Masaki Morishita

Development of Three-Dimensional Seismic Isolation Technology for Next Generation Nuclear Power Plant Applications

Mitigation of earthquake loads by seismic isolation technology is very promising for enhanced safety and economy of the next generation nuclear reactors, through rationalized and simplified design of structures, systems and components. The horizontal base isolation with laminated rubber bearings is a proven technology and its application has been widely spread including nuclear facilities. On the other hand, significantly increased benefit of mitigated seismic loads is expected with three-dimensional (abbreviated 3D) seismic isolation, since the earthquake loads are inherently three-dimensional and the vertical component of the earthquake load sometimes plays an important role in the structural design of reactor components. From these points of view, a research project has been undertaken for the development of 3D seismic isolation technology, under the sponsorship of the Ministry of Economy, Trade and Industry of the Japanese government. It was presented in a former conference that two types of 3D seismic isolation systems were applicable to the next generation nuclear power plants. One is 3D base isolation of a whole nuclear island, and the other is a vertical isolation system for main components with horizontal base isolation system. Among a number of proposed concepts, three were promising ideas for the 3D base isolation system (or device), i.e., “hydraulic 3D base isolation system”, “independent cable reinforced rolling-seal air spring”, and “rolling seal type air spring”. Then the last idea, i.e., “rolling seal type air spring”, was selected from above three ideas for further development. In this paper, current status of this R&D project are firstly shown. Next, the performance requirements for 3D isolation system and devices are shown. Then the developing targets for 3D isolation technology are shown. Furthermore, future plan of the project is provided.Copyright

Experimental Study on the Avoidance and Suppression Criteria for the Vortex-Induced Vibration of a Cantilever Cylinder

Experimental validation of the design guideline to prevent the failure of a thermometer well by vortex-induced vibration is presented, clarifying the effect of structure damping on displacement amplitudes of a cantilever cylinder. The available experimental data in piping are limited to those with small damping in water flow, because of the difficulty in increasing structure damping of the cantilever cylinders in experiments. In the present experiment, high-viscosity oil within cylinders is used to control their structure damping. Resulting values of reduced damping (C n ) are 0.49, 0.96, 1.23, 1.98, and 2.22. The tip displacements of the cylinder induced by vortex vibration were measured in the range of reduced velocity (V r ) from 0.7 to 5 (Reynolds number is 7.8×10 4 at V r = 1). Cylinders with reduced damping 0.49 and 0.96 showed vortex-induced vibration in the flow direction in the V r >1 region. However, in cases of reduced damping of 1.23, 1.98, and 2.22, the vibration was suppressed to less than 1 percent diameter It is confirmed that the criteria of V r 1.2 for the prevention of vortex-induced vibration is reasonably applicable to a cantilever cylinder in a water flow pipe.

Design Method of Vertical Component Isolation System

A structural concept of a vertical component isolation system for fast reactors, assuming a building adopting a horizontal base isolation system, has been studied. In this concept, a reactor vessel and major primary components are suspended from a large common deck supported by isolation devices consisting of large coned disk springs. A series of experiments using a simple model for the confirmation of the isolation effect, and a case study of vertical isolation device and plant layout are shown in this paper.© 2002 ASME

Elaboration of the System Based Code Concept — Activities in JSME and ASME: (4) Joint Efforts of JSME and ASME

This paper describes the ongoing activities at the Joint Task Group for System Based Code established in 2012 by the Japan Society of Mechanical Engineers (JSME) and the American Society of Mechanical Engineers (ASME) in the ASME Boiler and Pressure Vessel Code Committee. The Joint Task Group aims at developing alternative rules for ASME Boiler and Pressure Vessel Code Section XI Division 3, inservice inspection requirements for liquid metal reactors. The alternative rules will be developed based on the System Based Code concept which was originally proposed in Japan and is being elaborated both in JSME and ASME. The alternative rules are for sodium-cooled fast reactors where some of the components could have difficulties in conforming to the current requirements primarily due to accessibility. The alternative requirements would consist of a set of relieved requirements and a logic flow through which the applicability of them is judged. The logic flow considers both component structural integrity and the plant safety goals. The issuance of a Code Case is envisioned around 2016. Further efforts to integrate the process into a new framework being developed in Section XI would cover various types of reactors.Copyright

Development of the Guideline on Inelastic Analysis for Design

The guideline on inelastic analysis for design, one of the key items of Fast Reactor Design Standard (FDS), is being developed. The basic policies of this guideline are as follows: (a) to emphasis conservative analysis output rather than nominal value representing actual behavior, (b) to clarify the applicable area for assurance of conservative results. With such concepts, it would be possible that the guideline provides useful explanations on the manner of analysis and estimation in the form of concrete examples of design as well as general rules (somehow vague). As the first step of the guideline development, the following five issues to be solved were extracted: 1) applicable area, 2) selection of constitutive equation, 3) modeling method of the load history, 4) ratchet strain and creep fatigue damage evaluation methods by inelastic analysis and 5) example design problems to check users’ analysis quality and to complement the general rules. In parallel, inelastic analyses with the promising constitutive equations were applied by way of trial to obtain rough presumption on their effects on structural design of the components. As a result, all inelastic analyses provided smaller cumulative strains and equivalent strain ranges than the existing design method based on elastic analysis, suggesting advantage of introducing them into actual design.Copyright

Development of System Based Code for Structural Integrity of FBRS

This paper introduces a newly started study that aims to expand the basic concept of the system based code for structural integrity, which has been proposed by Asada et al [1,2], for application to fast breeder reactors (FBRs) [3]. The System Based Code for FBR (FBR System Based Code) offers more reasonable structural integrity assessment methods for relatively important components of FBR than current codes and standards do, to enable profound improvement both in reliability and lifetime cost-effectiveness. It will be realized by margin optimization through an integrated evaluation of all the technical fields that affect structural integrity at the stage of design. Those fields are the prerequisites of design, material, design analysis, fabrication and installation, preservice inspection, operation, inservice inspection, and repair and replacement. For margin optimization, three promising methods, failure probability assessment, application of quality assurance index, and the introduction of “systematized design factors” were proposed. The FBR System Based Code will consist of a control code and partial codes. The former takes care of margin optimization while the latter offers various options of technologies and engineering tools among which a designer can chose the most appropriate one according to various requirements to be fulfilled. Partial codes will be developed for each technical field that is dealt with in the code. Technologies that need to be developed for the development of the FBR system Based code were also clarified.Copyright

A Large Scale Ongoing R&D Project on Three-Dimensional Seismic Isolation for FBR in Japan

FBR is well known as a reactor that breeds the nuclear fuel. As Japan has little nuclear resources in its territory, the technologies those realize Fast Breeder Reactor (abbreviated FBR) have been developed for decades. The results of the development have been demonstrated through the construction and operation of the experimental fast reactor JOYO and the prototype FBR MONJU. In 1999, the R&D to realize the commercialized FBR has started as a national project. In the project, improving the economic competency of the commercialized FBR plant was set as one of the most important objectives as well as enhancing the safety.Copyright

Ultimate Strength of a Thin Wall Elbow for Sodium Cooled Fast Reactors Under Seismic Loads

With a purpose of identifying the failure mode and associating the ultimate strength of piping components against seismic integrity, many kinds of failure tests have been conducted for thick wall piping for light water reactors (LWRs). However, there are little failure test data on thin wall piping for sodium cooled fast reactors (SFRs). In this paper, a series of failure tests on thin wall elbows for SFRs is presented. Based on the tests, the failure mode of a thin wall piping component under seismic loads was identified to be fatigue. The safety margin included in the current design methodology was clarified quantitatively.

Introduction of a Research Activity on the Seismic Safety Evaluation of Nuclear Piping Systems Taking the Effect of Elastic-Plastic Behavior Into Account

According to investigations of several nuclear power plants (NPPs) hit by actual seismic events and a number of experimental researches on the failure behavior of piping systems under seismic loads, it is recognized that piping systems used in NPPs include a large seismic safety margin until boundary failure and the current code design allowable stresses are very conservative. Since the stress assessment based on the elastic analysis does not reflect actual response of piping systems including plastic region, rational procedures to estimate the elastic-plastic behavior of piping systems under a large seismic load are expected to be developed for piping seismic design applications.With the aim of establishing a procedure that takes into account the elastic-plastic behavior effect in the seismic safety estimation of nuclear piping systems, a research activity has been planned. Through the activity, the authors intend to establish two kinds of guidelines; 1) a guideline of a standard analysis procedure to evaluate elastic-plastic behavior of piping systems under extreme seismic loads with rational and conservative margins, and 2) a guideline that provide criteria for the seismic safety assessment of piping systems by the standard analysis to evaluate elastic-plastic behavior established by the above guideline. As the first step of making out the analysis guideline, benchmark analyses are conducted for a pipe element test and a piping system test.In this paper, the outline of the research activity and the preliminary results of benchmark analyses for a pipe element test are described.Copyright

Structural Materials and Code Development for Japanese Sodium-Cooled Fast Reactors

This paper describes the latest status on the development of elevated temperature materials and structural codes for Japanese sodium-cooled fast reactors (SFRs). Based on the extensive research and development activities in the last decades in Japan, two materials, 316FR and Modified 9Cr-1Mo steels were recently incorporated into the 2012 Edition of Fast Reactor Design and Construction Code of the Japan Society of Mechanical Engineers (JSME). Structural design methodologies are continuously being improved towards the next major revision planed in 2016 Edition where methodologies for a 60-year design of Japanese demonstration fast reactor will be provided. Codes and guidelines for fitness-for-service, leak-before-break evaluation and reliability assessment are concurrently being developed utilizing the System Based Code concept aiming at establishing an integrated code system that encompasses a life cycle of SFRs.Paper published with permission.Compilation Copyright