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Journal Articles

Generalized formulation of extended cross-section adjustment method based on minimum variance unbiased linear estimation

Yokoyama, Kenji; Kitada, Takanori*

Journal of Nuclear Science and Technology, 56(1), p.87 - 104, 2019/01

 Times Cited Count:1 Percentile:50.57(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Study on heterogeneous minor actinide loading fast reactor core concepts with improved safety

Ohgama, Kazuya; Oki, Shigeo; Kitada, Takanori*; Takeda, Toshikazu*

Proceedings of 21st Pacific Basin Nuclear Conference (PBNC 2018) (USB Flash Drive), p.942 - 947, 2018/09

Journal Articles

Comparative study on prediction accuracy improvement methods with the use of integral experiments for neutronic characteristics of fast reactors

Yokoyama, Kenji; Kitada, Takanori*

Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.1221 - 1230, 2018/04

no abstracts in English

Journal Articles

Dimension-reduced cross-section adjustment method based on minimum variance unbiased estimation

Yokoyama, Kenji; Yamamoto, Akio*; Kitada, Takanori*

Journal of Nuclear Science and Technology, 55(3), p.319 - 334, 2018/03

 Times Cited Count:2 Percentile:50.63(Nuclear Science & Technology)

A new formulation of the cross-section adjustment methodology with the dimensionality reduction technique has been derived. This new formulation is proposed as the dimension reduced cross-section adjustment method (DRCA). Since the derivation of DRCA is based on the minimum variance unbiased estimation (MVUE), an assumption of normal distribution is not required. The result of DRCA depends on a user-defined matrix that determines the dimension reduced feature subspace. We have examine three variations of DRCA, namely DRCA1, DRCA2, and DRCA3. Mathematical investigation and numerical verification have revealed that DRCA2 is equivalent to the currently widely used cross-section adjustment method. Moreover, DRCA3 is found to be identical to the cross-section adjustment method based on MVUE, which has been proposed in the previous study.

Journal Articles

Azimuthal flux distribution measurements around fuel rods in reduced-moderation LWR lattices

Yoshioka, Kenichi*; Kitada, Takanori*; Nagaya, Yasunobu

Progress in Nuclear Energy, 82, p.7 - 15, 2015/07

 Times Cited Count:1 Percentile:85.75(Nuclear Science & Technology)

A reduced-moderation LWR has been developed for the reduction of spent fuel and for the efficient utilization of uranium resources. The streaming channel concept to improve the negative void reactivity coefficient is one of the features of the reactor. This concept makes the fuel assembly more heterogeneous. The geometrical heterogeneity makes azimuthal neutron flux distribution of fuel rods steep. To validate azimuthal neutron flux distribution calculation, we measured the distribution around fuel rods in reduced moderation LWR lattices. These measurements were conducted in NCA with the foil activation method. The core consisted of the central triangular tight lattice zone and the outer driver zone of a square lattice. A pile of polystyrene plates for simulating void fraction was installed into the triangular tight lattice. Azimuthal neutron flux distributions were deduced from the activity of these small foils measured with plastic scintillators. Measurements were compared to calculations by the MVP code with JENDL-3.3. It was found that calculations agreed well with measurements.

Journal Articles

Void reactivity evaluation by modified conversion ratio measurements in LWR critical experiments

Yoshioka, Kenichi*; Kikuchi, Tsukasa*; Gunji, Satoshi*; Kumanomido, Hironori*; Mitsuhashi, Ishi*; Umano, Takuya*; Yamaoka, Mitsuaki*; Okajima, Shigeaki; Fukushima, Masahiro; Nagaya, Yasunobu; et al.

Journal of Nuclear Science and Technology, 52(2), p.282 - 293, 2015/02

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

We have developed a void reactivity evaluation method by using modified conversion ratio measurements in a light water reactor (LWR) critical lattice. Assembly-wise void reactivity is evaluated from the "finite neutron multiplication factor", $$k^ast$$, deduced from the modified conversion ratio of each fuel rod. The distributions of modified conversion ratio and $$k^ast$$ on a reduced-moderation LWR lattice, for which the improvement of negative void reactivity is a serious issue, were measured. Measured values were analyzed with a continuous-energy Monte Carlo method. The measurements and analyses agreed within the measurement uncertainty. The developed method is useful for validating the nuclear design methodology concerning void reactivity.

Journal Articles

Intra-pellet neutron flux distribution measurements in LWR critical lattices

Yoshioka, Kenichi*; Kikuchi, Tsukasa*; Gunji, Satoshi*; Kumanomido, Hironori*; Mitsuhashi, Ishi*; Umano, Takuya*; Yamaoka, Mitsuaki*; Okajima, Shigeaki; Fukushima, Masahiro; Nagaya, Yasunobu; et al.

Journal of Nuclear Science and Technology, 50(6), p.606 - 614, 2013/06

 Times Cited Count:1 Percentile:86.44(Nuclear Science & Technology)

We have developed an intra-pellet neutron flux and conversion ratio distribution measurement method. A foil activation method with special foils was used for the neutron flux distribution measurement. A $$gamma$$-ray spectrum analysis method with special collimators was used for the conversion ratio distribution measurement. Using the developed methods, intra-pellet neutron flux distributions and conversion ratio distributions were measured in critical experiments on a reduced-moderation LWR. Measured values were analyzed with a deterministic method and a Monte Carlo method. The neutron flux distribution measurements and analyses agreed within the range of 1% to 2%. The conversion ratio distribution measurements and analyses were consistent with each other. We found that the measurement methods are useful for the validation of neutron behavior in a fuel pellet, which is known as micro reactor physics.

Journal Articles

Development of the 4S and related technologies, 7; Summary of the FCA XXIII experiment analyses towards evaluation of prediction accuracies for the 4S core characteristics

Ueda, Nobuyuki*; Fukushima, Masahiro; Okajima, Shigeaki; Takeda, Toshikazu*; Kitada, Takanori*; Nauchi, Yasushi*; Kinoshita, Izumi*; Matsumura, Tetsuo*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9493_1 - 9493_9, 2009/05

A series of critical experiments were carried out in the JAEA fast critical facility (FCA) named FCA XXIII cores with placing emphases on the reflector reactivity worth and the sodium void reactivity which are especially important from the view point of safety features of the 4S. The analyses of those physics mockup experiments have been carried out by the neutron transport calculation methods with continuous energy Monte Carlo code MVP and 70 energy-group discrete ordinate P0-S8 transport code DANTSYS using libraries processed from JENDL-3.3 data file. The results showed that combination of the stochastic and deterministic transport calculation methods (Monte Carlo and Sn) provided good prediction bases for criticality, reflector worth, sodium void reactivity, reaction rate ratios and absorber reactivity worth for the 4S nuclear design.

Journal Articles

Prediction accuracy improvement of neutronic characteristics of a breeding light water reactor core by extended bias factor methods with use of FCA-XXII-1 critical experiments

Kugo, Teruhiko; Ando, Masaki; Kojima, Kensuke; Fukushima, Masahiro; Mori, Takamasa; Nakano, Yoshihiro; Okajima, Shigeaki; Kitada, Takanori*; Takeda, Toshikazu*

Journal of Nuclear Science and Technology, 45(4), p.288 - 303, 2008/04

 Times Cited Count:6 Percentile:54.81(Nuclear Science & Technology)

The effectiveness of the extended bias factor methods, the LC and PE methods, is numerically investigated by applying them to a breeding light water reactor core as a target core with use of FCA-XXII-1 critical experiments. The present study numerically verifies the features of the extended bias factor methods. Both the methods can improve the prediction accuracy the most by using all the experiments. The PE method always improves the prediction accuracy with any combination of experiments. The PE method is always superior to the LC method for improvement of the prediction accuracy. From the present study, the followings are found. The experiments on multiplication factor are more applicable to a reaction rate ratio of $$^{238}$$U capture to $$^{239}$$Pu fission (C28/F49) of the target core than the experiments on C28/F49. Combinations of the experiments on multiplication factor is more effective to a void reactivity of the target core than those of the experiments on void reactivity though those on void reactivity are superior to those on multiplication factors in the case of using a single experiment. From these results, we conclude that the experiments on multiplication factor are more effective than the other experiments for all the neutronic characteristics of the target core. From these results, it is concluded that the PE method is promising to complement full mockup experiments for various future nuclear systems by using a number of existing and future benchmark experiments.

Journal Articles

Development of 3-D detailed FBR core calculation method based on method of characteristics

Takeda, Toshikazu*; Imai, Hideki*; Kitada, Takanori*; Nishi, Hiroshi; Ishibashi, Junichi; Kitano, Akihiro

Proceedings of International Topical Meeting on Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications (M&C 2005) (CD-ROM), 12 Pages, 2005/09

A new detailed 3-D transport calculation method taking into account the heterogeneity of fuel assemblies has been developed in hexagonal-z geometry by combining the method of characteristics and the nodal transport method. From the nodal transport calculation which uses assembly homogenized cross sections, the axial leakage is calculated, and it is used for the MOC calculation which treats the heterogeneity of fuel assemblies. Series of homogeneous MOC calculations which use assembly homogeneous cross sections are carried out of obtain effective cross sections, which preserve assembly reaction rates. This effective cross sections are again used in the 3-D nodal transport calculations. The numerical calculations have been performed to verify 3-D radial calculations of FBR assemblies and partial core calculations. Results are compared with the reference Monte-Carlo calculations. A good agreement has been achieved. It is shown that the present method has an advantage in calculating reaction tates in small region.

Journal Articles

Preliminary evaluation of reduction of prediction error in breeding light water reactor core performance

Kugo, Teruhiko; Kojima, Kensuke; Ando, Masaki; Okajima, Shigeaki; Mori, Takamasa; Takeda, Toshikazu*; Kitada, Takanori*; Matsuoka, Shogo*

Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 10 Pages, 2005/05

We have preliminarily evaluated the reduction of prediction errors of the core characteristics of the breeding light water reactor core based on the bias factor method by utilizing the FCA critical experiments carried out for MOX fueled tight lattice light water reactor cores. The prediction uncertainty of k$$_{eff}$$ is reduced from 0.62% to 0.39% by utilizing the FCA-XV-2 (65V) result. As for the reaction rate ratio of $$^{238}$$U capture and $$^{239}$$Pu fission, it is found that the FCA XXII-1 (95V) and XV (95V) results are suitable for the upper core and the upper blanket of the real core and the FCA XXII-1 (65V) and XV-2 (65V) results are suitable for the lower core and the internal blanket.

Journal Articles

Analysis of benchmark results for reactor physics of LWR next generation fuels

Kitada, Takanori*; Okumura, Keisuke; Unesaki, Hironobu*; Saji, Etsuro*

Proceedings of International Conference on Physics of Fuel Cycles and Advanced Nuclear Systems; Global Developments (PHYSOR 2004) (CD-ROM), 8 Pages, 2004/04

Burnup calculation benchmark has been carried out for the LWR next generation fuels aiming at high burnup up to 70 GWd/t with UO$$_{2}$$ and MOX. Based on the submitted results by many benchmark participants, the present status of calculation accuracy has been confirmed for reactor physics parameters of the LWR next generation fuels, and the factors causing the calculation differences were analyzed in detail. Moreover, the future experiments and research subjects necessary to reduce the calculation differences were discussed and proposed.

Journal Articles

Update status of benchmark activity for reactor physics study of LWR next generation fuels

Unesaki, Hironobu*; Okumura, Keisuke; Kitada, Takanori*; Saji, Etsuro*

Transactions of the American Nuclear Society, 88, p.436 - 438, 2003/06

In order to investigate the calculation accuracy of the nuclear characteristics of LWR next generation fuels, the Research Committee on Reactor Physics organized by JAERI has proposed "Reactor Physics Benchmark for LWR Next Generation Fuels". The next generation fuels aim at very high burn-up of about 70GWd/t in PWR or BWR with UO$$_{2}$$ or MOX fuels whose fissile enrichments may exceed the Japanese regulatory limitations for the current LWR fuels such as 5wt.% U-235. Until now, twelve organizations have pariticipated in the benchmark activity. From the comparison with the cell burn-up calculation results using different codes and library data, status of the calculation accuracy and future subjects are clarified.

Journal Articles

3D Transport Theory Method Based on MOC for Analyzing Integral Dta of Transmutation

Takeda, Toshikazu*; Hamada, Yuzuru*; Kitada, Takanori*; Nishi, Hiroshi; Ishibashi, Junichi; Kitano, Akihiro

Proceedings of International Conference on Advanced Nuclear Energy and Fuel Cycle Systems (GLOBAL 2003) (CD-ROM), p.1005 - 1010, 2003/00

A new 3-D transport calculation method taking into account the heterogeneity of fuel assemblies has been developed by combining the characteristics method and the nodal transport method. In the axial direction the nodal transport method is applied, and the characteristics method is applied to take into account the radial heterogeneity of fuel assemblies. The numerical calculations have been performed to verify 2-D radial calculations of FBR assemblies and partial core calculations. Results are compared with the reference Monte-Carlo calculations. A good agreement has been achieved. It is shown that the present method has an advantage in calculating reaction rates in a small region.

Journal Articles

Benchmark results of burn-up calculation for LWR next generation fuels

Okumura, Keisuke; Unesaki, Hironobu*; Kitada, Takanori*; Saji, Etsuro*

Proceedings of International Conference on the New Frontiers of Nuclear Technology; Reactor Physics, Safety and High-Performance Computing (PHYSOR 2002) (CD-ROM), 10 Pages, 2002/10

In order to investigate the calculation accuracy of the nuclear characteristics of LWR next generation fuels, the Research Committee on Reactor Physics organized by Japan Atomic Energy Research Institute has proposed "Reactor Physics Benchmark for LWR Next Generation Fuels". The next generation fuels aim at very high burn-up of about 70GWd/t in PWR or BWR with UO2 or MOX fuels whose fissile enrichments may exceed the Japanese regulatory limitations for the current LWR fuels such as 5wt.% U-235. Twelve organizations have carried out the analyses of the benchmark problems with different codes and data, and their submitted results have been compared. As a result, status of accuracy with the current data and method and some problems to be solved in the future were clarified.

JAEA Reports

Study on improvement of reactor physics analysis method for FBRs with various core concept (2)

Takeda, Toshikazu*; Tagawa, Akihiro*; *; Kitada, Takanori*; *

JNC-TJ9400 2001-009, 239 Pages, 2001/02

JNC-TJ9400-2001-009.pdf:8.71MB

Investigation was made on the following three themes as a part of the improvement of reactor physics analysis method for FBR with various core concepts. [Part 1: Investigations on Improvement of Neutron Spectrum Evaluation by the Use of Co-variance Matrices and Bias Corrections] In order to improve the neutron spectrum unfolding method used in the experimental fast reactor JOYO, investigation was made on the bias corrections to the initial neutron spectrum and error evaluation of nuclear data with the co-variance matrices. The error estimation was done by accumulating each bias correction factor and the co-variance matrix. It was concluded that the accumulated error for the initial neutron spectrum is relatively small, and a considerable improvement was achieved by the use of bias corrections. [Part 2: Evaluation of Neutron Streaming in Gas Cooled Fast Reactors by the use of Monte Carlo Method] As a part of investigations on the evaluation of the anisotropic diffusion coefficients for gas cooled fast reactors, a new tally function was added to a Monte Carlo code so that the neutron streaming can be calculated with heterogeneous core configurations. It was found that the neutron streaming becomes larger when the heterogeneous model was used. The tendency was more distinct in lower energy range. The same types of comparison was also done for the difference of core calculation models and the transport/diffusion theory. The final result shows that the transport/diffusion error has positive values in higher energy range, and the heterogeneous/homogeneous error has negative values in lower energy range. [Part 3: Investigation on the Calculation Method for Nuclear Converters with Neutron Moderators] A new calculation system which can deals with the target assemblies with neutron moderators was proposed. This concept has been investigated as a device to achieve high conversion rate for long life fission products. It was concluded that the characteristics method is ideal, wh

JAEA Reports

Study on improvement of reactor physics analysis method for FBRs with various core concept

*; Kitada, Takanori*; Tagawa, Akihiro*; *; Takeda, Toshikazu*

JNC-TJ9400 2000-006, 272 Pages, 2000/02

JNC-TJ9400-2000-006.pdf:9.69MB

Investigation was made on the follwing three themes as a part of the improvement of reactor physics analysis method for FBR with various core concept. Part 1: Investigation of Error Estimation of Neutron Spectra in FBR and Suggestions to Improve the Accuracy. In order to improve the spectrum unfolding method used in fast experimental reactor JOYO, a trial was made to evaluate the error in the estimated neutron spectrum, cause by cause. And the evaluated errors were summed up to obtain the most probable and reasonable error as possible. The summed up error was found relatively small compared to the error caused by the uncertainty of cross section data: most of the error in the spectrum unfolding method can be attributed to the error in cross sections. It was also found that the error due to the fission spectrum causes a considerable error in the high energy neutron spectrum which is over several MeV. Part 2: Study on Reactor Physics Analysis Method for Gas-Cooled FBR. In gas-cooled FBR, the portion of coolant channels in core volume is larger than sodium-cooled FBR. This leads to strong neutron streaming effects. For sodium-cooled FBR, several methods were proposed to evaluate the neutron streaming effect, however, these methods can not be used directly to gas-cooled reactor because the direction dependent diffusion coefficient becomes infinitive along the direction pararel to the coolant chammel. In this study, a new method is proposed to evaluate the neutron streaming effect, based on the method taking the axial buckling into consideration, which method was originally proposed by K$"o$hler. Part 3: Study on Reactor Physics Analysis Method for Water-Cooled FBR An investigation was made on low-moderated water-cooled FBR, on the point that the ordinary used analysis method for FBR may give considerable difference in results in such core. In light water reactors, it is well known that the space dependence of self-shielding effect of heavy nuclides are considerably ...

JAEA Reports

Improvement of numerical analysis method for FBR core characteristics (IV)

Takeda, Toshikazu*; *; Kitada, Takanori*

JNC-TJ9400 99-002, 171 Pages, 1999/03

JNC-TJ9400-99-002.pdf:4.44MB

Investigation was made on the following three themes as a part of the improvement of numerical analysis method for FBR core characteristics. [Part 1: Improvement of Reaction Rate Calculation Method in Blanket Region.] A method to calculate multiband parameters directly from the precise energy structure of a cross section was established. This method can treat the precise neutron balance equation including the inter-band scattering. The procedure to treat the multiband effect with the conventional Sn code by the use of direction-dependent microscopic cross sections is shown. This procedure was applied to reaction rate distribution analysis on MONJU. As the result, the reaction rates increased in the blanket regions: the maximum increase was 1O% for U-238 cap, 12% for Pu-239 fis, 12% for U-235 fis, and 1% for U-238 fis. It became clear that the effect of the use of direction-dependent microscopic cross sections is small; thus most of the effect can be attributed to the change of microscopic cross section itself. [Part 2: Improvement of Reactivity Calculation by the Use of Monte Carlo Perturbation Method.] Applicability of the two Monte Carlo perturbation methods; the correlated sampling method and the derivative operator sampling method, to the perturbation problems with large change in material densities was investigated through numerical calculations. The result shows the fact that the present derivative operator sampling method, which treats only the first order term, is insufficient and higher order terms should be considered for such a large perturbation. Investigation was made on the new method which can treat the effect of adjoint flux change in energy and space distribution due to the perturbation, and a new formulation based on the exactperturbation was derived. [Part 3: Investigation of Error Estimation of Neutron Spectra in FBR and Suggestions to Improve the Accuracy.] A comparison was made between the two methods to improve the accuracy ...

JAEA Reports

Improvement of numerical analysis method for FBR core characteristics (III)

Takeda, Toshikazu*; Kitada, Takanori*; *; *

PNC-TJ9605 98-001, 267 Pages, 1998/03

PNC-TJ9605-98-001.pdf:11.65MB

As the improvement of numerical analysis method for FBR core characteristics, studies on several topics have been conducted; multiband method, Monte Carlo perturbation and nodal transport method. This report is composed of the following three parts. Part 1: Improvement of Reaction Rate Calculation Method in the Blanket Region Based on the Multiband Method. A method was developed for precise evaluation of the reaction rate distribution in the blanket region using the multiband method. With the 3-band parameters obtained from the ordinary fitting method, major reaction rates such as U-238 capture, U-235 fission, Pu-239 fission and U-238 fission rate distributions were analyzed. As for the nuclides to be analyzed, the elements of structure material, such as iron, nickel, chrome and sodium were considered. By the present method, all the reactions became larger at the deep region in the blanket. The maximum correction amounted as much as 5%. This tendency lessen the disagreement between the ordinary calculation and the experiment. It was made clear that the treatment in inter-band scattering term is veryimportant because it has large sensitivity on the result. An alternative method to determine the multiband parameters whieh method is based on more direct approach and is free from drawbacks in the present method, was also investigated. Part 2 : Improvement of Estimation Method for Reactivity Based on Monte-Carlo Perturbation Theory. Perturbation theory based on Monte-Carlo perturbation theory have been investigated and introduced into the calculational code. The continuous energy Monte-Carlo perturbation code has been developed by using not only the correlated sampling method which is already used before, but also the derivative operator sampling method. The Monte-Carlo perturbation code was applied to MONJU core and the calculational results were compared to the reference. The change of eigenvalue caused by the change of sodium density in the GEM or dummy ...

JAEA Reports

Improvement of numerical analysis method for FBR core characteristics (II)

Takeda, Toshikazu*; *; Kitada, Takanori*; *

PNC-TJ9605 97-001, 100 Pages, 1997/03

PNC-TJ9605-97-001.pdf:2.82MB

This report is composed of the following two parts and appendix. (I)Improvement of the Method for Evaluating Reactivity Based on Monte Carlo Perturbation Theory (II)Improvement of Nodal Transport Method for 3-D Hexagonal Geometry (Appendix) Effective Cross Section of $$^{238}$$U Samples for Analyzing Doppler Reactivity in Fast Reactors Part I. Improvement of the Method for Evaluating Reactivity Bascd on Monte Carlo Perturbation Theory. Theoretical formulation in Monte Carlo perturbation method had been checked, and then introduced into a calculation code. The increase of CPU time is about 10 to 20 % compared to that if normal Monte Carlo code, in the cases of same number of history. This Monte Carlo perturbation method found to be effective, because results are almost reasonable and deviations of the results are especially small, by using the Monte Carlo perturbation code. However, there are somc cases that the results of the change of eigenvalues becomes positive or negative by changing the estimator, and there is no reasonable difference in the results between the conventional method, which does not consider the change of neutron source distribution caused by a perturbation, and the new method, which consider that change. Thus it is still necessary to check the Monte Carlo pcrturbation code. Part II. Improvement of Nodal Transport Method for 3-D Hexagonal Geometry It is certain that we can accurately evaluate hexagonal geometry FBR core by nodal transport calculation code for hexagonal-Z geometry named 'NSHEX'. However it is also found that in very heterogeneous core the results is not good enough. Because the treatment of the transverse leakage to the radial direction, which is use for evaluating intra-nodal flux distribution, is not so accurate. For the treatment of the leakage distribution, it is necessary to estimate the nodal vertex flux. In conventional method, the vertex flux estimated by the surrounding node surfacc flux around that vertex. On the contrary,

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