Nuclear Engineering and Design最新文献

{"title":"Validation of the neutron cross section processing code MGGC3.0 via JOYO-70 reactor physics experiments","authors":"Teng Zhang,&nbsp;Xubo Ma,&nbsp;Xudong Ma,&nbsp;Zhulun Li,&nbsp;Fuxing Wang","doi":"10.1016/j.nucengdes.2025.114045","DOIUrl":"10.1016/j.nucengdes.2025.114045","url":null,"abstract":"<div><div>Fast neutron reactor is a critical design within the Generation IV nuclear reactor systems. In this study, a high-precision neutron cross-section processing code named MGGC3.0 was developed. It directly applies HFG (hyperfine group:∼400000) cross-section data for resonance calculations and utilizes problem-dependent HFG neutron energy spectrum for energy group merging to produce the UFG (ultrafine group:∼2000) cross-section to take into account the complicated resonance self-shielding effect between isotopes. The computation of UFG elastic scattering matrix is expedited through prefabricated scattering function method. For the production of few-group cross section, MGGC3.0 conduct critical buckling searches and employs a two-region approximation for fuel and non-fuel assemblies, respectively. This process calculates the neutron energy spectrum for energy group merging to obtain the few-group cross section. Initially, verification was conducted using three fuel assemblies: MOX, UO2, and U-TRU-Zr. This involved comparing the UFG macroscopic cross-sections produced by MGGC3.0 with those obtained from OpenMC calculations. Subsequently, the code underwent verification using a series of fast reactor benchmarks in ICSBEP. This entailed comparing the eigenvalues computed based on cross sections produced by MGGC3.0 with those calculated by RMC. Lastly, validation of the code was conducted using the JOYO MK-I series zero-power experimental setup. This involved comparing the calculated and experimental values of control rod worth, sodium void reactivity, and fuel replacement reactivity. The computational results of the verification and validation processes indicate that the neutron cross sections produced by the MGCC3.0 code exhibit high accuracy, thereby furnishing precise cross-sectional data for fast reactor.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114045"},"PeriodicalIF":1.9,"publicationDate":"2025-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143767984","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":"Thorium and fully ceramic microencapsulated TRISO fuel neutronics feasibility analysis in a gas cooled fast reactor: Enhancing transmutation of long-lived fission products","authors":"Shohanul Islam","doi":"10.1016/j.nucengdes.2025.114014","DOIUrl":"10.1016/j.nucengdes.2025.114014","url":null,"abstract":"<div><div>This study investigates the feasibility of using Fully Ceramic Microencapsulated (FCM) TRISO fuel and thorium fuel in gas-cooled fast reactor, focusing on enhancing the transmutation of long-lived fission products by performing neutronics analysis using the OpenMC Monte Carlo code. The implementation of FCM and modifications to the TRISO layer aim to decrease the moderation effect of the TRISO fuel and achieve a harder neutron spectrum. Four alternative FCM TRISO fuels were proposed by replacing the porous buffer, inner pyrolytic carbon, and outer pyrolytic carbon layers with SiC, ZrC, TiC, and Si₃N₄ in each case. For thorium fuel, two options were investigated-ThUC and ThPuC. The analysis of neutronics parameters revealed that all models achieved a harder neutron spectrum, with all FCM models displaying more harder neutron spectrum than others. This enhancement in neutron spectra and the robust safety of FCM came with a decrease in cycle length and a marginal increase in the power peaking factor due to a more non-uniform neutron flux. Nevertheless, the FCM models still achieved a satisfactory long core life and maintained power peaking factors within acceptable limits. In contrast, the thorium models, particularly ThUC, demonstrated a longer cycle length and an improved power peaking factor. To completely analyze the viability of all models a comprehensive reactivity parameters calculation was performed including reactivity swing, effective delayed neutron fraction, fuel temperature coefficient, power coefficient of reactivity, control rod worth, and shutdown margin. The findings revealed that all models achieved satisfactory results across all reactivity parameters. Notably, all FCM models exhibited improved power coefficient, control rod worth, and shutdown margin compared to the other models. This comprehensive neutronics analysis suggests that while all proposed models displayed satisfactory neutronics performance, the FCM models showed superior reactivity performance. Notably, the FCM model demonstrated significantly improved transmutation efficiency for four long-lived fission products: Nb-94, Pd-107, I-129, and Sm-151.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114014"},"PeriodicalIF":1.9,"publicationDate":"2025-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143767983","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 on the state of thermal hydraulics research on air ingress scenarios in High-Temperature Gas-cooled Reactors following a D-LOFC","authors":"Matthew Scheel ,&nbsp;Piyush Sabharwall ,&nbsp;Richard Schultz ,&nbsp;Daniele Ludovisi ,&nbsp;Gianluca Blois","doi":"10.1016/j.nucengdes.2025.113946","DOIUrl":"10.1016/j.nucengdes.2025.113946","url":null,"abstract":"<div><div>With the expectation of near-immediate carbon neutrality, widespread implementation of proven High-Temperature Gas-cooled Reactors (HTGRs) embodies a viable solution pathway given their inherent, passive safety features and high thermal efficiency. This study provides an overview of the current state of research involving the thermal hydraulics associated with air ingress from a depressurized loss of forced cooling (D-LOFC) in HTGRs. Accurately characterizing and predicting the physical phenomena underlying air ingress is of paramount concern, as the integrity of the fuel and core graphite support structures are threatened by the presence of oxygen. Broadly speaking, the air ingress scenario can be delineated into three main stages: (1) Depressurization, (2) Density-Driven Flow, and (3) Natural Convection. In tandem with the underlying fundamental theory, this review collates and synthesizes the existing body of contemporary research concerning the air ingress scenario following a D-LOFC. As evinced by this review, our current understanding and predictive abilities have benefited from extensive research, predominantly concentrated on the rate of air ingestion into the core. Additional research is necessary to holistically capture the phenomenology of an air ingress scenario following a D-LOFC by considering an additional variable: the oxygen content of the ingressing air. The latter variable requires investigation into the complex interactions of the fully integrated system. Additionally, while numerical tools are evolving domestically through the Nuclear Energy Advanced Modeling and Simulation program, a sufficiently validated code remains absent.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 113946"},"PeriodicalIF":1.9,"publicationDate":"2025-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143767987","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 investigation on quenching behavior across spacer grid during reflooding transient","authors":"Long Ji ,&nbsp;Xiaojing Liu ,&nbsp;Wei Xu ,&nbsp;Wei Zeng ,&nbsp;Jie Wang ,&nbsp;Hui He","doi":"10.1016/j.nucengdes.2025.114042","DOIUrl":"10.1016/j.nucengdes.2025.114042","url":null,"abstract":"<div><div>During a Loss Of Coolant Accident (LOCA), the reflooding transient involves complex two-phase flow heat transfer process betweeen dispersed liquid phase, continuous vapor phase and high-temperature wall. The presence of spacer grids along the entire rod bundle has significant effects on the reflooding heat transfer phenomena during reflooding transients by interacting with entrained droplets. Quenching behavior in different regions of the spacer grid during the reflooding transient is studied experimentally using the 2 × 2 rod bundle test facility. The axial and circumferential quenching behaviors of the heater rods upstream and downstream of the spacer grid are analyzed for different reflooding velocities and linear power densities. Experimental results show that earlier occurrence of quenching downstream of the grid spacer is observed under low reflooding velocities and high linear power densities due to the mechanism of droplet breakup by dry spacer grid. On the other hand, at high reflooding velocity and low linear power density, a wetted grid results in increased downstream droplet size and decreased heat transfer performance, causing the linear change of the quench front curve. The experimental results also indicate that the circumferential quenching process of the heater rod upstream and downstream of the spacer grid is inconsistent due to the influence of the inhomogeneous flow pattern and spacer grid wetting conditions. The experimental data is used to support the development and validation of the model of film boiling heat transfer coefficient considering droplet breakup. The comparison results show that the boron equilibrium model can predict the performance of the boron concentration in the rod bundle channel with an accuracy of 5 %.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114042"},"PeriodicalIF":1.9,"publicationDate":"2025-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143767986","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":"Optimization research of heat transfer coefficient prediction model for supercritical water based on Bayesian search algorithm","authors":"Ma Dongliang ,&nbsp;Zhou Tao ,&nbsp;Huang Yanping","doi":"10.1016/j.nucengdes.2025.114036","DOIUrl":"10.1016/j.nucengdes.2025.114036","url":null,"abstract":"<div><div>In order to make better use of machine learning algorithm to perform thermo-hydraulic analysis of supercritical water reactor, different intelligent algorithms are used to optimise the model parameters for predicting the heat transfer coefficient of supercritical water. The accuracy changes of stochastic search and Bayesian search algorithms in predicting the heat transfer coefficient under different parameter spaces are compared and analysed. The results show that the search space and the initial distribution assumption have a large impact on the results. The Bayesian search algorithm is relatively less affected by the search space and parameter distribution assumptions. The prediction accuracy obtained by Bayesian search is 1.25–2.88% higher than that obtained by random search. After optimising the model parameters, the average test accuracy of predicting the heat transfer coefficient of supercritical water is more than 96.4%. At the same time, the spatial distribution characteristics of the optimal parameter points obtained by different search algorithms are analysed.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114036"},"PeriodicalIF":1.9,"publicationDate":"2025-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143767985","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":"Characteristics of oxygen mass transfer in a pneumatic mass exchanger for solid-phase oxygen control in the lead bismuth","authors":"Zhen Yang,&nbsp;Haicai Lyu,&nbsp;Honglin Zhou,&nbsp;Fang Liu,&nbsp;Weihao Xing,&nbsp;Shengfei Wang,&nbsp;Wentao Guo,&nbsp;Zhangpeng Guo,&nbsp;Ruixian Liang,&nbsp;Fenglei Niu","doi":"10.1016/j.nucengdes.2025.114022","DOIUrl":"10.1016/j.nucengdes.2025.114022","url":null,"abstract":"<div><div>Oxygen concentration in lead–bismuth alloy (LBE) systems has a significant effect on the corrosion rate of structural materials. The corrosion of structural materials by LBE can be effectively mitigated by dynamically adjusting the dissolved oxygen concentration in liquid lead–bismuth. Solid-phase oxygen control technology comprised of a packed bed of lead oxide spheres is widely recognized as an effect and promising solution to regulate the dissolved oxygen concentration, in which how to quantitatively regulate the supplemental solid-phase PbO oxygen source is the key to solid-phase oxygen control technology. This paper innovatively designs a pneumatic mass exchanger for controllable oxygenation without moving parts, and experimentally investigates the oxygen mass transfer characteristics under different temperatures, relative flow velocities, and surface areas of oxygen source areas. By combining an empirical oxygen mass transfer model and the oxygen mass conservation equation, a theoretical prediction model for the reciprocating pneumatic mass exchanger is developed. The results indicate that temperature can rapidly adjust the oxygen dissolution rate, while the relative flow velocity can be used as an effective measure to precisely control the oxygen concentration. The oxygen dissolution rate is directly proportional to the oxygen source surface area. The average relative error of the oxygen concentration between the theoretical prediction model and experimental results is 1.65, with the deviation primarily attributed to discrepancies in oxygen diffusion and the fitting of the mass transfer empirical model. The oxygen concentration in lead–bismuth can be controlled within a reasonable range, thereby validating the oxygenation performance of the novel solid-state oxygen control device—the pneumatic mass exchanger.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114022"},"PeriodicalIF":1.9,"publicationDate":"2025-04-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143759373","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":"Simplified models for high burnup spent nuclear fuel rods and their comparison","authors":"Seyeon Kim,&nbsp;Sanghoon Lee","doi":"10.1016/j.nucengdes.2025.114032","DOIUrl":"10.1016/j.nucengdes.2025.114032","url":null,"abstract":"<div><div>Effective management of spent nuclear fuel requires maintaining its structural integrity with safety guidelines emphasizing protection of the cladding from mechanical and physical damage that could lead to significant fuel rod failure. According to NUREG-1864, the plastic strain that can cause cladding failure was observed to be in the range of 1.0 ∼ 4.0 %. The experimental data was obtained by pressurized tube tests with defueled cladding specimens and could not reflect the complicated stress states due to the pellet-clad interaction that occurs in real drop impacts. However, the failure criterion of irradiated cladding under more complicated stress states is not easily available. In this study, we perform a comparative analysis of the applicability of simplified models for an SNF rod based on two new fracture criteria and flexural rigidity modification proposed in NUREG-2224. The two new failure criteria utilize curvature based plastic bending strain and membrane plus bending stress through the thickness of the cladding, respectively. Of the three models, the simplified model based on membrane plus bending stress failure criterion was found to be the most conservative in dynamic impact applications.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114032"},"PeriodicalIF":1.9,"publicationDate":"2025-04-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143759446","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":"Investigation of stress corrosion crack propagation characteristics and life prediction for thick-walled double U-groove pipe welds in PWR steam generator","authors":"Baoyin Zhu ,&nbsp;Zheng He ,&nbsp;Lu Zhang ,&nbsp;Shuitao Gu ,&nbsp;Xiao Jin ,&nbsp;Dungui Zuo ,&nbsp;Gongye Zhang","doi":"10.1016/j.nucengdes.2025.114019","DOIUrl":"10.1016/j.nucengdes.2025.114019","url":null,"abstract":"<div><div>Stress corrosion cracking (SCC) is a critical concern in evaluating the structural integrity of pressurized water reactor (PWR) primary pressure boundaries. Material mismatch and welding-induced residual stresses introduce significant challenges in predicting crack propagation paths and service-induced defect lifetimes. This study examines the influence of welding residual stress on SCC in a thick-walled double U-groove pipe dissimilar steel welded joint of a PWR steam generator (SG), and assesses the propagation life of SCC from an engineering standpoint. First, the distribution and stress state of residual stresses in complex SG dissimilar steel joints were analyzed using the two-way thermal coupling finite element method to establish initial stress boundary conditions for simulating SCC propagation influenced by residual stresses. Next, the extended finite element method combined with the maximum principal stress criterion was employed to investigate the propagation direction and path of cracks originating from various initial positions due to welding residual stresses. Concurrently, the J-integral method was used to calculate the stress intensity factors of cracks at different depths along the propagation path. Finally, based on the modified Shoji model, the relationship between the stress intensity factor and SCC propagation rate was examined, allowing for predictions of crack propagation rates and service life for SCC with varying initial defects.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114019"},"PeriodicalIF":1.9,"publicationDate":"2025-04-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143759445","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":"Functionally adaptable isolated structures for small modular reactors based on design demands","authors":"Kui Yang ,&nbsp;Ping Tan ,&nbsp;Jiying Shang ,&nbsp;Yafei Zhang ,&nbsp;Jiaxi Li","doi":"10.1016/j.nucengdes.2025.114037","DOIUrl":"10.1016/j.nucengdes.2025.114037","url":null,"abstract":"<div><div>Nuclear power plants are gradually trending toward miniaturization and standardization owing to their expensive costs, lengthy design cycles, challenging siting, and hazardous issues. However, this development presents challenges such as reduced site adaptability and weakened seismic capacity. In response to the above challenges, this paper proposes a multifunctional combination-rail tension friction slip isolator (MC-RTFSI). Its application enables small modular reactors (SMRs) to satisfy the requirements of different sites and seismic intensities without the necessity of modifying the design. MC-RTFSI is a composite system comprising different types of sliding rails that exhibit distinct mechanical properties. The selection of suitable rails is based on the specific characteristics of different sites. Furthermore, it exhibits performance decoupling in both orthogonal directions and uplift-restraining properties. Various combinations of MC-RTFSI were mechanically investigated, leading to the construction of a mechanical hysteresis model. This model was subsequently simulated by employing series–parallel combinations of nonlinear units in current finite element analysis software (ETABS, SAP 2000). Based on the result of the refined finite element simulation, the effectiveness of the simulated units is confirmed. A number of nonlinear response history analyses were conducted. The results indicate that novel isolated structures for small-scale nuclear power plants possess the capacity to perform various activities and may be adapted to accommodate different sites and ground motion characteristics.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114037"},"PeriodicalIF":1.9,"publicationDate":"2025-04-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143748236","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":"Efficient adsorption of Cs(I) and Sr(II) ions from solution by xanthate modified sponge-like chitosan","authors":"Bowen Xu ,&nbsp;Tian Huiyu ,&nbsp;Chen Jianbo ,&nbsp;Wei Lifeng ,&nbsp;Wang Kai ,&nbsp;Jianlong Wang","doi":"10.1016/j.nucengdes.2025.114031","DOIUrl":"10.1016/j.nucengdes.2025.114031","url":null,"abstract":"<div><div>Efficient removal of radionuclides from radioactive wastewater is vital for ensuring the sustainable development of nuclear energy. Herein, xanthate-modified sponge-like chitosan-based adsorbent (CTS-SX) was synthesized through a facile method, which exhibited excellent adsorption performance for Sr(II) and Cs(I). The maximum adsorption capacity of Sr(II) and Cs(I) reached 76.21 and 133.15 mg·g⁻¹ respectively. Various models were used to fit the adsorption process, and the results indicated that the adsorption process of Sr(II) was best fitted by Freundlich isotherm model and phenomenological internal mass transfer kinetic model, while the adsorption of Cs(I) was best fitted by Langmuir isotherm model and Langmuir kinetics model. In addition, the adsorption performance in binary-component system, influence of pH and competing ions were also explored. In summary, the modification could improve the amount of active sites and hydrophobility of CTS-SX. With excellent adsorption performance and simplicity of separation, CTS-SX has potential in application for radioactive wastewater treatment.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114031"},"PeriodicalIF":1.9,"publicationDate":"2025-04-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143748235","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}