We have developed a novel theoretical method to obtain independent fission product yields and prompt neutron observables by connecting mass and total kinetic energy (TKE) distributions calculated by a four-dimensional Langevin dynamical model to a Hauser-Feshbach statistical decay model. In the Langevin calculations, mass distributions corresponding to the standard I and II modes were obtained separately and superposed to reproduce the fission fragment yield of spontaneous fission of 238,240,242Pu and thermal neutron-induced fission of 239Pu. This was achieved by using different neck parameters for these two modes in the two-center shell model shape parametrization, and a systematics of the superposing ratio was obtained as a function of (N−Z)/A of the fissioning nuclei. The Hauser-Feshbach calculations were performed using a nuclear reaction code TALYS for 239Pu(n,f) reaction in the incident energy range from thermal up to 5MeV, and the calculated prompt fission observables were compared with experimental and evaluated data. Although further improvements are needed for the most important nuclides, it turned out that the present methodology has the capability to prepare fission-related nuclear data for nuclides for which measurements are difficult.
{"title":"Connection of four-dimensional Langevin model and Hauser-Feshbach theory to describe statistical decay of fission fragments","authors":"Kazuki Fujio, Shin Okumura, Chikako Ishizuka, Satoshi Chiba, Tatsuya Katabuchi","doi":"10.1080/00223131.2023.2273470","DOIUrl":"https://doi.org/10.1080/00223131.2023.2273470","url":null,"abstract":"We have developed a novel theoretical method to obtain independent fission product yields and prompt neutron observables by connecting mass and total kinetic energy (TKE) distributions calculated by a four-dimensional Langevin dynamical model to a Hauser-Feshbach statistical decay model. In the Langevin calculations, mass distributions corresponding to the standard I and II modes were obtained separately and superposed to reproduce the fission fragment yield of spontaneous fission of 238,240,242Pu and thermal neutron-induced fission of 239Pu. This was achieved by using different neck parameters for these two modes in the two-center shell model shape parametrization, and a systematics of the superposing ratio was obtained as a function of (N−Z)/A of the fissioning nuclei. The Hauser-Feshbach calculations were performed using a nuclear reaction code TALYS for 239Pu(n,f) reaction in the incident energy range from thermal up to 5MeV, and the calculated prompt fission observables were compared with experimental and evaluated data. Although further improvements are needed for the most important nuclides, it turned out that the present methodology has the capability to prepare fission-related nuclear data for nuclides for which measurements are difficult.","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":" 8","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-11-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135291201","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2023-11-06DOI: 10.1080/00223131.2023.2278599
So Kamada, Masayuki IGASHIRA, Tatsuya Katabuchi, MIZUMOTO Motoharu
ABSTRACTThe neutron capture cross sections and capture γ-ray spectra of 74,76,78,80,82Se were measured in a region from 15 to 100 keV and around 550 keV. A neutron time-of-flight method was used with a ns-pulsed neutron source based on the 7Li(p,n)7Be reaction and a large anti-Compton NaI(Tl) γ-ray spectrometer. A pulse-height weighting technique was applied to the observed γ-ray pulse-height spectra to obtain capture yields. The capture cross sections of 74,76,78,80Se were derived with uncertainties from 4.0 to 5.5% and those of 82Se were derived with uncertainties of 6.5–27% by using the standard capture cross sections of 197Au. The present results of 78,82Se were the first experimental ones above the resolved resonance region. The present results were compared with previous measurements and the evaluated values in JENDL-5.0 and ENDF/B-VIII.0. The evaluations of JENDL-5.0 differ from the present results of 74,76,78,80Se and 82Se by 0.9–51% and 6.9–120%, respectively. The capture γ-ray spectra of 74,76,78,80,82Se were derived by unfolding the observed capture γ-ray pulse-height spectra. The present results were the first experimental ones in the keV region.KEYWORDS: Neutron capturecross sectionsgamma spectrakev rangeselenium 74selenium 76selenium 78selenium 80selenium 82selenium 79gold 197Anti-compton NaI(Tl) gamma-ray spectrometertime-of-flight methodpulse-height weighting techniqueDisclaimerAs a service to authors and researchers we are providing this version of an accepted manuscript (AM). Copyediting, typesetting, and review of the resulting proofs will be undertaken on this manuscript before final publication of the Version of Record (VoR). During production and pre-press, errors may be discovered which could affect the content, and all legal disclaimers that apply to the journal relate to these versions also. AcknowledgementThe present study was supported by a Grant-in-Aid (No. 19360423) of the Japan Ministry of Education, Culture, Sports, Science and Technology. This work was also supported by KAKENHI Grant-in-Aids (21K04580) for publication.Additional informationFundingThe work was supported by the a Grant-in-Aid of the Japan Ministry of Education, Culture, Sports, Science and Technology [19360423].
{"title":"Measurements of keV-Neutron capture cross sections and capture gamma-ray spectra of 74 ,76 ,78 ,80 ,82Se","authors":"So Kamada, Masayuki IGASHIRA, Tatsuya Katabuchi, MIZUMOTO Motoharu","doi":"10.1080/00223131.2023.2278599","DOIUrl":"https://doi.org/10.1080/00223131.2023.2278599","url":null,"abstract":"ABSTRACTThe neutron capture cross sections and capture γ-ray spectra of 74,76,78,80,82Se were measured in a region from 15 to 100 keV and around 550 keV. A neutron time-of-flight method was used with a ns-pulsed neutron source based on the 7Li(p,n)7Be reaction and a large anti-Compton NaI(Tl) γ-ray spectrometer. A pulse-height weighting technique was applied to the observed γ-ray pulse-height spectra to obtain capture yields. The capture cross sections of 74,76,78,80Se were derived with uncertainties from 4.0 to 5.5% and those of 82Se were derived with uncertainties of 6.5–27% by using the standard capture cross sections of 197Au. The present results of 78,82Se were the first experimental ones above the resolved resonance region. The present results were compared with previous measurements and the evaluated values in JENDL-5.0 and ENDF/B-VIII.0. The evaluations of JENDL-5.0 differ from the present results of 74,76,78,80Se and 82Se by 0.9–51% and 6.9–120%, respectively. The capture γ-ray spectra of 74,76,78,80,82Se were derived by unfolding the observed capture γ-ray pulse-height spectra. The present results were the first experimental ones in the keV region.KEYWORDS: Neutron capturecross sectionsgamma spectrakev rangeselenium 74selenium 76selenium 78selenium 80selenium 82selenium 79gold 197Anti-compton NaI(Tl) gamma-ray spectrometertime-of-flight methodpulse-height weighting techniqueDisclaimerAs a service to authors and researchers we are providing this version of an accepted manuscript (AM). Copyediting, typesetting, and review of the resulting proofs will be undertaken on this manuscript before final publication of the Version of Record (VoR). During production and pre-press, errors may be discovered which could affect the content, and all legal disclaimers that apply to the journal relate to these versions also. AcknowledgementThe present study was supported by a Grant-in-Aid (No. 19360423) of the Japan Ministry of Education, Culture, Sports, Science and Technology. This work was also supported by KAKENHI Grant-in-Aids (21K04580) for publication.Additional informationFundingThe work was supported by the a Grant-in-Aid of the Japan Ministry of Education, Culture, Sports, Science and Technology [19360423].","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":"11 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-11-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135679013","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2023-11-06DOI: 10.1080/00223131.2023.2278600
Kenichi Tada, Akio Yamamoto, Satoshi Kunieda, Chikara Konno, Ryoichi Kondo, Tomohiro Endo, Go Chiba, Michitaka Ono, Masayuki Tojo
ABSTRACTNuclear data processing is important to connect evaluated nuclear data libraries and radiation transport codes. The nuclear data processing code FRENDY version 1 was released in 2019 to generate ACE formatted cross section files with simple input data. After we released FRENDY version 1, many functions were developed, e.g., neutron multi-group cross section generation, explicit consideration of the resonance interference effect among different nuclides in a material, adaptive setting of the background cross sections, consideration of the resonance upscattering, ACE file perturbation, statistical uncertainty quantification of probability table, and modification of ENDF-6 formatted file. FRENDY version 2 was released including these new functions. It generates GENDF and MATXS formatted neutron multi-group cross section files from an ACE formatted cross section file or an evaluated nuclear data file. This paper explains the features of the new functions implemented in FRENDY version 2 and the verification of the neutron multigroup cross section generation function of this code.KEYWORDS: FRENDYnuclear data processingevaluated nuclear data libraryneutron multigroup cross section generationACEGENDFMATXSDisclaimerAs a service to authors and researchers we are providing this version of an accepted manuscript (AM). Copyediting, typesetting, and review of the resulting proofs will be undertaken on this manuscript before final publication of the Version of Record (VoR). During production and pre-press, errors may be discovered which could affect the content, and all legal disclaimers that apply to the journal relate to these versions also. Figure 1. Difference in the neutron multi-group cross section generation between a conventional code and FRENDY.Display full sizeFigure 2. Cross section of U-238 from JENDL-4.0 with and without resonance upscattering (MT=102, XMAS 172g).Display full sizeFigure 3. Secondary photon spectrum of Fe-56 from JENDL-4.0 (MT=102, Ein=12.5-12.8 MeV, VITAMIN-J 42g).Display full sizeFigure 4. Secondary photon spectrum of Fe-56 from JENDL-4.0 (MT=22, Ein=12.5-12.8 MeV, VITAMIN-J 42g).Display full sizeFigure 5. Example of the interpolation of energy with a ratio of 1:2 in the two-dimensional interpolation procedure.Display full sizeFigure 6. Comparison of secondary photon spectra of Fe-56 from JENDL-4.0 in each division number (MT=22, Ein=12.5-12.8 MeV, VITAMIN-J 42g)Display full sizeFigure 7. Secondary photon spectrum of Fe-56, MT=22, Ein=12 and 13 MeV from JENDL-4.0.Display full sizeFigure 8. Comparison of neutron spectra at the distance of 60 cm from the center of a 1 m radius Fe-56 sphere (JENDL-4.0, VITAMIN-B6 199g).Display full sizeFigure 9. Comparison of secondary photon spectra at the distance of 60 cm from the center of a 1 m radius Fe-56 sphere (JENDL-4.0, VITAMIN-J 42g).Display full size
{"title":"Development of nuclear data processing code FRENDY version 2","authors":"Kenichi Tada, Akio Yamamoto, Satoshi Kunieda, Chikara Konno, Ryoichi Kondo, Tomohiro Endo, Go Chiba, Michitaka Ono, Masayuki Tojo","doi":"10.1080/00223131.2023.2278600","DOIUrl":"https://doi.org/10.1080/00223131.2023.2278600","url":null,"abstract":"ABSTRACTNuclear data processing is important to connect evaluated nuclear data libraries and radiation transport codes. The nuclear data processing code FRENDY version 1 was released in 2019 to generate ACE formatted cross section files with simple input data. After we released FRENDY version 1, many functions were developed, e.g., neutron multi-group cross section generation, explicit consideration of the resonance interference effect among different nuclides in a material, adaptive setting of the background cross sections, consideration of the resonance upscattering, ACE file perturbation, statistical uncertainty quantification of probability table, and modification of ENDF-6 formatted file. FRENDY version 2 was released including these new functions. It generates GENDF and MATXS formatted neutron multi-group cross section files from an ACE formatted cross section file or an evaluated nuclear data file. This paper explains the features of the new functions implemented in FRENDY version 2 and the verification of the neutron multigroup cross section generation function of this code.KEYWORDS: FRENDYnuclear data processingevaluated nuclear data libraryneutron multigroup cross section generationACEGENDFMATXSDisclaimerAs a service to authors and researchers we are providing this version of an accepted manuscript (AM). Copyediting, typesetting, and review of the resulting proofs will be undertaken on this manuscript before final publication of the Version of Record (VoR). During production and pre-press, errors may be discovered which could affect the content, and all legal disclaimers that apply to the journal relate to these versions also. Figure 1. Difference in the neutron multi-group cross section generation between a conventional code and FRENDY.Display full sizeFigure 2. Cross section of U-238 from JENDL-4.0 with and without resonance upscattering (MT=102, XMAS 172g).Display full sizeFigure 3. Secondary photon spectrum of Fe-56 from JENDL-4.0 (MT=102, Ein=12.5-12.8 MeV, VITAMIN-J 42g).Display full sizeFigure 4. Secondary photon spectrum of Fe-56 from JENDL-4.0 (MT=22, Ein=12.5-12.8 MeV, VITAMIN-J 42g).Display full sizeFigure 5. Example of the interpolation of energy with a ratio of 1:2 in the two-dimensional interpolation procedure.Display full sizeFigure 6. Comparison of secondary photon spectra of Fe-56 from JENDL-4.0 in each division number (MT=22, Ein=12.5-12.8 MeV, VITAMIN-J 42g)Display full sizeFigure 7. Secondary photon spectrum of Fe-56, MT=22, Ein=12 and 13 MeV from JENDL-4.0.Display full sizeFigure 8. Comparison of neutron spectra at the distance of 60 cm from the center of a 1 m radius Fe-56 sphere (JENDL-4.0, VITAMIN-B6 199g).Display full sizeFigure 9. Comparison of secondary photon spectra at the distance of 60 cm from the center of a 1 m radius Fe-56 sphere (JENDL-4.0, VITAMIN-J 42g).Display full size","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":"9 2","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-11-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135678847","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2023-11-06DOI: 10.1080/00223131.2023.2279299
Motomu Suzuki, Yasunobu Nagaya
ABSTRACTWith the release of the latest Japanese evaluated nuclear data library JENDL-5, the prediction accuracy of JENDL-5 for neutronics parameters of the BEAVRS benchmark for the hot zero power condition was evaluated in this study. The criticality, control rod bank worth (CRW), isothermal temperature coefficient (ITC), and in-core detector signals were calculated and compared with the measured data for evaluation. For the criticality, the calculation-to-measurement (C/E) values varied between 1.0001 and 1.0045. Sensitivity analysis by replacing cross section data from the JENDL-4.0u1 with JENDL-5 revealed that 1H, 235U, 238U, and 16O significantly affected the criticality. The individual CRW agreed within 50 pcm, and total CRW also agreed within 100 pcm from the measured values. The ITC results calculated with a temperature deviation of 5.56 K case were negatively overestimated comparing with the measured values; whereas those of with 2.78 K were improved and agreed with the measured values within a standard deviation. The axial detector signals indicated a maximum relative error of 4.46% and the root mean squared error (RMSE) of 2.13%. The differences between the previous version of JENDL-4.0u1 and JENDL-5 were also investigated.DisclaimerAs a service to authors and researchers we are providing this version of an accepted manuscript (AM). Copyediting, typesetting, and review of the resulting proofs will be undertaken on this manuscript before final publication of the Version of Record (VoR). During production and pre-press, errors may be discovered which could affect the content, and all legal disclaimers that apply to the journal relate to these versions also. AcknowledgementsThe authors thank Dr. Kenichi Tada of JAEA for the support in the handling FRENDY code, such as formatted cross section generation and plotting.Figure 1. Fuel assembly (Asm.), burnable absorber, and control rod bank layout (quarter rotational symmetry) [Citation28].Display full sizeFigure 2. Instrument tube positions [Citation28].Display full sizeFigure 3. Whole core calculation model of the horizontal plane at the axial mid-plane.Display full sizeFigure 4. Whole core calculation model of the vertical plane at the core center.Display full sizeFigure 5. Comparison of criticality between JENDL-4.0u1 and JENDL-5 for six cases with different boron concentrations and control rod bank conditions.Display full sizeFigure 6. Comparison of neutron spectra between JENDL-4.0u1 and JENDL-5 in ARO (D = 213 steps) case.Display full sizeFigure 7. Nuclide Substitution Reactivity of JENDL-4.0u1 with JENDL-5.Display full sizeFigure 8. Comparison of scattering cross sections of TSL data for 1H in H2O between JENDL-4.0u1 and JENDL-5.Display full sizeFigure 9. Comparison of capture cross sections of 16O between JENDL-4.0u1 and JENDL-5.Display full sizeFigure 10. Comparison of fission and capture cross sections of 235U between JENDL-4.0u1 and JENDL-5.Display full sizeFigure 11. Comparison of
{"title":"Whole core analysis of BEAVRS benchmark for hot zero Power condition using MVP3 with JENDL-5","authors":"Motomu Suzuki, Yasunobu Nagaya","doi":"10.1080/00223131.2023.2279299","DOIUrl":"https://doi.org/10.1080/00223131.2023.2279299","url":null,"abstract":"ABSTRACTWith the release of the latest Japanese evaluated nuclear data library JENDL-5, the prediction accuracy of JENDL-5 for neutronics parameters of the BEAVRS benchmark for the hot zero power condition was evaluated in this study. The criticality, control rod bank worth (CRW), isothermal temperature coefficient (ITC), and in-core detector signals were calculated and compared with the measured data for evaluation. For the criticality, the calculation-to-measurement (C/E) values varied between 1.0001 and 1.0045. Sensitivity analysis by replacing cross section data from the JENDL-4.0u1 with JENDL-5 revealed that 1H, 235U, 238U, and 16O significantly affected the criticality. The individual CRW agreed within 50 pcm, and total CRW also agreed within 100 pcm from the measured values. The ITC results calculated with a temperature deviation of 5.56 K case were negatively overestimated comparing with the measured values; whereas those of with 2.78 K were improved and agreed with the measured values within a standard deviation. The axial detector signals indicated a maximum relative error of 4.46% and the root mean squared error (RMSE) of 2.13%. The differences between the previous version of JENDL-4.0u1 and JENDL-5 were also investigated.DisclaimerAs a service to authors and researchers we are providing this version of an accepted manuscript (AM). Copyediting, typesetting, and review of the resulting proofs will be undertaken on this manuscript before final publication of the Version of Record (VoR). During production and pre-press, errors may be discovered which could affect the content, and all legal disclaimers that apply to the journal relate to these versions also. AcknowledgementsThe authors thank Dr. Kenichi Tada of JAEA for the support in the handling FRENDY code, such as formatted cross section generation and plotting.Figure 1. Fuel assembly (Asm.), burnable absorber, and control rod bank layout (quarter rotational symmetry) [Citation28].Display full sizeFigure 2. Instrument tube positions [Citation28].Display full sizeFigure 3. Whole core calculation model of the horizontal plane at the axial mid-plane.Display full sizeFigure 4. Whole core calculation model of the vertical plane at the core center.Display full sizeFigure 5. Comparison of criticality between JENDL-4.0u1 and JENDL-5 for six cases with different boron concentrations and control rod bank conditions.Display full sizeFigure 6. Comparison of neutron spectra between JENDL-4.0u1 and JENDL-5 in ARO (D = 213 steps) case.Display full sizeFigure 7. Nuclide Substitution Reactivity of JENDL-4.0u1 with JENDL-5.Display full sizeFigure 8. Comparison of scattering cross sections of TSL data for 1H in H2O between JENDL-4.0u1 and JENDL-5.Display full sizeFigure 9. Comparison of capture cross sections of 16O between JENDL-4.0u1 and JENDL-5.Display full sizeFigure 10. Comparison of fission and capture cross sections of 235U between JENDL-4.0u1 and JENDL-5.Display full sizeFigure 11. Comparison of ","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":"248 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-11-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135678745","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2023-11-06DOI: 10.1080/00223131.2023.2278598
Valentina Semkova, Naohiko Otuka, Arjan J. M. Plompen
The full energy peak efficiencies and covariances of a high purity germanium (HPGe) detector determined by the regression method and Monte Carlo method were compared. The impact of the obtained results on the neutron activation cross sections measured relative to monitor cross sections was determined. In the regression method, the efficiencies measured for a set of calibration point sources were analysed by the least-squares analysis. In the Monte Carlo method, the efficiencies for the calibration point sources were calculated by MCNP. The covariances of the efficiencies determined by the regression method were calculated analytically. Perturbation analysis was performed to estimate the covariances of the efficiencies calculated by the Monte Carlo method. Positive correlations higher than 0.8 were found in the uncertainties of the MCNP data for point-like sources. In the case of the regression method, the correlation matrix contains both positive and negative terms. The efficiencies and their covariances were estimated by both methods to take into account sample effects such as geometrical effect and gamma-ray self-absorption and found considerable differences in the cross sections and their uncertainties for reaction products quantified with low energy gammas. The efficiencies and covariances were clearly affected by the properties of the sample.
{"title":"Characterization of a HPGe detector response for activation cross section measurements: regression method versus Monte Carlo method","authors":"Valentina Semkova, Naohiko Otuka, Arjan J. M. Plompen","doi":"10.1080/00223131.2023.2278598","DOIUrl":"https://doi.org/10.1080/00223131.2023.2278598","url":null,"abstract":"The full energy peak efficiencies and covariances of a high purity germanium (HPGe) detector determined by the regression method and Monte Carlo method were compared. The impact of the obtained results on the neutron activation cross sections measured relative to monitor cross sections was determined. In the regression method, the efficiencies measured for a set of calibration point sources were analysed by the least-squares analysis. In the Monte Carlo method, the efficiencies for the calibration point sources were calculated by MCNP. The covariances of the efficiencies determined by the regression method were calculated analytically. Perturbation analysis was performed to estimate the covariances of the efficiencies calculated by the Monte Carlo method. Positive correlations higher than 0.8 were found in the uncertainties of the MCNP data for point-like sources. In the case of the regression method, the correlation matrix contains both positive and negative terms. The efficiencies and their covariances were estimated by both methods to take into account sample effects such as geometrical effect and gamma-ray self-absorption and found considerable differences in the cross sections and their uncertainties for reaction products quantified with low energy gammas. The efficiencies and covariances were clearly affected by the properties of the sample.","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":"268 3","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-11-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135678928","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2023-10-31DOI: 10.1080/00223131.2023.2271895
G.N. Pikulina, S.M. Taova
{"title":"EXFOR-Editor package for entering, processing and representing Nuclear reaction data in the EXFOR format","authors":"G.N. Pikulina, S.M. Taova","doi":"10.1080/00223131.2023.2271895","DOIUrl":"https://doi.org/10.1080/00223131.2023.2271895","url":null,"abstract":"","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":"34 4","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135870047","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The Particle and Heavy Ion Transport code System (PHITS) is a general-purpose Monte Carlo radiation transport code that can simulate the behavior of most particle species with energies up to 1 TeV (per nucleon for ions). Its new version, PHITS3.33, was recently developed and released to the public. In the new version, the compatibility with nuclear data libraries and the algorithm of the track-structure modes have been improved, and they are recommended to be used for certain simulation conditions. Some utility functions and software have been developed and integrated into the new PHITS package, such as PHITS Interactive Geometry viewer in 3D (PHIG-3D) and RadioTherapy packaged based on PHITS (RT-PHITS). With these upgraded features, PHITS can be applied in a wide diversity of fields – beyond traditional nuclear engineering domains – including cosmic-ray, environmental, medical, life, and material sciences. In this paper, we summarize the upgraded features of PHITS3.33 with respect to the physics models, utility functions, and application software introduced since the release of PHITS3.02 in 2017.
{"title":"Recent improvements of the particle and Heavy Ion transport code system – PHITS version 3.33","authors":"Tatsuhiko Sato, Yosuke Iwamoto, Shintaro Hashimoto, Tatsuhiko Ogawa, Takuya Furuta, Shin-Ichiro Abe, Takeshi Kai, Yusuke Matsuya, Norihiro Matsuda, Yuho Hirata, Takuya Sekikawa, Lan Yao, Pi-En Tsai, Hunter N. Ratliff, Hiroshi Iwase, Yasuhito Sakaki, Kenta Sugihara, Nobuhiro Shigyo, Lembit Sihver, Koji Niita","doi":"10.1080/00223131.2023.2275736","DOIUrl":"https://doi.org/10.1080/00223131.2023.2275736","url":null,"abstract":"The Particle and Heavy Ion Transport code System (PHITS) is a general-purpose Monte Carlo radiation transport code that can simulate the behavior of most particle species with energies up to 1 TeV (per nucleon for ions). Its new version, PHITS3.33, was recently developed and released to the public. In the new version, the compatibility with nuclear data libraries and the algorithm of the track-structure modes have been improved, and they are recommended to be used for certain simulation conditions. Some utility functions and software have been developed and integrated into the new PHITS package, such as PHITS Interactive Geometry viewer in 3D (PHIG-3D) and RadioTherapy packaged based on PHITS (RT-PHITS). With these upgraded features, PHITS can be applied in a wide diversity of fields – beyond traditional nuclear engineering domains – including cosmic-ray, environmental, medical, life, and material sciences. In this paper, we summarize the upgraded features of PHITS3.33 with respect to the physics models, utility functions, and application software introduced since the release of PHITS3.02 in 2017.","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":"50 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135765824","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
ABSTRACTIn this research, we used a self-developed compact D-D neutron generator and designed a neutron radiation biological device. The research results showed that lead was the material with the best neutron permeability; thus, hard lead was selected as the structural material for this device. Our research disclosed a phenomenon that differed from conventional wisdom, namely, the dose contribution of the neutrons reflected from the hard lead material at the top of the device was about twice that of the polyethylene material. Therefore, hard lead was also selected for the material of the top reflector in this device. The study also showed that the device had good dose uniformity, and the standard neutron absorption dose in each irradiation zone could be obtained according to the simulation results. This device may further promote the development of future research work such as neutron radiation biological effects and mutation breeding.KEYWORDS: D-D neutron generatorNeutronRadiation biodeviceAbsorption doseDisclaimerAs a service to authors and researchers we are providing this version of an accepted manuscript (AM). Copyediting, typesetting, and review of the resulting proofs will be undertaken on this manuscript before final publication of the Version of Record (VoR). During production and pre-press, errors may be discovered which could affect the content, and all legal disclaimers that apply to the journal relate to these versions also. AcknowledgementsThis work was Supported by the National Natural Science Foundation of China (No. 12075106), the Natural Science Foundation of Gansu Province (No. 20JR10RA607), and the Fundamental Research Funds for the Central Universities of China.Disclosure statementNo potential conflict of interest was reported by the author(s).Additional informationFundingThis work was supported by the Fundamental Research Funds for the Central Universities; National Natural Science Foundation of China [12075106]; Natural Science Foundation of Gansu Province [20JR10RA607].
{"title":"Research on the design of a radiation biodevice based on a compact D-D neutron generator","authors":"Dapeng Xu, Zhiqi Guo, Xu Yang, Xiaohou Bai, Zhiming Hu, Changqi Liu, Junrun Wang, Yu Zhang, Ze’en Yao","doi":"10.1080/00223131.2023.2276409","DOIUrl":"https://doi.org/10.1080/00223131.2023.2276409","url":null,"abstract":"ABSTRACTIn this research, we used a self-developed compact D-D neutron generator and designed a neutron radiation biological device. The research results showed that lead was the material with the best neutron permeability; thus, hard lead was selected as the structural material for this device. Our research disclosed a phenomenon that differed from conventional wisdom, namely, the dose contribution of the neutrons reflected from the hard lead material at the top of the device was about twice that of the polyethylene material. Therefore, hard lead was also selected for the material of the top reflector in this device. The study also showed that the device had good dose uniformity, and the standard neutron absorption dose in each irradiation zone could be obtained according to the simulation results. This device may further promote the development of future research work such as neutron radiation biological effects and mutation breeding.KEYWORDS: D-D neutron generatorNeutronRadiation biodeviceAbsorption doseDisclaimerAs a service to authors and researchers we are providing this version of an accepted manuscript (AM). Copyediting, typesetting, and review of the resulting proofs will be undertaken on this manuscript before final publication of the Version of Record (VoR). During production and pre-press, errors may be discovered which could affect the content, and all legal disclaimers that apply to the journal relate to these versions also. AcknowledgementsThis work was Supported by the National Natural Science Foundation of China (No. 12075106), the Natural Science Foundation of Gansu Province (No. 20JR10RA607), and the Fundamental Research Funds for the Central Universities of China.Disclosure statementNo potential conflict of interest was reported by the author(s).Additional informationFundingThis work was supported by the Fundamental Research Funds for the Central Universities; National Natural Science Foundation of China [12075106]; Natural Science Foundation of Gansu Province [20JR10RA607].","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":"208 ","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135929239","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
{"title":"Experimental and computational verifications of the dose calculation accuracy of PHITS for high-energy photon beam therapy","authors":"Naoya Kuga, Takuro Shiiba, Tatsuhiko Sato, Shintaro Hashimoto, Yasuyoshi Kuroiwa","doi":"10.1080/00223131.2023.2275737","DOIUrl":"https://doi.org/10.1080/00223131.2023.2275737","url":null,"abstract":"","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":"13 2-3","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134908441","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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