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Term
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Definition
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A
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B
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Beyond Design Basis Accident (BDBA)
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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.
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Boiling Water Reactor (BWR)
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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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C
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Combined License (COL)
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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).
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Containment Heat Removal System (CHRS)
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A passive heat removal system for the containment vessel. During a LOCA, when containment is pressurized, the CHRS removes heat and depressurizes the containment by a conduction and convection process which transfers energy to the reactor pool. The containment vessel is submerged in the reactor pool, so containment heat removal begins immediately when containment is heated and pressurized.
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D
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Decay Heat Removal System (DHRS)
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A passive heat removal system for the RCS. The DHRS provides secondary side reactor cooling for non-LOCA events when normal feed water is not available. The system is a closed loop, two-phase natural circulation cooling system.
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Dose (Radiation Dose)
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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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E
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Emergency Core Cooling System (ECCS)
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A passive reactor coolant recirculation cooling system. The ECCS is designed to maintain core cooling during a LOCA. It functions by inducing natural circulation flow between the CNV and the RPV. It consists of a set of valves and flow paths that allow communication between the Reactor Pressure Vessel and the Containment Pressure Vessel.
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F
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G
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General Design Criteria (GDC)
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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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H
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I
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J
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K
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L
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Licensing Topical Report (LTR)
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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).
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Light-Water Reactor (LWR)
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A type of thermal reactor that uses a light-water moderator to thermalize neutrons.
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Loss-Of-Coolant Accident (LOCA)
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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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M
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Multi-Application Small Light Water Reactor (MASLWR)
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A conceptually designed small, light water, natural circulation integral pressurized
water reactor.
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N
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NRC
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Nuclear Regulatory Commission
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Nuscale Power Module
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Ease NuScale Power Unit produces.>45 MWe and has its own turbine, condenser, and
feed water system and operates independent of the other units.
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Nuscale Power Plant
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An electrical or steam power generating station that uses one or more NuScale Power
Modules as the energy source.
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Nuclear Steam Supply System (NSSS)
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The NSSS is the set of components in a nuclear power plant which produce steam from
the core energy production.
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O
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P
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Pressurized- Water Reactor (PWR)
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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.
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Probabilistic Risk Assessment (PRA)
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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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Q
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Quality Assurance Program (QAP)
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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.
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Quality Management Plan (QMP)
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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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S
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Small Modular Reactor (SMR)
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A nuclear reactor that produces 300 MWe or less, featuring standardized, mass produced
components.
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T
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U
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U.S. Nuclear Regulatory Commission (NRC)
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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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V
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W
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X
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Y
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Z
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