Nuclear Engineering and Design最新文献

{"title":"Fuel performance code to code comparative analysis for the OECD/NEA MPCMIV benchmark","authors":"Quentin Faure ,&nbsp;Gregory Delipei ,&nbsp;Alessandro Scolaro ,&nbsp;Maria Avramova ,&nbsp;Kostadin Ivanov","doi":"10.1016/j.nucengdes.2024.113685","DOIUrl":"10.1016/j.nucengdes.2024.113685","url":null,"abstract":"<div><div>The recently developed Multi-Physics Pellet Cladding Mechanical Interaction Validation (MPCMIV) benchmark includes dedicated transient fuel performance exercises. In this work, three fuel performance codes (BISON, OFFBEAT, FAST) are used to perform supporting studies for the benchmark. The exercises consist of a three-year long base irradiation of a father rod in a boiling water reactor followed by a cold ramp transient for a fuel rodlet refabricated from the father rod. For the base irradiation, the results obtained are satisfactory in comparison to the measurements, with some discrepancies observed in the cladding outer diameter for OFFBEAT and BISON, which can be explained by the oxidation models implemented in both codes. Concerning the cold ramp, which consists of a very fast power increase with the linear heat rate going from zero to its maximal value in just a few seconds, all the codes tend to underpredict the cladding axial elongation temporal evolution. The observed discrepancies between predictions and measurements are both in the maximal amplitude and shape of the cladding axial elongation temporal evolution. This suggests that the phenomenology is not predicted accurately. Using a multi-physics coupling (Griffin, BISON, THM), involving reactor-physics, thermal–hydraulic, and fuel performance, the ramp is investigated and an estimation of the LHR is obtained. The OFFBEAT model is then updated with the new LHR. The cladding axial elongation is predicted with significant better agreement compared to the measurements. In single physics fuel performance modeling, the linear heat rate is obtained by calorimetric technique, which is not suited for fast transient, while in the multi-physics model, the linear heat rate is predicted instead by the multi-physics model and then is used as a source term in the fuel performance code. Analyzing further the obtained results, plastic strains mainly axially and with an amplitude one order of magnitude lower than the total strain at peak transient are observed on a small part of the cladding. Future work will focus on improving the BISON model by implementing frictional contact and to use the model for more multi-physics studies as well.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113685"},"PeriodicalIF":1.9,"publicationDate":"2024-11-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142661299","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":"Characterization and comparative appraisal of two novel annular channel designs with variable flow areas for supercritical heat transfer","authors":"Ashok Kumar Gond ,&nbsp;Tanuj Srivastava ,&nbsp;Amaresh Dalal ,&nbsp;Dipankar N. Basu","doi":"10.1016/j.nucengdes.2024.113679","DOIUrl":"10.1016/j.nucengdes.2024.113679","url":null,"abstract":"<div><div>Deterioration of heat transfer is a common concern for any supercritical heat transport system, regardless of the geometric orientation. Present study proposes two novel designs of annular channel with variable cross-sectional area and numerically assesses their respective performances, with the primary objective being the enhancement of overall heat transport characteristics in comparison with an equivalent traditional plain annular channel. Both the configurations exhibit substantial gain in terms of overall heat transfer coefficient and prevailing temperature level, while also eliminating deterioration over the entire parametric ranges explored here, with the diverging one demonstrating relatively superior characteristics. The converging channel generally maintains a flatter temperature profile and comparatively lower maximum temperature, and hence can be employed at larger power-to-mass-flux ratios. Taper angle is earmarked as the most influencing design variable, illustrating enhanced performance with greater tapering, albeit at the cost of nominal increase in pressure drop. Both the designs are found to be insensitive to flow acceleration, which is a primary reason of not encountering deterioration. Strong buoyancy effect can be present within the entrance region of the converging design, affecting its overall performance. Local thermalhydraulics have been noted to be contingent to the effective level of turbulence and distribution of specific heat in the radial direction, which also contribute to the favorable response from the diverging design.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113679"},"PeriodicalIF":1.9,"publicationDate":"2024-11-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142661301","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 analysis of Neutronics and Thermal-Hydraulic properties of Helical-Cruciform fuel assembly with Helium-Xenon gas mixture","authors":"Wenxuan Ju ,&nbsp;Kewei Ning ,&nbsp;Lin Xie ,&nbsp;Fulong Zhao ,&nbsp;Sichao Tan","doi":"10.1016/j.nucengdes.2024.113688","DOIUrl":"10.1016/j.nucengdes.2024.113688","url":null,"abstract":"<div><div>Helical-cruciform fuel is researched in new reactors due to its excellent thermal and hydraulic properties. A numerical simulation was conducted on the neutron, flow, and heat transfer characteristics of a 3 × 3 fuel assembly composed of helical-cruciform fuel with a Helium-Xenon (He-Xe) gas mixture. High-precision modeling of fuel rods was achieved through axial differential and geometric reconstruction, the power distribution characteristics were analyzed, the velocity and temperature distribution of HCF components were obtained, and the boundary layer properties were investigated. The calculation results show that the axial power density presents a cosine distribution. A significant difference appears in the velocity boundary layer thickness between the windward and leeward sides of the helical-cruciform fuel, the corresponding velocity of the viscous sublayer on the windward side is 1.38–2.72 times higher than that on the leeward side. A larger surface heat transfer coefficient appears on the windward side.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113688"},"PeriodicalIF":1.9,"publicationDate":"2024-11-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142661300","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":"Evaluation of probabilistic distribution of SCC growth rates obtained under the same test conditions in cold worked stainless steel","authors":"Dan Akazawa,&nbsp;Masato Koshiishi,&nbsp;Yasufumi Miura,&nbsp;Kenji Kako","doi":"10.1016/j.nucengdes.2024.113669","DOIUrl":"10.1016/j.nucengdes.2024.113669","url":null,"abstract":"<div><div>Probabilistic fracture mechanics (PFM) is a structural integrity assessment methodology that quantifies failure probability of components in nuclear power plants. In PFM analysis, probabilistic distributions expressed as probabilistic density functions are given to input parameters. This paper discusses the probabilistic distribution of SCC crack growth rates (CGRs) for cold worked Type 316L stainless steel in tests simulating a BWR environment. To assess the probabilistic distribution of these SCC CGRs, the associated data for 40 data of SCC CGRs were obtained from the single heat material under the same test conditions. Both normal and lognormal distributions were in agreement, and the standard deviation was much smaller than previously reported. The PFM results were strongly influenced by the standard deviation of the SCC CGR, suggesting that it is important to consider the SCC CGR probabilistic distribution for a reliable assessment of PFM.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113669"},"PeriodicalIF":1.9,"publicationDate":"2024-11-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142660872","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":"Improving operational flexibility of the integrated pressurized water reactor with the MED-TVC desalination system by control logic systems in the off-design mode","authors":"A. Naserbegi ,&nbsp;M. Aghaie ,&nbsp;Kh. Sadeghi ,&nbsp;S.H. Ghazaie ,&nbsp;E. Sokolova","doi":"10.1016/j.nucengdes.2024.113702","DOIUrl":"10.1016/j.nucengdes.2024.113702","url":null,"abstract":"<div><div>In many regions of the world, nuclear power plants play a vital role as an energy source for freshwater production. They can be coupled with various types of desalination plants to produce electricity and freshwater. In this paper, multi-effect desalination with thermal vapor compression, which is considered the most effective thermal desalination system, has been coupled with a pressurized water reactor. Thermodynamic modeling has been implemented based on energy analysis in the design and off-design modes of Thermoflex software, while economic modeling for freshwater production is conducted using DEEP software. To provide flexibility in freshwater production along with nuclear plant safety, appropriate control systems are incorporated at the design point with a certain degree of flexibility to ensure that variation in the main steam flow rate provides a steady energy source for the desalination system. The performance of the flow control valves was confirmed by applying flow fluctuations of ±30 % to the output steam from the steam generator. The design results showed that producing 15,000 m³/day at a cost of 1.232 $/m³ would reduce only 0.27 % of the net electrical efficiency in the base plant.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113702"},"PeriodicalIF":1.9,"publicationDate":"2024-11-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142660876","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":"Validation of assessment methods for creep crack growth rates in irradiated components from Nimonic PE16 Tie-Bar tests","authors":"Markian Petkov,&nbsp;Pierre-Alexandre Juan","doi":"10.1016/j.nucengdes.2024.113681","DOIUrl":"10.1016/j.nucengdes.2024.113681","url":null,"abstract":"<div><div>Creep crack growth is a life-limiting failure mechanism in high-temperature metallic components. The creep crack growth response is linked to the underlying creep deformation and failure properties, such as creep ductility. Irradiation effects such as helium (He) embrittlement in high-temperature reactor components trigger decreases in creep ductility. Creep crack growth (CCG) data obtained from unirradiated and irradiated Nimonic PE16 tie-bars confirm the accelerated crack growth rates under neutron irradiated conditions. The exact same data were also used to validate the analytical defect assessment procedure for transient creep crack growth in Code Case N-934 of ASME BPVC Section XI Division 2, developed from unirradiated material data. To do so, the mechanics-based creep crack growth law of Nikbin-Smith-Webster (NSW) is adopted to estimate the crack growth rates in unirradiated PE16, and then adjusted through changes in bulk mechanical properties in the irradiated case to estimate the changes in crack growth rates. The estimated <span><math><mrow><mover><mi>a</mi><mo>̇</mo></mover><mo>-</mo><msup><mrow><mi>C</mi></mrow><mrow><mo>∗</mo></mrow></msup></mrow></math></span> trend was employed as part of the analytical defect assessment procedure for transient creep crack growth in Code Case N-934 of ASME BPVC Section XI Division 2. Evaluation of results via the procedure and the estimated trends confirm the validity of the approach to predict increases in creep crack growth in neutron irradiated components, as observed in experimental data. The approach yields moderately conservative predictions for crack growth rates, and accurately captures the relative increase in creep crack growth rates for the irradiated case. The predictions were independently validated by predicting the C*-values expected in the loading configuration for a given recorded crack growth rate <span><math><mover><mi>a</mi><mo>̇</mo></mover></math></span>. The validation of the defect assessment method introduced in Code Case N-934 against both unirradiated and irradiated PE16 CCG data provides a practical path to its implementation in metallic components across high-temperature reactor designs. This is particularly useful where crack growth data in irradiated material may not be readily available. The technique also allows for propagating uncertainties in crack growth predictions which stems from compounding effects such as variability in unirradiated properties, degree of irradiation and corresponding embrittling effects, crack size and load perturbations. The relevance of mechanical properties accurately depicting aspects of the He embrittlement process for practical applications in plant and the importance of creep crack growth testing are discussed.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113681"},"PeriodicalIF":1.9,"publicationDate":"2024-11-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142660855","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":"Comparison of four strategies for separation of stable isotopes of iridium element by gas centrifugation method","authors":"Seyedeh Leila Mirmohammadi,&nbsp;Jaber Safdari,&nbsp;Mohammad Hasan Mallah","doi":"10.1016/j.nucengdes.2024.113671","DOIUrl":"10.1016/j.nucengdes.2024.113671","url":null,"abstract":"<div><div>This article presents four strategies for separating two stable isotopes of iridium with 99.9% purity. The cascades employed are tapered, square, and squared-off. To design and optimize these cascades, four computational codes were developed using the PSO algorithm. The findings indicate: (1) The tapered cascade strategy cannot achieve 99.90% separation for either type of centrifuge; (2) For the first centrifuge, both square and squared-off cascades strategy successfully separate the isotopes, achieving 99.90% purity, with recovery percentages of ¹⁹¹Ir and ¹⁹³Ir at 99% and 33%, respectively; (3) For the second centrifuge, only the square cascade strategy achieves 99.90% separation, with recovery percentages of approximately 99% for both isotopes.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113671"},"PeriodicalIF":1.9,"publicationDate":"2024-11-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142660856","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 tip injection on performance of highly loaded helium compressor in high-temperature gas-cooled reactor","authors":"Xinle Wang ,&nbsp;Zhitao Tian ,&nbsp;Adil Malik ,&nbsp;Yingqi Fan ,&nbsp;Hai Zhang ,&nbsp;Huawei Lu","doi":"10.1016/j.nucengdes.2024.113683","DOIUrl":"10.1016/j.nucengdes.2024.113683","url":null,"abstract":"<div><div>As one of the representatives of the fourth-generation advanced nuclear reactor, the continuous development and application of high-temperature gas-cooled reactor (HTGR) technology has promoted the technical progress of the whole nuclear energy field. Helium compressor is one of the core components of HTGR. The highly loaded helium compressor effectively solves the disadvantages of the low single-stage pressure ratio and numerous stages of the traditional helium compressor. But it also brings a more complex tip-leakage flow. Tip injection, which is an active control method, can effectively control the tip clearance leakage and inhibit the development of the leakage vortex. This paper analyses the effects of axial deflection angle and injection pitch angle on the rotor performance of a highly loaded helium compressor via numerical simulation and validates them via experiment. Results show that the leakage vortex can be blown to the pressure surface of the adjacent blade by proper axial deflection angle to reduce the vorticity of the leakage vortex. The injection pitch angle directly affects the intensity of the leakage vortex during its initiation and development. When the axial deflection angle is 60° and the injection pitch angle is 20°, the adiabatic compression efficiency and total pressure ratio increase by 0.554 % and 0.160 % respectively under the design condition, and by 0.822 % and 0.162 % respectively under near-stall conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113683"},"PeriodicalIF":1.9,"publicationDate":"2024-11-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142660854","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 single-flow method for efficient production of Cf-252 in high-flux fast reactor","authors":"Yu Xin ,&nbsp;Qingquan Pan ,&nbsp;Lianjie Wang ,&nbsp;Bangyang Xia ,&nbsp;Yun Cai ,&nbsp;Xiaojing Liu ,&nbsp;Jinbiao Xiong","doi":"10.1016/j.nucengdes.2024.113680","DOIUrl":"10.1016/j.nucengdes.2024.113680","url":null,"abstract":"<div><div>Targets are irradiated in high-flux reactors to produce transplutonium isotope. High neutron flux favors the production of transplutonium isotope. Fast reactors can achieve a higher neutron flux operating at the same power density, with the advantages of high neutron energy, controllability of energy spectrum and large irradiation volume. We investigate transplutonium isotopes production in High-Flux Fast Reactor (HFFR) focusing on ²⁵²Cf, proving that fast reactor has better production economy. Thermalized neutron spectrum promotes the production of transplutonium isotopes. Constructing a thermal neutron environment in fast reactor produces a significant quantity of thermal neutrons, and the thermal neutron overflow from the irradiation channel results in an increased fission power in the fuel region. We proposed a neutron single-flow method that restricts thermal neutrons in the irradiation channel from entering the fuel region and achieves a partition distribution of neutrons. This method has achieved a maximum reduction of 83% in fission power of fuel region around the irradiation channel without affecting neutron flux in the irradiation channel. The neutron single-flow method offers technical support for producing transplutonium isotopes in fast reactors.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113680"},"PeriodicalIF":1.9,"publicationDate":"2024-11-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142660853","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 numerical study on metallic melt infiltration in porous media and the effect of solidification","authors":"Liang Chen,&nbsp;Yan Xiang,&nbsp;Di Fang,&nbsp;Weimin Ma","doi":"10.1016/j.nucengdes.2024.113687","DOIUrl":"10.1016/j.nucengdes.2024.113687","url":null,"abstract":"<div><div>The melt infiltration in porous debris is of importance to severe accident prediction and mitigation in nuclear power plants (NPPs), but its mechanism remains elusive. In this study, a computational fluid dynamics (CFD) model is proposed to simulate the evolution of melt infiltration within porous media, incorporating both solidification and melting processes. The CFD model is validated against the experiment (REMCOD facility) and Moving Particle Semi-implicit (MPS) simulation results. Building upon this validated model, the influence of the melt superheat, the initial particle temperature, and its surface wettability on melt infiltration dynamics are mainly analyzed. It is found that increased initial melt superheat enhances melt infiltration length and rate; higher initial particle temperatures promote deeper and faster infiltration, while lower temperatures may result in solidification that blocks further infiltration. Additionally, the wettable particulate bed can enhance melt relocation and heat transfer, but it also accelerates the solidification of the melt, which complicates the infiltration process. Furthermore, phase changes could intensify melt flow instability. This work may expand our understanding of melt infiltration dynamics and pave the way to severe accident modeling in NPPs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113687"},"PeriodicalIF":1.9,"publicationDate":"2024-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142660852","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}