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Journal Articles

Development of FRENDY nuclear data processing code; Generation capability of multi-group cross sections from ACE file

Yamamoto, Akio*; Endo, Tomohiro*; Tada, Kenichi

Transactions of the American Nuclear Society, 122(1), p.714 - 717, 2020/06

A generation capability of multi-group cross sections from point-wise cross sections in ACE files is being developed as a function of the nuclear data processing code FRENDY. This presentation describes features of this function and comparison of the processing results between this function and GROUPR module in NJOY.

Journal Articles

Implementation of random sampling for ACE-format cross sections using FRENDY and application to uncertainty reduction

Kondo, Ryoichi*; Endo, Tomohiro*; Yamamoto, Akio*; Tada, Kenichi

Proceedings of International Conference on Mathematics and Computational Methods applied to Nuclear Science and Engineering (M&C 2019) (CD-ROM), p.1493 - 1502, 2019/00

A perturbation capability of ACE formatted cross section files was developed using the modules of FRENDY. Uncertainty quantification using MCNP was carried out for the Godiva critical experiment by the RS method. We verified the results of the RS method by comparing with those obtained by the conventional sensitivity analyses. Moreover, uncertainty reduction using the bias factor method with the RS technique was applied to kinetic parameter, i.e., neutron generation time.

Journal Articles

Estimation of sensitivity coefficient based on lasso-type penalized linear regression

Katano, Ryota; Endo, Tomohiro*; Yamamoto, Akio*; Tsujimoto, Kazufumi

Journal of Nuclear Science and Technology, 55(10), p.1099 - 1109, 2018/10

 Times Cited Count:2 Percentile:33.17(Nuclear Science & Technology)

In this study, we propose the penalized regression "adaptive smooth-lasso" for the estimation of sensitivity coefficients of the neutronics parameters. The proposed method estimates the sensitivity coefficients of the neutronics parameters using the variation of the microscopic cross sections and the neutronics parameter obtained by random sampling. The proposed method utilizes only the forward calculations. Thus, the proposed method can be applied for the complex reactor analysis for which the application of the adjoint method is difficult. In this study, we proposed a penalty term considering the characteristics of the sensitivity coefficients of the neutronics parameter to the microscopic multi-group cross sections. Through verification calculation, we show that the proposed method achieves high accuracy with less computational cost compared to the method based on random sampling proposed in the previous studies.

Journal Articles

Sensitivity analysis for neutron multiplication parameters of accelerator driven subcritical system

Chiba, Go; Nishihara, Kenji; Endo, Tomohiro*

Proceedings of International Conference on Mathematics and Computational Methods applied to Nuclear Science and Engineering (MC 2011) (CD-ROM), 12 Pages, 2011/05

Sensitivity calculations are carried out for some neutronic parameters on neutron multiplication in subcritical systems. Sensitivities of the subcriticality multiplication rate $$k$$$$_{rm sub}$$ and the generation-wise neutron multiplication rate $$k$$$$_{i}$$, which is a new neutronic parameter proposed in the present paper, are calculated, and these sensitivities are compared with the $$k$$$$_{rm eff}$$ sensitivities which have been conventionally used for sensitivity works. The present sensitivity calculations show that the sensitivities of $$k$$$$_{rm sub}$$ and $$k$$$$_{i}$$ in small values of $$i$$ are significantly different from the $$k_{eff}$$ sensitivities. This result indicates that the sensitivity analyses focusing only on the solution of the eigenvalue equation cannot provide full information on the core properties of subcritical systems, and that the sensitivities of the other neutronic parameters describing real neutron multiplication are essential.

Journal Articles

Improvement of Tone's method with two-term rational approximation

Yamamoto, Akio*; Endo, Tomohiro*; Chiba, Go

Journal of Nuclear Science and Technology, 48(2), p.263 - 271, 2011/02

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

An improvement of Tone's method, which is a resonance calculation method based on the equivalence theory, is proposed. In order to increase calculation accuracy the two-term rational approximation is incorporated for the representation of neutron flux. Furthermore, some theoretical aspects of Tone's method, i.e., its inherent approximation and choice of adequate multi-group cross section for collision probability estimation, are also discussed. The validity of the improved Tone method is confirmed through a verification calculation in a irregular lattice geometry, which represents part of a LWR fuel assembly. The calculation result clarifies the validity of the present method.

Journal Articles

The H-Invitational Database (H-InvDB); A Comprehensive annotation resource for human genes and transcripts

Yamasaki, Chisato*; Murakami, Katsuhiko*; Fujii, Yasuyuki*; Sato, Yoshiharu*; Harada, Erimi*; Takeda, Junichi*; Taniya, Takayuki*; Sakate, Ryuichi*; Kikugawa, Shingo*; Shimada, Makoto*; et al.

Nucleic Acids Research, 36(Database), p.D793 - D799, 2008/01

 Times Cited Count:51 Percentile:74.53(Biochemistry & Molecular Biology)

Here we report the new features and improvements in our latest release of the H-Invitational Database, a comprehensive annotation resource for human genes and transcripts. H-InvDB, originally developed as an integrated database of the human transcriptome based on extensive annotation of large sets of fulllength cDNA (FLcDNA) clones, now provides annotation for 120 558 human mRNAs extracted from the International Nucleotide Sequence Databases (INSD), in addition to 54 978 human FLcDNAs, in the latest release H-InvDB. We mapped those human transcripts onto the human genome sequences (NCBI build 36.1) and determined 34 699 human gene clusters, which could define 34 057 protein-coding and 642 non-protein-coding loci; 858 transcribed loci overlapped with predicted pseudogenes.

Oral presentation

Development of generalized area ratio method for subcriticality measurement

Katano, Ryota; Nishihara, Kenji; Tsujimoto, Kazufumi; Endo, Tomohiro*

no journal, , 

no abstracts in English

Oral presentation

Implementation of random sampling for ACE-format cross sections using FRENDY

Kondo, Ryoichi*; Endo, Tomohiro*; Yamamoto, Akio*; Tada, Kenichi

no journal, , 

The random sampling module for ACE format cross sections are developed using modules of FRENDY. This module perturbs the cross sections and other parameters in ACE format cross section library using covariance data. The GODIVA reactor is used and the calculation results of TSUNAMI-1D are compared to verify this module.

Oral presentation

Multi-group cross section generation capability using ACE format file for MCNP

Yamamoto, Akio*; Endo, Tomohiro*; Tada, Kenichi

no journal, , 

This presentation explains the multi-group generation module. This module generates multi-group cross section library from the ACE file which is the cross section library for continuous energy Monte Carlo calculation codes. The processing results of this function are compared with those of NJOY to verify this module.

Oral presentation

Development of simultaneous analytical method for $$^{93}$$Zr and $$^{93}$$Mo based on solid phase extraction combined with ICP-MS/MS, 2; Spectral interference removal for measurement of $$^{93}$$Zr and $$^{93}$$Mo by ICP-MS/MS

Do, V. K.; Furuse, Takahiro; Murakami, Erina; Aita, Rena; Ota, Yuki; Tomitsuka, Tomohiro; Sano, Yuichi; Akimoto, Yuji*; Endo, Tsubasa*; Katayama, Atsushi; et al.

no journal, , 

The paper presents removal of possible interferences including from an isobar ($$^{93}$$Nb) and tailings of adjacent peaks for the quantification of $$^{93}$$Zr and $$^{93}$$Mo using an ICP-MS/MS (Agilent 8900). By using ammonia gas (NH$$_{3}$$) as a reaction gas, $$^{93}$$Zr and $$^{93}$$Mo can be separated from each other and from $$^{93}$$Nb owing to the different reactions of those elements with the reaction gas. Based on the characterization results, we propose a measurement scheme aiming at quantification of $$^{93}$$Zr and $$^{93}$$Mo in environmental samples collected at adjacent location of Fukushima Daiichi Nuclear Power Station.

Oral presentation

Development of simultaneous analytical method for $$^{93}$$Zr and $$^{93}$$Mo based on solid phase extraction combined with ICP-MS/MS, 1; Sequential chemical separation of Zr and Mo from Nb

Furuse, Takahiro; Do, V. K.; Aita, Rena; Ota, Yuki; Murakami, Erina; Tomitsuka, Tomohiro; Sano, Yuichi; Akimoto, Yuji*; Endo, Tsubasa*; Katayama, Atsushi; et al.

no journal, , 

In order to simplify the analysis of $$^{93}$$Zr and $$^{93}$$Mo in radioactive waste from conventional radiation measurement, we have considered analysis method combining solid-phase extraction and ICP-MS/MS. In this presentation, we report the results of a study on sequential chemical separation of Zr and Mo from Nb and sample matrix using ZR resin as a solid-phase extraction resin.

Oral presentation

Development of FRENDY/MG, 1; Outline of multi-group cross section generation capability

Yamamoto, Akio*; Endo, Tomohiro*; Chiba, Go*; Tada, Kenichi

no journal, , 

The multi-group cross section generation function FRENDY/MG is now under development. FRENDY/MG generates a multi-group cross section file from ACE file which is cross section file for a continuous energy Monte Carlo calculation code. Using FRENDY/MG and FRENDY version 1, the multi-group cross section file can be generated.

Oral presentation

Statistical error calculation method for probability table generation

Tada, Kenichi; Endo, Tomohiro*

no journal, , 

Probability table is used for consideration of self-shielding effect in the unresolved resonance region. The probability table is calculated by the Monte Carlo calculation method. However, statistical error of the probability table is not calculated. In this study, we developed statistical error calculation method for probability table generation and implemented this method to FRENDY.

Oral presentation

Handling the evaluated nuclear data format and ACE format in FRENDY

Tada, Kenichi; Yamamoto, Akio*; Endo, Tomohiro*

no journal, , 

The purpose of this consultancy meeting is to discuss the applicability of evaluated nuclear data format. FRENDY can handle evaluated nuclear data format, i.e., ENDF-6, and it will be handle the GNDS format in the near future. This presentation explains treatment of ENDF-6 and GNDS format in FRENDY and plotting and modification tools using FRENDY functions. FRENDY is now preparing the multi-group cross section generation function from the ACE file and the ACE file plotting and perturbation tools. This presentation also explain the treatment of the ACE file in FRENDY.

Oral presentation

Nuclear data processing code FRENDY Version 2

Tada, Kenichi; Yamamoto, Akio*; Endo, Tomohiro*; Chiba, Go*; Ono, Michitaka*; Tojo, Masayuki*

no journal, , 

FRENDY (From Evaluated Nuclear Data library to any application) is a nuclear data processing code for the evaluated nuclear data libraries JENDL, ENDF/B, JEFF, TENDL, and so on. The first version of FRENDY was released in 2019 as an open-source software under the 2-clause BSD license. FRENDY Ver. 1 generates ACE files which is used for the continuous energy Monte Carlo codes such as PHITS, Serpent, and MCNP. Many new modules, e.g., the multi-group neutron cross-section generation from the ACE file, are implemented in FRENDY Ver. 2. This presentation explains the characteristics of FRENDY and new capabilities implemented in FRENDY Ver. 2.

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