Neutronics and thermal hydraulic coupling analysis of integrated pressurized water reactor

Khan, Salah Ud‐Din; Peng, Minjun; Khan, Shahab Ud‐Din

doi:10.1002/er.2981

Cited by 26 publications

(6 citation statements)

References 9 publications

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“…Brief Description of IP200. IP200 is a newly developed IPWR by Harbin Engineering University, whose major design parameters could be referenced from previous papers [8][9][10][11]. Its main purpose is to generate electricity for the floating nuclear power plant (NPP).…”

Section: System Design and Modelingmentioning

confidence: 99%

Assessment of Severe Accident Management for Small IPWR under an ESBO Scenario

Peng

2019

Science and Technology of Nuclear Installations

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Interest in evaluation of severe accidents induced by extended station blackout (ESBO) has significantly increased after Fukushima. In this paper, the severe accident process under the high and low pressure induced by an ESBO for a small integrated pressurized water reactor (IPWR)-IP200 is simulated with the SCDAP/RELAP5 code. For both types of selected scenarios, the IP200 thermal hydraulic behavior and core meltdown are analyzed without operator actions. Core degradation studies firstly focus on the changes in the core water level and temperature. Then, the inhibition of natural circulation in the reactor pressure vessel (RPV) on core temperature rise is studied. In addition, the phenomena of core oxidation and hydrogen generation and the reaction mechanism of zirconium with the water and steam during core degradation are analyzed. The temperature distribution and time point of the core melting process are obtained. And the IP200 severe accident management guideline (SAMG) entry condition is determined. Finally, it is compared with other core degradation studies of large distributed reactors to discuss the influence of the inherent design characteristics of IP200. Furthermore, through the comparison of four sets of scenarios, the effects of the passive safety system (PSS) on the mitigation of severe accidents are evaluated. Detailed results show that, for the quantitative conclusions, the low coolant storage of IP200 makes the core degradation very fast. The duration from core oxidation to corium relocation in the lower-pressure scenario is 53% faster than that of in the high-pressure scenario. The maximum temperature of liquid corium in the lower-pressure scenario is 134 K higher than that of the high-pressure scenario. Besides, the core forms a molten pool 2.8 h earlier in the lower-pressure scenario. The hydrogen generated in the high-pressure scenario is higher when compared to the low-pressure scenario due to the slower degradation of the core. After the reactor reaches the SAMG entry conditions, the PSS input can effectively alleviate the accident and prevent the core from being damaged and melted. There is more time to alleviate the accident. This study is aimed at providing a reference to improve the existing IPWR SAMGs.

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Section: System Design and Modelingmentioning

confidence: 99%

Assessment of Severe Accident Management for Small IPWR under an ESBO Scenario

Peng

2019

Science and Technology of Nuclear Installations

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“…In order to correctly assess the temperature distribution of the BU, an accurate convective HTC should be required. The HTC correlations are determined by Nusselt ( Nu ), Reynolds ( Re ) and Prandtl ( Pr ) numbers . For helium coolant flow in the cooling channel, the Reynolds number is defined as

R e = \frac{ρ normalv D_{H}}{μ}

where D H is the hydraulic diameter of the cooling channel; v is the mean velocity of the helium coolant; ρ is the density of the helium coolant; μ is the dynamic viscosity of the helium coolant.…”

Section: Thermal Hydraulic Calculationmentioning

confidence: 99%

Thermal hydraulic analysis of the HECLIC blanket breeder unit for CFETR

Lei

Song

2014

Int. J. Energy Res.

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SUMMARYChina Fusion Engineering Test Reactor (CFETR) is proposed by the China National Integration Design Group. A heliumcooled lithium ceramic (HECLIC) blanket for CFETR has been designed by Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) and University of Science and Technology of China (USTC). The HECLIC blanket includes the tungsten armor, first wall (FW), breeder units (BUs), caps, stiffening grids and back plates. The BU consists of the beryllium pebble beds, lithium ceramic pebble beds and cooling plates. The stiffening grid reinforces the blanket structure and also cools the BU together with cooling plates. Thermal hydraulic analysis has been performed to assess the heat transfer coefficient (HTC) and temperature distribution. The results indicate that the velocity is fairly uniform in the cooling loops, and the HTC is high enough to remove the heat timely. The maximum temperatures are within the design temperature limits. Besides, optimization has been done to improve the layout of the cooling channels. In addition, thermal mechanical analysis has been carried out. The maximum stress can satisfy design limits, and it proves the feasibility of the design.

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“…The effective layout of components makes the IPWR have a compact structure. The detailed design parameters of IPWR are shown in the paper of Salah et al . Major design parameters of IPWR are summarized in Table .…”

Section: Research Modelmentioning

confidence: 99%

Analysis of load-following characteristics for an integrated pressurized water reactor

Xia

Peng

2013

Int. J. Energy Res.

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SUMMARYIntegrated pressurized water reactor (IPWR) usually be equipped with once-through steam generators (OTSGs). The OTSG has many advantages such as simple mechanical structure, smaller size, and higher heat transfer efficiency. It produces superheated steam but with less inventory in its secondary side. The steam pressure is easily affected by steam flow rate or feed water flow rate. This draws more attention to design advanced reactor control system. In this paper, a study has been carried out to analyze the thermal hydraulic performance of an advanced IPWR under steady-state and transient conditions by using a thermal hydraulic safety analysis code Relap5. An effective load-following control system is proposed. The steady-state operating characteristics of IPWR at different load conditions show that the average primary coolant temperature, steam pressure, and coolant mass flow rate are the most important control parameters. Pump frequency conversion strategy and OTSG grouping run strategy are used to study the transient operating characteristics of IPWR. Simulation results of the control system demonstrate its capability in regulating feedwater flow rate and reactor power to follow the change of steam flow rate. According to the results, the OTSG grouping run strategy is optimized to ensure the OTSG operates safely under low load conditions.

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Neutronics and thermal hydraulic coupling analysis of integrated pressurized water reactor

Cited by 26 publications

References 9 publications

Assessment of Severe Accident Management for Small IPWR under an ESBO Scenario

Assessment of Severe Accident Management for Small IPWR under an ESBO Scenario

Thermal hydraulic analysis of the HECLIC blanket breeder unit for CFETR

Analysis of load-following characteristics for an integrated pressurized water reactor

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