HOME Nuclear Acronyms

See also:
AECL's glossary page
Glossary of Nuclear Terms (US NRC)

AECL Atomic Energy of Canada Ltd.
AECB Atomic Energy Control Board
AESOP Atomic Energy Simulation of Optimization (computer code)
ASDV Atmospheric Steam Discharge Valve
ASSERT Advanced Solution of Subchannel Equations in Reactor Thermalhydraulics (computer code)
ASTM American Society for Testing Materials
BLC Boiler Level Control
BLW Boiling Light Water
BPC Boiler Pressure Controller
CCP Critical Channel Power
CHF Critical Heat Flux
CPR Critical Power Ratio
CRL Chalk River Laboratories
CRT Cathode Ray Tube
CSA Canadian Standards Association
CSDV Condenser Steam Discharge Valve
CSNI Canadian Standards for the Nuclear Industry
DBE Design Base Earthquake
DCC Digital Control Computer
DF-ET Drift Flux-Equal Temperature
DF-UT Drift Flux-Unequal Temperature
DNB Departure from Nucleate Boiling
ECC Emergency Core Cooling
ECI Emergency Core Injection
EFPH Effective Full Power Hours
EVET Equal Velocity Equal Temperature
EVUT Equal Velocity-Unequal Temperature
EWS Emergency Water Supply
FBR Feed, Bleed and Relief
FP Full Power
HEM Homogeneous Equilibrium Model
HTS Heat Transport System
HWP Heavy Water Plant
HYDNA Hydraulic Network Analysis (extinct computer code)
I&C Instrumentation and Control
IBIF Intermittent Buoyancy Induced Flow
ICRP International Commission on Radiological Protection
LOC Loss of Coolant
LOCA Loss of Coolant Accident
LOC/LOECC Loss of Coolant with Coincident Loss of Emergency Core Cooling
LOP Loss of Pumping
LOR Loss of Regulation
MCCR Ministry of Corporate and Consumer Relations
MCS Maintenance Cooling System
MHD Magneto hydrodynamics
milli-k Unit of reactivity for reactor physics
NPD Nuclear Power Demonstration
NPSH Net Positive Suction Head
NUCIRC Nuclear Circuits (computer code)
OECD Organization for Economic Co-operation & Development
OH Ontario Hydro
PGSA Pickering Generating Station A
PHTS Primary Heat Transport System
PHW Pressurized Heavy Water
PHWR Pressurized Heavy Water Reactor
PRESCON2 Pressure Containment (computer code)
QA Quality Assurance
RAMA Reactor Analysis Implicit Algorithm
R&M Reliability and Maintainability
RB Reactor Building
RIH Reactor Inlet Header
ROH Reactor Outlet Header
RTD Resistance Temperature Detectors
SDM Safety Design Matrices
SOPHT Simulation of Primary Heat Transport (computer code)
SRV Safety Relief Valve
TMI Three Mile Island
TOFFEA Two Fluid Flow Equation Analysis (computer code)
UVUT Unequal Velocity Unequal Temperature
VB Vacuum Building
VC Vacuum Chamber
WRE Whiteshell Research Establishment