[en] Monju, Japan’s prototype sodium cooled fast reactor of 280 MW(e) class, restarted its test operation in 2010 after a long stoppage since the sodium leak accident in 1995. The zero power system startup tests were successfully conducted. The major achievement of the tests was an accurate prediction of reactor physics parameters with a core having a complex fuel composition that includes americium-rich fuel. The hardware troubles recently experienced, none of them being safety significant, have been restored to make the plant ready for the next power increase tests. The reactor, however, has been put into a standby mode again since the Fukushima-Daiichi accident of 11 March 2011, at least until the national energy and environment programme is revised and a research plan using Monju is developed through 2013. We believe the roles of Monju will not change and comprise further enhancement of safety against severe accidents; demonstration of stable power generation and actinide burning; provision of technology and knowledge base for future sodium cooled fast reactors; and use of the plant as an international research facility. (author)

[en] MONJU is a prototype fast breeder reactor in Japan designed to have a 280-MW(electric) output. The Power Reactor and Nuclear Fuel Development Corporation (PNC) started its construction in the autumn of 1985 in Tsuruga. The loading of the core fuel assemblies was started in October 1993, and the preoperational test is ongoing. MONJU uses 198 mixed-oxide (MOX) fuel assemblies as core fuel and 172 depleted uranium assemblies as blanket fuel. Assemblies loaded in-core and stored in the ex-vessel storage tank (EVST) reside in liquid sodium. These plutonium-containing fuel assemblies, MOX, and irradiated depleted uranium are regarded as in the difficult-to-access area, and the flows of fuel assemblies into and out of the area must be verified. Flow is verified by fuel flow monitors measuring radiation, which can limit inspector attendance during fuel handling

[en] A long-life in-core neutron detector was developed from 1973 through 1978 and since then, it has been applied to the prototype Heavy Water Reactor ''FUGEN'' as a local power monitor. More than 10 year's operation experience of the detector proved its performance as was expected. The detector is a fission chamber of regenerative type whose tube is made of Zircaloy-2 metal for neutron economy. In-core neutron detectors for the Local Power Monitor are placed in heavy water moderator of 50⁰C. The in-core conditions bring a high level thermal neutron flux environment. It tends to hurt the linearity and shorten the life time of the detectors because of rapid burnup of the neutron conversion material (²³⁵U) in the detector compared to the light water reactors. These problems were solved by some improvements and it was demonstrated that the detector had the performance of high linearity and long-life characteristics

[en] The verification of the fuel assemblies flows is designed to be made with fuel flow monitors measuring radiation, which can abridge inspector attendance during the fuel handling. The monitors consist of ex-vessel radiation monitors (EVRMs) and an exit gate monitor (EXGM). EVRMs are linked to the ex-vessel transfer machine, which charges and discharges fuel assemblies into and out of the core and the EVST (a sodium tank located in an area adjacent to the reactor containment building which temporarily stores new and irradiated fuel assemblies). Flows of fuel assemblies into and out of the core and the EVST can be monitored by EVRMs, and the flows of irradiated fuel assemblies into the spent fuel pond (water pool) can be monitored by the EXGM, including the flow direction. These monitors have been developed through JASPAS (Japan Support Programme for Agency Safeguards) and through joint R and D by the Power Reactor and Nuclear Fuel Development Corporation and the US Department of Energy. As well as the above monitors, the IAEA installed an entrance gate monitor using a neutron coincidence counter and several optical surveillance cameras. Using this equipment the transfers of fuel assemblies inside the facility can be monitored in an unattended mode and routine inspections are made monthly. The paper describes details of the fuel transfers in the MONJU facility, their verification through flow monitors, and the functions of other safeguards equipment. (author). 3 refs, 5 figs, 1 tab

[en] Monju is a prototype fast breeder reactor designed to have an output of 280 MW (714 MWt), fueled with mixed oxides of plutonium and uranium and cooled by liquid sodium. The principal data on plant design and performance are shown in Table 1. Monju attained initial criticality in April 1994 and the reactor physics tests were carried out from May through November 1994. The reaction rate distribution measurement by the foil activation method was one of these tests and was carried out in order to verify the core performance and to contribute to the development of the core design methods. On the basis of the reaction rate measurement data, the Monju initial core breeding ratio and the power distribution were evaluated. (author)