[en] Korea Atomic Energy Research Institute, using the ANS low power and shutdown (LPSD) probabilistic risk assessment (PRA) Standard, evaluated the LPSD PSA model of the KSNP, Yonggwang Units 5 and 6, and identified the items to be improved. The evaluation results of human reliability analysis (HRA) of the post-accident human errors in the LPSD PSA model for the KSNP showed that 10 items among 19 items of supporting requirements for those in the ANS PRA Standard were identified as them to be improved. Thus, we newly carried out a HRA for post-accident human errors in the LPSD PSA model for the KSNP. Following tasks are the improvements in the HRA of post-accident human errors of the LPSD PSA model for the KSNP compared with the previous one: Interviews with operators in the interpretation of the procedure, modeling of operator actions, and the quantification results of human errors, site visit. Applications of limiting value to the combined post-accident human errors. Documentation of information of all the input and bases for the detailed quantifications and the dependency analysis using the quantification sheets The assessment results for the new HRA results of post-accident human errors using the ANS LPSD PRA Standard show that above 80% items of its supporting requirements for post-accident human errors were graded as its Category II. The number of the re-estimated human errors using the LPSD Korea Standard HRA method is 385. Among them, the number of individual post-accident human errors is 253. The number of dependent post-accident human errors is 135. The quantification results of the LPSD PSA model for the KSNP with new HEPs show that core damage frequency (CDF) is increased by 5.1% compared with the previous baseline CDF It is expected that this study results will be greatly helpful to improve the PSA quality for the domestic nuclear power plants because they have sufficient PSA quality to meet the Category II of Supporting Requirements for the post-accident human errors in the ANS LPSD PRA Standard

[en] The report describes the theoretical study and the experiment of ultrasonic guided waves which are attractive for the online and real time inspection of pipes installed in nuclear power plants. The following list presents the main contents of the report. The characteristics of guided wave, mode, dispersion curves, and Short Time Fourier Transform(STFT) which is necessary for the analysis of guided wave signal were studied and summarized. The dispersion curves of steam generator tube was calculated. The incidence angle of ultrasonic wave to generate the specific mode was also calculated. And the modes were identified through the measurements of group velocity and STFT. We also identified the mode conversion of guided wave when it passed through the bent part of the tube. We integrated FR signal, Fourier transform data, phase velocity dispersion curve, group velocity dispersion curve, and STFT into one comprehensive method that could identity and analyze the complicate modes of guided waves. We prepared the criteria which are necessary to the selection of the optimal mode for the detection of flaws, and selected the most suitable mode in the detection of flaws. Finally we validated it through the experiments

[en] This report describes the process development and optimization process of the barrier coating applicable to the inside of cladding and the development of innovative coating technology that can prevent interfacial reaction (FCCI) of metal nuclear fuel and cladding in SFR. Using the circulation system of the electro-plating the inner surface of the cladding was successfully coated by a chromium coating layer with uniform thickness within 13 μm and no defect inside. The diffusion couple test with a rare earth alloy (50Ce-50Nd) was set as a barrier property evaluation methodology, and a coating layer formed on the inner surface of the cladding proved to successfully perform a barrier function to prevent FCCI. Also, it was found that the thickness of the plating layer and the average crack length decrease as the off time is lower when the pulse current is applied. In order to develop a coating layer that can withstand high temperatures in the event of a reactor accident, a Cr/Cr₂N multi-coating layer was selected and developed. Cr and Cr/Cr₂N barrier coatings were applied to the inner-surface of cladding and their barrier properties were proved by diffusion couple tests using surrogate fuel materials. We also studied effects of cold working and pilger roll on cladding tube and liner materials for evaluation on manufacturing liner cladding by process parameters. Evaluate the characteristics of the prototype according to the pilger process and number of times. In the case of material Titanium Grade 2, plastic deformation processing was easier than Zircaloy-4. A titanium liner tube with a thickness of 100 μm could be fabricated by controlling the hardness characteristics by using heat treatment and repeated pilger process. Using the 9Cr-2W-NbVB series tube and zircaloy-4 tube, the manufacturing process with the cladding liner layer (thickness: 50 μ) was secured and the prototype was manufactured (length: 1 m)

[en] Well-ordered high crystalline LiNi_0.5Mn_1.5O₄ spinel has been readily synthesized by a molten salt method using a mixture of LiCl and LiOH salts. Synthetic variables on the synthesis of LiNi_0.5Mn_1.5O₄, such as synthetic atmosphere, LiCl salt amount, synthetic temperature, and synthetic time, were intensively investigated. X-ray diffraction (XRD) patterns and scanning electron microscopic (SEM) images showed that LiNi_0.5Mn_1.5O₄ synthesized at 900 and 950 deg. C have cubic spinel structure (Fd3-bar m) with clear octahedral dimension. LiNi_0.5Mn_1.5O₄ spinel phase began to decompose at around 1000 deg. C accompanied with structural and morphological degradation. LiNi_0.5Mn_1.5O₄ powders synthesized at 900 deg. C for 3 h delivered an initial discharge capacity of 139 mAh/g with excellent capacity retention rate more than 99% after 50 cycles

[en] The development of the ripple pattern during the ion beam sputtering (IBS) is expounded via the evolution of its constituent ripples. For that purpose, we perform numerical simulation of the ripple evolution that is based on Bradley–Harper model and its non-linear extension. The ripples are found to evolve via various well-defined processes such as ripening, averaging, bifurcation and their combinations, depending on their neighboring ripples. Those information on the growth kinetics of each ripple allow the detailed description of the pattern development in real space that the instability argument and the diffraction study both made in k-space cannot provide.

[en] The paper presents a proceduralised human reliability analysis (HRA) methodology, AGAPE-ET (A Guidance And Procedure for Human Error Analysis for Emergency Tasks), covering both qualitative error analysis and quantification of human error probability (HEP) of emergency tasks in nuclear power plants. The AGAPE-ET method is based on the simplified cognitive model. By each cognitive function, error causes or error-likely situations have been identified considering the characteristics of the performance of each cognitive function and influencing mechanism of the performance influencing factors (PIFs) on the cognitive function. Then, error analysis items have been determined from the identified error causes or error-likely situations and a human error analysis procedure based on the error analysis items is organised to help the analysts cue or guide overall human error analysis. The basic scheme for the quantification of HEP consists in the multiplication of the BHEP assigned by the error analysis item and the weight from the influencing factors decision tree (IFDT) constituted by cognitive function. The method can be characterised by the structured identification of the weak points of the task required to perform and the efficient analysis process that the analysts have only to carry out with the necessary cognitive functions. The paper also presents the application of AGAPE-ET to 31 nuclear emergency tasks and its results

[en] The conceptual designs for SG, MCP, CEDM to be installed in the integral reactor SMART were developed. Three-dimensional CAD models for the major components were developed to visualize the design concepts. Once-through helical steam generator was conceptually designed for SMART. Canned motor pump was adopted in the conceptual design of MCP. Linear pulse motor type and ballscrew type CEDM, which have fine control capabilities were studied for adoption in SMART. In parallel with the structural design, the electro-magnetic design was performed for the sizing motors and electro-magnet. Prototypes for the CEDM and MCP sub-assemblies were developed and tested to verify the performance. The impeller design procedure and the computer program to analyze the dynamic characteristics of MCP rotor shaft were developed. The design concepts of SG, MCP, CEDM were also invetigated for the fabricability

[en] The conventional Human Reliability Analysis (HRA) methods such as THERP/ASEP, HCR and SLIM has been criticised for their deficiency in analysing cognitive errors which occurs during operator's decision making process. In order to supplement the limitation of the conventional methods, an advanced HRA method, what is called the 2^nd generation HRA method, including both qualitative analysis and quantitative assessment of cognitive errors has been being developed based on the state-of-the-art theory of cognitive systems engineering and error psychology. The method was developed on the basis of human decision-making model and the relation between the cognitive function and the performance influencing factors. The application of the proposed method to two emergency operation tasks is presented

[en] It is expected that motivation and basic technologies for the future R and D plans can be provided from the results of this study. This study has been done to develop fundamentals for radiation applications based on the existing radiation technology, and to establish technical basis for enhancing efficacy of radiation utilization by studying the simultaneous application of ionizing radiation with another factor. Application of radiation technology together with the existing technologies to enhance the physical, chemical, biological characteristics through structural changes of biomolecules will exert a favorable influence on the creation of de novo scientific and industrial values. A theoretical model for the combined action of ionizing radiation with another factor can make it possible to predict a prior the maximum value of synergistic interaction and the conditions for it. Furthermore, the results of this study give a clues for establishment of fundamental theories associated with positive efficacy of radiation applications

[en] The experiments on the fission products release behavior from spent fuel at high temperature assuming reactor accident conditions have been carried out at Oak Ridge Nation Laboratory of USA in HI/VI tests, CEA of France in HEVA/VERCOS tests, AEA of England and CRNL of Canada in HOX test. The VEGA program to study the fission product release behavior from LWR irradiated fuel was recently initiated at JAERI. The key parameter affecting the fission product(FP) release behavior is temperature. In addition, other parameters such as fuel oxidation, burnup, pre-transient conditions are found to affect the FP releases considerably in the earlier tests. The atmosphere conditions such as oxidizing atmosphere (steam or air) or reducing atmosphere (hydrogen) can cause significant change of FPs release and transport behavior due to chemical forms of the reactive FPs which is dependent on the oxidation potential. The effect of fuel burnup on the Kr-85 or Cs-137 release showed that the release rates of these radionuclides increased with the increase of burnup, meaning that release rates are dominated by the atomic diffusions in the grains and they are primarily a function of temperature. However, the data on FPs release behavior using higher burnups above 50,000 MWD/MTU are not so many reported up to now. This report summarizes the test results of FPs release behavior in reactor accident conditions produced from other countries mentioned above. This review and analysis on earlier studies would be useful for predicting the release characteristics of FPs from domestic spent fuel. The release rates of fission gas or FPs from spent fuel at high temperature conditions during fabrication process of dry recycling fuel were also analyzed using many data obtained from earlier tests