Conference Agenda

The S-Allegro is a state-of-the-art, electrically heated, downscaled Integral Test Facility (ITF) of the ALLEGRO Gas-Cooled Fast Reactor (GFR) demonstrator, operated by CVR in Pilsen, Czech Republic. The facility is designed to investigate operational states and transients of the ALLEGRO GFR demonstrator and to serve as a platform for testing innovative systems and components for the gas-cooled reactor technology. Additionally, it aims to generate experimental data for the validation and verification of thermal-hydraulic (TH) codes and models used in further ALLEGRO research and development.
Two of the most critical components of the S-Allegro ITF are the primary (gas-gas) and secondary (gas-water) heat exchangers. This paper presents a summary of the TH benchmarking activities conducted as part of the Euratom project SafeG. Participants from VUJE, HUN-REN EK, CVR, BME and NCBJ developed TH models of the primary and secondary S-Allegro heat exchangers using the RELAP5-3D, CATHARE2, DYMOLA, ATHLET and APROS codes, and performed a complex steady-state qualification. The outcomes of these activities, along with recommendations for further benchmarking, are discussed in this paper.

ALLEGRO is a demonstrator for the large GFR2400 gas-cooled fast reactor selected by the Generation IV International Forum (GIF). These have been under development in Europe for more than two decades. The primary aims of ALLEGRO are to demonstrate helium technology and to provide some technological background to test the new refractory fuel in a fast-spectrum environment. Two main core configurations are envisaged in ALLEGRO. The first is the so-called driver core, which consists of MOX or UOX fuel with stainless steel cladding. The second is the refractory core aiming to utilise SiC-SiC cladding and carbide fuel. In this study, we carry out thermal-hydraulics calculations for the new refractory core, which was proposed in the SafeG EU project. Two transients are investigated with the CATHARE thermal hydraulics code. First, a 200% break at the hot duct is initiated, which does not lead to loss of coolant but causes a serious core bypass. The second transient describes the evolution of a LOCA transient at the cold duct. The results are compared to the simulations carried out for the former refractory core.

Nuclear power generation has advantages, such as high energy density, reliable power supply, and a reduction in greenhouse gas emissions. However, the potential risk of nuclear accidents requires increased reliability. High Temperature Gas-cooled Reactor (HTGR) is a new generation of reactors that operate at high temperatures above 750°C. This high thermal energy can be used not only for power generation but also for industrial heat applications and hydrogen production. HTGR improves safety with a Reactor Cavity Cooling System (RCCS), which is a fully passive system requiring no external power or coolant. When the active cooling system of the reactor core is off, the RCCS transfers decay heat from the reactor core to the concrete walls of the reactor cavity. In the RCCS, a vertical rectangular riser duct surrounds the reactor vessel at a certain distance, and a chimney connects to the riser duct. The riser duct receives the decay heat from the reactor vessel and the rising air is released to the external atmosphere by natural circulation, maintaining the safe temperature of the reactor. During this process, most of the heat is transferred in the form of radiation. In this study, a preliminary validation of the radiation model comparison for radiative heat transfer analysis of the air-cooled RCCS of MHTGR is performed by Computational Fluid Dynamics (CFD) analysis.

Based on the efficient thermoelectric conversion capability of Stirling engines, the Autonomous Circulation Micro Integrated Nuclear Reactor (ACMIR)is highly integrated and lightweight, making it a favorable candidate for deep space exploration, manned spaceflight, and other projects. However, in demonstrating the applicability and safety of ACMIR across various application scenarios, challenges arise due to the lack of simulation calculation models and modeling methods that account for multi-parameter physical coupling. Therefore, this study considers the heat source structure of the reactor core integrated within the Stirling engine to establish a refined system thermodynamic model. Additionally, it establishes a dynamic model for the pistons, considering the reciprocating motion of the gas-distribution piston and power piston in the Stirling engine. Subsequently, the transient neutron dynamics and the mathematical differential equations for the electromagnetism of the moving-coil linear generator are coupled and solved, completing the multi-physical parameter coupling calculation for the "nuclear-thermal-mechanical-electrical" aspects of ACMIR. By selecting appropriate mathematical algorithms for model solving, preliminary characteristic analysis of ACMIR under different load conditions is conducted. The analysis results indicate that the established simulation model can basically align with the operational states of the space reactor system under different mission conditions. The developed model can serve as a research reference for the next step in system control.

The S-ALLEGRO Integral Test Facility (ITF) is a downscaled version of the ALLEGRO Gas-cooled Fast Reactor (GFR) demonstrator. Benchmarking the measurements conducted on this facility is crucial for the safe and effective development of ALLEGRO. The benchmark activities require participants to create thermal-hydraulic models of the whole S-ALLEGRO system. Since the system consists of several different and innovative components, the modelling approach is to look at the different heat exchangers, blowers, and pipelines and create a standalone model for each. If measured data is available for the separate components, the validation of the standalone models is essential to get reasonable calculations for the whole facility.

The modelling of the heat exchangers is probably the most critical part of the benchmark from the calculations point of view. One of these heat exchangers is a shell and tube type with U-shaped tubes and baffles inside it. It is called the Secondary Heat Exchanger (SHX) in S-ALLEGRO, and it has helium on the tube side and water on the shell side. Having baffles on the shell side can make the waterside flow pattern complex, so special attention has to be paid to its modelling by a 1D code.

In this paper, a special way of modelling such a heat exchanger in the ATHLET code is presented which is supported by standalone measurements.

High Temperature Gas-cooled Reactor (HTGR) with high outlet temperature from 700°C to 800 ~ 1000°C is expected to be widely used for heat supply, hydrogen production, steelmaking, seawater desalination, thermal recovery of heavy oil, coal liquefaction and gasification. The 10 MW High Temperature gas-cooled test Reactor (HTR-10), with outlet temperature of 700°C, had been constructed and operated in China as a pilot plant to demonstrate the inherent safety features of the modular HTGR. Supported by Chinese National S&T Major Project and National Key R&D Program of China, some research on HTGR technology with much higher outlet temperature is carried out.

This paper presents results obtained for two Beyond Design Basis Accidents: (1) control rod withdrawal ATWS and (2) control rod withdrawal ATWS combined with DLOFC. Within a cooperation between the Nuclear Research Group (NRG) of Netherland and Institute of Nuclear and new Energy Technology (INET), Tsinghua University of China analysis was performed with two different codes, TINTE code, a thermal-hydraulic design and accident analysis tool for the Pebble-bed High Temperature Gas-cooled Reactor (HTGR), and the SPECTRA code, a thermal-hydraulic analysis code developed at the NRG.

The performed calculations showed that the fuel temperature will stay below the acceptable limits set for the DBA (1620ºC) during the accidents. The results show feasibility to increase the outlet helium temperature of the HTR-10 to 950°C.

This study presents an experimental investigation on compact diffusion-bonded steam generators, namely Printed Circuit Steam Generator (PCSG) designed for high-temperature gas reactors. The thermal performance of the PCSG was evaluated utilizing the High-Temperature Helium Experimental Facility at the University of Michigan, which enables the characterization of single-phase and two-phase flow heat transfer in the PCSG’s mini-zigzag channels. In this study, two PCSGs with different flow channel design, i.e, straight channels and zigzag channels, were tested under helium-to-water/steam heat transfer setup. The heat transfer characteristics of both the PCSGs were analyzed based on the measured parameters, including the system pressure, mass flow rate, and temperature data. The averaged two-phase heat transfer coefficient inside the cold channels was found to vary with the vapor quality at the cold channel outlet. A sharp drop in the two-phase heat transfer coefficient was observed when the cold channel outlet vapor quality was around 0.5 – 0.6 due to local dry-out of the thin liquid film at the annular flow regime. In addition, the zigzag channel PCSG exhibited enhanced convective boiling heat transfer, with a higher heat transfer coefficient compared to the straight-channel PCSG in the high vapor quality region. However, in the low vapor quality region, significant measurement uncertainties were observed due to the high sensitivity of the evaluated heat transfer coefficient to the helium-side single-phase heat transfer coefficient. The findings from this study provide valuable insights into the design optimization of compact steam generators for next-generation small modular reactors and micro modular reactors.