A ROSA by Any Other Name…

SAM-N – In Today’s World it is hard to imagine any emerging and promising technology to be so dominantly controlled by a single nation as nuclear power was by the United States during the first few decades of the Atomic Age. Had the technology arose originally for peaceful applications, the story might have been different. U.S. President Dwight Eisenhower’s Atom’s for Peace speech of 1953 was a first step to alleviate international concerns over the United States’ monopoly on nuclear technology.

Given the political climate of the time, it may have seemed strangely logical to select the country that had already suffered the most from the power of the atom, Japan, as the best place to advance this new technology openness. Japan’s entry as a nuclear power must have had its share of political intrigue between the survivors of the nuclear bombings at Hiroshima and Nagasaki and the post-war leadership. Nonetheless, a legal infrastructure supporting the development of nuclear power as a source of electricity was established by 1955 and with it the Japan Atomic Energy Research Institute (JAERI).

After several years considering gas reactor technology imported from British experience at Calder Hall, Japan’s electric power companies transitioned to light water reactor (LWR) technology in the early 1960s. Through subsequent investments in turnkey projects with both Westinghouse and General Electric, Japan’s domestic nuclear power industry would be firmly established by 1970. This included the Japanese government’s adoption of a nuclear regulatory framework from the United States and the establishment of a domestic research and development program.

Aware of the rapidly evolving nuclear policy debates occurring in the United States at that time, the Japanese government initiated the SAFE (Safety Assessment and Facilities Establishment) Project to assess the effectiveness of LWR engineered safeguards, in particular, emergency core cooling system (ECCS) designs. As part of this program, JAERI commissioned the ROSA (Rig Of Safety Assessment) program for the study of thermal-hydraulic response of LWRs during loss-of-coolant accidents (LOCAs) and operational transients. It was first dedicated for separate-effect tests mainly for fundamental blowdown phenomena and core heat transfer in the ROSA-I Program, similar to the earliest Semiscale testing in the United States (ROSA-I occupied only 1.7 m3). The ROSA Program then proceeded to confirm the ECCS effectiveness through integral systems tests for pressurized water reactors (PWRs) in the ROSA-II Program and for boiling water reactors (BWRs) in the ROSA-III Program. These early ROSA program tests utilized relatively small test facilities (1/400 volume-scaled in ROSA-II and -III) with a half-height core.

The Three Mile Island Unit-2 (TMI-2) reactor accident in 1979 led to a redirection of the ROSA-III program toward BWR small-break LOCAs and operational transients with emphasis on natural circulation involving two-phase stratified flows and counter-current flows. Test result were found to relate well to the expected BWR performance through comparison to best-estimate computer codes and similar tests performed at General Electric’s TLTA and FIST facilities. Unfortunately, with ROSA-III’s short core, smaller core power, the larger stored heat release from structure materials, large heat loss from the system, and the oversized recirculation loop volumes, similarity between the ROSA-III test results and the LOCA phenomena in the reference BWR was not ideal. Subsequently, JAERI began plans for a new test facility with system geometry more representative of a full-size prototype.

In 1980 JAERI initiated the ROSA-IV Program and in 1985 completed the construction of the Large Scale Test Facility (LSTF) emulating the design of a large Westinghouse PWR, departing from the BWR reference used for ROSA-III.  As with similar test programs at Babcock & Wilcox, General Electric, and Westinghouse ongoing at that time, the LSTF was designed with prototypical component elevations, large loop piping diameters, prototypical pressure levels, simulated system controls, and core power levels sufficient to simulate the decay power through long-term transients typical of small break LOCAs. The core region was represented by an impressive 1008 full-height, electrically-heated rods configured in a typical Westinghouse PWR 17x17 fuel assembly matrix. The total number of simulated fuel rods corresponded to a volume-scale of 1/48 of the reference PWR. Given the practical limits on utility resources, LSTF’s maximum core power of 10 MWt was equal to or less than 14% of the scaled rated power. This stands as the largest all-electric thermal-hydraulic test facility ever constructed and operated.

By the end of the ROSA-IV program in 1993 over 80 experiments had been completed. The result highlights included TMI-2 type experiments, LOCAs with various break conditions and plant recovery actions, an experiment for the OECD/NEA/CSNI international standard problem program (ISP-26), counter-part tests with France’s BETHSY facility, Germany’s PKL facility, and the United States’s Semiscale facility. It was also used to investigate the steam generator tube rupture (SGTR) accident at Japan's Mihama Unit-2 plant. Unlike the thermal-hydraulics test facilities in the United States, LSTF continues to support an active research program. Immediately following the end of the ROSA-IV program, JAERI began to host experiments at the LSTF under the ROSA-V program. This program has focused on accident management strategies for both design- and beyond-design-basis accidents and the effectiveness of engineered passive-safety features introduced with more recent advanced LWR designs (e.g., AP1000 and ESBWR).