Annals of Nuclear Energy最新文献

{"title":"Improving neutron economy of a Lead-Bismuth Eutectic-Cooled reactor using Modified-CANDLE axial fuel shuffling","authors":"Nina Widiawati ,&nbsp;R.Andika Putra Dwijayanto ,&nbsp;Fitria Miftasani ,&nbsp;Nuri Trianti ,&nbsp;Anni Nuril Hidayati ,&nbsp;Maryam Afifah ,&nbsp;Ratna Dewi Syarifah ,&nbsp;Zaki Su'ud","doi":"10.1016/j.anucene.2025.111328","DOIUrl":"10.1016/j.anucene.2025.111328","url":null,"abstract":"<div><div>LWRs do not effectively utilize natural uranium as fuel, harnessing less than 1% of its energy potential. Furthermore, initial studies show that conventional fast reactors require higher uranium enrichment and fuel reprocessing, which are expensive technology. The Modified-CANDLE strategy enables the direct use of natural uranium without enrichment or reprocessing. This study implements modified-CANDLE axial fuel shuffling in a lead–bismuth eutectic (LBE)-cooled fast reactor to examine its impact on reactor performance compared to radial fuel shuffling. The calculations use the SRAC code and the JENDL 4.0 nuclear data library. This research evaluates the effective multiplication factor (k-eff), power density, power peaking factor (PPF), and neutron leakage over a 15-year fuel cycle. The research indicates that axial fuel shuffling is more efficient at sustaining the reactor’s reactivity than radial shuffling. Therefore, axial fuel shuffling can improve neutron economy compared to radial fuel shuffling.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111328"},"PeriodicalIF":1.9,"publicationDate":"2025-03-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143550075","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}

{"title":"A review of the scientific contributions by Barry Ganapol","authors":"P. Ravetto ,&nbsp;P. Saracco","doi":"10.1016/j.anucene.2025.111253","DOIUrl":"10.1016/j.anucene.2025.111253","url":null,"abstract":"<div><div>Barry Ganapol has been an outstanding personality within the transport theory and the nuclear reactor physics communities throughout the past 50 years. He is universally recognized as one of the scientists who has given relevant contributions to the fields of mathematics and computations in nuclear science and he has shown an uncommon capability to export this knowledge to different areas, with a rare attitude towards interdisciplinarity. He has obtained significant achievements in the development of methods for the solution of the linear transport equation, in reactor physics and in radiative transport analysis. He is widely known and appreciated for having established a wide variety of accurate benchmarks for many problems within these fields. This paper reviews the scientific production of Barry Ganapol, highlighting what are believed to be his most important scientific accomplishments.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111253"},"PeriodicalIF":1.9,"publicationDate":"2025-03-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143550074","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}

{"title":"Integrated thermal power measurement in the modified STACY for the performance inspections","authors":"Shouhei Araki ,&nbsp;Eiju Aizawa ,&nbsp;Takahiko Murakami ,&nbsp;Yu Arakaki ,&nbsp;Yuta Tada ,&nbsp;Yutaka Kamikawa ,&nbsp;Kenta Hasegawa ,&nbsp;Tomoki Yoshikawa ,&nbsp;Masato Sumiya ,&nbsp;Masakazu Seki ,&nbsp;Junichi Ishii ,&nbsp;Kazuhiko Izawa ,&nbsp;Daiki Iwahashi ,&nbsp;Shigeki Shiba ,&nbsp;Satoshi Gunji","doi":"10.1016/j.anucene.2025.111323","DOIUrl":"10.1016/j.anucene.2025.111323","url":null,"abstract":"<div><div>Japan Atomic Energy Agency (JAEA) has modified the Static Experimental Critical Facility (STACY) to a heterogeneous system using fuel rods in order to obtain criticality characteristics of fuel debris. Thermal power measurement was required for the calibration of a power meter system in the modified STACY in order to conduct a series of performance inspections and operation. We measured the thermal power using the modified activation method that combined MVP and PHITS code because neutron flux distribution cannot be measured experimentally. Four operations were conducted for the thermal power measurement. The power meter was calibrated by using three operational data and tested with one operational data. It was found that the calibration of the power meter system was successfully agreed within 3% with the results measured by the activation method.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111323"},"PeriodicalIF":1.9,"publicationDate":"2025-03-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143550076","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}

{"title":"Sensitivity analysis of parameters and optimization of produced entropy in a nuclear reactor with equilateral fuel rod arrangement and variable angle turbulators","authors":"Abdullah A.A.A. Al-Rashed,&nbsp;Abdulwahab A. Alnaqi,&nbsp;Jalal Alsarraf","doi":"10.1016/j.anucene.2025.111310","DOIUrl":"10.1016/j.anucene.2025.111310","url":null,"abstract":"<div><div>This paper investigates the produced entropy (ENT), including thermal ENT, fluid loss, total ENT, and the Bejan number (Be) in a nuclear reactor with equilateral fuel rod arrangements. The fuel rods are placed within tubes that have diameters varying from 3.1 to 10.1 cm. The rods are vertically positioned in the reactor at a fixed distance, and nanofluids are used to cool them. Turbulator (TUR) blades with angles varying from 0 to 6 degrees and heights from 0 to 13 cm are employed in the nanofluid flow. To enhance the analysis, machine learning methods are used for sensitivity analysis of parameters and optimization of outputs. The results of this study indicate that increasing the tube diameter and the height of the TUR blades leads to an increase in fluid loss ENT, while changes in blade angle have a minimal effect on fluid loss ENT. Increasing the tube diameter and blade length significantly increases thermal ENT and total ENT. Blade angle also has an effect, with thermal ENT decreasing as the angle increases. Changes in blade height, angle, and tube diameter containing the fluid flow result in a 1.8 % variation in the Be. The results demonstrate that tube diameter, blade length, and TUR blade angle are all important parameters in determining the Be in the reactor. An increase in tube diameter significantly raises the Be, while an increase in blade length decreases it. The angle of the blades also affects the Be, with an increase in angle leading to a decrease in the Be.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111310"},"PeriodicalIF":1.9,"publicationDate":"2025-03-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143550073","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}

{"title":"Experimental study on flux allocation characteristics of annular fuel bundle channel","authors":"Jinyang Li ,&nbsp;Shouxu Qiao ,&nbsp;Sijia Hao ,&nbsp;Yu Zou ,&nbsp;Xupeng Li ,&nbsp;Sichao Tan ,&nbsp;Ruifeng Tian","doi":"10.1016/j.anucene.2025.111316","DOIUrl":"10.1016/j.anucene.2025.111316","url":null,"abstract":"<div><div>Annular fuel allows coolant to flow on both the inner and outer sides of the fuel, thereby enlarging the surface area-to-volume ratio for heat transfer and increases the power density. Due to the different resistance characteristics of the inner and outer channels, the flux allocation ratio varies with the change in inlet velocity, which directly affects the heat transfer efficiency of the annular fuel assembly. This study used the Particle Image Velocimetry (PIV) technique and the rotation integration method to directly measure the flux in the inner and outer channels of a 5 × 5 annular fuel bundle. The results show that the flux allocation ratio initially decreases rapidly, then more slowly, and finally increases as the inlet velocity increases. Under pulsating flow conditions, the flux allocation ratio exhibits a periodic variation. The pulsating acceleration and the flux allocation ratio are approximately linearly correlated.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111316"},"PeriodicalIF":1.9,"publicationDate":"2025-03-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143534352","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}

{"title":"Correcting nuclide concentration fields and optimizing monitoring locations based on environmental-monitoring data","authors":"Yue Qi ,&nbsp;Yang Jie ,&nbsp;Chen Jiachen ,&nbsp;Meng Binchi ,&nbsp;Shi Yikun ,&nbsp;Wu Feifei ,&nbsp;Lian Bing ,&nbsp;Wang Yan","doi":"10.1016/j.anucene.2025.111326","DOIUrl":"10.1016/j.anucene.2025.111326","url":null,"abstract":"<div><div>Radioactive materials can be released into the external environment after nuclear facility accidents. Therefore, it is necessary to rapidly assess the spatial distribution of radioactive nuclides. Using only the concentration distribution of radioactive nuclides predicted by the atmospheric dispersion models may result in deviations from the actual concentration distribution. To make the predicted distribution more consistent with the actual distribution, a method based on Kriging interpolation was developed to rapidly correct the predicted concentration distribution using a small amount of environmental aerosol monitoring data. In addition, a method for optimizing environmental aerosol monitoring locations based on a differential evolution algorithm was developed to guide the placement of more representative monitoring locations and improve the accuracy of the corrected concentration distribution. The results indicated that using optimized monitoring locations to obtain environmental monitoring data enhanced the accuracy of prediction in key areas in the constructed case-studies, reducing false-negative areas by 97%.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111326"},"PeriodicalIF":1.9,"publicationDate":"2025-03-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143534351","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}

{"title":"Prediction of composite neutron source spectra by combination of JENDL-5 and PHITS","authors":"Tatsuhiko Ogawa","doi":"10.1016/j.anucene.2025.111256","DOIUrl":"10.1016/j.anucene.2025.111256","url":null,"abstract":"<div><div>A novel robust method has been developed to simulate the performance of composite neutron sources composed of an alpha-emitting actinide and a light nucleus with low neutron separation energy. This method is based on the JENDL-5 cross-section data library and the Monte-Carlo radiation transport code PHITS. In contrast to previously devised methods, this approach can predict various quantities of the sources, such as actinide grain size dependence, absolute neutron emission intensity, energy spectra of neutrons and parasitic photons, neutron multiplicity, and time structure, with little approximation.</div><div>The accurate calculation of stopping power of alpha rays in actinide grains and light elements, as well as the use of (<span><math><mi>α</mi></math></span>, n) reaction evaluated cross sections, which is one of the unique features of PHITS Ver.3.34 and its later versions, are the essences of the method. This method allows for the calculation of quantities important for practical applications, such as detection signal frequency, coincidence event rate, and the impact of parasitic gamma-rays.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111256"},"PeriodicalIF":1.9,"publicationDate":"2025-03-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143534218","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}

{"title":"Numerical study on flow instability development of natural circulation in narrow channel under medium-to-high pressure","authors":"Yao Yao ,&nbsp;Tao Zhou ,&nbsp;Jianyu Tang ,&nbsp;Dongli Huang ,&nbsp;Zefeng Wang","doi":"10.1016/j.anucene.2025.111322","DOIUrl":"10.1016/j.anucene.2025.111322","url":null,"abstract":"<div><div>Natural circulation flow instability is an important phenomenon for nuclear system safety, especially in narrow channels which have a broad application prospect. However, the small gap of narrow channels can easily lead to the narrow-space effect, resulting in mechanisms and performance of instability different from those in conventional channels. Medium-to-high pressure is a common condition in nuclear systems. Therefore, it is necessary to study flow instability in narrow channels under medium-to-high pressure. The ultimate goal is to effectively avoid the instability and improve the safety of nuclear systems. This manuscript investigates the characteristics of natural circulation instability in a narrow channel with deionized water under medium-to-high pressure (7.0 MPa-15.0 MPa). Results provide new insights into the safety and reliability of nuclear reactor cooling systems, supporting the establishment of more accurate and reliable natural circulation systems. A numerical study establishes a model of natural circulation loop including a narrow channel via system code RELAP5 based on a well-validated natural circulation test facility. The instability development is divided into three stages, natural circulation stable flow, instability flow (no reverse flow), and instability flow (reverse flow) or periodic dryout instability, according to the amplitudes and periods of mass flow rate. The proposed RELAP5 model demonstrates the relationship between mass flow rate and pressure drop on a time scale of a few seconds in every stage. Mass flow rate, exit quality, two-phase length, and flow regime are examined throughout all stages. Results indicate that increasing system pressure and inlet subcooling have a significant inhibitory effect on instability oscillation.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111322"},"PeriodicalIF":1.9,"publicationDate":"2025-03-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143534219","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}

{"title":"Research on steady-state behavior of natural circulation with internally heated fluids of liquid-fueled molten salt reactor","authors":"Ying Cao ,&nbsp;Weishi Wan ,&nbsp;Chong Zhou","doi":"10.1016/j.anucene.2025.111320","DOIUrl":"10.1016/j.anucene.2025.111320","url":null,"abstract":"<div><div>Liquid-fueled molten salt reactors (LFMSR) utilize internally heated fuel-salt compounds as the working fluid for the first time and the heat source and heat sink conditions are crucial for the natural circulation phenomena. This paper explored the steady-state behaviors of natural circulation of LFMSR and analyzed the impact of internal heat. By incorporating a heat source term into the liquid phase, the natural circulation governing differential equations were modified and solved using a self-developed Python code. Then, the homogeneously volumetric heat model of LFMSR was compared with the core central heat model of a traditional solid fuel reactor, the LFMSR exhibited a smaller steady-state natural circulation flow reducing the natural circulation thermal effect. Subsequently, an extended RELAP5/SCDAPSIM/MOD4.0 code for LFMSR was applied to verify the self-developed code, and their quantitative steady-state parameters showed a good match, which can be used for more complex system analyses in future works.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111320"},"PeriodicalIF":1.9,"publicationDate":"2025-03-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143534353","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}

{"title":"Influence of narrow gaps between the adjacent components on graphite lifetime in molten salt reactors","authors":"Qi Wang ,&nbsp;Yu Zhong ,&nbsp;Chenggang Yu ,&nbsp;Jianhui Wu ,&nbsp;Zhichao Wang ,&nbsp;Wei Guo ,&nbsp;Jingen Chen ,&nbsp;Xiangzhou Cai","doi":"10.1016/j.anucene.2025.111317","DOIUrl":"10.1016/j.anucene.2025.111317","url":null,"abstract":"&lt;div&gt;&lt;div&gt;As one of the prospective Generation IV reactor designs, molten salt reactors (MSRs) are receiving increasing attention due to its unique advantages. Graphite is a preferred moderator in an MSR with the excellent moderation capability, good compatibility with the fuel salt and the stability under neutron radiation. However, the narrow gaps filled with fuel salt are existed between adjacent graphite assemblies due to the fabrication tolerances, the irradiation-induced deformation, and the different thermal expansion coefficients between the support structure and graphite. As the fuel salt acts as both a heat source and coolant, the effect of narrow gaps on the component performance and their impact on graphite lifespan is necessary to be evaluated. This study analyzes the specific effects of narrow gaps between the adjacent graphite components in a round channel assembly (RCA) and round groove assembly (RGA) of a small modular molten salt reactor (SM-MSR), respectively. The results show that the performances of graphite assemblies with the narrow gaps significantly differ from those of ideal assemblies with no gaps. In terms of the component structure, the variations on the neutronic and thermohydraulic performances between RGA and RCA are primarily caused by the distinct positions of the narrow gap in each component. One the one hand, the heat flux distribution indicates that the component structures have a significant impact on the overall cooling capacity, where RGA exhibited a more than 19.98 % reduction in the heat removal through the primary channel to the graphite compared to RCA with a 0.5 mm narrow gap. The effective cooling range extends from inlet to the outlet with the increased narrow gap width. On the other hand, the flow exchange between the fuel salt in channels is enhanced in RGA due to the connectivity of the primary and narrow gap channels. Furthermore, the hotspot in the component is mitigated by the increase in the mass flow rate of fuel salt through the narrow gap flow channel, which is more prominent with the widening of the narrow gap. Therefore, the adverse effects on the temperature distribution caused by the narrow gap can be alleviated in RGA compared to RCA. Due to the enhanced flow when the narrow gap width is increased from 0.5 mm to 3 mm, the temperature gradient is decreased by 68.56 % and 62.38 % for RCA and RGA, respectively. The graphite lifetime is also evaluated under the considerations of narrow gap width and component structure. When the narrow gap width is expanded to be 3 mm, the graphite lifetime of RGA is observed to be 7.47 % longer than that of RCA. The component lifespan is slightly improved with the increased width of the narrow gap. For the narrow gap width of 0.5 mm, the lifespan of RCA and RGA is reduced by 30.88 % and 22.43 %, respectively, compared to the ideal assembly with no narrow gap. Furthermore, when the narrow gap width of RGA reached to be 3 mm, the graphite lifespan is superi","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111317"},"PeriodicalIF":1.9,"publicationDate":"2025-03-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143550104","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}