Nuclear Engineering and Technology最新文献

英文中文

Thermodynamic study on the separation of strontium and barium from LWR spent fuel 从低温反应堆乏燃料中分离锶和钡的热力学研究

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.08.050

Jin-Mok Hur, Jungho Hur, Yung-Zun Cho, Chang Hwa Lee

The separation of high heat load fission products, such as alkaline earth metals, from nuclear spent fuel can significantly reduce the burden of spent fuel disposal. This study investigates the feasibility of separating strontium and barium from light water reactor spent fuel through non-aqueous processes. Process flows were developed for treating spent nuclear fuel by heating it at high temperatures to remove volatile nuclides, followed by chlorination with a chlorinating agent. The chlorinated products were then treated with a precipitating agent in LiCl-KCl molten salt for further separation. The remaining liquid was distilled to recover strontium and barium. Thermodynamic equilibrium calculations were conducted for the process flows. Under the conditions of the process flows, the chlorinating agents MgCl₂ and NH₄Cl both converted SrO and BaO entirely into SrCl₂ and BaCl₂, respectively. The precipitating agent Li₂CO₃ exhibited superior separation effectiveness compared to Li₃PO₄. Thermodynamic calculations indicate that strontium and barium recovered by MgCl₂ chlorination, Li₂CO₃ precipitation, and distillation will contain 0.18 %, 1.06 %, and 0.32 % impurities in terms of mass, radioactivity, and decay heat, respectively.

从核乏燃料中分离碱土金属等高热负荷裂变产物，可以大大减轻乏燃料处置的负担。本研究调查了通过非水工艺从轻水反应堆乏燃料中分离锶和钡的可行性。研究开发了处理乏核燃料的工艺流程，通过高温加热去除挥发性核素，然后用氯化剂进行氯化处理。氯化产品随后在 LiCl-KCl 熔盐中用沉淀剂进行处理，以进一步分离。剩余液体经蒸馏回收锶和钡。对工艺流程进行了热力学平衡计算。在工艺流程条件下，氯化镁和 NHCl 都能将 SrO 和 BaO 分别完全转化为 SrCl 和 BaCl。沉淀剂 LiCO 的分离效果优于 LiPO。热力学计算表明，通过氯化镁氯化、LiCO 沉淀和蒸馏回收的锶和钡在质量、放射性和衰变热方面分别含有 0.18 %、1.06 % 和 0.32 % 的杂质。

{"title":"Thermodynamic study on the separation of strontium and barium from LWR spent fuel","authors":"Jin-Mok Hur, Jungho Hur, Yung-Zun Cho, Chang Hwa Lee","doi":"10.1016/j.net.2024.08.050","DOIUrl":"10.1016/j.net.2024.08.050","url":null,"abstract":"<div><div>The separation of high heat load fission products, such as alkaline earth metals, from nuclear spent fuel can significantly reduce the burden of spent fuel disposal. This study investigates the feasibility of separating strontium and barium from light water reactor spent fuel through non-aqueous processes. Process flows were developed for treating spent nuclear fuel by heating it at high temperatures to remove volatile nuclides, followed by chlorination with a chlorinating agent. The chlorinated products were then treated with a precipitating agent in LiCl-KCl molten salt for further separation. The remaining liquid was distilled to recover strontium and barium. Thermodynamic equilibrium calculations were conducted for the process flows. Under the conditions of the process flows, the chlorinating agents MgCl<sub>2</sub> and NH<sub>4</sub>Cl both converted SrO and BaO entirely into SrCl<sub>2</sub> and BaCl<sub>2</sub>, respectively. The precipitating agent Li<sub>2</sub>CO<sub>3</sub> exhibited superior separation effectiveness compared to Li<sub>3</sub>PO<sub>4</sub>. Thermodynamic calculations indicate that strontium and barium recovered by MgCl<sub>2</sub> chlorination, Li<sub>2</sub>CO<sub>3</sub> precipitation, and distillation will contain 0.18 %, 1.06 %, and 0.32 % impurities in terms of mass, radioactivity, and decay heat, respectively.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 2","pages":"Article 103181"},"PeriodicalIF":2.6,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207229","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}

引用次数: 0

Analysis and optimization of LN2 two-phase flow in CRAFT NNBI cryopump CRAFT NNBI 低温泵中 LN2 两相流的分析与优化

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.08.047

Qianxu Wang , Yuanlai Xie , Huihui Hong , Yang Zhu , Fang Wang , Kun Tian , Bin Li

As an important component of Negative ion based Neutral Beam Injector (NNBI), the cryopump mainly provides a suitable vacuum environment for beam generation and transmission. In the paper, Liquid nitrogen (LN2) pipe structure models of two cryopumps were established for simulation and experimental studies. Thermal analysis of the thermal radiation shielding baffle (LN2 cooling) was carried out by Ansys steady-state thermal analysis software, while Fluent was used to simulate the two-phase flow in the LN2 pipeline, then the pressure drops, temperature, velocity, gas-phase volume fraction, and other parameters of the two pipeline structures were analyzed and compared. The simulation results show that the optimized structure B can reduce the pressure drop by 3 bar, the outlet temperature of structure A is 78.8 K, the outlet temperature of structure B is 79.2 K, the temperature difference is 0.4 K. The outlet velocity increases from 2.067 m/s to 4.947 m/s for Structure A, and from 1.096 m/s to 6.614 m/s for Structure B. The experiment results show that the outlet temperature of structure A is 79.4 K, meanwhile the outlet temperature of structure B is 80.1 K, the optimized structure B can reduce the pressure drop about 3 bar.

作为负离子中性束注入器（NNBI）的重要组成部分，低温泵主要为束流的产生和传输提供合适的真空环境。本文建立了两个低温泵的液氮（LN2）管道结构模型，并进行了模拟和实验研究。利用 Ansys 稳态热分析软件对热辐射屏蔽挡板（LN2 冷却）进行了热分析，同时利用 Fluent 对 LN2 管道中的两相流动进行了模拟，然后对两种管道结构的压降、温度、速度、气相体积分数等参数进行了分析和比较。模拟结果表明，优化后的结构 B 可使压降降低 3 bar，结构 A 的出口温度为 78.8 K，结构 B 的出口温度为 79.2 K，温差为 0.4 K，出口速度从结构 A 的 2.067 m/s 增加到 4.实验结果表明，结构 A 的出口温度为 79.4 K，而结构 B 的出口温度为 80.1 K，优化后的结构 B 可减少压降约 3 bar。

{"title":"Analysis and optimization of LN2 two-phase flow in CRAFT NNBI cryopump","authors":"Qianxu Wang , Yuanlai Xie , Huihui Hong , Yang Zhu , Fang Wang , Kun Tian , Bin Li","doi":"10.1016/j.net.2024.08.047","DOIUrl":"10.1016/j.net.2024.08.047","url":null,"abstract":"<div><div>As an important component of Negative ion based Neutral Beam Injector (NNBI), the cryopump mainly provides a suitable vacuum environment for beam generation and transmission. In the paper, Liquid nitrogen (LN2) pipe structure models of two cryopumps were established for simulation and experimental studies. Thermal analysis of the thermal radiation shielding baffle (LN2 cooling) was carried out by Ansys steady-state thermal analysis software, while Fluent was used to simulate the two-phase flow in the LN2 pipeline, then the pressure drops, temperature, velocity, gas-phase volume fraction, and other parameters of the two pipeline structures were analyzed and compared. The simulation results show that the optimized structure B can reduce the pressure drop by 3 bar, the outlet temperature of structure A is 78.8 K, the outlet temperature of structure B is 79.2 K, the temperature difference is 0.4 K. The outlet velocity increases from 2.067 m/s to 4.947 m/s for Structure A, and from 1.096 m/s to 6.614 m/s for Structure B. The experiment results show that the outlet temperature of structure A is 79.4 K, meanwhile the outlet temperature of structure B is 80.1 K, the optimized structure B can reduce the pressure drop about 3 bar.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 2","pages":"Article 103178"},"PeriodicalIF":2.6,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207241","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}

引用次数: 0

Development of CFD porous model parameters of Korean PWR spent nuclear fuel assemblies for thermal analysis

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.09.014

Do Yun Kim , Han-Seop Song , Jae-Ho Jeong , Seung-Hwan Yu

This study focused on developing and validating porous model parameters to accurately estimate the thermal behavior of spent nuclear fuel assemblies for computational fluid dynamics simulations. These assemblies are stored in Korean pressurized water reactors (PWRs) under dry storage conditions. Korean PWR fuel assemblies were categorized into four distinct types based on their configuration, facilitating the calculation of effective thermal conductivities and flow resistance coefficients. These parameters are crucial for simulating heat transfer and determining the peak cladding temperature (PCT), ensuring the thermal safety of stored nuclear fuel. The validation of the proposed porous model against explicit two-dimensional and three-dimensional simulations demonstrated its efficacy and reliability in enhancing the thermal analysis of spent fuel assemblies, highlighting the significance of accurately calculating the porous model parameters in the safety assessment of nuclear fuel storage. Furthermore, the study assessed the sensitivity of variables influencing temperature during spent nuclear fuel storage, focusing on flow resistance coefficients, control rod effect, and basket size. The findings revealed that variations in flow resistance and the presence of control rods minimally impacted PCT, whereas basket size significantly influenced temperature, underscoring its importance in the thermal analysis of nuclear fuel storage systems.

{"title":"Development of CFD porous model parameters of Korean PWR spent nuclear fuel assemblies for thermal analysis","authors":"Do Yun Kim , Han-Seop Song , Jae-Ho Jeong , Seung-Hwan Yu","doi":"10.1016/j.net.2024.09.014","DOIUrl":"10.1016/j.net.2024.09.014","url":null,"abstract":"<div><div>This study focused on developing and validating porous model parameters to accurately estimate the thermal behavior of spent nuclear fuel assemblies for computational fluid dynamics simulations. These assemblies are stored in Korean pressurized water reactors (PWRs) under dry storage conditions. Korean PWR fuel assemblies were categorized into four distinct types based on their configuration, facilitating the calculation of effective thermal conductivities and flow resistance coefficients. These parameters are crucial for simulating heat transfer and determining the peak cladding temperature (PCT), ensuring the thermal safety of stored nuclear fuel. The validation of the proposed porous model against explicit two-dimensional and three-dimensional simulations demonstrated its efficacy and reliability in enhancing the thermal analysis of spent fuel assemblies, highlighting the significance of accurately calculating the porous model parameters in the safety assessment of nuclear fuel storage. Furthermore, the study assessed the sensitivity of variables influencing temperature during spent nuclear fuel storage, focusing on flow resistance coefficients, control rod effect, and basket size. The findings revealed that variations in flow resistance and the presence of control rods minimally impacted PCT, whereas basket size significantly influenced temperature, underscoring its importance in the thermal analysis of nuclear fuel storage systems.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 2","pages":"Article 103211"},"PeriodicalIF":2.6,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143099601","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}

引用次数: 0

Analysis of Naturally Occurring Radioactive Material (NORM) in shifting earth layers in Piyungan, Bantul

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.08.055

Iqbal Nurrahim , Andang Widi Harto , Yohannes Sardjono , Isman Mulyadi Triatmoko , Zuhdi Ismail , Gede Sutresna Wijaya

In recent decades, understanding the correlation between radon gas emissions and geodynamic processes has been pivotal in enhancing geophysical monitoring. This study advances current methodologies by refining radon data analysis and employing mathematical models to further elucidate this relationship. We developed a radon well geometry for Piyungan using Finite Element Method principles, and conducted simulations via GNU Octave to assess normal and anomalous conditions. The results indicate a basal radon concentration of 2.2420 Bq/L at a depth of 14.2308 m. Notably, radon levels inversely correlate with the water volume and diffusivity, yet proportionally increase with soil moisture. These findings offer a nuanced understanding of radon behavior in response to geodynamic activity, paving the way for improved environmental monitoring.

引用次数: 0

Differentiated influences of perceived benefit types of nuclear power on its acceptance: Evidence from South Korea

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.09.017

Jin Won Lee , Hailin Zhang , Seungkook Roh

Perceived benefits and risk of nuclear power have received considerable attention as influencing factors in public acceptance of nuclear power. The validation of the differential effects of these perceptions across contexts can provide implications for public communication strategies. This study is based on the similarity and dissimilarity between nuclear benefits and risk in their consequences. It posits that the pattern of the effect of environmental benefit perception on nuclear power acceptance is more similar to that of risk perception than to that of energy supply benefit perception, which is another type of benefit perception. Upon analyzing a national survey data from South Korea, we discovered the following: (1) regarding nuclear power as a non-dispreferred option, all the benefit perceptions and the risk perception exhibited a significant effect; (2) however, regarding nuclear power as the favorite option, the perception of environmental benefit and that of risk demonstrated a significant effect, but that of energy supply benefit did not. These findings have implications for both public communication strategies and theoretical considerations in future studies.

{"title":"Differentiated influences of perceived benefit types of nuclear power on its acceptance: Evidence from South Korea","authors":"Jin Won Lee , Hailin Zhang , Seungkook Roh","doi":"10.1016/j.net.2024.09.017","DOIUrl":"10.1016/j.net.2024.09.017","url":null,"abstract":"<div><div>Perceived benefits and risk of nuclear power have received considerable attention as influencing factors in public acceptance of nuclear power. The validation of the differential effects of these perceptions across contexts can provide implications for public communication strategies. This study is based on the similarity and dissimilarity between nuclear benefits and risk in their consequences. It posits that the pattern of the effect of environmental benefit perception on nuclear power acceptance is more similar to that of risk perception than to that of energy supply benefit perception, which is another type of benefit perception. Upon analyzing a national survey data from South Korea, we discovered the following: (1) regarding nuclear power as a non-dispreferred option, all the benefit perceptions and the risk perception exhibited a significant effect; (2) however, regarding nuclear power as the favorite option, the perception of environmental benefit and that of risk demonstrated a significant effect, but that of energy supply benefit did not. These findings have implications for both public communication strategies and theoretical considerations in future studies.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 2","pages":"Article 103214"},"PeriodicalIF":2.6,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143099940","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}

引用次数: 0

Development of simplified treatment methods for removal of Ra-226 and Ra-228 from sludge waste produced in natural gas processing plants

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.08.044

Ayman M. Sharaf , M.A. El-Nahal , Islam M. Nabil , I.H. Saleh , Mohamed Elsafi

The present study aims to establish a new treatment method for removing or reducing the concentration of ²²⁶Ra and ²²⁸Ra in sludge waste produced due to natural gas production at natural gas processing Plants. The main radionuclides present in sludge waste of natural gas processing are Radium and its daughters due to their low solubility in the water they can form compounds with sulfate ions, carbonate ions, silicate ions, then precipitation scale or sludge. The suggested method is designed to be simple, applicable, economically beneficial, and environmentally safe, by using internal available resources like emitted hydrogen sulfide and sodium hydroxide, also reduction of chemicals that used in radioactivity treatment to be sulfuric acid or sodium hydroxide instead of multiple different chemicals that used in other treatment procedures. The effects of the used chemical reagents sulfuric acid and sodium hydroxide in various concentrations at different reaction temperatures on the sludge waste to extract the radium isotopes have been studied, evaluated, and compared. The study proved that removing radium isotopes is increased by elevating the reagents concentration and sludge temperature. The removal effects of the used reagents are approximately similar.

{"title":"Development of simplified treatment methods for removal of Ra-226 and Ra-228 from sludge waste produced in natural gas processing plants","authors":"Ayman M. Sharaf , M.A. El-Nahal , Islam M. Nabil , I.H. Saleh , Mohamed Elsafi","doi":"10.1016/j.net.2024.08.044","DOIUrl":"10.1016/j.net.2024.08.044","url":null,"abstract":"<div><div>The present study aims to establish a new treatment method for removing or reducing the concentration of <sup>226</sup>Ra and <sup>228</sup>Ra in sludge waste produced due to natural gas production at natural gas processing Plants. The main radionuclides present in sludge waste of natural gas processing are Radium and its daughters due to their low solubility in the water they can form compounds with sulfate ions, carbonate ions, silicate ions, then precipitation scale or sludge. The suggested method is designed to be simple, applicable, economically beneficial, and environmentally safe, by using internal available resources like emitted hydrogen sulfide and sodium hydroxide, also reduction of chemicals that used in radioactivity treatment to be sulfuric acid or sodium hydroxide instead of multiple different chemicals that used in other treatment procedures. The effects of the used chemical reagents sulfuric acid and sodium hydroxide in various concentrations at different reaction temperatures on the sludge waste to extract the radium isotopes have been studied, evaluated, and compared. The study proved that removing radium isotopes is increased by elevating the reagents concentration and sludge temperature. The removal effects of the used reagents are approximately similar.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 2","pages":"Article 103175"},"PeriodicalIF":2.6,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143099941","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}

引用次数: 0

Fluence-to-dose conversion coefficients in a voxel rat model for external neutron irradiation 体外中子辐照大鼠模型中的通量-剂量转换系数

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.08.056

Xiaomin Zhang , Xu Xu , Yong Yuan, Jing Ning, Dawei Li, Yunlong Ji

Fluence-to-dose conversion coefﬁcients based on a developed rat model have been calculated for neutrons with energies <20 MeV using Monte Carlo N-particle code (MCNP), for the purpose of neutron radiation effect evaluation. The calculation was conducted respectively for 27 monodirectional monoenergetic neutron beams in the energy range 10⁻⁹ MeV to 20 MeV, under four different irradiation conditions: left lateral, right lateral, dorsal–ventral and ventral–dorsal. The neutron ﬂuence-to-dose conversion coefﬁcients for selected organs were presented in the study and can be used to determine the organ dose of the rats experimentally exposed to external neutron irradiation. The results in this work were compared with the published data based on a mouse model to investigate the effect of size and weight difference on neutron organ dose. The comparison results showed the fluence-to-dose conversion coefficients of the rat model have the similar energy dependency and sensitivity to irradiation conditions compared with that of the mouse model, and the weight and size difference in individuals could lead to different levels of neutron organ dose difference depending on neutron energy, irradiation conditions as well as the location of organs.

为了评估中子辐射效应，我们使用蒙特卡洛 N 粒子代码（MCNP）计算了能量小于 20 MeV 的中子的通量-剂量转换系数，该系数基于开发的大鼠模型。在左外侧、右外侧、背-腹侧和腹-背侧四种不同的辐照条件下，分别对能量范围为 10 MeV 至 20 MeV 的 27 个单向单能中子束进行了计算。研究中提出了选定器官的中子辐照剂量转换系数，可用于确定大鼠在外部中子辐照实验中的器官剂量。这项工作的结果与已发表的基于小鼠模型的数据进行了比较，以研究体型和体重差异对中子器官剂量的影响。比较结果表明，与小鼠模型相比，大鼠模型的通量-剂量转换系数具有相似的能量依赖性和对辐照条件的敏感性，而个体的体重和体型差异会因中子能量、辐照条件和器官位置的不同而导致不同程度的中子器官剂量差异。

{"title":"Fluence-to-dose conversion coefficients in a voxel rat model for external neutron irradiation","authors":"Xiaomin Zhang , Xu Xu , Yong Yuan, Jing Ning, Dawei Li, Yunlong Ji","doi":"10.1016/j.net.2024.08.056","DOIUrl":"10.1016/j.net.2024.08.056","url":null,"abstract":"<div><div>Fluence-to-dose conversion coefﬁcients based on a developed rat model have been calculated for neutrons with energies <20 MeV using Monte Carlo N-particle code (MCNP), for the purpose of neutron radiation effect evaluation. The calculation was conducted respectively for 27 monodirectional monoenergetic neutron beams in the energy range 10<sup>−9</sup> MeV to 20 MeV, under four different irradiation conditions: left lateral, right lateral, dorsal–ventral and ventral–dorsal. The neutron ﬂuence-to-dose conversion coefﬁcients for selected organs were presented in the study and can be used to determine the organ dose of the rats experimentally exposed to external neutron irradiation. The results in this work were compared with the published data based on a mouse model to investigate the effect of size and weight difference on neutron organ dose. The comparison results showed the fluence-to-dose conversion coefficients of the rat model have the similar energy dependency and sensitivity to irradiation conditions compared with that of the mouse model, and the weight and size difference in individuals could lead to different levels of neutron organ dose difference depending on neutron energy, irradiation conditions as well as the location of organs.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 2","pages":"Article 103187"},"PeriodicalIF":2.6,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142226264","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}

引用次数: 0

Accurate pulse time distribution determination using MLEM algorithm in integral experiments 在积分实验中使用 MLEM 算法精确测定脉冲时间分布

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.08.060

S.Y. Zhang , Y.B. Nie , Y.Y. Ding , Q. Zhao , K.Z. Xu , X.Y. Pan , H.T. Chen , Q. Sun , Z. Wei

Integral experiments play a crucial role in advancing nuclear science and technology by providing critical data that validate theoretical models and enhance reactor designs. This study presents a novel approach to accurately determine pulse time distribution in integral experiments conducted with pulsed accelerators. By strategically placed monitors and shields at angles of 0° and 90° relative to the beam direction, neutron flight times from the target are measured, and a response matrix for neutron emission at different times is constructed through simulation. The Maximum Likelihood Expectation Maximization (MLEM) algorithm is employed for pulse time reconstruction, with the gamma ray flight time spectrum from monitors used as the initial spectrum to streamline the computational process. Experimental validation using a standard polyethylene sample and n-p scattering cross-sections confirms the accuracy of the method. Results are compared across multiple nuclear databases such as CENDL-3.2, ENDF/B-VIII.0, JENDL-5.0, and JEFF-3.3 libraries. The developed method significantly enhances the precision of pulse time distribution determination, thereby improving the quality and reliability of experimental data obtained from integral experiments conducted with pulsed accelerators.

积分实验通过提供验证理论模型和改进反应堆设计的关键数据，在推动核科学与技术发展方面发挥着至关重要的作用。本研究提出了一种新方法，用于在使用脉冲加速器进行的积分实验中精确测定脉冲时间分布。通过在与光束方向成 0° 和 90° 角的位置战略性地放置监测器和防护罩，测量了来自目标的中子飞行时间，并通过模拟构建了不同时间中子发射的响应矩阵。采用最大似然期望最大化（MLEM）算法进行脉冲时间重建，并将监测器中的伽马射线飞行时间频谱作为初始频谱，以简化计算过程。使用标准聚乙烯样品和 n-p 散射截面进行的实验验证证实了该方法的准确性。比较了多个核数据库（如 CENDL-3.2、ENDF/B-VIII.0、JENDL-5.0 和 JEFF-3.3 库）的结果。所开发的方法大大提高了脉冲时间分布测定的精确度，从而提高了利用脉冲加速器进行积分实验所获得的实验数据的质量和可靠性。

{"title":"Accurate pulse time distribution determination using MLEM algorithm in integral experiments","authors":"S.Y. Zhang , Y.B. Nie , Y.Y. Ding , Q. Zhao , K.Z. Xu , X.Y. Pan , H.T. Chen , Q. Sun , Z. Wei","doi":"10.1016/j.net.2024.08.060","DOIUrl":"10.1016/j.net.2024.08.060","url":null,"abstract":"<div><div>Integral experiments play a crucial role in advancing nuclear science and technology by providing critical data that validate theoretical models and enhance reactor designs. This study presents a novel approach to accurately determine pulse time distribution in integral experiments conducted with pulsed accelerators. By strategically placed monitors and shields at angles of 0° and 90° relative to the beam direction, neutron flight times from the target are measured, and a response matrix for neutron emission at different times is constructed through simulation. The Maximum Likelihood Expectation Maximization (MLEM) algorithm is employed for pulse time reconstruction, with the gamma ray flight time spectrum from monitors used as the initial spectrum to streamline the computational process. Experimental validation using a standard polyethylene sample and n-p scattering cross-sections confirms the accuracy of the method. Results are compared across multiple nuclear databases such as CENDL-3.2, ENDF/B-VIII.0, JENDL-5.0, and JEFF-3.3 libraries. The developed method significantly enhances the precision of pulse time distribution determination, thereby improving the quality and reliability of experimental data obtained from integral experiments conducted with pulsed accelerators.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 2","pages":"Article 103191"},"PeriodicalIF":2.6,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207216","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}

引用次数: 0

Estimation of neutral beam injector power transferred to KSTAR plasmas 估算传输到 KSTAR 等离子体的中性束注入器功率

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.08.057

Jong-Gu Kwak, J.H. Jeong, S.C. Hong

KSTAR shows the long pulse capability based on superconducting magnet and NBI plays a crucial role in sustaining the plasma via plasma heating and current driving. So, the accurate NBI power measurement transferred to KSTAR plasma is important for the analysis of plasma transport as well as plasma performance parameters. In a long pulse operation, the water coolant temperature is at the steady state condition in longer pulse more than 20s and the coupled neutral beam power to KSTAR plasma during the tokamak experiments is rechecked by using water flow calorimetric method after the experiment. The effect of beam duct scraper loss which was not considered at the neutral beam power calibration process is less than 5 % in terms of neutral beam power. However, in long pulse operation of NBI in KSTAR experiments, high strength of stray PF affects the beam path, the neutral beam power registered on mds + using beam current method is over estimated and it is calculated up to a few percent in terms of neutral beam power using calorimetric method. Therefore, it is necessary to consider the beam power variation by PF effect to interpret the plasma performance degradation in long pulse operation. Lastly. when the ion source tun on or turn off in condition other ion sources are operated, the beam transmission power is also affected because of sharing the beam box for ion sources so that the careful power estimation is necessary for such kind of beam power modulation experiments in KSTAR.

KSTAR 显示了基于超导磁体的长脉冲能力，而 NBI 在通过等离子体加热和电流驱动维持等离子体方面发挥着至关重要的作用。因此，对 KSTAR 等离子体进行精确的 NBI 功率测量对于分析等离子体传输和等离子体性能参数非常重要。在长脉冲运行中，水冷却剂温度在超过 20 秒的长脉冲中处于稳定状态，在托卡马克实验过程中，我们在实验结束后使用水流量热法重新检查了传递到 KSTAR 等离子体的耦合中性束功率。中性束功率校准过程中未考虑的束管刮刀损耗对中性束功率的影响小于 5%。然而，在 KSTAR 实验中的 NBI 长脉冲运行中，高强度的杂散 PF 会影响光束路径，使用束流法在 mds + 上记录的中性束功率被高估，而使用量热法计算的中性束功率仅为百分之几。因此，有必要考虑 PF 效应带来的束流功率变化，以解释长脉冲运行时等离子体性能的下降。最后，当离子源在其他离子源运行的条件下开启或关闭时，由于离子源共用光束箱，光束传输功率也会受到影响，因此在 KSTAR 进行此类光束功率调制实验时，有必要进行仔细的功率估算。

{"title":"Estimation of neutral beam injector power transferred to KSTAR plasmas","authors":"Jong-Gu Kwak, J.H. Jeong, S.C. Hong","doi":"10.1016/j.net.2024.08.057","DOIUrl":"10.1016/j.net.2024.08.057","url":null,"abstract":"<div><div>KSTAR shows the long pulse capability based on superconducting magnet and NBI plays a crucial role in sustaining the plasma via plasma heating and current driving. So, the accurate NBI power measurement transferred to KSTAR plasma is important for the analysis of plasma transport as well as plasma performance parameters. In a long pulse operation, the water coolant temperature is at the steady state condition in longer pulse more than 20s and the coupled neutral beam power to KSTAR plasma during the tokamak experiments is rechecked by using water flow calorimetric method after the experiment. The effect of beam duct scraper loss which was not considered at the neutral beam power calibration process is less than 5 % in terms of neutral beam power. However, in long pulse operation of NBI in KSTAR experiments, high strength of stray PF affects the beam path, the neutral beam power registered on mds + using beam current method is over estimated and it is calculated up to a few percent in terms of neutral beam power using calorimetric method. Therefore, it is necessary to consider the beam power variation by PF effect to interpret the plasma performance degradation in long pulse operation. Lastly. when the ion source tun on or turn off in condition other ion sources are operated, the beam transmission power is also affected because of sharing the beam box for ion sources so that the careful power estimation is necessary for such kind of beam power modulation experiments in KSTAR.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 2","pages":"Article 103188"},"PeriodicalIF":2.6,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207221","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}

引用次数: 0

Enhanced formula-based method to estimate burnup, initial enrichment, and cooling time using limited gamma-ray source data of reference PWR spent nuclear fuel assemblies

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.09.025

Sang Hyeok Lee, Jae Hak Cheong

This study proposes a framework to develop a model for estimating burnup (BU), initial enrichment (IE), and cooling time (CT) of spent nuclear fuel (SNF) overcoming the limitations of existing models in determining IE. Database (DB) of gamma-ray sources for PWR SNF was generated using ORIGAMI. Target gamma radionuclides were selected based on the ease and credibility of measurement. The concept of Product of Powers of Radioactivity (PPR) was suggested, and PPRs for modeling were selected through evaluation of all possible PPRs. Functions established for PPRs were applied to estimate BU, IE, and CT (BIC) of PWR SNF assemblies. Optimized PPR combinations for estimating BU, IE, and CT respectively, as well as three variables simultaneously with the highest accuracy, were suggested. The estimation model proposed in this study enhanced IE estimations, maintaining the reliability of BU and CT estimations. The above framework enables construction of intuitive BIC estimation models using only a small amount of data. This can help hasten the development of estimation models for SNF from various reactor types. Consequently, it is expected to contribute to the safety and safeguards of SNF management by improving the verification that SNF meets acceptance criteria and conforms to the declared information.

{"title":"Enhanced formula-based method to estimate burnup, initial enrichment, and cooling time using limited gamma-ray source data of reference PWR spent nuclear fuel assemblies","authors":"Sang Hyeok Lee, Jae Hak Cheong","doi":"10.1016/j.net.2024.09.025","DOIUrl":"10.1016/j.net.2024.09.025","url":null,"abstract":"<div><div>This study proposes a framework to develop a model for estimating burnup (BU), initial enrichment (IE), and cooling time (CT) of spent nuclear fuel (SNF) overcoming the limitations of existing models in determining IE. Database (DB) of gamma-ray sources for PWR SNF was generated using ORIGAMI. Target gamma radionuclides were selected based on the ease and credibility of measurement. The concept of Product of Powers of Radioactivity (PPR) was suggested, and PPRs for modeling were selected through evaluation of all possible PPRs. Functions established for PPRs were applied to estimate BU, IE, and CT (BIC) of PWR SNF assemblies. Optimized PPR combinations for estimating BU, IE, and CT respectively, as well as three variables simultaneously with the highest accuracy, were suggested. The estimation model proposed in this study enhanced IE estimations, maintaining the reliability of BU and CT estimations. The above framework enables construction of intuitive BIC estimation models using only a small amount of data. This can help hasten the development of estimation models for SNF from various reactor types. Consequently, it is expected to contribute to the safety and safeguards of SNF management by improving the verification that SNF meets acceptance criteria and conforms to the declared information.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 2","pages":"Article 103222"},"PeriodicalIF":2.6,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143099582","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}

引用次数: 0

首页上一页

下一页尾页

类型

全部化学•材料生命科学医学物理工程技术环境•农林材料科学地球科学法学管理学化学环境科学与生态学计算机科学教育学经济学农林科学人文科学生物学数学物理与天体物理心理学综合性期刊其他工业工程理学历史学农学文学信息工程

数据库

全部 ACS Publications Elsevier ieeexplore Springer The Royal Society of Chemistry Wiley

期刊

Nuclear Engineering and Technology

全部 Acc. Chem. Res. ACS Applied Bio Materials ACS Appl. Electron. Mater. ACS Appl. Energy Mater. ACS Appl. Mater. Interfaces ACS Appl. Nano Mater. ACS Appl. Polym. Mater. ACS BIOMATER-SCI ENG ACS Catal. ACS Cent. Sci. ACS Chem. Biol. ACS Chemical Health & Safety ACS Chem. Neurosci. ACS Comb. Sci. ACS Earth Space Chem. ACS Energy Lett. ACS Infect. Dis. ACS Macro Lett. ACS Mater. Lett. ACS Med. Chem. Lett. ACS Nano ACS Omega ACS Photonics ACS Sens. ACS Sustainable Chem. Eng. ACS Synth. Biol. Anal. Chem. BIOCHEMISTRY-US Bioconjugate Chem. BIOMACROMOLECULES Chem. Res. Toxicol. Chem. Rev. Chem. Mater. CRYST GROWTH DES ENERG FUEL Environ. Sci. Technol. Environ. Sci. Technol. Lett. Eur. J. Inorg. Chem. IND ENG CHEM RES Inorg. Chem. J. Agric. Food. Chem. J. Chem. Eng. Data J. Chem. Educ. J. Chem. Inf. Model. J. Chem. Theory Comput. J. Med. Chem. J. Nat. Prod. J PROTEOME RES J. Am. Chem. Soc. LANGMUIR MACROMOLECULES Mol. Pharmaceutics Nano Lett. Org. Lett. ORG PROCESS RES DEV ORGANOMETALLICS J. Org. Chem. J. Phys. Chem. J. Phys. Chem. A J. Phys. Chem. B J. Phys. Chem. C J. Phys. Chem. Lett. Analyst Anal. Methods Biomater. Sci. Catal. Sci. Technol. Chem. Commun. Chem. Soc. Rev. CHEM EDUC RES PRACT CRYSTENGCOMM Dalton Trans. Energy Environ. Sci. ENVIRON SCI-NANO ENVIRON SCI-PROC IMP ENVIRON SCI-WAT RES Faraday Discuss. Food Funct. Green Chem. Inorg. Chem. Front. Integr. Biol. J. Anal. At. Spectrom. J. Mater. Chem. A J. Mater. Chem. B J. Mater. Chem. C Lab Chip Mater. Chem. Front. Mater. Horiz. MEDCHEMCOMM Metallomics Mol. Biosyst. Mol. Syst. Des. Eng. Nanoscale Nanoscale Horiz. Nat. Prod. Rep. New J. Chem. Org. Biomol. Chem. Org. Chem. Front. PHOTOCH PHOTOBIO SCI PCCP Polym. Chem.

﹀