Won Jae Lee

A STUDY OF A NUCLEAR HYDROGEN PRODUCTION DEMONSTRATION PLANT

The current energy supply system is burdened by environmental and supply problems. The concept of a hydrogen economy has been actively discussed worldwide. KAERI has set up a plan to demonstrate massive production of hydrogen using a VHTR by the early 2020s. The technological gap to meet this goal was identified during the past few years. The hydrogen production process, a process heat exchanger, the efficiency of an I/S thermochemical cycle, the manufacturing of components, the analysis tools of VHTR, and a coated particle fuel are key areas that require urgent development. Candidate NHDD plant designs based on a 200 MWth VHTR core and I/S thermochemical process have been studied and some of analysis results are presented in this paper.

PERSPECTIVES OF NUCLEAR HEAT AND HYDROGEN

Nuclear energy plays an important role in world energy production by supplying 6% of the world’s current total electricity production. However, 86% of the energy consumed worldwide to produce industrial process heat, to generate electricity and to power the transportation sector still originates in fossil fuels. To cope with dwindling fossil fuels and climate change, it is clear that a clean alternative energy that can replace fossil fuels in these sectors is urgently required. Clean hydrogen energy is one such alternative. Clean hydrogen can play an important role not only in synthetic fuel production but also through powering fuel cells in the anticipated hydrogen economy. With the introduction of the high temperature gas-cooled reactor (HTGR) that can produce nuclear heat up to 950℃ without greenhouse gas emissions, nuclear power is poised to broaden its mission beyond electricity generation to the provision of nuclear process heat and the massive production of hydrogen. In this paper, the features and potential of the HTGR as the energy source of the future are addressed. Perspectives on nuclear heat and hydrogen applications using the HTGR are discussed.

THE BENCHMARK CALCULATIONS OF THE GAMMA+ CODE WITH THE HTR-10 SAFETY DEMONSTRATION EXPERIMENTS

KAERI (Korea Atomic Energy Research Institute) has developed the GAMMA+ code for a thermo-fluid and safety analysis of a VHTR (Very High Temperature Gas-Cooled Reactor). A key safety issue of the VHTR design is to demonstrate its inherent safety features for an automatic reactor power trip and power stabilization during an anticipated transient without scram (ATWS) accident such as a loss of forced cooling by a trip of the helium circulator (LOFC) or a reactivity insertion by a control rod withdrawal (CRW). This paper intends to show the ATWS assessment capability of the GAMMA+ code which can simulate the reactor power response by solving the point-kinetic equations with six-group delayed neutrons, by considering the reactivity changes due to the effects of a core temperature variation, xenon transients, and reactivity insertions. The present benchmark calculations are performed by using the safety demonstration experiments of the 10 MW high temperature gascooled- test module (HTR-10) in China. The calculation results of the power response transients and the solid core temperature behavior are compared with the experimental data of a LOFC ATWS test and two CRW ATWS tests by using a 1mk-control rod and a 5mk-control rod, respectively. The GAMMA+ code predicts the power response transients very well for the LOFC and CRW ATWS tests in HTR-10.

Computational Fluid Dynamics Assessment of the Local Hot Core Temperature in a Pebble-Bed Type Very High Temperature Reactor

Assessment of the local hot core temperature during normal operation in a pebble-bed type very high temperature reactor has been carried out by using the computational fluid dynamic (CFD) method for which the boundary conditions were obtained from the results of a macroscopic analysis of the core using a system thermal analysis code, GAMMA. Three pebble arrangements are selected, which are simple cubic (SC), body-centered cubic, and face-centered cubic. The results showed that the SC arrangement having the lowest porosity gives the highest fuel temperature of 1237°C but still below the normal operational fuel limit of 1250°C. Comparison of the CFD results with an empirical correlation was made for the pressure drop and Nusselt number. Both results showed a similar tendency that the pressure drop and the Nusselt number increases as the porosity decreases but there were large differences in their absolute values. The benchmark calculation for the pressure drop of the packed particles in a square channel indicated that the correlation for the full core used in the system code is not appropriate for the prediction of a local thermal-fluid behavior in an ordered pebble arrangement.

CFD Assessment of the Local Hot Core Temperature in a Pebble-Bed Type Very High Temperature Reactor

Assessment of the local hot core temperature during normal operation in a pebble-bed type of Very High Temperature Reactor (VHTR) has been carried out by using the Computational Fluid Dynamic (CFD) method for which the boundary conditions were obtained from the results of a macroscopic analysis of the core using a system thermal analysis code, GAMMA. Three pebble arrangements are selected, which are Simple Cubic (SC), Body-Centered Cubic (BCC), and Face-Centered Cubic (FCC). Results showed that the SC arrangement having the lowest porosity gives the highest fuel temperature of 1237°C but still below the normal operational fuel limit of 1250°C. Comparison of the CFD results with an empirical correlation was made for the pressure drop and the Nusselt number but there were large differences between them. The benchmark calculation of a pressure drop for packed particles in a square channel indicated that the correlation for the full core used in the system code is not appropriate for the prediction of a local thermal fluid behavior.Copyright

Experimental Validation for Specific Thermal Hydraulic Phenomena of SMART

SMART (System-integrated Modular Advanced ReacTor) is an integral light water reactor developed by KAERI for both electricity generation and various applications. SMART can produce 90MWe of electricity and 40,000 tons/day of sea-water desalination, which is deemed suitable for a city population of 100,000. The SMART project aims at standard design approval. The experimental validation for SMART specific thermal hydraulic phenomena is to generate background information required for SMART licensing.The objective of the experimental validation is to demonstrate the safety and performance of the SMART-specific system and component designs, and to provide a database for the validation of the design tools, such as a system analysis code. From the experimental results, it is proved that SMART was designed properly and the data are used to approve the validation of the system analysis code. Also, the experimental results were submitted to a regulatory body as technical background information of the SMART standard design, and they enhance the licensing possibility of this design.Copyright