Glossary of Terms and Definitions Used by the DCISC

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.
As used in “main bank transformer” or “main transformer bank” references refers to a set of installed electric transformers.
is the act of reviewing and evaluating practices at other nuclear plants, which are known for excellence in a specific area, for incorporation or improvement at one’s plant
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.
Cross-cutting Aspect
A nuclear plant activity that affects most or all of NRC’s safety cornerstones, which include the plant’s corrective action program, human performance, and “safety-conscious work environment." A Substantive Cross-cutting Issue refers to a performance deficiency characteristic that compromises more areas than just the specific situation in which it occurred.
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.
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 IPEEE is 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.
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.
– with reference to the Hot Leg or Cold Leg refers to piping trains leading to or from the reactor vessel. The Hot Leg removes heat and the Cold Leg provides cooling water to the vessel and nuclear core.
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.
Line Organization refers to the direct reporting supervisory chain in an organization through which orders and information flow. It is also known as the “chain of command.”
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.
means a positionable component, such as a valve, placed or left out of the required position for existing plant conditions when the component’s required position is tracked by a station status control tool, such as a procedure, drawing, or valve list.
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.
formerly known as an “Action Request” or “AR” is a document, which is used to identify and track resolution of a problem and incorporate it into the Corrective Action Program.
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.
Primary Side and Secondary Side
refer, respectively, to the Reactor Coolant System, which is used to remove heat from the nuclear reactor and the Main Steam and Feedwater Systems which provide cooling to the Steam Generators and generate and provide steam to the Turbines.
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 is 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.
Reactor Oversight Process
is the process by which the NRC monitors and evaluates the performance of commercial nuclear power plants. Designed to focus on those plant activities that are most important to safety, the process uses inspection findings and performance indicators to assess each plant’s safety performance.
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.
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.
Single Point Vulnerability (SPV)
is an individual component, which does not have a significant level of component redundancy and whose failure alone could adversely impact the system or plant performance. DCPP defines a SPV as “a High-Critical component whose failure results in a plant trip or derate > 2%.
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.
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.
refers to individual functional lines of system piping, components, or wiring which are usually independent of other parallel lines, which have the same redundant function.
(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.

For more information about DCISC contact:

Diablo Canyon Independent Safety Committee
Office of the Legal Counsel
857 Cass Street, Suite D, Monterey, California 93940
Telephone: in Califonia call 800-439-4688; outside of California call 831-647-1044
Send E-mail to: