Pub Date : 2025-02-19DOI: 10.1016/j.anucene.2025.111262
S.N. Nazrin , Seema Thakur , Camellia Doroody , Halimah Badioze Zaman , Siti Nur Aida Mohd Nashruddin , Mongi Amami , Faznny Mohd Fudzi
The melt-quenching approach was used to produce Gd2O3-doped tellurium zinc borate glass system with composition {[(TeO2)0.7(B2O3)0.3]0.7(ZnO)0.3}1−x(Gd2O3)x where x = 0.01, 0.02, 0.03, 0.04 as well as 0.05 M fraction and assess its optical and radiation shielding characteristics. Phy-X/PSD software was used to calculate parameters such as mass attenuation coefficient (MAC), linear attenuation coefficient (LAC), half value layer (HVL), tenth value layer (TVL), and mean free path (MFP) in order to evaluate the effectiveness of radiation shielding. Glass containing 0.05 M fraction of Gd2O3 exhibited the greatest shielding efficacy, surpassing conventional concretes such as chromite and serpentine. The refractive index dropped from 2.518 to 2.352 while the direct and indirect optical band gaps increased from 3.239 to 3.519 eV and from 2.587 to 3.172 eV, respectively. The amorphous nature of the glass was validated by XRD, while TeO3, TeO4, BO3, and BO4 vibrational groups were identified by infrared spectra demonstrating the potential of the prepared glass for optical and radiation shielding applications.
采用熔淬法制备了掺杂 Gd2O3 的硼酸碲锌玻璃体系，其组成为 {[(TeO2)0.7(B2O3)0.3]0.7(ZnO)0.3}1-x(Gd2O3)x 其中 x = 0.01、0.02、0.03、0.04 以及 0.05 M 分数，并对其光学和辐射屏蔽特性进行了评估。使用 Phy-X/PSD 软件计算质量衰减系数 (MAC)、线性衰减系数 (LAC)、半值层 (HVL)、十值层 (TVL) 和平均自由路径 (MFP) 等参数，以评估辐射屏蔽的有效性。含 0.05 M 分量 Gd2O3 的玻璃显示出最大的屏蔽效果，超过了铬铁矿和蛇纹石等传统混凝土。折射率从 2.518 降至 2.352，直接和间接光带隙分别从 3.239 eV 和 2.587 eV 增至 3.519 eV 和 3.172 eV。XRD 验证了这种玻璃的非晶性质，而红外光谱则确定了 TeO3、TeO4、BO3 和 BO4 的振动基团，这表明制备的玻璃具有光学和辐射屏蔽应用的潜力。
{"title":"Investigation of optical and radiation shielding characteristics of Gd2O3-doped tellurium zinc borate glass via melt-quenching method","authors":"S.N. Nazrin , Seema Thakur , Camellia Doroody , Halimah Badioze Zaman , Siti Nur Aida Mohd Nashruddin , Mongi Amami , Faznny Mohd Fudzi","doi":"10.1016/j.anucene.2025.111262","DOIUrl":"10.1016/j.anucene.2025.111262","url":null,"abstract":"<div><div>The melt-quenching approach was used to produce Gd<sub>2</sub>O<sub>3</sub>-doped tellurium zinc borate glass system with composition {[(TeO<sub>2</sub>)<sub>0.7</sub>(B<sub>2</sub>O<sub>3</sub>)<sub>0.3</sub>]<sub>0.7</sub>(ZnO)<sub>0.3</sub>}<sub>1−</sub><em><sub>x</sub></em>(Gd<sub>2</sub>O<sub>3</sub>)<em><sub>x</sub></em> where <em>x</em> = 0.01, 0.02, 0.03, 0.04 as well as 0.05 M fraction and assess its optical and radiation shielding characteristics. Phy-X/PSD software was used to calculate parameters such as mass attenuation coefficient (MAC), linear attenuation coefficient (LAC), half value layer (HVL), tenth value layer (TVL), and mean free path (MFP) in order to evaluate the effectiveness of radiation shielding. Glass containing 0.05 M fraction of Gd<sub>2</sub>O<sub>3<!--> </sub>exhibited the greatest shielding efficacy, surpassing conventional concretes such as chromite and serpentine. The refractive index dropped from 2.518 to 2.352 while the direct and indirect optical band gaps increased from 3.239 to 3.519 eV and from 2.587 to 3.172 eV, respectively. The amorphous nature of the glass was validated by XRD, while TeO<sub>3</sub>, TeO<sub>4</sub>, BO<sub>3</sub>, and BO<sub>4</sub> vibrational groups were identified by infrared spectra demonstrating<!--> <!-->the<!--> <!-->potential of the prepared glass for optical and radiation shielding applications.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111262"},"PeriodicalIF":1.9,"publicationDate":"2025-02-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143445447","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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 : 2025-02-19DOI: 10.1016/j.anucene.2025.111278
Yuan Xiaoxiao, Li Guodong, Yuhao Zhang, Zhao Haiqi, Lu Daogang
The Fast Flux Test Facility (FFTF) is a 400MWth loop-type fast neutron reactor, in which 13 sets of experiments are loss of flow without scram (LOFWOS) accident conditions without emergency shutdown. The safe shutdown behavior of FFTF in emergency state is one of the focuses of its safety assessment. In the accident condition, a natural circulation is developed in the reactor, but due to the complex core structure, the characteristics of the natural circulation are not clear yet. The system analysis code can simulate the overall change of parameters over time during the accident process, but it can hardly obtain the detailed three-dimensional thermal–hydraulic phenomena in the sodium pool. Therefore, based on porous media and detailed fuel bundles modeling methods pairs, the natural circulation characteristics in FFTF were simulated with reduced number of grids. Then two kinds of numerical simulations were conducted on FFTF LOFWOS Test#13 benchmark experiment, so that the applicability and uncertainty of above modeling methods can be revealed by the comparison results. It indicates that overall thermal–hydraulic characteristics could be obtained by both simplified models. However, it is difficult for porous media to capture detailed parameters, while more detailed inter-wrapper flow and natural circulation can be observed inside the core for the fuel bundle model, as well as more precise core temperature distribution, with the average error less than 5.0%. These findings not only provide guidance for optimizing calculation and modeling methods, but also reveal key thermal–hydraulic characteristics for the reactor under the loop type unprotected loss accident.
{"title":"Three-dimensional numerical simulation on natural circulation characteristics in FFTF under unprotected loss of flow accident","authors":"Yuan Xiaoxiao, Li Guodong, Yuhao Zhang, Zhao Haiqi, Lu Daogang","doi":"10.1016/j.anucene.2025.111278","DOIUrl":"10.1016/j.anucene.2025.111278","url":null,"abstract":"<div><div>The Fast Flux Test Facility (FFTF) is a 400MW<sub>th</sub> loop-type fast neutron reactor, in which 13 sets of experiments are loss of flow without scram (LOFWOS) accident conditions without emergency shutdown. The safe shutdown behavior of FFTF in emergency state is one of the focuses of its safety assessment. In the accident condition, a natural circulation is developed in the reactor, but due to the complex core structure, the characteristics of the natural circulation are not clear yet. The system analysis code can simulate the overall change of parameters over time during the accident process, but it can hardly obtain the detailed three-dimensional thermal–hydraulic phenomena in the sodium pool. Therefore, based on porous media and detailed fuel bundles modeling methods pairs, the natural circulation characteristics in FFTF were simulated with reduced number of grids. Then two kinds of numerical simulations were conducted on FFTF LOFWOS Test#13 benchmark experiment, so that the applicability and uncertainty of above modeling methods can be revealed by the comparison results. It indicates that overall thermal–hydraulic characteristics could be obtained by both simplified models. However, it is difficult for porous media to capture detailed parameters, while more detailed inter-wrapper flow and natural circulation can be observed inside the core for the fuel bundle model, as well as more precise core temperature distribution, with the average error less than 5.0%. These findings not only provide guidance for optimizing calculation and modeling methods, but also reveal key thermal–hydraulic characteristics for the reactor under the loop type unprotected loss accident.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111278"},"PeriodicalIF":1.9,"publicationDate":"2025-02-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143436653","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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 : 2025-02-18DOI: 10.1016/j.anucene.2025.111280
Quan Han , Yuntao Song , Kun Lu , Weibin Xi , Chen Liu , Kaizhong Ding , Qingquan Zhang , Xufeng Liu , Shuangsong Du , Jinxing Zheng
The primary function of CFETR magnet current feeder is to supply and discharge the current for the magnets. The current feeder is directly related to the safe operation of the magnets and has important research significance. In this paper, the design structure of the CFETR magnet current feeder was presented. The engineering design of main components has been nearly completed, including high temperature superconducting current leads(HTSCLs） and busbars, and other critical components. Key analyses and optimization results are provided, followed by discussions and conclusions. The main analysis and optimization are as follows. First, the cooling scheme of the current feeder was analyzed and determined. Second, the mechanical analysis of the current feeder was conducted, confirming that its structure meets the mechanical strength requirements. Third, an electrical performance analysis of the insulation structure was carried out, verifying that its insulation properties satisfy the design specifications. Fourth, the heat load analysis was performed, determining both the static and dynamic heat loads. Last, the test of 55 kA HTSCL was performed to verify some analysis results. The design and analysis of magnet current feeder will provide technical guidance for the design, manufacture and testing of CFETR feeder system in the future.
{"title":"Design and analysis of CFETR magnet current feeder","authors":"Quan Han , Yuntao Song , Kun Lu , Weibin Xi , Chen Liu , Kaizhong Ding , Qingquan Zhang , Xufeng Liu , Shuangsong Du , Jinxing Zheng","doi":"10.1016/j.anucene.2025.111280","DOIUrl":"10.1016/j.anucene.2025.111280","url":null,"abstract":"<div><div>The primary function of CFETR magnet current feeder is to supply and discharge the current for the magnets. The current feeder is directly related to the safe operation of the magnets and has important research significance. In this paper, the design structure of the CFETR magnet current feeder was presented. The engineering design of main components has been nearly completed, including high temperature superconducting current leads(HTSCLs） and busbars, and other critical components. Key analyses and optimization results are provided, followed by discussions and conclusions. The main analysis and optimization are as follows. First, the cooling scheme of the current feeder was analyzed and determined. Second, the mechanical analysis of the current feeder was conducted, confirming that its structure meets the mechanical strength requirements. Third, an electrical performance analysis of the insulation structure was carried out, verifying that its insulation properties satisfy the design specifications. Fourth, the heat load analysis was performed, determining both the static and dynamic heat loads. Last, the test of 55 kA HTSCL was performed to verify some analysis results. The design and analysis of magnet current feeder will provide technical guidance for the design, manufacture and testing of CFETR feeder system in the future.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111280"},"PeriodicalIF":1.9,"publicationDate":"2025-02-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143428658","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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 : 2025-02-18DOI: 10.1016/j.anucene.2025.111254
Yandong Hou , Yiliang Dong , Chuntian Gao , Bowen Chen , Chao Zhang , Weichao Li , Yan Xiang
Helical Cruciform Fuel (HCF) embodies advancement in the fusion of unique geometric design with state-of-the-art metallic alloy materials. This innovative design leverages the optimized heat transfer characteristics of its distinctive geometry to potentially achieve elevated power output levels. Additionally, the employment of U-50Zr fuel contributes significantly to reducing the risk of potential accidents. The operation of nuclear fuel is a typical multi physics process, and accurate evaluation and prediction require advanced research methods. The open-source, parallel finite element framework MOOSE, a renowned software platform, is integral to the effective modeling and simulation of these intricate processes. Based on the MOOSE framework, simulate the operational behavior of HCF under high burnup conditions in pressurized water reactor environment and challenging scenarios of loss of coolant accident (LOCA). The calculation results indicate that U-10Zr experiences excessive swelling during the initial burnup period, and stress will concentrate at the concave arc position of the cladding. The swelling of U-50Zr gradually increases with stress, rendering it a more suitable alternative fuel for HCF. During LOCA accidents, the mechanical behavior of the fuel assembly, particularly the cladding, undergoes a sharp decrease in stress after an increase.
Notably, the minimum axial stress post-cladding stress drop occurs near the central height. Furthermore, the two axial helices of the concave and convex arcs of the cladding exhibit opposing characteristics during such accidents. A comparative analysis between LB-LOCA and SB-LOCA reveals a significant lag in the reduction of cladding stress in the case of SB-LOCA.
{"title":"Performance analysis of U-50Zr helical cruciform fuel during loss-of-coolant accidents Based on MOOSE framework","authors":"Yandong Hou , Yiliang Dong , Chuntian Gao , Bowen Chen , Chao Zhang , Weichao Li , Yan Xiang","doi":"10.1016/j.anucene.2025.111254","DOIUrl":"10.1016/j.anucene.2025.111254","url":null,"abstract":"<div><div>Helical Cruciform Fuel (HCF) embodies advancement in the fusion of unique geometric design with state-of-the-art metallic alloy materials. This innovative design leverages the optimized heat transfer characteristics of its distinctive geometry to potentially achieve elevated power output levels. Additionally, the employment of U-50Zr fuel contributes significantly to reducing the risk of potential accidents. The operation of nuclear fuel is a typical multi physics process, and accurate evaluation and prediction require advanced research methods. The open-source, parallel finite element framework MOOSE, a renowned software platform, is integral to the effective modeling and simulation of these intricate processes. Based on the MOOSE framework, simulate the operational behavior of HCF under high burnup conditions in pressurized water reactor environment and challenging scenarios of loss of coolant accident (LOCA). The calculation results indicate that U-10Zr experiences excessive swelling during the initial burnup period, and stress will concentrate at the concave arc position of the cladding. The swelling of U-50Zr gradually increases with stress, rendering it a more suitable alternative fuel for HCF. During LOCA accidents, the mechanical behavior of the fuel assembly, particularly the cladding, undergoes a sharp decrease in stress after an increase.</div><div>Notably, the minimum axial stress post-cladding stress drop occurs near the central height. Furthermore, the two axial helices of the concave and convex arcs of the cladding exhibit opposing characteristics during such accidents. A comparative analysis between LB-LOCA and SB-LOCA reveals a significant lag in the reduction of cladding stress in the case of SB-LOCA.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111254"},"PeriodicalIF":1.9,"publicationDate":"2025-02-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143428659","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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 : 2025-02-18DOI: 10.1016/j.anucene.2025.111273
Yong Yang , Fanwei Wang , Lingjiao Kang , Huairong Zhou , Dongliang Wang , Zongliang Fan
The utilization of spent nuclear energy inevitably generates high-level waste (HLW), and the development of safe and efficient treatment technologies for HLW constitutes a significant bottleneck in advancing nuclear energy cycles and scaling the associated industries. Among the available treatment methods, HLW vitrification in glass stands out as the only technology that has achieved practical engineering application. This article provides a comprehensive review of the current HLW vitrification state of glass formulations, the HLW melting process, and the Joule Heating Ceramic Melter (JHCM). The design and optimization of glass formulations serve as the foundational step in HLW vitrification. Borosilicate and phosphate glasses are the most widely employed glass matrices; however, the formulation tailored to specific HLW streams remains an area requiring urgent development and optimization. Substantial progress has been made in understanding the physicochemical reactions and heat and mass transfer characteristics that occur during the HLW melting process. Notably, the cold cap zone, a critical component of the melting process, presents considerable opportunities for enhancement through both experimental investigations and numerical modeling. JHCM encompasses two primary configurations for batch and continuous furnaces. While mathematical modeling has been undertaken across various scales, there remains a pressing need for the development and application of suitable furnace bricks and electrode materials. Furthermore, multi-scale coupled JHCM models require further refinement to facilitate the JHCM design and process optimization. Such progress will provide essential support data and modeling frameworks for the practical HLW vitrification.
{"title":"Research progress on high-level waste vitrification based on Joule heating ceramic melter","authors":"Yong Yang , Fanwei Wang , Lingjiao Kang , Huairong Zhou , Dongliang Wang , Zongliang Fan","doi":"10.1016/j.anucene.2025.111273","DOIUrl":"10.1016/j.anucene.2025.111273","url":null,"abstract":"<div><div>The utilization of spent nuclear energy inevitably generates high-level waste (HLW), and the development of safe and efficient treatment technologies for HLW constitutes a significant bottleneck in advancing nuclear energy cycles and scaling the associated industries. Among the available treatment methods, HLW vitrification in glass stands out as the only technology that has achieved practical engineering application. This article provides a comprehensive review of the current HLW vitrification state of glass formulations, the HLW melting process, and the Joule Heating Ceramic Melter (JHCM). The design and optimization of glass formulations serve as the foundational step in HLW vitrification. Borosilicate and phosphate glasses are the most widely employed glass matrices; however, the formulation tailored to specific HLW streams remains an area requiring urgent development and optimization. Substantial progress has been made in understanding the physicochemical reactions and heat and mass transfer characteristics that occur during the HLW melting process. Notably, the cold cap zone, a critical component of the melting process, presents considerable opportunities for enhancement through both experimental investigations and numerical modeling. JHCM encompasses two primary configurations for batch and continuous furnaces. While mathematical modeling has been undertaken across various scales, there remains a pressing need for the development and application of suitable furnace bricks and electrode materials. Furthermore, multi-scale coupled JHCM models require further refinement to facilitate the JHCM design and process optimization. Such progress will provide essential support data and modeling frameworks for the practical HLW vitrification.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111273"},"PeriodicalIF":1.9,"publicationDate":"2025-02-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143436652","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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 : 2025-02-17DOI: 10.1016/j.anucene.2025.111276
Aljawhara H. Almuqrin , M.I Sayyed , M. Elsafi
This work investigated the gamma-ray attenuation capability of four glass compositions labeled as BTBZC-1, BTBZC-2, BTBZC-3 and BTBZC-4, with fixed amount of Bi2O3 and TeO2 (5 and 20 mol%, respectively), while the CaO and B2O3 decrease from 17 to 8 mol% (in 3 mol% decrement) and 38 to 32 mol% (in 2 mol% decrement), and the ZnO increases from 20 to 35 mol% (in 5 mol% increment). The data was evaluated experimentally using narrow beam method by different point sources and semiconductor detector (HPGe-detector). Experimental data are compared with theoretical predictions (Phy-X software), demonstrating strong agreement with R2 values of 1. The LAC at 0.059 MeV is in the 12.332–13.146 cm−1 range, decreasing at 0.662 MeV to 0.331–0.345 cm−1, thus revealing the prepared glasses to exhibit high shielding efficiency at low energies, primarily due to the photoelectric effect’s dominance, while Compton scattering at higher energies leads to reduced attenuation. The addition of ZnO improves the shielding performance by increasing glass density and Zeff, as confirmed by lower HVL, TVL, and MFP values. The investigation highlights these glasses’ potential for radiation protection, with BTBZC-4 exhibiting superior performance due to its higher ZnO content.
{"title":"Experimental investigation for radiation shielding performance of B2O3-TeO2-Bi2O3-ZnO-CaO glass system","authors":"Aljawhara H. Almuqrin , M.I Sayyed , M. Elsafi","doi":"10.1016/j.anucene.2025.111276","DOIUrl":"10.1016/j.anucene.2025.111276","url":null,"abstract":"<div><div>This work investigated the gamma-ray attenuation capability of four glass compositions labeled as BTBZC-1, BTBZC-2, BTBZC-3 and BTBZC-4, with fixed amount of Bi<sub>2</sub>O<sub>3</sub> and TeO<sub>2</sub> (5 and 20 mol%, respectively), while the CaO and B<sub>2</sub>O<sub>3</sub> decrease from 17 to 8 mol% (in 3 mol% decrement) and 38 to 32 mol% (in 2 mol% decrement), and the ZnO increases from 20 to 35 mol% (in 5 mol% increment). The data was evaluated experimentally using narrow beam method by different point sources and semiconductor detector (HPGe-detector). Experimental data are compared with theoretical predictions (Phy-X software), demonstrating strong agreement with R<sup>2</sup> values of 1. The LAC at 0.059 MeV is in the 12.332–13.146 cm<sup>−1</sup> range, decreasing at 0.662 MeV to 0.331–0.345 cm<sup>−1</sup>, thus revealing the prepared glasses to exhibit high shielding efficiency at low energies, primarily due to the photoelectric effect’s dominance, while Compton scattering at higher energies leads to reduced attenuation. The addition of ZnO improves the shielding performance by increasing glass density and Zeff, as confirmed by lower HVL, TVL, and MFP values. The investigation highlights these glasses’ potential for radiation protection, with BTBZC-4 exhibiting superior performance due to its higher ZnO content.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111276"},"PeriodicalIF":1.9,"publicationDate":"2025-02-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143428655","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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 : 2025-02-17DOI: 10.1016/j.anucene.2025.111275
Rasito Tursinah , Sidik Permana , Zaki Su’ud , Alan Maulana , Abu Khalid Rivai , Muhammad Subekti , Suyatno , Achmad Ramdhani , Fahrurrozi Akbar , M. Refai Muslih , Wahyudi , Istanto , Djoko Prakoso , M.F. Ramadhani , Tri C. Laksono , Bunawas
Neutron spectrum characterization of RSG-GAS neutron radiography was carried out using a single-moderator-based neutron spectrometer with a passive type Au and In activation foils neutron detector. The neutron spectrum was obtained using Unfolding Maxed-Gravel (UMG) technique based on the activity of 198Au and 116mIn after irradiation. The neutron spectrum obtained was then compared with the simulation results using the MCNPX 2.7 program. Neutron spectrum measurements using In foil provide flux values that are closer to the simulation results, there is a difference with the simulation of 25 %. The total neutron flux is (6.65 ± 0.08) × 106 n/cm2·s with thermal neutron of 83 %. The results of neutron spectrum characterization using measurements and simulations show that the RSG-GAS radiography neutron beam has a spectrum dominated by thermal neutrons with a flux of the order of 106 n/cm2·s so it is very suitable as a thermal radiography neutron facility.
{"title":"Characterization of neutron spectrum at the neutron radiography of RSG-GAS reactor using a passive single-cylindrical neutron spectrometer","authors":"Rasito Tursinah , Sidik Permana , Zaki Su’ud , Alan Maulana , Abu Khalid Rivai , Muhammad Subekti , Suyatno , Achmad Ramdhani , Fahrurrozi Akbar , M. Refai Muslih , Wahyudi , Istanto , Djoko Prakoso , M.F. Ramadhani , Tri C. Laksono , Bunawas","doi":"10.1016/j.anucene.2025.111275","DOIUrl":"10.1016/j.anucene.2025.111275","url":null,"abstract":"<div><div>Neutron spectrum characterization of RSG-GAS neutron radiography was carried out using a single-moderator-based neutron spectrometer with a passive type Au and In activation foils neutron detector. The neutron spectrum was obtained using Unfolding Maxed-Gravel (UMG) technique based on the activity of<!--> <sup>198</sup>Au and<!--> <sup>116m</sup>In after irradiation. The neutron spectrum obtained was then compared with the simulation results using the MCNPX 2.7 program. Neutron spectrum measurements using In foil provide flux values that are closer to the simulation results, there is a difference with the simulation of 25 %. The total neutron flux is (6.65 ± 0.08) × 10<sup>6</sup> <!-->n/cm<sup>2</sup>·s with thermal neutron of 83 %. The results of neutron spectrum characterization using measurements and simulations show that the RSG-GAS radiography neutron beam has a spectrum dominated by thermal neutrons with a flux of the order of 10<sup>6</sup> <!-->n/cm<sup>2</sup>·s so it is very suitable as a thermal radiography neutron facility.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111275"},"PeriodicalIF":1.9,"publicationDate":"2025-02-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143428657","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In this paper, the tensor Canonical-Polyadic (CP) decomposition method is applied to represent few-group homogenized cross sections. The results of the method are analyzed when applied to two assemblies typical of VVER reactor (in Russian Vodo-Vodyanoi Energetichesky Reaktor meaning Water-Water Power Reactor). In particular, the presented analysis focuses on the 22UA assembly, represented by a dataset of 2-group assembly-homogenized cross sections, and on the 430GO assembly, represented by a dataset of 20-group pin-by-pin homogenized cross sections. Both assemblies’ specifications are taken from the X2VVER benchmark.
Errors in microscopic and macroscopic cross sections, are evaluated to assess the accuracy of the method to reproduce reference few-group homogenized cross sections. Furthermore, the reactivity error as well the error on fission-production and on absorption rate distributions are discussed by two fictive simple 3-D colorset simulations to evaluate the impact of the reconstructed cross sections on reactor core simulations. Additionally, the potential effectiveness of incorporating the CP decomposition into standard two-step neutronics schemes is studied and analyzed to effectively take into account the thermal-hydraulics feedback during core simulations. Preliminary results demonstrate the capability of the method to sensibly reduce the data storage as well as the memory usage and the computational time during cross-sections’ update.
{"title":"Representation of homogenized cross section data by Canonical-Polyadic decomposition","authors":"Dinh Quoc Dang Nguyen, Emiliano Masiello, Daniele Tomatis","doi":"10.1016/j.anucene.2025.111244","DOIUrl":"10.1016/j.anucene.2025.111244","url":null,"abstract":"<div><div>In this paper, the tensor Canonical-Polyadic (CP) decomposition method is applied to represent few-group homogenized cross sections. The results of the method are analyzed when applied to two assemblies typical of VVER reactor (in Russian Vodo-Vodyanoi Energetichesky Reaktor meaning Water-Water Power Reactor). In particular, the presented analysis focuses on the 22UA assembly, represented by a dataset of 2-group assembly-homogenized cross sections, and on the 430GO assembly, represented by a dataset of 20-group pin-by-pin homogenized cross sections. Both assemblies’ specifications are taken from the X2VVER benchmark.</div><div>Errors in microscopic and macroscopic cross sections, are evaluated to assess the accuracy of the method to reproduce reference few-group homogenized cross sections. Furthermore, the reactivity error as well the error on fission-production and on absorption rate distributions are discussed by two fictive simple 3-D colorset simulations to evaluate the impact of the reconstructed cross sections on reactor core simulations. Additionally, the potential effectiveness of incorporating the CP decomposition into standard two-step neutronics schemes is studied and analyzed to effectively take into account the thermal-hydraulics feedback during core simulations. Preliminary results demonstrate the capability of the method to sensibly reduce the data storage as well as the memory usage and the computational time during cross-sections’ update.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111244"},"PeriodicalIF":1.9,"publicationDate":"2025-02-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143428656","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-17DOI: 10.1016/j.anucene.2025.111269
Haiqi Qin , Daogang Lu , Dawen Zhong , Qiong Cao
As an important throttling component for Sodium-cooled Fast Reactor (SFR), the orifice plate throttles are widely used for the pressure adjustment of coolant pipelines. The complicated flow phenomena are the common issues encountered in similar engineering applications, which need to be further investigated. In this work, a new orifice plate throttle applied to outlet pressure adjustment of main vessel cooling system is specifically designed and manufactured based on the design requirements of SFR. The full-scale verification experiment and numerical simulation are employed to investigate its hydraulic characteristics and geometrical parameter effect. The experiment results indicate that the orifice plate throttle with the throttling diameter ratio of 0.516 meets the design requirements, where this improved design has also been verified. Moreover, several empirical correlations of resistance coefficient and outflow coefficient related to the geometrical parameters are obtained by regression analysis of experiment data, which can be used to preliminarily predict the hydraulic characteristics of orifice plate throttle. As a comprehensive supplement, the numerical results with the SST turbulence model are in good agreement with the experiment results, as the maximum error less than 8%. Furthermore, the effect of geometrical parameters on the hydraulic characteristics has been numerically analyzed, such as the throttling diameter ratio, throttling orifice shape and chamfer angle. It is evident that the throttling effect of the orifice plate throttle with the upstream chamfer is superior to other. Meanwhile, reasonable chamfer angle configuration can improve the throttling effect of orifice plate throttle. After evaluation, increasing the chamfer angle by 5° can result in an increase of approximately 3% in pressure drop. As for the upstream chamfer orifice plate throttle, the chamfer angle of 45° has excellent throttling effect. This investigation reports a promising design modification, as the potential structure optimization for the orifice plate throttles of SFR.
{"title":"Experimental and numerical investigation on the hydraulic characteristics of orifice plate throttle for sodium-cooled fast reactor","authors":"Haiqi Qin , Daogang Lu , Dawen Zhong , Qiong Cao","doi":"10.1016/j.anucene.2025.111269","DOIUrl":"10.1016/j.anucene.2025.111269","url":null,"abstract":"<div><div>As an important throttling component for Sodium-cooled Fast Reactor (SFR), the orifice plate throttles are widely used for the pressure adjustment of coolant pipelines. The complicated flow phenomena are the common issues encountered in similar engineering applications, which need to be further investigated. In this work, a new orifice plate throttle applied to outlet pressure adjustment of main vessel cooling system is specifically designed and manufactured based on the design requirements of SFR. The full-scale verification experiment and numerical simulation are employed to investigate its hydraulic characteristics and geometrical parameter effect. The experiment results indicate that the orifice plate throttle with the throttling diameter ratio of 0.516 meets the design requirements, where this improved design has also been verified. Moreover, several empirical correlations of resistance coefficient and outflow coefficient related to the geometrical parameters are obtained by regression analysis of experiment data, which can be used to preliminarily predict the hydraulic characteristics of orifice plate throttle. As a comprehensive supplement, the numerical results with the SST turbulence model are in good agreement with the experiment results, as the maximum error less than 8%. Furthermore, the effect of geometrical parameters on the hydraulic characteristics has been numerically analyzed, such as the throttling diameter ratio, throttling orifice shape and chamfer angle. It is evident that the throttling effect of the orifice plate throttle with the upstream chamfer is superior to other. Meanwhile, reasonable chamfer angle configuration can improve the throttling effect of orifice plate throttle. After evaluation, increasing the chamfer angle by 5° can result in an increase of approximately 3% in pressure drop. As for the upstream chamfer orifice plate throttle, the chamfer angle of 45° has excellent throttling effect. This investigation reports a promising design modification, as the potential structure optimization for the orifice plate throttles of SFR.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111269"},"PeriodicalIF":1.9,"publicationDate":"2025-02-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143420853","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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 : 2025-02-16DOI: 10.1016/j.anucene.2025.111252
Sipeng Du, Qingmin Zhang, Yaodong Sang, Shiyu Liu
Due to the advantages of high-temperature tolerance, strong radiation resistance, tiny size, and self-powered feature, Self-Powered Neutron Detectors (SPNDs) have been widely used in the 2nd and 3rd generation reactors. The typical 4th generation reactors have higher temperature, higher neutron flux, stronger irradiation, harder neutron energy spectrum and more ununiform neutron/gamma distribution than the 3rd generation reactors. These features may affect SPND’s response performance when SPND is utilized in the 4th generation reactors. Due to its advantages in the sustainable development of nuclear fission energy, Sodium-cooled fast reactor (SFR) is chosen for this study. It is found in the study that resonant absorption plays an important role due to the harder energy spectrum in the fast reactors. In the meanwhile, current amplitude, current components, and emitter burn-up have been compared for common emitter materials, suggesting Hf is more suitable for long-term use in SFR. Finally, Hf-SPND’s response dependence on deployment position has been studied and it’s found that the current amplitude and component vary considerably with positions, indicating that position correction is required.
{"title":"Evaluation for self-powered neutron detector used in sodium-cooled fast reactor","authors":"Sipeng Du, Qingmin Zhang, Yaodong Sang, Shiyu Liu","doi":"10.1016/j.anucene.2025.111252","DOIUrl":"10.1016/j.anucene.2025.111252","url":null,"abstract":"<div><div>Due to the advantages of high-temperature tolerance, strong radiation resistance, tiny size, and self-powered feature, Self-Powered Neutron Detectors (SPNDs) have been widely used in the 2<sup>nd</sup> and 3<sup>rd</sup> generation reactors. The typical 4<sup>th</sup> generation reactors have higher temperature, higher neutron flux, stronger irradiation, harder neutron energy spectrum and more ununiform neutron/gamma distribution than the 3<sup>rd</sup> generation reactors. These features may affect SPND’s response performance when SPND is utilized in the 4<sup>th</sup> generation reactors. Due to its advantages in the sustainable development of nuclear fission energy, Sodium-cooled fast reactor (SFR) is chosen for this study. It is found in the study that resonant absorption plays an important role due to the harder energy spectrum in the fast reactors. In the meanwhile, current amplitude, current components, and emitter burn-up have been compared for common emitter materials, suggesting Hf is more suitable for long-term use in SFR. Finally, Hf-SPND’s response dependence on deployment position has been studied and it’s found that the current amplitude and component vary considerably with positions, indicating that position correction is required.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111252"},"PeriodicalIF":1.9,"publicationDate":"2025-02-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143421052","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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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