Preliminary design and assessment of a heat pipe residual heat removal system for the reactor driven subcritical facility

Wenwen Zhang,Kaichao Sun,Chenglong Wang,Dalin Zhang,Wenxi Tian,Suizheng Qiu,G.H. Su 한국원자력학회 2021 Nuclear Engineering and Technology Vol.53 No.12

A heat pipe residual heat removal system is proposed to be incorporated into the reactor drivensubcritical (RDS) facility, which has been proposed by MIT Nuclear Reactor Laboratory for testing anddemonstrating the Fluoride-salt-cooled High-temperature Reactor (FHR). It aims to reduce the risk of thesystem operation after the shutdown of the facility. One of the main components of the system is an aircooled heat pipe heat exchanger. The alkali-metal high-temperature heat pipe was designed to meet theoperation temperature and residual heat removal requirement of the facility. The heat pipe modeldeveloped in the previous work was adopted to simulate the designed heat pipe and assess the heattransport capability. 3D numerical simulation of the subcritical facility active zone was performed by thecommercial CFD software STAR CCM þ to investigate the operation characteristics of this proposedsystem. The thermal resistance network of the heat pipe was built and incorporated into the CFD model. The nominal condition, partial loss of air flow accident and partial heat pipe failure accident weresimulated and analyzed. The results show that the residual heat removal system can provide sufficientcooling of the subcritical facility with a remarkable safety margin. The heat pipe can work under therecommended operation temperature range and the heat flux is below all thermal limits. The facilitypeak temperature is also lower than the safety limits

Prediction of flow boiling heat transfer coefficient in horizontal channels varying from conventional to small-diameter scales by genetic neural network

Jing Zhang,Yichao Ma,Mingjun Wang,Dalin Zhang,Suizheng Qiu,Wenxi Tian,Guanghui Su 한국원자력학회 2019 Nuclear Engineering and Technology Vol.51 No.8

Three-layer back propagation network (BPN) and genetic neural network (GNN) were developed in this study to predict the flow boiling heat transfer coefficient (HTC) in conventional and small-diameter channels. The GNN has higher precision than BPN (with root mean square errors of 17.16% and 20.50%, respectively) and other correlations. The inputs include vapor quality x, mass flux G, heat flux q, diameter D and physical parameter φ, and the predicted flow boiling HTC is set as the outputs. Influences of input parameters on the flow boiling HTC are discussed based on the trained GNN: nucleate boiling promoted by a larger saturated pressure, a larger heat flux and a smaller diameter is dominant in small channels; convective boiling improved by a larger mass flux and a larger vapor quality is more significant in conventional channels. The HTC increases with pressure both in conventional and small channels. The HTC in conventional channels rises when mass flux increases but remains almost unaffected in small channels. A larger heat flux leads to the HTC growth in small channels and an increase of HTC was observed in conventional channels at a higher vapor quality. HTC increases inversely with diameter before dry out.

Water film covering characteristic on horizontal fuel rod under impinging cooling condition

Zhang Penghui,Wang Bowei,Chen Ronghua,Su G.H.,Tian Wenxi,Qiu Suizheng 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.11

Jet impinging device is designed for decay heat removal on horizontal fuel rods in a low temperature heating reactor. An experimental system with a fuel rod simulator is established and experiments are performed to evaluate water film covering capacity, within 0.0287e0.0444 kg/ms mass flow rate, 0 e164.1 kW/m2 heating flux and 13.8e91.4C feeding water temperature. An effective method to obtain the film coverage rate by infrared equipment is proposed. Water film flowing patterns are recoded and the film coverage rates at different circumference angles are measured. It is found the film coverage rate decreases with heating flux during single-phase convection, while increases after onset of nucleate boiling. Besides, film coverage rate is found affected by Marangoni effect and film accelerating effect, and surface wetting is significantly facilitated by bubble behavior. Based on the observed phenomenon and physical mechanism, dry-out depth and initial dry-out rate are proposed to evaluate film covering potential on a heating surface. A model to predict film coverage rate is proposed based on the data. The findings would have reliable guide and important implications for further evaluation and design of decay heat removal system of new reactors, and could be helpful for passive containment cooling research.

Numerical study of oxygen transport characteristics in lead-bismuth eutectic for gas-phase oxygen control

Wang, Chenglong,Zhang, Yan,Zhang, Dalin,Lan, Zhike,Tian, Wenxi,Su, Guanghui,Qiu, Suizheng Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.7

One-dimensional oxygen transport relation is indispensable to study the oxygen distribution in the LBE-cooled system with an oxygen control device. In this paper, a numerical research is carried out to study the oxygen transport characteristics in a gas-phase oxygen control device, including the static case and dynamic case. The model of static oxygen control is based on the two-phase VOF model and the results agree well with the theoretical expectation. The model of dynamic oxygen control is simplified and the gas-liquid interface is treated as a free surface boundary with a constant oxygen concentration. The influences of the inlet and interface oxygen concentration, mass flow rate, temperature, and the inlet pipe location on the mass transfer characteristics are discussed. Based on the results, an oxygen mass transport relation considering the temperature dependence and velocity dependence separately is obtained. The relation can be used in a one-dimensional system analysis code to predict the oxygen provided by the oxygen control device, which is an important part of the integral oxygen mass transfer models.

Performance analysis of automatic depressurization system in advanced PWR during a typical SBLOCA transient using MIDAC

Sun, Hongping,Zhang, Yapei,Tian, Wenxi,Qiu, Suizheng,Su, Guanghui Korean Nuclear Society 2020 Nuclear Engineering and Technology Vol.52 No.5

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.

Assessment of ECCMIX component in RELAP5 based on ECCS experiment

Song, Gongle,Zhang, Dalin,Su, G.H.,Chen, Guo,Tian, Wenxi,Qiu, Suizheng Korean Nuclear Society 2020 Nuclear Engineering and Technology Vol.52 No.1

ECCMIX component was introduced in RELAP5/MOD3 for calculating the interfacial condensation. Compared to other existing components in RELAP5, user experience of ECCMIX component is restricted to developmental assessment applications. To evaluate the capability of the ECCMIX component, ECCS experiment was conducted which included single-phase and two-phase thermal mixing. The experiment was carried out with test sections containing a main pipe (70 mm inner diameter) and a branch pipe (21 mm inner diameter) under the atmospheric pressure. The steam mass flow in the main pipe ranged from 0 to 0.0347 kg/s, and the subcooled water mass flow in the branch pipe ranged from 0.0278 to 0.1389 kg/s. The comparison of the experimental data with the calculation results illuminated that although the ECCMIX component was more difficult to converge than Branch component, it was a more appropriate manner to simulate interfacial condensation under two-phase thermal mixing circumstance, while the two components had no differences under single-phase circumstance.

Design and heat transfer optimization of a 1 kW free-piston stirling engine for space reactor power system

Dai, Zhiwen,Wang, Chenglong,Zhang, Dalin,Tian, Wenxi,Qiu, Suizheng,Su, G.H. Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.7

The Free-Piston Stirling engine (FPSE) is of interest for many research in aerospace due to its advantages of long operating life, higher efficiency, and zero maintenance. In this study, a 1-kW FPSE was proposed by analyzing the requirements of Space Reactor Power Systems (SRPS), of which performance was evaluated by developing a code through the Simple Analysis Method. The results of SAM showed that the critical parameters of FPSE could satisfy the designed requirements. The heater of the FPSE was designed with the copper rectangular fins to enhance heat transfer, and the parametric study of the heater was performed with Computational Fluid Dynamics (CFD) software STAR-CCM⁺. The Performance Evaluation Criteria (PEC) was used to evaluate the heat transfer enhancement of the fins in the heater. The numerical results of the CFD program showed that pressure drop and Nusselt number ratio had a linear growth with the height of fins, and PEC number decreased as the height of fins increased, and the optimum height of the fin was set as 4 mm according to the minimum heat exchange surface area. This paper can provide theoretical supports for the design and numerical analysis of an FPSE for SRPSs.

Assessment of the severe accident code MIDAC based on FROMA, QUENCH-06&16 experiments

Shihao Wu,Yapei Zhang,Dong Wang,Wenxi Tian,Suizheng Qiu,G.H. Su 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.2

In order to meet the needs of domestic reactor severe accident analysis program, a MIDAC (Module Invessel Degraded severe accident Analysis Code) is developed and maintained by Xi'an Jiaotong University. As the accuracy of the calculation results of the analysis program is of great significance for theformulation of severe accident mitigation measures, the article select three experiments to evaluate theupdated severe accident models of MIDAC. Among them, QUENCH-06 is the international standardNo.45, QUENCH-16 is a test for the analysis of air oxidation, and FROMA is an out-of-pile fuel rod meltingexperiment recently carried out by Xi'an Jiaotong University. The heating and melting model withlumped parameter method and the steam oxidation model with Cathcart-Pawel and Volchek-Zvonarevcorrelations combination in MIDAC could better meet the needs of severe accident analysis. Although theinfluence of nitrogen still need to be further improved, the air oxidation model with NUREG still has theability to provide guiding significance for engineering practice

Numerical simulation of three-dimensional flow and heat transfer characteristics of liquid lead-bismuth

He, Shaopeng,Wang, Mingjun,Zhang, Jing,Tian, Wenxi,Qiu, Suizheng,Su, G.H. Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.6

Liquid lead-bismuth cooled fast reactor is one of the most promising reactor types among the fourth-generation nuclear energy systems. The flow and heat transfer characteristics of lead-bismuth eutectic (LBE) are completely different from ordinary fluids due to its special thermal properties, causing that the traditional Reynolds analogy is no longer recommended and appropriate. More accurate turbulence flow and heat transfer model for the liquid metal lead-bismuth should be developed and applied in CFD simulation. In this paper, a specific CFD solver for simulating the flow and heat transfer of liquid lead-bismuth based on the k - 𝜀 - k_𝜃 - 𝜀_𝜃 model was developed based on the open source platform OpenFOAM. Then the advantage of proposed model was demonstrated and validated against a set of experimental data. Finally, the simulation of LBE turbulent flow and heat transfer in a 7-pin wire-wrapped rod bundle with the k - 𝜀 - k_𝜃 - 𝜀_𝜃 model was carried out. The influence of wire on the flow and heat transfer characteristics and the three-dimensional distribution of key thermal hydraulic parameters such as temperature, cross-flow velocity and Nusselt number were studied and presented. Compared with the traditional SED model with a constant Pr_t = 1.5 or 2.0, the k - 𝜀 - k_𝜃 - 𝜀_𝜃 model is more accurate on predicting the turbulence flow and heat transfer of liquid lead-bismuth. The average relative error of the k - 𝜀 - k_𝜃 - 𝜀_𝜃 model is reduced by 11.1% at most under the simulation conditions in this paper. This work is meaningful for the thermal hydraulic analysis and structure design of fuel assembly in the liquid lead-bismuth cooled fast reactor.

Design and analysis of a free-piston stirling engine for space nuclear power reactor

Dai, Zhiwen,Wang, Chenglong,Zhang, Dalin,Tian, Wenxi,Qiu, Suizheng,Su, G.H. Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.2

The free-piston Stirling engine (FPSE) has been widely used in aerospace owing to its advantages of high efficiency, high reliability, and self-starting ability. In this paper, a 20-kW FPSE is proposed by analyzing the requirements of space nuclear power reactor. A code was developed based on an improved simple analysis method to evaluate the performance of the proposed FPSE. The code is benchmarked with experimental data, and the maximum relative error of the output power is 17.1%. Numerical results show that the output power is 21 kW, which satisfies the design requirements. The results show that: a) reducing the pressure shell's thickness can improve the output power significantly; b) the system efficiency increases with the wire porosity, while the growth of system efficiency decreases when the porosity is higher than 80%, and system efficiency exhibits a linear relationship with the temperatures of the cold and hot sides; c) the system efficiency increases with the compression ratio; the compression ratio increases by 16.7% while the system efficiency increases by 42%. This study can provide valuable theoretical support for the design and analysis of FPSEs for space nuclear power reactors.