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":null,"pages":null},"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":null,"pages":null},"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}
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":null,"pages":null},"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}
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":null,"pages":null},"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}
{"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":null,"pages":null},"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}
ABSTRACTNeutron energy spectra down to thermal energy were measured using a Bonner sphere spectrometer (BSS) for various thicknesses of concrete and steel shielding at the CERN/CHARM facility, where high-energy neutrons were produced by 24-GeV/c protons incident on a thick copper target. The thicknesses of the concrete and steel shielding blocks ranged from 40 cm to 200 cm and from 20 cm to 80 cm, respectively. The BSS consisted of a spherical 3He proportional counter and five polyethylene moderators with diameters of 7.62 cm, 10.2 cm, 12.7 cm, 17.8 cm, and 24.1 cm, respectively. In addition, polyethylene moderators combined with a lead or copper inner shell were used to increase the sensitivity to high-energy neutrons. The neutron energy spectra were deduced using an unfolding method. The initial guesses were obtained using the PHITS code for each experimental geometry. The response function for the BSS was determined using the MCNP6.2 code with JENDL-4.0/HE. The neutron energy spectra over the entire energy region from 10−4 eV to 10 GeV were successfully obtained for the different shielding conditions. The validity of the response function and the contribution of each moderator are discussed referring to previous studies and tests at the standard neutron fields of AIST.KEYWORDS: Neutronsconcrete shieldingsteel shieldinghigh-energy neutronsBonner sphere spectrometerunfoldingDisclaimerAs 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. Additional informationFundingThis study was supported by the Tsukuba Innovation Arena (TIA) collaborative research program “Kakehashi”.
{"title":"Measurement of neutron spectra for various thicknesses of concrete and steel shielding at 24-GeV/c proton beam facility using Bonner sphere spectrometer","authors":"Tetsuro Matsumoto, Akihiko Masuda, Eunji Lee, Toshiya Sanami, Takahiro Oyama, Tsuyoshi Kajimoto, Noriaki Nakao, Hiroshi Yashima, Seiji Nagaguro, Yoshitomo Uwamino, Seiya Manabe, Nobuhiro Shigyo, Hideki Harano, Robert Froeschl, Elpida Iliopoulou, Angelo Infantino, Stefan Roesler, Markus Brugger","doi":"10.1080/00223131.2023.2274933","DOIUrl":"https://doi.org/10.1080/00223131.2023.2274933","url":null,"abstract":"ABSTRACTNeutron energy spectra down to thermal energy were measured using a Bonner sphere spectrometer (BSS) for various thicknesses of concrete and steel shielding at the CERN/CHARM facility, where high-energy neutrons were produced by 24-GeV/c protons incident on a thick copper target. The thicknesses of the concrete and steel shielding blocks ranged from 40 cm to 200 cm and from 20 cm to 80 cm, respectively. The BSS consisted of a spherical 3He proportional counter and five polyethylene moderators with diameters of 7.62 cm, 10.2 cm, 12.7 cm, 17.8 cm, and 24.1 cm, respectively. In addition, polyethylene moderators combined with a lead or copper inner shell were used to increase the sensitivity to high-energy neutrons. The neutron energy spectra were deduced using an unfolding method. The initial guesses were obtained using the PHITS code for each experimental geometry. The response function for the BSS was determined using the MCNP6.2 code with JENDL-4.0/HE. The neutron energy spectra over the entire energy region from 10−4 eV to 10 GeV were successfully obtained for the different shielding conditions. The validity of the response function and the contribution of each moderator are discussed referring to previous studies and tests at the standard neutron fields of AIST.KEYWORDS: Neutronsconcrete shieldingsteel shieldinghigh-energy neutronsBonner sphere spectrometerunfoldingDisclaimerAs 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. Additional informationFundingThis study was supported by the Tsukuba Innovation Arena (TIA) collaborative research program “Kakehashi”.","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-10-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"136381339","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}
ABSTRACTUncertainty in neutron reaction rates after penetrating a thick copper benchmark assembly was estimated based on two different kinds of Total Monte Carlo methods under random sampling methodology. 500 random nuclear data files were generated for 63Cu by Bayesian Monte-Carlo method by perturbing underlying nuclear model parameters. In the first method, these files were used directly to generate processed libraries preserving all the correlations among different physical quantities. In the second method, the random files generated in the first method were used but the angular distributions of elastic scattering were kept fixed to those of the non-perturbed nominal one. It was found that the two methods gave the same neutron reaction rate after 608 mm penetration of a copper. However, the uncertainty of the first method was smaller than that of the second method. It shows that the correlation between angular distribution of elastic scattering at 0 degrees and total cross section, which stems from Wick’s inequality, affects uncertainty of the calculated neutron reaction rate. It could be concluded that the uncertainty obtained by using the covariance files given in the ENDF-6 format may not give correct results for the uncertainty of neutron penetration calculations.KEYWORDS: Uncertainty of nuclear datacovariancecross-sections14MeV neutron63Cutotal Monte Carlo methodcorrelation of different physical quantitiesfusion neutronicsDisclaimerAs 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. Competing interestsThe authors declare that they have no competing interests.Figure 1. Calculation scheme of cross section and associated covariance (T6, upper part), and the Total Monte Carlo (TMC) method (lower part). TANES and TAFIS in T6 were not used in the present work.Display full sizeFigure 2. 63Cu neutron total cross sections estimated by T6 system (gray lines) compared with ENDF/B-VIII.0 (red line) and experimental data and the errors [Citation15–20].Display full sizeFigure 3. 63Cu neutron elastic cross sections estimated by T6 system (gray lines) compared with ENDF/B-VIII.0 (red line) and experimental data [Citation21–24].Display full sizeFigure 4. Computational geometry model of FNS copper benchmark experiment.Display full sizeFigure 5. Comparison of 90Zr(n,2n)89Zr reaction rate distribution and the standard deviation between T6 and the result ignoring perturbation of angular distribution.Display full sizeFigure 6. Comparison of 59Co(n,α)56Mn reaction rate distribution and the standard deviation between T6 and the result ignoring perturbation of angular distribution.Display fu
摘要在随机抽样方法下，采用两种不同的Total Monte Carlo方法估计了穿透厚铜基准组件后中子反应速率的不确定性。采用贝叶斯蒙特卡罗方法，通过扰动底层核模型参数，生成了500个随机核数据文件。在第一种方法中，这些文件被直接用于生成保留不同物理量之间所有相关性的处理库。第二种方法利用第一种方法产生的随机文件，但保持弹性散射角分布不变，不变为无扰动标称散射角分布。结果表明，两种方法在铜的608 mm穿透后的中子反应速率相同。然而，第一种方法的不确定度小于第二种方法。结果表明，由Wick不等式引起的0度弹性散射角分布与总截面的相关性影响了中子反应速率计算的不确定性。可以得出结论，用ENDF-6格式给出的协方差文件得到的不确定度不能给出中子侵彻计算的不确定度的正确结果。关键词:核数据的不确定性协方差横截面14mev中子总量蒙特卡罗方法不同物理量的相关性聚变中子免责声明作为对作者和研究人员的服务，我们提供此版本的已接受稿件(AM)。在最终出版版本记录(VoR)之前，将对该手稿进行编辑、排版和审查。在制作和印前，可能会发现可能影响内容的错误，所有适用于期刊的法律免责声明也与这些版本有关。竞争利益作者声明他们没有竞争利益。图1所示。截面及相关协方差的计算方案(T6，上半部分)和Total Monte Carlo (TMC)方法(下半部分)。T6中的TANES和TAFIS在本工作中未被使用。显示完整尺寸图2T6系统估算的63Cu中子总截面(灰线)与ENDF/B-VIII.0的比较(红线)和实验数据及误差[Citation15-20]。显示完整尺寸图3T6系统估算的63Cu中子弹性截面(灰线)与ENDF/B-VIII.0的比较(红线)和实验数据[Citation21-24]。显示完整尺寸图4。FNS铜基准实验计算几何模型。显示完整尺寸图5比较90Zr(n,2n)89Zr反应速率分布及T6与忽略角分布扰动结果的标准差。显示完整尺寸图659Co(n，α)56Mn反应速率分布及T6与忽略角分布扰动结果的标准差比较。显示完整尺寸图7。比较56Fe(n,p)56Mn反应速率分布及T6与忽略角分布扰动结果的标准差。显示完整尺寸图8。64Zn(n,p)64Cu反应速率分布及T6与忽略角分布扰动结果的标准差比较。显示完整尺寸图9。115In(n,n ')115mIn反应速率分布与T6与忽略角分布扰动结果的标准差比较。显示完整尺寸图10比较55Mn(n，γ)56Mn的反应速率分布和T6与忽略角分布扰动的结果之间的标准差。显示完整尺寸图11。63Cu(n，γ)64Cu的反应速率分布及T6与忽略角分布扰动结果的标准差比较。显示完整尺寸图12T6生成的1000个随机文件的63Cu总截面与0度弹性散射截面的相关性入射中子能量为14兆电子伏。显示完整尺寸图13T6生成的1000个随机文件的63Cu总截面与0度弹性散射截面的相关性入射中子能量为2.2 MeV。显示完整尺寸图14。T6生成的1000个随机文件的63Cu总截面与0度弹性散射截面的相关性入射中子能量为1.0兆电子伏。显示完整尺寸图15。T6生成的1000个随机文件的63Cu总截面与0度弹性散射截面的相关性入射中子能量为0.2兆电子伏。全尺寸显示
{"title":"Effects of correlations in uncertainties of total cross section and elastic angular distribution for a deep-penetration of 14-MeV neutrons in Cu","authors":"Naoki Yamano, Satoshi Chiba, Tsunenori Inakura, Chikako Ishizuka","doi":"10.1080/00223131.2023.2272759","DOIUrl":"https://doi.org/10.1080/00223131.2023.2272759","url":null,"abstract":"ABSTRACTUncertainty in neutron reaction rates after penetrating a thick copper benchmark assembly was estimated based on two different kinds of Total Monte Carlo methods under random sampling methodology. 500 random nuclear data files were generated for 63Cu by Bayesian Monte-Carlo method by perturbing underlying nuclear model parameters. In the first method, these files were used directly to generate processed libraries preserving all the correlations among different physical quantities. In the second method, the random files generated in the first method were used but the angular distributions of elastic scattering were kept fixed to those of the non-perturbed nominal one. It was found that the two methods gave the same neutron reaction rate after 608 mm penetration of a copper. However, the uncertainty of the first method was smaller than that of the second method. It shows that the correlation between angular distribution of elastic scattering at 0 degrees and total cross section, which stems from Wick’s inequality, affects uncertainty of the calculated neutron reaction rate. It could be concluded that the uncertainty obtained by using the covariance files given in the ENDF-6 format may not give correct results for the uncertainty of neutron penetration calculations.KEYWORDS: Uncertainty of nuclear datacovariancecross-sections14MeV neutron63Cutotal Monte Carlo methodcorrelation of different physical quantitiesfusion neutronicsDisclaimerAs 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. Competing interestsThe authors declare that they have no competing interests.Figure 1. Calculation scheme of cross section and associated covariance (T6, upper part), and the Total Monte Carlo (TMC) method (lower part). TANES and TAFIS in T6 were not used in the present work.Display full sizeFigure 2. 63Cu neutron total cross sections estimated by T6 system (gray lines) compared with ENDF/B-VIII.0 (red line) and experimental data and the errors [Citation15–20].Display full sizeFigure 3. 63Cu neutron elastic cross sections estimated by T6 system (gray lines) compared with ENDF/B-VIII.0 (red line) and experimental data [Citation21–24].Display full sizeFigure 4. Computational geometry model of FNS copper benchmark experiment.Display full sizeFigure 5. Comparison of 90Zr(n,2n)89Zr reaction rate distribution and the standard deviation between T6 and the result ignoring perturbation of angular distribution.Display full sizeFigure 6. Comparison of 59Co(n,α)56Mn reaction rate distribution and the standard deviation between T6 and the result ignoring perturbation of angular distribution.Display fu","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135315854","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 : 2023-10-20DOI: 10.1080/00223131.2023.2269974
Y. Sun, M. Kumagai, M. Jin, E. Sato, M. Aoki, Y. Ohishi, K. Kurosaki
ABSTRACTAdvanced nuclear fuels are designed to offer improved performance and accident tol- erance, with an emphasis on achieving higher thermal conductivity. While promising fuel candidates like uranium nitrides, carbides, and silicides have been widely stud- ied, the majority of uranium compounds remain unexplored. To search for potential candidates among these unexplored uranium compounds, we incorporated machine learning to accelerate the material discovery process. In this study, we trained a multiclass classification model to predict a compound’s thermal conductivity based on 133 input features derived from element properties and temperature. The initial training data consists of over 160,000 processed thermal conductivity records from the Starrydata2 database, but a skewed data class distribution led the trained model to underestimate compound’s thermal conductivity. Consequently, we addressed the issue of class imbalance by applying Synthetic Minority Oversampling TEchnique and Random UnderSampling, improving the recall for materials with thermal con- ductivity higher than 15 W/mK from 0.64 to 0.71. Finally, our best model is used to identify 119 potential advanced fuel candidates with high thermal conductivity among 774 stable uranium compounds. Our results underscore the potential of ma- chine learning in the field of nuclear science, accelerating the discovery of advanced nuclear materials.KEYWORDS: Advanced nuclear fuelsMachine learningthermal conductivityDisclaimerAs 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. Disclosure statementNo potential conflict of interest was reported by the author(s).Data availability statementThe data that support the findings of this study are openly available at https://github.com/AzarashiYifan/classification-uranium-thermal-conductivity.Additional informationFundingThis work was supported by MEXT Innovative Nuclear Research and Development Program Grant Number JPMXD0220354330 and JPMXD0222682541.
{"title":"A multiclass classification model for predicting the thermal conductivity of uranium compounds","authors":"Y. Sun, M. Kumagai, M. Jin, E. Sato, M. Aoki, Y. Ohishi, K. Kurosaki","doi":"10.1080/00223131.2023.2269974","DOIUrl":"https://doi.org/10.1080/00223131.2023.2269974","url":null,"abstract":"ABSTRACTAdvanced nuclear fuels are designed to offer improved performance and accident tol- erance, with an emphasis on achieving higher thermal conductivity. While promising fuel candidates like uranium nitrides, carbides, and silicides have been widely stud- ied, the majority of uranium compounds remain unexplored. To search for potential candidates among these unexplored uranium compounds, we incorporated machine learning to accelerate the material discovery process. In this study, we trained a multiclass classification model to predict a compound’s thermal conductivity based on 133 input features derived from element properties and temperature. The initial training data consists of over 160,000 processed thermal conductivity records from the Starrydata2 database, but a skewed data class distribution led the trained model to underestimate compound’s thermal conductivity. Consequently, we addressed the issue of class imbalance by applying Synthetic Minority Oversampling TEchnique and Random UnderSampling, improving the recall for materials with thermal con- ductivity higher than 15 W/mK from 0.64 to 0.71. Finally, our best model is used to identify 119 potential advanced fuel candidates with high thermal conductivity among 774 stable uranium compounds. Our results underscore the potential of ma- chine learning in the field of nuclear science, accelerating the discovery of advanced nuclear materials.KEYWORDS: Advanced nuclear fuelsMachine learningthermal conductivityDisclaimerAs 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. Disclosure statementNo potential conflict of interest was reported by the author(s).Data availability statementThe data that support the findings of this study are openly available at https://github.com/AzarashiYifan/classification-uranium-thermal-conductivity.Additional informationFundingThis work was supported by MEXT Innovative Nuclear Research and Development Program Grant Number JPMXD0220354330 and JPMXD0222682541.","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-10-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135616877","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}
ABSTRACTThe Japan Atomic Energy Agency has been developing 9Cr-oxide dispersion strengthened (ODS) steel as a fuel cladding material for sodium-cooled fast reactors (SFRs). Previous studies have formulated the creep rupture equation for 650°C–850°C. However, little data have been obtained above 850°C, and no equation has been formulated. This study conducted creep tests to evaluate creep strength at 700°C–1000°C. Two creep test methods, the internal pressure and ring creep tests under development, were used, and the validation of the ring creep test method was conducted. The results showed that 9Cr-ODS steel undergoes almost no strength change due to the matrix’s phase transformation, and a single equation can express a creep rupture strength from 700°C to 1000°C. In validating the ring creep test method, analysis clarified the effect of stress concentration on the specimen. Plastic deformation occurs at high initial stress and may lead to early rupture. The results will be essential for future creep testing and evaluation of neutron-irradiated 9Cr-ODS steel.KEYWORDS: Oxide dispersion strengthened steelfuel cladding tube,creep strengthcreep straininternal creep testring creep testDisclaimerAs 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. AcknowledgmentsThe authors would like to express their sincere gratitude to Dr. Tomoyuki Uwaba for his valuable guidance on finite element simulation.Additional informationFundingMEXT Innovative Nuclear Research and Development Program Grant Number JPMXD0219214482, Ministry of Education Culture, Sports, Science, and Technology, supported this work.
{"title":"Creep deformation and rupture behavior of 9Cr-ODS steel cladding tube at high temperatures from 700°C to 1000°C","authors":"Yuya Imagawa, Ryuta Hashidate, Takeshi Miyazawa, Takashi Onizawa, Satoshi Ohtsuka, Yasuhide Yano, Takashi Tanno, Takeji Kaito, Masato Ohnuma, Masatoshi Mitsuhara, Takeshi Toyama","doi":"10.1080/00223131.2023.2269178","DOIUrl":"https://doi.org/10.1080/00223131.2023.2269178","url":null,"abstract":"ABSTRACTThe Japan Atomic Energy Agency has been developing 9Cr-oxide dispersion strengthened (ODS) steel as a fuel cladding material for sodium-cooled fast reactors (SFRs). Previous studies have formulated the creep rupture equation for 650°C–850°C. However, little data have been obtained above 850°C, and no equation has been formulated. This study conducted creep tests to evaluate creep strength at 700°C–1000°C. Two creep test methods, the internal pressure and ring creep tests under development, were used, and the validation of the ring creep test method was conducted. The results showed that 9Cr-ODS steel undergoes almost no strength change due to the matrix’s phase transformation, and a single equation can express a creep rupture strength from 700°C to 1000°C. In validating the ring creep test method, analysis clarified the effect of stress concentration on the specimen. Plastic deformation occurs at high initial stress and may lead to early rupture. The results will be essential for future creep testing and evaluation of neutron-irradiated 9Cr-ODS steel.KEYWORDS: Oxide dispersion strengthened steelfuel cladding tube,creep strengthcreep straininternal creep testring creep testDisclaimerAs 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. AcknowledgmentsThe authors would like to express their sincere gratitude to Dr. Tomoyuki Uwaba for his valuable guidance on finite element simulation.Additional informationFundingMEXT Innovative Nuclear Research and Development Program Grant Number JPMXD0219214482, Ministry of Education Culture, Sports, Science, and Technology, supported this work.","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-10-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135730090","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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