Analysis of the Thermal-Hydraulic and Mechanical Performance of Intermediate Heat Exchangers Used in Systems With Very High Temperature Nuclear Reactors

Intermediate heat exchangers are one of the most critical devices in the safety of facilities with very high temperature nuclear reactors. In this application, the printed circuit heat exchanger (PCHE) design has been the one that has shown the greatest advantages in terms of heat transfer, compaction and structural strength. In this work, a thermal-hydraulic and mechanical model of the PCHE was developed using computational fluid dynamics (CFD) techniques and finite element methods, respectively. The CFD model was validated by comparison with experimental data and empirical correlations of Nusselt number and friction factor published by other authors. A methodology was proposed to evaluate the operation of the exchanger based on the analysis of capital and operating costs. As a relevant aspect of this methodology, the relationship between the maximum Von Misses stress in the structure and the time of operation was considered. In the structural calculations it was observed that increasing the temperature gradient between the channels caused by the increase of the mass flows of gases, causes the displacement of the solid region and the maximum stress increase. The Taguchi method was applied to identify the dimensions that have the greatest influence on the operation of the PCHE and to obtain an optimized design of the device.

Additive Manufacturing of Spent Fuel Storage Rack Model by Selective Laser Melting

Additive Manufacturing (AM) can fabricate 3D complex functional parts, which can reduce material waste and increase manufacturing efficiency significantly. These benefits make AM technique suitable for some critical industry applications. Confident application of the AM technique requires whole understanding of AM parts’ properties. Safety and economics are essential to nuclear power plant. In this study, an innovative 316L stainless steel spent fuel storage rack with integrative structure was designed, and a small model of this rack was fabricated by selective laser melting (SLM), mechanical properties of the 316L stainless steel manufactured by SLM technique are studied and discussed. Key technical issues of application of AM to manufacturing nuclear parts are also discussed.

Research and Practice of Nuclear Equipment Procurement and Supply Management Based on “Internet+” Mode

The procurement and supply management of nuclear power equipment are carried out offline in tradition, the mode is mono and inflexible, the organizational process assets is poorly accumulated, the implementation process is not transparent, the tracking control effect is poor, the efficiency is relatively low. Adopting the information technology has acquired certain effect in improving organizational process assets accumulation although, the information is fragmented, efficiency improvement is not obvious. Study on “Internet + Procurement” was carried out since the year of 2015, compared with the traditional model, all the business and supplier information of procurement management of nuclear power equipment were interconnected, the whole process of procurement and supply management are through the internet, it is a innovation of the procurement and supply management mode using information technology. a wealth of data was accumulated for further analysis, an interconnected network was built between buyer and supply, significant change was brought out to the field of nuclear power equipment procurement, the role of this synergistic effect, contributes to the shortening of the construction period of nuclear power projects.

Thermal Analysis Safety Margins Using ABAQUS for the MP-2 Experiment in the Advanced Test Reactor

The Mini-Plate 2 (MP-2) irradiation test is a fueled experiment designed for irradiation in multiple test locations in the Advanced Test Reactor (ATR). MP-2 is considered a non-instrumented drop-in test where small aluminum-clad fuel plate samples are cooled directly by the ATR Primary Coolant System (PCS) water. The MP-2 fuel plate experiment will be irradiated in several different irradiation locations of the ATR. This fueled experiment contains aluminum-clad fuel mini plates consisting of monolithic U-Mo. Four different types of fuel plates were analyzed. A thermal analysis has been performed on the MP-2 experiment to be irradiated in the ATR at Idaho National Laboratory (INL). A new technique for calculating Departure from Nucleate Boiling Ratio (DNBR) and Flow Instability Ratio (FIR) using the commercial finite element and heat transfer code ABAQUS is demonstrated. This new technique calculates DNBR for the fuel plate surfaces and FIR for all water components for each finite element surface and node. Pressure drop data is fed into the calculations in order to geometrically calculate the water saturation temperature. Results from the DNBR and FIR calculations are displayed with the ABAQUS post processor named Viewer. By calculating these parameters at each location in the finite element model, conservatism is replaced with accuracy. This allows for a greater margin for the thermal hydraulic safety parameters.

The Features of Closed Brayton Cycle and Sub-Critical Combined Cycles Coupled With (Very) High Temperature Gas-Cooled Reactor

High Temperature Gas-cooled Reactor (HTR) and Very High Temperature Gas-cooled Reactor (VHTR), are the most promising and achievable fourth-generation nuclear reactor for its inherent safety. In this paper, the performance of Closed Brayton Cycle (CBC) and two sub-critical combined cycles were investigated and compared. The CBC is a recuperated and inter-cooling closed Brayton cycle. Two combined cycles include the sub-critical Rankine cycle without steam reheating (Simple Combined Cycle, SCC) and a sub-critical reheated Rankine cycle (Reheated Combined Cycle, RCC). The topping cycles of SCC and RCC are both a simple Brayton cycle, and connect with the bottoming cycles by a sub-critical heat recovery steam generator (HRSG). Physical and mathematical models of three different thermodynamic cycles were established. Within the temperature range of the HTR and VHTR, the effects and mechanism of key parameters, such as reactor outlet temperature, steam temperature and pressure, on features of three different cycles were investigated. The results showed the elevated reactor outlet temperature could obviously enhance efficiency of three cycles. The results showed that RCC had the highest efficiency while SCC had the lowest efficiency, and the efficiency of CBC is slightly lower than that of RCC. The results could be helpful to understand and develop the power conversion system coupled with (V)HTR in the future.

Analysis of Loss-of-Flow Accidents in Pre-Cooler and Inter-Cooler of HTR-10GT

Closed Brayton cycle (CBC) coupled with High Temperature Gas-cooled Reactor (HTGR) has potential application due to its compact configuration, high power generation efficiency and inherent safety. It is also one of the major power conversion methods for Generation IV advanced nuclear power systems. The typical CBC has several helium-water heat exchangers, including pre-cooler and inter-cooler. These helium-water heat exchangers have important influence on the performance of power conversion system, especially in loss-of-flow accidents (LOFAs). A system model including the reactor and the energy conversion system was established in this paper. The 10MW High Temperature Gas-cooled reactor-test Module helium Gas Turbine (HTR-10GT) was taken as the example to show the consequences of LOFAs. The results showed that LOFAs led to the rising of water temperature out of heat exchangers. With the reduction of water flow rate, the maximum water temperature would increase sharply, and the water temperature in pre-cooler was higher than that in inter-cooler. At low water flow rate, the water temperature would exceed the boiling point. LOFAs also made the rising of helium temperature. It had impacts on the performance of helium compressors. The elevated inlet temperature of helium compressors changed the corrected speed and corrected flow rate, therefore caused the deterioration of compressor’s performance. Furthermore, the LOFAs caused the reactor inlet temperature increasing. In low water flow rate, it would make the reactor inlet temperature beyond the temperature limitation of reactor pressure vessel and influence the safety of reactor. And the LOFAs also reduced the output work of cycle. This paper provides insights of features of CBC in LOFAs and will be helpful to the design and safety operation of closed Brayton cycle coupled with HTGR.