Progress in Nuclear Energy最新文献

{"title":"A general correlation for saturated flow boiling heat transfer in vertical narrow rectangular channel based on test data from specialized experiment","authors":"Bo Kuang ,&nbsp;Pengfei Liu ,&nbsp;Jian Deng ,&nbsp;Shuhua Ding ,&nbsp;Dan Wu","doi":"10.1016/j.pnucene.2025.105668","DOIUrl":"10.1016/j.pnucene.2025.105668","url":null,"abstract":"<div><div>Due to the special geometry of narrow rectangular channels, their heat transfer characteristics differ from conventional channels. Heat transfer data and visualization images within a wide parameter range are obtained in this study, through heat transfer tests and visualization tests. The boiling curves of rectangular narrow channel under different parameter conditions are given, and the effect of parameters on the heat transfer coefficient is analyzed. It is indicated that mass flux, pressure, and gap size have an impact on the heat transfer coefficient. In addition, the boiling heat transfer mechanism exhibits a dominant mode transition from nucleate boiling to forced convective vaporization. In response to the asymmetric confinement effect on the vapor phase in narrow rectangular channels, the vapor velocity correction coefficient and vapor size correction coefficient are obtained through vapor force balance analysis, to characterize the effects of convection in liquid film evaporation and asymmetric bubble growth, respectively. By combining the liquid film confinement number and mechanism transition function applicable to rectangular narrow channels, a heat transfer correlation suitable for rectangular narrow channels is obtained, and the results showed that the new equation was in good agreement with the experimental results.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"183 ","pages":"Article 105668"},"PeriodicalIF":3.3,"publicationDate":"2025-02-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143445640","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":"Development of novel oval-twisted helical tube once-through steam generator: Part I: Single-phase laminar flow","authors":"Scott Wahlquist ,&nbsp;Amir Ali ,&nbsp;Kyle Schroeder ,&nbsp;Su-Jong Yoon ,&nbsp;Piyush Sabharwall","doi":"10.1016/j.pnucene.2025.105667","DOIUrl":"10.1016/j.pnucene.2025.105667","url":null,"abstract":"<div><div>The Helically coiled tube Once-Through Steam Generator (H-OTSG) offers enhanced overall thermal performance due to its high surface area per unit volume and secondary flow generation. The H-OTSG design (circular cross-sectional flow area) is considered for various Small Modular Reactor (SMR) designs due to its compactness, practical manufacturability, and higher thermal efficiency than current OTSG technologies (e.g., shell and tube). A novel conceptual H-OTSG tube design, known as the Oval-Twisted Helically coiled tube Once-Through Steam Generator (OTH-OTSG), is proposed for nuclear applications. The concept is a helically twisted coil with an oval cross-sectional flow area. The OTH-OTSG geometry simultaneously enhances the heat transfer performance through double-induced secondary flow. One through the tube twist generated swirling flow, and the second by the centrifugal force offered by the helical shape. Technology development requires two significant steps. The first is investigating the thermal performance of single-phase flow, which is the focus of this study, to develop heat transfer and friction loss models. In this step, the single-phase heat transfer performance is evaluated and compared with the H-OTSG. The OTH-OTSG and H-OTSG are evaluated for various mean coil diameters (<span><math><mrow><msub><mi>D</mi><mi>c</mi></msub></mrow></math></span> = 75–225 mm), laminar single-phase flow regime (<span><math><mrow><mi>R</mi><mi>e</mi></mrow></math></span> = 200–2000), and isothermal wall conditions (<span><math><mrow><msub><mi>T</mi><mrow><mi>w</mi><mi>a</mi><mi>l</mi><mi>l</mi></mrow></msub></mrow></math></span> = 320–350 K). Over the range of investigated parameters, the new proposed OTH-OTSG geometry shows an enhanced heat transfer rate (<span><math><mrow><msub><mi>q</mi><mi>o</mi></msub></mrow></math></span>/ <span><math><mrow><msub><mi>q</mi><mi>C</mi></msub></mrow></math></span> = 28.29–59.78%), heat transfer coefficient (<span><math><mrow><msub><mrow><mi>N</mi><mi>u</mi></mrow><mi>o</mi></msub><mo>/</mo><msub><mrow><mi>N</mi><mi>u</mi></mrow><mi>c</mi></msub></mrow></math></span> = 19.14–39.06%) and increased friction coefficient (<span><math><mrow><msub><mi>f</mi><mi>o</mi></msub><mo>/</mo><msub><mi>f</mi><mi>c</mi></msub></mrow></math></span> = 15.07–41.84%) over the single-phase flow in H-OTSG with a circular cross-sectional flow area. The presented heat transfer (<span><math><mrow><mi>N</mi><mi>u</mi></mrow></math></span>) and friction loss (<span><math><mrow><mi>f</mi></mrow></math></span>) models are within <span><math><mrow><mo>±</mo></mrow></math></span> 9 and <span><math><mrow><mo>±</mo></mrow></math></span> 13% with all data, respectively. The heat transfer and hydrodynamic models developed through this stage will be implemented in a two-phase flow study for the single-phase flow regime at the early stage of the innovative OHT-OTSG study, which is the next step in the development process.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"183 ","pages":"Article 105667"},"PeriodicalIF":3.3,"publicationDate":"2025-02-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143438230","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":"Design of fuzzy adaptive fractional-order controller for total power control of AHWR","authors":"Devbrat Gupta ,&nbsp;Jitendra Kumar ,&nbsp;Vishal Goyal ,&nbsp;K.N.V. Sairam","doi":"10.1016/j.pnucene.2025.105641","DOIUrl":"10.1016/j.pnucene.2025.105641","url":null,"abstract":"<div><div>In this article, a fuzzy adaptive fractional-order proportional plus derivative controller for total power control of an Advanced Heavy Water Reactor is designed in the presence of thermal hydraulics feedback. In addition to a conventional Fuzzy Logic Control layer (FLC), the proposed controller uses a self-tuning FLC layer for run-time adaption of the controller output. A genetic algorithm (GA) based optimization method is used for tuning the controller parameters. Transient Simulation studies are carried out to demonstrate the efficacy of the proposed controller for set-point changes and feedwater temperature disturbances. Results of the proposed controller are compared with those of the traditional proportional plus derivative (PD) and fractional-order proportional plus derivative (FOPD) controllers using standard performance indices. It is observed that the AFOPD controller has 13.99% performance improvement over the FOPD controller and 20.53% improvement over the PD controller for considered objective function. At various operating conditions also, the AFOPD controller shows similar superiority over FOPD and PD controllers.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"183 ","pages":"Article 105641"},"PeriodicalIF":3.3,"publicationDate":"2025-02-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143438231","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":"Post-irradiation examination of silicide fuel (U3Si2-Al) with 4.8 gU/cm3 for RSG-GAS","authors":"Aslina Br Ginting ,&nbsp;Rohmad Sigit Eko Budi Prasetyo ,&nbsp;Boybul ,&nbsp;Sungkono ,&nbsp;Supardjo ,&nbsp;Surian Pinem ,&nbsp;Wahid Luthfi","doi":"10.1016/j.pnucene.2025.105673","DOIUrl":"10.1016/j.pnucene.2025.105673","url":null,"abstract":"<div><div>The GA Siwabessy Multipurpose Reactor (RSG-GAS) has a nominal thermal power of 30 MWth and uses silicide fuel (U₃Si₂-Al) with a 2.96 gU/cm³ uranium density. To extend the reactor operating cycle, the RSG-GAS fuel can be converted to use a higher density fuel such as silicide fuel with a 4.8 gU/cm³ density. Irradiation tests of silicide fuel with 4.8 gU/cm³ density in the RSG-GAS core have been carried out to achieve burnup fractions of 20%, 40%, and 60% burned U-235. As part of a study for the development of new research reactor fuels, the post-irradiation examination (PIE) has been done for mini fuel plates that have been irradiated at the RSG-GAS. Several aspects of the PIE that were considered were physical changes, fuel swelling, and burnup distribution on the fuel plate. An in-depth analysis was conducted to verify the performance of 4.8 gU/cm³ silicide fuel so it can be used in the reactor core. Non-destructive testing in the form of visual inspection showed that the dimensions of the fuel plate did not change and there were no physical anomalies that could affect the performance of the silicide fuel plate for up to 60% burnup fraction. Measurement of the axial burnup distribution on the fuel using destructive tests, either based on the isotopic masses of Cs-137 or U-235 showed a consistency. In general, the physical properties of the silicide fuel plate that has been burned up to 60% burnup fraction are still quite relevant so the RSG-GAS fuel plate with a 4.8 gU/cm³ uranium density can be a candidate to replace the typical 2.56 gU/cm³ silicide fuel in RSG-GAS which currently has an average discharge burnup of 56%.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"183 ","pages":"Article 105673"},"PeriodicalIF":3.3,"publicationDate":"2025-02-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143438232","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":"Hydraulic characteristics investigation on a novel multi-plate throttle entry tube of fast reactor control rod subassembly","authors":"Haiqi Qin ,&nbsp;Daogang Lu ,&nbsp;Dawen Zhong ,&nbsp;Qiong Cao ,&nbsp;Xiaowei Li","doi":"10.1016/j.pnucene.2025.105669","DOIUrl":"10.1016/j.pnucene.2025.105669","url":null,"abstract":"<div><div>As the only reactivity control method, the Control Rod Subassembly (CRSA) is employed to control the startup, shutdown and power regulation of Sodium-cooled Fast Reactor (SFR). The entry tube is an essential structure component of CRSA, serving not only as the inlet of liquid sodium but also as the support and guidance for CRSA. To address the shortcomings of traditional straight-through entry tube in terms of throttling effect and experiment cost, a novel Multi-plate Throttle Entry Tube (MTET) is designed and manufactured in this investigation. MTET hydraulic characteristics and geometrical parameter effect are investigated detailed by hydraulic experiments and numerical simulations. Based on the similarity analysis, the alternative hydraulic experiments are conducted on MTETs with different geometrical parameters. Experiment results reveal that MTET can achieve 299.12 kPa at the rated mass flow rate of 1.20 kg/s, with a relative error of −0.29%, which also verifies the rationality of this new MTET. In view of the design requirements of SFR, the optimal geometrical parameters are determined as three throttling holes of 10 mm in diameter and five throttling plates of 41 mm in width. Additionally, numerical results provide an evident that the geometrical parameters are pivotal influence factors to the MTET hydraulic characteristics, such as throttling plate width, throttling hole arrangement, connecting rod length, and throttling plate pitch. Increasing the throttling plate width and extending the throttling plate pitch can both improve the throttling effect of MTET. By comparison, the throttling effect of MTET is more sensitive to the throttling hole arrangement rather than the connecting rod length. It's important to note that the modification of connecting rod length needs to be carefully evaluated. Meanwhile, the non-uniform arrangement in multi-plate throttle is another solution to adjust the MTET hydraulic characteristics, which only needs to modify the width of single throttling plate. Compared to the fifth throttling plate (along the coolant flow direction), the adjustment efficiency of the first throttling plate increases by more than 4%. Therefore, the significant pressure drop should be adjusted by modifying the widths of upstream throttling plates, while the modification of downstream throttling plates is more suitable for the minor pressure drop adjustment. This innovative design demonstrates the improved throttling effect of MTET, as the promising structure optimization for CRSA and other subassembly of SFR.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"182 ","pages":"Article 105669"},"PeriodicalIF":3.3,"publicationDate":"2025-02-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143429035","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":"Analysis schemes to evaluate HELB-induced blast waves","authors":"Tae-Yong Kim,&nbsp;Yoon-Suk Chang","doi":"10.1016/j.pnucene.2025.105665","DOIUrl":"10.1016/j.pnucene.2025.105665","url":null,"abstract":"<div><div>High energy line breaks (HELBs) have been postulated as a design basis event for protection of major structures, systems and components in nuclear power plants (NPPs). However, it was recently pointed out that plausible phenomena such as blast wave and structural resonance caused by jet impact need to be carefully evaluated. In this context, the present study investigates analysis schemes for evaluating the blast wave effects during design stage of new reactors. At first, three types of comparative analyses were conducted with a focus on the steam generator (SG) in an operating 1400 MWe NPP under a representative HELB scenario, from which the most conservative but reasonable one was identified. As the next step, the adequate scheme was applied for parametric analyses of the SG in a developing 60 MWe NPP considering nine hypothesized conditions. Ultimately, structural behaviors and integrity of the smaller SG according to design acceptance criterion as well as impacts of different pipe diameters and rupture distances on dynamic responses were assessed to discuss capability withstanding the HELB-induced blast waves.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"182 ","pages":"Article 105665"},"PeriodicalIF":3.3,"publicationDate":"2025-02-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143420134","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 investigation on flow-induced vibration of LBE-cooled wire-wrapped rod and rod bundle","authors":"Sipeng Wang ,&nbsp;Jingbo Hui ,&nbsp;Zhuoyao Li ,&nbsp;Hang Zhang ,&nbsp;Bao-Wen Yang ,&nbsp;Kun Zhuang","doi":"10.1016/j.pnucene.2025.105664","DOIUrl":"10.1016/j.pnucene.2025.105664","url":null,"abstract":"<div><div>Among the Generation-IV nuclear reactors, LBE-cooled reactors stand out as particularly promising due to their high neutron economy, high thermal conductivity, and high boiling point. However, the high density of lead-bismuth coolants and the mixing effect of helical structures lead to flow-induced vibrations, which can cause fuel rod vibrations leading to the leakage of radioactive materials and posing a significant threat to the safe operation of nuclear reactors. Consequently, this study employs CFD method to conduct an in-depth analysis of the flow-induced vibration characteristics of LBE-cooled wire-wrapped rod and rod bundles under various flow velocities to elucidate the intrinsic link between fluid excitation and structural vibration characteristics. Through multi-scenario analysis, this study not only clarifies the direct relationship between fluid excitation frequencies and fuel rod vibration responses but also integrates micro-scale fluid dynamic characteristics with macro-scale structural vibration characteristics, offering a novel perspective for understanding the physical mechanisms behind flow-induced vibrations. The results indicate that the main low-frequency components in the fluid excitation have induced low-frequency forced vibrations in the fuel rods that correspond to their response frequencies, which is consistent with numerous studies related to vibration. There is a significant difference in vibration behavior between single rod and rod bundle. As for rod bundle, due to the distinct pressure distribution characteristics of center channel and corner channel, the edge rods vibrate in diagonal directions and away from the center with a large amplitude, and the amplitude increase as flow velocity increases. However, flow velocity is not the definitive factor for vibration intensity. The vibration intensity and amplitude of the central rod increase as flow velocity increases, and fuel rods resonate in the studied flow velocities. In this study, not only the relationship between fluid excitation and structural vibration was revealed, but also the vibration characteristics of fuel rods are obtained. This provides a reference for further research and engineering applications in nuclear field.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"182 ","pages":"Article 105664"},"PeriodicalIF":3.3,"publicationDate":"2025-02-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143420133","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 of the dominant frequency of pressure oscillations and condensing pattern in vertically upward steam bubbles jets under rolling conditions","authors":"Pengbo Wei,&nbsp;Zhenghang Luo,&nbsp;Weixiong Chen,&nbsp;Daotong Chong,&nbsp;Junjie Yan","doi":"10.1016/j.pnucene.2025.105655","DOIUrl":"10.1016/j.pnucene.2025.105655","url":null,"abstract":"<div><div>Steam bubbles direct jet condensing technology, known for their exceptional performance in both the transfers of heat or mass, is popular in static and marine industrial energy systems. Nevertheless, it causes intense pressure oscillations, which can be exacerbated by rolling motions in the marine environments. This can severely impact the safe operation of steam bubbles jet equipment. Therefore, the experimental study is carried out to study the effect of rolling motions on condensing pattern and pressure oscillation frequency of vertically upward bubbles jets. The results reveal that, firstly, the condensing pattern of the vapor bubbles changed, and the rolling motions force the vapor bubbles patterns to roll in the same direction as the rolling motions as compared to the static environment. This rolling behavior of the bubbles causes instability during necking and detachment of bubbles and enhances the heat transfers between the bubbles and the subcooled water, which expedites the condensation rate of the bubbles. This phenomenon explains the second finding: the condensing pressure oscillation dominant frequency of vapor bubbles under rolling conditions is larger than that under static conditions. Meanwhile, with a decrease in these rolling periods and an increasing of the maximum angles of rolling, this dominant frequency of bubbles condensing pressure oscillations increases gradually. Finally, a correlation equation is developed to predict dominant frequency of bubbles condensing pressure oscillations for rolling motions with a bias of ±25%.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"182 ","pages":"Article 105655"},"PeriodicalIF":3.3,"publicationDate":"2025-02-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143403462","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":"Thermal-hydraulic performance of natural circulation system with narrow rectangular channel under heaving condition based on 1D/3D coupling analysis","authors":"Qiang Lian ,&nbsp;Luteng Zhang ,&nbsp;Simiao Tang ,&nbsp;Longxiang Zhu ,&nbsp;Bin Liu ,&nbsp;Zaiyong Ma ,&nbsp;Wan Sun ,&nbsp;Liangming Pan","doi":"10.1016/j.pnucene.2025.105663","DOIUrl":"10.1016/j.pnucene.2025.105663","url":null,"abstract":"<div><div>As the demand for ocean exploration grows, nuclear reactors utilizing natural circulation for power generation have become a significant propulsion force for long-term operations. However, the coolant within the reactor system is influenced by ocean conditions, characterized by passive flow fluctuations. In this study, a one-dimensional (1D)/three-dimensional (3D) coupling analysis method is developed and applied to investigate the impact of heaving conditions on the flow and heat transfer performance of natural circulation with a narrow rectangular channel. The flow and heat transfer characteristics within the narrow rectangular channel, which represents the reactor core, are analyzed in detail using a three-dimensional approach, while the overall natural circulation performance of the system is assessed using a one-dimensional method. Validation is conducted based on pulsating flow experiments and natural circulation experiments. Subsequently, detailed thermal-hydraulic parameters for both the rectangular channel and the natural circulation system are obtained. The results indicate that flow fluctuations caused by the heaving conditions significantly influence the transient performance of both the rectangular channel and the entire natural circulation system. However, the time-averaged flow resistance and heat transfer capability remain unaffected by the heaving conditions. An increase in the heaving period results in a decrease in flow fluctuations within the natural circulation system, and the phase delay between heaving displacement and mass flow flux also diminishes. Conversely, as the heaving amplitude increases, flow fluctuations rise, but this has no effect on the phase delay. Due to the ring effect introduced by the heaving motion, the difference between the main flow zone and the wall zone dominates the transient thermal-hydraulic performance in the narrow rectangular channel. Additionally, the heaving motion does not significantly influence the cycle-averaged flow resistance or the cycle-averaged Nusselt number.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"182 ","pages":"Article 105663"},"PeriodicalIF":3.3,"publicationDate":"2025-02-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143403659","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":"Thermal-hydraulic analysis of sequential disposal scenarios in a spent nuclear fuel repository","authors":"Samuel Park ,&nbsp;Pilhyeon Ju ,&nbsp;Taehoon Park ,&nbsp;Richard I. Foster ,&nbsp;GyoSoon Kim ,&nbsp;Sungyeol Choi","doi":"10.1016/j.pnucene.2025.105662","DOIUrl":"10.1016/j.pnucene.2025.105662","url":null,"abstract":"<div><div>In order to address issues related to Spent Nuclear Fuel (SNF), nations using nuclear energy have considered disposal of SNF in a Deep Geological Repository (DGR) located 500m below ground as most promising option. The Korean government has considered the multi-barrier disposal concept developed by Finland and Sweden and has developed a Korean Reference disposal System (KRS). The main consideration of KRS is how effectively it can isolate SNF from ecosystem, including considerations for thermal, mechanical and material degradations such as illitization of bentonite and canister corrosion. Among these criteria, temperature plays a significant role in maintaining repository integrity, influencing both mechanical and material degradation behavior. Consequently, various trials, including analytical and numerical solutions, have been employed to estimate temperature evolutions. However, a significant temperature difference between the analytical and numerical solutions arises due to the simulation region. Specifically, the analytical solution considers a multi-canister geometry, while current numerical solutions only account for a single-canister due to computational cost. As a result, the single canister model does not accurately reflect heat flux from the adjacent canisters. In this study, a quantitative comparison between analytical and numerical solutions using multi-canister geometries was conducted, and the results showed that the analytical and numerical solutions exhibit similar results. Additionally, four different SNF disposal methods were investigated to identify the optimized disposal method in terms of peak temperature. According to the results, it is suggested that the plop-plop method is the optimized disposal method.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"181 ","pages":"Article 105662"},"PeriodicalIF":3.3,"publicationDate":"2025-02-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143388086","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}