Annals of Nuclear Energy最新文献

{"title":"The nucleation characteristics of geyser boiling in sodium heat pipes","authors":"","doi":"10.1016/j.anucene.2024.111010","DOIUrl":"10.1016/j.anucene.2024.111010","url":null,"abstract":"<div><div>Many studies have been carried out on the geyser boiling phenomenon in liquid metal heat pipes, but most of them are phenomenological. Considering that nucleation should be the trigger of geyser boiling, occurrence of geyser boiling should be predicted by studying related nucleation characteristics. In this paper, the experimental value of nucleation superheat were gained and analyzed. The results showed that the main parameters affecting the nucleation superheat of sodium heat pipe were heating power and pressure, and the influences of wick mesh number, inclination angle, liquid filling ratio etc. were not important. By fitting the experimental data with heat flux and wall cavity vapor pressure, a correlation which could predict the data well was proposed. Further investigation showed that there may be materials with bad wetting properties in the wall cavities, whose surface thermo-physical properties could be greatly affected by temperature, resulting in the decrease of nucleation contact angle with increase in wall temperature.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526953","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":"European research reactor strategy derived in the scope of the towards optimized use of research reactors (TOURR) project","authors":"","doi":"10.1016/j.anucene.2024.110963","DOIUrl":"10.1016/j.anucene.2024.110963","url":null,"abstract":"<div><div>Nuclear research reactors ( RR ) are essential facilities in countries implementing nuclear power plants and are used for experiments necessary for commercial reactor development, training and education programs, and many other applications not related to nuclear energy production (e.g., isotope production, neutron sources, materials science). Europe has a broad and very diverse landscape of RRs, many of which have been in operation for 30-60 years, are well maintained and regularly modernized. However, financial pressures caused by a combination of declining interest and the lack of a sound financial model have led to the closure of many of them (e.g. OSIRIS in Saclay, JEEP II research reactor at IFE Kjeller and BER2 in Berlin). These negative trends called for coordinated European action to assess the impact of the declining number of RRs. The Towards Optimized Use of Research Reactors (TOURR) project was a response to this challenge. Its main objective was to assess the status of the EU RR fleet and to develop a strategy for the refurbishment and construction of new RR in Europe. The assessment was based on analysed data obtained through extensive questionnaires sent to all operating European RR. The analysis revealed gaps in terms of lack of long-term funding, lack of manpower and lack of communication between RRs and their customers. It also showed threats of further European RR closures. Regarding the long-term EU RR strategy, the main recommendations of the TOURR project are to build (at least) two RRs, a medium-flux multipurpose reactor and a flexible zero-power facility. Both reactor cores could be part of a single facility built at the European level and accessible to all EU Member States.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142527009","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":"Analysis of internal flow excitation characteristics of reactor coolant pump based on DMD","authors":"","doi":"10.1016/j.anucene.2024.111011","DOIUrl":"10.1016/j.anucene.2024.111011","url":null,"abstract":"<div><div>This paper presents a study on the internal flow characteristics of reactor coolant pumps using Dynamic Mode Decomposition (DMD) technology. As a core component of nuclear power plants, the internal flow characteristics of reactor coolant pumps play a crucial role in the performance and stability of the pumps. This paper initially introduces the application of DMD and Proper Orthogonal Decomposition (POD) methods in fluid mechanics, emphasizing the effectiveness of DMD in analyzing the dynamic characteristics of flow fields. A computational model of the reactor coolant pump was constructed, and numerical simulation of the internal flow field under non-uniform inflow conditions was conducted. The impact of the lower chamber of the steam generator on the pump’s inlet conditions was evaluated. The numerical simulation results were analyzed using DMD technology, extracting flow characteristics and revealing the main flow modes and dynamic behaviors in the flow field. The results demonstrate that the DMD technology can accurately capture the time-dynamic characteristics within the flow field, providing crucial insights for optimizing performance and preventing faults in the reactor coolant pump.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142527010","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":"On the Neutron Kinetics during a Promptcritical Accident in a Heavy Liquid Metal Fast Reactor and the Importance of Low-Energy Neutrons","authors":"","doi":"10.1016/j.anucene.2024.110975","DOIUrl":"10.1016/j.anucene.2024.110975","url":null,"abstract":"<div><div>Accidental scenarios that involve degradation of a fast reactor core and/or relocation of its fuel material call for particular attention with respect to promptcritical reactivity events. The dynamics of a power transient during such an event is governed by the Prompt Neutron Generation Time (PNGT), a parameter that is sensitive to the moderating power of the system. Since Heavy Liquid Metals (HLMs) have a boiling point that is higher than the melting point of stainless steel, the first degradation mechanism to occur in a Heavy Liquid Metal Fast Reactor (HLMFR) is likely to be the loss of structural material. This sequence holds the potential to create conditions for considerable spectrum softening and may thus have important implications for the value of the PNGT. In the framework of this study, a (postulated) complete absence of structural material in and around the reactor core of an HLMFR is demonstrated to lead to an increase in the PNGT by an order of magnitude.</div><div>The sensitivity of the fission energy release during a promptcritical event to the value of the PNGT is further investigated by employing the severe accident code SIMMER-III. To correctly model a degraded reactor core characterized by a considerably softer neutron spectrum when compared to the spectrum of its intact configuration, a new neutron data set is generated. This is done by introducing new energy groups to the already existing 11-energy-group structure and collapsing multigroup cross-section data by employing a weighting spectrum representative of a degraded core configuration of an HLMFR. Subsequent simulations demonstrate that an increase in the PNGT by a factor of ∼4 yields an increase in the fission energy release during a Core Disruptive Accident (CDA) by ∼50 %. It is therefore established that low-energy neutrons may play an important role during a promptcritical reactivity transient in an HLMFR.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142527014","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":"Research on simulation of hydrogen diffusion behavior based on CONTHAC-3D code","authors":"","doi":"10.1016/j.anucene.2024.111003","DOIUrl":"10.1016/j.anucene.2024.111003","url":null,"abstract":"<div><div>An in-house code called CONTHAC-3D was developed to investigate the fundamental thermal–hydraulic phenomena occurred in the containment under severe accidents for NPPs. The code included specific models to simulate the special systems of HPR1000 and ACP100. The classical backward-facing step flow benchmark and BMC HYJET helium jet experiments were selected to investigate the code’s capability of simulating hydrogen diffusion process. The results showed that the difference between the calculated and experimental results could be negligible. The code was then applied to investigate hydrogen diffusion and distribution for HPR1000. The results showed that the hydrogen released from the break rises vertically and rapidly to the containment dome, then the gas diffused into the dome and lower compartments. As the time went by, the hydrogen concentration in lower compartments seemed to be higher than that in the containment dome. The results could provide foundation for the arrangement of hydrogen risk mitigation measures.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142527102","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":"Machine-learned force fields for thermal neutron scattering law evaluations","authors":"","doi":"10.1016/j.anucene.2024.110978","DOIUrl":"10.1016/j.anucene.2024.110978","url":null,"abstract":"<div><div>A new method is presented to use machine-learned interatomic potentials (MLPs) to generate material models for thermal neutron scattering laws (TSLs). MLPs are computationally efficient models of <em>ab initio</em> force fields that can be used in the creation of a vibrational spectrum as an input to TSL generation. MLP-based molecular dynamics introduces temperature effects into the vibrational spectrum, which have been neglected in most modern TSLs. Yttrium hydride (<span><math><msub><mrow><mi>YH</mi></mrow><mrow><mi>x</mi></mrow></msub></math></span>) is used to illustrate this new MLP technique. The MLP approach is shown to predict temperature effects in the vibrational spectrum observed in experiment and improve on key features of the oscillatory scattering cross section of <span><math><msub><mrow><mi>YH</mi></mrow><mrow><mi>x</mi></mrow></msub></math></span> when compared to current temperature-independent, <em>ab initio</em> techniques.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142527101","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":"Neutron transmission measurements for silica glass at the KURNS-LINAC","authors":"","doi":"10.1016/j.anucene.2024.111017","DOIUrl":"10.1016/j.anucene.2024.111017","url":null,"abstract":"<div><div>Silica glass has been used as a base and host material in vitrified radioactive waste and lithium glass scintillators for neutron detection because of its superb transparency, high heat resistance, and excellent chemical inertness. Therefore, an accurate total cross section of the silica glass is crucial to evaluate the criticality safety of vitrified wastes and understand the neutron response for lithium glass scintillators. This study performed neutron transmission measurements for silica glass using a pulsed neutron beam with the time-of-flight method at the Kyoto University Institute for Integrated Radiation and Nuclear Science − Linear Accelerator to provide an accurate total cross section in the thermal and epithermal energy range. We obtained the neutron total cross section of the silica glass in the energy region from 0.002–25 eV. The results were compared and discussed with previous results and evaluated data.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142527011","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":"HZP and HFP rod ejection analysis in a SMART-like reactor model using the GUARDYAN-SUBCHANFLOW coupled code system","authors":"","doi":"10.1016/j.anucene.2024.110988","DOIUrl":"10.1016/j.anucene.2024.110988","url":null,"abstract":"<div><div>GUARDYAN is a dynamic 3D Monte Carlo reactor physics code with continuous energy handling developed for GPU hardware that has recently been coupled to the SUBCHANFLOW (SCF) subchannel thermal hydraulics solver. In this paper two control rod ejection accident scenarios will be presented in a Small Modular Reactor (SMR) geometry: a transient starting from Hot Zero Power (HZP), and one starting from Hot Full Power (HFP) conditions, both of them using Beginning of Cycle (BOC) material composition. Both the time dependent core-wise data and the node-wise data at certain times calculated by the GUARDYAN-SCF coupled code system exhibit the tendencies expected during such transients, with the thermal hydraulic properties mostly inside their safe limits. Relative variances estimated from 8 independent realisations suggest the results are credible. To further support our findings the HZP results are presented alongside data from PARCS-SCF and Serpent2-SCF calculations provided by Karlsruhe Institute of Technology (KIT), while for the HFP case we were able to compare some of the quantities to PARCS-SCF results.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142527013","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":"Measurement of gamma field inside the biological concrete shielding of VVER-1000 Mock-Up at the LR-0 reactor","authors":"","doi":"10.1016/j.anucene.2024.110999","DOIUrl":"10.1016/j.anucene.2024.110999","url":null,"abstract":"<div><div>The long-term operation of existing nuclear power reactors is a crucial concern due to the complexities and expenses associated with replacing key components, such as the reactor pressure vessel and reactor internals. Gamma radiation, a byproduct of nuclear reactions and radioactive decay, significantly influences the lifetime of these components. This radiation is responsible for various degradation pathways leading to void swelling in steel reactor components and cracking or other radiation damage in concrete structures.</div><div>A study conducted at a full-scale mock-up of the VVER-1000 reactor at the LR-0 zero-power reactor employed HPGe and stilbene measurements to analyze gamma spectra behind the reactor pressure vessel and within concrete biological shielding. While simulations behind the reactor pressure vessel aligned with measurements, notably, a<!--> <!-->marked overestimation of stilbene spectrum calculations occurred deep in concrete, suggesting potential inaccuracies in radiation predictions for power plant structures.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526955","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":"Investigation on radioisotopes evolution in the fuel of Lead-Bismuth eutectic (LBE) cooled SPARK-NC core","authors":"","doi":"10.1016/j.anucene.2024.110998","DOIUrl":"10.1016/j.anucene.2024.110998","url":null,"abstract":"<div><div>SPARK-NC, a 10 MW(e) lead–bismuth eutectic (LBE) cooled fast reactor design, exhibits promising characteristics like inherent gamma shielding, natural circulation, and a high boiling point. Following detailed neutronic studies, a thorough investigation of nuclear safety necessitates a detailed analysis of the core radionuclide inventory. This information is particularly crucial for source term calculations, which play a vital role in assessing the potential radiological consequences. This study establishes the life-cycle inventory of SPARK-NC using two independent computational systems: ORIGEN2.2 and NECP-SARAX. ORIGEN2.2, equipped with a reactor-specific library generated by NECP-MCX, is used for average whole-core inventory analysis. NECP-SARAX, on the other hand, explicitly considers core heterogeneity in terms of enrichment, specific power, and burn-up. This work presents the radionuclide inventories and the relative calculation differences observed between the codes. Actinides like uranium and curium display minimal code dependence, while plutonium isotopes exhibit a maximum relative difference of 8 %. Fission products generally agree within 5 %, except for I-131, which shows a discrepancy of around 10 %. The activity of I-131 and Cs-137 are estimated to be approximately 1 × 10¹⁶ Bq and 3 × 10¹⁵ Bq, respectively. Additionally, the photon source strength is 10¹⁷/s at 1 MeV, dropping to 10¹⁶/s below 6 MeV. Fission products and actinides contribute a decay heat of 0.65 MW. Assembly-wise analysis reveals a direct proportionality between radionuclide inventory and peaking factor, with the average assembly inventory being roughly 25 % lower than the peak assembly inventory. Rare earth elements (Ce, Sm, Pm, Pr, Nd, La, Y) exhibit a maximum mass of approximately 8.5 kg with a 3 % relative difference between the codes. Conversely, halogens (I, Br) have a minimum mass of around 0.2 kg with a 13 % relative difference. These findings, alongside the quantification of radionuclides, provide valuable insights into the code selection for future analyses of SPARK-NC and similar reactor systems.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526931","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}