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.

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 describes the design process followed by Westinghouse Electric South Africa for the insertion of hot internals into the Reactivity Control System (RCS) and Reserve Shutdown System (RSS) Units Under Test (UUTs) at the Helium Test Facility (HTF) at Pelindaba, South Africa. The aim of the UUTs is to allow the validation of the high temperature operation of the RCS and RSS systems for implementation into the proposed Demonstration Power Plant of the PBMR. The units use electrical heaters to obtain pebble-bed reactor thermal conditions for both the control rods and small absorber spheres (SAS) under a pressurized helium environment. Design challenges include providing for strength under elevated temperatures (900°C maximum); pressure boundary integrity (9MPa maximum); separation of different volumes (representing core barrel, reactor citadel and other Reactor Pressure Vessel (RPV) volumes); thermal protection of carbon steel vessels by using thermal insulation; allowing for diverse thermal expansion coefficients of different materials; allowing for depressurization events within the insulation and internals; having access for temperature, pressure, stress and proximity sensors and electrical wiring through high pressure penetrations; and provision for assembly of the hot internals both on and off-site. Several thermal analyses using Computational Fluid Dynamics (CFD) were performed to evaluate both worst-case and operational conditions of the UUTs. Factors that were considered include thermal insulation properties, heat transfer modes (internal radiation, external radiation and natural convection, forced internal convection for cooling) and operating pressure (ranging from 1 to 9MPa). The thermal design uses elements originally proposed for hot gas duct design. The results obtained show that the proposed design satisfies ASME VIII requirements of the pressure boundary and that all challenges are successfully met.

By altering the coolant flow direction in a pebble bed reactor from axial to radial, the pressure drop can be reduced tremendously. In this case the coolant flows from the outer reflector through the pebble bed and finally to flow paths in the inner reflector. As a consequence, the fuel temperatures are elevated due to the reduced heat transfer of the coolant. However, the power profile and pebble size in a radially cooled pebble bed reactor can be optimized to achieve lower fuel temperatures than current axially cooled designs, while the low pressure drop can be maintained. The radial power profile in the core can be altered by adopting multi-pass fuel management using several radial fuel zones in the core. The optimal power profile yielding a flat temperature profile is derived analytically and is approximated by radial fuel zoning. In this case, the pebbles pass through the outer region of the core first and each consecutive pass is located in a fuel zone closer to the inner reflector. Thereby, the resulting radial distribution of the fissile material in the core is influenced and the temperature profile is close to optimal. The fuel temperature in the pebbles can be further reduced by reducing the standard pebble diameter from 6 cm to a value as low as 1 cm. An analytical investigation is used to demonstrate the effects on the fuel temperature and pressure drop for both radial and axial cooling. Finally, two-dimensional numerical calculations were performed, using codes for neutronics, thermal-hydraulics and fuel depletion analysis, in order to validate the results for the optimized design that were obtained from the analytical investigations. It was found that for a radially cooled design with an optimized power profile and reduced pebble diameter (below 3.5 cm) both a reduction in the pressure drop (Δp = −2.6 bar), which increases the reactor efficiency with several percent, and a reduction in the maximum fuel temperature (ΔT = −50 °C) can be achieved compared to present axially cooled designs.

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 also by the Republic of Korea for the Nuclear Hydrogen Development and Demonstration (NHDD) project. One of the key long-lead items for the VHTR is the Reactor Pressure Vessel (RPV). In the absence of active vessel cooling, the RPV temperature during normal operation is determined by the design point selected for the primary coolant inlet temperature and the design of the reactor internal components, including the physical location of riser channels that route the coolant flow to the plenum above the reactor core. For the VHTR, 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. For the RPV, both SA-508/533 steel and higher alloy steels with higher temperature capability (e.g., 9Cr-1Mo-V steel) are being considered. Because of its extensive experience base as an ASME Section III code-approved material for Light Water Reactor (LWR) pressure vessels, SA-508/533 steel is emerging as a strong candidate for the VHTR RPV. However, in order to use this material, the RPV temperature must be maintained below ASME code limits, which are 371°C during normal operation and 538°C for up to 1000 h during accident conditions.

This paper compared the performance of very high temperature reactor (VHTR) plants with direct and indirect closed Brayton Cycles (CBCs) and investigated the effect of the molecular weight of the CBC working fluid on the number of stages in and the size of the single shaft turbomachines. The CBC working fluids considered are helium (4 g/mole) and He-Xe and He-N 2 binary mixtures (15 g/mole). Also investigated are the effects of using LPC and HPC with inter-cooling, cooling the reactor pressure vessel with He bled off at the exit of the compressor, and changing the reactor exit temperature from 700°C to 950°C on the plant thermal efficiency, CBC pressure ratio and the number of stages in and size of the turbo-machines. Analyses are performed for reactor thermal power of 600 MW, shaft rotation speed of 3000 rpm, and IHX temperature pinch of 50 °C.

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.

A low decay heat (implying Spent Fuel (SF) pebbles older than 8–9 years) bulk dry storage section is proposed to supplement a 12-tank wet storage section. Decay heat removal by passive means must be guaranteed, taking into account the fact that dry storage vessels are under ground and inside the building footprint. Cooling takes place when ambient air (drawn downwards from ground level) passes on the outside of the 6 tanks’ vessel containment (and gamma shielding), which is in a separate room inside the building, but outside PBMR building confinement and open to atmosphere. Access for loading / unloading of SF pebbles is only from the top of a tank, which is inside PBMR building confinement. No radioactive substances can therefore leak into atmosphere, as vessel design will take into account corrosion allowance. In this paper, it is shown (using CFD (Computational Fluid Dynamics) modelling and analytical analyses) that natural convection and draught induced flow combine to remove decay heat in a self-sustaining process. Decay heat is the energy source, which powers the draught inducing capability of the dry storage modular cell system: the more decay heat, the bigger the drive to expel heated air through a higher outlet and entrain cool ambient air from ground level to the bottom of the modular cell.

Cooling helium of HTRs is expected to contain a low level of impurities: oxidizing gasses and carbon-bearing species. Reference structural materials for pipes and heat exchangers are chromia-former nickel base alloys — typically alloys 617 and 230 — and, as is generally the case in any high temperature process, their long term corrosion resistance relies on the growth of a surface chromium-oxide that can act as a barrier against corrosive species. This implies that the HTR environment must allow for oxidation of these alloys to occur, while it remains not too oxidizing against in-core graphite. First, studies on the surface reactivity under various impure helium containing low partial pressures of H2, H2O, CO and CH4 show that alloys 617 and 230 oxidize in many atmospheres from intermediate temperatures up to 890–970°C, depending on the exact gas composition. However when heated above a critical temperature, the surface oxide becomes unstable: it was demonstrated that at the scale/alloy interface the surface oxide interacts with the carbon from the material. These investigations have established an environmental area that promotes oxidation. When expose in oxidizing HTR helium, alloys 617 and 230 actually develop a sustainable surface scale over thousands of hours. On the other hand if the scale is destabilized by reaction with the carbon, the oxide is not protective anymore and the alloy surface interacts with gaseous impurities. In the case of CH4-containg atmospheres, this causes rapid carburization in the form of precipitation of coarse carbides on the surface and in the bulk. Carburization was shown to induce an extensive embrittlement of the alloys. In CH4-free helium mixtures, alloys decarburize with a global loss of carbon and dissolution of the pre-existing carbides. As carbides take part to the alloy strengthening at high temperature, it is expected that decarburization impacts the creep properties. Carburization and decarburization degrade rapidly the alloy properties and thus result in an unacceptably high risk on the material integrity at high temperature. Therefore, the purification system shall control the gas composition in order to make this unique helium atmosphere compatible with the in-core graphite as well as with structural materials. This paper reviews the data on the corrosion behavior of structural material in HTR and draws some conclusion on appropriate helium chemistry regarding the material compatibility at high temperature.