Nuclear Engineering and Design最新文献_第5页

Preparation and performance research of TRISO particle with niobium layer

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-02-03 DOI: 10.1016/j.nucengdes.2025.113896

Guoqiang Wang , Lihua Guo , Linyuan Lu , Feng Zhang , Xu Yang , Jun Lin

{"title":"Preparation and performance research of TRISO particle with niobium layer","authors":"Guoqiang Wang , Lihua Guo , Linyuan Lu , Feng Zhang , Xu Yang , Jun Lin","doi":"10.1016/j.nucengdes.2025.113896","DOIUrl":"10.1016/j.nucengdes.2025.113896","url":null,"abstract":"<div><div>In this study, a niobium (Nb) layer was prepared by spouted-bed chemical vapor deposition (CVD) to replace the outer pyrolytic carbon (OPyC) layer in TRISO particles as a novel configuration for enhancing fuel performance. Such a configuration in TRISO particles is favorable for the following reasons: Nb has good mechanical properties, which can relieve the tensile stress on the silicon carbide (SiC) layer better with applied compressive stress compared with the OPyC layer. Furthermore, owing to its better thermal conductivity and good contact with the SiC layer, the Nb layer in the TRISO particle can transport the heat generated from the kernel to the coolant in a more efficient manner. In addition, the Nb layer is believed to protect SiC from mechanical damage better with its ductility behavior during the fuel element fabrication, while the OPyC layer is a brittle material. The deposited Nb layer has a thickness of 10 μm, indicating the designed characteristics such as high density and a pure phase. The interface between the SiC and Nb layers exhibits tighten bonding without any visible debonding or gaps. The nanoindentation results suggest that the average hardness of the Nb layer is approximately 5 GPa and the average Young’s modulus is about 110 GPa, which are remarkably higher than the OPyC layer. Simultaneously, the stress and failure fractions for both designs are calculated. The new configuration confirms that the Nb layer exerts a better compressive stress effect on the SiC and thus the failure fraction of fuel particles can be lowered to some extent. In addition, finite element calculations indicate that the application of the Nb layer in TRISO particle is conducive to reducing the peak temperature by about 50 K. Based on the above advantages, the performance of the reactor can be efficiently improved.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"433 ","pages":"Article 113896"},"PeriodicalIF":1.9,"publicationDate":"2025-02-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143169296","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}

引用次数: 0

NEAMS IRP challenge problem 2: Thermal striping of reactor Internals

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-02-03 DOI: 10.1016/j.nucengdes.2025.113879

Emilio Baglietto , John Acierno , Annalisa Manera , Quynh M. Nguyen , Victor Petrov , Monica Pham , Yu-Jou Wang , Yue Jin , Jinyong Feng , Wayne Strasser , Dillon Shaver , Elia Merzari

{"title":"NEAMS IRP challenge problem 2: Thermal striping of reactor Internals","authors":"Emilio Baglietto , John Acierno , Annalisa Manera , Quynh M. Nguyen , Victor Petrov , Monica Pham , Yu-Jou Wang , Yue Jin , Jinyong Feng , Wayne Strasser , Dillon Shaver , Elia Merzari","doi":"10.1016/j.nucengdes.2025.113879","DOIUrl":"10.1016/j.nucengdes.2025.113879","url":null,"abstract":"<div><div>Oscillatory mixing of non-isothermal liquid coolant streams in advanced reactors can lead to thermal fatigue damage to fuel and reactor components. The NEAMS IRP CP2 has been seeking the development of an accurate, yet computationally affordable, turbulence modeling option for thermal-striping predictions. Efficient mixing of coolant streams in upper internal structures, lower plena, and heat exchangers significantly impacts the design of these systems, as well as their operation, and maintenance. This challenge problem generalizes specific needs related to the TerraPower and General Atomics designs by developing a set of benchmarks to advance and quantify the accuracy of the thermal striping modeling approach. The focus of the activities is to advance and demonstrate a modeling practice capable of accurately representing the performance of the structural reactor components under the influence of thermal striping. Assessment against adiabatic quasi-2D jets striping has demonstrated great promise for the proposed turbulence approaches, with 2 orders of magnitude acceleration from the reference LES solution. More recent efforts have extended the quasi-2D validation to diabatic conditions, leveraging the demonstrated accuracy of the highly resolved LES methods, and with a new set of experimental data for heated parallel round jets. Further upscaling through the use of reduced order models is being evaluated, and a two-step machine learning approach has been demonstrated on thermal striping for 3 parallel jets.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"433 ","pages":"Article 113879"},"PeriodicalIF":1.9,"publicationDate":"2025-02-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143169289","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}

引用次数: 0

Power reactor noise simulation and analysis by developing Time-Domain neutron noise Simulator: iPWR case study

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-02-02 DOI: 10.1016/j.nucengdes.2025.113894

Ali Kolali, Mohsen Ghafari, Naser Vosoughi

{"title":"Power reactor noise simulation and analysis by developing Time-Domain neutron noise Simulator: iPWR case study","authors":"Ali Kolali, Mohsen Ghafari, Naser Vosoughi","doi":"10.1016/j.nucengdes.2025.113894","DOIUrl":"10.1016/j.nucengdes.2025.113894","url":null,"abstract":"<div><div>In nuclear industry, the main challenges of nuclear power plants are to improve safety and reduce maintenance periods which lead to optimized operation costs. These challenges are also present for new-generation power plants especially small modular reactors (SMR). Implementation of online condition monitoring methods such as power reactor noise analysis that investigates symptoms caused by various perturbations in the in-core sensors can be useful to minimize these challenges. In this study, a time-domain power reactor noise analysis simulator will be presented to investigate perturbations in power reactors. The simulator is structured based on the nodal expansion method to discretization of neutron noise equations in two energy groups and three-dimensional rectangular space, whereas the implicit-difference discretization approach is employed to treat the time evolution. The developed steady-state module is verified by evaluating the large-scale and small-scale benchmark problems, and comparing with the PARCS code results. To verify the outcome, a noise analysis is also carried out for both the KWU and SMART cores, and the results are compared against the estimates by the CORE SIM and SD-HACNEM frequency-domain power reactor noise simulators. Simulation and analysis have been conducted for perturbations including control rod random vibrating and vibration with combining frequencies. The approach proved superior to the conventional frequency-domain simulators over the advantage of random perturbation simulation without removing part of the in-core sensor response. Furthermore, the results show that the framework is specifically suited for dealing with perturbation sources including random vibrating of the absorbers and absorbers with variable strength.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"433 ","pages":"Article 113894"},"PeriodicalIF":1.9,"publicationDate":"2025-02-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143169290","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}

引用次数: 0

Development of a coarse-mesh subchannel CFD model for prediction of core thermal–hydraulics in natural circulation conditions

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-02-01 DOI: 10.1016/j.nucengdes.2024.113738

Erina Hamase , Yasuhiro Miyake , Yasutomo Imai , Norihiro Doda , Ayako Ono , Masaaki Tanaka

{"title":"Development of a coarse-mesh subchannel CFD model for prediction of core thermal–hydraulics in natural circulation conditions","authors":"Erina Hamase , Yasuhiro Miyake , Yasutomo Imai , Norihiro Doda , Ayako Ono , Masaaki Tanaka","doi":"10.1016/j.nucengdes.2024.113738","DOIUrl":"10.1016/j.nucengdes.2024.113738","url":null,"abstract":"<div><div>For the safety enhancement of a pool-type sodium-cooled fast reactor (SFR) developed in Japan, the core-cooling performance of natural circulation (NC) decay heat removal systems with a dipped-type direct heat exchanger (D-DHX) installed in the hot pool of a reactor vessel (RV) has been investigated. During the D-DHX operation in the NC conditions, core–plenum interactions reduce the maximum core temperature, comprising of the penetration of the coolant from the D-DHX into assemblies and a narrow gap between them (interwrapper flow), and a heat transfer through a wrapper tube among assemblies (radial heat transfer). A computational fluid dynamics (CFD) analysis with a dense mesh arrangement can precisely predict these three-dimensional phenomena in the RV. However, RV modeling using the CFD code (RV-CFD) with a relatively coarse-mesh arrangement under a reasonable computational load has an advantage in SFR design studies. Focused on assembly modeling, the CFD model was developed based on a subchannel analysis with meshes for each subchannel (subchannel CFD (SC) model). Then, the RV-CFD with the SC model was validated by analyzing a sodium experiment in a sodium test facility, PLANDTL-1. In this study, to achieve a lower computational cost while maintaining the prediction accuracy for the design study, we developed a coarse-mesh subchannel CFD (CMSC) model for various velocities, including a rated operation to NC conditions. The CMSC model is applied to the core of the RV-CFD in the PLANDTL-1 analysis. Two numerical simulation cases are performed with and without the D-DHX operation under NC conditions. One case focused on radial heat transfer, whereas the other focused on core–plenum interactions. The result shows that the RV-CFD with the CMSC model applies to the core–plenum interactions during the D-DHX operation under NC conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113738"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167283","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}

引用次数: 0

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

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-02-01 DOI: 10.1016/j.nucengdes.2024.113822

Aref Rahimian, MohammadHadi Porhemmat

{"title":"Assessment of critical heat flux (CHF) on irradiated and non-irradiated nano-composite surfaces","authors":"Aref Rahimian, MohammadHadi Porhemmat","doi":"10.1016/j.nucengdes.2024.113822","DOIUrl":"10.1016/j.nucengdes.2024.113822","url":null,"abstract":"<div><div>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-Emmett-Teller (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 TiO<sub>2</sub>-ZrO<sub>2</sub> nano-composite coated plate at 300 kGy increased from 1646 to 2258 kW/m<sup>2</sup>, 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.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113822"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167578","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}

引用次数: 0

Characterization of neutrons emitted by an expected small amount of fuel debris in a trial retrieval from Fukushima Daiichi Nuclear Power Station

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-02-01 DOI: 10.1016/j.nucengdes.2024.113791

Taichi Matsumura , Keisuke Okumura , Masahiro Sakamoto , Kenichi Terashima , Eka Sapta Riyana , Kazuhiro Kondo

{"title":"Characterization of neutrons emitted by an expected small amount of fuel debris in a trial retrieval from Fukushima Daiichi Nuclear Power Station","authors":"Taichi Matsumura , Keisuke Okumura , Masahiro Sakamoto , Kenichi Terashima , Eka Sapta Riyana , Kazuhiro Kondo","doi":"10.1016/j.nucengdes.2024.113791","DOIUrl":"10.1016/j.nucengdes.2024.113791","url":null,"abstract":"<div><div>Retrieving objects with a small amount of fuel debris, such as a few grams, will begin soon at the Fukushima Daiichi Nuclear Power Station (1F) at the start of decommissioning. Objects retrieved from the primary containment vessel are not necessarily fuel debris; fuel debris is an object from which neutrons are emitted because it contains nuclear-fuel material. However, the characteristics of the neutrons emitted by fuel debris are unknown. Fuel debris was categorized into five types according to the elapsed time from the accident, burnup, and fuel type (UO<sub>2</sub> or mixed oxide). The number and energy spectra of (α, <em>n</em>) and spontaneous fission neutrons emitted from 1 g of each fuel debris type were estimated using the SOURCES 4C code to obtain the neutron characteristics. The results showed that the average neutron energy is approximately 2.1 MeV, regardless of the type of fuel debris. However, the intensities of neutrons emitted from the fuel debris in 1F Units 2 and 3 varied by four orders of magnitude according to the fuel debris type.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113791"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168373","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}

引用次数: 0

Proposal of evaluation methodology of performance loss function for piping system subjected to seismic load

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-02-01 DOI: 10.1016/j.nucengdes.2024.113744

Akihiro Mano, Takuya Sato, Masakazu Ichimiya, Naoto Kasahara

{"title":"Proposal of evaluation methodology of performance loss function for piping system subjected to seismic load","authors":"Akihiro Mano, Takuya Sato, Masakazu Ichimiya, Naoto Kasahara","doi":"10.1016/j.nucengdes.2024.113744","DOIUrl":"10.1016/j.nucengdes.2024.113744","url":null,"abstract":"<div><div>The risk concept is the fundamental concept for the safety of nuclear facilities. In the structural field, risk concept is defined for a component as the multiplication of failure occurrence probability and failure consequence on safety performance of a component. For the risk analysis for a component, fragility analysis is employed. It is noted that the analysis considers only failure occurrence probability. Failure consequence is assumed to be complete loss of safety performance of components. Thus, the risk may be overestimated for a failure of component with tiny safety performance degradation. If such analysis result is applied for planning of countermeasures for risk reduction, countermeasures may become ineffective. Based on the background, we previously proposed the new concept named “performance loss function” regarding risk analysis for a component considering both failure occurrence probability and its consequence. In this paper, we propose a quantitative evaluation methodology for the proposed concept for a piping system with cooling performance subjected to seismic load. In the evaluation methodology, flow rate of in the piping system is employed as the numerical index to represent the degradation level of the cooling performance. The failure consequence on the safety performance is quantified by considering the relationship between geometrical changes caused by the failure mode and the flow rate reduction. The failure consequence is combined with occurrence probability of a failure mode to become the performance loss function.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113744"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143169074","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}

引用次数: 0

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

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-02-01 DOI: 10.1016/j.nucengdes.2024.113794

İbrahim Hakan Ünal , Gökhan Durmuş

{"title":"Investigation of radiation and water impermeability attitudes of heavy concrete produced with hematite and chromite for nuclear reactor shielding structure","authors":"İbrahim Hakan Ünal , Gökhan Durmuş","doi":"10.1016/j.nucengdes.2024.113794","DOIUrl":"10.1016/j.nucengdes.2024.113794","url":null,"abstract":"<div><div>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 × 150 × 150 mm<sup>3</sup> in size, and plate samples, 200 × 200 × 50 mm<sup>3</sup> 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<sup>−1</sup> and 0.25 cm<sup>−1</sup> for gamma, 0.16 cm<sup>−1</sup>, 0.17 cm<sup>−1</sup> and 0.21 cm<sup>−1</sup> for neutron in HHH-40-1. The maximum water absorption was detected in CCC-40-2 with 12.3 mm. In C50 samples, the highest UCS is 72.6 MPa, attenuation coefficients are 0.25 cm<sup>−1</sup> and 0.3 cm<sup>−1</sup> for gamma, 0.16 cm<sup>−1</sup>, 0.19 cm<sup>−1</sup> and 0.23 cm<sup>−1</sup> 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.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113794"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143169077","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}

引用次数: 0

A nuclear-driven combined cooling and power system based on supercritical and transcritical CO2 cycles: energy, exergy, and exergoeconomic analysis, multi-objective optimization and reactor startup analysis

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-02-01 DOI: 10.1016/j.nucengdes.2024.113816

Mengxuan Wang , Jiangfeng Wang , Fang Luo , Juwei Lou , Junyi Zhang

{"title":"A nuclear-driven combined cooling and power system based on supercritical and transcritical CO2 cycles: energy, exergy, and exergoeconomic analysis, multi-objective optimization and reactor startup analysis","authors":"Mengxuan Wang , Jiangfeng Wang , Fang Luo , Juwei Lou , Junyi Zhang","doi":"10.1016/j.nucengdes.2024.113816","DOIUrl":"10.1016/j.nucengdes.2024.113816","url":null,"abstract":"<div><div>Energy cascade utilization is deemed as one of the most efficient paths to cope with energy crises and improve energy efficiency. To further analyze the application and performance of the multi-generation system in nuclear energy cascade utilization, this paper establishes a combined power and cooling system comprising a supercritical CO<sub>2</sub> Brayton cycle with a transcritical CO<sub>2</sub> refrigeration cycle. The energy, exergy, and exergoeconomic (3E) analyses of the combined system are conducted and multi-objective optimization is carried out to examine the optimum system performance. A startup strategy for the supercritical CO<sub>2</sub> direct-cooled reactor is proposed and the reactor’s startup transient process is analyzed to examine the coupling characteristics between the reactor and its secondary circuit coolant system. The results show that the increasing the main compressor and throttle valve outlet pressure as well as turbine inlet temperature are conducive to the thermodynamic and exergoeconomic performance of the system. When the reactor thermal power is 5 MW, the maximum system thermal efficiency is 42.50 %, while <em>COP</em> and total product unit cost are 3.53 and 14.01 $·GJ<sup>−1</sup>, respectively. The comparative analysis with a similar system reveals the superior performance of the combined system. Under identical operating conditions, the thermal efficiency and exergy efficiency of the combined system are 2.50 % and 3.36 % higher, respectively, compared to the reference system, with the total product unit cost being 11.62 % lower. Moreover, the response time for the reactor to reach low power steady state and full power steady state is about 487 min and 236 min, respectively.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113816"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143169081","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}

引用次数: 0

The minimum burnup requirement analysis for molten chlorine-salt fast reactor

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-02-01 DOI: 10.1016/j.nucengdes.2025.113831

Yu Peng , Guifeng Zhu , Rui Xue , Minyang Ma , Ying Song , Ailin Yao , Chengfei Tao

{"title":"The minimum burnup requirement analysis for molten chlorine-salt fast reactor","authors":"Yu Peng , Guifeng Zhu , Rui Xue , Minyang Ma , Ying Song , Ailin Yao , Chengfei Tao","doi":"10.1016/j.nucengdes.2025.113831","DOIUrl":"10.1016/j.nucengdes.2025.113831","url":null,"abstract":"<div><div>To investigate the discharge burnup of a chlorine salt fast reactor operating in breed-and-burn (B&B) mode, this study used the neutron balance method and selected five commonly used chloride salts. This article examines the effects of heavy metal density, discharge cycle time and online fuel salt removal schemes on the minimum burnup requirement, acceptable reactor core neutron loss conditions for maintaining B&B modes, and the effect of <sup>37</sup>Cl enrichment ratio on minimum burnup requirement. The results show that heavy metal density significantly influences the minimum burnup requirement, while online removal of fission gases and insoluble noble metals significantly improves the neutron economy in the fuel salt system and increases the acceptable neutron loss conditions for B&B operations. Increasing enrichment of <sup>37</sup>Cl can also increase neutron economy by reducing parasitic absorption of neutrons. However, the effect is not as obvious as the simultaneous online removal of fission gases and insoluble noble metals.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113831"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167945","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}

引用次数: 0