Japan Electric Association Research Articles

Impact of Inelastic Seismic Design Using Elastic–Plastic Pipe Supports on Vibration Response of Piping Systems

Abstract This study presents the response mitigation effect of piping systems by inelastic seismic design based on elastic–plastic property of steel pipe supports. The inelastic seismic design to control vibration by absorbing energy using elastic–plastic properties of materials can be one of useful ideas. The design idea to use the elastic–plastic behavior of pipe supports is addressed in Technical Code for Seismic Design of Nuclear Power Plants (JEAC4601) published by the Japan Electric Association in Japan. Here, the component named an elastic–plastic pipe support is proposed as an energy-absorbing element. However, in order to put the inelastic seismic design using the elastic–plastic pipe supports into practical use, it is necessary to accumulate more findings related to the seismic response and the application range. This study aims to investigate the applicability of the inelastic seismic design taking the elastic–plastic pipe supports in the piping systems and to increase the basic findings. In this study, the seismic response analysis using three-dimensional piping system with an elastic–plastic pipe support was conducted. As a result, it was found that the elastic–plastic pipe support affected the seismic response largely. Additionally, the vibration characteristics, the response acceleration, and the load generated in the piping system were discussed relating to the plastic deformation and the plasticity rate of the elastic–plastic pipe support.

Guideline on Probabilistic Fracture Mechanics Analysis for Japanese Reactor Pressure Vessels

Abstract In Japan, to prevent nil-ductile fracture of reactor pressure vessels (RPVs) due to neutron irradiation embrittlement, deterministic fracture mechanics evaluation in accordance with the codes provided by the Japan Electric Association is performed for assessing the structural integrity of RPVs under pressurized thermal shock (PTS) events considering neutron irradiation embrittlement. In recent years, a structural integrity assessment methodology based on probabilistic fracture mechanics (PFM) has been introduced into the regulations in the United States and a few European countries. PFM is a rational methodology for evaluating the failure frequency of important pressure boundary components by considering the probabilistic distributions of various influence factors related to aged degradation due to the long-term operation. In Japan Atomic Energy Agency (JAEA), a PFM analysis code called PASCAL has been developed to evaluate the failure frequency of RPVs considering neutron irradiation embrittlement and PTS events. In addition, we have developed a guideline for structural integrity assessment of RPVs based on PFM to improve the applicability of PFM in Japan and enable persons who have knowledge on fracture mechanics to perform PFM analyses and evaluate through-wall cracking frequency (TWCF) of RPVs easily. The guideline consists of a main body, explanation, and several supplements. The technical basis for PFM analysis is provided, and the latest knowledge is included in the guideline. In this paper, an overview of the guideline and some typical analysis results obtained based on the guideline and the Japanese database related to PTS evaluation are presented.

Statistical arguments towards the development of an advanced embrittlement correlation method for reactor pressure vessel materials

ABSTRACT An embrittlement correlation method is one of the most important techniques used to ensure the integrity of pressure vessel steels in nuclear power plants. In Japan, the embrittlement correlation method is being addressed in accordance with the Japan Electric Association Code (JEAC 4201), which was developed using actual measured data on the irradiation embrittlement of pressure vessel steels. In the present study, to develop more reliable methodologies, statistical arguments were made concerning the embrittlement data. With regard to a set of residual data defined in JEAC 4201 as a collection of differences between the measured and calculated (ductile-to-brittle transition temperature shift) values, a statistical relationship between the population and samples was found, and then, the sampling errors in the mean values of the residuals were identified as key for establishing a more reliable correlation method. Using this relationship, it was noted that when predicting the amount of irradiation embrittlement for a particular plant using the JEAC 4201 correlation method, the deviation associated with sampling errors needed to be corrected. Based on this finding, a more appropriate interpretation was found for the so-called MC correction in JEAC 4201, and moreover, a new correction method was developed within the Bayesian estimation framework.

Numerical Study on Inelastic Seismic Design of Piping Systems Using Damping Effect Based on Elastic–Plastic Property of Pipe Supports

This study describes inelastic seismic design of piping systems considering the damping effect caused by elastic–plastic property of a pipe support which is called an elastic–plastic support. Though the elastic–plastic support is proposed as inelastic seismic design framework in the Japan Electric Association code for the seismic design of nuclear power plants (JEAC4601), the seismic responses of the various piping systems with the support are unclear. In this study, the damping coefficient of a piping system is focused on, and the relation between seismic response of the piping system and elastic–plastic behavior of the elastic–plastic support was investigated using nonlinear time history analysis and complex eigenvalue analysis. The analysis results showed that the maximum seismic response acceleration of the piping system decreased largely in the area surrounded by pipe elbows including the elastic–plastic support which allowed plastic deformation. The modal damping coefficient increased a maximum of about sevenfold. Furthermore, the amount of the initial stiffness of the elastic–plastic support made a difference in the increasing tendency of the modal damping coefficient. From the viewpoint of the support model in the inelastic seismic design, the reduction behavior for the seismic response of the piping system was little affected by the 10% variation of the secondary stiffness. These results demonstrated the elastic–plastic support is a useful inelastic seismic design of piping systems on the conditions where the design seismic load is exceeded extremely.

Applicability of Miniature Compact Tension Specimens for Fracture Toughness Evaluation of Highly Neutron Irradiated Reactor Pressure Vessel Steels

The applicability of miniature compact tension (Mini-C(T)) specimens to fracture toughness evaluation of neutron-irradiated reactor pressure vessel (RPV) steels was investigated. Three types of RPV steels neutron-irradiated to a high-fluence region were prepared and manufactured as Mini-C(T) specimens according to Japan Electric Association Code (JEAC) 4216-2015. Through careful selection of the test temperature by considering previously obtained mechanical properties data, valid fracture toughness, and reference temperature (To) was obtained with a relatively small number of specimens. Comparing the fracture toughness and To values determined using other larger specimens with those determined using the Mini-C(T) specimens, To values of both unirradiated and irradiated Mini-C(T) specimens were found to be the acceptable margin of error. The scatter of 1T-equivalent fracture toughness values of both unirradiated and irradiated materials obtained using Mini-C(T) specimens did not differ significantly from the values obtained using larger specimens. The correlation between the Charpy 41 J transition temperature (T41J) and the To values agreed very well with that of the data in the literature, regardless of specimen size and fracture toughness of the materials before irradiation. Based on these findings, it was concluded that Mini-C(T) specimens can be applied to fracture toughness evaluation of neutron-irradiated materials without significant specimen size dependence.

A REVIEW ON DEVELOPING INDUSTRIAL STANDARDS TO INTRODUCE DIGITAL COMPUTER APPLICATION FOR NUCLEAR I&C AND HMIT IN JAPAN

A comprehensive review on the technical standards about human factors (HF) design and software reliability maintenance for digital instrumentation and control (I&C) and human-machine interface technology (HMIT) in Japanese light water reactor nuclear power plants (NPPs) was given in this paper mainly by introducing the relevant activities at the Japan Electric Association to set up many industrial standards within the traditional framework of nuclear safety regulation in Japan. In Japan, the Fukushima Daiichi accident that occurred on March 11, 2011 has great impact on nuclear regulation and nuclear industries where concerns by the general public about safety have heightened significantly. However for the part of HF design and software reliability maintenance of digital I&C and HMIT for NPP, the author believes that the past practice of Japanese activities with the related technical standards can be successfully inherited in the future, by reinforcing the technical preparedness for the prevention and mitigation against any types of severe accident occurrence.

리튬이차전지 출력측정법의 원리 및 측정법간 비교 연구

리튬이차전지의 수요가 소형 휴대용 전자기기에서 중대형 전기자동차와 에너지저장장치로 변화함에 따라, 요구되는 전지 특성도 크게 달라지고 있다. 10년 이상의 장기 신뢰성, 팩 수준의 안전성, 가격 경쟁력 확보 등 기존 소형 전지보다 크게 강화된 요구 조건뿐만 아니라, 고출력 특성도 함께 필요로 한다. 그런데, 출력을 측정하기 위해서는 전압과 전류가 동시에 측정 및 제어되어야 하기 때문에, 기존의 단위셀 용량 측정보다 훨씬 어렵다. 또한, 측정 방법의 선택뿐만 아니라 출력유지시간, 충전상태(State-of-Charge, SOC), Cut-off 조건, 측정온도 등의 측정 조건에 따라서도 그 측정값이 크게 달라지는 경향이 있다. 본 논문에서는 HPPC (Hybrid Pulse Power Characterization)법, J-pulse(Japan Electric Vehicle Association Standards, JEVS D 713)법, 정출력 측정법(Constant Power Measurement)의 측정 원리 및 실제 실험에 어떻게 적용해야 하는지 설명하고, HEV(Hybrid Electric Vehicle)용 단위셀을 이용해 출력법간 측정 결과를 비교 분석한다. As the market of lithium secondary batteries moves from mobile IT devices to large-format electric vehicles or energy storage systems, the strengthened battery specifications such as long-term reliability longer than 10 years, pack-level safety and tough competitive price have been required. Moreover, even though high power properties should also be achieved for hybrid electric vehicles, it is not easy to measure accurate power values at various conditions. Because it is difficult to choose a proper measurement method and its experimental condition is more complex comparing to capacity measurement. In addition, the power values are very sensitive to power duration time, state-of-charge (SOC) of cells, cut-off voltages, and temperatures, whereas capacity values are not. In this paper, we introduce three kinds of power measurement methods, hybrid pulse power characterization (HPPC) suggested by US FreedomCar, so-called J-pulse by Japan electric vehicle association standards (JEVS) and constant power measurement, respectively. Moreover, with pouch-type unit cells for HEV, experimental power data are discussed in order to compare each power measurement.

Industry Practice for the Neutron Irradiation Embrittlement of Reactor Pressure Vessels in Japan

Neutron irradiation embrittlement of reactor pressure vessel steels (RPVs) is one of the important material aging issues. In Japan, almost 40 years have past since the first plant started its commercial operation, and several plants are expected to become beyond 40 years old in the near future. Thus, the safe operation, based on the appropriate recognition of the neutron irradiation embrittlement, is inevitable to ensure the structural integrity of RPVs. The amount of the neutron irradiation embrittlement of RPV steels has been monitored and predicted by the complemental use of the surveillance program and embrittlement correlation method. Recent surveillance data suggest some discrepancies between the measurements and predictions of the embrittlement in some old boiling water reactor (BWR) RPV steels with high impurity content. Some discrepancies of pressurized water reactor (PWR) RPV surveillance data from the predictions have also been recognized in the embrittlement trend. Although such discrepancies are basically within a scatter band, the increasing necessity of the improvement of the predictive capability of the embrittlement correlation method has been emphasized to be prepared for the future long term operation. Regarding the surveillance program, on the other hand, only one original surveillance capsule, except for the reloaded capsules containing Charpy broken halves, is available in some BWR plants. This situation strongly pushed establishing a new code for a new surveillance program, where the use of the reloading and reconstitution of the tested specimens is specified. The Japan Electric Association Code, JEAC 4201-2007 “Method of Surveillance Tests for Structural Materials of Nuclear Reactors,” was revised in December 2007, in order to address these issues. A new mechanism-guided embrittlement correlation method was adopted. The surveillance program was modified for the long term operation of nuclear plants by introducing the “long term surveillance program,” which is to be applied for the operation beyond 40 years. The use of the reloading, reirradiation, and reconstitution of the tested Charpy/fracture toughness specimens is also specified in the new revision. This paper reports the application and practice of the JEAC4201-2007 in terms of the prediction of embrittlement and the use of tested surveillance specimens in Japan.

Philosophy and practice of the aseismic design of nuclear power plants — Summary of the guidelines in Japan

Aseismic design is considered to be one of the most important factors for the safety of the nuclear power plants built in zones of high seismicity such as Japan. All structures, equipment and piping are classified in accordance with the importance of their radioactive safety to the plant, and the dynamic analysis and/or factored seismic coefficient analysis are applied accordingly. Site and ground conditions, as well as seismicity, should be studied thoroughly in order to estimate the intensities of the design earthquake and the safety margin check earthquake. Dynamic analyses of buildings and structures are performed using the multi-lumped-mass-system supported by soil springs with time history analysis conceptions. This idea is also applied to the design of equipment and piping by coupled system to the major structure or by the floor response spectra criteria. Tolerances are applied to damping factors although some experiments show more realistic results. Allowable stresses of ferrous metals for equipment and piping during earthquakes are more scientifically precise. This report summarizes a guideline for aseismic design of nuclear power plants. The guideline was prepared by the Japan Electric Association in April, 1970, after three years laborious work. In sect. 1, the philosophy and criteria are described. All components of a plant should be classified into three classes in accordance with their contributions to reactor safeties. Design to earthquake loadings should be based on “design basis earthquake” which is decided in consideration of local seismicity. In sect 2, site selection and review for ground are described in the sense of seismic aspects. In sect 3, deciding the earthquake motion for design is discussed. In Japan, semi-statistical approaches are used in normal practice. In sect. 4, design philosophy and practice of building structures and containment vessels are described. They are designed under statical seismic forces, and the design of the class “A” structures should be checked by a dynamic response technique. In sect. 5, design philosophy and practice of piping, vesels and equipment are described. Those which belong to class “A” items should be designed in a dynamic sense. Several programs for dynamic analyses of these items are prepared. Allowable stress under earthquake conditions is discussed in relation to other codes, for example, ASME Section III. The greater part of the philosophy and design criteria have been adopted to all nuclear power plants which have been and are currently being built in Japan.