The Next Generation Nuclear Plant (NGNP), a very High temperature Gas-Cooled Reactor (VHTR) concept, will provide the first demonstration of a closed-loop Brayton cycle at a commercial scale, producing a few hundred megawatts of power in the form of electricity and hydrogen. The power conversion unit (PCU) for the NGNP will take advantage of the significantly higher reactor outlet temperatures of the VHTRs to provide higher efficiencies than can be achieved with the current generation of light water reactors. Besides demonstrating a system design that can be used directly for subsequent commercial deployment, the NGNP will demonstrate key technology elements that can be used in subsequent advanced power conversion systems for other Generation IV reactors. In anticipation of the design, development and procurement of an advanced power conversion system for the NGNP, the system integration of the NGNP and hydrogen plant was initiated to identify the important design and technology options that must be considered in evaluating the performance of the proposed NGNP. As part of the system integration of the VHTRs and the hydrogen production plant, the intermediate heat exchanger is used to transfer the process heat from VHTRs to the hydrogen plant. Therefore, the design and configuration of the intermediate heat exchanger is very important. This paper will include analysis of one stage versus two stage heat exchanger design configurations and simple stress analyses of a printed circuit heat exchanger (PCHE), helical coil heat exchanger, and shell/tube heat exchanger.

Energy is at the top of the political agenda. Energy technology, including nuclear, has a role to play in addressing the associated challenges. The Euratom research programme in the area of nuclear fission is contributing to the development of innovative reactor concepts that are more sustainable, intrinsically safe and proliferation resistant, economic and capable of co-generating both electricity and process heat. The need for such Generation-IV reactors is also stressed in the EU’s recently published Strategic Energy Technology Plan, and the development of these advanced systems is the focus of the Sustainable Nuclear Energy Technology Platform established in September 2007, which is at the cornerstone of the EU’s strategy in nuclear R&D and brings together all the key European nuclear research and industrial players in this field. The paper provides a summary of the present situation and prospects in these areas, with an emphasis on V/HTR technology.

Loss of primary coolant flow test is under planning by using the High Temperature engineering Test Reactor (HTTR). In this test, all the gas circulators are tripped and the position of all control rods keeps its initial one. The new calculation model was developed to perform the preliminary analysis for the test. This model is so improved that an equivalent fuel channel model based on one point kinetics code and a whole reactor model based on two-dimensional thermal analysis code are coupled to simulate the reactor performance during the loss of coolant flow. Both calculation codes were used in the safety evaluation of the HTTR licensing. The improved calculation model was validated by comparison between the calculated result and the experimental one obtained from the coolant flow reduction test in the HTTR. The loss of primary coolant flow test simulates the depressurization accident and the data obtained from the test is useful for the validation and improvement of the calculation code applied to the safety analysis in the future HTGR such as Very High Temperature Reactor which is selected as one of candidates of the generation IV reactor system.

Simulation of some fluid phenomena associated with Generation IV reactors requires the capability of modeling mixing in two- or three-dimensional flow. At the same time, the flow condition of interest is often transient and depends upon boundary conditions dictated by the system behavior as a whole. Computational fluid dynamics (CFD) is an ideal tool for simulating mixing and three-dimensional flow in system components, whereas a system analysis tool is ideal for modeling the entire system. This paper presents the reasoning which has led to coupled CFD and systems analysis code software to analyze the behavior of advanced reactor fluid system behavior. In addition, the kinds of scenarios where this capability is important are identified. The important role of a coupled CFD/systems analysis code tool in the overall calculation scheme for a Very High Temperature Reactor is described. The manner in which coupled systems analysis and CFD codes will be used to evaluate the mixing behavior in a plenum for transient boundary conditions is described. The calculation methodology forms the basis for future coupled calculations that will examine the behavior of such systems at a spectrum of conditions, including transient accident conditions, that define the operational and accident envelope of the subject system. The methodology and analysis techniques demonstrated herein are a key technology that in part forms the backbone of the advanced techniques employed in the evaluation of advanced designs and their operational characteristics for the Generation IV advanced reactor systems.