Annals of Nuclear Energy最新文献

{"title":"An interpretable deep transfer learning method for fault diagnosis of nuclear power plants under multiple power level conditions","authors":"Zhanguo Ma ,&nbsp;Wenhao Jia ,&nbsp;Long Tian ,&nbsp;Jing Cui ,&nbsp;Dihao Zheng ,&nbsp;Ziyang Cui","doi":"10.1016/j.anucene.2025.111582","DOIUrl":"10.1016/j.anucene.2025.111582","url":null,"abstract":"<div><div>Nuclear power plants (NPPs) operations under different power level conditions (i.e., different operating modes) often exhibit non-independent and identically distributed (non-IID) characteristics in their fault-related parameters, posing significant challenges to traditional data-driven fault diagnosis methods. To address this issue, the study proposes a fault diagnosis approach that combines deep transfer learning with an interpretable multi-variable gated recurrent unit (IMV-GRU) model. The proposed approach incorporates a hybrid loss strategy integrating adaptive focal loss (AFL) and maximum mean discrepancy (MMD) to improve cross-power-level feature transfer capability. The interpretability of IMV-GRU is demonstrated through its autonomous quantification of multi-variable contribution degrees, enabling feature selection to optimize computational efficiency and mitigate interference from non-key variables. Experimental results demonstrate that the proposed method is effective in cross-power-level fault diagnosis, with particularly significant accuracy improvements under sparse data conditions. Furthermore, the effectiveness of extracting multi-variable contribution degrees is validated, highlighting its value in fault diagnosis.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"222 ","pages":"Article 111582"},"PeriodicalIF":1.9,"publicationDate":"2025-06-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144253532","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":"Power distribution reconstruction from in-core detector measurements using gaussian process regression","authors":"Artur Souza e Silva,&nbsp;Alessandro da Cruz Gonçalves,&nbsp;Willian Vieira de Abreu,&nbsp;Adilson Costa da Silva,&nbsp;Aquilino Senra Martinez","doi":"10.1016/j.anucene.2025.111581","DOIUrl":"10.1016/j.anucene.2025.111581","url":null,"abstract":"<div><div>In order to safely and efficiently operate pressurized water reactors, it is essential to monitor neutron flux and power distributions throughout the operation cycle. Most of currently operated reactors have in-core instrumentation systems capable of taking assembly-wise axially integrated flux measurements. However, commercial reactors typically possess few fuel assemblies that contain guide tubes for in-core instrumentation and it is necessary to estimate power values at the remaining fuel assemblies. This work employs gaussian process regression, a non-parametric supervised learning method, to predict the power distribution over the entire core using only measured data. The OpenMC Monte Carlo code was employed to emulate detector signals from instrumented assemblies as well as to produce reference values at non-instrumented assemblies against which the regression method was validated. The mean and maximum relative discrepancies were below 0.8% and 2.5%, respectively, in a scenario with full detector availability. Considering the failure of a single detector, the maximum relative discrepancy did not exceed 5.3%, showing the feasibility of the model as an operation monitoring system capable of real-time reconstruction of the core power distribution.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"223 ","pages":"Article 111581"},"PeriodicalIF":1.9,"publicationDate":"2025-06-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144220949","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":"Ex-vessel stabilization of corium: An analysis of corium-concrete interaction with top flooding for siliceous concrete","authors":"Florian Fichot","doi":"10.1016/j.anucene.2025.111584","DOIUrl":"10.1016/j.anucene.2025.111584","url":null,"abstract":"&lt;div&gt;&lt;div&gt;In case of severe accident without possibility of in-vessel retention, the corium must be stabilized outside the vessel, either in a designed core-catcher or directly on the concrete basemat of the reactor building, after spreading. In this last case, the stabilization strategy must be efficient enough to avoid the progression of corium through the concrete basemat and to the environment.&lt;/div&gt;&lt;div&gt;In order to ensure corium stabilization, the physical processes leading to quench some part of the liquid corium must be fast enough, compared to the erosion of concrete. One of the possible strategies consists in spreading the corium over a rather large area (i.e. the reactor pit plus one or two adjacent rooms) and flood it with water on top. Two processes have been identified experimentally for a potential quenching of corium: melt eruption and water ingression. Melt eruption is a process that is driven by the flow of gases (CO&lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mrow&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt; and H&lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mrow&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt;O) generated by concrete erosion: the gas flow is likely to entrain droplets of corium into the overlying water, quenching the droplets quickly. Water ingression is a process where cracks are generated by the contraction of the top crust due to cooling: this process enables the flow of water into the cracks and the propagation of the cracking front deep into the corium crust. When the basemat is made of siliceous concrete, stabilization depends mostly on water ingression because there is little gas content in such concrete and melt eruption may only have a limited effect.&lt;/div&gt;&lt;div&gt;This paper presents the analysis of a situation where corium is spread over a rather large area (around 80 m&lt;span&gt;&lt;math&gt;&lt;msup&gt;&lt;mrow&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/mrow&gt;&lt;/msup&gt;&lt;/math&gt;&lt;/span&gt;) in order to reduce the corium thickness after spreading. Of course, this thickness depends on the reactor design (total volume of corium divided by total spreading area). After spreading, corium is flooded with water on top. First, the melt eruption process is investigated: it is shown that it has a limited efficiency in case of siliceous concrete but it plays an important role in triggering the water ingression earlier. Then, the process of water ingression is examined, as it is the process leading to complete stabilization. The maximum heat flux extracted by water ingression (CHF) is the most important parameter in this process and needs to be evaluated accurately. Therefore, a theoretical determination of the CHF is made, as a function of the concrete content in the corium, based on the existing Lister-Epstein model, on data obtained earlier at Argonne National Laboratory and on data measured in volcanic rocks. In addition, a simple modeling of transient heat conduction through the concrete basemat is proposed, to determine a criterion of stop of erosion that can be used in a lumped-parameter code. As a","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"223 ","pages":"Article 111584"},"PeriodicalIF":1.9,"publicationDate":"2025-06-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144230267","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":"Mechanistic CHF model development based on the liquid sublayer dryout model for flow boiling phenomena under IVR-ERVC conditions","authors":"Min Suk Lee ,&nbsp;Jun Yeong Jung ,&nbsp;Yong Hoon Jeong","doi":"10.1016/j.anucene.2025.111627","DOIUrl":"10.1016/j.anucene.2025.111627","url":null,"abstract":"<div><div>In this work a mechanistic critical heat flux (CHF) model was developed based on the liquid sublayer dryout (LSD) model, designed to predict CHF in downward-facing flow boiling conditions, particularly the operational environment of In-Vessel Retention through External Reactor Vessel Cooling (IVR-ERVC). For accurate CHF prediction using the LSD model, several key parameters were properly defined: relative velocity between liquid and vapor, slug length, liquid velocity, and liquid sublayer thickness. In this study, these parameters were interpreted from the perspective of vapor slug dynamics, which have been well characterized through prior experimental observations. The developed CHF model was validated against previous well-established experimental results, including ULPU, FIRM, MIT, and KAIST experiments, with the CHF model demonstrating excellent predictability with a root mean square error of 14.53%.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"222 ","pages":"Article 111627"},"PeriodicalIF":1.9,"publicationDate":"2025-06-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144222131","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 different TRU compositions on system response during an unprotected station blackout in small lead-cooled reactors","authors":"Fredrik Dehlin,&nbsp;Janne Wallenius","doi":"10.1016/j.anucene.2025.111586","DOIUrl":"10.1016/j.anucene.2025.111586","url":null,"abstract":"<div><div>The dynamic response to an Unprotected Station Blackout (USBO) has been evaluated for a small, lead-cooled reactor when fuelled with two different actinide compositions: one sourced from spent light water reactor (LWR) fuel and the other from UN fuel discharged from a small LFR. We demonstrate that a reduction in the delayed neutron fraction, primarily due to the addition of americium, leads to lower peak temperatures during phase one of the USBO. This reduction could help with ensuring cladding integrity despite an increased internal gas pressure resulting from helium production during the decay of ²⁴²Cm. It is also shown that the coolant volume required to buffer decay heat until vessel air cooling becomes effective must be increased to ensure the integrity of the fuel cladding. We conclude by demonstrating that (U,Pu)N fuel, with negligible ²⁴¹Pu content, offers the best properties to ensure cladding integrity during the USBO.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"222 ","pages":"Article 111586"},"PeriodicalIF":1.9,"publicationDate":"2025-06-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144222136","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":"Performance of the borate glasses doped with some rare earth compounds during the neutron irradiation: Analysis for the mass removal cross section of fast neutrons and secondary gamma-ray emission","authors":"O.L. Tashlykov ,&nbsp;K.A. Mahmoud","doi":"10.1016/j.anucene.2025.111630","DOIUrl":"10.1016/j.anucene.2025.111630","url":null,"abstract":"<div><div>Three zinc borate glasses doped with a small concentration of various rare earth oxides (La₂O₃, Eu₂O₃, and Dy₂O₃) were fabricated using the melt quenching method. The fabrication of glass samples aims to investigate the impact of rare earth elements on the secondary gamma-ray emission following neutron flux irradiation. To achieve the desired goal, the prepared glass samples were placed in a capsule, which is suspended within a dry channel of the nuclear reactor IVV-2 M. The prepared samples were irradiated with fast neutron flux of 8.11E + 11 n/(cm².s) and a thermal neutron flux of 8156E + 12 n/(cm².s) for four hours. The absorbed doses by the prepared glass samples reach 7.41E + 04, 3.06E + 04, and 5.44E + 04 Gy for glasses doped with La₂O₃, Eu₂O₃, and Dy₂O₃, respectively. Immediately after the irradiation process ended, analysis for the secondary gamma-emission was performed using a HPGe detector from the CANBERRA gamma spectrometer equipped with Genie-2000 software. The analysis shows that the highest total activity reaches 14.45 MBq, which is released from glasses-doped La₂O₃, while the lowest activity reaches 6.26 MBq, released from glasses-doped Eu₂O₃. Additionally, the dose rate from the irradiated glass samples was estimated at various distances surrounding the irradiated samples. Also, the dose rate across various periods of time was estimated for the irradiated glass samples.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"222 ","pages":"Article 111630"},"PeriodicalIF":1.9,"publicationDate":"2025-06-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144222129","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":"Fault detection for nuclear power plant based on improved moving window and sparse autoencoder","authors":"Shaomin Zhu ,&nbsp;Wenzhe Yin ,&nbsp;Hong Xia","doi":"10.1016/j.anucene.2025.111626","DOIUrl":"10.1016/j.anucene.2025.111626","url":null,"abstract":"<div><div>Due to factors such as component performance degradation and changes in operating conditions, nuclear power plants (NPPs) equipment exhibits significant time-varying characteristics during operation, leading to the failure of fault detection models. Therefore, this study proposes a fault detection method based on an improved moving window and sparse autoencoder to enhance the adaptability of the detection method to the time-varying data of NPPs. This method establishes a sparse autoencoder as a fault detection model, determining the operating status of equipment by calculating the statistical relationship between test data and reconstructed data. In this process, the traditional moving window update strategy is optimized based on Euclidean distance, and the improved moving window strategy enables effective model updating. Finally, the effectiveness of the proposed method is verified using data from a nuclear power plant reactor coolant pump. The results show that the proposed method performs well in terms of fault detection rate and false alarm rate.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"222 ","pages":"Article 111626"},"PeriodicalIF":1.9,"publicationDate":"2025-06-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144213437","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 energy collapsing on the effective neutron lifetime","authors":"S. Dulla ,&nbsp;N. Abrate ,&nbsp;P. Ravetto ,&nbsp;P. Saracco ,&nbsp;M. Carta ,&nbsp;V. Fabrizio ,&nbsp;V. Peluso","doi":"10.1016/j.anucene.2025.111550","DOIUrl":"10.1016/j.anucene.2025.111550","url":null,"abstract":"<div><div>The effective mean prompt neutron generation time (or effective lifetime) is an integral parameter that is introduced in the point kinetic model for nuclear reactor time-dependent analysis. Although such a model requires a strong simplification of the neutron kinetic process, the value of the effective neutron lifetime can give an immediate and useful information on the physical characteristics of a multiplying system and on its time response to perturbations. Furthermore, point kinetics is still used for the simulation of control and transient situations and, especially, in an inverse fashion for the interpretation of neutronic experiments. Based on the standard separation-projection mathematical procedure to derive point kinetics equations, the effective lifetime is defined as the ratio between the total instantaneous importance within the system and the total importance generated by fission per unit time. The evaluation of the effective lifetime can be carried out by both deterministic and stochastic computational tools. Relevant differences can be observed if different physical models are used. In this paper the attention is particularly focused on the energy structure employed. In the first part some analytical analyses for simplified configurations are carried out, in order to gain some physical insight on the effects associated with the detail of the energy group structure on the computed value of the parameter. Then, some more detailed numerical studies allow to investigate more complex configurations. The study includes both critical and subcritical, source-driven systems.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"222 ","pages":"Article 111550"},"PeriodicalIF":1.9,"publicationDate":"2025-06-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144222239","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":"Deconvolution approach for determining absolute prompt γ-ray intensities of 52V yielded from the 51V(n, γ)52V reaction at the Dalat Nuclear Research Reactor","authors":"Thanh Tai Chau ,&nbsp;Ngoc Son Pham ,&nbsp;Thien Thanh Tran ,&nbsp;Van Tao Chau","doi":"10.1016/j.anucene.2025.111587","DOIUrl":"10.1016/j.anucene.2025.111587","url":null,"abstract":"<div><div>All previous research relied on equipment such as Compton suppression spectrometers or coincidence spectrometers to determine the absolute prompt <span><math><mi>γ</mi></math></span>-ray intensities of ⁵²V obtained from ⁵¹V(n, <span><math><mi>γ</mi></math></span>) reaction. In the present work, we utilize a deconvolution method to determine the absolute prompt <span><math><mi>γ</mi></math></span>-ray intensities of the ⁵²V measured by the HPGe detector of the PGNA spectrometer at horizontal channel No. 2 of the Dalat Nuclear Research Reactor (DNRR). This new method is based on the validated Geant4 simulation to deconvolute the prompt <span><math><mi>γ</mi></math></span>-ray spectrum of ⁵²V and an in-beam saturated neutron capture reaction in a vanadium foil as the research sample and self-internal standard. Subsequently, by counting the 1.434 MeV delayed <span><math><mi>γ</mi></math></span>-ray peak emitted by ⁵²V, the absolute prompt <span><math><mi>γ</mi></math></span>-ray intensities can be obtained. These <span><math><mi>γ</mi></math></span>-ray intensities were then compared to those in the ENSDF database and other references, which show a good agreement.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"222 ","pages":"Article 111587"},"PeriodicalIF":1.9,"publicationDate":"2025-06-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144195832","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":"Optimizing sorosilicate crystals for γ-ray shielding: Role of density and SiO2 content","authors":"Z.Y. Khattari ,&nbsp;H. Tekin ,&nbsp;Y. Rammah","doi":"10.1016/j.anucene.2025.111618","DOIUrl":"10.1016/j.anucene.2025.111618","url":null,"abstract":"<div><div>This study investigates the radiation shielding properties of a series of sorosilicate crystals by examining their LAC values at photon energies of 0.015 to 15.0 MeV. Detailed analyses were conducted to explore the effects of chemical composition, particularly SiO₂ weight percentage, on the LAC value of these crystals. The results reveal that the LAC value is significantly influenced by both the density and the SiO₂ content of the materials. Notably, the LAC value increases with density and exhibits a non-linear cubic relationship with SiO₂ wt%, with maximum attenuation observed at approximately 25 % SiO₂ wt% and E = 0.015 MeV for the Hemimorphite crystal. These findings suggest that sorosilicate crystals with optimized SiO₂ content and density offer promising potential for advanced radiation shielding applications. The study provides crucial insights into the material selection and design for radiation protection, emphasizing the importance of compositional control in enhancing shielding efficiency.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"222 ","pages":"Article 111618"},"PeriodicalIF":1.9,"publicationDate":"2025-06-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144195831","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}