In this study several air and water ingress scenarios for the PBMR [1] were simulated by means of the dynamic reactor code TINTE ( TI me-dependent N eutronics and TE mperatures) [2]. The Power Conversion Unit (PCU) and other sub-systems cannot be modelled with the TINTE code and therefore air ingress rates were obtained from Computational Fluid Dynamics (CFD) analysis performed by utilizing the FLUENT code [3]. The use of the TINTE code was previously validated with simulations of the NACOK corrosion experiments [4], [5], [6]. The validations were performed at Forschungszentrum Juelich, however the results are not yet published. The rates of chemical reactions between graphite and gases like O 2 , CO 2 , H 2 O and H 2 are negligible below 400°C. Air and water ingress into the PBMR core at high temperatures can result in corrosion of the PBMR fuel spheres and a possible increase in the fission product release rate. The air ingress scenarios included in this study are; a break in the core outlet pipe at the turbine inlet location, which results in air ingress from the outlet plenum, and a break in a pipe that is connected to the top of the Reactor Pressure Vessel (RPV), which results in air ingress from top of the core. For both transients it is assumed that a Depressurized Loss of Forced Cooling (DLOFC) event takes place prior to the air ingress. The DLOFC leads to high fuel and reflector temperatures that allow higher oxidation rates. The results show that the oxidation of graphite structures in the core is more severe in the case of the outlet pipe break transient. A break in the Core Conditioning System (CCS) heat exchanger circuit during a maintenance mode or following a reactor trip could result in water ingress of up to 1000 kg into the core (the primary system is depressurized at this stage). During the water ingress the CCS continuously cools down the core. Due to the low water ingress rate and lower fuel temperatures, the water ingress transient is not as severe as the air ingress transients.

A control and operation method of a High Temperature Gas Cooled Reactor (HTGR) power plant with Steam Cycle has been investigated, and an adequacy of the method has been examined by plant dynamics analyses. In this plant, the steam cycle system with regenerative and reheat cycle is employed, where the heat generated in the reactor is transferred to the secondary coolant (water and steam) through the steam generator (SG), which has the steam turbine and generator, is installed in the secondary system. The reactor and steam cycle system control methods were evaluated. Analytical results through system modeling concluded that the reactor power changes could be effectively achieved by utilizing primary system helium flow rate control, and overall plant system responsiveness and control could be achieved by employing feed water flow control system and turbine governor control system. The plant dynamics analysis code ASURA, which was developed by MHI, can simulate the major components and systems of this plant. It was used for this examination. In order to adequate of the control and operation method, some cases of plant dynamics analysis by ASURA was conducted in this study.

For direct cycle gas cooled high temperature reactor designs, operating conditions may be limited as a result of excessive maintenance dose rates caused by the Ag-110m source term on the turbine. It is therefore important to accurately predict silver release from fuel during reactor operation. Traditionally diffusion models were used to derive transport parameters from limited irradiation testing of fuel materials and components. Best estimates for all applicable German fuel irradiation tests with defendable uncertainty ranges were never derived. However, diffusion theory and current parameters cannot account for all irradiation and heat-up test results, and for some tests, it appears unacceptably conservative. Other transport mechanisms have been suggested, and alternative calculation models are being considered. In this paper the applicable German irradiation test results are evaluated with a classic diffusion model as well as an alternative model called the Molecular Vapour transport Release (MVR) model. New transport models and parameters for silver in fuel materials are suggested and compared.

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 paper deals with the issue of increasing efficiency of nuclear power plants with the modular high-temperature helium reactor (HTGR) and direct gas turbine cycle. It should be noted that only this combination can highlight the advantages of the HTGR, namely the ability to heat helium to about 1000°C, in comparison with other reactor plants for electricity generation. The HTGR has never been used in the direct gas turbine cycle. At present, several designs of such commercial plants are at the stage of experimental validation of main technical features. In Russia, “OKB Mechanical Engineering” together with “General Atomics” (USA) are developing the GT-MHR project with the reactor power of 600 MW, reactor outlet helium temperature of 850 °C, and efficiency of about 45.2%; the South African Republic is developing the PBMR project with the reactor power of 400 MW, reactor outlet helium temperature of 900 °C, and efficiency of about 42%; and Japan is developing the GTHTR-300 project with the reactor power of 600 MW, reactor outlet helium temperature of 850°C, and efficiency of about 45.6%. As it has been proven by technical and economic estimations, one of the most important factors for successful promotion of reactor designs is their net efficiency, which must be not lower than 47%. A significant advantage of a reactor plant with the HTGR and gas-turbine power conversion unit over the steam cycle is considerable simplification of the power unit layout and reduction of the required equipment and systems (no steam generators, no turbine hall including steam lines, condenser, deaerator, etc.), which makes the gas-turbine power conversion unit more compact and less costly in production, operation and maintenance. However, in spite of this advantage, it seems that in the projects currently being developed, the potential of the gas-turbine cycle and high-temperature reactor to more efficiently generate electricity is not fully used. For example, in modern reactor plants with highly recuperative steam cycle with supercritical heat parameters, the net efficiency of electricity generation reaches 50–55%. There are three methods of Brayton cycle carnotization: regeneration, helium cooldown during compression, and heat supply during expansion. These methods can be used both separately and in combination, which gives a total of seven complex heat flow diagrams. Besides, there are ways to increase helium temperature at the reactor inlet and outlet, to reduce hydraulic losses in the helium path, to increase the turbomachine (TM) rotation speed in order to improve the turbine and compressor efficiency, to reduce helium leaks in the circulation path, etc. The analysis of GT-MHR, PBMR and GTHTR-300 development experience allows identification of the main ways of increasing the efficiency by selecting optimal parameters and design solutions for the reactor and power conversion unit. The paper estimates the probability of reaching the maximum electricity generation efficiency in reactor plants with the HTGR and gas turbine cycle with account of the up-to-date development status of major reactor plant components (reactor, vessels, turbocompressor (TC), generator, heat exchange equipment, and structural materials).

Three systems have been proposed for advanced high temperature gas-cooled reactors (HTGRs): a supercritical carbon dioxide (S-CO 2 ) gas turbine power conversion system; a new M icro C hannel H eat E xchanger (MCHE); and a once-through-then-out (OTTO) refueling scheme with burnable poison (BP) loading. An S-CO 2 gas turbine cycle attains higher cycle efficiency than a He gas turbine cycle due to reduced compression work around the critical point of CO 2 . Considering temperature lowering at the turbine inlet by 30°C through the intermediate heat exchange, the S-CO 2 indirect cycle achieves efficiency of 53.8% at turbine inlet temperature of 820°C and turbine inlet pressure of 20 MPa. This cycle efficiency value is higher by 4.5% than that (49.3%) of a He direct cycle at turbine inlet temperature of 850°C and 7 MPa. A new MCHE has been proposed as intermediate heat exchangers between the primary cooling He loop and the secondary S-CO 2 gas turbine power conversion system; and recuperators of the S-CO 2 gas turbine power conversion system. This MCHE has discontinuous “S”-shape fins providing flow channels with near sine curves. Its pressure drop is one-sixth reference to the conventional MCHE with zigzag flow channel configuration while the same high heat transfer performance inherits. The pressure drop reduction is ascribed to suppression of recirculation flows and eddies that appears around bend corners of zigzag flow channels in the conventional MCHE. An optimal BP loading in an OTTO refueling scheme eliminates the drawback of its excessively high axial power peaking factor, reducing the power peaking factor from 4.44 to about 1.7; and inheriting advantages over the multi-pass scheme because of the lack of fuel handling and integrity checking systems; and reloading. Because of the power peaking factor reduction, the maximum fuel temperatures are lower than the maximum permissible values of 1250°C for normal operation and 1600°C during a depressurization accident.

Due to problems with the availability and the price of water, and the concerns relating to adverse environmental effects of wet cooling systems, the need for water conserving cooling systems has been increasing. Presently, dry cooling accounts for over 30,000 MWe of capacity in more than 30 countries. GT-MHR is specially suited for use of dry cooling due to 1) high efficiency, 2) high heat rejection temperatures and 3) large temperature difference between the turbine inlet and heat rejection temperatures. Higher efficiency means the amount of energy rejected to the cooling per MWe is less. The majority of heat is rejected in precooler and intercooler at helium temperature of more than 100 °C. This results in higher temperature difference for heat rejection. Also due to large temperature difference between the turbine inlet and heat rejection temperatures, changes in ambient temperature have a smaller effect on overall thermal efficiency. Preliminary evaluation shows that pure dry cooling is economical for GT-MHR for water cost of more than 0.8$/m 3 and power cost of 3.5 c/kWh. A combination of dry and wet cooling can reduce large percentage of the water use without affecting the efficiency.

The Pebble Bed Modular Reactor (PBMR), under development in South Africa, is an advanced helium-cooled graphite moderated high-temperature gas-cooled nuclear reactor. The heat output of the PBMR is primarily suited for process applications or power generation. In addition, various desalination technologies can be coupled to the PBMR to further improve the overall efficiency and economics, where suitable site opportunities exist. Several desalination application concepts were evaluated for both a cogeneration configuration as well as a waste heat utilization configuration. These options were evaluated to compare the relative economics of the different concepts and to determine the feasibility of each configuration. The cogeneration desalination configuration included multiple PBMR units producing steam for a power cycle, using a back-pressure steam turbine generator exhausting into different thermal desalination technologies. These technologies include Multi-Effect Distillation (MED), Multi-Effect Distillation with Thermal Vapor Compression (MED-TVC) as well as Multi-Stage Flash (MSF) with all making use of extraction steam from backpressure turbines. These configurations are optimized to maximize gross revenue from combined power and desalinated water sales using representative economic assumptions with a sensitivity analysis to observe the impact of varying power and water costs. Increasing turbine back pressure results in a loss of power output but a gain in water production. The desalination systems are compared as incremental investments. A standard MED process with minimal effects appears most attractive, although results are very sensitive with regards to projected power and water values. The waste heat utilization desalination configuration is based on the current 165 MWe PBMR Demonstration Power Plant (DPP) to be built for the South African utility Eskom. This demonstration plant is proposed at the Koeberg Nuclear site and utilizes a direct, single shaft recuperative Brayton Cycle with helium as working fluid. The Brayton Cycle uses a pre-cooler and inter-cooler to cool the helium before entering the low-pressure compressor (LPC) and the high-pressure compressor (HPC) respectively. The pre-cooler and intercooler rejects 218 MWt of waste heat at 73°C and 63°C, respectively. This waste heat is ideally suited for some low temperature desalination processes and can be used without negative impact on the power output and efficiency of the nuclear power plant. These low temperature processes include Low Temperature Multi-Effect Distillation (LT-MED) as well Reverse Osmosis (RO) with pre-heated water. The relative economics of these design concepts are compared as add-ons to the PBMR-DPP and the results include a net present value (NPV) study for both technologies. From this study it can be concluded that both RO as well LT-MED provide water at reasonable production rates, although a final study recommendation would be based on site and area specific requirements.

A power-generating unit with the high-temperature helium reactor (GT-MHR) has a turbomachine (TM) that is intended for both conversion of coolant thermal energy into electric power in the direct gas-turbine cycle, and provision of helium circulation in the primary circuit. The vertically oriented TM is placed in the central area of the power conversion unit (PCU). TM consists of a turbocompressor (TC) and a generator. Their rotors are joined with a diaphragm coupling and supported by electro-magnetic bearings (EMB). The complexity and novelty of the task of the full electromagnetic suspension system development requires thorough stepwise experimental work, from small-scale physical models to full-scale specimen. On this purpose, the following is planned within the framework of the GT-MHR Project: investigations of the “flexible” rotor small-scale mockup with electro-magnetic bearings (“Minimockup” test facility); tests of the radial EMB; tests of the position sensors; tests of the TM rotor scale model; tests of the TM catcher bearings (CB) friction pairs; tests of the CB mockups; tests of EMB and CB pilot samples and investigation of the full-scale electromagnetic suspension system as a part of full-scale turbocompressor tests. The rotor scale model (RSM) tests aim at investigation of dynamics of rotor supported by electromagnetic bearings to validate GT-MHR turbomachine serviceability. Like the full-scale turbomachine rotor, the RSM consist of two parts: the generator rotor model and the turbocompressor rotor model that are joined with a coupling. Both flexible and rigid coupling options are tested. Each rotor is supported by one axial and two radial EMBs. The rotor is arranged vertically. The RSM rotor length is 10.54 m, and mass is 1171 kg. The designs of physical model elements, namely of the turbine, compressors, generator and exciter, are simplified and performed with account of rigid characteristics, which are identical to those of the full-scale turbomachine elements.

The experimental program that is being conducted at the Matched Index-of-Refraction (MIR) Flow Facility at Idaho National Laboratory (INL) to obtain benchmark data on measurements of flow phenomena in a scaled model of a prismatic gas-cooled reactor lower plenum using 3-D Particle Image Velocimetry (PIV) is presented. A description of the scaling analysis, experimental facility, 3-D PIV system, measurement uncertainties and analysis, experimental procedures and samples of the data sets that have been obtained are included. Samples of the data set that will be presented include the mean velocity field in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic gas-cooled reactor (GCR) similar to a General Atomics Gas-Turbine-Modular Helium Reactor (GTMHR) design. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. The flow in the lower plenum consists of multiple jets injected into a confined cross flow — with obstructions. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to approximate flow scaled to that expected from the staggered parallel rows of posts in the reactor design. The model is fabricated from clear, fused quartz to match the refractive-index of the mineral oil working fluid. The benefit of the MIR technique is that it permits high-quality measurements to be obtained without locating intrusive transducers that disturb the flow field and without distortion of the optical paths. An advantage of the INL MIR system is its large size which allows obtaining improved spatial and temporal resolution compared to similar facilities at smaller scales. Results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). Inlet jet Reynolds numbers (based on the jet diameter and the time-mean average flow rate) are approximately 4,300 and 12,400. The measurements reveal developing, non-uniform flow in the inlet jets and complicated flow patterns in the model lower plenum. Data include three-dimensional vector plots, data displays along the coordinate planes (slices) and charts that describe the component flows at specific regions in the model. Information on inlet velocity profiles is also presented.

The possibility of fuel and graphite degradation due to chemical attack is a perennial issue for HTR’s. For the direct cycle used in the PBMR design, only air ingress is a problem that merits serious attention. Initially, and as reported at a previous conference, investigation of the problem was tackled by assuming worst case conditions for a break at the core outlet pipe to determine what the grace time would be, before counter measures need be taken. The current work identified worst case break positions, quantifying air ingress rates, assuming a Guillotine break. These calculations include first order corrosion reactions in the bottom reflector and the core. Taking the worst possible large break location and the maximum initial air ingress as a determinant, a period of 24 hours was determined to be sufficient to prevent both serious fuel and core structure degradation. The acceptability of the extent of corrosion will be determined by the Safety Analysis Report (SAR), which is under preparation. However, it was realized that a more realistic specification and analysis of the problem was required to enable design decisions to be made, and a more detailed model of the break and the Main Power System (MPS) cavities was developed. This includes the maximum movement of large piping postulating a Double Ended Guillotine Break (DEGB) at worst possible locations. Further calculations on the improved model are described that investigate the influence of various pipe separations i.e. 50 mm and 500 mm at the turbine inlet. A strong correlation between the opening size and total core corrosion rate was confirmed. The simulation also established an approximate duration for air to be expelled to stop further ingress and the volume flow requirements for the inert gas system using helium or nitrogen.