High Temperature Reactor Technology Research Articles

Simplified Reactor Model for Microreactor Coupled with Helium Closed Brayton Cycle

Microreactors comprise a new actively developing class of very small advanced reactors that have the potential to be an alternative to carbon-intensive energy technologies. A microreactor based on high-temperature gas reactor (HTGR) technology is a very promising advanced reactor with inherent safety, and it can couple with a closed Brayton cycle for higher efficiency. Since dynamics characteristics are fundamental to analyzing a power generation system and a reactor is the main source of the dynamics characteristics of a system, it is necessary to study a microreactor model suitable for system analysis. The main goal is to simulate the performance of the previously mentioned integrated system, focusing on the details of the power conversion unit while still ensuring acceptable calculation times. Hence, a simplified reactor model is needed that could supply sufficiently accurate values of pressure drop and heat transfer across the core. In this paper, by simplifying the physical processes in a microreactor, a dynamic model described by differential algebraic equations is obtained based on the lumped parameter modeling methodology and the basic conservation of fluid mass, momentum, and energy. Coupling thermal hydraulics with neutron kinetics, the temperature coefficient of reactivity and xenon poisoning are considered. Finally, the model is programmed and calculated using Modelica language. The transient responses of the main parameters under typical perturbations are analyzed, and the results show that the responses are correct. Because of the effect of reactivity feedback, fluctuations of the main parameters caused by microperturbations eventually tend to stabilize. In addition, the effects of negative reactivity introduced by xenon poisoning under two typical dynamic processes are analyzed. In power regulation, excess reactivity is required to compensate for the negative reactivity introduced by 135Xe. The model and results can properly predict the systematic parameters and serve as a basis for system analysis of microreactor coupling with the helium closed Brayton cycle.

Preface

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INFORMATION PROCESSING IN THE REACTOR PROTECTION SYSTEMS OF HIGH TEMPERATURE GAS-COOLED REACTORS

Reactor protection systems (RPS) transform process variable signals from the sensors into initiation and actuation signals to trip the reactor if the signal's value exceeds the predefined trip setpoints of the RPS. Information on the current value of the process variables signals and the trip setpoint should be displayed properly on the visual display unit (VDU) in order to maintain the situation awareness of the operators in main control rooms (MCR). In addition, it is also helpful for them to investigate the cause of an accident after the reactor trip and to mitigate the accident based on the appropriate emergency operating procedures. This paper investigates how the information is processed in the RPS of Experimental Power Reactor (EPR) based on high temperature reactor (HTR) technology, and how the information is displayed on the human machine interface (HMI) of the MCR of the EPR. It is conducted by classifying the RPS into three layers based on its components and their functions, followed by the investigation of the type and the information processing in each layer. The results show that the form of the information has been changed throughout the RPS, started from the sensors and until it is displayed on the VDU. The results of the investigation are necessary to understand the concept of RPS, especially for new operators, and to prepare the mitigation actions based on the process variable that cause the reactor trip.Keywords: Experimental power reactor, Reactor protection system, Human machine interface, Information processing, Situation awareness

Scaling Methodology for Integral Effects Tests in Support of Fluoride Salt–Cooled High-Temperature Reactor Technology

The capability to validate integral transient response models is of critical importance for licensing new reactor designs. The Kairos Power testing program has developed a methodology to design scaled experiments that predict the thermal fluid behavior of the Kairos Power Fluoride Salt–Cooled High-Temperature Reactor (KP-FHR). Such experiments will be used as part of the assessment base of evaluation models supporting KP-FHR safety analysis. The Hierarchical Two-Tiered Scaling (H2TS) methodology was selected for Kairos Power scaling efforts that will be applied to integral effects tests (IETs) for system-level testing. In this paper, the scaling methodology is presented for thermal fluid IETs that will model the KP-FHR primary heat transport system under normal operations and transient conditions involving transition to natural circulation. This paper provides a basis for using surrogate fluids for testing that requires the thermal fluid performance of the KP-FHR primary coolant lithium fluoride–beryllium fluoride [2LiF/BeF2 (Flibe)] to be replicated. Kairos Power intends to use heat transfer oil as a surrogate fluid for Flibe in thermal fluid IETs. This paper demonstrates that this class of surrogate fluids is an acceptable substitute for Flibe salt for some types of scaled IETs and that the principal thermal fluid properties can be properly scaled with minor distortions over the range of conditions expected for both normal and off-normal operating conditions of the KP-FHR.

Thermal hydraulics modeling of On-line refueling for the Advanced High Temperature Reactor (AHTR)

As part of the effort to move forward the Fluoride salt-cooled High-temperature Reactor (FHR) technology, the Advanced High Temperature Reactor (AHTR) is being developed at Oak Ridge National Laboratory (ORNL) and several other institutions. Due to its plank fuel design and inherently low heavy metal loading, which challenges fuel utilization, online refueling is considered as an option for improving the economic viability of this reactor.This work presents the thermal hydraulic modeling of the online refueling for the AHTR. Analyses at the single channel and fuel assembly level have been performed in order to develop a model of the reactor. The reactor modeling has been integrated with CFD studies for characterizing the steady state conditions and the refueling transient. The simulations of the transient have been complemented by the analysis of the stability of the refueling operation.This work demonstrates the viability of online refueling from a thermal hydraulic standpoint and develops a modeling approach for this type of operational transient.

On-line refueling for the Advanced High Temperature Reactor (AHTR)

Abstract Several academic and commercial organizations around the world are developing the Fluoride salt-cooled High-temperature Reactor (FHR) technology, due to its safety features and potential to generate high temperature energy for electricity and process heat applications. The Advanced High Temperature Reactor (AHTR) being considered in this study is a FHR design developed at Oak Ridge National Laboratory (ORNL). It is based on the use of graphite as moderator, FLiBe as coolant, and hexagonal fuel elements with fuel plates (or “planks”) composed of TRISO particles embedded in a carbonaceous matrix. The AHTR reference design is based on traditional batch refueling approach, which requires to shut down the reactor and replace/reshuffle a certain amount of fuel assemblies in the core at a specific frequency. Several options have been evaluated in the design process, in order to maximize the cycle length and optimize the use of fuel. However, the relatively short cycle and poor fuel utilization are intrinsic features of this family of reactors, due to the low heavy metal loading in the core and insufficient moderation, which are competing aspects in terms of core volume fraction. Since the fuel is expected to be more expensive than the fuel of light water reactors (LWR), this issue might challenge the economic viability of the AHTR. In order to eliminate or ameliorate this issue, a novel approach to refueling has been developed and proposed, consisting of continuous on-power refueling, or on-line refueling, in which the refueling procedure is performed at full power or partially reduced power (the reactor is not shut down) and a single assembly is removed for each refueling operation. A systematic neutronic and thermal-hydraulic analysis approach has been developed and performed to assess the viability and safety of the refueling operations, followed by the evaluation of the core design requirements and a quantification of the economic advantages resulting from the implementation of this procedure.

Phenomena Identification and Ranking Table (PIRT) study for metallic structural materials for advanced High-Temperature reactor

The Fluoride High-Temperature Reactor (FHR) technology promises many benefits including passive safety, proliferation-resistant waste forms, and improved economics. However, selection of reliable structural materials and identification of the possible degradation mechanisms for these is important for the licensure and the safe operation of FHRs. In order to address this task, the Georgia Tech led Integrated Research Project (IRP) hosted a Phenomena Identification and Ranking Table (PIRT) panel of experts to address degradation mechanisms and other materials related issues of importance to the FHRs. Materials, ones that come in contact with FLiBe or FLiNaK molten salts or other related environments like high temperature steam etc., were considered in this PIRT. Focus of this PIRT was the metallic alloys, especially the ones that are permitted for the construction of elevated temperature Class A components by the ASME code. Degradation mechanisms considered in this PIRT included chemical degradation, mechanical degradation, radiation degradation, and synergistic effect of these mechanisms that may negatively impact operations or cause some safety concerns for the major structural components of FHRs. Main components which were considered included vessel and primary piping, primary heat exchangers, steam generator vessel, steam generator tubes, intermediate loop piping, valves and pumps. Welds in all structural components were identified as an important class of material, which varies in composition and properties, and needs more attention. Importance of impurity control in molten fluorides considered for FHR was highlighted throughout PIRT panel discussions. This paper gives a summary of important results from the PIRT panel discussions and report.

High temperature reactor technology development in India

High Temperature Reactor technology development programme was initiated in India with an aim to provide high temperature process heat for nuclear hydrogen production by splitting water. As high efficiency hydrogen production needs process heat at temperatures around 1123 K, a challenging technology development goal for the high temperature reactors was set to achieve coolant temperature of 1273 K. Currently development is in progress for a Compact High Temperature Reactor (CHTR), and a 600 MWth Innovative High Temperature Reactor (IHTR). Current design version of CHTR has 235U based TRISO (TRistructural-ISOtropic) coated particle fuel, Beryllium oxide (BeO) as moderator, graphite as reflector, and lead-bismuth eutectic (LBE) as the coolant. The design incorporates many passive safety features for reactor heat removal. Current design version of IHTR is based on pebble bed fuel configuration with molten salt as coolant. For both the reactors, reactor heat is removed passively by natural circulation of the coolant. Technology development for these reactors include development of TRISO coated particle fuel, lead-bismuth eutectic and molten salt coolant technologies, BeO and graphite, oxidation resistant coatings, high creep strength alloys compatible to these coolants, high temperature instrumentation for these coolants, as well as high efficiency hydrogen production and electricity generation technologies.

Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges

A high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions. Each loop will also include an interchangeable high-temperature test section that can be customized to address specific research issues associated with each working fluid. This paper also discusses needs and challenges associated with advanced instrumentation for the multi-loop facility, which could be further applied to advanced high-temperature reactors. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. A preliminary design configuration of the ARTIST facility will be presented with the required design and operating characteristics of the various components. 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. The salt loop will be thermally integrated with the steam/water loop operating at PWR conditions. Experiment design challenges include identifying suitable materials and components that will withstand the required loop operating conditions. The instrumentation needs to be highly accurate (negligible drift) in measuring operational data for extended periods of times, as data collected will be used for code and model verification and validation, one of the key purposes for the loop. The experimental facility will provide a much-needed database for successful development of advanced reactors and provide insight into the needs and challenges in instrumentation for advanced high-temperature reactors.

Feasibility study of high temperature reactor utilization in Czech Republic after 2025

High temperature reactors (HTRs) were examined as an option to intended future broadening of the nuclear energy production in Czech Republic. The known qualities as the inherent safety, high thermal utilization and non-electrical applications have been assessed in years 2009–2011 during the survey funded by Czech Ministry of Industry and Trade. The survey of high temperature reactors with spherical fuel was initiated by reason of mature state of the art of this technology type in South Africa and in China, where in both countries pilot plants were planned. Unfortunately, the global financial crisis caused the decision of stopping the governmental support in South African programme was made. In China, however, the development still continues. Czech Republic has almost 60 years nuclear research history and the knowledge of operation of gas cooled and heavy water moderated reactor has been gained in the past. Nevertheless, the design of light water reactors was more developed in former Soviet Union, which provided Czech scientists by initial knowledge base; hence the research has been reoriented to this technology. But, the demands on future nuclear reactors application are still growing and the same or even higher living standard of next generations have to be taken into consideration. Therefore the systems, which can produce more energy and less waste, are getting into foreground of interest of Czech decision makers. The high temperature reactor technology seems to be the successful representative of the GEN IV reactor types, which will be operated commercially in the near future. The broad spectrum of utilization enables this system to be an option after 2030, when the electricity demand is planned to be covered from about 50% by nuclear in our country.

Theoretical and experimental modelling of a heat pipe heat exchanger for high temperature nuclear reactor technology

High temperature heat sources are becoming an ever-increasing imperative in the processing industries for the production of various plastics, fertilisers, coal-to-liquid fuel and hydrogen generation. Current high temperature reactor technology is capable of producing reactor coolant temperatures in excess of 950 °C. At these temperatures, tritium which is a radioactive contaminant found in the reactor coolant stream, is able to contaminate the secondary stream by diffusing through the steel retaining wall of the heat exchanger between the reactor coolant and secondary process coolant stream. Current regulations therefore require an extra intermediate heat transfer loop to ensure no cross contamination. A novel heat pipe heat exchanger design is presented which circumvents the need for an intermediate coolant loop. This is done by physically separating the reactor coolant and secondary coolant by two pipe walls and a vapour section and a liquid section. A theoretical transient heat transfer model of such a device is presented. The model uses separate hot gas heating fluid and cold water cooling fluid control volumes, and for the internal working fluid a control volume consisting of a liquid and its vapour in equilibrium with each other. A 2 kW rated experimental model was constructed and tested, using Dowtherm-A as working fluid, to validate the heat pipe heat exchanger theoretical model and design. By determining the boiling heat transfer coefficient through the use of an experimentally formulated correlation it was shown that the theoretical model is indeed able to simulate the characteristic chaotic behaviour, due to the boiling and condensation processes, of the device to within a reasonable level of accuracy. It is concluded that the theoretical simulation model can be used to predict the performance of a higher temperature sodium-charged heat pipe heat exchanger, provided suitable boiling and condensation heat transfer coefficients are used.

The current status of fluoride salt cooled high temperature reactor (FHR) technology and its overlap with HIF target chamber concepts

The fluoride salt cooled high temperature reactor (FHR) is a class of fission reactor designs that use liquid fluoride salt coolant, TRISO coated particle fuel, and graphite moderator. Heavy ion fusion (HIF) can likewise make use of liquid fluoride salts, to create thick or thin liquid layers to protect structures in the target chamber from ablation by target X-rays and damage from fusion neutron irradiation. This presentation summarizes ongoing work in support of design development and safety analysis of FHR systems. Development work for fluoride salt systems with application to both FHR and HIF includes thermal-hydraulic modeling and experimentation, salt chemistry control, tritium management, salt corrosion of metallic alloys, and development of major components (e.g., pumps, heat exchangers) and gas-Brayton cycle power conversion systems. In support of FHR development, a thermal-hydraulic experimental test bay for separate effects (SETs) and integral effect tests (IETs) was built at UC Berkeley, and a second IET facility is under design. The experiments investigate heat transfer and fluid dynamics and they make use of oils as simulant fluids at reduced scale, temperature, and power of the prototypical salt-cooled system. With direct application to HIF, vortex tube flow was investigated in scaled experiments with mineral oil. Liquid jets response to impulse loading was likewise studied using water as a simulant fluid. A set of four workshops engaging industry and national laboratory experts were completed in 2012, with the goal of developing a technology pathway to the design and licensing of a commercial FHR. The pathway will include experimental and modeling efforts at universities and national laboratories, requirements for a component test facility for reliability testing of fluoride salt equipment at prototypical conditions, requirements for an FHR test reactor, and development of a pre-conceptual design for a commercial reactor.

EUROPAIRS: The European project on coupling of High Temperature Reactors with industrial processes

Developers of High Temperature Reactors (HTR) worldwide acknowledge that the main asset for market breakthrough is its unique ability to address growing needs for industrial cogeneration of heat and power (CHP) owing to its high operating temperature and flexibility, adapted power level, modularity and robust safety features. A strong alliance between nuclear and process heat user industries is a necessity for developing such a nuclear system for the conventional process heat market, just as the electro-nuclear development required a close partnership with utilities. Initiating such an alliance is one of the objectives of the EUROPAIRS project (www.europairs.eu) presently on-going in the frame of the Euratom 7th Framework Programme (FP7). Although small and of short duration (21 months), EUROPAIRS is of strategic importance: it generates the boundary conditions for rapid demonstration of collocating HTR with industrial processes as proposed by the European High Temperature Reactor Technology Network (HTR-TN). This paper presents the main goals, the organization and the working approach of EUROPAIRS. It also presents the status of the viability assessment studies for coupling HTR with industrial end-user systems as one of the main pillars of the project. The main goal of the viability assessment is to identify developments required to remove the last technological and licensing barriers for a viable coupling scheme. The study is expected to result in guidelines for directing the choice of an industrial scale prototype.

Improvements in the fabrication of HTR fuel elements

The application of High Temperature Reactor (HTR) Technology in the course of the continuously increasing world wide demand on energy is taken more and more under serious consideration in the power supply strategy of various countries. Especially for the emerging nations the HTR Technology has become of special interest because of its inherent safety feature and due to the alternative possibilities of applications, e.g. in the production of liquid hydrocarbons or the alternative application in H2 generation.The HTR fuel in its various forms (spheres or prismatic fuel blocks) is based on small fuel kernels of about 500μm in diameter. Each of these uranium oxide or carbide kernels are coated with several layers of pyrocarbon (PyC) as well as an additional silicon carbide (SiC) layer. While the inner pyrocarbon layer is porous and capable to absorb gaseous fission products, the dense outer PyC layer forms the barrier against fission product release. The SiC layer improves the mechanical strengths of this barrier and considerably increases the retention capacity for solid fission products that tent to diffuse at these temperatures. Especially the high quality German LEU TRISO spherical fuel based on the NUKEM design, has demonstrated the best fission product release rate, particular at high temperatures. The ∼10% enriched uranium triple-coated particles are embedded in a moulded graphite sphere. A fuel sphere consists of approximately 9g of uranium (some 15,000 particles) and has a diameter of 60mm.As the unique safety features, especially the inherent safety of the HTR is based on the fuel design, this paper shall reflect the complexity but also developments and economical aspects of the fabrication processes for HTR fuel elements.

New Temperature Monitoring Devices for High-Temperature Irradiation Experiments in the High Flux Reactor Petten

Within the European High Temperature Reactor Technology Network (HTR-TN) and related projects a number of HTR fuel irradiations are planned in the High Flux Reactor Petten (HFR), The Netherlands, with the objective to explore the potential of recently produced fuel for even higher temperature and burn-up. Irradiating fuel under defined conditions to extremely high burn-ups will provide a better understanding of fission product release and failure mechanisms if particle failure occurs. After an overview of the irradiation rigs used in the HFR, this paper sums up data collected from previous irradiation tests in terms of thermocouple data. Some R&D for further improvement of thermocouples and other on-line instrumentation will be outlined.

High and very high temperature reactor research for multipurpose energy applications

Ten years ago, the European High Temperature Reactor (HTR) Technology Network (HTR-TN) launched a programme for developing HTR Technology, which expanded so far through 4 successive Euratom Framework Programmes. Many projects have been performed – in particular the RAPHAEL project in the 6th Euratom Framework Programme and presently ARCHER in the 7th – in line with the Network strategy that identified cogeneration of process heat and power as the main specific mission of HTR. HTR can indeed address the growing energy needs of industry presently fully relying on fossil fuel combustion with a CO 2-lean generation technology, thanks to its high operating temperature and to its unique flexibility obtained from its large thermal inertia and its low power. Relying on the legacy of the former European leadership in HTR technology, this programme has addressed specific developments required for industrial process heat applications and for increasing HTR performances (higher temperatures and fuel burn-up). Decisive achievements have been obtained concerning fuel manufacturing and irradiation behaviour, key components and their materials, safety, computer code validation and specific HTR waste (fuel and graphite) management. Key experiments have been performed or are still ongoing: irradiation of graphite, fuel and vessel materials and the corresponding post-irradiation examinations, safety tests and isotopic analyses; thermal-hydraulic tests of an Intermediate Heat Exchanger mock-up in helium; air ingress experiments for a block type core, etc. Through Euratom participation in the Generation IV International Forum (GIF), these achievements contribute to international cooperation. HTR-TN strategy has been recently integrated by the “Sustainable Nuclear Energy Technology Platform” (SNE-TP) as one of the 3 “pillars” of its global nuclear strategy. It is also in line with the orientations and the timing of the “Strategic Energy Technology Plan (SET-Plan)” for the development of CO 2-lean energy technologies, and thus strengthens the nuclear option in a future European energy mix. Nuclear cogeneration for industrial process heat applications is a major innovation and a major challenge, requiring large-scale demonstration to prove its industrial viability. To enable demonstration, it is necessary not only to develop an appropriate nuclear heat source, but also to develop coupling technologies and to adapt industrial processes to the coupling with a HTR. This requires a close partnership between the conventional and the nuclear technology holders as the base of a Nuclear Cogeneration Industrial Initiative. Recently the project EUROPAIRS initiated by HTR-TN together with process heat user industries has set the bases of such a strategic partnership.

HTR-TN Achievements and Prospects for Future Developments

It is already 10 years since the (European) High Temperature Reactor Technology Network (HTR-TN) launched a program for development of HTR technology, which expanded through three successive Euratom framework programs, with many projects in line with the network strategy. Widely relying in the beginning on the legacy of the former European HTR developments (DRAGON, AVR, THTR, etc.) that it contributed to safeguard, this program led to advances in HTR/VHTR technologies and produced significant results, which can contribute to the international cooperation through Euratom involvement in the Generation IV International Forum (GIF). the main achievements of the European program, performed in complement to efforts made in several European countries and other GIF partners, are presented: they concern the validation of computer codes (reactor physics, as well as system transient analysis from normal operation to air ingress accident and fuel performance in normal and accident conditions), materials (metallic materials for vessel, direct cycle turbines and intermediate heat exchanger, graphite, etc.), component development, fuel manufacturing and irradiation behavior, and specific HTR waste management (fuel and graphite). Key experiments have been performed or are still ongoing, like irradiation of graphite and of fuel material (PYCASSO experiment), high burn-up fuel PIE, safety test and isotopic analysis, IHX mock-up thermohydraulic test in helium atmosphere, air ingress experiment for a block type core, etc. Now HTR-TN partners consider that it is time for Europe to go a step forward toward industrial demonstration. In line with the orientations of the “Strategic Energy Technology Plan (SET-Plan)” recently issued by the European Commission that promotes a strategy for development of low-carbon energy technologies and mentions Generation IV nuclear systems as part of key technologies, HTR-TN proposes to launch a program for extending the contribution of nuclear energy to industrial process heat applications addressing (1) the development of a flexible HTR that can be coupled to many different process heat and cogeneration applications with very versatile requirements, (2) the development of coupling technologies for such coupling, (3) the possible adaptations of process heat applications required for nuclear coupling, and (4) the integration and optimization of the whole coupled system. As a preliminary step for this ambitious program, HTR-TN endeavors to create a strategic partnership between nuclear industry and R&D and process heat user industries.

A contribution on the investigation of the dynamic behavior of rotating shafts with a Hybrid Magnetic Bearing Concept (HMBC) for blower application

Within a subproject of the RAPHAEL-Program, which was part of the 6th EURATOM Framework Program supervised by the European Commission, it was investigated whether the use of a Hybrid Magnetic Bearing Concept (HMBC) will be beneficial for a blower application. Within the RAPHAEL program, the subproject “Component Development” is dealing with R&D of components of High Temperature Reactor Technology (HTR), where a major focus is on safety- and reliability-related issues. That implies special requirements for the support of high speed rotating shafts in HTR-Applications that only can be satisfied by using Active Magnetic Bearings (AMB). Regarding safety and competitiveness, AMBs are considered to be key components for the support of rotating HTR-components due to their technical features. AMBs are characterized by an electromagnetic actuator that is generating the bearing force depending on the clearance between stator and rotor, in which the rotor is levitated. Therefore an active control of the coil current is necessary. Furthermore, Touch Down Bearings (TDB) are needed to avoid damages in case of an emergency shut down or in case of energy supply losses. This contribution provides an internal insight on the advantages of a Hybrid Magnetic Bearing Concept that is characterized by a completely Active Magnetic Bearing-supported vertical arranged rotor and an additional permanent magnetic radial orientated bearing. One benefit of the HMBC is an additional radial guidance of the shaft that may reduce the loads while dropping into the Touch Down Bearings e.g. in case of energy supply losses of the AMBs. Reduced loads on the TDBs will increase their life cycle and the availability of the AMB supported component. The scope of this R&D-Project, which will be described more detailed in this contribution, includes: the analytical modeling and simulation of the dynamic behavior of the Hybrid Magnetic Bearing System; the modification of the completely AMB-supported test facility FLP 500 with a radial PMB; the experimental tests and validation of the analytical models to provide recommendations for the investigated blower application as a HTR-component. Furthermore, the effects occurring during the modification of the test facility and the approach that was necessary to solve unexpected problems will be described.

Postirradiation Testing of High Temperature Reactor Spherical Fuel Elements Under Accident Conditions

A new furnace for accident condition testing of spherical high temperature reactor fuel elements has been installed and now operates in the hot cells of the Institute for Transuranium Elements (ITU) Karlsruhe. The recent apparatus was constructed on the basis of a former development by Forschungszentrum Jülich (Schenk, Pitzer, and Nabielek, 1988, “Fission Product Release Profiles From Spherical HTR Fuel Elements at Accident Temperatures,” Jülich Report No. 2234), where it was named KüFA, the German acronym for cold finger apparatus. In a preceding publication (Kostecka, Ejton, de Weerd, and Toscano, 2004, “Post-Irradiation Testing of HTR-Fuel Elements Under Accident Conditions, Part 1 and 2,” Second International Topical Meeting on High Temperature Reactor Technology, Beijing, China) the general concept and details of the device were described. The present paper reports on the first operation under hot conditions, and the calibration of the fission gas measurement and of the efficiency of the cold finger, which is used to plate out solid fission products. Finally the results of fission product release and analysis of two heating tests on two fuel elements from the high temperature reactor K6 irradiation experiment (Nabielek, Conrad, Roellig, and Meyers, 1993, “Fuel Irradiation Experiments on HFR-K6 and HFR-B1 With Intermittent Water Vapour Injections,” Technical Committee Meeting on Response of Fuel, Fuel Elements and Gas Cooled Reactor Cores Under Accidental Air or Water Ingress Conditions, Beijing, China, Oct. 25–27) are presented and discussed.

A Comparison Between Recent Advances in Cylindrical Nodal Diffusion Methods

The resurgence of high temperature reactor (HTR) technology has prompted the development and application of modern calculation methodologies, many of which are already utilized in the existing power reactor industry, to HTR designs. To this end, the use of nodal diffusion methods for full core neutronic analysis is once again considered for both their performance and accuracy advantages. Recently a number of different approaches to two-dimensional and 3D multigroup cylindrical nodal diffusion methods were proposed by various institutions for use in HTR and, specifically, pebble-bed modular reactor (PBMR) calculations. In this regard, we may mention the NEM code from the Pennsylvania State University based on the nodal expansion method and the OSCAR-4 code from NECSA, utilizing a conformal mapping approach to the analytic nodal method. In this work we will compare these two approaches in terms of accuracy and performance. Representative problems, selected to test the methods thoroughly, were devised and based on both a modified version of the PBMR 400 MW benchmark problem and a “cylindrisized” version of the IAEA two-group problem. The comparative results between OSCAR-4 and NEM are given, focusing on global reactivity estimation, as well as power and flux errors as compared with reference finite-difference solutions. These results indicate that both OSCAR-4 and NEM recover the global reference solution for the IAEA problem and show power errors, which are generally acceptable for nodal methods. For the PBMR problem the accuracy is similar, but some convergence difficulties are experienced at the outer boundaries of the system due to the very large dimensions of the reflector (when compared with typical water-moderated reactors). For both codes a significant performance increase was found, as compared with finite-difference calculations, which is the method currently employed by the PBMR (Pty) Ltd. In conclusion it seems that nodal methods have potential for use in the HTR analysis and, specifically, the PBMR calculational arena, although cylindrical geometry based nodal methods will have to develop toward maturity before becoming the industry standard.