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Nuclear engineering and design
North-Holland Pub. Co.
Nuclear engineering and design

North-Holland Pub. Co.

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0029-5493

Nuclear engineering and design/Journal Nuclear engineering and designSCIISTPEI
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    Flow distribution in the primary circuit of a fast reactor: Impact with reduced number of subassemblies in the core

    Vikram, G.Chauhan, Amit KumarRajendrakumar, M.Natesan, K....
    1.1-1.14页
    查看更多>>摘要:The primary circuit of a typical pool-type sodium-cooled fast reactor (SFR) is a complex flow network with multiple flow paths. The resistances of these flow paths vary significantly from very low to very high values. Estimation of flow fractions in these paths is essential for the design and analysis of various components in the primary circuit under various operating conditions. In the present work, the primary circuit of a typical mediumsized pool-type SFR has been modeled using the Flownex code, a commercial system dynamics code. The steadystate flow distribution in the primary circuit has been first studied and an overall flow balance has been established. Out of the total flow supplied by the primary pumps, 91 % flows through the core, and 93 % flows through the Intermediate Heat Exchanger (IHX). Notably, in the storage locations of the core, there is no leakage from the Grid Plate (GP, the structure on which the core subassemblies (SAs) are supported) to the bottom Core Support Structure (CSS) plenum. Instead, flow is in the reverse direction due to the high resistance offered by the sleeve holes in the storage locations. Then, the effect of removing SAs from the fuel and storage locations of the core has been studied. The empty sleeves in the GP were modeled in 3D using Ansys (R) Fluent and have been coupled with the Flownex model of the primary circuit. Two different flow configurations were observed in the empty GP sleeves when the fuel SAs were removed and when the storage SAs were removed. When fuel SAs are removed, the sodium flows downwards in the empty sleeve bottom opening. However, when storage SAs are removed, the sodium flows upwards in the empty sleeve bottom opening. This happens because of the high hydraulic resistance offered by the storage SA sleeve holes. As a result, when fuel SAs are removed, the flow rates in the paths fed by the CSS plenum (main vessel cooling system path, shielding SA flow path, etc.) increase. When storage SAs are removed, the flows in these paths decrease. There is no significant change in the pump operating point when a single fuel or storage SA is removed from the core. However, when more fuel SAs are removed from the core, a change in the pump operating point is observed. For instance, the pump flow decreases by 4.1 % when seven fuel SAs are removed from the core.

    Review on flow and heat transfer processes in rod bundle channels of water-cooled nuclear reactor

    Du, LipengChen, XiangZhang, WenchaoSun, Jianchuang...
    1.1-1.24页
    查看更多>>摘要:The thermal-hydraulic analysis of coolant flow in the reactor core plays a significant role in the optimized design of nuclear fuel assemblies and nuclear safety. This article reviews the experimental and Computational Fluid Dynamics (CFD) methods for the thermal-hydraulic characteristics of subchannels with water as the coolant in different rod bundle geometries and flow conditions over the past few decades. It summarizes the effects of design parameters such as rod spacing and spacer grid type, as well as flow parameters like pressure, mass flow rate, and heat flux on fluid flow and heat transfer. Additionally, various models used for numerical simulations of rod bundle channels are introduced, providing valuable references for research and practice in related fields.

    Investigation of radiation and water impermeability attitudes of heavy concrete produced with hematite and chromite for nuclear reactor shielding structure

    Unal, Ibrahim HakanDurmus, Gokhan
    1.1-1.14页
    查看更多>>摘要:In this study, two different high performance heavy concrete (HPHC) (C40, C50) were produced using three different aggregates: limestone, hematite, and chromite. During fabrication of HPHC, 18 different concrete mixtures were obtained by using 3 different aggregate size (0-4 mm, 4-8 mm, 8-16 mm) in accordance with TS EN 802, 2016. The water/cement ratios were maintaned at 0.45 and 0.44 for C40 and C50 samples, respectively. Cube samples, 150 x 150 x 150 mm3 in size, and plate samples, 200 x 200 x 50 mm3 in size were produced. Ultrasonic pulse velocity (UPV), uniaxial compressive strength (UCS), water permeability and radiation attenuation tests were performed on these samples. In addition, the radiation shielding and water impermeability properties of HPHC in the nuclear reactor shield structure were investigated. According to the test results, the highest UCS in C40 samples is 63.4 MPa, attenuation coefficients are 0.23 cm-1 and 0.25 cm-1 for gamma, 0.16 cm- 1 , 0.17 cm- 1 and 0.21 cm- 1 for neutron in HHH-40-1. The maximum water absorption was detected in CCC40-2 with 12.3 mm. In C50 samples, the highest UCS is 72.6 MPa, attenuation coefficients are 0.25 cm- 1 and 0.3 cm- 1 for gamma, 0.16 cm- 1 , 0.19 cm- 1 and 0.23 cm- 1 for neutron in the HHH-50-1. The maximum water absorption was detected in CCC-50-2 with 10 mm. When all the data were examined, HHH-40 mixture demostrates the best value in UCS and radiation shielding, CCC-40 mixture gave the best value in water impermeability in the C40 class. The HHH-50 mixture demostrated the highest values in UCS and radiation shielding, and the CCC-50 mixture achieved the best values in water impermeability within the C50 class. Consequently, hematite mixtures were characterized by UCS and radiation shielding, while chromite mixtures were characterized by water impermeability.

    A comprehensive analysis of neutronic properties of annular dispersed particle fuel

    Li, SongLiu, LeiZhang, YongfaHao, Jianli...
    1.1-1.14页
    查看更多>>摘要:This work made a comprehensive analysis of the neutronic performance of annular dispersed particle fuel (ADF), and the results were compared with annular ceramic fuel (ACF), cylindrical dispersed particle fuel (CDF), and cylindrical ceramic fuel (CCF). The four types of cells are simulated by in-house code, and then the effective multiplication factor, neutron flux distribution, depletion characteristics, nuclides composition variation, and temperature coefficient of each fuel rod are compared and analyzed. Moreover, the characteristics of fuel assembly consisting of the four types of fuels are analyzed, including the effective multiplication factor, normalized pin power distribution, etc. The calculation results show that ADF has obvious advantages in all the indicators mentioned above. The calculation and analysis conducted in this paper could compare the advantages of the new annular dispersed particle fuel rod over the traditional fuel rod, which provides a certain reference for the application of the new fuel rod in engineering.

    Comparative radiological impact of LOCA and RDD scenarios: An AI-enhanced assessment using HotSpot code

    Maglas, Najeeb N. M.Najar, MerouaneQiang, ZhaoAli, Mohsen M. M....
    1.1-1.13页
    查看更多>>摘要:This study provides a comprehensive radiological assessment of two hypothetical incidents: a Loss of Coolant Accident (LOCA) at a nuclear reactor and a Radiological Dispersal Device (RDD) detonation, both simulated in Dhamar City, Yemen, using the HotSpot Health Physics Code. We evaluated the dispersion of radioactive materials under consistent atmospheric conditions to assess their environmental and human health impacts. Our analysis was based on two parameters: sampling time and exposure duration. For sampling time, the Total Effective Dose Equivalent (TEDE) was measured at specific intervals. After 2000 min, the TEDE for LOCA was 47 Sv, significantly higher than 0.0033 Sv for the RDD within a 1 km2 area. In the initial moments of the explosion, the doses were 340 Sv for LOCA and 0.042 Sv for RDD, showing a dramatic decrease over time. For exposure duration, the LOCA scenario, results in a TEDE of 150 Sv after one year. In contrast, the RDD leads to a TEDE of 0.17 Sv after the same period. The LOCA scenario results in higher radiation doses due to multiple radionuclides with varying decay rates, causing a rapid increase in dose. In contrast, the RDD scenario shows a slower dose accumulation due to the long half-life of 137 Cs. This study introduces an AI-enhanced approach to radiological assessments of LOCA and RDD incidents, using an Artificial Neural Network (ANN) model comprised of classification and regression sub-models. The classification sub-model accurately identifies the nature of the radiation event, while the regression sub-model estimates the distance of the explosion within 80 km radius from the explosion epicenter. With a predictive accuracy of 100 % in classification and over 99 % in regression, the model significantly improves the effectiveness and speed of emergency response strategies, offering critical advancements in radiological safety measures. The impact on human organs was more severe in LOCA, with doses to the liver, skin, lungs, thyroid gland, brain, and kidneys exceeding those from the RDD by factors ranging from 55 to 6000. The findings stress the need for strong safety measures, long-term monitoring, and preparedness, especially in regions like Yemen, while highlighting the potential long-term environmental and health impacts of nuclear incidents and the importance of effective response and recovery plans.

    Impact assessment of the integration of a generic PEM electrolyser facility into Rivne nuclear station

    Diaz-Pescador, EduardViebach, MarcoGamaleja, FlorianLippmann, Wolfgang...
    1.1-1.15页
    查看更多>>摘要:This manuscript presents the outcomes from the impact assessment applied to the onsite integration of a 40 MW hydrogen production plant (HPP) into the protected area of Rivne nuclear power plant (NPP) in Ukraine. The scope of this work is limited to frequency estimation and blast loading characterization applied to three cases of study, representative of the worst-case scenario at different HPP locations. In the framework of the Euratom NPHyCo project, the NPP operator has proposed integration locations and provided corresponding distances to nearby structures with corresponding fragility criteria. The presented study assumes HPP operation at full capacity with allocated NPP electricity as a feedstock. Hazardous points within the HPP are identified through HAZID (HAZard IDentification) methodology, which is assisted with system reliability analysis based on component failure rates. The identification of accident sequences in the HPP and frequency estimation is accomplished by means of event tree analysis (ETA). The study identifies as the worst-case scenario within the HPP a hydrogen explosion due to the destructive potential over far-range distances via shock wave propagation. In the electrolyser system, an unintended release of hydrogen may trigger a vapour cloud explosion (VCE), whereas in the adjacent hydrogen buffer tank a physical explosion may trigger a shock wave with subsequent projectile generation. The results show that the highest frequency of a hydrogen VCE is in the order of 10- 7 per year following a 1% leak in the separator vessel upstream flange within the gas processing area. The blast load characterization shows that a hydrogen VCE within the electrolyser facility would not exceed the 10 kPa fragility criterion of nearby safety-related structures. Contrarily, a physical explosion in the 30 kg buffer tank may lead to projectile generation with the potential to reach the standby diesel generators, cooling towers, and turbine building at a distance up to 500 m. The vessel fragments are deemed as soft missiles with high deformability upon impact. Nevertheless, means of protection are proposed to reduce the risk of the coupled facility. The outcomes of the study are meant as recommendations for a subsequent comprehensive safety assessment by the NPP operator.

    Assessment of critical heat flux (CHF) on irradiated and non-irradiated nano-composite surfaces

    Rahimian, ArefPorhemmat, Mohammadhadi
    1.1-1.13页
    查看更多>>摘要:This study employs an electrophoretic deposition (EPD) technique to fabricate a uniform nano-composite thin film coating on boiling thin steel plates. Two primary methodologies for creating composite nano coatings are explored: the simultaneous method and the sequential method, each comprising three distinct modes. To assess the impact of gamma irradiation on critical heat flux (CHF), test specimens were irradiated in a gamma cell at doses ranging from 100 to 300 kGy, followed by scanning electron microscopy (SEM) and Brunauer-EmmettTeller (BET) analysis. Contact angle and capillary length measurements were conducted for each coated specimen. Subsequently, the specimens underwent testing in a boiling pool to determine CHF and boiling heat transfer coefficients. The results indicate that both nano-composite coating and gamma irradiation significantly reduce the maximum pore diameter while enhancing porosity, pore surface area, and pore volume. Among the coating techniques, the sequential method with a double ratio of the outer to inner layer demonstrated superior performance in CHF enhancement. Notably, the CHF of the irradiated TiO2-ZrO2 nano-composite coated plate at 300 kGy increased from 1646 to 2258 kW/m2, representing a 37 % improvement. This enhancement in CHF is attributed to increased capillary effects resulting from the structural modifications induced by the coating and irradiation processes.

    Development of numerical code for an in-depth energy, exergy, exergoeconomic (3-E) assessments, and sensitivity analysis of NS Savannah marine propulsion: A pre-optimization-focused approach

    Delgarm, NavidVarnousfaaderani, Mahmoud RostamiFarrokhfal, HamidArdeshiri, Sajad...
    1.1-1.32页
    查看更多>>摘要:The maritime sector heavily relies on heavy fuel oil as its primary energy source for ship propulsion, which has a significant adverse impact on the environment, underscoring the urgent need for clean, eco-friendly propulsion alternatives. This paper aims to develop an in-depth thermoeconomic modeling of a real-world nuclear marine propulsion system, employing the nuclear ship Savannah as the baseline benchmark. Given the limited availability of data, this study utilizes innovative ideas to precisely characterize the thermodynamic properties and energy flows within the nuclear propulsion, enabling a comprehensive exergoeconomic performance assessment. Upon validating the developed nuclear ship Savannah propulsion model, an extensive analysis is undertaken from the energy, exergy, and exergoeconomic viewpoints, combining the principles of the first and second laws of thermodynamics with the specific exergy costing technique. The nuclear propulsion model is subsequently integrated with both local sensitivity analysis and global sensitivity analysis to examine how output variables respond to variations in different design parameters. Four key thermoeconomic performance indexes of nuclear propulsion including the energy efficiency, exergy efficiency, propulsion power, and total product exergy cost rate are considered as system output variables. The study employs a one-at-a-time approach for local sensitivity analysis and utilizes the variance-based Sobol method for global sensitivity analysis. Within the local sensitivity analysis framework, a novel indicator, termed "dispersion sensitivity index" is introduced to precisely quantify the overall sensitivity of outputs to inputs. This is subsequently compared with the total sensitivity index obtained from the global sensitivity analysis. The energy analysis demonstrates that the high-pressure and lowpressure steam turbines achieve mechanical power outputs of 6.881 MW and 8.209 MW, respectively, with the overall nuclear propulsion efficiency determined to be 26.18 %. The high-pressure steam generator is identified as the primary source of exergy destruction, with a value of 7161.03 kW, while the condensers exhibit the lowest exergy efficiency, around 30.68 %. Additionally, the exergoeconomic evaluation highlights that the high-pressure steam generator bears the highest exergy costs for both fuel and product, at $1490.20 and $1600.70 per hour, respectively, and the highest total operational cost of $110.60 per hour. The sensitivity analysis reveals that the steam flow rate at the high-pressure turbine inlet exerts the greatest influence on energy efficiency, exergy efficiency, and propulsion power with total sensitivity index of 28.5 %, 42.2 %, and 50.34 %, respectively. Conversely, the heat transfer surface area of high-pressure steam generator has the most significant effect on total product exergy cost rate, with a substantial total sensitivity index of 59.34 %. The integration of sensitivity analysis with exergoeconomic modeling of nuclear marine propulsion enables the identification of critical design parameters that have a substantial impact on system performance. This approach facilitates targeted improvements in energy and exergy efficiency, propulsion power, and economic costs. Furthermore, the analysis assesses system robustness by evaluating the variability of outputs with respect to parameter changes, thereby prioritizing optimization efforts to achieve maximum efficiency and cost-effectiveness while minimizing the need for trial-and-error iterations.

    Numerical and experimental study of shut-off rod assembly of a typical Indian research reactor under multi-support seismic excitation

    Kiran, A. RaviAgrawal, M. K.Sinha, S. K.
    1.1-1.14页
    查看更多>>摘要:The Shut-off Rod (SoR) assembly is a critical safety-related system in the majority of nuclear reactors. For its seismic qualification, it is required to ensure that the Shut-off Rod falls freely through the guide tube in the stipulated time under seismic excitation. Due to this stringent requirement, full-scale testing is usually performed for seismic qualification. The present work demonstrates the methodology for the evaluation of functionality and integrity of SoR assembly under multi-support seismic excitation. Experimental and numerical studies are carried out on the SoR assembly of a typical Indian Research Reactor under multi-support seismic excitation. In the present work, the adequacy of numerical simulation for seismic qualification of the SoR assembly is explored. The present study provides two conditions to ensure the sufficiency of numerical simulation for the seismic qualification of the SoR assembly. The first condition is to ensure the structural integrity of the lattice guide tube and the second one is to ensure the free fall of the SoR. This paper provides a comprehensive overview of both experimental and numerical results, along with the sufficiency of numerical simulation for seismic qualification of the SoR assembly.

    Effect of fouling deposition on thermal performance degradation of steam generators

    Jing, LiuZhen, LiZhenqin, XiongYuan, Gao...
    1.1-1.16页
    查看更多>>摘要:To investigate the deposition and distribution of Fe3O4 and its impact on the thermal performance of steam generators (SG), a three-dimensional porous medium numerical simulation model for SG is developed coupling thermal hydraulics and fouling deposition. The thermal power predicted agrees well the designed value with error 0.2 % under no fouling condition. The fouling deposition model considering the variable porosity and aging effect are addressed. The reduction of thermal power is smaller when the porosity of fouling is increased. The thermal power predicted agrees best with the actual power after 5 years operation when the average porosity of fouling is 0.6. After the 5-year operation, the portion of fouling thermal resistance ranges from 1.4 % to 16.3 % on the hot leg, while it ranges from 0.14% to 4.8 % on the cold leg. The proportion of fouling thermal resistance increases with the operation time, especially the hot side. The maximum proportion of fouling thermal resistance reaches 23.7 % and 8.4 % for hot leg and cold leg, respectively. The fouling accumulation on the hot leg is significantly higher compared to the cold leg which impairs the heat transfer performance on hot leg but this phenomenon is alleviated on the cold side due to increase of primary fluid temperature on cold leg. The simulation model established in this study can be utilized for the study of fouling characteristics in steam generators and prediction of their impact on thermal performance.