Annals of Nuclear Energy最新文献

{"title":"Doppler effect evaluation of the TRU-based Moroccan TRIGA Mark II reactor: Applying novel methodologies","authors":"Mohamed Mira ,&nbsp;O. El Hajjaji ,&nbsp;Jordi Freixa ,&nbsp;T. El Bardouni ,&nbsp;H. Boukhal ,&nbsp;Y. Boulaich ,&nbsp;Tarik El Ghalbzouri","doi":"10.1016/j.anucene.2025.111353","DOIUrl":"10.1016/j.anucene.2025.111353","url":null,"abstract":"<div><div>The Doppler effect, commonly referred to as fuel temperature coefficient (<span><math><msub><mrow><mi>α</mi></mrow><mrow><mi>T</mi></mrow></msub></math></span>), is a critical parameter that quantifies the reactivity feedback resulting from changes in fuel temperature. Understanding and accurately quantifying <span><math><msub><mrow><mi>α</mi></mrow><mrow><mi>T</mi></mrow></msub></math></span> is essential for evaluating and adopting innovative fuel compositions. In this study, we present a comprehensive evaluation of <span><math><msub><mrow><mi>α</mi></mrow><mrow><mi>T</mi></mrow></msub></math></span> for transuranic (TRU)-based Moroccan TRIGA Mark II reactor, employing advanced simulation techniques and methodologies based on the OpenMC code. Two novel methods are introduced and validated. The first method analyzes the contributions of the fast fission factor (<span><math><mi>ϵ</mi></math></span>), resonance escape probability factor (<span><math><mi>p</mi></math></span>), thermal utilization factor (<span><math><mi>f</mi></math></span>), and reproduction factor (<span><math><mi>η</mi></math></span>) to <span><math><msub><mrow><mi>α</mi></mrow><mrow><mi>T</mi></mrow></msub></math></span>. The second evaluates the Upscattering and isotope-wise contributions to gain deeper insights into the behavior of these factors. Additionally, <span><math><msub><mrow><mi>α</mi></mrow><mrow><mi>T</mi></mrow></msub></math></span> is evaluated as a function of fuel burnup. For comparative purposes, the standard U-ZrH_1.65 fuel is also analyzed to benchmark the behavior of the TRU-based fuel against conventional fuel types.</div><div>The results reveal that TRU-based fuel, despite its high fissile content, it maintains a negative <span><math><msub><mrow><mi>α</mi></mrow><mrow><mi>T</mi></mrow></msub></math></span> across the operational temperature range and throughout the reactor’s lifetime. This negative reactivity feedback is attributed to the significant thermal capture resonances of TRU-based isotopes, such ²⁴⁰Pu, and ²⁴²Pu, and the unique properties of the zirconium hydride matrix ZrH_1.65. Although the magnitude of <span><math><msub><mrow><mi>α</mi></mrow><mrow><mi>T</mi></mrow></msub></math></span> for TRU-based fuel is less negative compared to the standard U-ZrH_1.65 fuel, it remains sufficiently negative to guarantee safe reactor operation.</div><div>These findings are promising and suggest that further exploration of TRU-based fuels could support the development of reliable, clean, and sustainable nuclear energy systems.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111353"},"PeriodicalIF":1.9,"publicationDate":"2025-03-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143643707","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 newly and efficient virtual mesh based CMFD acceleration method for pebble-bed HTGR with cylindrical geometry","authors":"Yuchen Wen,&nbsp;Chen Hao,&nbsp;Yizhen Wang","doi":"10.1016/j.anucene.2025.111349","DOIUrl":"10.1016/j.anucene.2025.111349","url":null,"abstract":"<div><div>Three-dimensional method of characteristic (3D-MOC) codes have been developed to conduct neutronic simulation for pebble-bed high temperature gas-cooled rector (HTGR). Although parallel strategy such as domain decomposition and acceleration method such as coarse mesh finite difference (CMFD) are applied in these codes, whole core simulation still faces huge computational burden which needs more efficient acceleration method to enhance simulation efficiency. Moreover, considering the cylindrical geometry of pebble-bed HTGR, differential length of coarse meshes will gradually increase along radial direction which will lead convergence problem in conventional CMFD method. In this paper, a virtual mesh based CMFD method is proposed to overcome this issue. Virtual coarse meshes are generated by further dividing CMFD coarse meshes along circumferential direction to reduce the optical thickness of peripheral coarse meshes. The performance of proposed virtual coarse mesh based CMFD acceleration method is verified in a simplified cylindrical pebble-bed model and HTR-10 whole core problem. Numerical results show that virtual coarse mesh based CMFD method can reduce around 30% source iterations of MOC compared with original CMFD method.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111349"},"PeriodicalIF":1.9,"publicationDate":"2025-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143632043","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":"Object-oriented petri nets for reliability assessment and sensitivity analysis of safety-critical nuclear power plant systems","authors":"Kabiratna Behera ,&nbsp;Vinay Kumar ,&nbsp;Binod Kumar Singh","doi":"10.1016/j.anucene.2025.111364","DOIUrl":"10.1016/j.anucene.2025.111364","url":null,"abstract":"<div><div>A Safety-critical system (SCS) is a complex system consisting of various components functioning cooperatively that demands rigorous analysis to ensure reliability and effectiveness in mitigating potential hazards. In several previous studies, the ordinary Petri net (PN) models are used to model and analyze the reliability of SCSs. However, these models lack modularity and flexibility and have state explosion problems. On the other hand, the application of Object-Oriented Petri Nets (OOPNs) in the field of dependability analysis of SCSs is limited. Thus, we propose a framework using OOPNs to model the complex SCSs and perform reliability and sensitivity analysis of such systems. Unlike ordinary PN models, the OOPN models offer better modularity and flexibility and also reduce the state explosion problems. We apply our framework to the Reactor Core Isolation Cooling (RCIC) system in Nuclear Power Plants (NPPs) to demonstrate its effectiveness. We have validated our framework and achieved an accuracy of 99.97% in reliability prediction, which proves its efficacy.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111364"},"PeriodicalIF":1.9,"publicationDate":"2025-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143642125","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":"Toward environmental sustainability: What role does Russian nuclear technology play?","authors":"Shujaat Abbas","doi":"10.1016/j.anucene.2025.111368","DOIUrl":"10.1016/j.anucene.2025.111368","url":null,"abstract":"<div><div>Nuclear energy technology has enormous untapped potential to generate stable and clean energy required to achieve net-zero carbon emissions. This study explores the impact of Russian nuclear technology exports on environmental sustainability in 21 major nuclear power-consuming countries from 2002 to 2021. This objective is realized by using panel inverse quantile and instrumental variable quantile with smoothed generalized methods of moments technique. The estimated result reveals that Russian nuclear technology exports have a significant negative impact on GHG emissions in all quantiles. Similarly, environmental policy stringency has also emerged as an important policy instrument to enhance environmental sustainability in lower quantiles (Qtile-10 to Qtile-50), while insignificant impact is observed in higher quantiles. Furthermore, rapid urbanization has emerged as a major contributor to GHG emissions and environmental degradation, while the increase in economic globalization is instrumental in enhancing environmental sustainability only at higher quantiles.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111368"},"PeriodicalIF":1.9,"publicationDate":"2025-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143632041","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":"Assessing sensitivity and uncertainty in rod ejection accidents: A combined PCE-POD approach","authors":"Yijun Zhang ,&nbsp;Wenhuai Li ,&nbsp;Sitao Peng ,&nbsp;Jinggang Li ,&nbsp;Ting Wang ,&nbsp;Qingyun He ,&nbsp;Tao Wang ,&nbsp;Haoliang Lu ,&nbsp;Zeng ling","doi":"10.1016/j.anucene.2025.111362","DOIUrl":"10.1016/j.anucene.2025.111362","url":null,"abstract":"<div><div>This study investigates the sensitivity and uncertainty of outcomes related to rod ejection accidents in nuclear reactors, focusing on the impact of perturbations in neutron parameters. By perturbing the macroscopic cross-sections of fuel assemblies, we assess how these variations influence reactor dynamics during transient events. We employ Polynomial Chaos Expansion (PCE) combined with Proper Orthogonal Decomposition (POD) to create an efficient analytical framework for quantifying uncertainty. Using Monte Carlo simulations as a benchmark, we demonstrate that PCE can accurately estimate the mean and variance of global peaking parameters with significantly fewer samples than traditional methods. Our findings reveal that only a limited number of POD modes can capture the majority of the energy distribution, facilitating a more streamlined analysis. This research enhances our understanding of reactor behavior under uncertainty and contributes to improving safety assessments in nuclear engineering.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111362"},"PeriodicalIF":1.9,"publicationDate":"2025-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143632042","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 of a new on-site semi-passive design to improve the safety of Generation II/II+ Pressurized water reactors","authors":"Wei Li ,&nbsp;Zhenhua Wu ,&nbsp;Chengcheng Deng ,&nbsp;Junmei Wu","doi":"10.1016/j.anucene.2025.111370","DOIUrl":"10.1016/j.anucene.2025.111370","url":null,"abstract":"<div><div>Generation II/II+ (Gen II/II+) Pressurized Water Reactors (PWRs) still account for a significant proportion of nuclear reactors worldwide. However, some Gen II/II+ PWR plants face a significant safety risk, as current accident management typically covers Station BlackOut (SBO) or Total Loss of FeedWater (TLFW) accidents individually. It is challenging to address the combination of SBO and TLFW (TMLB’), which can occur under extreme circumstances. Numerical simulations of severe accidents have indicated that this combination can lead to reactor core melting in as little as 2.5 h after the accident. To enhance the safety of Gen II/II+ PWRs during TMLB’, a new strategy called semi-passive cyclic cooling was proposed. This approach aims to delay accident progression by fully utilizing on-site resources, such as the cooling water inside the deaerator, the high-pressure steam in the Steam Generators (SG) as the driving force, and the feedwater pipeline in the secondary loop. This paper presents an uncertainty and sensitivity analysis of the semi-passive cyclic cooling performance during TMLB’. The fundamental characteristics of the semi-passive cyclic cooling strategy were studied using a best-estimate thermal–hydraulic model for the CPR1000, an improved Chinese Gen II/II+ PWR. An uncertainty analysis was then performed to assess the impact of various operating parameters of the reactor system and the semi-passive cyclic cooling system on the safety benefit, specifically in terms of delayed reactor core uncovery. Sobol sequence sampling-based simulations demonstrated that the semi-passive cyclic cooling system can significantly delay reactor core uncovery, with delays of 9.3173 to 9.606 h for the 5th percentile value, depending on the mode of deaerator pressurization (1,000 samples for each mode). Finally, a sensitivity analysis with both local and global methods indicated that the initial water level in the deaerator is the most critical parameter affecting the time to reactor core uncovery.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111370"},"PeriodicalIF":1.9,"publicationDate":"2025-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143632070","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 study on friction factor correlation considering the torsion effect in helical pipe flow","authors":"Thi Thanh Thuy Nguyen,&nbsp;Thi Huong Vo,&nbsp;Dinh Hung Cao,&nbsp;Nhu Viet Ha Pham","doi":"10.1016/j.anucene.2025.111290","DOIUrl":"10.1016/j.anucene.2025.111290","url":null,"abstract":"<div><div>A new correlation of friction factors in helical tubes for fully laminar flow was developed, covering curvature ratios from 0.05 to 0.15 and torsion parameters from 0.91 to 1.90. This correlation was separately analyzed the effects of curvature and torsion through regression analysis of experimental data and numerical results. The correlation was verified using experimental data and discussed on the influences of curvature ratio, torsion parameter, and Reynolds number. The friction factor decreased with a decrease in the curvature ratio and an increase the Reynolds number. At low Reynolds numbers, the friction factor decreased more rapidly compared to higher Reynolds numbers. It is confirmed that the friction factor formula for toroidal tubes was not applicable for torsion parameters exceed 1.0, especially in low Reynolds numbers flow. These findings highlight the importance of torsion effects, especially in large curvature coils, for accurate friction factor and pressure drop predictions.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111290"},"PeriodicalIF":1.9,"publicationDate":"2025-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143632040","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":"MCNPX simulations of NE213 scintillator neutron response functions and detection efficiencies up to 1 GeV","authors":"M. Tajik","doi":"10.1016/j.anucene.2025.111369","DOIUrl":"10.1016/j.anucene.2025.111369","url":null,"abstract":"<div><div>This study employed the MCNPX code to simulate the neutron response functions and detection efficiencies of an NE213 scintillator for high-energy neutrons. The MCNPX code’s PTRAC card facilitated the calculation of the scintillator’s neutron response function for mono-energetic neutrons across an energy range of 15 to 1000 MeV. MCNPX accurately reproduced the measured response for neutrons below 200 MeV. However, discrepancies emerged for higher-energy neutrons (&gt;200 MeV), particularly in regions with low light output. In these regions, the simulated values were higher than the experimental data. The calculated neutron response functions and detection efficiencies were compared using MCNPX, SCINFUL-QMD, and CECIL codes. The MCNPX simulations showed better agreement with the measured data compared to the results generated by SCINFUL-QMD and CECIL. The uncertainties associated with the detection efficiencies calculated by MCNPX were estimated to be below 8 % for neutrons in the sub-100 MeV energy region and within 15 % for the 100–600 MeV range.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111369"},"PeriodicalIF":1.9,"publicationDate":"2025-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143632071","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":"High Accuracy analytic expressions for the critical dimensions of reflected spheres and modified asymptotic diffusion theory","authors":"Shay I. Heizler","doi":"10.1016/j.anucene.2025.111319","DOIUrl":"10.1016/j.anucene.2025.111319","url":null,"abstract":"<div><div>The success of the asymptotic diffusion approximation in calculating the critical thickness or radii in different geometries with high accuracy has been well-known for decades. The high accuracy is achieved by taking into account the radius of curvature in the boundary condition in curvilinear coordinate systems, such as spherical or cylindrical systems. In reflected systems, as the simplest case of heterogeneous media, the asymptotic diffusion fails due to the continuous conditions on the boundary between the core and the reflector. Discontinuous asymptotic diffusion approximation improves dramatically the accuracy of the calculated critical thickness or radii. In this work, we study the importance of the radius of curvature correction, which is applied to the discontinuous jump conditions between the core and the reflector in simple mono-energetic (one-velocity) reflected spheres. We find a new one-velocity high-accuracy analytic expression for the critical radii, coated by a general-depth reflector, in spherical geometry. The accuracy of the analytic expression is better than 1% accuracy compared to the calculated exact transport critical radii. The radius of curvature corrected discontinuous conditions give rise to a new modified diffusion-like equation that reproduces the high accuracy of the critical radii of reflected systems. The new modified equation is tested via numerical simulations, yielding high accuracy with the analytic expression results.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111319"},"PeriodicalIF":1.9,"publicationDate":"2025-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143632039","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":"Neutron moderation in Pr3+ doped sodium lead calcium borate glasses for radiography and shielding","authors":"Prashant N. Patil ,&nbsp;M.M. Hosamani ,&nbsp;K.L. Susheela ,&nbsp;N.M. Badiger","doi":"10.1016/j.anucene.2025.111373","DOIUrl":"10.1016/j.anucene.2025.111373","url":null,"abstract":"<div><div>In this work, the physically examined sodium lead calcium borate glasses doped with Pr³⁺ at different concentrations have been examined qualitatively to deduce the neutron moderation parameters for collimating the neutron beam in radiography. Physically examined glass samples have a nominal composition matrix of 20Na₂O₃-10PbO-(60-x)B₂O₃-xPr₂O₃ with 10Ca₂O, 10Ba₂O,<!--> <!-->and 10Sr₂O substitutions. The neutron moderation properties such as effective removal cross section, relative thermal lethargy, average logarithmic energy decrement, absorption cross sections, moderation power, moderation efficiency, shared energy at each collision, and average number of collisions for all the elements, and above-mentioned glass samples have been deduced. In addition, the effective or equivalent atomic number and atomic mass for neutron interaction with the glass samples has been estimated. Obtained an interesting trend with the density of the material and analysed with a code.</div><div>PACS No.: 28.20. − v; <span><span>07.05.Tp</span><svg><path></path></svg></span>; <span><span>28.20.Cz</span><svg><path></path></svg></span>; <span><span>28.20.Gd</span><svg><path></path></svg></span>; <span><span>28.20.Pr</span><svg><path></path></svg></span>; 28.41.Ak; <span><span>28.41.Pa</span><svg><path></path></svg></span>.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111373"},"PeriodicalIF":1.9,"publicationDate":"2025-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143632072","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}