[en] Gamma/neutron measurement can be performed in a shielded glove-box in a case of small spent fuel samples because it has a much lower radioactivity than that of the fuel assembly or rod. Although there are several measurement methods in examining the spent fuel characteristics, gamma and neutron counting methods are most frequently used as the NDT method. In this study, a gamma/neutron combined system was designed and manufactured for the neutron coincidence counting and gamma isotopic analysis for small spent fuel samples before measuring pyro-processing products. After manufacturing all the components of the gamma and neutron measuring system, inactive performance tests were carried out by using a checking and standard gamma and neutron sources, and then the gamma/neutron combined system was installed in the shielded glove-box. Active tests were performed by using some spent fuel samples, and some problems found in the course of the tests were modified and improved. As a result of the active performance test, burnup and the Pu/U ratio determined by gamma spectrometry resulted in good agreement with those of chemical analysis within 5% and 0.8 % difference, respectively. Therefore, the precision of the gamma measuring system was verified. If this technique is used with some NDT methods, it will be able to be usefully applied to nuclear material accounting and safeguards in a spent fuel processing facility in the near future

[en] In this study, the development of an active neutron coincidence counting has been described as one of non-destructive assay method for accounting of nuclear materials in a pyroprocess. Using previously developed passive neutron coincidence counting system, we modified it, added an interrogation source of a radio-isotope or a neutron generator, and evaluated its performances. Based on the MCNPX simulation results, we prepared proper moderators for an interrogation source and performed experiments with natural and low enriched uranium oxide powder samples. For the neutron generator as an interrogation source, the average specific coincident counts (per unit mass of ²³⁵U) was 2.64 cps/g-²³⁵U in the range of 0.5 - 3.5 kg natural uranium oxide ( up to 21.7g of ²³⁵U). Average background standard deviation was 9.57 cps, signal standard deviation after background subtraction was 13.8 cps. In case of using ²⁵²Cf neutron source, the error was 6% to 10% for 15 ∼ 30 g-²³⁵U, 5.2% for 67.2 g-²³⁵U, and 3.86% for 100.8 g-²³'5U. The experimental results showed larger error for the case of neutron generator as an interrogation source than the case of isotope source, and there should be more efforts to reduce the error. If we overcome this challenge and use its advantage, the safeguards techniques for a spent fuel management process will be expanded and have a high confidence by combining with other conventional non-destructive assay methods

[en] A consistent general order nodal method for solving the 3-D neutron diffusion equation in (x-y-z) geometry has been derived by using a weighted integral technique and expanding the spatial variables by the Legendre orthogonal series function. The equation set derived can be converted into any order nodal schemes. It forms a compact system for general order of nodal moments. The method utilizes the analytic solutions of the transverse-integrated quasi-one dimensional equations and a consistent expansion for the spatial variables so that it renders the use of an approximation for the transverse leakages no necessary. Thus, we can expect extremely accurate solutions and the solution would converge exactly when the mesh width is decreased or the approximation order is increased since the equation set is consistent mathematically

[en] A new general high order consistent nodal method for solving the 3-D multigroup neutron kinetic equations in (x-y-z) geometry has been derived by expanding the flux in a multiple polynomial series for the space variables without the quadratic fit approximations of the transverse leakage and for the time variable and using a weighted-integral technique. The derived equation set is consistent mathematically, and therefore, we can expect very accurate solutions and less computing time since we can use coarse meshes in time variable as well as in spatial variables and the solution would converge exactly in fine mesh limit

[en] The objective of this project is to analyze the safeguard ability of pyroprocess facility and to establish the safeguards system for pyroprocess by developing the technology of nuclear material accounting for unit process, surveillance technology and nuclear characteristics analysis technology which are needed to demonstrate the safeguards technology of pyroprocess. In order to establish the nuclear material accountancy for PRIDE the unified NDA was designed by integrating neutron detection, gamma ray detection, and mass measurement. The surveillance system of PRIDE includes gamma ray detector system for tracing the position of nuclear material in PRIDE and the safeguards system was designed considering the characteristics of nuclear material accountancy and surveillance equipment and monitoring the main factors of process equipment. Based on the design of safeguards system for PRIDE a simulation program for the integrated accounting and surveillance information system has been developed and tested. The safeguard ability analysis code for pyroprocessing facility has been designed to develop a Pyroprocessing Material flow and Material Unaccounted For Uncertainty Simulation (PYMUS) program based on the result from safeguard ability analysis by ROK IAEA MSSP. For novel technologies development of nuclear material accountancy by domestic and international cooperation, the application of XRF, SINRD, PNAR, FRAM, LIBS to NMA has been studied. The safeguards system of PRIDE will be referred to verify the safeguards approach and implementation techniques for pyroprocessing facility on international cooperation(ROK US Joint Fuel Cycle Study and ROK IAEA Member State Support Program). The results of this project will contribute to increase the nuclear transparency for realizing the pyroprocessing technology of the ROK as well as to establish the safeguards technology for pyroprocessing facility

[en] This project is aimed at the development of nuclear material accounting and safeguards technology. Nuclear material accountancy technology for an each unit process and nuclear characteristic analysis technology to demonstrate the safeguards technology for a pyroprocessing facility have been developed during the first phase of the project. A study for analyzing the safeguardability of pyroprocessing facility and preliminary evaluation has also been carried out. The safeguards technology system for electro-reduction process has been established to develop the unit process nuclear material accountancy technology through nuclear material accounting in ACPF and performance test of surveillance equipment using spent fuels and neutron source. Nuclear material accountancy measure for electro-refining process where various kinds of nuclear materials are generated compared to electro-reduction process has been constructed, and its performance test has been conducted as well. A Gamma/neutron integrated system has been developed as a nuclear characteristic analysis technology for pyroprocess nuclear material, the possibility of Pu and U measurement has been analyzed using FRAM, and fundamental experiment has been performed to examine whether LIBS technology is applicable to nuclear material accountancy of pyroprocess. A preliminary concept design of safeguards system for pyroprocessing facility and basic design of computer code for analyzing the safeguardability have been carried out to perform the safeguardability analysis and preliminary evaluation for pyroprocessing facility, and a study for analyzing the safeguardability of KAPF which has scale of 100 MT-HM/year has been conducted in collaboration with LANL. Also, an IAEA Member State Support Program for establishing safeguards approach technology for pyroprocessing facility ('Support for Development for a Safeguards Approach for a Pyroprocessing Plant') has been conducting in cooperation with KINAC. It is expected that the results of this project would contribute to increase the nuclear transparency for realizing the pyroprocessing technology of the ROK as well as to establish the safeguards technology for pyroprocessing facility

[en] During the first phase of R and D program conducted from 1997 to 1999, nuclear material safeguards studies system were performed on the technology development of DUPIC safeguards system such as nuclear material measurement in bulk form and product form, DUPIC fuel reactivity measurement, near-real-time accountancy, and containment and surveillance system for effective and efficient implementation of domestic and international safeguards obligation. For the nuclear material measurement system, the performance test was finished and received IAEA approval, and now is being used in DUPIC Fuel Fabrication Facility(DFDF) for nuclear material accounting and control. Other systems being developed in this study were already installed in DFDF and being under performance test. Those systems developed in this study will make a contribution not only to the effective implementation of DUPIC safeguards, but also to enhance the international confidence build-up in peaceful use of spent fuel material. (author)

[en] DUPIC safeguards R and D in the second phase has focused on the development of nuclear material measurement system and its operation and verification, the development of nuclear material control and accounting system, and the development of remote and unmanned containment/surveillance system. Of them, the nuclear material measurement system was authenticated from IAEA and officially used for IAEA and domestic safeguards activities in DFDF. It was also verified that the system could be used for quality control of DUPIC process. It is recognised that the diagnostic software using neural network and remote and unmanned containment/surveillance system developed here could be key technologies to go into remote and near-real time monitoring system. The result of this project will eventually contribute to similar nuclear fuel cycles like MOX and pyroprocessing facility as well as the effective implementation of DUPIC safeguards. In addition, it will be helpful to enhance international confidence build-up in the peaceful use of spent fuel material

[en] The scope of the project includes steam explosion experiments using prototypic materials, hydrogen combustion experiments for the development of a quenching screen, the development of a debris coolability model and the development of a new core catcher concept. In the steam explosion research, 15 steam explosion experiments were performed using reactor materials including prototypic partially oxidized corium (UO₂ + ZrO₂ + Fe + Zr). The layer inversion experiments using partially oxidized corium were also conducted among those tests and confirmed previous works. The evaluation technique for the thermodynamic phase diagram using the GEMINI code was also established. In addition, the material effect on the occurrence of a steam explosion was examined by establishing the physical and chemical analysis on the debris. For the precise measurement of the very high temperature and the very fast phenomena and so on, the two-color pyrometry and the axial tomography were improved. Finally, the measures drawn to confine the reactor core melt were registered as and applied to national and international patents. For the development of a quenching mesh for the protection of the instruments from hydrogen combustion in the reactor, about 30 experiments were performed to develop a model for the estimation of the quenching mesh and finally a quenching mesh was developed. For the establishment of corium coolability estimation techniques, the models in a steam explosion code for the material effect, multi- dimensional effect and melt breakup were modified to estimate reactor cases. Along with collection and estimation of the experimental data obtained from OECD/MCCI project, the methodology for the MCCI analysis was improved by using the data and the estimation technique for the reactor case was established

[en] This report describes the fundamentals of the Laser Induced Breakdown Spectroscopy(LIBS), and it describes the quantitative analysis method in the vacuum condition to obtain a high measurement accuracy. The LIBS system employs the following major components: a pulsed laser, a gas chamber, an emission spectrometer, a detector, and a computer. When the output from a pulsed laser is focused onto a small spot on a sample, an optically induced plasma, called a laser-induced plasma (LIP) is formed at the surface. The LIBS is a laser-based sensitive optical technique used to detect certain atomic and molecular species by monitoring the emission signals from a LIP. This report was described a fundamentals of the LIBS and current states of research. And, It was described a optimization of measurement condition and characteristic analysis of a LIP by measurement of the fundamental metals. The LIBS system shows about a 0.63 ∼ 5.82% measurement errors and calibration curve for the 'Cu, Cr and Ni'. It also shows about a 5% less of a measurement errors and calibration curve for a Nd and Sm. As a result, the LIBS accuracy for a part was little improved than preexistence by the optimized condition