Musashi Reactor Research Articles

Benchmark Analysis of TRIGA Mark II Reactivity Experiment Using a Continuous Energy Monte Carlo Code MCNP

The benchmark analysis of reactivity experiments in the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by a three-dimensional continuous-energy Monte Carlo code MCNP4A. The reactivity worth and integral reactivity curves of the control rods as well as the reactivity worth distributions of fuel and graphite elements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated values of integral reactivity curves of the control rods were in agreement with the experimental data obtained by the period method. The integral worth measured by the rod drop method was also consistent with the calculation. The calculated values of the fuel and the graphite element worth distributions were consistent with the measured ones within the statistical error estimates. These results showed that the exact core configuration including the control rod positions to reproduce the fission source distribution in the experiment must be introduced into the calculation core for obtaining the precise solution. It can be concluded that our simulation model of the TRIGA-II core is precise enough to reproduce the control rod worth, fuel and graphite elements reactivity worth distributions.

Neutron beam design for neutron radiography at the Musashi reactor

Studies were carried out for the design of a thermal, an epithermal, and a fast neutron beam for neutron radiography (NRG) in the horizontal beam port or in the thermalizing column at the Musashi reactor (TRIGA-II, 100 kW). Thermal, epithermal and fast neutron energy ranges were selected as <0.4 eV, 0.4 eV–10 keV, and > 10 keV, respectively. To produce a good neutron beam in terms of intensity and quality, several materials (bismuth (Bi), silicon (Si) and alumina (Al 2O 3)) were examined as neutron filters. The MCNP (Monte Carlo Neutron Photon) computations showed that a thermal, an epithermal, and a fast neutron beam could be produced using single-crystal Si, Al 2O 3 and Bi filters, respectively. Such neutron beams, if built into the Musashi reactor, could increase extensive application of NRG in Japan.

武蔵工大炉医療用照射場の中性子ビーム設計

Design studies were carried out for the Musashi reactor (TRIGA-II, 100kW) with a view to generating epithermal and thermal neutron beams applicable to boron neutron capture therapy (BNCT). By Monte Carlo calculations, the thermal neutron beam intensity 1.1×109n/cmcm2·s and the epithermal neutron beam intensity 4.1×108n/cmcm2·s at the operation power of 100kW have been obtained at the irradiation ports in the thermal and thermalizing columns, respectively. These beams are little adulterated with fast neutrons and gamma-rays, and are relatively forward-directed. Depth-dose calculations in a tissue phantom showed that above thermal and epithermal neutron beams would be useful to treat superficial (-4cm) and deeper (-6.5cm) tumors, respectively. The full-dose treatment time is estimated about 2.5h. These neutron beam designs suggest how TRIGA reactors could be modified for BNCT, and the produced neutron beams, if built into the Musashi reactor, will significantly increase the chance of success of BNCT in Japan.

Design Studies of an Epithermal Neutron Beam for Neutron Capture Therapy at the Musashi Reactor

Studies were carried out to design an epithermal neutron beam for neutron capture therapy at the Musashi reactor, a TRIGA-II of 100kW. The idea in this design is to use spent fuel elements as a converter assembly to convert thermal neutrons from the core to fission neutrons which would then be moderated to epithermal neutrons. In the design, 43 elements were placed outside of the graphite reflector but inside the tank. These elements could be cooled by water in the reactor tank and also be easily removed for inspection. Monte Carlo computations indicated that by using a 63 cm-thick A12O3 moderator, an epithermal neutron beam with an intensity of 0.34x109 n·cm−2·s−1 and fast neutron and γ-doses per epithermal neutron of 4.3 x 10−11 cGy·cm2·n −1 and 0.3x10−11 cGy·cm2·n −1, respectively, could be produced. The multiplication factor of the 43 spent fuel elements was <0.95 and negligible reactivity coupling to the core by the spent fuel was calculated. Such an epithermal neutron beam, if built into the Musashi reactor, would significantly Increase the chance or success of Boron Neutron Capture Therapy (BNCT) in Japan.

Design Studies of an Epithermal Neutron Beam for Neutron Capture Therapy at the Musashi Reactor.

Studies were carried out to design an epithermal neutron beam for neutron capture therapy at the Musashi reactor, a TRIGA-II of 100kW. The idea in this design is to use spent fuel elements as a converter assembly to convert thermal neutrons from the core to fission neutrons which would then be moderated to epithermal neutrons. In the design, 43 elements were placed outside of the graphite reflector but inside the tank. These elements could be cooled by water in the reactor tank and also be easily removed for inspection. Monte Carlo computations indicated that by using a 63 cm-thick A12O3 moderator, an epithermal neutron beam with an intensity of 0.34x109 n·cm−2·s−1 and fast neutron and γ-doses per epithermal neutron of 4.3 x 10−11 cGy·cm2·n −1 and 0.3x10−11 cGy·cm2·n −1, respectively, could be produced. The multiplication factor of the 43 spent fuel elements was <0.95 and negligible reactivity coupling to the core by the spent fuel was calculated. Such an epithermal neutron beam, if built into the Musashi...

武蔵工大炉の冷却水中の放射能 燃料要素被覆材の影響

The replacement of Al cladded fuel elements with stainless steel ones was undertaken in the Musashi Reactor (TRIGA-II, 100kW) in 1985. The change of radioactivity in the cooling water before and after the replacement has been discussed in this report.The radioactivity of 19 nuclides in the cooling water has been measured in the present study. The activity of 51Cr, 54Mn, 56Mn, 51Co, 60Co, 122Sb and 124Sb increased by about 100 to 10 times and 24Na, 72Ga, 99mTc and 187W decreased by about 0.1 times that before the replacement. The increase and decrease of their nuclides, except for 99mTc and 187W were related to the amount of each element, which produced the radioactive nuclides, in the clad of fuel elements. Fission products such as 131I, 132I, 133I, 134I, 135I and 132Te showed only slight decreases by about 0.5 times in spite of the large difference of U contents in the cladding material between new and old fuel elements.Release processes have been considered based on two mechanisms, chemical dissolution and direct recoil of activated atoms, using the radioactivity data.

Present status of the medical irradiation facility at the Musashi reactor.

Boron neutron capture therapy (BNCT) of malignant brain tumors has been efficiently performed since March 1977, and the first human case of malignant melanoma was also successfully treated on July 1987 in our reactor (Musashi reactor, TRIGA-II, 100 kW). To obtain both good irradiation field characteristics and a better irradiation facility, some tests and developments have been continued in accordance with the study of medical and biological irradiations. The results of these evaluations and a new approach are presented.

Three-Dimensional Neutron Streaming Analysis Using the Monte Carlo Coupling Technique in a Medical Irradiation Room of the Musashi Reactor: Comparison with Experiments

Measurements of thermal and nonthermal neutron streaming were taken throughout the medical irradiation room and the two-legged labyrinth of the Musashi Institute of Technology Research Reactor (the Musashi reactor) by using a rem counter. The length of the measured line was 8 m. The measurements were also analyzed by using the Monte Carlo coupling technique. The contribution of nonthermal neutrons was obtained with a cadmium-covered rem counter and that of thermal neutrons was obtained from the difference between the responses measured with and without the cadmium cover. The response ratio of total neutrons to nonthermal neutrons is constant for the straight part of the duct and increases rapidly around the bent portion. The constant values of the response ratio are 2, 3, and 5 for the first, second, and third legs, respectively. The value of 1.5 count/s = 1 mrem/h was used as the coefficient for conversion to the neutron dose rate. The discrepancies between the measured and calculated results are within about 50% for the nonthermal neutron response, and within a factor of 2 for total neutron response. The fractional standard deviations of the Monte Carlo calculations are 0.07 to 0.12 and 0.13 to 0.24 in the first leg,more » 0.07 to 0.18 and 0.13 to 0.44 in the second leg, and 0.12 to 0.38 and 0.17 to 0.56 in the third leg for nonthermal and total neutron dose rates, respectively.« less

わが国の大学における原子炉共同利用の最近の実態

In the present data, the actual status of the visiting researchers program on the Nuclear Reactor; i.e. Japan Material Test Reactor, YAYOI, JRR-2, -3, -4, Musashi Reactor, Rikkyo Reactor, KUR, KUCA; is reported from the individual facilities.