ABSTRACTA molten chloride salt fast reactor with inherent safety features is being studied in order to utilize spent fuel discharged from light-water reactors effectively and to reduce environmental burdens. A hard neutron spectrum is required to achieve the transmutation of TRU efficiently, which can be realized by using a molten chloride salt. However, it was found that the criticality analysis shows a large difference of 2%Δk among the effective multiplication factors (keff) obtained using several evaluated nuclear data libraries. A sensitivity analysis was performed to investigate the cause of the difference in keff and it was clarified that the difference of 2%Δk/k reactivity was due to the difference in cross sections of the 35Cl(n,p)35S reaction. Such a large difference in keff makes a reliable core design difficult and strongly affects important core characteristics such as the TRU transmutation rate, conversion ratio, and so on. Therefore, this paper has been prepared to make a strong request for measurements and re-evaluations of the (n,p) cross section of 35Cl.KEYWORDS: Molten salt reactorchloridefast reactortransmutationTRU35Cl(np) reactionDisclaimerAs 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. AcknowledgmentsWe would like to thank Assistant Prof. Chikako Ishizuka at the Tokyo Institute of Technology for her helpful discussions, encouragement and also for computer maintenance to proceed calculations.Fig. 1 Conceptual design for 700-MWt reactor core with four heat exchangers (unit in meters).Display full sizeFig. 2 Bird’s-eye view of the reactor with four loops and a stainless steel reflector.Display full sizeFig. 3 Calculation model of the MCSFR.Display full sizeFig. 4 Average neutron spectrum in the core of the MCSFR.Display full sizeFig. 5 Sensitivities of keff to the reaction cross sections of 35Cl calculated using Serpent 2Display full sizeand ENDF/B-VIII.0.(n,p), (n,α), and (n,γ) cross sections(n,n) and (n,n’) cross sectionsFig. 6 Relative cross-section differences of 35Cl between ENDF/B-VIII.0 and JENDL-5.Display full sizeThe reference of the relative difference is ENDF/B-VIII.0.(n,p), (n,α), and (n,γ) cross sections(n,n) and (n,n’) cross sectionsFig. 7 Contribution of the relative difference in the (n,p) cross sections for 35Cl between ENDF/B-VIII.0 and JENDL-5 to reactivity.Display full sizeThe reference of the relative difference is ENDF/B-VIII.0.Fig. 8 Comparison of nuclear data libraries for the 35Cl (n,p) 35S reaction cross sections with the experimental values. There are three experimental values around 14 MeV: one of Schantl at 14.7 MeV (1970)
{"title":"Request for <sup>35</sup> Cl(n,p) reaction cross-section measurements and re-evaluations from the standpoint of molten chloride salt fast reactor design","authors":"Yoshihisa Tahara, Haruka Hirano, Satoshi Chiba, Hiroyasu Mochizuki, Tatsuya Katabuchi","doi":"10.1080/00223131.2023.2282553","DOIUrl":"https://doi.org/10.1080/00223131.2023.2282553","url":null,"abstract":"ABSTRACTA molten chloride salt fast reactor with inherent safety features is being studied in order to utilize spent fuel discharged from light-water reactors effectively and to reduce environmental burdens. A hard neutron spectrum is required to achieve the transmutation of TRU efficiently, which can be realized by using a molten chloride salt. However, it was found that the criticality analysis shows a large difference of 2%Δk among the effective multiplication factors (keff) obtained using several evaluated nuclear data libraries. A sensitivity analysis was performed to investigate the cause of the difference in keff and it was clarified that the difference of 2%Δk/k reactivity was due to the difference in cross sections of the 35Cl(n,p)35S reaction. Such a large difference in keff makes a reliable core design difficult and strongly affects important core characteristics such as the TRU transmutation rate, conversion ratio, and so on. Therefore, this paper has been prepared to make a strong request for measurements and re-evaluations of the (n,p) cross section of 35Cl.KEYWORDS: Molten salt reactorchloridefast reactortransmutationTRU35Cl(np) reactionDisclaimerAs 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. AcknowledgmentsWe would like to thank Assistant Prof. Chikako Ishizuka at the Tokyo Institute of Technology for her helpful discussions, encouragement and also for computer maintenance to proceed calculations.Fig. 1 Conceptual design for 700-MWt reactor core with four heat exchangers (unit in meters).Display full sizeFig. 2 Bird’s-eye view of the reactor with four loops and a stainless steel reflector.Display full sizeFig. 3 Calculation model of the MCSFR.Display full sizeFig. 4 Average neutron spectrum in the core of the MCSFR.Display full sizeFig. 5 Sensitivities of keff to the reaction cross sections of 35Cl calculated using Serpent 2Display full sizeand ENDF/B-VIII.0.(n,p), (n,α), and (n,γ) cross sections(n,n) and (n,n’) cross sectionsFig. 6 Relative cross-section differences of 35Cl between ENDF/B-VIII.0 and JENDL-5.Display full sizeThe reference of the relative difference is ENDF/B-VIII.0.(n,p), (n,α), and (n,γ) cross sections(n,n) and (n,n’) cross sectionsFig. 7 Contribution of the relative difference in the (n,p) cross sections for 35Cl between ENDF/B-VIII.0 and JENDL-5 to reactivity.Display full sizeThe reference of the relative difference is ENDF/B-VIII.0.Fig. 8 Comparison of nuclear data libraries for the 35Cl (n,p) 35S reaction cross sections with the experimental values. There are three experimental values around 14 MeV: one of Schantl at 14.7 MeV (1970)","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-11-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134901628","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 mutual separation of Am and Cm is conducted using an alkyl-diamide amine (ADAAM(EH)) extractant. ADAAM(EH) exhibits extremely high separation factor (SF) with respect to Am and Cm separation (SF = 5.9) in a HNO3-n-dodecane system. The batch-wise multistage extractions are performed using a system containing 0.2 M ADAAM(EH) and 1.5 M HNO3, with their distribution ratios (D) of the system for Am and Cm equal to 2.55 and 0.434, respectively. In this multistage extraction, 3 stages of feed, 9 stages involving scrubbing, and 11 stages involving the use of an organic solvent give 96.5% and 1.06% yields of Am and Cm, respectively. After the mutual separation, the aqueous and the organic phases that includes actinides increases four times in volume compared to the feed solution. Additional extraction steps are to reduce the volume of the solutions including actinides. We use tetraethyl diglycolamide (TEDGA)/water for Am concentration, tetraoctyl diglycolamide/n-dodecane for the intermediate extraction, and TEDGA/water for stripping in the Cm concentration step. Taking the additional steps, Am and Cm can be recovered in just two or three stages in the aqueous phases.KEYWORDS: Solvent extractionAmCmseparationADAAMDisclaimerAs 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 gratefully acknowledge Dr. Y. Kitatsuji at JAEA for safety management with respect to actinides and Dr. M. Arisaka at JAEA for ICP-measurements.Declaration of Interest StatementNo potential conflict of interest was reported by the authors.Figure 1 Extractant, masking agents, and other reagents used in this studyDisplay full sizeFigure 2 The conditions of batch-wise multistage extraction. (a) Am and Cm separation, (b) Am concentration, (c) Cm concentrationDisplay full sizeFigure 3 Extraction behavior of metals using alkyl-diamide amine (ADAAM). (a) D(Ln, An) against HNO3 concentration, (b) D(An) against ADAAM concentration, (c) D(Ln) against ADAAM concentration, (d) D(Ln) against their atomic number, (e) D(FP) against HNO3 concentration, and (f) D(FP) against ADAAM concentrationDisplay full sizeFigure 4 Ball-and-stick description of [M(ADAAM)]3+, wherein the black, red, blue, and brown spheres show metal, oxygen, nitrogen, and carbon atoms, respectively, and hydrogen atoms are not shown to avoid confusion, with the M3+–Namine and M3+–Oamide bond lengths in the Å unit shown as Am3+/Cm3+Display full sizeFigure 5 Representatives of 5f-type α-MOs in the valence region for [M(ADAAM)]3+Display full size Figure 6 pH and masking agent dependence on D(An, Ln) in alkyl-diamide amine (ADA
摘要采用烷基二胺(ADAAM(EH))萃取剂对Am和Cm进行了相互分离。ADAAM(EH)在hno3 -n-十二烷体系中具有极高的Am和Cm分离系数(SF = 5.9)。采用含有0.2 M ADAAM(EH)和1.5 M HNO3的体系进行分批多级提取，Am和Cm的分配比(D)分别为2.55和0.434。在这个多级萃取中，3级进料，9级洗涤，11级有机溶剂的使用，分别得到96.5%和1.06%的Am和Cm收率。相互分离后，包括锕系元素在内的水相和有机相的体积比进料溶液增加了四倍。附加的萃取步骤是减少溶液的体积，包括锕系元素。在Cm浓度步骤中，采用四乙基二甘醇酰胺(TEDGA)/水萃取Am浓度，四辛基二甘醇酰胺/正十二烷萃取中间部分，TEDGA/水萃取。通过采取额外的步骤，Am和Cm可以在水相中只需要两到三个阶段就可以回收。关键词:溶剂萃取- amcm -分离- adaam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。作者感谢日本原子能机构的Y. Kitatsuji博士对锕系元素的安全管理和日本原子能机构的M. Arisaka博士对icp测量的研究。利益声明作者未报告潜在的利益冲突。图1本研究中使用的萃取剂、掩蔽剂和其他试剂显示完整尺寸图2分批多级萃取条件(a) Am和Cm的分离，(b) Am浓度，(c) Cm浓度。图3烷基二胺(ADAAM)萃取金属的行为。(a) D(Ln, An)对HNO3浓度的影响，(b) D(An)对ADAAM浓度的影响，(c) D(Ln)对ADAAM浓度的影响，(D) D(Ln)对它们的原子序数的影响，(e) D(FP)对HNO3浓度的影响，(f) D(FP)对ADAAM浓度的影响显示完整尺寸图4 [M(ADAAM)]3+的球棒描述，其中黑色、红色、蓝色和棕色球体分别表示金属、氧、氮和碳原子，氢原子不显示，以避免混淆。图5 [M(ADAAM)]3+价区5f型α-MOs的代表图6烷基二胺(ADAAM)萃取剂中pH和掩掩剂对D(An, Ln)的依赖条件:(a) 0.5 M ADAAM−2.5 M HNO3 + 1.3 M en + 1m MAc + 0.1 M DTBA (b) 0.5 M ADAAM−1.5 M HNO3 + 0.05 M DGAc + 0.5 M en + 1m MAc (c) 0.5 M ADAAM−1.5 M HNO3 + TMDGA或NTAamide(C2)显示全尺寸图7多级提取结果(a)水相各段An的回收率(%)(b)有机相各段An的回收率(%)萃取条件:0.2 M ADAAM/n-十二烷，1.5 M HNO3多段萃取条件:有机相、进料水相、磨砂水相的段数:11、3、9。表2 (b) ADAAM萃取后丙二酸水溶液pH值*下载csv显示表
{"title":"Mutual separation of Am and Cm using ADAAM (alkyl DiAmide AMine) and Reduction of Volumes for Liquid Waste generated via batch wise multistage extractions","authors":"Yuji Sasaki, Masashi Kaneko, Yasutoshi Ban, Hideya Suzuki","doi":"10.1080/00223131.2023.2283214","DOIUrl":"https://doi.org/10.1080/00223131.2023.2283214","url":null,"abstract":"ABSTRACTThe mutual separation of Am and Cm is conducted using an alkyl-diamide amine (ADAAM(EH)) extractant. ADAAM(EH) exhibits extremely high separation factor (SF) with respect to Am and Cm separation (SF = 5.9) in a HNO3-n-dodecane system. The batch-wise multistage extractions are performed using a system containing 0.2 M ADAAM(EH) and 1.5 M HNO3, with their distribution ratios (D) of the system for Am and Cm equal to 2.55 and 0.434, respectively. In this multistage extraction, 3 stages of feed, 9 stages involving scrubbing, and 11 stages involving the use of an organic solvent give 96.5% and 1.06% yields of Am and Cm, respectively. After the mutual separation, the aqueous and the organic phases that includes actinides increases four times in volume compared to the feed solution. Additional extraction steps are to reduce the volume of the solutions including actinides. We use tetraethyl diglycolamide (TEDGA)/water for Am concentration, tetraoctyl diglycolamide/n-dodecane for the intermediate extraction, and TEDGA/water for stripping in the Cm concentration step. Taking the additional steps, Am and Cm can be recovered in just two or three stages in the aqueous phases.KEYWORDS: Solvent extractionAmCmseparationADAAMDisclaimerAs 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 gratefully acknowledge Dr. Y. Kitatsuji at JAEA for safety management with respect to actinides and Dr. M. Arisaka at JAEA for ICP-measurements.Declaration of Interest StatementNo potential conflict of interest was reported by the authors.Figure 1 Extractant, masking agents, and other reagents used in this studyDisplay full sizeFigure 2 The conditions of batch-wise multistage extraction. (a) Am and Cm separation, (b) Am concentration, (c) Cm concentrationDisplay full sizeFigure 3 Extraction behavior of metals using alkyl-diamide amine (ADAAM). (a) D(Ln, An) against HNO3 concentration, (b) D(An) against ADAAM concentration, (c) D(Ln) against ADAAM concentration, (d) D(Ln) against their atomic number, (e) D(FP) against HNO3 concentration, and (f) D(FP) against ADAAM concentrationDisplay full sizeFigure 4 Ball-and-stick description of [M(ADAAM)]3+, wherein the black, red, blue, and brown spheres show metal, oxygen, nitrogen, and carbon atoms, respectively, and hydrogen atoms are not shown to avoid confusion, with the M3+–Namine and M3+–Oamide bond lengths in the Å unit shown as Am3+/Cm3+Display full sizeFigure 5 Representatives of 5f-type α-MOs in the valence region for [M(ADAAM)]3+Display full size Figure 6 pH and masking agent dependence on D(An, Ln) in alkyl-diamide amine (ADA","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-11-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134901633","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 : 2023-11-14DOI: 10.1080/00223131.2023.2282551
Yuki Sato, Yuta Terasaka, Masatoshi Oura
In the decommissioning of the Fukushima Daiichi Nuclear Power Station, understanding the distribution of radioactive substances and dose-equivalent rates is crucial to develop detailed decontamination plans and minimize worker exposure. In this study, we remotely visualized radioactive hotspots and dose-equivalent rate distribution in Unit 1 reactor building of the station using a Mecanum wheel robot equipped with a Compton camera, simultaneous localization and mapping device, and survey meter. We successfully visualized high-concentration radioactive hotspots on the U-shaped piping of the drywell humidity control system and the atmospheric control piping in the ceiling in front of the transverse in-core probe room. Furthermore, the hotspot location was identified in three dimensions using the Compton camera used to analyze the atmospheric control piping. By simultaneously analyzing the dose-equivalent rate data acquired by the survey meter and the hotspot locations visualized by the Compton camera, it was confirmed that the hotspots caused elevated dose-equivalent rates in the surrounding area. In the future, if this robotic system is used in unexplored areas, such as the upper floors of reactor buildings, it can provide information about the locations of radioactive hotspots and the distribution of dose-equivalent rates.
{"title":"Detailed visualization of radioactive hotspots inside the unit 1 reactor building of the Fukushima Daiichi Nuclear Power Station using an integrated radiation imaging system mounted on a mecanum wheel robot","authors":"Yuki Sato, Yuta Terasaka, Masatoshi Oura","doi":"10.1080/00223131.2023.2282551","DOIUrl":"https://doi.org/10.1080/00223131.2023.2282551","url":null,"abstract":"In the decommissioning of the Fukushima Daiichi Nuclear Power Station, understanding the distribution of radioactive substances and dose-equivalent rates is crucial to develop detailed decontamination plans and minimize worker exposure. In this study, we remotely visualized radioactive hotspots and dose-equivalent rate distribution in Unit 1 reactor building of the station using a Mecanum wheel robot equipped with a Compton camera, simultaneous localization and mapping device, and survey meter. We successfully visualized high-concentration radioactive hotspots on the U-shaped piping of the drywell humidity control system and the atmospheric control piping in the ceiling in front of the transverse in-core probe room. Furthermore, the hotspot location was identified in three dimensions using the Compton camera used to analyze the atmospheric control piping. By simultaneously analyzing the dose-equivalent rate data acquired by the survey meter and the hotspot locations visualized by the Compton camera, it was confirmed that the hotspots caused elevated dose-equivalent rates in the surrounding area. In the future, if this robotic system is used in unexplored areas, such as the upper floors of reactor buildings, it can provide information about the locations of radioactive hotspots and the distribution of dose-equivalent rates.","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-11-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134901478","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}
ABSTRACTPreviously we reported a simple algorithmic method of spectral determination method (SDM), which is based on the first principle that a γ-ray spectrum obtained for a sample is a linear superposition of individual spectra of the radioactive nuclides included in the sample and demonstrated that the method is valid for γ-ray determination. Here we apply it to the spectra obtained by liquid scintillation counter (LSC). In LSC measurements quenching is generally observed and we at first developed its correction method to standard spectra. The SDM code reported in the previous investigation is used to analyze the LSC spectra. Based on the analyses done by using the measured spectra, we concluded that the SDM method is valid in the LSC spectra similarly to the γ-ray spectra studied in the previous investigation.KEYWORDS: Decomposition of a liquid-scintillation-counter spectrumfull-spectrum analysisleast-squares fitquenching correctiondetermination accuracyDisclaimerAs 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 JAEA Nuclear Energy S&T and Human Resource Development Project through concentrating wisdom Grant Number JPJA20P20333366.
{"title":"Application of the spectral determination method to liquid scintillation spectra","authors":"Masumi Oshima, Jun Goto, Takehito Hayakawa, Hirofumi Shinohara, Katsuyuki Suzuki, Yuichi Sano, Masato Asai, Tomoko Haraga","doi":"10.1080/00223131.2023.2282552","DOIUrl":"https://doi.org/10.1080/00223131.2023.2282552","url":null,"abstract":"ABSTRACTPreviously we reported a simple algorithmic method of spectral determination method (SDM), which is based on the first principle that a γ-ray spectrum obtained for a sample is a linear superposition of individual spectra of the radioactive nuclides included in the sample and demonstrated that the method is valid for γ-ray determination. Here we apply it to the spectra obtained by liquid scintillation counter (LSC). In LSC measurements quenching is generally observed and we at first developed its correction method to standard spectra. The SDM code reported in the previous investigation is used to analyze the LSC spectra. Based on the analyses done by using the measured spectra, we concluded that the SDM method is valid in the LSC spectra similarly to the γ-ray spectra studied in the previous investigation.KEYWORDS: Decomposition of a liquid-scintillation-counter spectrumfull-spectrum analysisleast-squares fitquenching correctiondetermination accuracyDisclaimerAs 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 JAEA Nuclear Energy S&T and Human Resource Development Project through concentrating wisdom Grant Number JPJA20P20333366.","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-11-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"136346660","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 : 2023-11-13DOI: 10.1080/00223131.2023.2282550
Ke Sun, Zhitao Tian, Yingqi Fan, Huawei Lu, Jianchix Xin
ABSTRACTThe highly-loaded design method of helium compressors can effectively solve the difficulty in compressing helium in High Temperature Gas-cooled Reactors (HTGR). But it also causes obviously different attack angle characteristics of blade surface loads in a highly-loaded helium compressor compared to air compressors. This difference inevitably affects separation characteristics and flow loss within the compressor. In the current study, the effects of highly-loaded design methods and changes in attack angle on the separation characteristics of the compressor cascade are analyzed by applying a numerical simulation method firstly. Then the influence of Mach number on the loss characteristics of the cascade for a highly-loaded helium compressor is systematically analyzed. Finally, the effect of differences in the material properties of working fluid on the separation characteristics is discussed. The results indicate that the proportion of secondary flow loss to the total loss in highly-loaded compressor cascades is 2.46 times larger than that in conventionally-loaded ones. While properties of working fluid have an effect on the performance of the compressor cascade, their effects on the weight factor of vortex loss are highly limited.KEYWORDS: Helium compressorHigh temperature gas-cooled reactor (HTGR)HTGR type reactorhighly-loaded design methodSecondary flow lossTurbineClosed Brayton cycleSpecific heat ratioComputational fluid dynamicsDisclaimerAs 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 work is financially supported by the National Natural Science Foundation of China (No. 52206042), Natural Science Foundation of Liaoning Province (No. 2022-BS-096), Basic Research Project of Liaoning Provincial Department of Education(LJKMZ20220364).NomenclatureTableDisplay TableDeclaration of interest statementNo potential conflict of interest was reported by the author(s).Figure 1 Construction of highly-loaded helium compressor velocity triangleDisplay full sizeFigure 2 Schematic diagram of the geometric parameters of cascadesDisplay full sizeFigure 3 Schematic diagram of S3 sectional positionDisplay full sizeFigure 4 Schematic diagram of the calculation meshDisplay full sizeFigure 5 Axial vorticity diagrams for different grid numbers of exit sectionsDisplay full sizeFigure 6 Validation of numerical simulation methodDisplay full sizeFigure 7 Variation of total loss coefficients and pressure ratios with attack angle for helium compressor cascades of different design methodsDisplay full sizeFigure 8 Static pressure curves of blade surface with 50% blad
{"title":"Spatial evolution mechanism of vortex structure in the highly-loaded helium compressor cascade","authors":"Ke Sun, Zhitao Tian, Yingqi Fan, Huawei Lu, Jianchix Xin","doi":"10.1080/00223131.2023.2282550","DOIUrl":"https://doi.org/10.1080/00223131.2023.2282550","url":null,"abstract":"ABSTRACTThe highly-loaded design method of helium compressors can effectively solve the difficulty in compressing helium in High Temperature Gas-cooled Reactors (HTGR). But it also causes obviously different attack angle characteristics of blade surface loads in a highly-loaded helium compressor compared to air compressors. This difference inevitably affects separation characteristics and flow loss within the compressor. In the current study, the effects of highly-loaded design methods and changes in attack angle on the separation characteristics of the compressor cascade are analyzed by applying a numerical simulation method firstly. Then the influence of Mach number on the loss characteristics of the cascade for a highly-loaded helium compressor is systematically analyzed. Finally, the effect of differences in the material properties of working fluid on the separation characteristics is discussed. The results indicate that the proportion of secondary flow loss to the total loss in highly-loaded compressor cascades is 2.46 times larger than that in conventionally-loaded ones. While properties of working fluid have an effect on the performance of the compressor cascade, their effects on the weight factor of vortex loss are highly limited.KEYWORDS: Helium compressorHigh temperature gas-cooled reactor (HTGR)HTGR type reactorhighly-loaded design methodSecondary flow lossTurbineClosed Brayton cycleSpecific heat ratioComputational fluid dynamicsDisclaimerAs 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 work is financially supported by the National Natural Science Foundation of China (No. 52206042), Natural Science Foundation of Liaoning Province (No. 2022-BS-096), Basic Research Project of Liaoning Provincial Department of Education(LJKMZ20220364).NomenclatureTableDisplay TableDeclaration of interest statementNo potential conflict of interest was reported by the author(s).Figure 1 Construction of highly-loaded helium compressor velocity triangleDisplay full sizeFigure 2 Schematic diagram of the geometric parameters of cascadesDisplay full sizeFigure 3 Schematic diagram of S3 sectional positionDisplay full sizeFigure 4 Schematic diagram of the calculation meshDisplay full sizeFigure 5 Axial vorticity diagrams for different grid numbers of exit sectionsDisplay full sizeFigure 6 Validation of numerical simulation methodDisplay full sizeFigure 7 Variation of total loss coefficients and pressure ratios with attack angle for helium compressor cascades of different design methodsDisplay full sizeFigure 8 Static pressure curves of blade surface with 50% blad","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-11-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"136346289","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}
ABSTRACTTo systematically study the methyl iodide (CH3I) adsorption performance and adsorption mechanism of Ag ion exchange NaY zeolite (AgY), AgY adsorbents with different Ag content were selected. 16.6% AgY was capable of storing 150 mg/g and 275 mg/g of CH3I when the penetrated CH3I reached 1‰ and 100% inlet concentration. TEM and EDS results presented a high relationship between Ag dispersion and CH3I adsorption capability. The formation of a stable AgI complex rationalized the CH3I adsorption mechanism on AgY. Compared to AgX, AgY was verified to possess a strong acid resistance by FT-IR, SEM and XRD. The effect of three key parameters (temperature, humidity and gas velocity) on CH3I adsorption was carefully investigated by orthogonal experiment design. The experimental results showed that AgY had better iodine removal performance at a high temperature, low humidity, and low flow rate. This work provides technical reference for further engineering applications of 16.6%AgY in spent fuel reprocessing plants.KEYWORDS: Zeolitespent fuelradioactive iodineadsorption AcknowledgementThis research was supported by the State Administration of Science, Technology and Industry for National Defence (KY20007) and the National Natural Science Foundation of China (U1967215).Disclosure statementNo potential conflict of interest was reported by the author(s).Additional informationFundingThis work was supported by the the National Natural Science Foundation of China [U1967215]; the State Administration of Science, Technology and Industry for National Defense [KY20007].
{"title":"Study on iodine adsorption capacity and adsorption mechanism of AgY","authors":"Cheng Zhang, Zexiang Chen, Jianrong Hou, Hongzheng Ren, Zhen Wang, Xin Li, Yongguo Li, Haixia Kong, Haifeng Yu, Jie Yu","doi":"10.1080/00223131.2023.2276397","DOIUrl":"https://doi.org/10.1080/00223131.2023.2276397","url":null,"abstract":"ABSTRACTTo systematically study the methyl iodide (CH3I) adsorption performance and adsorption mechanism of Ag ion exchange NaY zeolite (AgY), AgY adsorbents with different Ag content were selected. 16.6% AgY was capable of storing 150 mg/g and 275 mg/g of CH3I when the penetrated CH3I reached 1‰ and 100% inlet concentration. TEM and EDS results presented a high relationship between Ag dispersion and CH3I adsorption capability. The formation of a stable AgI complex rationalized the CH3I adsorption mechanism on AgY. Compared to AgX, AgY was verified to possess a strong acid resistance by FT-IR, SEM and XRD. The effect of three key parameters (temperature, humidity and gas velocity) on CH3I adsorption was carefully investigated by orthogonal experiment design. The experimental results showed that AgY had better iodine removal performance at a high temperature, low humidity, and low flow rate. This work provides technical reference for further engineering applications of 16.6%AgY in spent fuel reprocessing plants.KEYWORDS: Zeolitespent fuelradioactive iodineadsorption AcknowledgementThis research was supported by the State Administration of Science, Technology and Industry for National Defence (KY20007) and the National Natural Science Foundation of China (U1967215).Disclosure statementNo potential conflict of interest was reported by the author(s).Additional informationFundingThis work was supported by the the National Natural Science Foundation of China [U1967215]; the State Administration of Science, Technology and Industry for National Defense [KY20007].","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-11-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135037313","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}
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":null,"pages":null},"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":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":"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":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":"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":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":"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}
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