ASME Code Considerations for the Compact Heat Exchanger

Control of Advanced Reactor-Coupled Heat Exchanger System: Incorporation of Reactor Dynamics in System Response to Load Disturbances

A new high-temperature multi-fluid, multi-loop test facility for advanced nuclear applications is under development at the Idaho National Laboratory. The facility will include three flow loops: high-temperature helium, molten salt, and steam/water. Molten salts have been identified as excellent candidate heat transport fluids for primary or secondary coolant loops, supporting advanced high temperature and small modular reactors (SMRs). Details of some of the design aspects and challenges of this facility, which is currently in the conceptual design phase, are discussed. A preliminary design configuration will be presented, with the required characteristics of the various components. The loop will utilize advanced high-temperature compact printed-circuit heat exchangers (PCHEs) operating at prototypic intermediate heat exchanger (IHX) conditions. The initial configuration will include a high-temperature (750°C), high-pressure (7 MPa) helium loop thermally integrated with a molten fluoride salt (KF-ZrF4) flow loop operating at low pressure (0.2 MPa) at a temperature of ∼450°C. Experiment design challenges include identification of suitable materials and components that will withstand the required loop operating conditions. Corrosion and high temperature creep behavior are major considerations. The facility will include a thermal energy storage capability designed to support scaled process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will also provide important data for code verification and validation (V&V) related to these systems.

After heat removal by a closed secondary cooler circuit on the example of the advanced high temperature reactor AHTR 500 with central graphite column

Design of multifunctional lattice-frame materials for compact heat exchangers

Thermal–hydraulic performance of fin-and-oval tube compact heat exchangers with innovative design of corrugated fin patterns

Study on tritium permeation behavior in primary and second circuits of the hydrogen production system by methane steam reforming using HTGR

The Phenomena Identification and Ranking Table (PIRT) technique was used to identify safety-relevant/safety-significant phenomena and assess the importance and related knowledge base of high-temperature structural materials issues for the Next Generation Nuclear Plant (NGNP), a very high temperature gas-cooled reactor (VHTR). The major aspects of materials degradation phenomena that may give rise to regulatory safety concern for the NGNP were evaluated for major structural components and the materials comprising them, including metallic and nonmetallic materials for control rods, other reactor internals, and primary circuit components; metallic alloys for very high-temperature service for heat exchangers and turbomachinery, metallic alloys for high-temperature service for the reactor pressure vessel (RPV), other pressure vessels and components in the primary and secondary circuits; and metallic alloys for secondary heat transfer circuits and the balance of plant. These materials phenomena were primarily evaluated with regard to their potential for contributing to fission product release at the site boundary under a variety of event scenarios covering normal operation, anticipated transients, and accidents. Of all the high-temperature metallic components, the one most likely to be heavily challenged in the NGNP will be the intermediate heat exchanger (IHX). Its thin, internal sections must be able to withstand the stresses associated with thermal loading and pressure drops between the primary and secondary loops under the environments and temperatures of interest. Several important materials-related phenomena related to the IHX were identified, including crack initiation and propagation; the lack of experience of primary boundary design methodology limitations for new IHX structures; and manufacturing phenomena for new designs. Specific issues were also identified for RPVs that will likely be too large for shop fabrication and transportation. Validated procedures for on-site welding, post-weld heat treatment (PWHT), and inspections will be required for the materials of construction. High-importance phenomena related to the RPV include crack initiation and subcritical crack growth; field fabrication process control; property control in heavy sections; and the maintenance of high emissivity of the RPV materials over their service lifetime to enable passive heat rejection from the reactor core. All identified phenomena related to the materials of construction for the IHX, RPV, and other components were evaluated and ranked for their potential impact on reactor safety.

A compact and efficient heat exchanger for exhaust gas recovery energy was needed to raise the total efficiency of a thermo-photovoltaic system TPV (Thermo-Photo-Voltaic) for automotive applications (see [1]). In order to respect the strict condition of a high heat transfer surface to volume ratio, a heat exchanger configuration with a plate compact multi-pass counter flow and locally cross-flow recuperator has been chosen. The goal of this work is the understanding of the behaviour of the heat exchanger with numerical and experimental analysis for different geometrical and operating conditions. A high number of dimensions and manufacturing constraints was evaluated before reaching a definite design of a compact and efficient heat exchanger to be tested in the lab for initial experiments. The experimental work was needed in order to validate the numerical model. As the material needed for the real application could not be easily manufactured and instrumented in a workshop, a simplified real model, made of brass, was built, in order to compare numerical results and experimental findings. It was supposed that results obtained in this way would be sufficient to be considered valid when extrapolated in the real heat exchanger high temperature operating conditions and manufacturing material. The experimental results have been successfully compared with numerical ones obtained with the Fluent CFD code (release 6.2.16) Curves of performance (ε-NTU diagram plotted as a function of the ratio between the minimum and the maximum thermal capacities of the flows and pressure drop -mass flowrate diagram as a function of the average temperature) have been obtained and were useful to choose the adequate configuration for different applications, depending on the requested heat transfer and maximum allowable pressure drop. The output of the investigation was: heat transfer, outlet temperatures for both air flows, heat exchanger efficiency, differential pressure drop for both hot and cold sides. After this validation final numerical simulations have been carried out in order to understand the dependence of the heat exchanger efficiency on other geometrical parameters and operating conditions such as plates dimensions, numbers and height of vanes, operating pressure and so on.

Beginning with the concept of micro heat exchangers and its advantages and disadvantages, we illustrate the notion of heat transfer intensification by several innovative designs of mini-scale heat exchangers proposed during our research work. Distinct from other approaches, we do not seek extra fine channel size. On the contrary, we work on how to effectively manage the hydrodynamic aspects and the geometric organization of heat transfer surface to intensify heat transfer with acceptable increase of total pressure drop, for example, using internal (chaotic) mixing, multi-passage configuration and multi-scale geometries. Other influencing factors such as materials, flow maldistribution and fabricating techniques are also discussed for a global consideration of efficient and compact heat exchanger designs.

ABSTRACTA commercial very high-temperature gas-cooled reactor (VHTR) hydrogen cogeneration system named gas turbine high temperature reactor 300-cogeneration (GTHTR300C) is designed and developed in Japan Atomic Energy Agency (JAEA). Moreover, it has been planned that hydrogen production system and gas turbine system is connect to high-temperature engineering test reactor (HTTR). The reactor system is required to continue a stable and safety operation as well as a stable power supply in the case that thermal-load is fluctuated by the occurrence of abnormal event in the heat utilization system like as the hydrogen production facility. Then, it is necessary to confirm that the thermal-load fluctuation could be absorbed by the reactor system so as to continue the stable and safety operation. The thermal-load fluctuation absorption tests using the HTTR were planned to clarify the absorption characteristics of the HTGR system. However, it is difficult to clarify the phenomenon due to many kinds of fluctuation in nuclear thermal power in the reactor core. Moreover, the actual data regarding how the delay of the temperature response is effective for the reactor system had been gained quantitatively.The thermal-load fluctuation absorption tests without nuclear heating were planned and conducted in JAEA to clarify the absorption characteristic of thermal-load fluctuation mainly by the reactor and by the intermediate heat exchanger (IHX). The absorption characteristics of thermal-load fluctuation can be revealed with sufficient temperature fluctuation. So the tests were conducted with the primary coolant temperature 120 °C which is the start-up temperature of the HTTR. As a result it was revealed that the reactor has the larger absorption capacity of thermal-load fluctuation than the expected one and that the IHX can be contributed to the absorption of the thermal-load fluctuation generated in the heat utilization system in the reactor system. It was confirmed from their results that the reactor and the IHX has effective absorption capacity of the thermal-load fluctuation generated in the heat utilization system. Moreover, the calculation with the safety evaluation code based on RELAP5/MOD3 was performed. It was confirmed that the calculated temperature for the reactor is almost same to the measured one with the new analysis model. On the other hand, it was confirmed that the calculated temperature for the IHX decreased faster than the measured one due to smaller absorption capacity in the calculation model than that in actual one. It can be considered that the calculation for the IHX produces the conservative result. It was summarized that the safety evaluation code can represent the thermal-load fluctuation absorption behavior conservatively.

Experimental and numerical study of flow distributor of intermediate heat exchanger in high temperature gas-cooled reactor

This study demonstrates an innovative approach to the automatic design of compact gyroid heat exchangers using the advanced engineering software nTopology, which is based on implicit modeling technology. The aim is to provide a modern enhancement to traditional 'Shell and Tube' type heat exchangers. Utilizing functions in implicit modeling and parametric design features, complex internal gyroid structures can be used as an alternative heat transfer interface in 'Shell and Tube' type heat exchangers. The most striking aspect of the methodology is its ability to fully automate the design process. By consolidating specific parameters into a single function block and entering scalar values, a fast and flexible workflow is activated, automatically generating the final geometry. Gyroid structures with high thermal performance and fluid dynamics are automatically adapted for various volumes and geometries. These structures can improve the overall efficiency of heat exchangers and offer significant advantages, especially in specialized application areas such as aerospace and space industries. In the design and sizing process of the gyroid heat exchanger, if manufacturing is planned, it is necessary to pay attention to the design principles for metal additive manufacturing. In conclusion, this study demonstrates that the advanced engineering software known as nTopology can create a synergistic effect in the rapid and easy design of gyroid heat exchangers and in establishing the automation of the design process.

The analysis of specific control strategies and dynamic behavior of the supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle has been extended to the two reactor types selected for continued development under the Generation IV Nuclear Energy Systems Initiative; namely, the Very High Temperature Reactor (VHTR) and the Sodium-Cooled Fast Reactor (SFR). Direct application of the standard S-CO{sub 2} recompression cycle to the VHTR was found to be challenging because of the mismatch in the temperature drop of the He gaseous reactor coolant through the He-to-CO{sub 2} reactor heat exchanger (RHX) versus the temperature rise of the CO{sub 2} through the RHX. The reference VHTR features a large temperature drop of 450 C between the assumed core outlet and inlet temperatures of 850 and 400 C, respectively. This large temperature difference is an essential feature of the VHTR enabling a lower He flow rate reducing the required core velocities and pressure drop. In contrast, the standard recompression S-CO{sub 2} cycle wants to operate with a temperature rise through the RHX of about 150 C reflecting the temperature drop as the CO{sub 2} expands from 20 MPa to 7.4 MPa in the turbine and the fact that the cycle is highly recuperated such that the CO{sub 2} entering the RHX is effectively preheated. Because of this mismatch, direct application of the standard recompression cycle results in a relatively poor cycle efficiency of 44.9%. However, two approaches have been identified by which the S-CO{sub 2} cycle can be successfully adapted to the VHTR and the benefits of the S-CO{sub 2} cycle, especially a significant gain in cycle efficiency, can be realized. The first approach involves the use of three separate cascaded S-CO{sub 2} cycles. Each S-CO{sub 2} cycle is coupled to the VHTR through its own He-to-CO{sub 2} RHX in which the He temperature is reduced by 150 C. The three respective cycles have efficiencies of 54, 50, and 44%, respectively, resulting in a net cycle efficiency of 49.3 %. The other approach involves reducing the minimum cycle pressure significantly below the critical pressure such that the temperature drop in the turbine is increased while the minimum cycle temperature is maintained above the critical temperature to prevent the formation of a liquid phase. The latter approach also involves the addition of a precooler and a third compressor before the main compressor to retain the benefits of compression near the critical point with the main compressor. For a minimum cycle pressure of 1 MPa, a cycle efficiency of 49.5% is achieved. Either approach opens up the door to applying the SCO{sub 2} cycle to the VHTR. In contrast, the SFR system typically has a core outlet-inlet temperature difference of about 150 C such that the standard recompression cycle is ideally suited for direct application to the SFR. The ANL Plant Dynamics Code has been modified for application to the VHTR and SFR when the reactor side dynamic behavior is calculated with another system level computer code such as SAS4A/SYSSYS-1 in the SFR case. The key modification involves modeling heat exchange in the RHX, accepting time dependent tabular input from the reactor code, and generating time dependent tabular input to the reactor code such that both the reactor and S-CO{sub 2} cycle sides can be calculated in a convergent iterative scheme. This approach retains the modeling benefits provided by the detailed reactor system level code and can be applied to any reactor system type incorporating a S-CO{sub 2} cycle. This approach was applied to the particular calculation of a scram scenario for a SFR in which the main and intermediate sodium pumps are not tripped and the generator is not disconnected from the electrical grid in order to enhance heat removal from the reactor system thereby enhancing the cooldown rate of the Na-to-CO{sub 2} RHX. The reactor side is calculated with SAS4A/SASSYS-1 while the S-CO{sub 2} cycle is calculated with the Plant Dynamics Code with a number of iterations over a timescale of 500 seconds. It is found that the RHX undergoes a maximum cooldown rate of {approx} -0.3 C/s. The Plant Dynamics Code was also modified to decrease its running time by replacing the compressible flow form of the momentum equation with an incompressible flow equation for use inside of the cooler or recuperators where the CO{sub 2} has a compressibility similar to that of a liquid. Appendices provide a quasi-static control strategy for a SFR as well as the self-adaptive linear function fitting algorithm developed to produce the tabular data for input to the reactor code and Plant Dynamics Code from the detailed output of the other code.

The IHX (Intermediate Heat Exchanger) of a VHTR (Very High Temperature Reactor) transfers 950°C heat generated from the VHTR to a hydrogen production plant. The Korea Atomic Energy Research Institute (KAERI) has manufactured a 70 kW class lab-scale PCHE (Printed Circuit Heat Exchanger) prototype made of SUS316L under consideration as a candidate. In this study, as a part of a structural integrity evaluation of the lab-scale PCHE prototype, an elastic structural analysis including structural analysis modeling and a thermal/elastic structural analysis was carried out under the test conditions of a helium experimental loop (HELP). In addition, a structural integrity evaluation considering the weld material properties was conducted.

Investigating the effect of properties variation in optimum design of compact heat exchanger using segmented method