Beyond Design Basis Accident (BDBA)

A range of possible accident sequences that are possible but are not fully considered in the design process because they were judged to be too improbable. As the regulatory process strives to be as thorough as possible, "beyond design-basis" accident sequences are analyzed to fully understand the performance of a design.

Boiling Water Reactor (BWR)

A nuclear power plant characterized by the production of steam, which is used to drive the turbines, in the upper part of the reactor core.

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Combined License (COL)

A combined construction permit and operating license with conditions for a nuclear power facility, issued under 10 CFR Part 52. See 10 CFR 52.1(a).

Containment Heat Removal System (CHRS)

A system consisting of reactor vent valves, reactor recirculation valves, the reactor pressure vessel, the containment vessel, and the reactor pool, which provides a means of removing core decay heat by venting steam produced in the core to containment via the reactor vent valves, condensing the steam on the inside surface of the containment vessel via the reactor pool outside of containment and returning the condensate to the core via the reactor recirculation valves.

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Decay Heat Removal System (DHRS)

A system consisting of inlet components that provide water from the reactor pool to the inlet of the steam generators, and outlet components that provide a vent path from the outlet of the steam generators to a sparger located in the reactor pool. These components allow circulation of containment cooling pool water and provide a means of removing core decay heat via the steam generators.

Dose (Radiation Dose)

A generic term that means absorbed dose, dose equivalent, effective dose equivalent, committed dose equivalent, committed effective dose equivalent, or total effective dose equivalent.

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Emergency Core Cooling System (ECCS)

A component of the reactor system designed to mitigate the effects of a loss of coolant accident (LOCA).

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General Design Criteria (GDC)

The minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have been issued by the NRC.

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Licensing Topical Report (LTR)

A licensing document that addresses a technical topic related to nuclear power plant safety submitted for review and approval by the NRC. The NRC reviews it and issues a Safety Evaluation Report (SER).

Light-Water Reactor (LWR)

A type of thermal reactor that uses a light-water moderator to thermalize neutrons.

Loss-Of-Coolant Accident (LOCA)

A hypothetical accident resulting from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant makeup system.

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Multi-Application Small Light Water Reactor (MASLWR)

A conceptually designed small, light water, natural circulation integral pressurized water reactor.

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United States Nuclear Regulatory Commission

Nuclear Steam Supply System (NSSS)

The NSSS is the set of components in a nuclear power plant which produce steam from the core energy production.

NuScale Power Module™

Each NuScale Power Module generates 50 MWe (gross) and has its own turbine generator, condenser, and feed water system and operates independent of the other units.

NuScale Power Plant

An electrical or steam power generating station that uses one or more NuScale Power Modules as the energy source.

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Pressurized-Water Reactor (PWR)

A nuclear power plant characterized by production of the steam used to drive the turbines occurring in the dedicated steam generator components separated hydraulically from the reactor coolant and by the condition that the coolant within the reactor remains in a subcooled condition.

Probabilistic Risk Assessment (PRA)

A systematic method for addressing the risk triplet as it relates to the performance of a complex system to understand likely outcomes, sensitivities, areas of importance, system interactions, and areas of uncertainty. The risk triplet is the set of three questions that the NRC uses to define "risk": (1) What can go wrong? (2) How likely is it? And (3) What are the consequences? NRC identifies important scenarios from such an assessment.

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Quality Assurance Program (QAP)

The NuScale quality assurance program as a whole, including the Quality Management Plan (QMP), Project Quality Plan (PQP), and all applicable sub-tier implementing documents.

Quality Management Plan (QMP)

The top level or Tier 0 document of the NuScale requirements document hierarchy. The Quality Management Plan (QMP) defines the quality management system. This document establishes governance, expectations, and rules and includes roles, responsibilities, accountabilities, and authorities (R2A2).

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Reactor Recirculation Valve (RRV)

An engineered safety component located at an elevation proximal to and above the top of the core that connects the primary system downcomer to the high pressure containment volume.

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Small Modular Reactor (SMR)

A nuclear reactor that produces 300 MWe or less, featuring standardized, mass produced components.

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United States Nuclear Regulatory Commission (NRC)

The regulatory authority for commercial nuclear power plants and other uses of nuclear materials, such as in nuclear medicine, through licensing, inspection and enforcement of its requirements.

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