HTRs, both prismatic block fuelled and pebble fuelled, feature a number of uniquely beneficial characteristics that will be discussed in this paper. In this paper the construction of an international experimental pebble bed reactor is proposed, possible experiments suggested and an invitation extended to interested partners for co-operation in the project. Experimental verification by nuclear regulators in order to facilitate licensing and the development of a new generation of reactors create a strong need for such a reactor. Suggested experiments include: • Optimized incineration of waste Pu in a pebble bed reactor: The capability to incineration pure reactor grade plutonium by means of ultra high burn-up in pebble bed reactors will be presented at this conference in the track on fuel and fuel cycles. This will enable incineration of the global stockpile of separated reactor grade Pu within a relatively short time span; • Testing of fuel sphere geometries, aimed at improving neutron moderation and a decrease in fuel temperatures; • Th/Pu fuel cycles: Previous HTR programs demonstrated the viability of a Th-232 fuel-cycle, using highly enriched uranium (HEU) as driver material. However, considerations favoring proliferation resistance limit the enrichment level of uranium in commercial reactors to 20%, thereby lowering the isotopic efficiency. Therefore, Pu driver material should be developed to replace the HEU component. Instead of deploying a (Th, Pu)O 2 fuel concept, the proposal is to use the unique capability offered by pebble bed reactors in deploying separate Th- and Pu-containing pebbles, which can be cycled differently; • Testing of carbon-fiber-carbon (CFC) structures for in-core or near-core applications, such as guide tubes for reserve shutdown systems, thus creating the possibility to safely shutdown reactors with increased diameter; • Development of very high temperature reactor components for process heat applications; • Advanced decay heat removal systems e.g. design specific air flow channels, or heat pipe designs external to the reactor pressure vessel; • Development of a plutonium fuelled peaking reactor with the proposed duel cycle; • A radial coolant flow pattern with increased power output; • Testing of carbon-fiber-carbon (CFC) core barrel applications. The design will facilitate ease of licensing by sacrificing performance in favor of safety and employing redundant defense-in-depth safety systems. Speedy licensing is therefore expected. The economic model will be based on a commercial expedition of the agreed experimental value to collaborating participants. Target costs will be minimized by exploiting known technology only and by utilizing off-the-shelf components as far as possible.

In order to present credible results in nuclear design and safety analysis, computer codes must adhere to stringent qualification procedures imposed by nuclear licensing authorities. Such procedures form the basis for a quality assured verification and validation process. This is particularly true for advanced nuclear systems of Generation IV type, where little licensing experience exists as well as little or no plant data is available. Qualification of nuclear design and analysis codes can be achieved in various ways, namely: comparison of results from a code with results from another code i.e. code to code benchmarking; comparison of results from a given code with experimental results, i.e. code to experiment benchmarking; comparison of results from a given code with operational plant data; and finally, comparison of the results of a given code with known analytical solutions. In this paper, a systematic qualification of the coupled neutron transport and thermal hydraulics code DORT-TD/THERMIX is presented. As part of developing this coupled code to the level where it can be used as an independent tool by both designers of pebble-bed High-Temperature Gas-cooled Reactors (HTGRs) and regulators, an effort has been made to verify the coupling scheme as well as the validity of application for this code package. At these initial stages a code to code comparison has been adopted as the qualification method of choice. This is done for both steady-state and transient benchmark problems, ranging from simplified to detailed models. As shown in the results section, all benchmarks have been successfully recalculated and generally show good to very good agreement with the “reference” solutions.

Safety demonstration tests on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) were conducted to verify the inherent safety characteristics of modular High Temperature Gas-cooled Reactors (HTGRs) as well as to obtain the reactor core and primary cooling system transient data for validation of HTGR safety analysis models and codes. As one of these safety demonstration tests, a simulated anticipated transient without scram (ATWS) test called loss of forced cooling by tripping the helium circulator without reactor scram was carried out at 100% rated power level in July, 2005. This paper simulates the reactor transient behaviour during the test by using the THERMIX code system. The reactor power transition and a comparison with the test result are presented. Owing to the negative temperature coefficient of reactivity, the reactor undergoes a self-shutdown after the stop of the helium circulator and keeps subcritical till the end of the test. Due to the loss of forced cooling, the residual heat is slowly transferred from the core to the Reactor Cavity Cooling System (RCCS) by conduction, radiation and natural convection. The thermal response of this heat removal process is investigated. The calculated and test temperature transients of the measuring points in the reactor internals are given and the differences are preliminarily discussed. With respect to the safety features of the HTR-10, it is of most importance that the maximum fuel center temperature is always lower than 1230 °C which is the limited value at the first phase of the HTR-10 project. The simulation and test results show that the HTR-10 has the built-in passive safety features, and the THERMIX code system is applicable and reasonable for simulating and analyzing the helium circulator trip ATWS test.

The institute of nuclear engineering and energy systems (IKE), University of Stuttgart, Germany has developed a new thermal hydraulic tool which can be used for three-dimensional thermal hydraulic analysis of pebble bed as well as block type HTRs. During nominal operation, the flow inside the gas-cooled High Temperature Reactor is essentially single-phase, compressible, and non-isothermal. So, at least one gas phase has to be considered beside the solid phase for thermal hydraulic analysis of HTRs. Each phase (e.g. solid, gas) is considered as a continuum which occupies only its respective fraction of the control volume. Thermal non-equilibrium is considered between phases and time dependent energy conservation equations for solid and gas phases are solved. Simplified momentum conservation equation for gas obtained from porous media approximation is solved along with the time dependent mass conservation equation. Provisions for simulating more than one gas component are available in this newly developed code TH3D which could be required for simulating some accident situations (e.g air / water ingress by pipes break). The interaction between phases is made by a set of constitutive equations which rely on semi-empirical correlations obtained from different experiments. Finite volume method with a staggered grid approach is used for spatial discretization and a fully implicit, time adaptive, multi step method is used for time-dependent discretization. A benchmark calculation which is oriented to the pebble type fuel reactor PBMR-400 and a 3D calculation were presented in HTR-2006 conference and will also be published in Nuclear Engineering and Design (NED) journal. In order to demonstrate the capabilities of TH3D for simulating all block type HTRs, a benchmark calculation which is proposed by IAEA CRP-3 and oriented to the Gas Turbine Modular Helium Reactor (GT-MHR) is performed. Calculations are performed for the steady state case (nominal operation) as well as for Loss of Forced Cooling (LOFC) with and without depressurization. The results obtained from TH3D are compared with the results obtained from several countries participated in this benchmark calculation program by using different code system. In this paper, results of this benchmark calculation and comparisons will be presented. A fuel model for pebble type fuel is implemented in TH3D where heterogeneity of heat production inside the fuel pebble is taken into account. The assumption of homogeneous heat production could be justified for steady state calculation or for slow transient but for fast transient calculation, the assumptions of homogeneous and heterogeneous heat production produce a huge difference for coupled thermal hydraulics and neutronics calculation. In order to show the capabilities of this newly developed code TH3D to couple with a neutronics system, it was coupled with a point kinetics model for a fast reactivity insertion case. In this case all control rods were withdrawn very quickly (with a velocity of 1 m/sec) to the end position. It was assumed that the scram signals were not activated when power or temperature was increased beyond a limiting value during this withdrawal process but the control rods system continued to be withdrawn up to the top position instead of getting down and the coolant flow was reduced by controlling the blowers. The neutronics feedback during this fast reactivity insertion case with homogeneous and heterogeneous fuel model will also be presented.

This paper presents a model developed for determining fuel particle and fuel pebble temperatures in normal operation and transient conditions based on multi-scale modelling techniques. This model is qualified by comparison with an analytical solution in a one-dimensional linear steady state test problem. Comparison is made with finite element simulations of an idealised “two-dimensional” pebble in transient conditions and with a steady state analytical solution in a spherical pebble geometry. A method is presented for determining the fuel temperatures in the individual batches of a multi-batch recycle refuelling regime. Implementation of the multi-scale and multibatch fuel models in a whole-core CFD model is discussed together with the future intentions of the research programme.

This paper presents a model developed for determining fuel particle and fuel block temperatures of a prismatic core modular reactor during both normal operation and under fault conditions. The model is based on multi-scale modeling techniques and has been qualified by comparison with finite element simulations for both steady state and transient conditions. Further, a model for determining the effective conductivity of the block fuel elements — important for heat removal in loss of flow conditions — is presented and, again, qualified by comparison with finite element simulations. A numerical model for predicting conduction heat transfer both within and between block fuel elements has been developed which, when coupled with the above multi-scale model, allows simulations of whole cores to be carried out whilst retaining the ability to predict the temperatures of individual coolant channels and individual coated particles in the fuel if required. This ability to resolve heat transfer on length scales ranging from a few meters down to a few microns within the same model is very powerful and allows a complete assessment of the fuel and structural temperatures within a core to be made. More significantly, this level of resolution facilitates interactive coupling with neutronics models to enable the strong temperature/reactivity feedbacks, inherent in such cores, to be resolved correctly.

The Next Generation Nuclear Plant (NGNP) will most likely produce electricity and process heat, with both being considered for hydrogen production. To capture nuclear process heat, and transport it to a distant industrial facility requires a high temperature system of heat exchangers, pumps and/or compressors. The heat transfer system is particularly challenging not only due to the elevated temperatures (up to ∼ 1300K) and industrial scale power transport (≥50 MW), but also due to a potentially large separation distance between the nuclear and industrial plants (100+m) dictated by safety and licensing mandates. The work reported here is the preliminary analysis of two-phase thermosyphon heat transfer performance with alkali metals. A thermosyphon is a device for transporting heat from one point to another with quite extraordinary properties. In contrast to single-phased forced convective heat transfer via ‘pumping a fluid’, a thermosyphon (also called a wickless heat pipe) transfers heat through the vaporization / condensing process. The condensate is further returned to the hot source by gravity, i.e. without any requirement of pumps or compressors. With this mode of heat transfer, the thermosyphon has the capability to transport heat at high rates over appreciable distances, virtually isothermally and without any requirement for external pumping devices. Two-phase heat transfer by a thermosyphon has the advantage of high enthalpy transport that includes the sensible heat of the liquid, the latent heat of vaporization, and vapor superheat. In contrast, single-phase forced convection transports only the sensible heat of the fluid. Additionally, vapor-phase velocities within a thermosyphon are much greater than single-phase liquid velocities within a forced convective loop. Thermosyphon performance can be limited by the sonic limit (choking) of vapor flow and/or by condensate entrainment. Proper thermosyphon requires analysis of both.

Good safety characteristics are an outstanding feature of High Temperature Reactors (HTR): • The high graphite inventory in the core provides significant thermal inertia. Graphite also has a high thermal conductivity, which facilitates the transfer of heat to the reflector, and it can withstand high temperatures; • The strongly negative Doppler coefficient gives a negative feedback, such that the reactor shuts down by itself in overpower accidental conditions; • The high quality of fuel elements — tri-isotropic (TRISO) coated particles — minimizes operational and accidental fission gas release. The materials selected have resistance to high temperatures; • The low power density enables stabilization of core temperature significantly below the maximum allowable, even in case of severe accidents (such as loss-of-coolant accident). Together, these aspects significantly reduce the risk of massive fission product release, which is one of the attractive features of HTRs. The fuel that is currently used in pebble bed reactors such as AVR, HTR-10 and soon PBMR is based on a homogeneous distribution of coated particles within a fuel pebble. This homogenizes power density in the pebble, but creates a radial temperature gradient across the fuel sphere. Fuel particles placed at its centre has the highest temperature. Reducing the average temperature of particles would help preserve their integrity and maintain the resistance of the first barrier against fission product release. As early as the 1970s, attempts were made to reduce the peak fuel temperature by means of so-called “wallpaper fuel”, in which the fuel is arranged in a spherical shell within a pebble. At that time, the production process was not sufficiently mature and had caused unacceptable damage to the (less performing) BISO particles, which is why this fundamentally promising concept was abandoned. In this paper, proposals will be put forward to improve the production process. This paper further exploits the wallpaper concept, not only from the point of view of temperature reduction, but also for enhanced neutronic performance through improved neutron economy, resulting in reduced fissile material and/or enrichment needs or providing the potential to achieve higher burn-up. Parameters modified were the density of the central fuel-free graphite zone and the packing fraction of the fuel zone. It is demonstrated that this fuel type impacts positively on the fuel cycle, reduces production of minor actinides (MA) and improves the safety-relevant parameters of the reactor. A comparison of these characteristics with PBMR-type fuel is presented. The calculations were performed using Monte Carlo neutron transport and depletion codes MCNP/MCB and the deterministic code WIMS. By comparison with PBMR fuel, the “wallpaper design” of the fuel pebble results in an effective neutron multiplication coefficient increase (by about 2%), which is combined with a decrease of between 3 and 15% in MA production. An improved neutron economy of the heterogeneous design enables enrichment of the “wallpaper type” of fuel to be reduced by more than 6%.

Within a subproject of the RAPHAEL-Program, which is part of the 6 th 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. As in the RAPHAEL program the subproject “Component Development” deals with R&D on components of High Temperature Reactor Technology (HTR), 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 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 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 FLP500 with a radial PMB and the experimental tests and validation of the analytical models to provide recommendations for the investigated blower application as an 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.

US and international applications for large onsite cogeneration (steam and power) systems are emerging as a near term market for the PBMR. The South African PBMR demonstration project applies a high temperature (900°C) Brayton cycle for high efficiency power generation. In addition, a number of new applications are being investigated using an intermediate temperature range (700–750°C) with a simplified heat supply system design. This intermediate helium delivery temperature supports conventional steam Rankine cycle designs at higher efficiencies than obtained from water type reactor systems. These designs can be adapted for cogeneration of steam, similar to the design of gas turbine cogeneration plants that supply steam and power at many industrial sites. This temperature range allows use of conventional or readily qualifiable materials and equipment, avoiding some cost premiums associated with more difficult operating conditions. As gas prices and CO 2 values increase, the potential value of a small nuclear reactor with advanced safety characteristics increases dramatically. Because of its smaller scale, the 400–500MW t PBMR offers the economic advantages of onsite thermal integration (steam, hot water and desalination coproduction) and of providing onsite power at cost versus at retail industrial rates avoiding transmission and distribution costs. Advanced safety characteristics of the PBMR support the location of plants adjacent to steam users, district energy systems, desalination plants, and other large commercial and industrial facilities. Additional benefits include price stability, long term security of energy supply and substantial CO 2 reductions. Target markets include existing sites using gas fired boilers and cogeneration units, new projects such as refinery and petrochemical expansions, and coal-to-liquids projects where steam and power represent major burdens on fuel use and CO 2 emissions. Lead times associated with the nuclear licensing process may support early applications in the 2018–2020 timeframe. This paper summarizes the design options likely to shape these emerging steam and cogeneration applications.

The Republic of South Africa is currently developing the Pebble Bed Modular Reactor (PBMR); an advanced, fourth-generation reactor that incorporates inherent safety features, which require no human intervention and which provide an unprecedented level of nuclear safety. In addition to electrical power generation, the reactor is uniquely suited for a variety of non-traditional nuclear applications including oil sands extraction, desalination, and hydrogen production. A state-of-the-art digital Protection System for the PBMR is currently being developed in conjunction with Westinghouse Electric Company (WEC). The Protection System provides for: • reactor shutdown using two different reactor trip methodologies (dropping of the control rods and insertion of Small Absorber Spheres (SASs) which are composed of boron carbide); • post-event monitoring; and • manual reactor shutdown, which is independent of software-based systems. The reactor shutdown and post-event instrumentation monitoring components of the Protection System are being implemented utilizing the WEC ‘Common Q’ platform, which is comprised of ‘commercially dedicated’ Programmable Logic Controllers (PLCs), colour-graphic Flat Panel Displays (FPDs) with integral touch screens, and high-speed data communication links. High reliability and availability are achieved through component redundancy, continuous automatic self-testing which is run online in a background mode, and implementation of a multi-channel system design which is tolerant to failures. The Protection System is also designed to support periodic surveillance testing through a suite of built-in computer-aided test facilities that are accessible via an FPD interface. These allow various system surveillance requirements to be readily performed in a convenient and systematic manner. This paper discusses the following topics with regard to the PBMR Protection System: development strategy, functional requirements, selection of applicable Codes and Standards, key design specifications, architectural configuration, design and implementation challenges, and unique opportunities that are provided by this type of Protection System.

The question arose whether the correlation for the pressure drop prescribed for cylindrical pebble bed reactors by the Nuclear Safety Commission (KTA) of Germany could still be applied to the proposed annular configuration of the Pebble Bed Modular Reactor (PBMR) currently being developed in South Africa. An approach is described which uses the extended Brinkman equation for fully developed flow together with the original KTA correlation, to account for the resistance of the pebbles, and an effective viscosity, to account for the effect of the walls. A cylindrical packed bed with the same hydraulic diameter as the annular core was first of all considered. The pressure drops for various Reynolds numbers were calculated using a correlation which accounts for the effect of the wall. The formulation of the correlation for an infinite bed was then used along with the Brinkman equation to determine the appropriate values of the effective viscosity to give the same pressure drops. It was then assumed that the effective viscosities obtained in this way could be applied to the annular configuration of the PBMR. The pressure drop through the annular core was then calculated for various Reynolds numbers employing the effective viscosities in the extended Brinkman equation. It was found that the friction coefficients that could be derived from these pressure drops were in good agreement with the friction coefficients obtained from physical experiments performed on a scale model of the PBMR annular core. It was therefore concluded that the strategy followed could be used with the necessary care to predict the pressure drop through the annular core of the PBMR.

The AVR (Arbeitsgemeinschaft Versuchsreaktor) is a pebble bed type helium cooled graphite moderated high temperature reactor that operated in Germany for 21 years and was closed down in December 1988 [1]. The AVR melt-wire experiments [2], where graphite spheres with melt-wires of different melting temperatures were introduced into the core, indicate that measured pebble temperatures significantly exceeded temperatures calculated with the models used at the time [3]. These discrepancies are often attributed to the special design features of the AVR, in particular the control rod noses protruding into the core, and to inherent features of the pebble bed reactor. In order to reduce the uncertainty in design and safety calculations the PBMR Company is re-evaluating the AVR melt-wire experiments with updated models and tools. 3-D neutronics thermal-hydraulics analyses are performed utilizing a coupled VSOP99-STAR-CD calculation. In the coupled system VSOP99 [4] provides power profiles on a geometrical mesh to STAR-CD [5] while STAR-CD provides the fuel, moderator and solid structure temperatures to VSOP99. The different fuel histories and flow variations can be modelled with VSOP99 (although this is not yet included in the model) while the computational fluid dynamics (CFD) code, STAR-CD, adds higher-order thermal and gas flow modelling capabilities. This coupling therefore ensures that the correct thermal feedback to the neutronics is included. Of the many possible explanations for the higher-than-expected melt-wire temperatures, flow bypassing the pebble core was identified as potentially the largest contributor and was thus selected as the first topic to study. This paper reports the bounding effects of bypass flows on the gas temperatures in the top of the reactor. It also presents preliminary comparisons between measured temperatures above the core ceiling structure and calculated temperatures. Results to date confirm the importance of correctly modelling the bypass flows. Plans on future model improvements and other effects to be studied with the coupled VSOP99-STAR-CD tool are also included.

A R&D programme has been launched addressing the needs of the development of an indirect cycle flexible modular HTR operating at 850°C for electricity generation and/or heat production for industrial processes. In the frame of this program, several significant technical challenges required to demonstrate the viability and performance of the system have been successfully addressed. Design and safety analysis needed the development of computational tools, therefore reactor physics, and thermo-fluid dynamics codes have been developed and are now in the process of being validated in the frame of international code-to-code and code to experiment benchmarks. Most importantly, the performance of the HTR/VHTR fuel identified as TRISO-coated particle must prove to be excellent in operating as well as accidental conditions. A manufacturing and quality control process has been developed and now fuel qualification based on irradiation and heating safety tests is being prepared on the basis of irradiation programs in France and in the frame of the GENERATION IV International Forum (GIF) as well as the development of fuel behaviour models including performance data, failure particle prediction and long-term integrity of the coating. Material and component technologies have been investigated in normal and accident conditions for V/HTR objectives. Significant progress has been made for vessel structures and reactor core structural elements. Major challenges still lie ahead for plate type compact intermediate heat exchangers, especially at temperatures above 850°C, but an alternative solution with helical tubes is also being developed. In order to demonstrate that materials have adequate performance over long service life under impure helium environment and constraints, the research programme focuses on microstructural and mechanical property data, long-term irradiation behaviour, corrosion, modelling and codification of design rules as well as qualification of components in representative helium test loops. The potential of this type of reactor for higher performances in terms of fuel burn-up and temperature (VHTR objective) has been explored, in particular for application to hydrogen production. The major research axes on hydrogen production technologies include the development and optimization of high temperature electrolysis and thermo-chemical water splitting processes such as sulphur/iodine or hybrid sulphur. Alternative thermo-chemical hydrogen generation processes operating at lower temperatures are also investigated. This paper addresses the R&D work performed since 2001 and the future work anticipated until 2012, where decisions about a demonstrator could be made at a European level within the Sustainable Nuclear Energy Technological Platform (SNE-TP). This program is strongly connected to the Euratom Framework Programmes as well as to GIF.

Along with the GFR another gas-cooled reactor identified in the Gen IV technology roadmap, the VHTR is studied in France. Some models have been developed at CEA relying on existing computational tools essentially dedicated to the prismatic block type reactor. These models simulate normal operating conditions and accidental reactor transients by using neutronic [1], thermal-hydraulic, system analysis codes [2], and their coupling [3, 4]. In the framework of the European RAPHAEL project, this paper presents the results of the preliminary investigations carried out on the VHTR design. These studies aimed at understanding the physical aspects of the annular core and to identify the limits of a standard block type VHTR with regard to a degradation of its passive safety features. Analysis was performed considering various geometrical scales: fuel cell and fuel column located at the core hot spot, 2D and 3D core configurations including the coupling between neutronic and thermal-hydraulic. From the thermal analysis performed at the core hot spot, the capability to reduce the maximum fuel temperature by modifying the design parameters such as the fuel compact and the fuel block geometry was assessed. The best performances are obtained for an annular fuel compact geometry with coolant flowing inside and outside the fuel compact (ΔT > 50°C). The reliability of such design option should however be addressed with respect to its performance during the LOFC transient (the residual decay heat will be evacuated by radiation during the transient instead of conduction through graphite). As far as the fuel element geometry is concerned, a gain of approximately 50°C can be achieved by making limited changes on the fuel compact distribution in the prismatic block: reduction of the number of fuel compact in the outer ring of the fuel element where the average ratio between coolant channels and fuel compact is smaller. On the other hand, the adopted modifications should also be evaluated with respect to the maximum temperature gradient achieved in the fuel (amoeba effect). In the end, calculations performed on the full core configuration taking into account the thermal feedback showed that the radial positioning of the fuel elements allows to reduce significantly the power peaking factor and the maximum fuel temperature. The gain on the fuel temperature, which varies during the core irradiation, is in the range 100 – 150°C. Several modifications such as the increase of the bypass fraction and the replacement of a part of the graphite reflector by material with better thermal properties were also addressed in this paper.

It is already 10 years since the (European) HTR Technology Network (HTR-TN) launched a programme for the development of HTR Technology, which expanded through 3 successive Euratom Framework Programmes, with many coordinated projects in line with the strategy of the Network. Widely relying in the beginning on the legacy of the former European HTR developments (DRAGON, AVR, THTR...) that it contributed to safeguard, this programme led to advances in HTR/VHTR technologies and produced significant results, which can benefit to the international HTR community through the Euratom involvement in the Generation IV International Forum (GIF). The main achievements of the European programme performed in complement to national efforts in Europe and already taking into consideration the complementarity with contributions of other GIF partners are presented: they concern the validation of computer codes (reactor physics, system transient analysis from normal operation to air ingress accident and fuel performance in normal and accident conditions), materials (metallic materials for the vessel, the direct cycle turbines and the intermediate heat exchanger, graphite...), component development, fuel manufacturing and irradiation behaviour and specific HTR waste management (irradiated fuel and graphite). Key experiments have been performed or are still ongoing, like irradiation of graphite to high fluence, fuel material irradiation (PYCASSO experiment), high burn-up irradiated fuel PIE, safety test and isotopic analysis, IHX mock-up thermo-hydraulic 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 towards industrial demonstration. In line with the orientations of the “Strategic Energy Technology Plan (SET-Plan)” recently issued by the European Commission, which promotes a strategy for the deployment of low carbon energy technologies and mentions Generation IV nuclear systems as one of the key contributors to this strategy, HTR-TN proposes to launch a programme for extending the contribution of nuclear energy to industrial process heat applications addressing jointly 1) The development of a flexible HTR able to be coupled to many different process heat and cogeneration applications with very versatile requirements 2) The development of coupling technologies with industrial processes 3) The possible adaptations of process heat applications which might be needed for coupling with a HTR and 4) The integration and optimisation of the whole coupled system. As a preliminary step for this ambitious programme, HTR-TN endeavours presently to create a strategic partnership between nuclear industry and R&D and process heat user industries.

The PBMR Fuel Plant (PFP), to be constructed at the Pelindaba site near Johannesburg will fuel the first South African Pebble Bed Modular Reactor. The qualification of the PBMR fuel shall be based on past experience with fuel which was produced in the German NUKEM/HOBEG plant and irradiated in the German AVR reactor. Accordingly, the PFP must produce the same fuel as the German plant did, and consequently, the design of the PFP has in essence to be a copy of the NUKEM/HOBEG plant. As a reminder this plant had been operated in accordance with the German regulatory rules which were defined in the years 1970/80. Since then, the requirements with regard to radiological protection, criticality safety and emission control have been significantly tightened, and of course the PFP must be designed in accordance with the most advanced international norms and standards. The implications which follow from these two potentially conflicting requirements, as defined above, are highlighted, and technical solutions are presented. Hence, the change from administrative criticality safety control to technical control, i.e. the application of safe geometry as far as possible, and the introduction of technical solutions for the remaining safe mass regime will be described. A lot of equipment in the Kernel area and in the recycling areas needed to be redesigned in safe geometry. The sensitive processes for Kernel Calcining, for the Coating and the Overcoating remain under safe mass regime, but the safety against criticality is completely independent from staff activities and based on technical measures. A new concept for safe storage of large volumes of Uranium-containing liquids has been developed. Also, the change from relatively open handling of Uranium to the application of containment enclosures wherever release of radioactivity into the room atmosphere is possible, will be addressed. This change required redesign of all process steps requiring the handling of dry Uranium oxides and uncoated Kernels. Finally, the introduction of processes for the near-total recycling of Uranium and chemicals, as well as for decontamination and purification of liquid and gaseous effluents will be presented. These processes were not available from NUKEM/HOBEG fuel facility and needed to be developed now, also following the above mentioned requirements, with respect to criticality safety and radiological protection.

The Very High Temperature Reactor (VHTR) has been selected by the U.S. as the Generation IV technology for the Next Generation Nuclear Plant (NGNP), and both the U.S. and Japan have been developing VHTR concepts based on a prismatic, block-type core design. For these VHTR concepts, the primary coolant (helium) inlet temperature is expected to be in the range 490°C to 590°C and the outlet temperature is expected to be in the range 850°C to 950°C. Passive safety is one of the fundamental requirements for the VHTR, and the VHTR is designed to be passively safe even during Loss of Coolant Accidents (LOCAs) and Loss of Flow Accidents (LOFAs). For the VHTR, these two transient events are referred to as a Low-Pressure Conduction Cooldown (LPCC) and High-Pressure Conduction Cooldown (HPCC), respectively. During both events, the decay heat is conducted through the graphite to the vessel. The heat is transferred from the vessel by thermal radiation and natural convection to a passive Reactor Cavity Cooling System (RCCS). In this paper, we describe parametric studies of LPCC and HPCC events using a 30-degree sector, 3-dimensional ANSYS model of the VHTR, which includes a detailed radiation exchange model between the RPV and RCCS.

Since the last decade, Tractebel Engineering has been involved in several consecutive projects in the field of High Temperature Gas Reactor (HTGR). The objectives of the present project called RAPHAEL ( www.raphael-project.org ) is to provide R&D results in order to consolidate available data on generic V/HTR technologies and to develop innovative solutions to further contribute to the improvement of HTR performances. One of the objectives of the RAPHAEL Sub-project Safety is to qualify tools for performing safety analyses and supporting the safety approach and demonstration. One of the work packages concerns the validation of the existing thermal-hydraulic system codes capabilities needed to perform transient analysis in V/HTR. This validation is carried out by benchmarking against experimental data and by comparing simulation results given by several codes. The current paper presents the work performed at Tractebel Engineering on the simulation of the HE-FUS3 experimental loop — ENEA facility, Brasimone (Italy) — with the MELCOR v.1.8.6 code. The HE-FUS3 loop contains a wide range of components characteristic of a V/HTR like compressor, pipes, diffusers, valves, heaters and heat exchangers. Even if the loop characteristics/configuration is not prototypical of a V/HTR design, the loop is useful to assess the objectives identified by the Project, i.e. helium operating fluid, design pressure and temperature set respectively at 10.5 MPa and 530 °C. The experimental data of the HE-FUS3 loop made available for the benchmark are a set of steady state tests for the thermal-hydraulic characterization of the loop and two transient tests — Loss Of Flow Accidents (LOFA). Moreover, to assess the characteristics of the compressor, data have also been provided from a compressor test campaign. From the code-to-experiment comparison the ability of MELCOR v.1.8.6 to reproduce the experimental results is judged.

The electrical utility in South Africa (Eskom) plan to construct a first of a kind Pebble Bed Modular Reactor (PBMR). It has been recognized that there is a need to adapt the licensing process for the PBMR to ensure that credible and effective licensing process be developed and implemented for this technology. This paper will briefly outline the regulatory framework within South Africa, explain the licensing process adopted and present the challenges that the South African National Nuclear Regulator (NNR) was facing in developing and implementing the licensing process and how these are being addressed. The paper will discuss the update of the regulatory framework and the gaps identified in terms of regulatory requirements needed for such a project. The scope of the regulatory assessment for the licensing of the PBMR is based on the licensing requirements and criteria defined by the NNR in regulatory documents that expand on the current legislative requirements. In addition guidance is provided on selected issues in regulatory guidance documents and position papers. The requirements comprise, besides the general requirements to respect good engineering practice and the ALARA and defense-in-depth principle, specific risk criteria and radiation dose limits. These are categorized for normal operation and operational occurrences as well as for design basis events and beyond design basis events for workers and the public. Additional requirements and recommendations are stipulated by the NNR on safety important areas like quality and safety management, qualification of the nuclear fuel and the core structures, core design, verification and validation of computer codes, source term analysis and others. Selected NNR Position Papers have been developed to elaborate and provide further clarification on NNR requirements. For preparation of the PBMR safety case so-called Key Licensing Issues have been defined and agreed with the applicant. Discussions relating to these Key Licensing Issues allow important nuclear safety aspects identified for the PBMR demonstration plant to be clarified in advance of the safety case submittal.