High Temperature Gas-cooled Reactor Plant Research Articles

Design analysis of a printed circuit heat exchanger for HTGR using a 3D finite elements model

The intermediate heat exchanger (IHX) is a critical component in most of the High Temperature Gas-cooled Reactor plant layouts. The printed circuit heat exchanger (PCHE) has been studied as an advantageous candidate for this application last years. The extreme operating conditions of the IHX and the higher structural integrity of the PCHE over other heat exchanger technologies have motivated these studies. In this work, the structural assessment of a zigzag channel PCHE under IHX conditions was carried out by means of computational fluid dynamics (CFD) and Finite Element Analysis (FEA). The ASME Boiling and Pressure Vessel Code, Sections III and VIII, were applied to evaluate the safety operation of the PCHE. The temperature field in steady state was obtained using a 3D CFD model. Then it was used as thermal load to calculate thermal stress using a FEA model. The influence of the zigzag angle and the bend radius on the thermal and mechanical stresses was analyzed. Thus, the values studied for the zigzag angle were 0° (straight), 15°, 30 and 45°, while for the bend radius were 0.5 mm, 2 mm, 4 mm and 6 mm. The importance of the 3D temperature field to calculate thermal stress in a zigzag channel was evidenced due to the appearance of local temperature differences in the channel cross-sections. The change in flow direction created areas of higher local temperatures where the flow lines change their directions impacting the channel wall. A significant thermal stress concentration was found in these areas, and this effect intensified with increasing zigzag angle. The bend radius was decisive to reduce the thermal stress concentration when the zigzag angle was higher than 15°.

Power-pressure coordinated control of modular high temperature gas-cooled reactors

Modular high temperature gas cooled reactor (mHTGR) adopts helium as coolant and graphite as both the moderator and structural material, whose fuel element is built by embedding thousands of TRISO coated particles into a graphite matrix. Since helium is an ideal gas, the dynamics of reactor coolant temperature is tightly coupled with that of primary pressure. To further improve the operation performance, it is meaningful to develop the coordinated control of power-level and primary pressure for mHTGRs. In this paper, a passivity-based power-pressure coordinated control is proposed for mHTGRs, and the sufficient condition for globally asymptotic closed-loop stability is given. This newly-built power-pressure coordinated control law is applied to the mHTGR of six-modular high temperature gas-cooled reactor plant HTR-PM600, and simulation results in the case of power ramping decrease and increase as well as performance comparison not only verify the theoretical result about closed-loop stability but also illustrate the influence of control parameters as well as the stronger performance relative to the classical proportional integral/differential (PI/D) controller.

Mechanical Reliability Analysis of Control Rod Assembly for HTR-PM

The article analyzes the forces and weaknesses of the control rod assembly in the high temperature gas cooled reactor (HTGR) demonstration project-pebble bed modular HTGR plant (HTR-PM), using the the mechanical reliability theory and mechanical simulation, and identifies the weaknesses at the connection of the control rod assembly. In the reliability-life calculation method based on the probabilistic fracture mechanics, in combination with the influence of irradiation on the fatigue life model and in consideration of the stress, the authors establish the reliability-life model to calculate the reliability-life of HTR-PM control rod assembly at the irradiation attenuation coefficient of 0.7. The calculation results show that at the reliability level of 0.99, the life of HTR-PM control rod assembly is about 150,000 action cycles, and that the irradiation has a great influence on the life of the control rod assembly. In view of what mentioned above, this paper can provide reference for the equipment reliability management of the HTR-PM control rod assembly.

Safety Investigation of Hazardous Materials Released from the Combined High Temperature Gas Cooled Reactor – Hydrogen Production Plant Using ALOHA Software

Safety Investigation of Hazardous Materials Released from The Combined High Temperature Gas Cooled Reactor (HTGR) – Hydrogen Production Plant Using ALOHA software has been carried out. Currently, most of studies for HTGR-hydrogen plant are focused only on the impact of hydrogen presence to the HTGR plant safety. Therefore, the objective of this study was to investigate the effect of the presence of natural gas and synthetic gas from Steam Methane Reforming hydrogen plant on the combine HTGR-Hydrogen production system using ALOHA software. Three selected hazardous materials: CH 4 , CO and H 2 were analyzed. The selected potential hazards of the hazardous materials after leaking from the pipe were downwind suffocation/toxication, flammable area and blast area from vapor cloud explosion. Two types of parameter, i.e., meteorological dispersion (including wind speed, temperature, humidity, nuclear building air changes for day and night) and source release parameters (including pipeline length, and distance from the reactor building to the hydrogen plant), were selected for this study. The effects of the parameters on the hazard distance were then analyzed. The study shows that hydrogen detector needs to be installed at the plant to ensure safety of field operator. Furthermore, CO adsorber and H 2 recombiner should be installed at the Reactor HVAC system for CO poisoning and H 2 fire protection. Provision of a separation distance of more than 250 meters or construction of a blast barrier between the reactor building and the hydrogen plant is also recommended to protect the reactor from H 2 explosion hazard.

Dynamical modeling and simulation of the six-modular high temperature gas-cooled reactor plant HTR-PM600

The modular high temperature gas-cooled reactor (MHTGR) is a typical small modular reactor (SMR) with inherent nuclear safety, which has a low average power density about 3 MW/m3, uses TRISO particle based fuel elements, and adopts helium as coolant and graphite as both moderator and structure material. Based upon the multimodular scheme, i.e. multiple MHTGR-based nuclear steam supply system (NSSS) modules driving a single turbine/generator system, the MHTGR-based nuclear power plants (NPPs) with inherent safety as well as any desired power ratings can be realized. However, since multiple NSSS modules are coupled together by the common secondary-circuit, the operation and control of multimodular NPPs is quite different from those classical single modular NPPs. For the design and verification of operation and control strategies of a NPP, it is necessary to develop the dynamical model of this NPP, and to study its open-loop and closed-loop dynamics. In this paper, a lumped-parameter dynamical model of the six-modular MHTGR-based NPP HTR-PM600 is proposed based on the conservation laws of mass, energy and momentum, which is composed of the MHTGR-based NSSS modules and the common secondary-circuit. Then, both the open-loop responses to exterior disturbances and the closed-loop behavior in power-level maneuver are given by numerical simulation, which shows not only the feasibility of this model but also the coupling effect of the modules caused by the common secondary-circuit.

OS7-09 高温ガス炉用再生熱交換器の開発(OS7 高温・高效率・発電・エネルギー貯蔵技術)

This paper describes a feasibility study of the Recuperator with fine off-set plate fin in High-temperature gas-cooled reactor plant. This fin is made by pressing process. If the fin pitch is too tight.it is possible to have some cracks in pressed fin plate. And the lower limit size of the fin pitch is about 1.2mm. The heat exchange and presseure loss characteristics in fine off-set fin plate are estimated with experimental equations. For the compactness of Recuperator, required heat exchanger density is more than 24MW/m^3, and required total core pressure loss is less than 100kPa. When thickness 0.2mm plate is pressed by a die with 1.2mm fin height, measured fin height is from 1.0mm to 1.2mm. If fin height is 1.0mm, total core pressure loss might be more than 100kPa. It is important to find proper pressing condition and keep fin height for getting higher heat exchanger characteristics and lower total core pressure loss.

Analyses of static energy conversion systems for small nuclear power plants

Small, Gas Cooled Reactor (GCR) nuclear power plants with static energy conversion could meet the energy mix in underdeveloped countries, including electricity, residential and industrial space heating, seawater desalination, and/or high temperature process heat or steam for industrial uses. Analyses are performed of one high-temperature GCR and three intermediate-temperature GCR power plants with co-generation options to calculate and compare the total energy utilizations of the plants. Conversion of the reactor thermal power to electricity is accomplished using static energy conversion units comprised of an Alkali Metal Thermal-To-Electric topping cycle and a Thermoelectric bottom cycle, as well as segmented thermoelectric converters. The calculated total energy utilization of these plants of 95% includes low-voltage (∼ 400 V) DC electrical power at a net plant efficiency of up to 32.7%, and co-generation for residential and industrial space heating at < 400 K, desalination at 365 K, and/or high temperature process heat or steam at > 500 K. Results indicate that the intermediate-temperature GCR plants with a small coolant temperature rise of 100 – 160 K in the reactor core could deliver 2–3 times the electricity of a GCR plant with a higher temperature rise of 340 K, and that all plants could provide for both seawater desalination and high temperature process heat/steam concurrently. In addition, depending on the number of co-generation options employed, the fraction of the reactor thermal power used for co-generation in the high-temperature GCR plant is ∼ 10–30% higher than in the intermediate-temperature GCR plants with a smaller temperature rise in the reactor.

A Simulation Model of the Fort St. Vrain HTGR Plant Suitable for Control and Related Studies

The paper discusses a simulation model of the Fort St. Vrain High-Temperature Gas-Cooled Reactor (HTGR), which has been developed to aid in the startup of the plant and in commissioning the plant control systems. The mathematical equations have been solved using both digital and hybrid computing techniques. These methods are discussed in the paper and in particular the usefulness of a System Simulation Language (SYSL) is emphasized. The features of the control system are discussed and typical plant control transients are given.

Effect on non-linear soil-structure interaction due to base slab uplift on the seismic response of a high-temperature gas-cooled reactor (HTGR)

In high seismic regions it has often been the practice to use oversized base slabs for the major nuclear power plant structures in order to prevent, or at least minimize, the amount of dynamic base slab uplift which will result from the overturning moments developed during seismic ground motion. Two major reasons have been expressed as to why dynamic base slab uplift should be minimized: (1) As nuclear power plants are normally designed for seismic loadings based upon linear analysis, and since soil-structure interaction becomes nonlinear when only a portion of the base slab is in contact with the soil, linear elastic analysis may be unacceptable if base slab uplift occurs (as the resultant design loads may be incorrect), and (2) substantial uplift could cause excessive toe pressures in the supporting soil and significant impact forces when the slab recontacts the soil. The primary purpose of this paper is to evaluate the importance of the nonlinear soil-structure interaction effects resulting from substantial base slab uplift occurring during a seismic excitation. The structure considered for this investigation consisted of the containment building and prestressed concrete reactor vessel (PCRV) for a typical HTGR plant. A simplified dynamic mathematical model was utilized consisting of a conventional lumped mass structure with soil-structure interaction accounted for by translational and rotational springs whose properties are determined by elastic half space theory. Three different site soil conditions (a rock site, a moderately stiff soil, and a soft soil site) and two levels of horizontal ground motion (0.3 and 0.5 g earthquakes) were considered. Based upon the parametric cases analyzed in this investigation, it may be concluded that linear analysis (which ignores the nonlinear soil-structure interaction effects of base slab uplift) can be used to conservatively estimate the important behavior of the base slab even under conditions of substantial base slab uplift. For all cases investigated here, linear analysis resulted in higher base overturning moments, greater toe pressures, and greater heel uplift distances than nonlinear analyses. It may also be concluded that the nonlinear effect of uplift does not result in any significant lengthening of the fundamental period of the structure. Also, except in the short period region (period less than half of the fundamental period) only negligible differences exist between in-structure response spectra based on linear analysis and those based on nonlinear analysis. Finally, it may be concluded that for sites in which soil-structure interaction is not significant, as for the rock site, the peak structural response (shears and moments) at all locations above the base mat are not significantly influenced by the nonlinear effects of base slab uplift. However, for the two soil sites the peak shears and moments are, in a few instances, significantly different between linear and nonlinear analyses. As a result, linear analysis may be used to determine all structural response for rock sites even when there is substantial base slab uplift. However, for soil sites, nonlinear analyses are necessary if substantial base slab uplift occurs.

HTGR plant safety design bases

The high temperature gas-cooled reactor (HTGR) has inherent and design safety features that are sifnificant and unique, requiring a number of safety criteria and approaches that differ markedly from other reactor types. This paper briefly reviews the design of HTGR plants that have been built and are being offered in the United States. It then reviews the safety considerations involved in the design of the plants being offered. The unique features, their development, and their effects on safety criteria are described. The design bases of the prestressed concrete reactor vessel (PCRV) are given particular attention. Operating characteristics of the HTGR and plant response to transient conditions are discussed. The design-basis depressurization accident evolution and related HTGR safety requirements are discussed. Characteristics of the HTGR with respect to technical specifications are discussed, with particular emphasis on the PCRV and the core safety limit.

HTGR plant dynamics analysis under accident conditions

Important features of high temperature gas-cooled reactor (HTGR) systems related to plant dynamics and accident analysis are discussed. Because of the basic simplicity of the HTGR system, it is possible to analyze the full reactor plant (core, helium circulators, steam generators and reheaters, feedwater controls, turbine controls, and plant protective action) in a single computer code. Representative dynamics analysis is presented for the Fort St. Vrain Power Station.

Statatus of the High-Temperature Gas-Cooled Reactor

General Atomic's High-Temperature Gas-Cooled Reactor (HTGR) is discussed, with special reference to the 40-Mwe (electrical megawatts) prototype HTGR plant to be built at Peach Bottom, Pa. This plant will supply steam at 1,000 F (degrees Fahrenheit) and 1450 psia (pounds per square inch absolute); will have a fuel burn-up of 50,000 mw-davs per tonne; and will operate for 3 years without refueling, Recent developments in the prototype HTGR prograni include coated fuel particles, the single-wall fuel element, and measurements of the Doppler coefficient, using the HTGR critical assembly.