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[en] In this study, a deterministic/probabilistic fracture mechanics analysis program for reactor pressure vessel, PROFAS-RV, is developed. This program can evaluate failure probability of RPV using recent radiation embrittlement model of 10CFR50.61a and stress intensity factor calculation method of RCC-MRx code as well as the required basic functions of PFM program. Applications of some new radiation embrittlement model, material database, calculation method of stress intensity factors, and others which can improve fracture mechanics assessment of RPV are introduced. The purpose of this study is to develop a probabilistic fracture mechanics (PFM) analysis program for RPV considering above modification and application of newly developed models and calculation methods. In this paper, it deals with the development progress of the PFM analysis program for RPV, PROFAS-RV. The PROFAS-RV is being tested with other codes, and it is expected to revise and upgrade by reflecting the latest model and calculation method continuously. These efforts can minimize the uncertainty of the integrity evaluation for the reactor pressure vessel.
[en] The precipitation and sensitization characteristics in AISI436 weld metal were investigated in different chemical composition ranges of Ti and Nb content. We manufactured four welding wires made of 0-0.2 wt% of Ti and 0-1.0 wt% of Nb and did flux cored arc welding. After heat treatment at 900 °C for 20 hours, we made a Double Loop Electrochemical Potentiokinetic Reactivation (DL-EPR) test, Electron Backscattering Diffraction and SEM. The DL-EPR test revealed that as the amount of addition of Ti and Nb rose, the degree of sensitization fell. The microstructure became more refined, and Cr carbide formed at the grain boundary that had no addition of Ti and Nb. Furthermore, in the specimen with the addition of Ti, Nb, the Ti, Nb carbide and nitride were precipitated in the intergranular boundary, and the laves phase was precipitated at the grain boundary.
[en] Lugs, brackets, stiffners and other attachments may be welded, bolted and studded to the outside or inside of piping and the local stress arise because of the radial thermal expansion of the piping, the dilatation of the piping due to its internal pressure, the circumferential contraction of the pipe as a results of an axial tensile force (Poisson's effect), etc., being constrained by those. So the evaluation of the local stress for the piping constrained by those. So the evaluation of the local stress for the piping constrained by the attachments in accordance with the ASME Section III NB-3651.3 is required for the nuclear class 1 piping. In this report the local stress analysis procedure for the nuclear class 1 piping welded to the seal plate was established. SInce the stress analysis for the nuclear class 1 piping require the fatigue analysis procedure was established using the commercial code ANSYS 5.1 for the structure analysis in this report. The procedure which was developed in this report can be used very effectively for the design for the seal plate and the local stress analysis of the nuclear class 1 piping welded to the seal plate. 14 figs., 5 refs. (Author) .new
[en] The round robin project was proposed by the PFM Research Subcommittee of the Japan Welding Engineering Society to Asian Society for Integrity of Nuclear Components (ASINCO) members, which is designated in Korea as Phase 2 of A-Pro2. The objective of this phase 2 of RR analysis is to compare the scheme and results related to the assessment of structural integrity of RPV for the events important to safety in the design consideration but relatively low fracture probability. In this study, probabilistic fracture mechanics analysis was performed for the round robin cases using PROFAS-RV code. The effects of key parameters such as different transient, fluence level, Cu and Ni content, initial RT_N_D_T and RT_N_D_T shift model on the failure probability were systematically compared and reviewed. These efforts can minimize the uncertainty of the integrity evaluation for the reactor pressure vessel
[en] Reactor pressure vessel (RPV) in nuclear power plant is one of the most important components because it is not replaceable and the materials under irradiation. Recent efforts have been made to apply newly developed model and extend applicable range based on enhanced database and model along with structural integrity evaluation of RPV using advanced FE modeling technique by the number of researchers. In this paper, recent progress in research on structural integrity of RPV is analyzed with a focus on latest PVP conference papers. Issues on fracture toughness curve, nozzles in the beltline region, PTS evaluation and eXtended Finite Element Method (XFEM) are dealt with and discussed for the future work. In this paper, recent issues on RPV integrity researches, such as fracture toughness model, effect of the beltline region to the P-T limit curve and XFEM technique are reviewed
[en] A finite element model was developed for the thermal analysis of a stud-to-plate laser brazing joint, and the transient temperature fields were analysed by using a three-dimensional model. The finite element program ABAQUS, together with a few user subroutines, was employed to perform the numerical approximation. Temperature-dependent thermal properties, effect of latent heat, and the convection and radiative heat losses were considered. The brazing parts used were AISI 304 stainless steel stud and aluminium A1 5052 plate, and the brazing alloy 88 A1-12 Si was used as filler metal. A pseudo-TM01 mode of the cw CO2 laser beam was used as heat source, for which TM00 mode generated by beam oscillator was optically modulated using axicon lens. Re-location of the filler metal during the brazing process including its wetting and spreading was examined by using a high speed motion analyser, and the results were incorporated inn the FEM modelling for defining the solution domain and boundary conditions. The numerical results were obtained for typical process parameters, and were compared with experimental ones determined by using the infrared and thermocouple measurements. 11 figs., 30 refs. (Author)
[en] For design and safety analysis of the steam generator tube rupture (SGTR) under severe accident conditions, a creep rupture model for the steam generator tube material is needed, and it is essential to evaluate the mechanical properties in the as-manufactured material conditions. However, transverse tensile properties of steam generator tube cannot be obtained directly from the tensile test using ring-type tensile specimen due to its geometry limitation. In this paper, the circumferential tensile properties were derived using finite element analysis and optimization technique. The purpose of this study is to derive the tensile properties of the alloy 690 steam generator tube which is used as the basic data for evaluating the creep rupture properties of the steam generator tube under severe accident condition. The difference of the tensile properties according to the two directions will be investigated by microstructural analysis. And the gauge length position will be compared with the current 9 o’clock position and 12 o’clock position. In addition, the effects of bending of specimen during the test and initial gap between specimen and mandrels on the results will be analyzed in detail.
[en] During the Nuclear Regulatory Commission (NRC) audit process of Doosan HF Control HFC-6000 safety system 2009, cyber security assessment was a major audit process. The result of the assessment was favorably satisfied. As preventing digital I and C systems from being hijacked by malicious software a major goal for the NRC, audit process of actual digital I and C implementations such as the HFC-6000 safety system which provides already strong cyber security measures is mutually beneficial to both the NRC and the vendor: NRC can enhance their set of cyber security assessments and vendors such as Doosan HFC can also augment their cyber security measures. The NRC Safety Evaluation Report (SER) for the HFC-6000 system was released in April 2011 qualifying the system to be used as safety systems in US nuclear power plants. This paper provides the summary of the cyber security assessment of the complete software life cycle of HFC-6000 Safety System. Lessons learned in each life cycle phase are provided. In addition, alternate measures or recommendations for enhancing the cyber security in each life cycle phase are also described
[en] Arteriovenous malformation (AVM) of female pelvic organ is a rare disease of unknown cause. The authors report a case of pelvic AVM which was incidentally found during US examination of the patient with choriocarcinoma after chemotherapy. The real-time sonography revealed several cystic lesions around the uterus with adjacent dilated tortuous vessels. The color Doppler sonography depicted abundant blood flow mixed with red and blue colors within the cystic lesions and rapid turbulent systolic and diastolic flows. CT showed well-enhancing round vascular lesions with elongated vessels in the pelvis, and MRI depicted signal-void cystic lesions on both T1 and T2 weighted images with small portions of high intensity with the lesions on T2 weighted image. The angiography revealed pelvic AVM fed by tortuous uterine and vaginal arteries with a dilated draining vein
[en] With 2 mm thick CT scanning during the rapid infusion of contrast material(TICT), cerebral aneurysms arising from the circle of Willis and adjacent vessels can be directly visualized. Twenty five patients who had cerebral aneurysm confirmed by surgery were examined with TICT and digital subtraction angiography. The authors examined TICT prospectively to assess the detection rate of the cerebral aneurysms and to evaluation the clinical usefulness of TICT. The detection rates of aneurysms by TICT and digital subtraction angiography were 68% and 84%, respectively. TICT is a rapid, safe and reliable method in the evaluation of patients with suspected cerebral aneurysm, permitting direct visualization of the aneurysm