed on a schedule to prevent poor performance or failure.
- Residual Heat Removal (RHR)
- is the removal of the residual heat generated in the reactor fuel after reactor shutdown to prevent the fuel overheating and possibly melting. The heat removal is performed by a set of pumps, piping, valves and heat exchange equipment circulating water by the fuel while the reactor is shut down.
- Safety System Functional Audit and Review (SSFAR)
- is an investigation of a single plant safety system from all perspectives such as design basis, operations, maintenance, engineering, testing, materials, problems and resolutions, quality control, etc. The review is performed by a multi-functional team and can last several months.
- Simulator
- is a simulated nuclear power reactor control room with gauges, instruments and controls connected to a computer. The computer is programmed to behave like a nuclear reactor and respond to operator actions and commands. The simulator is used in training nuclear operators in controlling the reactor and responding to simulated transients and accidents.
- Spent Fuel Pool (SFP)
- is an in-plant stainless-steel-lined concrete pool of water into which highly radioactive spent nuclear fuel is stored when it has been discharged from the reactor. The spent fuel is maintained in the pool until its ultimate disposal is determined.
- Steam Dump Valve
- is a device to discharge (dump) steam from the power plant piping to lower its pressure and reduce the energy in the line. This is done to permit faster shutdowns.
- Steam Generator
- is a large, vertical, inverted-U-tube-and-shell heat exchanger with hot reactor coolant on its tube side transferring heat to and boiling the non-nuclear feedwater to form steam on the shell side. Besides transferring heat, the steam generator is important as a barrier between the nuclear and non-nuclear coolants.
- Surveillance
- is the process of testing, inspecting, or calibrating components and systems to assure that the necessary quality is maintained, operation is within safety limits, and operation will be maintained within limiting conditions.
- Technical Specifications (TS)
- are the rules and limitations by which the plant is operated. They consist of safety limits, limiting safety system and control settings, limiting conditions for operation, surveillance requirements, description of important design features, administrative controls, and required periodic and special notifications and reports.
- Technical Support Center (TSC)
- is the in-plant facility which directs plant activities in mitigating accidents and minimizing their effects.
- Trip (or scram)
- is the shutting down of the nuclear reactor by inserting control rods which shut down the nuclear fission process. An automatic trip is initiated by plant monitoring systems when one or more parameters differ from preset limits. A manual trip is initiated by plant operators in an off-normal event to prevent preset limits from being exceeded or as a backup to the automatic system.
- Vital Area
- is an area inside the plant within the Protected Area which contains equipment vital for safe operation.