Nuclear Engineering and Design最新文献

{"title":"Financial feasibility of optimizing the GA Siwabessy reactor-utilization for enhanced radioisotope production","authors":"Mudjiono ,&nbsp;Nuryanti ,&nbsp;Sudi Ariyanto ,&nbsp;Kurnia Anzhar ,&nbsp;Pini Wijayanti ,&nbsp;Yus Rusdian Akhmad ,&nbsp;Amil Mardha ,&nbsp;Djati Hoesen Salimy ,&nbsp;Jupiter Sitorus Pane ,&nbsp;Rizki Firmansyah Setya Budi ,&nbsp;Elok Satiti Amitayani ,&nbsp;Donny Nurmayady ,&nbsp;Nurlaila ,&nbsp;Ewitha Nurulhuda ,&nbsp;Nur Hasanah ,&nbsp;Imam Bastori ,&nbsp;Anis Rohanda ,&nbsp;Sufiana Solihat","doi":"10.1016/j.nucengdes.2025.114011","DOIUrl":"10.1016/j.nucengdes.2025.114011","url":null,"abstract":"<div><div>The GA Siwabessy multipurpose reactor (RSG-GAS) is a strategic facility in Indonesia, that supports various research applications and medical treatments by producing radioisotopes and radiopharmaceuticals. As the demand for critical isotopes such as I-131, Mo-99, and Ir-192 continues to grow, there is a pressing need to enhance the reactor’s capabilities to meet these requirements efficiently. This research focuses on optimizing the reactor’s utilization by prioritizing its role in isotope production and addressing operational and strategic challenges. A systematic cost-benefit analysis framework is developed to evaluate the financial feasibility of this optimization. The framework incorporates comprehensive cost components, including revitalization investments and operational expenses, while benefits are assessed in terms of increased isotope production, and economic returns. Four reactor’s irradiation channel utilization are analyzed at 15 MW and 20 MW, spanning single-product and multi-product scenarios. Results indicate that the current operational condition is financially unsustainable, while the proposed optimization scenario offer significant economic benefit. Among the configurations, dedicated I-131 production (Option 1) yields the highest financial returns, while the multi-product scenario (Option 4) balances profitability with diverse national needs. This research underscores the strategic importance of reactor revitalization in achieving Indonesia’s long-term independence in nuclear medicine and industrial applications.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114011"},"PeriodicalIF":1.9,"publicationDate":"2025-03-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143725210","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 the rewetting velocity on dry out surface due to stepwise boundary condition changes","authors":"A. Satou,&nbsp;Y. Wada,&nbsp;Y. Sibamoto","doi":"10.1016/j.nucengdes.2025.114020","DOIUrl":"10.1016/j.nucengdes.2025.114020","url":null,"abstract":"<div><div>Post-boiling transition (post-BT) heat transfer is essential for analyzing the duration of surface dryout and peak cladding temperature during abnormal transients and accidents in light water reactors. The rewetting phenomenon, in which the liquid film front propagates to the dryout surface and becomes wet again due to a decrease in heat flux or an increase in flow rate, is very important for evaluating the dryout duration. Although rewetting velocity models have been proposed mainly for reflooding processes, mechanical rewetting velocity models applicable to much higher flow and heat flux conditions than reflooding processes have been developed in recent years. However, due to the lack of an experimental database on rewetting velocities under such a wide range of thermal–hydraulic conditions, sufficient data for model development and validation do not exist. Therefore, a database on rewetting velocities caused by stepwise boundary condition changes under a wide range and multiple combination of thermal–hydraulic conditions was obtained using a single-tube experimental apparatus. Based on this database and the characteristics of rewetting velocities obtained, an empirical correlation for rewetting velocity was proposed. This correlation predicts the rewetting velocity accurately by taking the change in the mass flux of the liquid or gas phase with stepwise transients as a parameter. This suggested that the change in the mass flux of the gas or liquid phase near the liquid film front has a strong influence on the rewetting under extremely high mass flux conditions compared to the reflooding process.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114020"},"PeriodicalIF":1.9,"publicationDate":"2025-03-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143725208","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":"Enhancing the strength and toughness of high-Cr high-Si F/M steel for nuclear power applications by optimizing hot rolling deformation amount","authors":"Yingqi Liu ,&nbsp;Qianfu Pan ,&nbsp;Runkun Shi ,&nbsp;Xiaochang Xu","doi":"10.1016/j.nucengdes.2025.114006","DOIUrl":"10.1016/j.nucengdes.2025.114006","url":null,"abstract":"<div><div>High-Cr high-Si ferritic/martensitic (F/M) steel has emerged as a key candidate material for nuclear power equipment due to its excellent radiation resistance and corrosion resistance. However, the high Cr and Si contents lead to a greater proportion of ferrite, which can reduce the material’s strength and toughness and affect the type and distribution of second phase particles, thereby limiting its application in nuclear industry environments. This study optimizes the microstructure of High-Cr high-Si F/M steel by adjusting the hot rolling deformation amount, achieving the best match of strength and toughness. The mechanisms of strength enhancement are also investigated to clarify the relationship between microstructure and performance. The results show that a deformation amount of 40% during hot rolling yields the best overall performance, characterized by a low ferrite content, fine MX-type carbides with a high number density, and excellent strength and toughness. Rolling at the austenitization temperatures significantly improves the material’s comprehensive properties, providing strong support for the application of high-Cr high-Si F/M steel in the nuclear power sector.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114006"},"PeriodicalIF":1.9,"publicationDate":"2025-03-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143725207","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":"Practical approach to Hazard-Consistent fragility curve estimates using Bayesian updating","authors":"Nevena Šipčić ,&nbsp;Pablo García de Quevedo Iñarritu ,&nbsp;Mohsen Kohrangi ,&nbsp;Dimitrios Vamvatsikos ,&nbsp;Paolo Bazzurro","doi":"10.1016/j.nucengdes.2025.114029","DOIUrl":"10.1016/j.nucengdes.2025.114029","url":null,"abstract":"<div><div>Seismic fragility curves provide the probability of exceedance of a given damage state, should different levels of ground motion intensity be experienced at the site where the structure, or component, is located. Such curves are often derived via multiple nonlinear response history analyses (NLRHA) using sets of “suitable” ground motions that, in line with the best practice, should be consistent with the seismic hazard at the site. Based on the selected sets of records, one can estimate fragility functions that are often assumed to follow a lognormal distribution defined by two parameters, i.e., the logarithmic mean (µ) and the logarithmic standard deviation (β). Our focus is on estimating them using a state-of-the-art approach that involves hazard-consistent record selection via Conditional Spectrum and multiple stripe analysis. However, this approach usually requires many NLRHAs, with high computational costs, especially for the complex structural models typical of the nuclear industry. This study investigates the optimal number of ground motions and intensity levels required to keep the computational burden acceptable without compromising accuracy. To do so, we adopt a Bayesian framework with Markov chain Monte Carlo simulation and Metropolis–Hasting sampling. Our findings show that this approach effectively helps analysts best allocate computational resources while ensuring acceptable accuracy in estimating the probability of reaching or exceeding the considered damage states.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114029"},"PeriodicalIF":1.9,"publicationDate":"2025-03-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143715479","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":"keff behaviour of a potential PWR assembly loaded disposal canister in simplified corrosion scenarios","authors":"M. Frankl ,&nbsp;L. Berry ,&nbsp;A. Vasiliev ,&nbsp;D. Rochman ,&nbsp;H. Ferroukhi ,&nbsp;N. Diomidis ,&nbsp;M. Wittel","doi":"10.1016/j.nucengdes.2025.114007","DOIUrl":"10.1016/j.nucengdes.2025.114007","url":null,"abstract":"<div><div>The Swiss National Cooperative for the Disposal of Radioactive Waste (Nagra) and the Center for Nuclear Engineering and Sciences at the Paul Scherrer Institute (PSI) are investigating how potential long-term changes to the geometry and material composition of the current final disposal canister (BE-ELB) design may affect the neutron multiplication factor (<em>k_eff</em>) of a canister loaded with PWR spent nuclear fuel assemblies (FAs). Several conservative corrosion scenarios were formulated, modeled, simulated and analyzed using the Monte-Carlo codes MCNP6.2® and Serpent2.2. The corrosion-induced effects, such as the replacement of moderator by magnetite, differ significantly from typical fuel lattice configurations used in reactor or storage pool safety analyses. This paper therefore focuses on underlying physical phenomena causing the observed <em>k_eff</em> changes, including analyses of neutron currents and spectra, the ‘6-factor formula’, and the sensitivity of <em>k_eff</em> to specific regions, materials, and nuclides. These analyses showed the canister wall and the corrosion product magnetite to act as a heavy reflector, enhancing the backscattering of neutrons in the epithermal and fast energy ranges. Furthermore, the critical role of the water distribution in all the potential scenarios was revealed. Water inside the FAs clearly increases reactivity by moderation, water outside the remainders of the steel basket, however, has an inhibiting effect on neutron multiplication. All simulation results heavily depend on the specific preliminary ELB design. To that end, the results of this study can help to optimize the ELB design and the Swiss concept for the final disposal of high-level radioactive waste.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114007"},"PeriodicalIF":1.9,"publicationDate":"2025-03-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143706421","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":"Coupled analysis system development on heat pipe reactor","authors":"Sung Nam Lee,&nbsp;Sung Hoon Choi,&nbsp;Nam-il Tak,&nbsp;Hong-sik Lim,&nbsp;Chan Soo Kim","doi":"10.1016/j.nucengdes.2025.114003","DOIUrl":"10.1016/j.nucengdes.2025.114003","url":null,"abstract":"<div><div>This study outlines the design tool and coupled analysis for heat transport in a space heat pipe reactor to provide power in a space environment for 10 years without fuel replacement. Korea Atomic Energy Research Institute (KAERI) has developed the design analysis tools to investigate the temperature distribution and maximum temperature in the reactor core. The neutronics code, McCARD, provides the power profile of the fuel compact in the core. The heat transport code, HEPITOS, predicts the temperature profiles coupled with the heat pipe analysis code, LUHPIS.</div><div>Since each code has different physics, the numerical calculation can be performed by the explicit method or the coupled method. Most studies have been analyzed using the coupled calculations of HEPITOS and LUHPIS. The coupled analysis with the neutronic code has been done to find out the feasibility of the coupled system development. The calculated results will be a reference for the evaluation of the thermal margins of the components.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114003"},"PeriodicalIF":1.9,"publicationDate":"2025-03-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143706420","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":"CHF enhancement in downward-facing boiling surface using shrouds for calandria vessel during severe accident in PHWRs","authors":"P.K. Verma ,&nbsp;P.P. Kulkarni ,&nbsp;A.K. Nayak","doi":"10.1016/j.nucengdes.2025.114010","DOIUrl":"10.1016/j.nucengdes.2025.114010","url":null,"abstract":"<div><div>The orientation of the heated surface significantly affects the boiling process. Boiling on a downward-facing surface is particularly challenging because bubble detachment is hindered, leading to longer bubble residence times and unique interactions than on vertical or inclined surfaces. This study investigates boiling on a large downward-facing flat surface (100 × 400 mm), focusing on critical heat flux (CHF) phenomenon. During the postulated severe accident, the situation arises in Pressurised Heavy Water Reactors (PHWRs) due to multiple failures of cooling systems and safety systems. The pressure tubes and calandria tubes have the potential to break, resulting in hot debris that falls to the bottom of the calandria vessel. The calandria vessel has a large curvature due to its larger diameter, and the bottommost portion is like a flat plate. To contain the hot debris or molten corium inside the vessel and maintain the integrity of the calandria vessel at a higher temperature is crucial to arrest the progress of a severe accident. The cooling of the vessel from outside without occurring CHF at the bottom location is important. Historically, downward-facing boiling has received limited attention, as it is normally not used in industrial applications owing to lower heat transfer and CHF values due to adverse buoyancy. Nonetheless, it is important to investigate because of the severe accident situation in PHWRs. Incorporating a simple technique of shrouds surrounding the calandria vessel can enhance the CHF by enhancing the buoyancy. This paper investigates the potential enhancement of CHF through the use of shrouds.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114010"},"PeriodicalIF":1.9,"publicationDate":"2025-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143706417","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":"Impact of γ irradiation and thermal aging on the swelling pressure of GMZ bentonite","authors":"Wei Liu ,&nbsp;Dong Liang ,&nbsp;Zhongtian Yang ,&nbsp;Chao Gao ,&nbsp;Jingli Xie ,&nbsp;Meilan Jia","doi":"10.1016/j.nucengdes.2025.114012","DOIUrl":"10.1016/j.nucengdes.2025.114012","url":null,"abstract":"<div><div>To investigate the combined effects of γ irradiation and thermal aging on the swelling pressure of buffer materials used in the geological disposal of high-level radioactive waste, Gaomiaozi (GMZ) bentonite from Inner Mongolia was herein selected. Samples were irradiated by γ rays and exposed to a combination of γ irradiation followed by thermal aging at 90℃. Swelling pressure tests were then conducted. Results showed that irradiation and thermal aging reduced the swelling pressure of GMZ bentonite. Notably, after irradiating and storing at (25 ± 3)℃ for 360 d, the swelling pressure increased by 51.59 % compared with that of the γ-irradiated sample; the swelling pressure decreased by 75.9 % for the γ irradiation–thermal sequentially aged sample (heated for 180 d). Partial chemical-bond destruction, the transformation of structure trivalent to divalent iron, and the water redistribution in the bentonite were the primary drivers of the decrease in the swelling pressure of GMZ bentonite.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114012"},"PeriodicalIF":1.9,"publicationDate":"2025-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143706418","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":"Assessment and predicting the axial power distribution effect on the thermal-mechanical parameters of the NuScale nuclear reactor core loaded with TVS-2 M fuel assemblies as well as axial Offset optimizing for load-following operation","authors":"M.H. Zahedi yeganeh,&nbsp;G.R. Ansarifar,&nbsp;H.Zayermohammadi Rishehri","doi":"10.1016/j.nucengdes.2025.114009","DOIUrl":"10.1016/j.nucengdes.2025.114009","url":null,"abstract":"<div><div>This study evaluates and examines the thermal–mechanical behavior of a NuScale reactor core which utilizes TVS-2 M hexagonal fuel assemblies. The efficiency of the fuel rods is validated using the FRAPCON code. Initially, the reactor’s core is modeled with the MCNP code to locate the control banks. The design phase ensures the capability to shut down the reactor in two scenarios. In the Hot Zero Power (HZP) scenario, MCNP simulation reveals a sub-critical state with a multiplication factor of 0.94481 ± 0.00023. In the Cold Zero Power (CZP) scenario, the multiplication factor of 0.9935 ± 0.00023 confirms the adequacy of control assemblies. Subsequently, a thermal–mechanical analysis is conducted on the fuel rod over 1330 days, confirming its acceptable design and operational effectiveness in the core. Also, one of the parameters that can be examined during reactor control and load-following operations is Axial Offset (AO). Therefore, the study investigates the impact of AO on fuel rod’s thermal–mechanical changes. The MCNP code was used to simulate control rod inputs and obtain power distribution data for each AO deviation. Based on assessments regarding the association between AO and the thermal–mechanical characteristics of fuel, it has been determined that the impact of power distribution increases significantly over time, particularly towards the end of the operational period. Afterward, based on FRAPCON results, an artificial neural network (ANN) estimator is developed to predict thermal–mechanical parameters at the beginning of the cycle (BOC). The ANN proves to be a powerful method for estimation. By employing the ANN estimator and exploring different cost functions based on thermal–mechanical parameters, the optimal AO is determined using a genetic algorithm, which enhances the reactor’s performance, particularly in load-following operations. The attained optimal AO value for various cost functions are as follows: −0.10316, −0.19635, and −0.25817. This approach allows for the selection of the most efficient AO, leading to improved performance of the NuScale reactor core loaded with TVS-2 M hexagonal fuel assemblies. Indeed, optimization of AO is very important and useful for load-following operation.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114009"},"PeriodicalIF":1.9,"publicationDate":"2025-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143706419","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":"Analyses of the MELCOR capability to simulate integral PWR using passive systems in a DBA scenario","authors":"M. Principato ,&nbsp;F. Giannetti ,&nbsp;M. Imperatori ,&nbsp;M. D’Onorio ,&nbsp;M. Garcia ,&nbsp;L.E. Herranz ,&nbsp;A. Bersano ,&nbsp;F. Mascari","doi":"10.1016/j.nucengdes.2025.114004","DOIUrl":"10.1016/j.nucengdes.2025.114004","url":null,"abstract":"<div><div>This paper is focused on a thermal–hydraulic transient analysis of a generic 300 MWe integral Pressurized Water Reactor (iPWR) based on a passive safety mitigation strategy and dry containment. The main target of the paper is to analyze the MELCOR code’s capability to simulate the operation of the passive systems in Small Modular Reactor (SMR) configuration and the consequent plant behavior. The nodalization and the modeling approach used in the MELCOR code have been described in the present work, and a double-ended rupture of the Direct Vessel Injection (DVI) line has been postulated as initiating event.</div><div>The results of the analysis demonstrated that the MELCOR code can qualitatively replicate the dominant phenomena that drive the passive mitigation strategy’s operation. Additionally, the core reflooding was made possible by the operation of the available safety systems, which was adequate to prevent severe accident conditions during the entire simulated transient. The activity has been developed in the framework of SASPAM-SA Horizon Euratom project as the base for further derivative activities focused on assessing the capability of the MELCOR code to simulate the phenomena taking place in postulated plausible severe accident scenarios in iPWR.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114004"},"PeriodicalIF":1.9,"publicationDate":"2025-03-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143682450","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}