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Graphite is used as a structural material, neutron moderator and reflector in some designs of nuclear reactors. The core of a prismatic or pebble bed type High Temperature Reactor (HTR) is also comprised of graphite components. In addition to HTRs, it is also used as a structural material in the Molten Salt Breeder Reactor (MSBR). It’s response to postulated accidental condition is a critical aspect of probabilistic safety assessment studies for such futuristic reactor designs. Generation of graphite aerosol particles has been shown as a possible hazard in the case of an air ingress accident for HTRs. These aerosol particles act as a carrier for radioactive source term and their characteristics decide the impact of accident. This study aims to measure and interpret number and mass characteristics of graphite particles generated by exposing graphite to varying high temperature conditions. An exclusive sophisticated facility is developed, tuned and employed for measuring the aerosol characteristics at extreme temperature and concentration conditions. Apart from measuring number and mass size distributions of generated particles at different temperatures and flows, emission of CO and CO2 was also monitored. Particle concentration and CO gas concentration was found to be maximum at 700 °C, the transition temperature for the generation process. Evaporation-condensation and transport of unburnt carbon particles could be linked to the observed particles in nucleation and accumulation mode size ranges, respectively. Mechanism of particle generation and graphite particle transformation at different temperatures is also discussed.

To identify the distribution of fuel debris remaining in the reactor vessel and/or the containment vessel of Fukushima Daiichi NPS, we focused on the inverse estimation of radioactive source distribution using the measured radiation counts. The M aximum L ikelihood- E xpectation M aximization (ML-EM) and the M oore- P enrose M atrix I nverse (MPMI) methods are examined. The ML-EM method has been used for the image reconstruction of computed tomography, and the MPMI method is one of the solution methods for simultaneous linear equations with the underdetermined condition. A simple calculation model simulating a containment vessel was constructed including detectors and radiation sources. In an actual situation, a sufficient number of radiation measurement positions would not be available owing to the complexity of structures inside the containment vessel. Thus, the number of radiation measurement points (number of constraints) is smaller than that of radiation source positions. It means that an underdetermined inverse problem should be solved. The detection probability of radiation (neutron or photon) is calculated by the adjoint transport calculation since the detection probability is used as the coupling coefficient between radiation counts at a detector and a radioactive source. The result of estimation using the ML-EM or the MPMI method indicates that the accuracy of estimation depends on the distance between a radiation source and a detector, and measurement positions of radiation count. The ML-EM and the MPMI methods show different prediction accuracy depending on the prediction condition. It is found that reasonable prediction accuracy would be obtained when the detectors are placed at the vicinity of radiation sources of interest.

Considerable efforts have been taken to estimate the radioactive source term by using integrated computer models, but uncertainties exist in using these models, partly due to the fact that the results cannot be validated using real-world large-scale experiments. The purpose of this research is to identify the MAAP4 phenomenological model parameters with large influences in predicting environmental releases. The identified parameters were compared for different accident scenarios and plant configurations. Two-step screening of the influential model parameters through the Latin hypercube-rank correlation and the one-at-time method was used for the comparative analysis. From the integral analyses, it was found that the source term error propagation was much larger in the containment and in the secondary system. However, error propagation was not necessarily larger inside the containment than in the secondary system. There are common phenomenological parameters with significant influences on the source term simulations: primary depressurization, molten core-concrete interaction, and cladding oxidation models. More focused research regarding the effect of aforementioned identified phenomenological models on the source term calculations would be necessary to better predict the amount of environmental radionuclides release for the severe accidents. However, the detailed uncertainty and sensitivity analysis involving phenomenological model parameters may have to be done in a plant-specific way. Some phenomenological model parameters that showed significant impact on the global source term uncertainty in one plant type did not necessarily have the same impact on another plant. A sensitivity study involving a typical PWR may not have same implications in other designs, even if a same accident scenario with similar accident phenomena is analyzed. Such sensitivity study would be much more beneficial for standardized plant designs.

This study reviewed the current status of disused sealed radioactive source (DSRS) in Korea and analyzed its characteristics according to three factors, i.e., IAEA recommended classification system of radioactive waste based on the half-life of radionuclides, potential risk index based on activity divided by the corresponding radionuclide-specific ‘D value’ and legal classification scheme of radioactive waste based on the concentration of radionuclides. The optimum management procedure was derived based on above three factors above and applied to DSRS which are currently being stored in the centralized temporary storage facility for the radioisotope (RI) waste operated by Korea Radioactive Waste Agency (KORAD). The results of this study would be used as technical background document to draw up an optimum management plan of DSRS in the near future.

During severe accidents, fission products as aerosols are released into the containment and consequently deposited on walls. The condensate flow on walls could wash the deposited aerosols downwards to the sump. Therefore, the aerosol wash-down process on walls is significant to evaluate radioactive source term distribution in nuclear containment. In this paper, the containment code COCOSYS coupled with aerosol wash-down model AULA and condensate coverage model CONRAG is used to analyze the aerosol wash-down process of THAI-AW3 experiment and ‘Generic Containment’. THAI-AW3 experiment was conducted by Becker Technologies to investigate insoluble aerosol wash-down behavior from the vertical, quasi-horizontal surfaces, and a small puddle. The COCOSYS calculated results shows the aerosol wash-down is predicted well, as well as the thermal-hydraulics in THAI-AW3 experiment. There are no significant differences of aerosol wash-down efficiency between the calculation and measurement. The acceptable COCOSYS is to enable further analysis on the aerosol wash-down process of the Generic Containment within SARNET2 activity under a full severe accident scenario in order to test the aerosol wash-down model. The evolutions of the containment’ thermal-hydraulics are analysed. The calculation offers the aerosol distribution, the aerosol wash-down mass, and the aerosol distribution inside containment over time. The comparisons with previous COCOSYS simulations indicate that the modification of COCOSYS, e.g. the coupling with condensate coverage model CONRAG, is necessary to investigate the aerosol wash-down.

A gamma scattering method for determining the water content in concrete sample applicable to any porous building material is described. The experimental setup consists of collimated 137Cs radioactive source and a high resolution HPGe detector based energy dispersive gamma ray spectrometer. Concrete samples of uniform density ≈2.4 g/cm3 are chosen for the study and the scattered intensities found to vary with the amount of water present in the specimen. The scattered intensities are corrected for attenuation effects and the results obtained with reference to a dry sample are compared with those obtained by gravimetrical and gamma transmission methods. A good agreement is seen between gamma scattering results and those obtained by gravimetric and transmission methods within accuracy of 6% and

Adequate knowledge of burn up levels of fuel elements within a research reactor is of great importance for its optimum operation. Such knowledge is required for the monitoring of reactivity parameters and flux and power distributions throughout the reactor core, the estimation of the radioactive source term needed in accidental situations analysis, the evaluation of the amount of fissile materials present at any moment within the fuel for safeguards purposes and the estimation of cooling and shielding requirements for interim storage or transport of spent fuel elements. This paper presents the approach of fuel burn up evaluation used at the NUR research reactor. The approach is essentially based upon the utilization of BURNUR.SYS code, an in-house developed software. BURNUR.SYS is an object oriented program under DELPHI 7 that integrates the cell calculation code WIMSD-4 and the core calculation code CITVAP. BURNUR.SYS calculates the evolution in time of pertinent quantities such as: the concentrations of U235 and others actinides, the concentrations of major poisons (Xe135 and Sm149), the distributions of power densities and burn up levels within fuel elements, the effective multiplication factor and core reactivity. The results are displayed in user friendly graphical and numerical formats.

In this paper, the angular exposure dose build up factor for combined photon energies of 6.13 and 7.12 MeV, emitted from a N-16 pipe radiator penetrating single layer shields of concrete and aluminium is calculated. Experimentally determined angular flux spectra were analysed. The angular exposure build up factor is defined for the combined energy source photons and the comparison with the values of angular build up factor for the same combined energy source of disc geometry is considered.

Scattered energy spectra produced from the interaction of multi-energy photons, emitted from an N-16 radioactive source, with concrete, steel and lead single layer shields will be examined. The influence of the material atomic number and the bremsstrahlung contribution to the examined spectra are also considered.

In this paper, experimentally determined angular flux spectra for combined source energies of 6.13 and 7.12 MeV, penetrating single layer shields of concrete, steel and lead are examined. A method for the calculation of the scattered photons energy, when multi-energy photons are emitted from a radioactive source, is described. Changes in the total scattered photon energy spectrum, as a function of polar angle and shield thickness are discussed.

The dry storage pit for spent fuel samples is located in the reception area leading to the shielded hot cells in the hot laboratory at Inchas. It is designed to contain six storing cavities, each provided with a metallic basket, and symmetrically arranged in a cylindrical lead container, which is mechanically rotated under a concrete cover. Each storing cavity allows for safe shielding and adequate storage of gamma sources equivalent to 5000 MeV-Ci. The radioactive source is transferred from the transporting container to the storing cavity through the concrete cover by the rotation of the storage lead container. The storing cavity is ready for receiving the radioactive source when it faces the cavity in the transporting container. The detailed design data are given.