The major test facilities used in developing an understanding of nuclear power plant transients and nuclear reactor safety are briefly describe here in alphabetic order. The list while extensive is hy no means exhaustive.

BETHSY - France

Design specifications call for a l/l00 scale (by volume) representation of a three-loop pressurized water reactor (PWR). The BETHSY system used three identically sized primary coolant loops each with a full-elevation steam generator and scaled pump. The electrically heated core consisted of 419 full-elevation heater rods capable of simulating power levels up to 3MW (10% of scaled, full power]. The vessel employed an external pipe downcomer like that used in the Idaho Semiscale facility (described below). Test program focused on simulating small breaks, Steam generator tube rupture and operational transients.

CCTF - Japan

The Cylindrical Core Test Facility (CCTF) was an experimental test facility, located in Tokai, designed to model a full-height core section and major components of a four-loop pressurized water reactor. The facility was to provide information on system thermal-hydraulics during the refill and reflood phases of a large break loss of coolant accident (LOCA) in a PWR. The core contained 2000 electrically heated rods (32 rod assemblies) in a cylindrical array. The facility was part of the 2D/3D joint research program among Japan, Germany, and the United States.

FEBA - West Germany

The Flooding Experiments with Blocked Arrays (FEBA} facility, located in Karlsruhe, was used to conduct experiments with 25 electrically heated rod bundles to examine heat transfer effects of coolant channel blockages and mid-plane spacer grids. The test bundle consisted of 1 x 5 and 5 x 5 arrays with full length indirectly heated rods. Both plate and sleeve blockages were used as flow blockage devices. Eight test series for the 5 x 5 rod arrays were conducted with various combinations of blockage shapes and geometries.

FIST - United States (General Electric)

The Full-height Integral Simulation Test (FIST) facility was cosponsored by the U.S. Nuclear Regulatory Commission (NRC), General Electric Company (GE), and the Electric Power Research Institute (EPRI). The facility was located at San Jose, California, and operated by GE. FIST was capable of simulating large and small break LOCAs and operational power transients in boiling water reactors (BWRS). It had one 8 x 8 fuel bundle, and was volume scaled to 1/624 of a full size BWR. The bundle was electrically heated and surrounded by a prototypical Zircaloy channel. The facility had two external recirculation loops, two jet pumps, and all other major Nuclear Steam Supply System (NSSS) components.

FIX-II - Sweden

FIX-II was a 1/777-scaled Swedish BWR loop facility, located at Studsvik, that was primarily designed for LOCA blowdown experiments. The fuel rod simulators were full length. The experimental program consisted of studying guillotine breaks, split breaks, and pump trips.

FLECHT-SEASET - United States

The Full Length Emergency Cooling Heat Transfer - Separate Effects and System Effects Test (FLECHT-SEASET) Program was jointly sponsored by the USNRC, Westinghouse and EPRI. The experiments were conducted by Westinghouse, in Pittsburgh, Pennsylvania. The purpose of the program was to provide experimental reflood heat transfer and other two-phase data in simulated PWR geometries for postulated LOCA conditions. The program performed separate effects tests on a steam generator component and 21, 161, and 163 electrically heated rod bundles. System tests were performed on a two-loop integral facility. The tests ranged from forced and gravity feed reflood with both unblocked and blocked rod arrays to system natural circulation tests.

FR2 - West Germany

FR2 was a research reactor with inpile loops located at Karlsruhe. It was used for materials testing and inpile fuel rod behavior experiments. In particular, tests were performed to investigate fuel rod ballooning and burst behavior under simulated PWR-LOCA reflood conditions. The reactor used 97 cm long PWR single rods with a burnup between 2500 and 35,000 MWd/t and 45 to 90 bar rod internal pressures.

FRIGG - Sweden

FRIGG was a Swedish ASEA-ATOM facility sponsored by a consortium of Scandinavian Atomic Energy groups. FRIGG was a separate effects test facility used to gather hydrodynamic data for the MARVIKEN facility under various system conditions. The facility had a single pass loop, pressure vessel with no downcomer region, a steam separator, a condenser, and a natural circulation loop with a forced circulation bypass. Primary areas investigated by FRIGG were mass flux vs. input power, two-phase pressure drop, temperature and void distributions, and dynamic measurements.

GERDA/OTIS - United States

GERDA was a single loop experimental system designed to investigate thermal-hydraulic behavior. The facility was sponsored by Brown-Bovery (West Germany) and operated by Babcock and Wilcox {B&W) at its Alliance, Ohio, research center. The system includes a pressure vessel with a core heat simulator, a 19 tube once-through steam generator (OTSG), and associated loop piping. GERDA was designed primarily to simulate small break LOCA and transient events involving natural circulation heat removal. Thus, the system had no coolant pump and was restricted to decay heat core power. Approximately 50 tests were performed, and the facility was later upgraded in 1984 and renamed OTIS.

GOTA - Sweden

The GOTA test facility was a separate effects facility scaled in the ratio of 1/676 to investigate the effectiveness of the Emergency Core Cooling (ECC) system used in Swedish BWRs. It had a pressure vessel, a pressurizer, circulation pump, and necessary piping and peripheral equipment to perform reflood experiments. The heated bundle consisted of 63 electrically heated rods.


HDR was a superheated steam cooled reactor used as a blowdown test facility. Located at Karlsruhe, the facility's objectives were to study critica1 flow, discharge flow rates in pipes and valves, and propagation of pressure waves inside the system, including deformation of the core barrel. Operational behavior of valves under accident conditions were also studied.


The High Pressure Test Rig was used at Winfrith for critical flow studies at sub- and super-critical pressures to generate information relevant to pressurized transients.

LOBI - Italy

The LOBI test facility operated by the EURAT0M research center in Ispra, was jointly sponsored by the Commission of the European Communities and the West German government. The system was a 1/700-scale model of a four-loop 1300 MWe pressurized water reactor of Kraftwerk Union (KWU) design. The full-length simulated core used 64 electrically heated pins with a maximum power of 53 MW. The two coolant loops, one simulating three, the other simulating one PWR coolant loop, both had a coolant circulation pump and full-length, active steam generator. The system could be configured to perform hot leg, cold leg, and pump suction breaks as well as steam generator U-tube ruptures and stuck open pressurizer power operated relief valve (PORV) transient. The original LOBI system, completed in 1979, was designed primarily to investigate large and intermediate breaks. The successor facility , L0B1-Mod 2, included a High Pressure Injection System (HPIS) and an Auxiliary Feedwater System (AFS), and extensively instrumented steam generators. This facility focused on small breaks and a variety of operational transients.

L0FT - United States

The Loss-of-Fluid-Test (LOFT) facility, sponsored first by the USNRC and later by the Organization for Economic Conservation and Development (OECD), was operated by EG&G Idaho, Inc. for the U.S., Department of Energy at the Idaho National Engineering Laboratory (INEL). It was the only nuclear-powered, integral test facility in the free world. LOFT was modeled after a four-loop pressurized water reactor, and was capable of simulating large breaks, small breaks, and operational transients. The system had two coolant loops, one active loop (with pumps and steam generator) representing three of the four PWR loops, and one inactive loop (with resistance simulators for pump and steam generator). The core consisted of approximately 1300 fuel rods (5.5 ft long), with a maximum thermal power output of 55 MW.

MIST - United States

The Multi-loop Integral System Test (MIST) facility was a successor to the GERDA/OTIS facility. It was planned under the joint sponsorship of B&W, NRC, EPRI, and the B&W Owner's group. MIST was designed to emulate B&W 2 x 4 design, including scaled coolant pumps.


The Marviken facility, located at the oil-fired Marviken Power Station in Sweden, was one of the largest research facilities in the world used to study nuclear pressure vessel, containment, blowdown and pipe rupture phenomena. This facility was used in many internationally organized programs to understand the behavior of reactor containments and reactor system blowdown under different conditions. The experimental facility grew from an actual BWR nuclear vessel which was never charged with nuclear fuel.

MERLIN - England

MERLIN was a rig containing a bundle of electrically heated fuel rod simulators and was used to investigate reflooding. It was located at the Springfield Nuclear Laboratory, within the United Kingdom Atomic Energy Authority (UKAEA).

PKL - West Germany

The PKL facility was a 1/134 scale model of a four-loop 1200 MWe pressurized water reactor. The facility featured two single coolant loops and one double loop. All three loops employed a full-length steam generator but had no pumps. The reactor vessel contained a core simulator with 340, indirectly heated rods capable of a maximum power of 1.5 MW. The primary coolant system operating pressure was 4 MPa. The PKL facility was primarily used to study core and system behavior during the refill/reflood phase for large breaks. It also was used for extensive investigation of small break LOCAs including effects of alternate ECC systems, heat transfer mechanisms in the core and the steam generator, and effects of noncondensible gas. In PKL was upgraded in 1983 to included modifications for examining end-of-blowdown transients.

REBEKA - West Germany

The REBEKA facility was used for investigating rod ballooning behavior during reflood following a large break LOCA in a PWR. The test bundle consisted of an indirectly heated 1 x 3 (later a 5 x 5) rod array with an internal rod pressure of 70 bar.

ROSA - III - Japan

The Rig of Safety Assessment (ROSA)-III facility was a volumetrically scaled (1/424) BWR system with electrically heated core designed to study BWR system response during a large or smallb re ak LOCA. There were four half-length, electrically heated power bundles in the core. One bundle simulated a high-power channel while the other three represent average-power channels. The facility included all the other major components of a BWR system.

ROSA - IV (LSTF) - Japan

The Large Scale Test Facility (LSTF) is a 1/48 scale model (by volume of a four-loop pressurized water reactor, featuring two equal-sized primary coolant loops. Each loop has a full-length, active steam generator and coolant pump (restricted to 20% of scaled, full flow). The LSTF core was full-height and consisted of 1080 electrically heated rods and 104 unheated rods. Maximum power was 10 MW, or 14% of scaled, full-power. The system primarily constructed to investigate small breaks (less than 10% of pipe area) in the hot leg, cold leg, or vessel; steam generator tube ruptures, and feedwater and steam line transients.

SCTF - Japan

The Slab Core Test Facility (SCTF) was a full-length (12 ft), full-radius (8 bundles in a row), electrically heated core test facility. It was built to study the two-dimensional thermal-hydraulic effects of emergency core cooling in the reactor vessel during the refill and reflood phases of a loss-of-coolant accident in a PWR. The facility was part of the 2D/3D joint research program among Japan, Germany, and the United States.

SEMISCALE (Mod-2B) - United States

Semiscale, located at the INEL, was a l/l700 scale representation of a four-loop PWR, and utilized a full-length electrically heated core simulator (maximum power 2 MW]. The primary coolant system consisted of a pressure vessel, two circulation coolant loops (1:3 ratio) and associated ECCS systems (accumulators, high and low-pressure injection) . Elevation relationships were scaled 1:1 with the full-scale plant. The system was highly instrumented and featured an external pipe downcomer on the vessel permitting direct measurement of coolant density in the core bundle. Experiments include large breaks, intermediate breaks, small breaks, steam generator tube rupture, and secondary breaks.

SSTF - United States

The Steam Sector Test Faciltiy (SSTF) was a full scale, 30-degree-sector mockup of a BWR. The simulated system included multidimensional phenomena during the refill/reflood phase of a BWR LOCA, and other transients where the automatic depressurization system was activated. The facility was located in Lynn, Massachusetts . The program was jointly sponsored by the U.S. NRC, GE, and EPRI . The core had 58 simulated fuel bundles and all other major NSSS components.

TBL - Japan

The Two Bundle Loop (TBL), located at the Hitachi Research Laboratory, was a BWR thermal-hydraulic systems test facility . It was scaled to a BWR/5-251 plant and consisted of two full size electrically heated bundles in a pressure vessel two full length jet pumps and two recirculation lines. Each bundle had 63 heater rods and one water rod in an 8 x 8 array. The maximum power of each bundle was approximately 6 MW. Several small and large break LOCA transients were run at the TBL.

THETIS - England

THETIS was an electrically heated test facility, located at UKAEA in Winfrith, capable of operating at 1000 psi and at decay power levels. The facility's purpose was to examine pin bundle heat transfer under transient, two-phase conditions. Additionally, it used to investigate forced and gravity-feed reflooding and level swell in a simulated ballooned cluster of 49 full length fuel rods.

TITAN - England

TITAN is an electrically heated, full-length full-power PWR facility at UKAEA, Winfrith. TITAN operated at a total power of 9 MW and pressures up to 2600 psi and was used to examine heat transfer under transient and normal operation, including critical heat flux and dryout during depressurization and low flow, low pressure conditions.


The CLUSTER LOOP was a completely self contained inpile loop within the steam-generating heavy water moderated reactor (SGHWR) at UKAEA in Winfrith. It typically contained a full length (12 ft) 57 pin bundle of fission heated rods and was rated at 1000 psi. The facility was used for a variety of thermal-hydraulics experiments related to SGHWR (e.g., ECCS performance during severe LOCA conditions}. A second inpile loop containing approximately three pins was also installed in SGHWR.

TLTA - United States

The Two Loop Test Apparatus (TLTA), located in San Jose, California, was a 1/624-volume-scaled, non-nuclear GE facility for the simulation of thermal-hydraulic phenomena in a BWR. The original design used one 7 x 7 electrically heated bundle. A later modification incorporated an 8 x 8 bundle. The principal components of TLTA included the vessel jet pumps, recirculation loops, blowdown lines, and feedwater and ECC injection systems. The facility simulated Design Basis Accidents, steam-line breaks and small-break LOCAs. TLTA was the predecessor of the FIST facility.

UPTF - West Germany

The Upper Plenum Test Facility (UPTF) was a 4 Loop 1:1 PWR test facility located in Mannheim, West Gexrnauy. The construction of the facility and the experiments was carried out by Kraftwerk Union (KWU) and was sponsored as part of the international 2D/3D program. UPTF had a steam and hot water operated core simulator including simulation of steam generator feedback. The facility's main objective was to investigate primary system behavior during end of blowdown refill, and reflood for different ECCS injection modes with emphasis on core-upper plenum interaction, downcomer behavior, and hot and cold leg behavior.

Test Facility Summary from:

Anticipated and Abnormal Plant Transients in Light Water Reactors, Proceedings from an American Nuclear Society Topical Meeting, September 26-29, 1983, Jackson, Wyoming.