Jianqiang Shan

• Ph.D
• Professor (Full) at Xi'an Jiaotong University

To meet the flexible and reliable power supply requirements of specialized scenarios, such as deep space and deep-sea exploration, a novel megawatt nuclear power system that couples a heat pipe cooled reactor with a supercritical CO2 Brayton cycle was proposed. During the early stages of system design, the start-up process represents a critical tra...

Integration of the supercritical CO2 (S-CO2) Brayton cycle with conventional power plants and fourth-generation nuclear reactors yields promising energy conversion technologies, including a load-following S-CO2 Brayton cycle cooled reactor system. However, the response of cycle components and reactor core changes significantly during the partial-lo...

Heat pipe reactors are one of the ideal reactor types for Unmanned Underwater Vehicles (UUVs) due to high energy density, long lifecycle, modularity, and compact structure. In the reactor concept design stage, simulation and analysis of typical accidents on heat pipe reactors are key to assessing their inherent safety. The loss of heat sink acciden...

The high-temperature heat pipe (HTHP) cooled reactor has the advantages of compact, safety and reliability, rendering it a promising technology for application in Unmanned Underwater Vehicles (UUVs). In this paper, a Passive Residual Heat Removal (PRHR) system is designed for a marine nuclear power plant proposed by the Chinese Academy of Engineeri...

The Supercritical CO2 (S-CO2) Brayton cycle-cooled nuclear reactor system presents characteristics such as higher efficiency and a compact design, which are significant for improving the competitiveness and sustainability of nuclear energy. Additionally, it is capable of supporting load-following operations. Nevertheless, the system dynamics underg...

The Supercritical CO2 (S-CO2) Brayton cycle-cooled nuclear reactor system offers advantages such as high efficiency and a compact configuration, both essential to enhance the competitiveness and sustainability of nuclear energy. Robust control strategies are required to maintain its performance and stability during operational and accidental condit...

The Supercritical CO2 (S-CO2) Brayton cycle-cooled nuclear reactor system is characterized by its high efficiency, small size, and ability to adapt to changes in load swiftly. The use of modern control methods to enhance its stability and efficiency necessitates using a simulation tool that enables their implementation. In this study, an integrated...

A megawatt-class nuclear power system has been developed by coupling a heat pipe reactor with a supercritical carbon dioxide (S-CO2) Brayton cycle. This system offers advantages in terms of high safety, power density, and compactness. [Purpose] This study aims at the operation characteristics of this power system with high efficiency and compactnes...

Heat pipe cooled reactors (HPRs) have been considered as one of the most promising candidates for deep space and deep-sea missions due to their advantages of simple structure, high power density and high reliability, etc. To investigate the transient characteristics of such heat pipe cooled reactors, including startup, shutdown, power transients an...

To analyze the potential impact of accident tolerant fuel (ATF) on reactor safety under pump shaft-stuck accident, China's improved three-loop pressurized water reactor CPR1000 was used as reference power station to carry out second development based on the system analysis code NUSOL-SYS. The performance of CPR1000 with different ATF combinations u...

Based on a compact heat pipe-cooled reactor carried by an unmanned underwater vehicle, a complete safety analysis model of heat pipe cooled reactor is established and optimized in this paper, which mainly includes core power transient model, cold start-up model of high temperature heat pipe and two-dimensional heat pipe grid model. The passive resi...

Metal fuel is considered to be the future of the Sodium-cooled Fast Reactor due to its advantage. However, the lower melting point is limit for its application, accuracy prediction of the fuel behavior is urgent. In this paper, based on the principle of FEM, the coupled model of thermal conduction and stress-strain analysis is established. The two...

In severe accidents, sodium-cooled fast reactors (SFRs) are more concerned with core disruptive accidents (CDAs), and in-vessel retention (IVR) is an important mitigation measure to ensure the integrity of the reactor vessel. The distribution and accumulation uniformity of the melt in the lower chamber seriously affects the residual heat discharge...

The fuel shape and flow path structure of the plate-type fuel reactor are different from those of the conventional rod-bundle reactor, which leads to the difference between the thermal-hydraulic phenomenon and thermal conduction in the rod bundle assembly. Therefore, in this thesis, the subchannel analysis model and code applied to the plate-type f...

Nuclear safety is the lifeline for the development and application of nuclear energy. In severe accidents of pressurized water reactor (PWR), aerosols, as the main carrier of fission products, are suspended in the containment vessel, posing a potential threat of radioactive contamination caused by leakage into the environment. The gas-phase aerosol...

For severe accidents, in-vessel retention (IVR) is a very effective and crucial severe accident mitigation measure. The lower head of the reactor pressure vessel plays a vital role in the IVR strategy. The failure of the lower head may lead to the release of radioactive substances into the environment. During the implementation of IVR, the lower he...

The internally and externally cooled annular fuel is an innovative fuel geometry proposal for advanced PWR, which could provide a substantial increase of power density while maintaining or improving safety margins. The quenching behavior of annular fuel during reflood phase in LOCA is more complicated than cylindrical solid fuel, owing to the geome...

Deformed fuel rods can cause a partial blockage of the flow area in a subchannel. Such flow blockage will influence the core coolant flow and further the core heat transfer during the reflooding phase and subsequent severe accidents. Nevertheless, most of the system analysis codes simulate the accident process based on the assumed flow blockage rat...

The Melt-Fuel-Coolant Interaction (MFCI) is one of the important phenomena during reactor severe accidents. In particular, the fragmentation process of the molten fuel in the metal coolant is a complicated multi-phase and multi-component heat transfer phenomenon accompanied by solid-liquid coupling and phase change, which is of great significance t...

The externally and internally cooled annular fuel is considered to have a good prospect with the shorter heat transfer path, larger heat transfer surface and higher burnup depth, which can effectively improve the safety of the reactor. This paper takes the nuclear power plant CPR1000, which is representative in China, as the reference power plant....

An experiment was carried out to obtain data on critical heat flux (CHF) in vertical internally heated annuli cooled with Refrigerant (R-134a) at high-pressure subcooled and low-quality conditions (outlet pressure:1.78–2.72 MPa, mass flux: 587–2135kg/m2s, critical quality: −0.64–0.12). The test section has two forms, concentric, eccentric (ε = 0.43...

The grand challenge of “net-zero carbon” emission calls for technological breakthroughs in energy production. The traveling wave reactor (TWR) is designed to provide economical and safe nuclear power and solve imminent problems, including limited uranium resources and radiotoxicity burdens from back-end fuel reprocessing/disposal. However, qualific...

Prediction of pressure drop by wall friction and at geometric discontinuities is important in nuclear thermal hydraulics simulation. Pressure-drop models were established and upgraded for decades; however, errors in predictions of experimental data are up to 40%, namely in low flow and two-phase conditions and particularly at geometric discontinuit...

SuperCritical Water Reactor(SCWR) applies water beyond the thermodynamic critical point as the coolant, which aims to achieve high efficiency around 45% compared to 33% for existing commercial light water reactors. In order to raise the reactor operating temperature and reactor criticality, the existing SCWR core designs are quite different from th...

Water heat transfer effectiveness during reflooding process in rod bundle geometry is very important for reactor safety under loss of coolant accident. A series of reflooding experiment is carried out at the REFB_NUSOL test facility for a 5 × 5 rod bundle to study the effect of several key parameters on overall reflooding process, including subcool...

Background Heat pipe cooled reactors (HPR) have good inherent safety. In the early stage of core design, heat pipe failure accident is usually one of the design basis accidents that need to be considered.PurposeThis study aims to analyze the neutronic-thermalhydraulic coupling performance of a new type of megawatt heat pipe reactor.Methods Firstly...

Jacobian‐Free Newton‐Krylov (JFNK) method is a stable and high‐efficiency method to solve the multi‐physics coupling problem for the modeling and simulation (M&S) of nuclear reactors. However, for the two‐fluid two‐phase flow model, the large number of constitutive models for different flow regimes as well as their discontinuities between different...

A new simulation code (NUSOL-LMR) is developed to simulate the thermal–hydraulic response to hypothetical accidents in liquid metal fast reactors. The basic field equations of NUSOL-LMR consist of the single-phase continuity equation, momentum equations and energy equations. Heat conduction model of heat structure is also included. A Newton-based f...

This paper reviews in detail the existing visualized experiments and mechanistic models of DNB-type CHF in forced convective flow boiling. And summarizes and reviews the existing CHF mechanistic models from the perspective of whether the NVG (Net Vapor Generation) point is established or not, from both the micro and macro aspects, and the relations...

The global climate shift and the grand challenge of "zero carbon" energy resources call for a technological breakthrough in the energy production industry. The travelling wave reactor (TWR) design targets to solve the imminent problems of economical and safe nuclear power, limitations on uranium resources and radiotoxicity burden from the post proc...

The miniaturization of nuclear reactor power is the direction of nuclear energy development in the future. A nuclear power system with a small solid core and cooled by a high-temperature alkali metal heat pipe has excellent safety performance. The start-up of the heat pipe cooled reactor depends on the heat pipe start-up characteristics from the fr...

Heat-pipe-cooled microreactors (HPMR) use a passive high-temperature alkali metal heat pipe to directly transfer the heat of solid core to the hot end of the intermediate heat exchanger or thermoelectric conversion device, thus avoiding a single point failure. To analyze and evaluate the transient safety characteristics of an HPMR system under acci...

目前以两流体三流场两相流模型为数学模型的核反应堆安全分析程序大都采用半隐数值算法， 数值稳 定性受声速库伦特值的影响。少数以两流体三流场模型和全隐数值算法为基础的程序， 采用经典牛顿迭代法 求解， 雅克比矩阵形成具有一定的难度。为了改善数值算法的稳定性且避免书写雅克比矩阵， 一种无需形成雅 克比矩阵的牛顿-Krylov 迭代法 (Jacobian-free Newton-Krylov， JFNK) 被用于两相流全隐数值算法。两流体三流 场两相流模型分别对汽相、 液相和液滴相建立守恒方程， 使用基于交错网格和有限体积差分全隐式离散守恒方 程， 线性方程组使用 JFNK 算法求解， 当相缺失时， 给缺失相一个很小的份额， 以解决使用三流场模型计算单相、 两相两流场时遇到的数值问题。程序模拟了 R...

During a loss-of-coolant accident (LOCA) in the pressurized water reactor (PWR), there is a possibility that high temperature and internal pressure of the fuel rods lead to ballooning of the cladding, which causes a partial blockage of flow area in a subchannel. Such flow blockage would influence the core coolant flow, thus affecting the core heat...

Thermal stratification of pool-type SFR under transient accident conditions has a very significant impact on the performance of key components of the primary loop and the establishment of natural circulation of the residual heat removal system. In this paper, combined with the mechanism of thermal stratification, a three-dimensional (3-D) system an...

The Accident Tolerant Fuel (ATF) is a new concept of fuel, which can not only withstand the consequences of the accident for a longer time, but also maintain or improve the performance under operating conditions. ISAA is a self-developed severe accident analysis code, which uses modular structures to simulate the development processes of severe acc...

Space nuclear reactor power system (SNRPS) is a priority technical solution to meet the future space power requirement of high-power, low-mass, and long-life. The thermoelectric conversion subsystem is the key component of SNRPS, which greatly affects the performance, quality, and volume of SNRPS. Among all kinds of proposed thermoelectric conversi...

In this paper an experimental study on two-phase flow pressure drop characteristics has been performed in a triangular-array rod bundle geometry under various flow conditions (P = 5–9 MPa and G = 100–350 kg·m⁻²·s⁻¹). Total, gravitational and frictional pressure drops are obtained by measurement methods. The effects of pressure, mass flux and thermo...

Cladding ballooning and rupture are the important phenomena at the early stage of a severe accident (SA). For example, the SA caused by the loss of coolant accident (LOCA), the decrease in primary loop pressure and the increase in core temperature will lead to the cladding ballooning and rupture. The cladding deformation affects the flow distributi...

An experiment was carried out to obtain data on critical heat flux (CHF) in an 8-mm vertical heated tube cooled with Refrigerant (R-134a) at high-pressure subcooled and low-quality conditions. Parametric trends of the CHF data were examined and have been shown consistent with those obtained in experiments with water flow. Variations of CHF depends...

Sodium cooled fast reactor (SFR) is the most comprehensively studied and the most extensively operated reactor-type in the 4th generation of nuclear power plants. One of the key points of the transient analysis in pool-type SFR is refined analysis of the transient thermal-hydraulic characteristics of the 3D sodium pool. In this paper, according to...

The performance of the droplets is vital of importance to the annular-mist flow and the heat transfer for the reflooding conditions. In this research, a two-fluid three-field model for the vertical upward flow is developed. The field equations of the vapor, continuous liquid phase and dispersed liquid droplet phase are established. The set of equat...

It is well know that the two-fluid single pressure model is currently widely used to analyze reactor transient accidents such as LOCA. Current mainstream reactor safety analysis codes such as RALAP5 and CATHARE are based on this single pressure model. However this model has been proved to be ill posed in the sense that the equation system is non-hy...

Natural circulation refers to the phenomenon of fluid circulation driven by the driving force formed by the density difference and height difference. Natural circulation is very common in nuclear power system, which has an important impact on reactor safety. Therefore, accurate simulation of natural circulation phenomenon is one of the most importa...

In this study, a critical heat flux (CHF) experiment using R-134a as working fluid was carried out in a vertical single rod geometry. A unique feature of this experiment are high speed visualization studies obtained at a viewing port located at the CHF location. The 9.5 mm O.D. heated rod was placed in a 19 x19mm square channel to simulate the Pres...

Dual cooled annular fuel is a novel fuel design, which has the potential to improve the reactor power density while maintaining or improving its safety margin. The effects of tight-lattice geometry, fuel burnup, fuel expansion, coolant channel blockage on the thermal hydraulic performance of annular fuel is studied to illustrate its special feature...

Cladding rupture is an important demarcation point for severe accidents. Most severe accident analysis codes are based on simple parameter models to determine the damage of the cladding. This paper uses the thermal–mechanical model to develop a core Fuel Rod Thermal-Mechanical Behavior analysis (FRTMB) module to analyze the fuel mechanical behavior...

Cladding ballooning and rupture are the important phenomena at the early stage of a severe accident. Most severe accident analysis codes determine the cladding rupture based on simple parameter models. In this paper, a FRTMB module was developed using the thermal-mechanical model to analyze the fuel mechanical behavior. The purpose is to judge the...