As a preliminary study of cost estimates for nuclear hydrogen systems, the hydrogen production costs of the nuclear energy sources benchmarking GT-MHR and PBMR are estimated in the necessary input data on a Korean specific basis. G4-ECONS was appropriately modified to calculate the cost for hydrogen production of SI process with VHTR as a thermal energy source rather than the LUEC. The estimated costs presented in this paper show that hydrogen production by the VHTR could be competitive with current techniques of hydrogen production from fossil fuels if CO 2 capture and sequestration is required. Nuclear production of hydrogen would allow large-scale production of hydrogen at economic prices while avoiding the release of CO 2 . Nuclear production of hydrogen could thus become the enabling technology for the hydrogen economy. The major factors that would affect the cost of hydrogen were also discussed.

The alloys Haynes 230 and Inconel 617 are potential candidates for the intermediate heat exchangers (IHX) of (V)-HTR reactors. The behaviour under corrosion of these alloys by the (V)-HTR coolant (impure helium) is an important selection criterion because it defines the service life of these components. At high temperature, the Haynes 230 is likely to develop a chromium oxide on the surface. This layer protects from the exchanges with the surrounding medium and thus confers certain passivity on metal. At very high temperature, the initial microstructure made up of austenitic grains and coarse intra and intergranular M 6 C carbide grains rich in W will evolve. The M 6 C carbides remain and some M 23 C 6 richer in Cr appear. Then, carbon can reduce the protective oxide layer. Then, the alloy loses its protective coating and can corrode quickly. Experimental investigations were performed on these nickel based alloys under an impure helium flow [1]. To predict the surface reactivity of chromium under impure helium, it is necessary to determine its chemical activity in a temperature range close to the operating conditions of the heat exchangers (T ≈ 1273 K). For that, high temperature mass spectrometry measurements coupled to multiple effusion Knudsen cells are carried out on several samples: Haynes 230, Inconel 617 and model alloys 1178, 1181, 1201. This coupling makes it possible thermodynamic equilibrium to be obtained between the vapour phase and the condensed phase of the sample. The measurement of the chromium ionic intensity ( I ) of the molecular beam resulting from a cell containing an alloy provides the values of partial pressure according to the temperature. This value is compared to that of the pure substance (Cr) at the same temperature. These calculations provide thermodynamic data characteristic of the chromium behaviour in these alloys. These activity results call into question those previously measured by Hilpert [2], largely used in the literature.

The coated particles (CP) performance computer code GOLT (Russian abbreviation of Gas-Cooled Fuel) is under development at the A. A. Bochvar All-Russia Research Institute of Inorganic Materials. The main goal of the code is supporting development of fuel for the Gas-Turbine Modular Helium Reactor (GT-MHR). The first version GOLT-v1 has capable to calculate temperature distribution along particle radius, fuel kernel swelling, development of internal pressure under coating due to formation of gaseous fission products and CO, development of stresses and deformation in each coating layer. For TRISO-type particles special probabilistic failure model was developed. According to the failure model integrated probability of silicon carbide failure depends on probability of each dense pyrocarbon layer failure. Probabilistic version GOLT-v2 takes into account possibility of gap formation between buffer and inner dense pyrocarbon layer or between kernel and buffer that influences on maximal fuel temperature and stresses distribution in coating. More detail model of buffer performance at irradiation was developed and included in the code. List of probable coating failure mechanisms was extended. The ability of coating failure due to Kernel-Coating Mechanical Interaction (KCMI) as well as model of failure due to kernel migration was added. Thermo-dynamical code ASTRA is used in some tasks as supporting tool for calculating internal pressure and chemical interaction between SiC coating and fission products and CO. The version GOLT-v3 has accumulated all capabilities of previous versions and included Monte-Carlo analysis for estimation of fraction of failed particles with account of statistical dispersion of structural, materials and operating parameters. In the paper short description of capabilities of last versions of the code is presented. Main attention is putted to results of development version GOLT-v2a for evaluation fuel performance during accidents.

Several fissile kernels are considered for the TRISO particles of (Very) – High Temperature Reactors (V)-HTR. Considering uranium as the fissile nucleus, the basic chemical composition of the TRISO fuel is always made of UO 2 which can be added by some UC 2 carbide. The high level operating temperature of this fuel implies to determine the products formed by the interactions between these two previous compounds. Some UO 2 kernels embedded in black carbon were heated in the 1250–1400°C temperature range in order to determine the kinetics of gaseous species formation [1] and the relative stability of the oxide and carbide phases. After High Temperature Mass Spectrometry (HTMS) experiments, the products formed during the interaction between uranium oxide (UO 2 ) and carbon powders were characterized by various global and punctual analysis methods. The XRD diagram showed the presence of UO 2 and UC phases. The contrasts of density observed by SEM in the Quadrant Back-Scattering Detector (QBSD) mode also allowed to highlight both oxide and carbide phase distributions within the TRISO kernels. During SEM observations, some particles showed particular profiles resulting from “non-uniform” reactional mechanisms as already described by Lindemer [2]. In other singular cases, the interaction between UO 2 and carbon led to the formation of the UC phase in the middle of the kernel, the UO 2 phase remaining at the outside part. Complementary EDS analyses confirmed the results on both oxide and carbide phases. By considering the interference energies between the K-ray of carbon and the N-ray of uranium, the study of the ray intensities consolidated the contrast distributions observed in SEM. Thanks to these results, some assumptions are also advanced concerning the dissolution of oxygen in the UC crystalline structure.

A fuel performance analysis code for a very high temperature gas-cooled reactor (VHTR) COPA ( Co ated Pa rticle) is being developed at the Korea Atomic Energy Research Institute (KAERI). The COPA code consists of nine modules: BURN, TEMTR, TEMPEB, TEMBL, MECH, FAIL, FPREL, ABAQ, and MPRO. The BURN determines neutron flux and fluence at a location of a reactor core, and then calculates a fuel burnup, a fission rate per volume and a fission product inventory throughout a fuel particle and a fuel element. The TEMTR, TEMPEB and TEMBL calculate the temperature distributions in a coated fuel particle, a pebble and a fuel block by using a one-dimensional finite difference method, respectively. The MECH performs mechanical calculations on a coated fuel particle by using a finite element method. The FAIL performs probabilistic calculations to estimate the failure probabilities of the coating layers during an experiment or a reactor operation. The FPREL estimates the migrations of gaseous and metallic fission products through a fuel particle and a fuel element by using a one-dimensional finite difference method. The ABAQ performs the analysis of the crack and debonding in a coated fuel particle. The MPRO calculates the material properties of the kernel, low-density pyrocarbon, high-density pyrocarbon, silicon carbide, matrix graphite, and structural graphite. Each module is used to produce input data for other modules or is inserted into other modules. The COPA code is one of the computer codes taking part in the IAEA-CRP-6 benchmarking program. The stresses and failure fractions calculated by the COPA-MECH and COPA-FAIL showed good agreements with the results by the other countries’ codes. In order to establish a good database of the related material properties, KAERI is participating in an international irradiation experiment, is planning its own irradiation and post-irradiation experiments, and will perform ab-initio calculations on the fuel materials.

The strong correlation between the density and the physical and, mechanical properties of graphite suggests that the method of nondestructive density evaluation could be developed into a characterization technique of great value for the overall improvement of safety of graphite moderator reactors. In this study, the oxidation-induced density changes in nuclear graphite for VHTR were determined by a conventional destructive bulk density measurement method (BM), and by a new non-destructive method based on acoustic microscopy and image processing (AM). The results were compared in order to validate the applicability of the latter method. For a direct comparison of the results from both measurements, two specimens were prepared from a cylindrical graphite sample (1 inch diameter and 1 inch height, oxidized to 10% weight loss at 973 K in air for 5 hours). The specimens were used for characterization by BM and AM methods, respectively. The results show that, even with a large standard deviation of the AM, the density changing trend from both methods appeared the same. This observation may be attributed to the fact that AM images reflect characteristic density changes of the graphite sample through the acoustic impedance changes. This study demonstrates the possibility of using AM as a nondestructive technique for the evaluation of density changes in graphite when a database is prepared through a systematic series of experiments.

An experimental system for in situ high temperature measurements of spectral emissivity of VHTR materials has been designed and constructed. The design consists of a cylindrical block of silicon carbide with several machined cavities for placement of test samples, as well as a black body cavity. The block is placed inside a furnace for heating to temperatures up to 1000°C. A shutter system allows for selective exposure of any given test sample for emissivity measurements. An optical periscope guides the thermal radiation from the sample to a Fourier Transform Infra Red (FTIR) spectrometer which is used for real-time measurements of spectral emissivity over a wavelength range of 0.8μm to 10μm. To specifically address the needs of VHTR applications, the system has been designed for studies with VHTR grade helium environments and air transients. Inlet and outlet gas compositions are measured using a gas chromatograph, which in conjunction with ex situ analysis of the samples by electron microscopy and x-ray diffraction will allow for the correlation of surface corrosion of the materials and their spectral emissivities under different operating and accident conditions.

Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source for a nuclear hydrogen generation. The VHTR can produce hydrogen from heat and water by using a thermo-chemical process or from heat, water, and natural gas by steam reformer technology. Korea Atomic Energy Research Institute (KAERI) is in the process of carrying out a nuclear hydrogen system by considering the indirect cycle gas cooled reactors that produce heat at temperatures in the order of 950°. The nuclear hydrogen system is planning to produce hydrogen by using nuclear energy and a thermo-chemical process. Helium gas is the choice for the coolant of the nuclear hydrogen system since it is an inert gas, with no affinity to a chemical or nuclear activity; therefore a radioactivity transport in the primary circuit of the nuclear hydrogen system is minimal under a normal operation. Moreover, its gaseous nature avoids problems related to a phase change and water-metal reactions and therefore improves its safety. A coaxial double-tube hot gas duct (HGD) is a key component connecting the reactor pressure vessel and the intermediate heat exchanger (IHX) for the nuclear hydrogen system. In this study, a preliminary design analysis for the primary and secondary HGDs of the nuclear hydrogen system was carried out. These preliminary design activities include a preliminary decision on the geometric dimensions, a preliminary strength evaluation and an appropriate material selection. A preliminary decision on the geometric dimensions of the HGDs was undertaken based on three engineering concepts, such as a constant flow velocity model (CFV model), a constant flow rate model (CFR model), a constant hydraulic head model (CHH model), and also based on a heat balanced model (HB model). We compared the geometric dimensions and their preliminary strength evaluation results from the various models.

There is currently renewed interest in high temperature nuclear fission power reactors. The Pebble Bed Modular Reactor (PBMR) is one of several high temperature gas-cooled reactors being investigated by researchers. The South African design of the PBMR is based on the original German design, with the fuel particles (called TRISO particles) being small multilayer spheres.

In a modern high-temperature nuclear reactor, safety is achieved by encapsulating the fuel elements by CVD-layers of pyrolytic carbon and silicon carbide (SiC) to prevent the fission products release. Some studies have raised doubts on the effectiveness of SiC layer as a diffusion barrier to fission fragments due to 110m Ag released from the coated particle at high temperatures ranging from 1500°C to 1600°C [1].

The high-temperature gas-cooled reactor technology is the only nuclear technology capable of achieving coolant temperatures as high as 950 °C and at the same time ensuring safe and efficient production of both electricity and hydrogen. OKBM and GA started independent research in this area in the 1990s. In 1995, OKBM in cooperation with GA started development of the GT-MHR design which combines a safe modular reactor and a power conversion unit based on the high-efficiency Brayton cycle. The power conversion unit in the GT-MHR design has integral configuration, with vertical arrangement of the turbomachine consisting of a synchronous generator and a turbocompressor. Active electromagnetic bearings are used as supports. In order to select optimal technical solutions, the effect of the following factors on the design was considered: vertical or horizontal arrangement, submerged or remote generator with oil bearings, and different turbomachine rotor speeds. Application of electromagnetic bearings and diaphragm coupling between the rotors, integral arrangement of the turbomachine inside the power conversion system vessel, and use of helium as coolant required performance of comprehensive analyses and experiments. For this purpose, the helium turbomachine technology demonstration program was developed and is currently being implemented. This technology demonstration program aims at validating the quantitative and qualitative characteristics of such turbomachine components as electromagnetic and catcher bearings, control system, computer codes, generator, diaphragm coupling, turbocompressor, etc. At the concluding stage of the technology demonstration program, a full-scale turbocompressor model will be tested at a helium test facility. The present paper lists the main parameters of the GT-MHR turbomachine and describes the status of experimental validation of its components.

In order to ensure HTR fuel qualification, as well as reactor safety, particles need to satisfy a set of specifications including particle integrity. To achieve this goal, AREVA NP has been engaged for several years in a R&D program aiming at the development of innovative industrial non destructive evaluation methods for HTR fuel as alternatives to destructive methods. After investigating a number of potential techniques, development has been focused on vision and eddy currents, both aiming at crack detection. High resolution Phase Contrast X-Ray imaging was also studied for structural defects characterization. For all these techniques, besides the development of HTR fuel dedicated control methods, equipment and probes were specifically designed, tested and optimized thanks to experiments conducted on real and artificial flaws, yielding for some of the methods to potential industrialization and quality control performed over 100% of the fuel production.

Considering the need to reduce waste production and greenhouse emissions by still keeping high energy efficiency, various 4 th generation nuclear energy systems have been proposed. As far as graphite moderated reactors are concerned, one of the key issues is the large volumes of irradiated graphite encountered (1770 m 3 for fuel elements and 840 m 3 for reflector elements during the lifetime (60 years) of a single reactor module [1]). With the objective to reduce volume of waste in the HTR concept, it is very important to be able to separate the fuel from low level activity graphite. This requires to separate TRISO particles from the graphite matrix with the sine qua non condition to not break TRISO particles in case of future embedding of particles in a matrix for disposal. According to National Regulatory Systems, in case of limited graphite waste production or of short duration HTR projects (e.g. in Germany), direct disposal without separation is acceptable. Nevertheless, in case of large scale deployment of HTR technology, such approach is not economical and sustainable. Previous attempts in graphite management (furnace, fluidised bed and laser incinerations and encapsulation matrices) dealt with graphite matrix only. These are the reasons why we studied the management of irradiated compact-type fuel element. We simulated the presence of fuel in the particles by using ZrO 2 kernels. Compacts with ZrO 2 TRISO particles were manufactured by AREVA NP. Two original methods have been studied. First, we tested high pressure jet to erode graphite and clean TRISO particles. Best erosion rate reached about 0.18 kg/h for a single nose ending. Examination of treated graphite showed a mixture of undamaged TRISO particles, particles that have lost the outer pyrolytic carbon layer and ZrO 2 kernels. Secondly, we studied the thermal shock method by immerging successively graphite into liquid nitrogen and hot water to cause fracturing of the compact. This produced particles and graphite fragments with diameter ranging from several centimetres to less than 500 μm. This relatively simple and economic method may potentially be considered as a pretreatment step and be coupled with other method(s) before reprocessing and recycling for example.

In the last years considerable efforts have been made at the Institute for Transuranium Elements (ITU) in order to reestablish European knowledge and ability in safety testing of irradiated high temperature reactor (HTR) Fuel Elements. In the framework of the 6th European framework programme a cold finger apparatus (Ku¨FA) furnace, formerly installed at FZ-Ju¨lich (FzJ), has been installed in a hot cell at ITU [Freis 2008] in order to test fission product release under high temperature and non-oxidising conditions. Several analytical methods (e.g. Gamma-spectrometry, mass-spectrometry) have been applied in order to analyse different isotopes released during Ku¨FA tests. After the heating tests, examinations of the fuel elements were performed including scanning electron microscopy (SEM) and micro-hardness testing of coated particles. Individual coated particles were object of heating tests in a Knudsen cell with a coupled mass spectrometer measuring all released species. In order to cover more accident scenarios, a second furnace for oxidising-conditions (air- or water-ingress) was constructed and installed in a cold lab. Furthermore a disintegration apparatus, based on anodic oxidation, was constructed and fuel elements were dissolved obtaining thousands of individual coated particles for further examination. A fully automated irradiated microsphere gamma analyzer (IMGA) is under construction and will be used, in particular, to identify and sort out failed particles.

Understanding of the fission product transport in TRISO fuel particles is fundamental to improving the safety and performance of high temperature gas cooled reactors. Previous experiments showing silver release from TRISO fuel have focused on release measurements and not direct observation of the fission product transport. The possible diffusion of Ag via a grain boundary diffusion mechanism is being examined. By characterizing the SiC grain boundary structure according the coincidence site lattice scheme and detecting diffusion along specific grain boundaries, insight into the relationship between SiC microstructure and Ag release may be obtained. In addition computer modeling is being used to investigate the diffusion of silver through SiC. We employ a multi-scale approach based on ab initio techniques, molecular dynamics, and continuum rate equations in order to establish relationships between complex microstructures and diffusion rates. Initial work has begun on transport through bulk SiC and on building realistic models of grain boundaries in SiC.