Okuda, Yukihiko; Takito, Kiyotaka; Nishida, Akemi; Li, Y.
Mechanical Engineering Journal (Internet), 12 Pages, 2024/00
After the Great East Japan earthquake and the accident at the TEPCO's Fukushima Daiichi Nuclear Power Stations in March 2011, the regulation for nuclear power plants (NPPs) has been enhanced to take countermeasures against beyond-design-basis events. To improve the seismic safety of nuclear facilities against earthquakes that exceed the design input ground motion, the importance of seismic probabilistic risk assessment (PRA) has drawn much attention. It is essential to evaluate the realistic seismic response of the equipment and piping in NPPs for fragility assessment in seismic PRA. In particular, since piping systems have plant-specific complex route geometries, it is known that the arrangement and stiffness of piping support structures have a significant impact on seismic response characteristics of the entire piping system. To construct a realistic seismic response analysis method for excessive input ground motion exceeding the elastic response, it is desired to develop an elastic-plastic response analysis method that can estimate the realistic response of piping systems including pipe support structures. In this study, the applicability of the method is confirmed by the simulation analysis of the elasto-plastic response for the piping support structure loading test previously reported. Moreover, based on the good correlation between the ductility factor and the damage status obtained from the test results and simulation analysis results, it is shown that the ductility factor is effective as a damage evaluation index for piping support structures.
Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi
Mechanical Engineering Journal (Internet), 10(4), p.22-00387_1 - 22-00387_20, 2023/08
For nuclear power plants, probabilistic risk assessment (PRA) should be performed not only against earthquake and tsunami, which are critical events especially in Japan, but also other external hazards such as strong wind. The aim of the present study is to develop a practical PRA methodology for sodium-cooled fast reactors (SFRs) against strong wind, paying attention to the final heat sink, ambient air, that removes decay heat under accident conditions. First, this study used Gumbel distributions to estimate hazard curves of the strong wind based on weather data recorded in Japan. Second, it identified important structures, systems and components (SSCs) for decay heat removal, and developed an event tree that results in core damage, focusing on the impacts of missiles (e.g., steel pipes) caused by strong wind. It also identified missiles that can reach SSCs at elevated places, and calculated the fragility of the SSCs against the missiles as a product of two probabilities. One is a probability of the missiles that would enter an inlet or outlet of the decay heat removal system, and another is a probability of failure caused by missile impacts. Finally, it quantified conditional decay heat removal failure probabilities by introducing the fragilities into the event tree. The core damage frequency (CDF) was estimated at about 5x10-10/y. The dominant sequence is that strong wind causes offsite power loss and missiles, the missiles penetrate the diesel fuel tank, cause a fire, and the fire increases air temperature around the reactor building where air cooler inlets of decay heat removal systems are installed, leads to loss of power for the diesel generator for forced circulation cooling, resulting in loss of decay heat removal. Through the above, this study has developed the practical PRA methodology for SFRs against strong wind.
Yamashita, Susumu; Uesawa, Shinichiro; Ono, Ayako; Yoshida, Hiroyuki
Mechanical Engineering Journal (Internet), 10(4), p.22-00485_1 - 22-00485_25, 2023/08
A detailed evaluation for air cooling of fuel debris in actual reactors will be essential in fuel debris retrieval under dry conditions. To understand the heat transfer in and around fuel debris, which is assumed as a porous medium in the primary containment vessel (PCV) mechanistically, we newly applied the porous medium model to the multiphase and multicomponent computational fluid dynamics code named JUPITER (JAEA Utility Program for Interdisciplinary Thermal-hydraulics Engineering and Research). We applied the Darcy-Brinkman model as for the porous medium model. This model has high compatibility with JUPITER because it can treat both a pure fluid and a porous medium phase simultaneously in the same manner as the one-fluid model in multiphase flow simulation. We addressed the case of natural convection with a high-velocity flow standing out nonlinear effects by implementing the Forchheimer model, including the term of the square of the velocity as a nonlinear effect to the momentum transport equation of JUPITER. We performed some simple verification and validation simulations, such as the natural convection simulation in a square cavity and the natural convective heat transfer experiment with the porous medium, to confirm the validity of the implemented model. We confirmed that the result of JUPITER agreed well with these simulations and experiments. In addition, as an application of the updated JUPITER, we performed the preliminary simulation of air cooling of fuel debris in the condition of the Fukushima Daiichi Nuclear Power Station unit 2 including the actual core materials. As a result, JUPITER calculated the temperature and velocity field stably in and around the fuel debris inside the PCV. Therefore, JUPITER has the potential to estimate the detailed and accurate thermal-hydraulics behaviors of fuel debris.
Choi, B.; Nishida, Akemi; Shiomi, Tadahiko; Kawata, Manabu; Li, Y.; Ota, Akira*; Sonobe, Hideaki*; Ino, Susumu*; Ugata, Takeshi*
Mechanical Engineering Journal (Internet), 10(4), p.23-00026_1 - 23-00026_11, 2023/08
In the seismic evaluation of nuclear facility buildings, basemat uplift-the phenomenon during which the bottom of the basemat of a building partially rises from the ground owing to overturning moments during earthquakes-is a very important aspect because it affects not only structural strength and integrity, but also the response of equipment installed in the building. However, there are not enough analytical studies on the behavior of buildings with a low ground contact ratio due to basemat uplift during earthquakes. In this study, we conducted a simulation using a three-dimensional finite element model from past experiments on basemat uplift; further, we confirmed the validity of this approach. In order to confirm the difference in the analytical results depending on the analysis code, the simulation was performed under the same analytical conditions using the three analysis codes, which are E-FrontISTR, FINAS/STAR and TDAPIII, and the obtained analysis results were compared. Accordingly, we investigated the influence of the difference in adhesion on the structural response at low ground contact ratio. In addition, we confirmed the effects of significant analysis parameters on the structural response via sensitivity analysis. In this paper, we report the analytical results and insights obtained from these investigations.
Lu, K.; Takamizawa, Hisashi; Li, Y.; Masaki, Koichi*; Takagoshi, Daiki*; Nagai, Masaki*; Nannichi, Takashi*; Murakami, Kenta*; Kanto, Yasuhiro*; Yashirodai, Kenji*; et al.
Mechanical Engineering Journal (Internet), 10(4), p.22-00484_1 - 22-00484_13, 2023/08
Yamano, Hidemasa; Futagami, Satoshi; Ando, Masanori
Mechanical Engineering Journal (Internet), 10(4), p.23-00043_1 - 23-00043_12, 2023/08
This study has conducted a detailed structural analysis of a reactor vessel (RV) in a loop-type sodium-cooled fast reactor using a general-purpose finite element analysis code, FINAS/STAR, to understand its deformation behavior under extremely high temperature conditions and to identify the areas which should be focused to mitigate impacts of failure. The RV was heated from the normal operation condition to the sodium boiling temperature in the upper sodium plenum during 20 hours assuming depressurization. The analysis has revealed less significant stress and strain which were sufficiently lower than failure criteria. The upper body of RV was identified as the important area in terms of mitigation of structural failure. The RV was eventually deformed downward about 16 cm, resulting in no failure. This effect contributes to maintaining RV sodium level in a long term, thereby enhancing the RV resilience.
Okajima, Satoshi; Mori, Takero; Kikuchi, Norihiro; Tanaka, Masaaki; Miyazaki, Masashi
Mechanical Engineering Journal (Internet), 10(4), p.23-00042_1 - 23-00042_12, 2023/08
In this paper, we propose the simplified procedure to estimate failure probability of components subjected to thermal transient for the design optimization. Failure probability can be commonly used as an indicator of component integrity for various failure mechanisms. In order to reduce number of analyses required for one estimation, we have adopted the First Order Second Moment (FOSM) method as the estimation method of failure probability on the process of the optimization, and an orthogonal table in experiment design method is utilized to define conditions of the analyses for the evaluation of the input parameters for the FOSM method. The superposition of ramp responses is also utilized to evaluate the time history of thermal transient stress instead of finite element analysis. Through the demonstration study to optimize thickness of cylindrical vessel subjected to thermal transient derived from shutdown, we confirmed that the procedure can evaluate the failure probability depending on the cylinder thickness with practical calculation cost.
Kubo, Kotaro; Tanaka, Yoichi; Hakuta, Yuto*; Arake, Daisuke*; Uchiyama, Tomoaki*; Muramatsu, Ken
Mechanical Engineering Journal (Internet), 10(4), p.23-00051_1 - 23-00051_17, 2023/08
The significance of probabilistic risk assessments (PRAs) of nuclear power plants against external events was re-recognized after the Fukushima Daiichi Nuclear Power Plant accident. Regarding the seismic PRA, handling correlated failures of systems, components, and structures (SSCs) is very important because this type of failure negatively affects the redundancy of accident mitigation systems. The Japan Atomic Energy Research Institute initially developed a fault tree quantification methodology named the direct quantification of fault tree using Monte Carlo simulation (DQFM) to handle SSCs' correlated failures in detail and realistically. This methodology allows quantifying the top event occurrence probability by considering correlated uncertainties related to seismic responses and capacities with Monte Carlo sampling. The usefulness of DQFM has already been demonstrated. However, improving its computational efficiency would allow risk analysts to perform several analyses. Therefore, we applied quasi-Monte Carlo and importance sampling to the DQFM calculation of simplified seismic PRA and examined their effects. Specifically, the conditional core damage probability of a hypothetical pressurized water reactor was analyzed with some assumptions. Applying the quasi-Monte Carlo sampling accelerates the convergence of results at intermediate and high ground motion levels by an order of magnitude over Monte Carlo sampling. The application of importance sampling allows us to obtain a statistically significant result at a low ground motion level, which cannot be obtained through Monte Carlo and quasi-Monte Carlo sampling. These results indicate that these applications provide a notable acceleration of computation and raise the potential for the practical use of DQFM in risk-informed decision-making.
Yada, Hiroki; Takaya, Shigeru; Morohoshi, Kyoichi*; Yokoi, Shinobu*; Miyagawa, Takayuki*
Mechanical Engineering Journal (Internet), 10(4), p.23-00044_1 - 23-00044_13, 2023/08
To develop rationalized maintenance plans for nuclear power plants, the characteristics of each plant must be considered. For sodium-cooled fast reactor (SFR) plants, constraints on inspections exist due to the specialty that equipment retaining sodium must be handled, which is one of the important points that must be considered in maintenance rationalization. In this study, we propose a maintenance optimization scheme, which is a design support tool, using risk information to develop a maintenance strategy based on the system based code (SBC) concept. The SBC concept intends to provide a theoretical procedure to optimize the reliability of structure, system and components (SSCs) by administrating every related engineering requirements throughout the life of the SSCs from design to decommissioning. ASME Boiler and Pressure Vessel Code, Code Case, N-875 was developed based on the SBC concept. The purpose of this study is to establish detailed procedures for the maintenance optimization scheme based on the procedure in Code Case N-875. Furthermore, a quantitative trial evaluation of the core support structure of the next SFR under development in Japan is also performed using the maintenance optimization scheme.
Okuda, Takahiro; Takahashi, Hideki*; Watakabe, Tomoyoshi
Mechanical Engineering Journal (Internet), 10(4), p.23-00075_1 - 23-00075_9, 2023/08
In recent years, to make the seismic design more rational for the piping systems in nuclear power plants, it has been expected to develop a design method considering plastic deformation and the accompanying energy dissipation of the piping itself. In this study, an extensive series of seismic response analyses was conducted to investigate the degree of influence of the plastic deformation of the pipe support structures on the seismic response of the entire piping system. The analyses include; plasticity is considered for (1) none, (2) the piping only, (3) the support structure only, and (4) both the piping and the support structure.
Okuda, Yukihiko; Kang, Z.; Nishida, Akemi; Tsubota, Haruji; Li, Y.
Mechanical Engineering Journal (Internet), 10(3), p.22-00370_1 - 22-00370_12, 2023/06
Many experimental studies have been reported on the impact resistance of reinforced concrete (RC) structures. However, most formulas were derived from impact tests based on normal impact to target structures using rigid projectiles that do not deform during impact. Therefore, this study develops a local damage evaluation method considering the rigidity of projectiles and oblique impacts that should be considered in realistic projectile impact phenomena. Specifically, we focused on scabbing, defined as the peeling off the back face of the target opposite the impact face, and conducted impact tests on RC panels to clarify the scabbing limit by changing the impact velocity in an oblique impact. The effects of the projectile rigidity and oblique impact on the scabbing limit were investigated based on the test results. This work presents the test conditions, equipment, results, and the scabbing limit on the local damage to RC panels subjected to oblique impacts.
Aihara, Jun; Kuroda, Masatoshi*; Tachibana, Yukio
Mechanical Engineering Journal (Internet), 9(4), p.21-00424_1 - 21-00424_13, 2022/08
It is important to improve oxidation resistance of fuel for huge oxygen ingress into core to improve safety of high temperature gas-cooled reactors (HTGRs), because almost volume of cores of HTGRs consist of graphite. In this study, simulated oxidation resistant fuel elements, of which matrix is mixture of SiC and graphite, has been fabricated by hot press method. In order to maintain structural integrity of fuel element under accident conditions, high-strength fuel elements should be developed. In order to identify optimal hot press conditions for preparing high-strength fuel elements, effect of hot press conditions on mechanical strength properties of fuel elements should be evaluated quantitatively. In the present study, response surface model, which represents relationship between hot press conditions and mechanical strength properties, has been constructed by introducing statistical design of experiments (DOE) approaches, and optimal hot press conditions were estimated by model.
Hamase, Erina; Miyake, Yasuhiro*; Imai, Yasutomo*; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki
Mechanical Engineering Journal (Internet), 9(4), p.21-00438_1 - 21-00438_15, 2022/08
To enhance the safety of sodium-cooled fast reactors, a dipped-type direct heat exchanger (D-DHX) has been investigated in a natural circulation decay heat removal system. During the D-DHX operation, the core-plenum interactions occurs and the thermal-hydraulics in the reactor vessel (RV) is complicated, the establishment of thermal-hydraulic analysis model in the RV for computational fluid dynamics code (RV-CFD) is required to simulate the thermal stratification in the upper plenum and thermal-hydraulics in the core. In this study, in terms of using RV-CFD for design study, the subchannel CFD model with low computational cost was adopted to the core of RV-CFD and the numerical simulation was carried out in comparison with the measured data in the sodium test facility named PLANDTL-1. As the result, the calculated sodium temperature in the core had good agreement with the experimental result and the applicability of the RV-CFD for the core-plenum interactions was confirmed.
Takai, Toshihide; Furukawa, Tomohiro; Watanabe, Shigeki*; Ishioka, Noriko*
Mechanical Engineering Journal (Internet), 9(4), p.21-00397_1 - 21-00397_11, 2022/08
For the mass production of astatine-211, a promising radiopharmaceutical for cancer treatment, the National Institute for Quantum and Radiological Science and Technology has proposed the innovative "Liquid Bismuth Target System." The target window in this system must be made from a material that resists the highly corrosive liquid bismuth environment. To meet this requirement, a promising target window material was selected in corrosion experiments performed in stagnant liquid bismuth. Based on knowledge of corrosion in liquid lead-bismuth eutectic gained during the development of fast reactors and accelerator-driven subcritical systems, experiments were carried out under saturated dissolved oxygen and low oxygen conditions, and the corrosion behaviors of the specimens were evaluated. The FeCrAl-alloy exhibited the most excellent corrosion resistance, followed by FeCrMo-alloy. Both materials are suitable candidates for the target window.
Sato, Ikken; Yamaji, Akifumi*; Li, X.*; Madokoro, Hiroshi
Mechanical Engineering Journal (Internet), 9(2), p.21-00436_1 - 21-00436_17, 2022/04
Shiro, Ayumi*; Okada, Tatsuya*; Shobu, Takahisa
Mechanical Engineering Journal (Internet), 8(6), p.21-00106_1 - 21-00106_8, 2021/12
The objective of the present study was to carry out observations of deformation and annealing processes of aluminum single crystals using a synchrotron radiation X-rays at SPring-8. Al single-crystalline samples having a 111 orientation parallel to the longitudinal direction were grown by a Bridgman method. The samples were deformed in tension to a nominal strain of 0.08 at room temperature using an in-line tensioning apparatus. Post-deformation annealing at 480 C was subsequently carried out in the same apparatus. A two-dimensional detector was used to detect multiple diffracted beams from the sample during the deformation and annealing processes. The volume irradiated by the X-ray beam was found to be composed of three regions having a small orientation difference, which was attributable to sub-grained microstructures of the sample. Detailed analyses of a diffraction spot intensity showed that the sub-grained microstructures were surpassed by dislocated microstructures with the increase in the tensile strain. During the post-deformation annealing, diffraction spots from a recrystallized grain first appeared at 180 s after the temperature reached 480 C. Coexistence of diffraction spots from the deformation matrix and recrystallized grain lasted only for about 22 s in the irradiated volume. The migration rate of the boundary between the deformation matrix and recrystallized grain was estimated to be of the order of several micrometers/s.
Aizawa, Kosuke; Hiyama, Tomoyuki; Nishimura, Masahiro; Kurihara, Akikazu; Ishida, Katsuji*
Mechanical Engineering Journal (Internet), 8(4), p.20-00547_1 - 20-00547_11, 2021/08
A sodium-cooled fast reactor has been designed to attain a high burn-up core in commercialized fast reactor cycle systems. The sodium-cooled fast reactor adopts a wire spacer between fuel pins. The wire spacer performs functions of securing the coolant channel and the mixing between subchannels. In high burn-up fuel subassemblies, the fuel pin deformation due to swelling and thermal bowing may decrease the local flow velocity in the subassembly and influence the heat removal capability. Therefore, understanding the flow field in a wire-wrapped pin bundle is important. This study performed particle image velocimetry (PIV) measurements using a wire-wrapped three-pin bundle water model to grasp the flow field in the subchannel under conditions, including the laminar to turbulent regions. In the region away from the wrapping wire, the maximum flow velocity was increased by decreasing the Re number. Accordingly, the PIV measurements using the three-pin bundle geometry without the wrapping wire were also conducted to understand the effect of the wrapping wires on the flow field in the subchannel. The results confirmed that the mixing due to the wrapping wire occurred, even in the laminar condition. These experimental results are useful not only for understanding the pin bundle thermal hydraulics, but also for the code validation.
Mechanical Engineering Journal (Internet), 8(4), p.21-00080_1 - 21-00080_15, 2021/08
It was reported that the long distance travel of temperature distribution causes a new type of thermal ratcheting, even in the absence of primary stress. When the distance of temperature travel is moderate, the accumulation of the plastic strain due to this mechanism is finally saturated. We have found the strong relationship between hoop-membrane distributions of accumulated plastic strain and residual stress in this saturated case. Focusing on this relationship, we have aimed to predict the saturated distribution of the plastic strain based on the residual stress distribution that is required for the elastic shakedown behavior. In this paper, based on classical shell theory, we formulated the plastic strain distribution that brings uniform hoop-membrane stress in the given region. The formulated strain distribution was validated by the comparison with the accumulated plastic strain distribution obtained by finite element analyses using an elastic-perfectly plastic material.
Kikuchi, Shin; Nakamura, Kinya*; Yamano, Hidemasa
Mechanical Engineering Journal (Internet), 8(4), p.20-00542_1 - 20-00542_13, 2021/08
In a postulated severe accidental condition of sodium-cooled fast reactor (SFR), eutectic melting between boron carbide (BC) and stainless steel (SS) may take place. Thus, kinetic behavior of BC-SS eutectic melting is one of the important phenomena to be considered when evaluating the core disruptive accidents in SFR. In this study, for the first step to obtain the fundamental information on kinetic feature of BC-SS eutectic melting, the thermal analysis using the pellet type samples of BC and Type 316L SS as different experimental technique was performed. The differential thermal analysis endothermic peaks for the BC-SS eutectic melting appeared from 1483K to 1534K and systematically shifted to higher temperatures when increasing heating rate. Based on this kinetic feature, apparent activation energy and pre-exponential factor for the BC-SS eutectic melting were determined by Kissinger method. It was found that the kinetic parameters obtained by thermal analysis were comparable to the literature values.
Takai, Toshihide; Furukawa, Tomohiro; Yamano, Hidemasa
Mechanical Engineering Journal (Internet), 8(4), p.20-00540_1 - 20-00540_11, 2021/08
In a core disruptive accident scenario, boron carbide, which is used as a control rod material, may melt below the melting temperature of stainless steel owing to the eutectic reaction with them. The eutectic mixture produced is assumed to extensively relocate in the degraded core, and this behavior plays an important role in significantly reducing the neutronic reactivity. However, these behaviors have never been simulated in previous severe accident analysis. To contribute to the improvement of the core disruptive accident analysis code, the thermophysical properties of the eutectic mixture in the solid state were measured, and regression equations that show the temperature (and boron carbide concentration) dependence are created.