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Dr. Xiang Wang
Harbin Engineering University

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Research Keywords & Expertise

0 Nuclear Energy
0 Neutron Physics- Generations and Applications,
0 nuclear reactor power
0 thermal-hydraulics
0 radiation shielding

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Journal article
Published: 29 June 2021 in Annals of Nuclear Energy
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The dual-fluid reactor (DFR) has been introduced into the nuclear community as a completely new concept for molten salt fast reactors (MSFRs). It has provided possibilities of having a reactor with inherent safety features, high fuel utility, fuel breeding, high efficiency, compact structures, online refueling, actinides burning, and minimum waste. DFR uses molten uranium–plutonium trichloride (UCl3 + PuCl3 [UPu]) as the default fuel salt and transuranium (TRU) fuel salt as the optional fuel salt. These reactors adopt liquid lead as the coolant in the reactor core. The fuel salt goes into a pyro-processing unit (PPU) for fuel re-freshening, and the liquid lead goes into a heat exchanger where it is cooled by a secondary gas coolant, which could be helium or super-critical carbon dioxide (sCO2). In this study, the system dynamics and transients of a DFR with 3000 MW (DFR3G) and 250 MW (DFR250M) are investigated. For this purpose, a model with three one-dimensional nodalized loops coupled with point-kinetic neutron dynamics is constructed based on the heat balance equations to investigate the system responses to the changes of various boundary conditions. The calculation provides a preview of the transient behavior of the DFR system, and the analysis is provided from a reactor safety point of view. The conclusion discusses the research and development status of the DFR for future improvements.

ACS Style

Mingyue Wang; Xun He; Rafael Macian-Juan; Xiang Wang. One-dimensional transient analysis of the dual-fluid reactor system. Annals of Nuclear Energy 2021, 162, 108481 .

AMA Style

Mingyue Wang, Xun He, Rafael Macian-Juan, Xiang Wang. One-dimensional transient analysis of the dual-fluid reactor system. Annals of Nuclear Energy. 2021; 162 ():108481.

Chicago/Turabian Style

Mingyue Wang; Xun He; Rafael Macian-Juan; Xiang Wang. 2021. "One-dimensional transient analysis of the dual-fluid reactor system." Annals of Nuclear Energy 162, no. : 108481.

Journal article
Published: 08 April 2021 in Expert Systems with Applications
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In this paper, a method based on a backpropagation neural network (BPNN) is proposed to calculate the exposure buildup factor (BD) of a point isotropic source in an infinite homogeneous medium under arbitrary energy and mean free path (mfp). The results obtained for aluminum, iron, lead, and concrete based on BPNN are compared to ANSI/ANS-6.4.3 standard data, the results calculated by MCNP 5 Monte Carlo code, and a geometric progression (G-P) fitting formula, and show that the BD calculated by the BPNN model is more consistent with the ANS standard data. This method improves the calculation and fitting effect of BD compared to other methods. This paper proposes a systematic process combining a Monte Carlo method and BPNN to calculate and predict the BD of new materials under different energy and mfp, thus replacing the G-P fitting formula and improving calculation accuracy.

ACS Style

Runkai Chen; Antonio Cammi; Marcus Seidl; Rafael Macian-Juan; Xiang Wang. Calculation of gamma-ray exposure buildup factor based on backpropagation neural network. Expert Systems with Applications 2021, 177, 115004 .

AMA Style

Runkai Chen, Antonio Cammi, Marcus Seidl, Rafael Macian-Juan, Xiang Wang. Calculation of gamma-ray exposure buildup factor based on backpropagation neural network. Expert Systems with Applications. 2021; 177 ():115004.

Chicago/Turabian Style

Runkai Chen; Antonio Cammi; Marcus Seidl; Rafael Macian-Juan; Xiang Wang. 2021. "Calculation of gamma-ray exposure buildup factor based on backpropagation neural network." Expert Systems with Applications 177, no. : 115004.

Journal article
Published: 15 December 2020 in Sustainability
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The unique design features of the molten salt fast reactor (MSFR) should enable higher coolant temperatures than in conventional water reactors, with a significant improvement in the achievable thermodynamic performance. The use of a molten salt as both fuel and coolant, however, poses several advanced heat transfer challenges, such as the design of innovative heat exchangers and energy conversion systems. In this work, we address a preliminary but quantitative analysis of the energy conversion system for the MSFR, based on reference design data from the SAMOFAR H2020-EURATOM project. We consider three main technologies, i.e., the supercritical steam cycle, the closed helium cycle and the helium/steam combined cycle. Preliminary design results are presented for each technology, based on a simplified modelling approach. The considered cycles show promising efficiency improvements, with the best performance being proven by the supercritical steam cycle. The analysis also highlights the critical issue related to the risk of freezing of the molten salts within the secondary heat exchangers, due to the low inlet temperatures of the working fluids. Results show potential incompatibility between the freezing point of molten salts and the temperatures typical of steam cycles, while helium cycles offer the best chances of freezing avoidance. The combined cycle promises intermediate performance in terms of thermodynamic efficiency and thermal compatibility with molten salts comparable with closed helium cycles.

ACS Style

Andrea Di Ronco; Francesca Giacobbo; Guglielmo Lomonaco; Stefano Lorenzi; Xiang Wang; Antonio Cammi. Preliminary Analysis and Design of the Energy Conversion System for the Molten Salt Fast Reactor. Sustainability 2020, 12, 10497 .

AMA Style

Andrea Di Ronco, Francesca Giacobbo, Guglielmo Lomonaco, Stefano Lorenzi, Xiang Wang, Antonio Cammi. Preliminary Analysis and Design of the Energy Conversion System for the Molten Salt Fast Reactor. Sustainability. 2020; 12 (24):10497.

Chicago/Turabian Style

Andrea Di Ronco; Francesca Giacobbo; Guglielmo Lomonaco; Stefano Lorenzi; Xiang Wang; Antonio Cammi. 2020. "Preliminary Analysis and Design of the Energy Conversion System for the Molten Salt Fast Reactor." Sustainability 12, no. 24: 10497.

Special issue research article
Published: 19 August 2020 in International Journal of Energy Research
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The super‐critical water‐cooled reactor (SCWR) is one of six types of Gen‐IV reactors. Based on the current mature pressurized water reactor technology, it has several advantages, including high conversion ratio, high breeding of nuclear fuel, high thermal efficiency, and a simplified system. However, due to the combination of water coolant and fast neutron spectrum, there is a need to address safety problems caused by the positive void coefficient. The purpose of this paper is to give a study on physical characteristics for the reactor core. Herein, the core modeling of an SCWR with a single‐pass flow is that seed assemblies and blanket assemblies are arranged in a tornado manner, in which MCNP and SERPENT are used for critical calculation. The assembly design is discussed from the aspects of the thickness of the solid moderator layer and cladding material. Based on the optimized assembly design, UO2, PuO2‐ThO2, and PuO2‐UO2 are selected for the sensitivity analysis of fuel composition. In addition, sensitivity analysis for axial coolant density distribution, water‐to‐fuel volume ratio, reflector thickness and fuel arrangement are carried out. Several fundamental system characteristics are discussed in this paper, including keff, void coefficient, Doppler coefficient, neutron spectrum, power distribution, one‐cycle burnup and conversion ratio. The results help in understanding the design of the SCWR better and provide fundamental research data for further study.

ACS Style

Mingyue Wang; Liping Huang; Qian Zhang; Marcus Seidl; Zijing Liu; Xiang Wang. Preliminary study on physical characteristics of single‐pass super‐critical water‐cooled reactor core. International Journal of Energy Research 2020, 45, 11807 -11821.

AMA Style

Mingyue Wang, Liping Huang, Qian Zhang, Marcus Seidl, Zijing Liu, Xiang Wang. Preliminary study on physical characteristics of single‐pass super‐critical water‐cooled reactor core. International Journal of Energy Research. 2020; 45 (8):11807-11821.

Chicago/Turabian Style

Mingyue Wang; Liping Huang; Qian Zhang; Marcus Seidl; Zijing Liu; Xiang Wang. 2020. "Preliminary study on physical characteristics of single‐pass super‐critical water‐cooled reactor core." International Journal of Energy Research 45, no. 8: 11807-11821.

Journal article
Published: 30 June 2020 in Annals of Nuclear Energy
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Burnup capability is developed on the ALPHA code with Sanchez-Pomraning MOC (Sanchez-MOC) approach for FCM fuel design. To better model the double heterogeneity (DH) of FCM fuel, ALPHA adopts a combination strategy of macroscopic discretization at pin level and microscopic discretization at particle level for burnup region. The burnup capability is validated by code-to-code comparison with Serpent for FCM fuel. In addition, the effect of spatial discretization of the burnup calculation on the macroscopic and microscopic level is analyzed. In this work, it is suggested that the FCM pin cell without poison or containing Er2O3 should implement the macroscopic 5-ring and microscopic non-discretization scheme. For B-bearing and Gd-bearing FCM pin cell, the macroscopic burnup region should reach 5 rings and 8 rings respectively, and their microscopic burnup region both require 2 regions.

ACS Style

Yunfei Zhang; Qian Zhang; Song Li; Yuechao Liang; Lei Lou; Xiang Wang; Qiang Zhao; Zhijian Zhang. Evaluation of burnup calculation for double-heterogeneity system based on Sanchez-MOC framework in LWR. Annals of Nuclear Energy 2020, 147, 107668 .

AMA Style

Yunfei Zhang, Qian Zhang, Song Li, Yuechao Liang, Lei Lou, Xiang Wang, Qiang Zhao, Zhijian Zhang. Evaluation of burnup calculation for double-heterogeneity system based on Sanchez-MOC framework in LWR. Annals of Nuclear Energy. 2020; 147 ():107668.

Chicago/Turabian Style

Yunfei Zhang; Qian Zhang; Song Li; Yuechao Liang; Lei Lou; Xiang Wang; Qiang Zhao; Zhijian Zhang. 2020. "Evaluation of burnup calculation for double-heterogeneity system based on Sanchez-MOC framework in LWR." Annals of Nuclear Energy 147, no. : 107668.

Short communication
Published: 26 June 2020 in Annals of Nuclear Energy
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The existing methods on the double heterogeneity self-shielding problem adopt the Markovian distribution to describe the spatial distribution of dispersed particles. However, the applicability of Markovian distribution for the practical Fully-Ceramic-Microencapsulated (FCM) fuel still needs to be discussed. This work aims to evaluate the effect caused by the approximation of Markovian distribution on the resonance self-shielding treatment in the situation of filled particles with different sizes in FCM fuel design. To be compared with the Markovian distribution, a series of practical chord length distributions are depicted by the Chord-Length-Sampling method. We implemented the analytical average Dancoff method and the Sanchez-Pomraning Subgroup method for 2D cylindrical FCM fuel. A series of average Dancoff factors and resonance cross-sections are calculated with the reference of results provided by the Monte Carlo method. The results show that the Markovian distribution can be accepted with the normal particle size. In contrast, the Markovian distribution affects the accuracy of resonance calculation severely in terms of extremely large particle size of the dispersed particles.

ACS Style

Yuechao Liang; Qian Zhang; Song Li; Liang Liang; Xiang Wang; Qiang Zhao; Lei Lou. Investigation of the chord length Markovian probability distribution for self-shielding treatment on double heterogeneity problem. Annals of Nuclear Energy 2020, 146, 107658 .

AMA Style

Yuechao Liang, Qian Zhang, Song Li, Liang Liang, Xiang Wang, Qiang Zhao, Lei Lou. Investigation of the chord length Markovian probability distribution for self-shielding treatment on double heterogeneity problem. Annals of Nuclear Energy. 2020; 146 ():107658.

Chicago/Turabian Style

Yuechao Liang; Qian Zhang; Song Li; Liang Liang; Xiang Wang; Qiang Zhao; Lei Lou. 2020. "Investigation of the chord length Markovian probability distribution for self-shielding treatment on double heterogeneity problem." Annals of Nuclear Energy 146, no. : 107658.

Journal article
Published: 02 March 2020 in Annals of Nuclear Energy
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The sensitivity of FCM parameters to the equivalent radius of the RPT model is evaluated by a quantitative method, and six important FCM parameters are identified. Then the relationship between the six parameters and the equivalent radius of the RPT model is analyzed. Finally, a rapid fitting RPT method is proposed. The new method can directly establish the RPT model by a fitting formula without relying on the results of the Monte Carlo high-fidelity model. The numerical results show that the fitting RPT method achieves high precision in most test cases, which demonstrates practicability for the lattice physics calculation of FCM fuel using conventional PWR lattice physics codes.

ACS Style

Yunfei Zhang; Qian Zhang; Jinchao Zhang; Xiang Wang; Lei Lou; Guoming Liu; Qiang Zhao; Zhijian Zhang. A comprehensive evaluation of the RPT method on FCM fuel in light water reactor. Annals of Nuclear Energy 2020, 142, 107434 .

AMA Style

Yunfei Zhang, Qian Zhang, Jinchao Zhang, Xiang Wang, Lei Lou, Guoming Liu, Qiang Zhao, Zhijian Zhang. A comprehensive evaluation of the RPT method on FCM fuel in light water reactor. Annals of Nuclear Energy. 2020; 142 ():107434.

Chicago/Turabian Style

Yunfei Zhang; Qian Zhang; Jinchao Zhang; Xiang Wang; Lei Lou; Guoming Liu; Qiang Zhao; Zhijian Zhang. 2020. "A comprehensive evaluation of the RPT method on FCM fuel in light water reactor." Annals of Nuclear Energy 142, no. : 107434.

Short communication
Published: 09 January 2020 in Annals of Nuclear Energy
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The impact from the neutronic-temperature coupling on the depletion calculation of fuel pellet in light water reactor configuration is investigated with coupling schemes between the Monte Carlo code Serpent and an in-house radial temperature solver. Two external coupling schemes are developed. One is the direct coupling of burnup and power on the whole pin level and the other one is the fine coupling of radial burnup and power distribution. Numerical results show that the neutronic-temperature coupling brings non-negligible effect on the kinf, power distribution and isotopic inventory in depletion calculation. The difference between two coupling scheme shows that the fine coupling of radial burnup and power distribution and the direct coupling obtain equal result from the aspect of lattice physics calculation.

ACS Style

Qian Zhang; Jian Yu; Jinchao Zhang; Xiang Wang; Liang Liang. Investigation on the depletion calculation with neutronic-temperature coupling in the fuel pellet of light water reactor. Annals of Nuclear Energy 2020, 140, 107297 .

AMA Style

Qian Zhang, Jian Yu, Jinchao Zhang, Xiang Wang, Liang Liang. Investigation on the depletion calculation with neutronic-temperature coupling in the fuel pellet of light water reactor. Annals of Nuclear Energy. 2020; 140 ():107297.

Chicago/Turabian Style

Qian Zhang; Jian Yu; Jinchao Zhang; Xiang Wang; Liang Liang. 2020. "Investigation on the depletion calculation with neutronic-temperature coupling in the fuel pellet of light water reactor." Annals of Nuclear Energy 140, no. : 107297.

Research article
Published: 01 September 2019 in International Journal of Energy Research
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A liquid‐fuel heat‐pipe reactor (LFHPR) is a novel fast heterogeneous reactor developed by Harbin Engineering University, China, on the basis of liquid‐fuel reactor designs and the heat‐pipe reactor concept. In the concept, the reactor abandons the graphite moderator and keeps neither fuel tubes arranged in the graphite nor fuel rings around the heat pipe. Instead, the reactor applies molten salt fuels, molten metallic eutectic fuels, or other fuels in liquid form. The heat generated in the reactor is removed by the heat pipes driven by liquid metals. With this change, an LFHPR is much more flexible in design and application and able to achieve several advanced features compared with conventional heat‐pipe reactors. In this paper, we describe the general reactor design of an LFHPR, discuss its potential advantages, and give a preliminary verification of the neutron physical feasibility for the reference case, which uses molten salt as the fuel, by using both Monte Carlo and deterministic methods. Results show that the LFHPR yields a hard neutron spectrum that brings a very good neutron economy and is a promising application for breeding. From our approach, we conclude that the proposed LFHPR has a very high power density and high negative temperature feedback coefficient.

ACS Style

Xiang Wang; Qian Zhang; Kun Zhuang; Xun He; Marcus Seidl; Rafael Macian‐Juan. Neutron physics of the liquid‐fuel heat‐pipe reactor concept with molten salt fuel—Static calculations. International Journal of Energy Research 2019, 1 .

AMA Style

Xiang Wang, Qian Zhang, Kun Zhuang, Xun He, Marcus Seidl, Rafael Macian‐Juan. Neutron physics of the liquid‐fuel heat‐pipe reactor concept with molten salt fuel—Static calculations. International Journal of Energy Research. 2019; ():1.

Chicago/Turabian Style

Xiang Wang; Qian Zhang; Kun Zhuang; Xun He; Marcus Seidl; Rafael Macian‐Juan. 2019. "Neutron physics of the liquid‐fuel heat‐pipe reactor concept with molten salt fuel—Static calculations." International Journal of Energy Research , no. : 1.

Journal article
Published: 31 December 2018 in Annals of Nuclear Energy
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The enhanced thermal conductivity oxide (ECO) nuclear fuel with BeO is analyzed from the aspect of neutronics. The Westinghouse 17 × 17 pressurized water reactor lattice model is selected as the reference and the influence on the neutronic features of different volume fractions of BeO is investigated by Monte Carlo depletion code Serpent. Results from criticality calculation of fresh fuel and depletion calculation show that increasing the fraction of BeO in the fuel will lead to more penalty on reactivity caused by extra neutron capture while the increasing moderating effect of BeO will increase the reactivity provided by the fuel. The two opposite effects compete with each other to decide the final impact the cycle length. As a result, different amount of enrichment of 235U is needed to maintain the reactivity evaluated in different models. The moderating effect leads to the softening of the spectrum and less accumulation of 239Pu when adding more BeO in the fuel. From the investigation of the radial profile, the spatially self-shielding effect decreased with the increasing of the volume fraction of BeO. Reactivity perturbation analysis show that adding BeO in the fuel causes the moderator coefficient to become less negative, which should be a concern for the reactor safety. In the analysis of the leading assembly with several standard UO2 fuel replaced with UO2-BeO, the pin power distributions of leading assemblies show the power defect at the position of the UO2-BeO fuel.

ACS Style

Qian Zhang; Xiang Wang; Yunfei Zhang; Xiaomiao Chi; Liang Liang; Kun Zhuang; Chao Wang. Neutronic analysis for potential Accident Tolerant Fuel UO2-BeO in the light water reactor. Annals of Nuclear Energy 2018, 127, 278 -292.

AMA Style

Qian Zhang, Xiang Wang, Yunfei Zhang, Xiaomiao Chi, Liang Liang, Kun Zhuang, Chao Wang. Neutronic analysis for potential Accident Tolerant Fuel UO2-BeO in the light water reactor. Annals of Nuclear Energy. 2018; 127 ():278-292.

Chicago/Turabian Style

Qian Zhang; Xiang Wang; Yunfei Zhang; Xiaomiao Chi; Liang Liang; Kun Zhuang; Chao Wang. 2018. "Neutronic analysis for potential Accident Tolerant Fuel UO2-BeO in the light water reactor." Annals of Nuclear Energy 127, no. : 278-292.

Journal article
Published: 25 October 2018 in Progress in Nuclear Energy
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This article focuses on the validation of the basic design of the distribution zone of the Dual Fluid Reactor (DFR), which is based on the Generation IV Molten-Salt Reactor concept and the Liquid-Metal Cooled Reactor. The distribution zone located under the core joins the core and input pipes of both fluids. On the top of the core, another distribution zone specified for output is also symmetrically set up; however it will not be mentioned in the paper. This distribution zone partitions the inlet flow of the salt, determines the pressure and velocity distribution of the core inlet and influences the neutronic behaviors in the core. This article uses COMSOL as the tool to develop models and to study the pressure and velocity differences of fuel salt and coolant lead in the “original” and “alternative” configurations/designs of the distribution zone in the DFR concept. COMSOL is also used to determine the initial/boundary conditions of configuration variations of the distribution zone by hydraulic modelling and calculations. The simulation results are analyzed and the optimized design is delivered.

ACS Style

Xiang Wang; Chunyu Liu; Rafael Macian-Juan. Preliminary hydraulic analysis of the distribution zone in the Dual Fluid Reactor concept. Progress in Nuclear Energy 2018, 110, 364 -373.

AMA Style

Xiang Wang, Chunyu Liu, Rafael Macian-Juan. Preliminary hydraulic analysis of the distribution zone in the Dual Fluid Reactor concept. Progress in Nuclear Energy. 2018; 110 ():364-373.

Chicago/Turabian Style

Xiang Wang; Chunyu Liu; Rafael Macian-Juan. 2018. "Preliminary hydraulic analysis of the distribution zone in the Dual Fluid Reactor concept." Progress in Nuclear Energy 110, no. : 364-373.

Research article
Published: 28 August 2018 in International Journal of Energy Research
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ACS Style

Xiang Wang; Rafael Macian-Juan. Steady-state reactor physics of the dual fluid reactor concept. International Journal of Energy Research 2018, 42, 4313 -4334.

AMA Style

Xiang Wang, Rafael Macian-Juan. Steady-state reactor physics of the dual fluid reactor concept. International Journal of Energy Research. 2018; 42 (14):4313-4334.

Chicago/Turabian Style

Xiang Wang; Rafael Macian-Juan. 2018. "Steady-state reactor physics of the dual fluid reactor concept." International Journal of Energy Research 42, no. 14: 4313-4334.