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Okuno, Hiroshi; Sato, Shohei; Kawasaki, Hiromitsu*
Journal of Nuclear Science and Technology, 46(12), p.1137 - 1144, 2009/12
Times Cited Count：3 Percentile：24.47(Nuclear Science & Technology)The critical concentrations of metal-SiO
and -HO mixtures were calculated for 26 actinides including 
U, 
Pu, 
Am, 
Cm and 
Cf, where the critical concentration was defined as the concentration that the infinite neutron multiplication factor, k
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 was ca. 1/5 of that of metal-HO for all the fissile nuclides investigated. The k's of the metal-SiO and metal-HO 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 
Cm, 
Cm and 
Cf.
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 
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 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.
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.
Sato, Shohei; Okuno, Hiroshi; Uchiyama, Gunzo
Journal of Nuclear Science and Technology, 46(3), p.268 - 277, 2009/03
Times Cited Count：0 Percentile：0.01(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 
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, fluctuates by a wide range and has a threshold, which can classify the fluctuation type of , despite the kind of the reflector. If H/L is above the threshold, 
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.
and additive in the most conservative concentration distributionOkuno, Hiroshi; Sato, Shohei; Sakai, Tomohiro*; Uchiyama, Gunzo
Journal of Nuclear Science and Technology, 45(11), p.1108 - 1115, 2008/11
Times Cited Count：2 Percentile：16.95(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) 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-enriched MOX, depleted UO 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 was fixed to 4.0 g/cm
. 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 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.
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.
Okuno, Hiroshi; Sato, Shohei; Kawasaki, Hiromitsu*
no journal, ,
no abstracts in English
Sato, Shohei; Okuno, Hiroshi
no journal, ,
no abstracts in English
and additive, 1; Background, formulation and calculation modelOkuno, Hiroshi; Sato, Shohei; Uchiyama, Gunzo; Sakai, Tomohiro*
no journal, ,
no abstracts in English
and additive, 2; Calculation code and results, and further investigationSato, Shohei; Okuno, Hiroshi; Uchiyama, Gunzo; Sakai, Tomohiro*
no journal, ,
no abstracts in English
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.
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.
Sato, Shohei; Okuno, Hiroshi
no journal, ,
From a viewpoint of nuclear criticality safety evaluation, it is important to comprehend temperature reactivity coefficient. In this paper, we calculate temperature reactivity coefficient of MOX fuel by fitting a quadric curve in the changing temperature system. As a result, it is indicated that temperature reactivity coefficient of MOX fuel is negative, and decreases with H/(U+Pu). Temperature reactivity coefficient does not depend on MOX density.
Okuno, Hiroshi; Sato, Shohei; Kawasaki, Hiromitsu*
no journal, ,
The critical masses of 23 minor actinides in metal and 7 minor actinides in metal-water mixtures were calculated using a combination of the Monte Carlo neutron transport code MCNP4C2 and the nuclear data library JENDL-3.3. The selected reflector conditions were bare, with a 30-cm-thick water and with a 30-cm-thick SS304. Comparisons of the calculated critical masses with the data in reference literature and those calculated using JENDL-3.2 were made. The effect of reflectors on the critical masses and the relations between the critical mass of actinide in metal and its neutron multiplication factor in the infinite media were discussed.
Matsumura, Tadayuki; Sumi, Hirotaka; Tokoro, Takeshi; Yamauchi, Sho; Sato, Shohei; Kano, Shigeru; Morikawa, Yo
no journal, ,
no abstracts in English
