[en] A major feature of the Magnox gas cooled reactor design is the graphite core, which acts as the moderator but also provides the physical structure for fuel, control rods, instrumentation and coolant gas channels. The lifetime of a graphite core is dependent upon two principal aging processes: irradiation damage and radiolytic oxidation. Irradiation damage from fast neutrons creates lattice defects leading to changes in physical and mechanical properties and the accumulation of stresses. Radiolytic oxidation is caused by the reaction of oxidizing species from the carbon dioxide coolant gas with the graphite, these species being produced by gamma radiation. Radiolytic oxidation reduces the density and hence the moderating capability of the graphite, but also reduces strength affecting the integrity of core components. In order to manage continued operation over the planned lifetimes of their power stations, BNFL needed to extend their database of the effects of these two phenomena on their graphite cores through an irradiation experiment. This paper will discuss the background, purpose, and the processes taken and planned (i.e. post irradiation examination) to ensure meaningful data on the graphite core material is obtained from the irradiation experiment. (author)

[en] The PEBBED technique provides a foundation for equilibrium fuel cycle analysis and optimization in pebble-bed cores in which the fuel elements are continuously flowing and, if desired, recirculating. In addition to the modern analysis techniques used in or being developed for the code, PEBBED incorporates a novel nuclide-mixing algorithm that allows for sophisticated recirculation patterns using a matrix generated from basic core parameters. Derived from a simple partitioning of the pebble flow, the elements of the recirculation matrix are used to compute the spatially averaged density of each nuclide at the entry plane from the nuclide densities of pebbles emerging from the discharge conus. The order of the recirculation matrix is a function of the flexibility and sophistication of the fuel handling mechanism. This formulation for coupling pebble flow and neutronics enables core design and fuel cycle optimization to be performed by the manipulation of a few key core parameters. The formulation is amenable to modern optimization techniques. (author)

[en] The 10 MWth high-temperature gas-cooled test reactor (termed HTR-10) went into criticality at the Institute of Nuclear Energy Technology (INET) of Tsinghua University in December 2000. As required by China nuclear safety authorities, we had developed nuclear emergency response plan and relevant technical procedures for the implementation of protective actions should an accident occur. This paper presents the technical basis for the development of the HTR-10 nuclear emergency plan. Firstly, it describes briefly the requirements of the China nuclear safety authorities about the nuclear emergency planning and preparedness for research reactors. Then, the paper focuses on the technical development of initiating conditions (ICs) and emergency action levels (EALs) for HTR-10. The ICs and EALs developed are tabulated in this paper. Finally, a brief presentation about the on-site emergency response exercise carried out before the first fuel loading on HTR-10 and other emergency preparedness activities conducted or being planned are given in this paper. (author)

[en] A theoretical model describing the coupling of neutronics, thermohydraulics and fluidization in a fluidized bed nuclear reactor is presented. Simulations of operational transient conditions are performed, viz. a decrease of coolant flow rate and a change of coolant inlet temperature. These simulations show that the fuel temperature remains below the maximum allowable temperature of TRISO fuel, therefore ensuring a safe operational transient. The maximum reactivity is inherently limited and is rapidly compensated by the passive feedback mechanism. (author)

[en] As part of the design calculations for the Pebble Bed Modular Reactor (PBMR), fuel inventories, neutron and gamma sources and decay heat needs to be determined for the fuel spheres. Using the SCALE4.4 code system, a PBMR specific cross section library was created for the ORIGEN-S depletion calculations, assuming a 10-pass refueling system for the PBMR. In this paper the rationale for the creation of the PBMR library is evaluated in terms of the spectrum dependence due to burn-up. The ORIGEN-S PBMR library was further evaluated comparing the results for different parameters calculated with the reactor analysis diffusion code VSOP and the Monte Carlo code MCNP4C. (author)

[en] The concept of a passive, inherently safe, and fail-safe design for an emergency control rod is presented. The functioning of the rod is based solely on inexorable physical laws. The operation of the rod in its emergency function does not require the intervention of a human operator, nor does it rely on any signal from a monitoring or safety system. Although the concept could be applicable to a variety of reactors (provided a normal temperature range is specified), in this paper, the concept is applied to the emergency shutdown of a pebble-bed reactor. The preliminary study presented here demonstrates that the proposed Electro-Magnetic Optimally Scramming Control Rod (EM-OSCR) naturally operates when needed. The rod is held out of the core region by the force of an electromagnet. The force is generated by a current carried by a conductor, a portion of which passes near or through the reactor core region. When the temperature in the conductor increases because of an increase in temperature in the reactor, the conductor resistivity increases. This, in turn, leads to a current decrease. When the current decreases below the level necessary to hold the rod up, the rod is released and it falls into the core under the effect of gravity. (author)

[en] This paper explores the susceptibility of Pebble-Bed Reactors (PBRs) to be used overtly or covertly for the production of plutonium for nuclear weapons. The basic assumption made for the consideration of overt production is that a country would purchase a PBR with the ostensible motive of producing electric power; then, after the power plant was built, the country would divert the facility entirely to the production of weapons material. It is assumed that the country would then have to manufacture production pebbles from natural uranium. The basic assumption made for covert production is that the country would obtain and use a PBR for power production, but that it would clandestinely feed plutonium production pebbles through the reactor in such small numbers that the perturbation on power plant operation would be very difficult to detect. This paper shows the potential rate of plutonium production under such constraints. It is demonstrated that the PBR is a very poor choice for either form of proliferation-intent use. (author)

[en] The basic parameters of critical assemblies created at the ASTRA critical facility of the RRC Kurchatov Institute for simulation of high-temperature reactors are given in the paper. Capability to simulate different HTGR reactors at the ASTRA critical facility is presented. Brief description of the experiments simulating the PBMR reactor being developed in Republic of South Africa is given. Spherical graphite elements with the outer diameter of 6 cm containing about 2.4 g of uranium with the enrichment of about 21% are used as fuel elements at the facility. Fuel particles (kernels) made of uranium dioxide with the diameter of about 500 micrometers, having four-layer coating are evenly distributed in the graphite matrix in the central part (5 cm in diameter) of each spherical fuel element. Graphite blocks are used as moderator in reflectors. Up to 46000 spherical elements were loaded in the core in the most recent experiments. Anticipatory computations of critical parameters of the assemblies performed at the RRC KI for the substantiation of criticality safety of the experiments are described in the paper. The calculations were performed with the Russian MCU program complex, which uses Monte-Carlo method. The MCU code was modified for calculations of the double heterogeneity of the HTGR fuel. Kernels of coated fuel particles were not smeared with graphite matrix of fuel elements, and their heterogeneous structure was directly taken into account in the calculations. Results of the implemented experiments are intended for validation of calculational codes used in HTGR designing and justification of neutronics of HTGRs under development. (author)

[en] Development and recent successful operations of the HTTR in JAERI are coordinated with programs to study applied systems of the promising reactor technology. One such program being carried out from 2001 to 2008 is design and development of the Gas Turbine High Temperature Reactor, GTHTR300, with objective to allow for demonstration of the plant in 2010s in Japan. GTHTR300 features an original design of fuel cycle based on improved HTTR fuel element and of simplified plant system in pursuit of targeted economics with minimum development requirements. The fuel cycle characterizes on high burnup, low power peaking factor and extended refueling interval. The plant system applies such simplified design features as conventional steel RPV, non-intercooled cycle, horizontal single-shaft gas turbine generator and distributed modular maintenance of the overall plant. Research and development essential to validating the design includes component development and control testing. This paper describes the reference plant design and associated continuing R and D activities in JAERI. (author)

[en] High temperature reactors are considered as future inherently safe and efficient energy sources. The presentations covered all the relevant aspects of the existing HTGRs and/or helium cooled pebble bed reactors. They were sorted into 7 sessions: HTR Projects and Programmes; Fuel and Fuel Cycle; Physics and Neutronics; Thermohydraulic Calculation; Engineering, Design and Applications; Materials and Components; Safety and Licensing