Kazuhiko Iigaki

Influence of Differences Between Seismic Safety Evaluation Methods for Equipment and Piping of a Nuclear Facility

The objective of this research was to investigate the influence of differences between methods for evaluating the seismic safety of the equipment and piping of a nuclear facility. For the input ground motion, one wave was chosen from among 200 waves of input ground motions of maximum acceleration of 700–1100 cm/s2 created for the Oarai District of the Ibaraki Prefecture. Seismic safety evaluations were performed using the conventional method, which relies on floor response spectrum data, and using the multi-input method. The differences between the two methods were summarized. The target equipment and piping system were cooling systems in a model plant. It was found that the response predicted by the multi-input method was approximately half of the response predicted by the conventional method. The third trial evaluation method using the floor response of a three-dimensional building model as input was also reported.Copyright

Structural Analysis for Assembly by Integrating Parts

Almost all industrial products are assembled from multiple parts, and this is true for all sizes of products. As an example, a nuclear facility is a large structure consisting of more than 10 million components. This paper discusses a method to analyze an assembly by gathering data on its component parts. Gathered data on component may identify ill conditioned meshes for connecting surfaces between components. These ill meshes are typified by nodal point disagreement in finite element discretization. A technique to resolve inconsistencies in data among the components is developed. By using this technique, structural analysis for an assembly can be carried out, and results can be obtained by the use of supercomputers, such as the K computer. Numerical results are discussed for components of the High Temperature Engineering Test Reactor of the Japan Atomic Energy Agency.© 2014 ASME

A Safety Evaluation of HTTR Core Components Against 2011 Tohoku Earthquake

On March 11th, 2011, the 2011 Tohoku Earthquake which is one of the largest earthquakes in japan occurred and the maximum acceleration in observed seismic wave in the HTTR exceeded the design value in a part of input seismic motions. Therefore, a visual inspection, a seismic analysis and a performance confirmation test of facilities were carried out in order to confirm the integrity of facility after the earthquake. The seismic analysis was carried out for the reactor core structures by using the response magnification factor method. As the results of the evaluation, the generated stress in the graphite blocks in the reactor core at the earthquake were well below the allowable values of safety criteria, and thus the structural integrity of the reactor core was confirmed. The integrity of reactor core was also supported by the visual inspections of facilities and the operation without reactor power in cold conditions of HTTR.Copyright

Preliminary Study on Thermal-Load Fluctuation Test Using HTTR

JAEA has designed and developed a commercial very high temperature reactor hydrogen cogeneration system named GTHTR300C. JAEA has investigated that a hydrogen production facility is treated as not a reactor grade facility but a general industrial plant grade facility. The reactor plant system should be designed absorbing the disturbance under AOO and becoming to steady state in the case that thermal-load of the facility is fluctuated or lost. Also, high-accuracy plant dynamics code should be developed to design the GTHTR300C.In order to demonstrate continuing the reactor operation in if the AOO is occurred in the hydrogen production facility and to obtain the validation data for high-accuracy plant dynamics code, the thermal-load fluctuation tests were planned using the HTTR in JAEA. In this study, it is concluded that the thermal-load fluctuation tests can be carried out without reactor scram and without modification of the HTTR facility and that the test data for validation of the code can be obtained sufficiently.Copyright

Improvement of Temperature Evaluation Model of Biological Shielding Concrete for HTTR Test Simulating LOFC With VCS Inactive

HTTR is a helium gas cooled graphite-moderated HTGR with the rated power 30 MWt and the maximum reactor outlet coolant temperature 950°C. The vessel cooling system (VCS), which is composed of thermal reflector plates, cooling panel composed of fins connected between adjacent water cooling tubes, removes decay heat from reactor core by heat transfer of thermal radiation, conduction and natural convection in case of loss of forced cooling (LOFC). The metallic supports are embedded in the biological shielding concrete to support the fins of VCS. To verify the inherent safety features of HTGR, the LOFC test is planned by using HTTR with the VCS inactive from an initial reactor power of 9 MWt under the condition of LOFC while the reactor shut-down system disabled. In this test, the temperature distribution in the biological shielding concrete is prospected locally higher around the support because of thermal conduction in the support. A 2-dimensional symmetrical model was improved to simulate the heat transfer to the concrete through the VCS support in addition to the heat transfer thermal radiation and natural convection. The model simulated the water cooling tubes setting horizontally at the same pitch with actual configuration. The numerical results were verified in comparison with the measured data acquired from the test, in which the RPV was heated up to around 110 °C without nuclear heating with the VCS inactive, to show that the temperature is locally high but kept sufficiently low around the support in the concrete due to sufficient thermal conductivity to the cold temperature region.Copyright

Impact Response Analysis of a Coaxial Double-Pipe Structure by Using Spectral Element Method

A coaxial double-pipe structure is to be used in the primary and auxiliary coolant system of a high-temperature gas-cooled reactor. In order to study the vibration characteristics of the coaxial double-pipe structure, hammering experiments were performed using specimens of the structure. Because the structural responses obtained in the experiments contained high-frequency components, impact response analysis was performed by using the spectral element method, which has high accuracy in the high-frequency region. A comparison between analysis results and experiment results showed good agreement between them. We also performed parametric studies on the damping properties of the specimens. The damping properties determined from the experiment results indicated that the inner and outer pipes had different damping properties.Copyright