Annals of Nuclear Energy最新文献

{"title":"Investigation of shielding and dosimetric parameters for neutrons emitted from various types of fission sources","authors":"Rahim Khabaz ,&nbsp;Harith Mohamed Al-Azri","doi":"10.1016/j.anucene.2025.111541","DOIUrl":"10.1016/j.anucene.2025.111541","url":null,"abstract":"<div><div>This study evaluates the shielding and dosimetric characteristics of neutrons emitted from various fission sources, utilizing the Monte Carlo simulations. Neutron emissions, crucial in nuclear applications and safety, are analyzed across a spectrum of fissionable isotopes. The work focuses on determining the mean fluence-to-dose conversion coefficients, average total macroscopic cross sections, as well as the flux and dose-equivalent buildup factors for various shielding materials. Results highlight that neutron flux and dose-equivalent buildup factors increase with shield thickness, influenced significantly by the shielding material’s effective mass number. Notably, concrete exhibits the highest buildup factor, contrasting with water, which shows the lowest. Furthermore, polynomial equations parameterized by source and material coefficients facilitate accurate flux and dose-equivalent buildup factor calculations. This study enhances the understanding of neutron shielding effectiveness and dosimetry, which is crucial for optimizing radiation protection measures in various nuclear contexts, from reactor operations to medical applications.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111541"},"PeriodicalIF":1.9,"publicationDate":"2025-05-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143903658","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":"Effect of permutation and combination patterns on the performances of shielding materials","authors":"Dengjian Wu ,&nbsp;Junjun Gong ,&nbsp;Zifu Hao ,&nbsp;Chengqiang Liang","doi":"10.1016/j.anucene.2025.111465","DOIUrl":"10.1016/j.anucene.2025.111465","url":null,"abstract":"<div><div>Composite materials have been extensively employed to enhance the performance of shielding materials under the condition of neutron and γ radiation fields. Researchers generally design the component based on the energy spectrum information of the radiation field with Monte Carlo software, then achieve the uniform mixing of components with specific technological processes, ultimately obtaining the composite shielding material with expected performance. However, due to the substantial differences in properties between the components of composite materials, the process of mixing them uniformly is complex and costly. This drawback may be unacceptable in specific situations such as scenarios requiring extensive use and complex on-site installation environments. Lead boron polyethylene (PbBPE), a widely studied and utilized composite material, was analyzed using the Geant4 software in conjunction with the K-Nearest Neighbors (KNN) algorithm to calculate and analyze the performance of laminated materials with the same mass fraction and total mass but with different permutations. The results indicate that the performance of the sandwich-like laminated material formed by filling boron polyethylene (BPE) between two layers of lead (Pb) plates may be superior to that of the composite material PbBPE.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111465"},"PeriodicalIF":1.9,"publicationDate":"2025-05-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143903659","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":"Influence of level density models on proton and deuteron-induced reactions using zinc target for the production of 66-68Ga medical radioisotopes","authors":"Bassam T. Al-Azraq ,&nbsp;Raghad I. Mahmood ,&nbsp;Lubna Abduljabbar Mahmood ,&nbsp;Radhwan Ch. Mohsin ,&nbsp;Rusul S. Jaffer","doi":"10.1016/j.anucene.2025.111535","DOIUrl":"10.1016/j.anucene.2025.111535","url":null,"abstract":"<div><div>Due to their medical importance in imaging and therapy, this paper examines proton- and deuteron-induced reactions on zinc targets from a theoretical perspective. Medically important ⁶⁶Ga, ⁶⁷Ga, and ⁶⁸Ga were simulated using TALYS 2.0 and different level density models. The relative variance technique was used to assess model agreement with experimental data given in EXFOR library. Key production parameters, including radionuclidic impurities, optimal energy range, theoretical yield, and target thickness, were derived from the selected theoretical model, which was chosen based on its agreement with experimental trends. Results confirmed the models’ predictive accuracy across a wide energy range in estimating the nuclear reaction cross-sections for ^66,67,68Ga production. For a number of studied reactions, the TGHFB model specifically agreed very well with the experimental data. Zinc targets proved effective and feasible for producing ^66,67,68Ga at low proton and deuteron energies, as typically available in medical cyclotrons.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111535"},"PeriodicalIF":1.9,"publicationDate":"2025-05-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143903660","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":"Lead-cooled fast reactor SGTR accident pressure wave analysis and structural response","authors":"Xi Huang ,&nbsp;Chong Qin ,&nbsp;Kefan Zhang ,&nbsp;Lixiang Zhang ,&nbsp;Hongli Chen","doi":"10.1016/j.anucene.2025.111489","DOIUrl":"10.1016/j.anucene.2025.111489","url":null,"abstract":"<div><div>Steam Generator Tube Rupture (SGTR) accidents are one of the most serious types of accidents in Lead-cooled Fast Reactors (LFRs). Due to the large pressure and temperature differences between the two sides of the steam generator heat transfer tubes in lead–bismuth fast reactors, combined with the corrosive effects of liquid lead–bismuth eutectic (LBE), steam generator heat transfer tube rupture accidents may occur. In this paper, the SGTR accident in the lead-cooled fast reactor M²LFR-1000 is taken as an example. The computational fluid dynamics (CFD) program is used to simulate the early stage of SGTR accident, and the correctness of the model is verified through experiments. The severity of the accident under different rupture sizes is also assessed. By coupling ANSYS Fluent with ANSYS Structural, the impact of a single heat exchanger tube rupture on neighboring tubes is calculated to evaluate the possibility of chain rupture accidents. The results of the study show that as the pressure wave peak increases with larger rupture sizes, the risk of localized fracture is higher in internally threaded tubes than in smooth tubes, as determined by stress assessment.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111489"},"PeriodicalIF":1.9,"publicationDate":"2025-05-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143903581","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":"Nuclear hydrogen demonstration project using the HTTR – Demarcation of nuclear-industrial laws and design standards","authors":"Takeshi Aoki,&nbsp;Atsushi Shimizu,&nbsp;Katsunori Ishii,&nbsp;Keisuke Morita,&nbsp;Naoki Mizuta,&nbsp;Kaoru Kurahayashi,&nbsp;Takanori Yasuda,&nbsp;Hiroki Noguchi,&nbsp;Yasunobu Nomoto,&nbsp;Kazuhiko Iigaki,&nbsp;Hiroyuki Sato,&nbsp;Nariaki Sakaba","doi":"10.1016/j.anucene.2025.111503","DOIUrl":"10.1016/j.anucene.2025.111503","url":null,"abstract":"<div><div>A high temperature gas cooled reactor has inherent safety features and potential to produce competitive and large amounts of carbon-free hydrogen. Japan Atomic Energy Agency has initiated the high temperature engineering test reactor (HTTR) Heat Application Test Project to establish coupling technologies for a nuclear hydrogen production system that connects a nuclear reactor plant and a hydrogen production plant. In order to ensure the public safety, appropriate laws and design standards should be applied to the hydrogen production plant because it handles combustible and toxic gases. The present study proposed a relative evaluation methodology using six metrics to characterize the superiority of candidates for the demarcation of applicable laws and design standards for a for the nuclear hydrogen production system. The proposed methodology was applied to the HTTR Heat Application Test Facility to select the most superior demarcation of applicable laws and design standards from six candidates for the HTTR Heat Application Test Facility. Candidates, applying nuclear law to all facilities, showed least superiority due to the higher cost of commercialized hydrogen production and the difficulty in entry for non-nuclear vendors to the business. The candidates, applying the High Pressure Gas Safety Act for the steam reformer and the Heat Application Test Facility (hydrogen production plant), showed least superiority in the feasibility of the system or the demonstration of the key equipment and technologies for commercialization depending on the installation of the secondary intermediate heat exchanger. On the other hand, a candidate applying the High Pressure Gas Safety Act to the Heat Application Test Facility (hydrogen production plant) and design standards established under the High Pressure Gas Safety Act to the steam reformer did not show the lowest category in any of the metrics, and was proposed as the most superior candidate for the demarcation of applicable laws and design standards the HTTR Heat Application Test Facility.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111503"},"PeriodicalIF":1.9,"publicationDate":"2025-05-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143900302","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 and CFD simulation of a printed circuit heat exchanger design","authors":"Muhammad Ali Khan ,&nbsp;Muhammad Ilyas ,&nbsp;Khalid Waheed ,&nbsp;Inamul Haq ,&nbsp;Fatih Aydogan","doi":"10.1016/j.anucene.2025.111510","DOIUrl":"10.1016/j.anucene.2025.111510","url":null,"abstract":"<div><div>A 4×4 mm straight square channel printed circuit heat exchanger is designed and manufactured for applications in nuclear facilities. Thermal and hydraulic characteristics of the heat exchanger are evaluated on a test rig using water as the coolant on both sides over a mass flux range from 62.5 kg/m²s to 562.5 kg/m²s. The pressure loss is measured across the hot and cold channels. The friction factor of the hot side is 4 times and the cold side is 3 times higher compared to the straight circular pipe. A friction factor correlation based on experimental data is developed for both sides of the channel. Heat transfer experiments are conducted with the hot inlet temperature varied from 331.15 K to 349.15 K. The heat transfer rate rises with an increase in hot inlet temperature and mass flux. The heat transfer rate increases linearly with the mass flux for a fixed hot inlet temperature. The heat transfer rate also depends on the hot inlet temperature. A 1.9 % and 4.6 % increase in the hot inlet temperature results in a 28 % and 72.2 % increase in the heat transfer rate, respectively, for a mass flux of 281.25 kg/m²s on the hot side and 562.5 kg/m²s on the cold side. Computational fluid dynamics simulations are performed for the printed circuit heat exchanger. The simulations are validated through experimental data. The simulation results are used to study the heat transfer characteristics of the heat exchanger. For the maximum mass flux and hot inlet temperature, the hot and cold sides exhibit a Nusselt number of 26.6 and 28.8, respectively. Correlations for the hot and cold side Nusselt number for the square channel are also developed. The friction factor and the Nusselt number are used to evaluate the Performance Evaluation Criteria. For the maximum operating conditions of the study, a Performance Evaluation Criteria of 8.5 is achieved on both the hot and cold sides, respectively. The heat exchanger design is highly compact and efficient in heat transfer, making it suitable for reducing the size and enhancing the efficiency of small modular reactors.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111510"},"PeriodicalIF":1.9,"publicationDate":"2025-05-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143900300","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Numerical investigation on ultimate pressure capacity of a SMR containment vessel","authors":"Min Jeong Park ,&nbsp;Han-Sang Woo ,&nbsp;Dong-Hyeon Choi,&nbsp;Yoon-Suk Chang","doi":"10.1016/j.anucene.2025.111530","DOIUrl":"10.1016/j.anucene.2025.111530","url":null,"abstract":"<div><div>The containment structure is a vital component ensuring the safety of nuclear power plants by preventing the release of radioactive materials under severe accident-induced conditions. As small modular reactors (SMRs) have been developing globally, steel containment vessels (CVs) have replaced conventional containment buildings and new assessment method should be established with regard to safety. In this study, ultimate pressure capacity (UPC) was assessed for the Innovative SMR (i-SMR) steel CV by using the finite element analysis software ABAQUS. Free-field areas were proposed with equations, through comparisons of UPCs at different locations. By using maximum principal, hoop, and membrane strain, the most appropriate strain parameter was suggested. Furthermore, temperature effect under the postulated severe accident-induced condition and buckling initiation pressure were assessed. The findings provide a comprehensive methodology for assessing the UPC of not only for the i-SMR but also for the advanced reactors utilizing steel CV.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111530"},"PeriodicalIF":1.9,"publicationDate":"2025-05-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143900303","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":"Progress in understanding fission product remobilization and hydrogen risk in water-cooled reactors: OECD/NEA THAI-3 project","authors":"S. Gupta ,&nbsp;M. Freitag ,&nbsp;E. Schmidt ,&nbsp;M. Colombet ,&nbsp;B. von Laufenberg ,&nbsp;G. Langrock ,&nbsp;F. Funke","doi":"10.1016/j.anucene.2025.111498","DOIUrl":"10.1016/j.anucene.2025.111498","url":null,"abstract":"<div><div>OECD/NEA THAI-3 project aimed to investigate hydrogen risk and source term issues with specific emphasis on representative boundary conditions as expected during a severe accident in light water reactors. The project was conducted between 2016 and 2019, hosted by Germany and supported by the signatories from 16 countries. The experimental program was conducted in the technical-scale THAI⁺ facility, comprising two interconnected vessels with a total volume of 80 m³.</div><div>The hydrogen combustion tests investigated the impact of varying initial flow conditions, gas composition, and burn direction on pressure buildup, flame front propagation, and jet-ignition effects during H₂-deflagration. One of the hydrogen deflagration tests (test HD-44) served as a benchmark for validating combustion models. Hydrogen recombiner (PAR) tests provided data on the onset of recombination, recombination rate, and hydrogen depletion efficiency under counter-current flow conditions, with test HR-49 serving as a code benchmark. The source term related experiments examined the re-entrainment of fission products (CsI, I₂) from water pools, showing CsI re-entrainment increased tenfold in the case of a water pool with reduced surface tension. The results of fission product resuspension tests revealed significant resuspension of fission products from paint and steel surfaces during hydrogen deflagration with the potential to remain gas-borne over a long time (“fine particles”), with notable release of organic iodine and decomposition of CsI-aerosol to gaseous I₂ in the high-temperature environment.</div><div>This paper presents key findings from the OECD/NEA THAI-3 project, highlighting their importance in mitigating hydrogen risk and source term in nuclear safety. It also summarizes the application of these results for code validation and reactor analyses to enhance severe accident management in LWRs. Additionally, the paper discusses remaining open issues identified through THAI-3, along with the associated experimental needs. The paper includes information about some of these unresolved issues addressed within the follow-up OECD/NEA THEMIS project.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111498"},"PeriodicalIF":1.9,"publicationDate":"2025-05-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143900301","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":"Verification of the nTRACER/CTF code system for full core high resolution cycle analysis with the OECD/NEA TVA Watts Bar Unit 1 benchmark","authors":"Sung Joon Kwon ,&nbsp;Marianna Papadionysiou ,&nbsp;Yeon Sang Jung ,&nbsp;Pascal Rouxelin ,&nbsp;Alexander Vasiliev ,&nbsp;Hakim Ferroukhi ,&nbsp;Mathieu Hursin ,&nbsp;Hyung Jin Shim","doi":"10.1016/j.anucene.2025.111532","DOIUrl":"10.1016/j.anucene.2025.111532","url":null,"abstract":"<div><div>This study aims to develop and verify the multi-physics high-resolution coupled code system nTRACER/CTF for full core depletion calculations in Cartesian geometry. The verifications are conducted on the basis of the OECD/NEA TVA Watts Bar 1 benchmark (TVA WB1), using multi-physics high-resolution results from the Virtual Environment for Reactor Applications (VERA), which includes MPACT and CTF. The paper presents the development of the nTRACER/CTF coupled system for depletion calculations in PWRs, which are based on the Cartesian geometry. The nTRACER/CTF model is verified under the Hot Zero Power Condition (HZP) at the beginning of the first cycle and the HFP case of TVA WB1. The Critical Boron Concentration (CBC) differences are less than 17 ppm in both cases, while the RMS differences in pin-wise power distributions are less than 0.7 % in both cases. The impacts of discrepancies between the nTRACER/CTF and VERA models are also quantified. nTRACER/CTF depletion calculations are performed for the full first cycle of TVA WB1 according to the simplified power curve presented in the benchmark specifications. The system is verified with VERA results for every burnup step of the cycle, in terms of CBC, axially averaged pin power distribution, outlet temperature distribution and axial power profile. The RMS of CBC differences after achieving a full power condition is 25 ppm. The RMS differences in pin-wise power and outlet temperature distributions are less than 1.12 % and 0.4 °C, which are within the target accuracies of 1.5 % and 2 °C for all depletion steps.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111532"},"PeriodicalIF":1.9,"publicationDate":"2025-05-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143900297","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":"Early fault detection method for nuclear power plants based on sparse denoising autoencoder and kernel principal component analysis","authors":"Wenzhe Yin,&nbsp;Hong Xia,&nbsp;Xueying Huang,&nbsp;Longfei Shan,&nbsp;Wenhao Ran,&nbsp;Zhujun Jia","doi":"10.1016/j.anucene.2025.111460","DOIUrl":"10.1016/j.anucene.2025.111460","url":null,"abstract":"<div><div>Effective fault detection technology is of great significance to the safety and economy of nuclear power plants (NPPs). To accurately identify early faults in NPPs, this study proposes a novel fault detection method based on sparse denoising autoencoder (SDAE) and kernel principal component analysis (KPCA). First, the operating data of NPPs is collected by numerous sensors, and the operating parameters are grouped according to physical properties. Then, the corresponding fault detection model is established according to each parameter group, and each detection model consists of the SDAE and KPCA. The case study evaluated four accident scenarios (LOCA, SLBIC, FHAIC, FHAIAB) across two development degrees (0–1 % and 0–0.1 %). The proposed method achieved fault detection rates of 99.07 %, 95.20 %, 99.73 %, and 99.60 % for the 0–1 % degree with zero false alarms. Even for the subtler 0–0.1 % degree, it maintained a 94.84 % average detection rate and no false alarms. Compared to traditional methods, its average fault detection rate was higher than that of PCA and KPCA by 62.9 % and 32.4 % (0–1 % degree), and by 89.5 % and 88 % (0–0.1% degree), demonstrating its potential application value in NPPs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111460"},"PeriodicalIF":1.9,"publicationDate":"2025-05-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143895778","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}