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Journal Articles

VHTR technology development in Japan; Progress of R&D activities for GIF VHTR system

Shibata, Taiju; Sato, Hiroyuki; Ueta, Shohei; Takegami, Hiroaki; Takada, Shoji; Kunitomi, Kazuhiko

2018 GIF Symposium Proceedings (Internet), p.99 - 106, 2020/05

no abstracts in English

Journal Articles

Conceptual plant system design study of an experimental HTGR upgraded from HTTR

Ohashi, Hirofumi; Goto, Minoru; Ueta, Shohei; Sato, Hiroyuki; Fukaya, Yuji; Kasahara, Seiji; Sasaki, Koei; Mizuta, Naoki; Yan, X.; Aoki, Takeshi*

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10

Conceptual design study of an experimental HTGR is performed to upgrade the plant system from Japanese High Temperature engineering Test Reactor (HTTR) to a commercial HTGR. Safety systems of HTTR are upgraded to demonstrate the commercial HTGR concept, such as a passive reactor cavity cooling system, a confinement, etc. An intermediate heat exchanger (IHX) is replaced by a steam generator (SG) for a process heat supply to demonstrate the technology for a commercial use. This paper describes the conceptual design study results of the plant system of the experimental HTGR.

Journal Articles

Beam commissioning of the linac for iBNCT

Naito, Fujio*; Anami, Shozo*; Ikegami, Kiyoshi*; Uota, Masahiko*; Ouchi, Toshikatsu*; Onishi, Takahiro*; Oba, Toshiyuki*; Obina, Takashi*; Kawamura, Masato*; Kumada, Hiroaki*; et al.

Proceedings of 13th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.1244 - 1246, 2016/11

The proton linac installed in the Ibaraki Neutron Medical Research Center is used for production of the intense neutron flux for the Boron Neutron Capture Therapy (BNCT). The linac consists of the 3-MeV RFQ and the 8-MeV DTL. Design average beam current is 10mA. Target is made of Beryllium. First neutron production from the Beryllium target was observed at the end of 2015 with the low intensity beam as a demonstration. After the observation of neutron production, a lot of improvement s was carried out in order to increase the proton beam intensity for the real beam commissioning. The beam commissioning has been started on May 2016. The status of the commissioning is summarized in this report.

Journal Articles

Characterization of high-energy quasi-monoenergetic neutron energy spectra and ambient dose equivalents of 80-389 MeV $$^{7}$$Li(p,n) reactions using a time-of-flight method

Iwamoto, Yosuke; Hagiwara, Masayuki*; Satoh, Daiki; Araki, Shohei*; Yashima, Hiroshi*; Sato, Tatsuhiko; Masuda, Akihiko*; Matsumoto, Tetsuro*; Nakao, Noriaki*; Shima, Tatsushi*; et al.

Nuclear Instruments and Methods in Physics Research A, 804, p.50 - 58, 2015/12

 Times Cited Count:17 Percentile:9.29(Instruments & Instrumentation)

We have measured neutron energy spectra for the 80, 100 and 296 MeV proton incident reactions at the RCNP cyclotron facility using time-of-flight method. The neutron energy spectrum consisted of the peak and continuum parts and the peak intensity was 0.9-1.1 $$times$$ 10$$^{10}$$ neutrons/sr/$$mu$$C. The ratio of peak intensity of the spectrum to the total intensity was between 0.38 and 0.48. To consider the correction required to derive a response in the peak region from the measured total response for neutron monitors, we proposed the subtraction method using energy spectra between 0$$^{circ}$$ and 25$$^{circ}$$. The normalizing factor k against the 25$$^{circ}$$ neutron fluence that equalizes the 0$$^{circ}$$ neutron fluence in the continuum region was from 0.74 to 1.02. With our previous results, we have obtained data for characterization of monoenergetic neutron field for the $$^{7}$$Li(p,n) reaction with 80$$sim$$389 MeV protons at the RCNP cyclotron facility.

JAEA Reports

Test plan using the HTTR for commercialization of GTHTR300C

Tachibana, Yukio; Nishihara, Tetsuo; Sakaba, Nariaki; Ohashi, Hirofumi; Sato, Hiroyuki; Ueta, Shohei; Aihara, Jun; Goto, Minoru; Sumita, Junya; Shibata, Taiju; et al.

JAEA-Technology 2009-063, 155 Pages, 2010/02

JAEA-Technology-2009-063.pdf:17.27MB

This report describes full scope of the feasible future test plan mainly using the HTTR. The test items cover fuel performance and radionuclide transport, core physics, reactor thermal hydraulics and plant dynamics, and reactor operations, maintenance, control, etc. The test results will be utilized for realization of Japan's commercial Very High Temperature Reactor (VHTR) system, GTHTR300C.

Journal Articles

Calculation of critical concentrations of actinides in an infinite medium of silicon dioxide

Okuno, Hiroshi; Sato, Shohei; Kawasaki, Hiromitsu*

Journal of Nuclear Science and Technology, 46(12), p.1137 - 1144, 2009/12

 Times Cited Count:2 Percentile:80.08(Nuclear Science & Technology)

The critical concentrations of metal-SiO$$_{2}$$ and -H$$_{2}$$O mixtures were calculated for 26 actinides including $$^{233, 235}$$U, $$^{239, 241}$$Pu, $$^{242m}$$Am, $$^{243, 245, 247}$$Cm and $$^{249, 251}$$Cf, where the critical concentration was defined as the concentration that the infinite neutron multiplication factor, k$$infty$$ being calculated to be 1.0. The calculations were performed using the Monte Carlo neutron transport calculation code MCNP5 combined with the evaluated nuclear data library JENDL3.3. The results showed that the critical actinide concentration of metal-SiO$$_{2}$$ was ca. 1/5 of that of metal-H$$_{2}$$O for all the fissile nuclides investigated. The k$$infty$$'s of the metal-SiO$$_{2}$$ and metal-H$$_{2}$$O at a half of the respective critical actinide concentration, which concentration was assumed as the subcritical actinide concentration limit, were found to be less than 0.8 for all the actinides considered. Applying a sum-of-fractions rule with respect to the ratios of actinide concentration to the subcritical actinide concentration limit for six fissile nuclides, subcriticality of high-level radioactive wastes was confirmed for a reported sample. The effect of different nuclear data libraries on the results of critical actinide concentrations was found large for $$^{242}$$Cm, $$^{247}$$Cm and $$^{250}$$Cf.

JAEA Reports

Calculation of the estimated criticality lower limit multiplication factor of MOX fuel systems based on the evaluation of calculation errors dependent on plutonium-240 isotopic fraction

Sato, Shohei; Okuno, Hiroshi

JAEA-Data/Code 2009-014, 19 Pages, 2009/11

JAEA-Data-Code-2009-014.pdf:3.03MB

The estimated criticality lower limit multiplication factor (hereafter, ECLLMF) is the upper limit of the neutron multiplication factor where the system may be judged subcritical through the calculation results of the same criticality calculation system applied to analogous fuel systems to be evaluated. Aiming to establish an effective method to find the rational ECLLMF of mixed uranium and plutonium oxide (MOX) fuel systems, this report investigated the classification of the critical experiments for the statistical processing, and evaluated the calculation errors with considering the dependence on $$^{240}$$Pu isotopic fraction within the classified experiments. In this evaluation, the criticality calculation code MVP and the evaluated nuclear data library JENDL-3.3 library were utilized, and the criticality experiments with MOX fuels registered in the international criticality safety benchmark evaluation project (ICSBEP) handbook were adopted. It was found that the dependency of the benchmark calculation results on the $$^{240}$$Pu isotopic fraction was enhanced by introducing a new fuel class: "dual heterogeneous fuel systems." As a result of this classification and error evaluation, it was confirmed that the calculated values of all the ECLLMFs were below the benchmark calculation results, and that the value of the ECLLMF was high compared with that obtained with the traditional method.

JAEA Reports

Calculation of the kinetic parameters for homogeneous fuel systems (MOX powder with zinc stearate and plutonium nitrate solution)

Sato, Shohei; Okuno, Hiroshi

JAEA-Data/Code 2009-006, 43 Pages, 2009/07

JAEA-Data-Code-2009-006.pdf:6.53MB

This report represents the kinetic parameters for homogeneous fuel systems obtained in the cooperative study with the Institut de Radioprotection et de Surete Nucleaire (IRSN) in France. The subject fuels for calculation are MOX powder mixed with zinc stearate and plutonium nitrate solution. The TWODANT code is utilized with 17 energy groups JENDL3.3 cross section collapsed by SRAC. As a result of the calculations, it was found that (1) The kinetic parameters of MOX powder is dependent on plutonium enrichment and the fraction of hydrogen, and is not dependent on the density of MOX powder and the fuel height except for the neutron lifetime, despite the kind of fuel system, (2) The kinetic parameters of plutonium nitrate solution depend on the concentration of plutonium; the temperature coefficient of which plutonium concentration is below 19g/l is positive.

Journal Articles

Fluctuation of the neutron multiplication factor induced by an oscillation of the fuel solution system

Sato, Shohei; Okuno, Hiroshi; Uchiyama, Gunzo

Journal of Nuclear Science and Technology, 46(3), p.268 - 277, 2009/03

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

This paper intends to figure out reactivity of the fuel solution system with a free surface. To fulfill this intension, criticality calculation with reflecting fluid calculation results have been carried out. For fluid calculation, the finite volume method and the VOF method are applied to track the free surface caused by an oscillation. For criticality calculation, we have applied the continuous energy Monte Carlo calculation method. As a result, three fluctuation types of $$k$$$$_{eff}$$ have been obtained depending on the oscillation frequency and the ratio of the solution height to the width of tank (H/L). If a sloshing motion is generated, $$k$$$$_{eff}$$fluctuates by a wide range and has a threshold, which can classify the fluctuation type of $$k$$$$_{eff}$$, despite the kind of the reflector. If H/L is above the threshold, ${it i.e.,}$ H/L=0.35, it fluctuates below the value of the static condition. The threshold value represented in this paper is smaller than that of the conventional one.

Journal Articles

Nuclear criticality safety evaluation of a mixture of MOX, UO$$_{2}$$ and additive in the most conservative concentration distribution

Okuno, Hiroshi; Sato, Shohei; Sakai, Tomohiro*; Uchiyama, Gunzo

Journal of Nuclear Science and Technology, 45(11), p.1108 - 1115, 2008/11

 Times Cited Count:1 Percentile:88.49(Nuclear Science & Technology)

For nuclear criticality safety evaluation of blenders at the mixed uranium-plutonium oxide (MOX) fuel plant, non-uniformity distributions of powders in three chemical components, i.e., MOX, uranium-dioxide (UO$$_{2}$$) and zinc-stearate, which is a fuel additive, should be taken into account. The model blender considered in this article contained a mixture of 33 wt% PuO$$_{2}$$-enriched MOX, depleted UO$$_{2}$$ and zinc-stearate in a shape of an upside-down truncated cone, which was surrounded by 30 cm-thick polyethylene. For a limitation of the number of calculation cases, the fissile plutonium mass of the mixture was fixed to 98 kg, and the total concentration of MOX and UO$$_{2}$$ was fixed to 4.0 g/cm$$^{3}$$. The most conservative fuel distribution in the aspect of nuclear criticality safety under these constraints was calculated with a two-dimensional optimum fuel distribution code OPT-TWO, so that the importance distribution of MOX and that of zinc-stearate should be individually flattened by conserving the mass of each component. The OPT-TWO calculation was followed by criticality calculation performed with the MCNP code to obtain the neutron multiplication factor of the fuel in the optimum fuel distribution. The most conservative fuel distribution obtained in this research was typically depicted as a shell of zinc-stearate embedded into the central MOX region surrounded by the peripheral UO$$_{2}$$ region. An increase in the neutron multiplication factor was found 25% at most; non-uniformity of plutonium enrichment concentration and that of zinc-stearate concentration contributed to it in almost equal and independent ways.

JAEA Reports

Examination on small-sized cogeneration HTGR for developing countries

Sakaba, Nariaki; Tachibana, Yukio; Shimakawa, Satoshi; Ohashi, Hirofumi; Sato, Hiroyuki; Yan, X.; Murakami, Tomoyuki; Ohashi, Kazutaka; Nakagawa, Shigeaki; Goto, Minoru; et al.

JAEA-Technology 2008-019, 57 Pages, 2008/03

JAEA-Technology-2008-019.pdf:8.59MB

The small-sized and safe cogeneration High Temperature Gas-cooled Reactor (HTGR) that can be used not only for electric power generation but also for hydrogen production and district heating is considered one of the most promising nuclear reactors for developing countries where sufficient infrastructure such as power grids is not provided. Thus, the small-sized cogeneration HTGR, named High Temperature Reactor 50-Cogeneration (HTR50C), was studied assuming that it should be constructed in developing countries. Specification, equipment configuration, etc. of the HTR50C were determined, and economical evaluation was made. As a result, it was shown that the HTR50C is economically competitive with small-sized light water reactors.

Journal Articles

Calculation of the pressure vessel failure fraction of fuel particle of gas turbine high temperature reactor 300C

Aihara, Jun; Ueta, Shohei; Mozumi, Yasuhiro; Sato, Hiroyuki; Motohashi, Yoshinobu*; Sawa, Kazuhiro

Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.416 - 422, 2007/09

In high temperature gas-cooled reactors (HTGRs), coated particles are used as fuels. For upgrading HTGR technologies, present SiC coating layer which is used as the 3rd layer could be replaced with ZrC coating layer which have much higher temperature stability in addition to higher resistance to chemical attack by fission product palladium than the SiC coating layer. The ZrC layer could deform plastically at high temperatures. Therefore, the Japan Atomic Energy Agency modified an existing pressure vessel failure fraction calculation code to treat the plastic deformation of the 3rd layer in order to predict failure fraction of ZrC coated particle under irradiation. Finite element method is employed to calculate the stress in each coating layer. The pressure vessel failure fraction of the coated fuel particles under normal operating condition of GTHTR300C is calculated by the modified code. The failure fraction is evaluated as low as 3.5$$times$$10$$^{-6}$$.

JAEA Reports

OPT-TWO; Calculation code for two-dimensional MOX fuel models in the optimum concentration distribution

Sato, Shohei; Sakai, Tomohiro*; Okuno, Hiroshi

JAEA-Data/Code 2007-017, 40 Pages, 2007/08

JAEA-Data-Code-2007-017.pdf:4.8MB

OPT-TWO is a calculation code which calculates the optimum concentration distribution, i.e., the most conservative concentration distribution in the aspect of nuclear criticality safety, of MOX (mixed uranium and plutonium oxide) fuels in the two-dimensional system. To achieve the optimum concentration distribution, we apply the principle of flattened fuel importance distribution with which the fuel system has the highest reactivity. Based on this principle, OPT-TWO takes the following 3 calculation steps iteratively to achieve the optimum concentration distribution with flattened fuel importance: (1) the forward and adjoint neutron fluxes, and the neutron multiplication factor, with TWOTRAN code which is a two-dimensional neutron transport code based on the SN method, (2) the fuel importance, and (3) the quantity of the transferring fuel. In OPT-TWO, the components of MOX fuel are MOX powder, uranium dioxide powder and additive. This report describes the content of the calculation, the computational method, and the installation method of the OPT-TWO, and also describes the application method of the criticality calculation of OPT-TWO.

Journal Articles

Thermodynamics of the UO$$_{2}$$ solid solution with magnesium and europium oxides

Fujino, Takeo*; Sato, Nobuaki*; Yamada, Kota*; Nakama, Shohei*; Fukuda, Kosaku; Serizawa, Hiroyuki; Shiratori, Tetsuo*

Journal of Nuclear Materials, 297(3), p.332 - 340, 2001/09

 Times Cited Count:3 Percentile:71.33(Materials Science, Multidisciplinary)

Oxygen potential of solid solution Mg$$_{y}$$Eu$$_{z}$$U$$_{1-y-z}$$O$$_{2+x}$$ was examined as a function of O/Metal ratio at 1000, 1100 and 1200 $$^{o}$$C. The O/Metal ratio which gave the steepest change of the oxygen potential(GOM) was 1.995 for y=0.05, z=0.1 and y=0.05, z=0.05. The position decreased to 1.908 at higher Mg$$^{2+}$$ concentration of y=0.1, z=0.05. The GOM did not change with temperature in a range 1000-1200 $$^{o}$$C. At GOM, a friction of 0.473 of total Mg$$^{2+}$$ occupy the interstitial position of the fluorite lattice.

Oral presentation

Calculation of critical and subcritical actinide concentrations in the infinite sea of silicon-dioxide or water

Okuno, Hiroshi; Sato, Shohei; Kawasaki, Hiromitsu*

no journal, , 

no abstracts in English

Oral presentation

Monte Carlo criticality analysis of solution fuel system with free surface

Sato, Shohei; Okuno, Hiroshi

no journal, , 

no abstracts in English

Oral presentation

Nuclear criticality calculations in considering non-uniformity of MOX, UO$$_{2}$$ and additive, 1; Background, formulation and calculation model

Okuno, Hiroshi; Sato, Shohei; Uchiyama, Gunzo; Sakai, Tomohiro*

no journal, , 

no abstracts in English

Oral presentation

Nuclear criticality caluculations in considering non-uniformity of MOX, UO$$_{2}$$ and additive, 2; Calculation code and results, and further investigation

Sato, Shohei; Okuno, Hiroshi; Uchiyama, Gunzo; Sakai, Tomohiro*

no journal, , 

no abstracts in English

Oral presentation

Improvement in error evaluation method of nuclear criticality safety calculation; Applicability to uranium fuel systems and mixed-oxide fuel systems

Okuno, Hiroshi; Sato, Shohei

no journal, , 

A formulation was investigated when errors in criticality calculation depend on parameters. Examples is shown when the errors can be expressed up to the 2nd order of uranium enrichment for 5 - 10wt% enriched uranium fuel systems. Parameter dependences are also examined for MOX fuel systems.

Oral presentation

Fluctuation of the neutron multiplication factor caused by the sloshing motion of the fuel solution system

Sato, Shohei; Okuno, Hiroshi

no journal, , 

The deformation of the fuel solution causes the reactivity. From the viewpoint of nuclear safety, it is important to comprehend the feature of this effect. Before now, it is found that the negative reactivity is induced, if the height of the fuel solution, which is more 0.5 is deformed. In the last report, we executed the combination computing between the fluid calculation and the criticality calculation, and calculated the time change of the neutron multiplication factor of the oscillated fuel solution. In this report, we investigate the relativity of the solution height for the fluctuation of the neutron multiplication factor using the same method. The fluid calculation is executed to calculate the fluctuation of the fuel solution contained in the slab tank oscillated with following the sine wave. The criticality calculation is executed to calculate the neutron multiplication factor deformed by the oscillation obtained from the fluid calculation. As the result, we found that there is a threshold, which can classify the fluctuation pattern of the neutron multiplication factor and the threshold is 0.4 which is expressed by the ratio of the solution height and the width of the tank. Also, we found that the threshold is smaller than before value.

28 (Records 1-20 displayed on this page)