[en] The report contains the studies on a molten corium-concrete interaction, which has been recognized as important phenomena of severe reactor accident, where APR-1400 plant has been selected as a reference plant. The purposes of report are to review phenomenological models related to the molten corium-concrete interaction, and to investigate modelling uncertainties by performing sensitivity analysis, and finally to determine that the domestic design requirements of relevant phenomena can be satisfied. Concrete properties, corium amount, corium distribution in the reactor cavity, debris bed configuration, debris power, and heat transfer to the overlying coolant or concrete are considered as important uncertain parameters affecting basemat melt-through. The relevant code modelling have been reviewed and the effects of these parameters are studied through sensitivity analysis

[en] Development of ultrashort high energy radiation source for femtosecond material study with the complementary contribution to advanced reactor neutron research - Generation of ultra-short and high-energy proton/ion beam: Energy : > 1 MeV, Yield : > 10⁸ ions/pulse, Repetition rate, pulse width : 10 Hz, -ps - Generation of ultra-short and high-energy proton beam, Energy : > 10 MeV, Yield : > 1 nC/pulse, Repetition rate, pulse width : 10 Hz, -ps - Generation of ultra-short and high-energy radiation: Energy : > 5 MeV, Yield : > 10⁷ photons/cm²/pulse, Repetition rate, pulse width : 10 Hz.

[en] Design of compact high power THz wave source and development of high speed imaging technology - Simulation and design fo 5 MeV compact microtron - Design and fabrication of high precision electromagnet for microtron - Simulation of electron beam optics and establishment of optimal electron beam injection condition into undulator - Design and fabrication of permanent magnet based quadrupole magnet for focusing optics - Evaluation of radiation dose for radiation shielding and design of shielding block - Structural analysis and optimization of magnetic field distributions for helical variable-period undulator - Design of low-loss THz wave circular waveguide cavity - Design of cavity mirror and investigation of its optical characteristics - Conceptual design of THz wave remote irradiation system.

[en] In this study, a MIDAS/PK code has been developed for analyzing the ATWS (Anticipated Transients Without Scram) which can be one of severe accident initiating events. The MIDAS is an integrated computer code based on the MELCOR code to develop a severe accident risk reduction strategy by Korea Atomic Energy Research Institute. In the mean time, the Chexal-Layman correlation in the current MELCOR, which was developed under a BWR condition, is appeared to be inappropriate for a PWR. So as to provide ATWS analysis capability to the MIDAS code, a point kinetics module, PKINETIC, has first been developed as a stand-alone code whose reference model was selected from the current accident analysis codes. In the next step, the MIDAS/PK code has been developed via coupling PKINETIC with the MIDAS code by inter-connecting several thermal hydraulic parameters between the two codes. Since the major concern in the ATWS analysis is the primary peak pressure during the early few minutes into the accident, the peak pressure from the PKINETIC module and the MIDAS/PK are compared with the RETRAN calculations showing a good agreement between them. The MIDAS/PK code is considered to be valuable for analyzing the plant response during ATWS deterministically, especially for the early domestic Westinghouse plants which rely on the operator procedure instead of an AMSAC (ATWS Mitigating System Actuation Circuitry) against ATWS. This capability of ATWS analysis is also important from the view point of accident management and mitigation

[en] This report describes a applicability of neutron generator to the development of neutron counters regarding with research project, 'development of nuclear material accounting and safeguards technologies'. Selection of the generator, installation, radiation shielding and dose evaluation of the operating area, and its basic performance test result are described in detail. The subject neutron generator is GENIE16GT model of EADS SODERN, France, and it is a D-D type generator with the maximum flux of 2E06 n/s of 2.5 MeV neutron and operating range of 80 μA and 110 kV. A neutron generator has several advantages over a radioisotope neutron source for development and performance test of neutron counting equipments. One of the important advantages is that it can control the emission. No emission of the turned-off neutron source is of importance in the view point of radiation safety, as well as its applicability of variable flux. Pulse mode emission is also applicable to develop the non-destructive technology based on delayed gamma rays or delayed neutrons. 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. We expect, however, the generator would be highly applicable to basic performance test of neutron counters for nuclear materials, to the situation in which radioisotope neutron source is not preferred, to the active neutron counting system whose measurement uncertainty for acceptance level is reasonably high, or to the situation in which the integrated system of passive/active neutron and gamma counting is needed. And, if we maximize its advantages of control the neutron emission, 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] To establish a robust technology for radiopharmaceutical development, we focused on the configuration of fundamental development of radiolabeled compounds for radioimmunotherapy and drug delivery as well as the development of bifunctional chelating agents and radiolabeling methods for the radiopharmaceuticals with highly specific activity to deliver sufficient number of radionuclides to the target site. In this project, we aim to improve the quality of life and the public welfare by fostering the medical application of radioisotopes for the effective treatment of malignant diseases and by developing efficient radiolabeling methods of specific bio-active materials with radioisotopes and new candidates for radiopharmaceutical application. We have established the procedure for the preparation of radiolabeled antibody and biotin with radioisotopes such as ¹⁶⁶Ho, ¹³¹I, ⁹⁰Y and ¹¹¹In for tumour targeting. In the future, these technologies will be applicable to development of radioimmunotherapeutic drug. The combination treatment of radioisotope with anti-cancer agents or chemotherapeutic agents may produce a synergistic static effects in the tumour and this synergism would be exerted via gene level through the activation of a cell death pathway. The combination therapy may be very beneficial for cancer treatment and this can overcome not only the hazards of unnecessary exposure to high radiation level during therapy, but also the tendency for drug resistance caused by chemotherapy. To develop new drug delivery system suitable for CT imaging agent, a chitosan derivative and radiolabed Folate-targeted polymer with ¹³¹I were synthesized. We also carried out the development of DTPA derivatives for CT imaging agent, radiolabeled precursor, and established a highly efficient radiolabeling methodology with lanthanide nuclide. In order to develop neuroreceptor targeting compounds, we synthesized WAY-100635 compound and ^99mTc(CO)₃ precursor from Chrysamine G derivatives. The prepared radiolabeled compounds with ^99mTc are potential radiotracers for evaluating the function of Central Nervous System. Moreover, we established the protocol for radio-conjugation using a new solid-phase reducing agent. In addition, we obtained an imaging candidate for estimating of tumor therapeutic effect. In order to develop a more effective chelating agents, we synthesized cystein- or histidine-based bifunctional chelating agents for labeling with M(CO)₃ precursor [M=^99mTc, ¹⁸⁸Re] and DTPA or DOTA derivatives for radiolabeling of lanthanides and others. For the creation of fusion technology using NT(Nano Technology) and RT(Radiation Technology), we developed manufacturing process of high efficiency radiotherapeutic agent using radiolabeled Carbon NanoTube(CNT). We also carried out in vitro tests to determine the efficacy of radiolabeled nano-material as a tumor indicator. This radioactive nano-sensor is expected to contribute to the development of a target-specific drug delivery system (DDS)

[en] RN2 package, which is one of two fission product-related package in MELCOR, has been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and data structure. To do this, data transferring methods of current MELCOR code are modified and adopted into the RN2 package. The data structure of the current MELCOR code using FORTRAN77 causes a difficult grasping of meaning of the variables as well as waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which lead to an efficient memory treatment and an easy understanding of the code. Restructuring of the RN2 package addressed in this paper includes module development, subroutine modification, and treats MELGEN, which generates data file, as well as MELCOR, which is processing a calculation. The validation has been done by comparing the results of the modified code with those from the existing code. As the trends are the similar to each other, it hints that the same approach could be extended to the entire code package. It is expected that code restructuring will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models

[en] Major purpose of the report is to develop a reactor kinetics model for analyzing the ATWS (Anticipated Transients Without Scram) which is one of severe accident initiating events, and to establish a coupling technology of the developed module with a large computer code for severe accident analysis. For this, a point kinetics module, PKINETIC, has been developed as a stand-alone code whose reference model is selected from the investigation of reactor kinetics models in the current accident analysis codes. In the next step, the MELCOR-PK code has been developed via coupling PKINETIC with the MELCOR (1.8.3) code interconnecting several thermal hydraulic parameters between two codes. In the mean time, the Chexal-Layman correlation in the current MELCOR, which was developed under a BWR condition, is appeared to be inappropriate for a PWR. The PKINETIC module and the MELCOR-PK are compared and verified with a RETRAN calculation. Also, MELCOR-PK is used to analyze the ATWS initiated by a TLOF (Total Loss of Feedwater) transient. (author). 9 refs., 61 tabs., 22 figs

[en] A voxel Monte Carlo code was developed for photon transport calculation by imploying CT and MRI data. The material data was obtained by internet provided by M.J.Berger in NIST, and the photoelectric effect, compton scattering, and pair prodiction were modeled. It was programmed under pentium-III 650 MHz computer with MS visual C++6.0. Mono energy photon was incident on 30X30X30 cm³ water phantom to calculate the depth dose distributions and the results were compared with MCNP results. In the comparison, for 10MeV incident photon the maximum discrepancy was 1.6%. The depth dose distributions were also calculated by varying the voxel size form 1 to 0.25 cm. The results are well agreed each other

[en] Fast neutron fluence was evaluated for KNGR RPV at BOC of the cycle 8 using 3-D discrete ordinates code, TORT. BUGLE96 library, distributed by ORNL, was used for multi-group cross sections. Gd library for poison material was obtained by using NJOY and AMPX processing codes, and added to the BUGLE96 library. Number density of the material was calculated by using CASMO. The results was used as an input of the GIP and material data for TORT input was produced. The power distribution calculated by ROCS code was used as neutron source term. As a result, the maximum neutron fluence was observed at 35 cm below of the center, at 34_.deg. in angle. The maximum neutron fluences at inner wall of the RPV, 1/4T, 1/2T, and 3/4 T were estimated as 3.48E10, 1.78E10, 9.1E9, an d 3.78E9 neutrons /cm², respectively