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Hideo Nakamura
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Proceedings Papers
Ikuo Kinoshita, Toshihide Torige, Michio Murase, Yoshitaka Yoshida, Takeshi Takeda, Akira Satou, Hideo Nakamura
Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A022, July 7–11, 2014
Paper No: ICONE22-30559
Abstract
The application of the Best Estimate Plus Uncertainty (BEPU) method is made to analysis of the “Intentional depressurization of steam generator secondary side” which is an accident management procedure in a small-break loss-of-coolant accident (SBLOCA) with high pressure injection (HPI) system failure. RELAP5/MOD3.2 is used as the analysis code. By applying the BEPU method, the uncertainties of the analysis results can be estimated quantitatively. However, the accuracy of the analysis results depends primarily on the base case result predicted by the best estimate code. In this study, in order to investigate the appropriate base case model, simulation analyses using the RELAP5/MOD3.2 were carried out for the ROSA Large Scale Test Facility (ROSA/LSTF) secondary-side depressurization tests. It was found that the code predicted well the major event progressions such as pressure responses, core liquid level responses, and rod surface temperatures, as well as important phenomena such as formation and clearing of loop seals, accumulation of water from condensation, and countercurrent flow limitation (CCFL) at the inlet of the U-tubes, which are characteristic features of this accident scenario.
Proceedings Papers
Proc. ASME. ICONE21, Volume 4: Thermal Hydraulics, V004T09A116, July 29–August 2, 2013
Paper No: ICONE21-16808
Abstract
A 3D steam flow within simulated fuel bundle of Large Scale Test Facility (LSTF), a PWR system simulator, has been investigated by Computational Fluid Dynamics (CFD) analysis with Ansys Fluent code to clarify influences of the steam flow on Core Exit Temperature (CET) response. A LSTF SBLOCA experiment with 1.5% hot leg break as the OECD/NEA ROSA-2 Project Test 3 was simulated by the CFD code to clarify relation between CET and fuel rod surface temperature. A portion of the LSTF core above the mixture level up to around CET sensors was modeled by taking into account high, medium and low heat-zone heater rod bundle, including internal structures such as end-box and upper core plate (UCP). Simulation of steady-state condition at a certain time when mixture level lowered to a certain position at around half of the core height (post-5) was carried out by considering relevant boundary conditions which were developed based on the LSTF Test 3 results. The calculation results revealed that inner structures of the core such as core spacer, end box and UCP indeed affect the CET due to heat transfer from hot steam to these cool structures. 3D flow mixing may also contribute to the final CET values and the delayed increase in the CET relative to the Peak Cladding Temperature (PCT) in the core.
Proceedings Papers
Proc. ASME. ICONE21, Volume 4: Thermal Hydraulics, V004T09A117, July 29–August 2, 2013
Paper No: ICONE21-16811
Abstract
RELAP5 code post-test analysis was performed on one of abnormal transient tests conducted with the ROSA/LSTF simulating a PWR station blackout (SBO) transient with the TMLB’ scenario in 1995. The RELAP5 code predicted well the overall trend of the major phenomena observed in the LSTF test, and indicated remaining problems in the predictions of reverse flow U-tubes in steam generator (SG) during long-term single-phase liquid natural circulation. Sensitivity analyses were done further to clarify effectiveness of depressurization of and coolant injection into SG secondary-side as accident management measures to maintain core cooling, based on the LSTF post-test analysis. SG secondary-side depressurization was initiated by fully opening the safety valve in one of two SGs with the incipience of core uncovery. Coolant injection was done into the secondary-side of the same SG at low pressures considering availability of fire engines. The SG depressurization with the coolant injection was found to well contribute to maintain core cooling by the actuation of accumulator system during a PWR SBO (TMLB’) transient.
Proceedings Papers
Proc. ASME. ICONE20-POWER2012, Volume 3: Thermal-Hydraulics; Turbines, Generators, and Auxiliaries, 343-353, July 30–August 3, 2012
Paper No: ICONE20-POWER2012-54427
Abstract
Separate-effect experiment simulating steam condensation on emergency core cooling system (ECCS) water in PWR cold legs during reflood phase of large-break loss-of-coolant accident (LBLOCA) was conducted in OECD/NEA ROSA Project using the Large Scale Test Facility (LSTF). A test section was furnished in the downstream of the LSTF break unit horizontally attached to the cold leg. The boundary test conditions were defined based on PWR LBLOCA analysis by RELAP5/MOD3.2.1.2 code considering typical Japanese safety analysis conditions. Significant condensation of steam appeared in a short distance from the simulated ECCS injection point, and the steam temperature in the test section decreased immediately after the initiation of the ECCS water injection. Fluid temperature distribution at 50 mm downstream from the ECCS injection point was significantly non-uniform, but became almost uniform in less than 350 mm. Total steam condensation rate estimated from the difference between steam flow rates at the test section inlet and outlet was in proportion to the simulated ECCS water mass flux until the complete condensation of steam. Inlet steam was completely condensed if inlet steam mass flux is less than 195 kg/(m 2 s) when the simulated ECCS water mass flux is 148 kg/(m 2 s); equivalent to full high-pressure injection with single-failure low-pressure injection conditions. Clear images of high-speed video camera were obtained on droplet behaviors through the viewer at 200 mm downstream from the ECCS injection point, especially for annular mist flow. The number of flowing droplets decreased with increasing distance from the ECCS injection point.
Proceedings Papers
Proc. ASME. ICONE20-POWER2012, Volume 5: Fusion Engineering; Student Paper Competition; Design Basis and Beyond Design Basis Events; Simple and Combined Cycles, 473-479, July 30–August 3, 2012
Paper No: ICONE20-POWER2012-54918
Abstract
As for the turbulent two-phase flow in the non-circular duct, the flow could show an anisotropic turbulence feature in liquid phase. In this study, the air-water bubbly turbulent upward flow experiment in the large square duct with the inside cross-section of 136mm×136mm has been conducted. Since the bubble size is very important for air-water bubbly flows, the bubble generating method was improved to get more uniform bubble size. After confirming the flow symmetry in the measuring cross-section, the distributions of local void fraction, bubble frequency and primary liquid velocity were measured by a hot-film anemometry, and the bubble behaviors were also investigated by using the high-speed video camera. The results show that the bubbles tend to accumulate to the wall region, where the liquid primary velocity shows the maximum especially at the corner.
Proceedings Papers
Proc. ASME. ICONE18, 18th International Conference on Nuclear Engineering: Volume 4, Parts A and B, 597-604, May 17–21, 2010
Paper No: ICONE18-29630
Abstract
The flow-induced void fraction transition phenomenon was observed in an upward air-water two-phase flow in a vertical pipe with inner diameter D = 200 mm and height z = 25 m. As the two-phase flow develops in a vertical pipe, the void fraction increases firstly in the flow direction in bubbly flow, then decreases in the flow direction, finally increase again. The flow-induced void fraction transition shows an N -shaped changing manner. The present experimental investigation revealed that this phenomenon was attributed to the formation and the growth of local dominant large bubbles in the flow. According to the bubble sizes and behaviors observed in the experiment, the flow regimes were classified into bubbly, churn and slug flows in a vertical large-diameter pipe. The drift velocities in the three flow regimes were measured in this paper. New constitutive equation for drift velocities in bubbly, churn and slug flows was proposed and confirmed in this study. The flow-induced void fraction transition in N -shaped manner can be predicted by using the drift flux model with the newly developed constitutive equations.
eBook Chapter
Series: ASME Press Select Proceedings
Publisher: ASME Press
Published: 2011
ISBN: 9780791859902
Abstract
Every enterprise management shall investigate return on investment whenever they need to implement protection measures for their machine and manufacturing system in order to reduce/ eliminate occurrence of industrial accidents. In general, enterprise management treat expenses related protection measures as additional cost, but these expenses shall be treated as investment to prevent from industrial accidents and to reduce/ eliminate additional cost by industrial accidents. Thus, it is important to clarify methods to measure return on investment. H. W. Heinrich announced analysis of probability of Industrial Accident Occurrence related with Return on Investment in 1926. Then, Simmons announced analysis of Cost caused by Industrial Accident related with Return on Investment in 1956. These analyses have been changed greatly due to changing times and environments. That's why this paper proposes a method of calculation of return on investment concerning with industrial accidents and its protection measures that shall be utilized in global. In this paper, we propose “Cost Characteristic of Occupational Accident” first. “Cost Characteristic of Occupational Accident” is a total cost caused by a maximum occupational accident by a machine and manufacturing system that does not implement any protection measures to prevent. The total cost shall include expenditure to recover from an occupational accident. Secondary, we define “Initial Investment for Protection Measures” that prevents a machine and manufacturing system form occurring occupational accidents. Then, compared with “Cost Characteristic of Occupational Accident” and “Initial Investment for Protection Measures”, relation between them shall be a guideline of return on investment for a specific machine and manufacturing system.
Proceedings Papers
Proc. ASME. ICONE17, Volume 3: Thermal Hydraulics; Current Advanced Reactors: Plant Design, Construction, Workforce and Public Acceptance, 287-296, July 12–16, 2009
Paper No: ICONE17-75287
Abstract
A series of experiments was performed for rewetting phenomena on dryed-out fuel rod surfaces under post-BT (Boiling Transition) conditions with high-pressure and high-water flow rate simulating anticipated operational occurrences of a BWR. An analytical model for rewetting velocity, defined by a propagation velocity of a quench front, has been developed on the basis of the experimental results. The rewetting for the post-BT condition is characterized by the faster propagation of the quench front than that for reflood phase conditions during a postulated large-break loss-of-coolant accident. In order to provide an explanation of this characteristic, the present analytical model took an effect of a precursory cooling into account by modifying the existing correlation by Sun-Dix-Tien [1] which is based on a one-dimensional analysis in a flow direction during the reflood phase. The present model demonstrates that the precursory cooling can significantly increase the rewetting velocity by more than an order of magnitude. Applying the experimental correlation developed in the separately conducted experiment into the heat transfer coefficient in the present model at a wet and a dry region with precursory cooling, our data of the rewetting velocity as well as the wall temperature profiles for the variable flow rates are successfully predicted.
Proceedings Papers
Proc. ASME. ICONE16, Volume 4: Structural Integrity; Next Generation Systems; Safety and Security; Low Level Waste Management and Decommissioning; Near Term Deployment: Plant Designs, Licensing, Construction, Workforce and Public Acceptance, 757-767, May 11–15, 2008
Paper No: ICONE16-48754
Abstract
Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary sides in case of core temperature excursion. Test 6-1 is the first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reasons of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection.
Proceedings Papers
Hideo Nakamura, Tadashi Watanabe, Takeshi Takeda, Hideaki Asaka, Masaya Kondo, Yu Maruyama, Iwao Ohtsu, Mitsuhiro Suzuki
Proc. ASME. ICONE16, Volume 3: Thermal Hydraulics; Instrumentation and Controls, 659-668, May 11–15, 2008
Paper No: ICONE16-48615
Abstract
The Japan Atomic Energy Agency (JAEA) started OECD/NEA ROSA Project in 2005 to resolve issues in the thermal-hydraulic analyses relevant to LWR safety through the experiments of ROSA/LSTF in JAEA. More than 17 organizations from 14 NEA member countries have joined the Project. The ROSA Project intends to focus on the validation of simulation models and methods for complex phenomena that may occur during DBEs and beyond-DBE transients. Twelve experiments are to be conducted in the six types. By utilizing the obtained data, the predictability of codes is validated. Nine experiments have been performed so far in the ROSA Project to date. The results of two out of these experiments; PV top and bottom small-break (SB) LOCA simulations are studied here, through comparisons with the results from pre-test and post-test analyses by using the RELAP5/MOD3.2 code as a typical and well-utilized/improved best estimate (BE) code. The experimental conditions were defined based on the pre-test (blind) analysis. The comparison with the experiment results may clearly indicate a state of the art of the code to deal with relevant reactor accidents. The code predictive capability was verified further through the post-test analysis. The obtained issues in the utilization of the RELAP5 code are summarized as well as the outline of the ROSA Project.
Proceedings Papers
Kenji Arai, Tomohisa Kurita, Mikihide Nakamaru, Yasunobu Fujiki, Hideo Nakamura, Masaya Kondo, Hiroyuki Obata, Rumi Shimada, Ken Yamaguchi
Proc. ASME. ICONE10, 10th International Conference on Nuclear Engineering, Volume 3, 741-747, April 14–18, 2002
Paper No: ICONE10-22442
Abstract
A passive containment cooling system (PCCS) using a shell-and-tube type heat exchanger has been developed as a decay heat removal system following a severe accident. A horizontal heat exchanger has been studied for the PCCS heat exchanger since it has several advantages over a vertical large diameter tube heat exchanger that was originally proposed for the SBWR. Based on the fundamental thermal-hydraulic test using a single horizontal U-tube, the feasibility of the horizontal tube type PCCS has been confirmed, and the analysis models for the steam condensation heat transfer with a non-condensable and the pressure loss with steam condensation have been established. In addition, the criterion for the film dryout type heat transfer deterioration in the boiling side has been clarified in the test. The heat exchanger performance is affected by the multi-dimensional thermal-hydraulic behavior in the cooling water pool and the interactions among the multiple heat transfer tubes. In order to clarify the multi-dimensional behavior, numerical analyses have been conducted employing a two-fluid model. From the analyses results, it has been confirmed that the horizontal PCCS heat exchanger meets the design requirements for both the heat removal and the pressure loss and there would be no film dryout type heat transfer deterioration occurred in the cooling pool.
Proceedings Papers
Masaya Kondo, Hideo Nakamura, Yoshinari Anoda, Sadanori Saishu, Hiroyuki Obata, Rumi Shimada, Shinichi Kawamura
Proc. ASME. ICONE10, 10th International Conference on Nuclear Engineering, Volume 3, 643-651, April 14–18, 2002
Paper No: ICONE10-22403
Abstract
A horizontal in-tube condensation heat exchanger is under investigation to be used for a passive containment cooling system (PCCS) of a next generation-type BWR. The flow conditions in the horizontal condenser tube were observed both visually and by local void fraction fluctuation. The observed flow regimes at a rated condition were annular flow at the tube inlet, and turned gradually into wavy flow and smooth stratified flow along the length of the tube. It was found further that frequency of the roll waves that appear on the liquid film in the annular flow is closely related to the measured local condensation heat transfer coefficient. Based on the flow observation, the roll wave frequency and measured condensation heat transfer coefficient, a model is proposed which predicts the condensation heat transfer coefficient particularly for annular flows around the tube inlet region. The proposed heat transfer model predicts well the influences of pressure, local gas-phase velocity and film thickness.
Proceedings Papers
Hideyuki Nakayama, Haruhiko Mori, Jun Shimizu, Takahiro Ito, Hideo Nakamura, Yoshiyuki Tsuji, Yutaka Kukita
Proc. ASME. ICONE12, 12th International Conference on Nuclear Engineering, Volume 3, 601-607, April 25–29, 2004
Paper No: ICONE12-49314
Abstract
Non-zero time averaged flows can be induced in fluctuating flows. Shimizu et al. found such time-averaged streaming to occur in stratified fluids excited in a vertical, stationary cylinder at a frequency close to the natural frequency of the axisymmetric interface wave [1]. PIV measurements have revealed intermittent formation of a stagnation region close the fluid-fluid interface that enables a jet-like flow departing from the center of the interface to be sustained over cycles. The jet-like flow induces a recirculating streaming flow that extends several diameters away from the interface. Difficulty still remains in quantifying experimentally the detailed flow structure and defining the driving mechanisms, since the streaming is more than one order of magnitude smaller than the forced oscillating flow. The present paper presents PIV data that show temporal evolution of the flow in each cycle, the effect of fluid viscosities and the excitation frequency and amplitude.
Proceedings Papers
Hideyuki Nakayama, Haruhiko Mori, Jun Shimizu, Takahiro Ito, Hideo Nakamura, Yoshiyuki Tsuji, Yutaka Kukita
Proc. ASME. ICONE12, 12th International Conference on Nuclear Engineering, Volume 1, 831-837, April 25–29, 2004
Paper No: ICONE12-49569
Abstract
Non-zero time averaged flows can be induced in fluctuating flows. Shimizu et al. found such time-averaged streaming to occur in stratified fluids excited in a vertical, stationary cylinder at a frequency close to the natural frequency of the axisymmetric interface wave [1]. PIV measurements have revealed intermittent formation of a stagnation region close the fluid-fluid interface that enables a jet-like flow departing from the center of the interface to be sustained over cycles. The jet-like flow induces a recirculating streaming flow that extends several diameters away from the interface. Difficulty still remains in quantifying experimentally the detailed flow structure and defining the driving mechanisms, since the streaming is more than one order of magnitude smaller than the forced oscillating flow. The present paper presents PIV data that show temporal evolution of the flow in each cycle, the effect of fluid viscosities and the excitation frequency and amplitude.
Proceedings Papers
Proc. ASME. ICONE14, Volume 2: Thermal Hydraulics, 25-32, July 17–20, 2006
Paper No: ICONE14-89052
Abstract
Since the transport of momentum, heat and mass tightly links with local interfacial characteristics it is essential to know the local interfacial parameters in various two-phase flows. The interfacial velocity plays a determinant role in determining the other interfacial parameters such as the interfacial area concentration and so on. It is accordingly one of the most important parameters in analyzing two-phase flow. However, it also is one of the most difficult parameters to measure up to now. Based on the application of the interfacial measurement theorem to several four-sensor probes, the present study established a theoretical foundation of the measurement method for the local instantaneous interfacial velocity in multidimensional two-phase flow by using three independent four-sensor probes. Since we can find three independent four-sensor probes in a multi-sensor probe, which has more than four sensors, by sharing the sensors of the first four-sensor probe with the sensors of the others, a five- or six-sensor probe including at least one set of three four-sensor combinations was recommended to measure the local instantaneous interfacial velocity, interfacial area concentration and so on in multidimensional two-phase flow. A six-sensor probe was developed and employed in the practical measurement in an air-water multi-dimensional two-phase flow in a pool. The six-sensor probe measurements were checked against the gas flow rate measurement using a rotameter and a manometer. The comparing results were very satisfactory.
Proceedings Papers
Masaya Kondo, Hideo Nakamura, Yutaka Kukita, Tomohisa Kurita, Kenji Arai, Toshihiko Okazaki, Ryousuke Inoue
Proc. ASME. ICONE14, Volume 2: Thermal Hydraulics, 685-693, July 17–20, 2006
Paper No: ICONE14-89652
Abstract
An experiment was performed using a horizontal condenser tube for the PCCS horizontal heat exchanger to confirm the performance and to estimate the thermal-hydraulic behavior in the tube thermally and visually. The PCCS, driven by the pressure difference between the drywell and wetwell, is a safety equipment to prevent the containment break for more than 1 day due to overpressurization at the containment spray failure situation. An annular flow, a wavy flow, a supercritical flow, a stratified flow and the liquid slug propagation are observed in the tube at the nominal condition. The flow regime transition, induced by the condensation, roughly agrees to the existing models. The observation also agrees with the condensate distribution in the tube circumference direction, which is suggested by the local thermal characteristics.
Proceedings Papers
Proc. ASME. ICONE14, Volume 2: Thermal Hydraulics, 695-702, July 17–20, 2006
Paper No: ICONE14-89672
Abstract
Series of out-of-pile experiments to obtain the knowledge on the transient void behavior during reactivity initiated accidents are in progress at JAEA. In the present series of experiments, the transient void behavior in a test section of 2 × 2 bundle geometry under atmospheric pressure condition was measured using an impedance technique. The measuring areas and the arrangement of electrodes for the impedance technique were defined on the basis of numerical analyses and scaled model experiments. The comparison was made between the impedance and differential pressure techniques for steady boiling experiments to estimate the accuracy of the impedance technique. The impedance technique showed a good agreement with the void fraction estimated from the differential pressure. The transient void behavior in the bundle geometry was measured using the impedance technique. It was clarified that the transient void behavior depends on both the subcooling of inlet water and the heat generation rate of simulated fuel rod. Local void fraction was influenced by the ratio of flow area to heat transfer area of the simulated fuel rod.
Proceedings Papers
Kazuhiro Itoh, Taro Itoh, Yutaka Kukita, Hiroyuki Koterazawa, Hiroo Kondo, Nobuo Yamaoka, Hiroshi Horiike, Mizuho Ida, Hideo Nakamura, Hiroo Nakamura, Takeo Muroga
Proc. ASME. ICONE14, Volume 2: Thermal Hydraulics, 463-468, July 17–20, 2006
Paper No: ICONE14-89437
Abstract
Waves on a liquid-lithium jet flow, simulating a proposed high-energy beam target design, have been measured using an optical technique based on specular reflection of a single laser beam on the jet surface. The streamwise and spanwise fluctuations of the local free-surface slope were least-square fitted with a sinusoidal curve to makeup the signals lost due to the constriction in the optical arrangement. The waveform was estimated with an assumption that wave phase speed can be calculated using the dispersion relation for linear capillary-gravity waves. The direction of propagation on the jet surface was also evaluated so that the wave amplitudes, calculated by integral of slope angle signal, agree consistently in streamwise and spanwise direction. These measurements and analyses show that the waves at the measurement location for a jet velocity of 1.2 m/s can best be represented by oblique waves with an inclination of 1.23 rad, a wavelength of 3.8 mm and a wave amplitude of about 0.05 mm.
