Pub Date : 2024-07-24DOI: 10.1016/j.net.2024.07.050
Peter J. Hiller MSc BSc, Caroline K. Pyke MSc BSc, Chris P. Lennon BSc, Olivia C.G. Tuck MMathStat, Caitlin A. Painter BSc
The disposal of radioactive waste within the UK is managed through a comprehensive regulatory framework. This framework requires radioactive waste to be sufficiently well characterized to ensure its disposal is compliant with the regulations and the acceptance criteria for any receiving facility. This is the responsibility of both the waste consignor and the receiving facility.
{"title":"The ability to manage uncertainty for solid radioactive waste characterization in the UK nuclear industry","authors":"Peter J. Hiller MSc BSc, Caroline K. Pyke MSc BSc, Chris P. Lennon BSc, Olivia C.G. Tuck MMathStat, Caitlin A. Painter BSc","doi":"10.1016/j.net.2024.07.050","DOIUrl":"https://doi.org/10.1016/j.net.2024.07.050","url":null,"abstract":"The disposal of radioactive waste within the UK is managed through a comprehensive regulatory framework. This framework requires radioactive waste to be sufficiently well characterized to ensure its disposal is compliant with the regulations and the acceptance criteria for any receiving facility. This is the responsibility of both the waste consignor and the receiving facility.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"14 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-07-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141771269","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 : 2024-07-24DOI: 10.1016/j.net.2024.07.051
Longxian Li, Min Zhu, Yan Li, Yanru Ren, Longfei Pu, Chengxuan Peng
A first-principles approach based on density-functional theory has been used to investigate the corrosion resistance of alpha-U in the CO environment. Calculations show that O molecules spontaneously dissociate on the uranium surface, and the two O atoms formed by dissociation tend to adsorb on the hole sites and bind to the surface in the form of a U-O bond to emit a large amount of heat. The CO molecules occur on the surface of uranium as a non-dissociative chemical. The mechanism of CO inhibiting the adsorption of O molecules stems from the fact that CO molecules occupy the optimal adsorption sites. Another inhibition mechanism, the combination of C atoms and O atoms to form bonds and consume oxygen atoms, has little effect on uranium corrosion.
我们采用基于密度泛函理论的第一原理方法研究了α-U 在 CO 环境中的耐腐蚀性。计算表明，O 分子在铀表面自发解离，解离形成的两个 O 原子倾向于吸附在空穴位点上，并以 U-O 键的形式与表面结合，从而释放出大量热量。CO 分子作为一种非解离化学物质出现在铀表面。CO 抑制 O 分子吸附的机制源于 CO 分子占据了最佳吸附位点。另一种抑制机制，即 C 原子和 O 原子结合成键并消耗氧原子，对铀腐蚀的影响很小。
{"title":"The role of CO on initial oxidation behavior of α-U(001) surface: a first principles study","authors":"Longxian Li, Min Zhu, Yan Li, Yanru Ren, Longfei Pu, Chengxuan Peng","doi":"10.1016/j.net.2024.07.051","DOIUrl":"https://doi.org/10.1016/j.net.2024.07.051","url":null,"abstract":"A first-principles approach based on density-functional theory has been used to investigate the corrosion resistance of alpha-U in the CO environment. Calculations show that O molecules spontaneously dissociate on the uranium surface, and the two O atoms formed by dissociation tend to adsorb on the hole sites and bind to the surface in the form of a U-O bond to emit a large amount of heat. The CO molecules occur on the surface of uranium as a non-dissociative chemical. The mechanism of CO inhibiting the adsorption of O molecules stems from the fact that CO molecules occupy the optimal adsorption sites. Another inhibition mechanism, the combination of C atoms and O atoms to form bonds and consume oxygen atoms, has little effect on uranium corrosion.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"40 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-07-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141771268","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 : 2024-07-23DOI: 10.1016/j.net.2024.07.047
Sung-Wook Kim, Hee-Man Yang, Hyung-Ju Kim
Surface decontamination agents must effectively capture particle contaminants that have been deposited on surfaces. This ability is particularly important in scenarios involving nuclear fallout particles. In this technical note, the particle-capturing ability of a hydrogel, specifically, the polyvinyl alcohol-borax complex, was investigated as a potential decontamination agent for treating radioactively contaminated surface. For the assessment, simulated nuclear fallout particles, prepared by melting consolidation at 1200 °C followed by mechanical milling, were fixed on a stainless-steel substrate. The results demonstrated that the hydrogel effectively removed the simulated fallout particles from the surface. Even after only 1 min of contact time, the surface was left thoroughly clean, demonstrating the effectiveness of the polyvinyl alcohol-borax complex as a surface decontamination agent.
表面去污剂必须能有效捕捉沉积在表面上的颗粒污染物。在涉及核尘埃粒子的情况下，这种能力尤为重要。在本技术说明中，研究了水凝胶（特别是聚乙烯醇-硼砂复合物）作为潜在去污剂处理放射性污染表面的粒子捕获能力。为了进行评估，在不锈钢基底上固定了通过 1200 °C 熔化固结和机械研磨制备的模拟核尘埃粒子。结果表明，水凝胶能有效去除表面的模拟核沉降物颗粒。即使只接触了 1 分钟，表面也被彻底清洁，这证明了聚乙烯醇-硼砂复合物作为表面去污剂的有效性。
{"title":"Evaluation of particle-capturing ability of a hydrogel-based surface decontamination agent using simulated nuclear fallout particles","authors":"Sung-Wook Kim, Hee-Man Yang, Hyung-Ju Kim","doi":"10.1016/j.net.2024.07.047","DOIUrl":"10.1016/j.net.2024.07.047","url":null,"abstract":"<div><div>Surface decontamination agents must effectively capture particle contaminants that have been deposited on surfaces. This ability is particularly important in scenarios involving nuclear fallout particles. In this technical note, the particle-capturing ability of a hydrogel, specifically, the polyvinyl alcohol-borax complex, was investigated as a potential decontamination agent for treating radioactively contaminated surface. For the assessment, simulated nuclear fallout particles, prepared by melting consolidation at 1200 °C followed by mechanical milling, were fixed on a stainless-steel substrate. The results demonstrated that the hydrogel effectively removed the simulated fallout particles from the surface. Even after only 1 min of contact time, the surface was left thoroughly clean, demonstrating the effectiveness of the polyvinyl alcohol-borax complex as a surface decontamination agent.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"56 12","pages":"Pages 5386-5395"},"PeriodicalIF":2.6,"publicationDate":"2024-07-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141771271","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 : 2024-07-23DOI: 10.1016/j.net.2024.07.049
Hyungi Byun , Han Gil Lee , Beom Kyu Kim , Geun Dong Song , Bongsoo Lee
This study developed a defect-monitoring system with an artificial intelligence model, YOLOv7, which is tailored for processing image data from an ultrasonic visualization system within sodium fast reactor (SFR) internal structures. For the safety of SFR internal structures, although it is a crucial inspection for defect monitoring, it is difficult to identify structural defects because of the invisible environment. Therefore, we applied the YOLOv7 model in this study; however, we encountered challenges including decreased accuracy with complex defect shapes and complications from data augmentation during pre-training. To solve these problems, we additionally applied the enhanced super-resolution generative adversarial network for higher resolution and a Sobel noise-filtering algorithm to enhance the defect detection accuracy. And we evaluated our system by comparing it with a confidence score. This underscores the effectiveness of the approach in enhancing the defect detection capabilities. Therefore, this defect-monitoring system should be designed to preemptively identify internal structure deformations and enhance SFR safety and maintenance practices.
{"title":"Defect monitoring system of the internal structures of a sodium fast reactor using an artificial intelligence model","authors":"Hyungi Byun , Han Gil Lee , Beom Kyu Kim , Geun Dong Song , Bongsoo Lee","doi":"10.1016/j.net.2024.07.049","DOIUrl":"10.1016/j.net.2024.07.049","url":null,"abstract":"<div><div>This study developed a defect-monitoring system with an artificial intelligence model, YOLOv7, which is tailored for processing image data from an ultrasonic visualization system within sodium fast reactor (SFR) internal structures. For the safety of SFR internal structures, although it is a crucial inspection for defect monitoring, it is difficult to identify structural defects because of the invisible environment. Therefore, we applied the YOLOv7 model in this study; however, we encountered challenges including decreased accuracy with complex defect shapes and complications from data augmentation during pre-training. To solve these problems, we additionally applied the enhanced super-resolution generative adversarial network for higher resolution and a Sobel noise-filtering algorithm to enhance the defect detection accuracy. And we evaluated our system by comparing it with a confidence score. This underscores the effectiveness of the approach in enhancing the defect detection capabilities. Therefore, this defect-monitoring system should be designed to preemptively identify internal structure deformations and enhance SFR safety and maintenance practices.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"56 12","pages":"Pages 5405-5413"},"PeriodicalIF":2.6,"publicationDate":"2024-07-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141771270","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 : 2024-07-23DOI: 10.1016/j.net.2024.07.045
Xicheng Wang, Govatsa Acharya, Dmitry Grishchenko, Pavel Kudinov
Boiling Water Reactor (BWR) employs the Pressure Suppression Pool (PSP) as a heat sink to prevent overpressure of the reactor vessel and containment. Steam can be injected into the PSP through spargers in normal and accident conditions and through blowdown pipes in case of a loss of coolant accident (LOCA). There is a safety limit on the maximum PSP temperature at which such steam injection might cause dynamic loads on the containment structures. The performance of the pool can be affected if thermal stratification is developed when temperature of the hot layer grows rapidly while cold layer remains inactive. Simulation of pool behavior during realistic accident scenarios requires validated models that can sufficiently address the interaction between phenomena, safety systems and operational procedures. Direct modeling of steam injection into a water pool in long-term transients is computationally expensive due to the need to resolve simultaneously the smallest space and time scales of individual steam bubbles and the scales of the whole PSP. To enable PSP analysis for practical purposes, Effective Heat source and Effective Momentum source (EHS/EMS) models have been proposed that avoid the need to resolve steam-water interface. This paper aims to implement mechanistic approaches previously developed by authors for the simulation of transient thermal stratification and mixing phenomena induced by steam injection through spargers in a Nordic BWR PSP. The latest version of the EHS/EMS models using the ‘Unit cell’ approach has been validated against integral effect pool tests and applied to plant simulations. Several scenarios with boundary conditions corresponding to postulated accident sequences were simulated to investigate the possibility of stratification development and the effects of activation of different systems (e.g., blowdown pipes, high momentum nozzle) on the pool behavior.
{"title":"CFD simulation of thermal stratification and mixing in a Nordic BWR pressure suppression pool","authors":"Xicheng Wang, Govatsa Acharya, Dmitry Grishchenko, Pavel Kudinov","doi":"10.1016/j.net.2024.07.045","DOIUrl":"10.1016/j.net.2024.07.045","url":null,"abstract":"<div><div>Boiling Water Reactor (BWR) employs the Pressure Suppression Pool (PSP) as a heat sink to prevent overpressure of the reactor vessel and containment. Steam can be injected into the PSP through spargers in normal and accident conditions and through blowdown pipes in case of a loss of coolant accident (LOCA). There is a safety limit on the maximum PSP temperature at which such steam injection might cause dynamic loads on the containment structures. The performance of the pool can be affected if thermal stratification is developed when temperature of the hot layer grows rapidly while cold layer remains inactive. Simulation of pool behavior during realistic accident scenarios requires validated models that can sufficiently address the interaction between phenomena, safety systems and operational procedures. Direct modeling of steam injection into a water pool in long-term transients is computationally expensive due to the need to resolve simultaneously the smallest space and time scales of individual steam bubbles and the scales of the whole PSP. To enable PSP analysis for practical purposes, Effective Heat source and Effective Momentum source (EHS/EMS) models have been proposed that avoid the need to resolve steam-water interface. This paper aims to implement mechanistic approaches previously developed by authors for the simulation of transient thermal stratification and mixing phenomena induced by steam injection through spargers in a Nordic BWR PSP. The latest version of the EHS/EMS models using the ‘Unit cell’ approach has been validated against integral effect pool tests and applied to plant simulations. Several scenarios with boundary conditions corresponding to postulated accident sequences were simulated to investigate the possibility of stratification development and the effects of activation of different systems (e.g., blowdown pipes, high momentum nozzle) on the pool behavior.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"56 12","pages":"Pages 5357-5376"},"PeriodicalIF":2.6,"publicationDate":"2024-07-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141771274","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 : 2024-07-22DOI: 10.1016/j.net.2024.07.044
Rashed MD. Sardar , Akhmed M. Baisov
The evaluation of available Look-Up Tables for prediction of heat transfer coefficient distribution in rod bundles cooled by supercritical water with the aim of their further use in computational analyses of various Fuel Assembles of Supercritical Water-Cooled Reactors is made. The comparison between the calculations based on Look-Up Tables with the values from empirical correlations and experimental data for smooth and wire-wrapped rod bundles was presented. The obtained results showed that Look-Up Table of the University of Ottawa, which was created to describe improved and deteriorated heat transfer regimes in round tubes, allows describing available data points with 30 % of the mean square deviation. It is noted that the presence of wire intensifies heat transfer exchange near pseudocritical temperature region but existing versions of Look-Up Tables cannot take into account this effect. Nevertheless, there is potential for further improvement in predicting the heat transfer coefficient using Look-Up Table by introducing additional correction factors.
{"title":"Assessment of Look-up tables for the prediction of heat transfer coefficient distribution in rod bundles cooled by supercritical water","authors":"Rashed MD. Sardar , Akhmed M. Baisov","doi":"10.1016/j.net.2024.07.044","DOIUrl":"10.1016/j.net.2024.07.044","url":null,"abstract":"<div><div>The evaluation of available Look-Up Tables for prediction of heat transfer coefficient distribution in rod bundles cooled by supercritical water with the aim of their further use in computational analyses of various Fuel Assembles of Supercritical Water-Cooled Reactors is made. The comparison between the calculations based on Look-Up Tables with the values from empirical correlations and experimental data for smooth and wire-wrapped rod bundles was presented. The obtained results showed that Look-Up Table of the University of Ottawa, which was created to describe improved and deteriorated heat transfer regimes in round tubes, allows describing available data points with 30 % of the mean square deviation. It is noted that the presence of wire intensifies heat transfer exchange near pseudocritical temperature region but existing versions of Look-Up Tables cannot take into account this effect. Nevertheless, there is potential for further improvement in predicting the heat transfer coefficient using Look-Up Table by introducing additional correction factors.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"56 12","pages":"Pages 5346-5356"},"PeriodicalIF":2.6,"publicationDate":"2024-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141771272","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 : 2024-07-22DOI: 10.1016/j.net.2024.07.046
Fan Miao , Bin Zhang , Tianci Xie , Hao Yang , Jianqiang Shan
The ACP100 is a small modular reactor (SMR) designed and built in China, featuring an integrated primary loop and passive safety systems. In SMRs, the interaction between the reactor coolant system, containment vessel, and other systems is highly interdependent. Therefore, it is necessary to use high-precision analysis codes, which can be achieved by coupling multiple codes. This paper is the first part of a study which investigates the LOCA in the ACP100 using a coupled platform. In this paper, a coupling platform was developed for transient analysis of SMR accidents, based on the system code NUSOL-SYS and the Integrated Severe Accident Analysis (ISAA) code. An inter-process communication module was developed adopting shared memory and event objects to exchange data between the two codes. A coupling interface defining the data to be exchanged was proposed. Both NUSOL-SYS and ISAA were modified to perform synchronous time step control. The coupling platform were validated through a hypothetical scenario and Edwards’ pipe blowdown experiment, demonstrating exact consistency and high accuracy. This coupling platform offers a new method for SMR accident analysis, providing a foundation for future work.
{"title":"Analysis of double-ended guillotine break accident of surge line in ACP100 based on coupling method: Development and validation of the coupling method","authors":"Fan Miao , Bin Zhang , Tianci Xie , Hao Yang , Jianqiang Shan","doi":"10.1016/j.net.2024.07.046","DOIUrl":"10.1016/j.net.2024.07.046","url":null,"abstract":"<div><div>The ACP100 is a small modular reactor (SMR) designed and built in China, featuring an integrated primary loop and passive safety systems. In SMRs, the interaction between the reactor coolant system, containment vessel, and other systems is highly interdependent. Therefore, it is necessary to use high-precision analysis codes, which can be achieved by coupling multiple codes. This paper is the first part of a study which investigates the LOCA in the ACP100 using a coupled platform. In this paper, a coupling platform was developed for transient analysis of SMR accidents, based on the system code NUSOL-SYS and the Integrated Severe Accident Analysis (ISAA) code. An inter-process communication module was developed adopting shared memory and event objects to exchange data between the two codes. A coupling interface defining the data to be exchanged was proposed. Both NUSOL-SYS and ISAA were modified to perform synchronous time step control. The coupling platform were validated through a hypothetical scenario and Edwards’ pipe blowdown experiment, demonstrating exact consistency and high accuracy. This coupling platform offers a new method for SMR accident analysis, providing a foundation for future work.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"56 12","pages":"Pages 5377-5385"},"PeriodicalIF":2.6,"publicationDate":"2024-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141784991","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 : 2024-07-20DOI: 10.1016/j.net.2024.07.031
Dany Mulyana , Topan Setiadipura , I Wayan Ngarayana
This technical policy study investigated the effect of heavy metal loading (HML) quantity in the pebble fuel on the special nuclear material (SNM) and waste quantities of an uprated Indonesian pebble bed reactor design, the Reaktor Daya Eksperimental (RDE). A set of Monte Carlo simulations, performed using OpenMC, was deployed to simulate the pebble fuel depletion using an infinite lattice reactor calculation. This study considers HML, fuel residence time, leftover 235U, total Pu, medium- and long-lived wastes, number of depleted pebbles, and volume of waste at different HML values, at attainable discharged fuel burnup, at a target burnup of 80 GWd/MTU and at different reactor powers (10 and 40 MWt). Due to an under-moderation effect, a higher HML quantity shortened the attainable discharged fuel burnup level, which also translated to a lower fuel utilization. An 80% higher HML quantity resulted in about a 25% lower burnup level. Assuming a five-pass refueling scheme, this resulted in approximately three times more leftover 235U, 2.4 times more Pu, 1.4 times more medium-lived waste, and 1.5 times more long-lived waste collections per year. The study showed that a power uprating by a factor of four increased the SNM and waste quantities by four times.
{"title":"Heavy metal loading effects on special nuclear material and waste of an uprated Indonesian Reaktor Daya Eksperimental","authors":"Dany Mulyana , Topan Setiadipura , I Wayan Ngarayana","doi":"10.1016/j.net.2024.07.031","DOIUrl":"10.1016/j.net.2024.07.031","url":null,"abstract":"<div><div>This technical policy study investigated the effect of heavy metal loading (HML) quantity in the pebble fuel on the special nuclear material (SNM) and waste quantities of an uprated Indonesian pebble bed reactor design, the Reaktor Daya Eksperimental (RDE). A set of Monte Carlo simulations, performed using OpenMC, was deployed to simulate the pebble fuel depletion using an infinite lattice reactor calculation. This study considers HML, fuel residence time, leftover <sup>235</sup>U, total Pu, medium- and long-lived wastes, number of depleted pebbles, and volume of waste at different HML values, at attainable discharged fuel burnup, at a target burnup of 80 GWd/MTU and at different reactor powers (10 and 40 MWt). Due to an under-moderation effect, a higher HML quantity shortened the attainable discharged fuel burnup level, which also translated to a lower fuel utilization. An 80% higher HML quantity resulted in about a 25% lower burnup level. Assuming a five-pass refueling scheme, this resulted in approximately three times more leftover <sup>235</sup>U, 2.4 times more Pu, 1.4 times more medium-lived waste, and 1.5 times more long-lived waste collections per year. The study showed that a power uprating by a factor of four increased the SNM and waste quantities by four times.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"56 12","pages":"Pages 5239-5247"},"PeriodicalIF":2.6,"publicationDate":"2024-07-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141839333","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 : 2024-07-20DOI: 10.1016/j.net.2024.07.042
Xiaoliang Zou , Yanting Sun , Qiusun Zeng , Xiaojian Wen , Xiaogang Cao , Xi Huang , Yibao Liu
The space nuclear reactor cooled by heat pipes has become the preferred choice for future space missions and deep space exploration missions. The use of low-enriched uranium (LEU) is promoted to achieve the goal of nuclear non-proliferation worldwide. In this study, a lithium heat pipe cooled space reactor with LEU (HP-LEU) was proposed based on Heat Pipes-Segmented Thermoelectric Module Converters (HP-STMCs), with the addition of moderators. The HP-LEU employs yttrium hydride (YH2) as the moderator and 19.9 % enriched uranium nitride (UN) as the fuel. The neutronics analysis has been performed on the HP-LEU reactor and the results have showed that the HP-LEU has a lifetime of more than 12 years. Two control systems have been applied in the reactor and have demonstrated the capacity to independently regulate and shut down the reactor. The total temperature reactivity coefficients are consistently negative, indicating that the HP-LEU reactor is inherently safe during operation. During normal operation, the temperatures of the materials are all acceptable. This study can serve as a reference for lithium heat pipe cooled space reactors with LEU.
用热管冷却的空间核反应堆已成为未来空间任务和深空探测任务的首选。为实现全球核不扩散的目标，低浓缩铀（LEU）的使用得到了推广。本研究基于热管-分段式热电模块转换器（HP-STMCs），提出了一种使用 LEU 的锂热管冷却空间反应堆（HP-LEU），并增加了慢化剂。HP-LEU 采用氢化钇（YH）作为慢化剂，19.9% 的浓缩氮化铀（UN）作为燃料。对 HP-LEU 反应堆进行了中子分析，结果表明 HP-LEU 的寿命超过 12 年。反应堆采用了两套控制系统，并证明其具有独立调节和关闭反应堆的能力。总温度反应系数始终为负值，表明 HP-LEU 反应堆在运行期间本质上是安全的。在正常运行期间，材料的温度都是可以接受的。这项研究可作为使用 LEU 的锂热管冷却空间反应堆的参考。
{"title":"Preliminary neutronics design and analysis of lithium heat pipe cooled space reactor with low-enriched uranium","authors":"Xiaoliang Zou , Yanting Sun , Qiusun Zeng , Xiaojian Wen , Xiaogang Cao , Xi Huang , Yibao Liu","doi":"10.1016/j.net.2024.07.042","DOIUrl":"10.1016/j.net.2024.07.042","url":null,"abstract":"<div><div>The space nuclear reactor cooled by heat pipes has become the preferred choice for future space missions and deep space exploration missions. The use of low-enriched uranium (LEU) is promoted to achieve the goal of nuclear non-proliferation worldwide. In this study, a lithium heat pipe cooled space reactor with LEU (HP-LEU) was proposed based on Heat Pipes-Segmented Thermoelectric Module Converters (HP-STMCs), with the addition of moderators. The HP-LEU employs yttrium hydride (YH<sub>2</sub>) as the moderator and 19.9 % enriched uranium nitride (UN) as the fuel. The neutronics analysis has been performed on the HP-LEU reactor and the results have showed that the HP-LEU has a lifetime of more than 12 years. Two control systems have been applied in the reactor and have demonstrated the capacity to independently regulate and shut down the reactor. The total temperature reactivity coefficients are consistently negative, indicating that the HP-LEU reactor is inherently safe during operation. During normal operation, the temperatures of the materials are all acceptable. This study can serve as a reference for lithium heat pipe cooled space reactors with LEU.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"56 12","pages":"Pages 5330-5338"},"PeriodicalIF":2.6,"publicationDate":"2024-07-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141784992","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 : 2024-07-20DOI: 10.1016/j.net.2024.07.043
Jowi Rapha P. Cruz , Alvie A. Astronomo , Gil Nonato Santos , Ryan U. Olivares
Since 1988, the Philippines lacked local access to a nuclear facility, creating a significant void in this field of study for Filipinos. However, after a hiatus of 34 years, this gap was addressed with the recent authorization granted to Philippine Research Reactor 1 (PRR-1) Subcritical Assembly for Training, Education, and Research (SATER), allowing it to resume operations. In this work, a PHITS-based computational model was developed for the recently commissioned PRR-1 SATER. The model utilized a simplified model of the Training, Research, Isotope, General Atomics (TRIGA) fuel that releases photons with 0.6617 MeV energy from the Cs-137 fission product in the fuel. Compared to previous works on photon transport mapping, which utilized average source definition, this study employed individually defined fuel intensities and compared them with the averaged source definition for the fuel. The two fuel source definitions showed noticeable differences inside the reactor tank which is relevant for mixed-field irradiation applications of a research reactor. However, defining the fuel rods by their average strength is sufficient for radiation protection purposes. Simulations were also performed for fuel source definitions based on the average and ±1 standard deviation of the gamma intensity. Gamma doses received by cylindrical phantoms positioned at 0.5 m from the surface of the reactor tank for 500 h were found to be 1 % of the radiation dose limits per year and 4 % of the average dose limit for 5 years as stipulated by the Code of Philippine Nuclear Research Institute (PNRI) Regulations. Loss of water accident was also analyzed based on a conservative exposure time of 500 h. This resulted in a dose value that is only 45.5 % of the dose identified as the emergency turnback guidance of the IAEA. Lastly, PHITS calculated values of gamma doses were found to agree well, with 0.98 ratio, when compared with gamma doses measured at specified locations in the reactor. Results of this study confirm the inherent safety of the PRR-1 SATER in terms of radiological shielding for Cs-137 photons.
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