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Conference Spotlight
2025 ANS Winter Conference & Expo
November 9–12, 2025
Washington, DC|Washington Hilton
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PR: American Nuclear Society welcomes Senate confirmation of Ted Garrish as the DOE’s nuclear energy secretary
Washington, D.C. — The American Nuclear Society (ANS) applauds the U.S. Senate's confirmation of Theodore “Ted” Garrish as Assistant Secretary for Nuclear Energy at the U.S. Department of Energy (DOE).
“On behalf of over 11,000 professionals in the fields of nuclear science and technology, the American Nuclear Society congratulates Mr. Garrish on being confirmed by the Senate to once again lead the DOE Office of Nuclear Energy,” said ANS President H.M. "Hash" Hashemian.
Mitsuhiro Suzuki, Kanji Tasaka, Yoshinari Anoda, Hiroshige Kumamaru, Hideo Nakamura, Masayoshi Shiba
Nuclear Technology | Volume 70 | Number 2 | August 1985 | Pages 189-203
Technical Paper | Nuclear Safety | doi.org/10.13182/NT85-A33643
Articles are hosted by Taylor and Francis Online.
Three loss-of-coolant accident (LOCA) tests were conducted at the Rig of Safety Assessment (ROSA)-III test facility, which simulates boiling water reactor (BWR)/6-251 with a volumetric scaling factor of 1/424. The fundamental features of the recirculation pump discharge line break LOCA and the effects of break areas on the features are investigated. It has been confirmed experimentally that the LOCA phenomena in the discharge line break are analogous to those in the suction line break with the same effective choking flow area, which is a sum of the least choking flow areas along the break flow paths and controls the system pressure responses. In general, the maximum effective choking flow area is (Aj + Ap) for discharge line breaks and (Aj + A0) for suction line breaks, where Aj, Ap, and A0 are the flow areas of the jet pump drive nozzles, the main recirculation pump discharge nozzle, and the break, respectively. The similarity between the ROSA-III test and a BWR LOCA has been confirmed in the key phenomena by the analyses using the RELAP5/MOD1 code. An atypical behavior is observed in the fuel rod surface temperature transient in the early phase of blowdown due to the limitation of the ROSA-III initial core power.
