Present status of a research on intermediate energy nuclear data, which are required for various applications, are reported. As an important foundation for data evaluation at intermediate energy region, nuclear reaction theories and models such as the preequilibrium model, intranuclear cascade model, quantum molecular dynamics, optical model potential and fission model, are reviewed. The model calculation codes are having been developed based on these theories and models. Status of these codes and results of a few benchmark tests are also described as well as the status of production of evaluated intermediate energy nuclear data files and improvement of hadron transport theories.

View full abstract

[in Japanese]

View full abstract

[in Japanese]

View full abstract

Based on the recognition that layout elements of a repository are important factors to govern total costs of the high-level radioactive waste disposal, the study on their feasibility has made as a trial from the view points of economics, shielding, strength against pressure, retention capability and especially mechanical interactions within near-field under some premises and conditions discribed below to establish an economical and rational repository concept.
(1) Host rocks: Granite (800m deep) and sedimentary (500m deep)
(2) Emplacement methods: Horizomtal
(3) Thickness of overpacks: 20cm
(4) Thickness of bentonite: 70cm
(5) Inner diameter of repository tunnel: 2.3m
(6) Center distance betweem repository tunnels: 20m
In addition, perceiving long-term behaviors of near field composed of waste package, overpack. bentonite and surrounding rock, some suggestions or tasks for each items above are also referred.
Among assumptions for these studies, there are some items, such as corrosion rate and performance of overpack, as well as mechanical properties of bentonite and so on, not comfirmed fully yet by experiments, and so studies are made under conservative assumptions for these items. Considering study results as a whole, the premises and conditions specified in these studies are concluded to be highly feasible.

View full abstract

The Intermediate Heat Exchanger (IHX) of the High Temperature Engineering Test Reactor (HTTR) is a helium-helium type heat exchanger with the heat capacity of 10MW. The internal structures such as heat transfer tubes made of Hastelloy XR are used normally at about 930°C because the IHX can supply the secondary helium gas with the maximum temperature of 905°C transferred from the primary helium gas of 950°C. They compose the primary boundary and their creep strain and creep damage evaluated conservatively based on the elastic analysis cannot meet the required criteria with enough margin. The design method based on the complete creep analysis was used for the primary coolant boundary components.
This report describes an evaluation procedure based on the creep analysis and its results. The results show that the creep strain and creep damage increased rapidly during a few initial cycles, however, increased far less after that. The results can meet the design limit for 10⁵h of the HTTR life time.

View full abstract

Numerical simulation of an unprotected loss-of-flow( ULOF) accident has been performed for a large liquid-metal-cooled fast breeder reactor (LMFBR) equipped with gas expansion modules (GEMs) in the radial periphery of the reactor core. The effectiveness of the GEMs in small fast reactors was demonstrated already in the passive safety testing in the Fast Flux Test Facility. According to neutronic calculations based on the transport theory, even in large reactors of electrical power 600 to 1, 300 MW, the reactivity worth of GEMs, which replace one layer of radial blanket fuel subassemblies, ranges from -1.9 to -1.4, depending on the size of the core. A simulation of ULOF transient was performed with a 5.5s flow-halving time in a 600 MWe LMFBR equipped with GEMs of -1.9$ reactivity worth. The result showed that, if 10% of the rated core coolant flow by pony motors was available following the main pump coastdown, the GEM reactivity alone could bring the reactor subcritical and the predicted maximum coolant temperature was substantially lower than the sodium boiling point. The reactivity worth calculations, a modeling of gas expansion behavior, and ULOF simulation together with needs of further development for the GEM application are described.

View full abstract

A new mechanical dispersion process to produce oxide dispersion strengthened (ODS) vanadium alloy has been successfully developed. In order to study effects of irradiation on microstructural evolution in ODS V alloy and unalloyed V, transmission electron microscopy observation and Vickers microhardness test were performed. ODS V alloy has five times as large Vickers microhardness as unalloyed V. ODS V alloy has twin structure. After irradiated with 3 MeV Au²⁺ ions up to 1.5×10²⁰ ions/mm² (about 60dpa) at 813 K, microcrystalline transformation and defect clusters were observed in unalloyed V. On the other hand, the twin structure was stable in ODS V alloy. After irradiated up to 6.Ox10²⁰ ions/mm² (about 250 dpa), many microvoids were observed in unalloyed V, while the twin structure in ODS V alloy was disappeared. However no voids could be observed in ODS V alloy. Postirradiation anneal hardening was observed in unalloyed V after irradiated by neutrons up to 8.3×10²²n/mm²(En> 1 MeV, 144 h). Nevertheless no radiation anneal hardening could be observed in ODS V alloy. Therefore ODS V alloy has a significant resistance to irradiation damage.

View full abstract