Hideo Nakamura

OVERVIEW OF RECENT EFFORTS THROUGH ROSA/LSTF EXPERIMENTS

JAEA started the LSTF experiments in 1985 for the fourth stage of the ROSA Program (ROSA-IV) for the LWR thermal-hydraulic safety research to identify and investigate the thermal-hydraulic phenomena and to confirm the effectiveness of ECCS during small-break LOCAs and operational transients. The LSTF experiments are underway for the ROSA-V Program and the OECD/NEA ROSA Project that intends to resolve issues in thermal-hydraulic analyses relevant to LWR safety. Six types of the LSTF experiments have been done for both the system integral and separate-effect experiments among international members from 14 countries. Results of four experiments for the ROSA Project are briefly presented with analysis by a best-estimate (BE) code and a computational fluid dynamics (CFD) code to illustrate the capability of the LSTF and codes to simulate the thermal-hydraulic phenomena that may appear during SBLOCAs and transients. The thermal-hydraulic phenomena dealt with are coolant mixing and temperature stratification, water hammer up to high system pressure, natural circulation under high core power condition, and non-condensable gas effect during asymmetric SG depressurization as an AM action.

Evaluation of Containment Failure Probability by Ex-Vessel Steam Explosion in Japanese LWR Plants

The containment failure probability due to ex-vessel steam explosions was evaluated for Japanese BWR and PWR model plants. A stratified Monte Carlo technique (Latin Hypercube Sampling (LHS)) was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The evaluation was made for three scenarios: a steam explosion in the pedestal area or in the suppression pool of a BWR model plant with a Mark-II containment, and in the reactor cavity of a PWR model plant. The scenario connecting the generation of steam explosion loads and the containment failure was assumed to be displacement of the reactor vessel and pipings, and failure at the penetration in the containment boundary. We evaluated the conditional containment failure probability (CCFP) based on the preconditions of failure of molten core retention within the reactor vessel, relocation of the core melt into the water pool without significant interference, and a strong triggering at the time of maximum premixed mass. The obtained mean and median values of the CCPF were 6.4x 10−2 (mean) and 3.9x 10−2 (median) for the BWR suppression pool case, 2.2x10−3 (mean) and 2.8x10−10 (median) for the BWR pedestal case, and 6.8X10−2 (mean) and 1.4x10−2 (median) for the PWR cavity case. The evaluation of CCFPs on the basis of core damage needs consideration of probabilities for the above-mentioned preconditions. Thus, the CCFPs per core damage should be lower than the values given above. The specific values of the probability were most dependent on the assumed range of melt flow rate and fragility curve that involved conservatism and uncertainty due to simplified scenarios and limited information.

Experiments on the Release of Gaseous Iodine from Gamma-Irradiated Aqueous CsI Solution and Influence of Oxygen and Methyl Isobutyl Ketone (MIBK)

Volatile iodine production due to radiation chemical effects is known to be an important uncertainty source in the evaluation of the source term for severe accidents of light water reactors (LWRs). The gaseous release fractions of molecular iodine and organic iodine from gamma-irradiated 10−4 mol/L cesium iodide aqueous solution were measuredwith the dose rate ∼ 7 kGy/h at room temperature. The solution was buffered with 0.1 mol/L boric acid and sodium hydroxide (pH ∼7) and contained up to 10−3 mol/L methyl isobutyl ketone (MIBK) as an organic additive. The concentrations of MIBK in the solution and oxygen in the sweep gas were changed as experimental parameters. The total iodine release fraction of the original aqueous inventory and the fraction released as organic iodine were 2–47 and 0.02–1.5%, respectively, at the end of 2 h of irradiation. They were dependent on both the aqueous MIBK concentration and oxygen concentration in the sweep gas. Under a constant cover gas condition, the total iodine release showed a decreasing trend and theorganic iodine release showed an increasing trend when the MIBK concentration increased. This behavior can be explained by the branching of the reaction path of radiolytic degradation of ketones depending on the availability of dissolved oxygen, and competition between iodine and organic compounds on the consumption of radicals produced by water radiolysis.

Upward air–water bubbly flow characteristics in a vertical square duct

In nuclear engineering fields, gas–liquid bubbly flows exist in channels with various shape and size cross-sections. Although many experiments have been carried out especially in circular pipes, those in a noncircular duct are very limited. To contribute to the development of gas–liquid bubbly flow model for a noncircular duct, detail measurements for the air–water bubbly flow in a square duct (side length: 0.136 m) were carried out by an X-type hot-film anemometry and a multi-sensor optical probe. Local flow parameters of the void fraction, bubble diameter, bubble frequency, axial liquid velocity and turbulent kinetic energy were measured in 11 two-phase flow conditions. These flow conditions covered bubbly flow with the area-averaged void fraction ranging from 0.069 to 0.172. A pronounced corner peak of the void fraction was observed in a quarter square area of a measuring cross-section. Due to a high bubble concentration in the corner, the maximum values of both axial liquid velocity and turbulent kinetic energy intensity were located in the corner region. It was pointed out that an effect of the corner on accumulating bubble in the corner region changed the distributions of axial liquid velocity and turbulent kinetic energy intensity significantly.

Small-scale Experiment on Subcooled Water Jet Injection into Molten Alloy by Using Fluid Temperature-Phase Coupled Measurement and Visualization

The plunging water jet behavior into a pool of a molten lead-bismuth alloy is experimentally investigated. The mixing and interactions of fluids were detected by measuring the fluid temperature as well as the fluid phase distinction by the newly developed bifunctional probes. In general knowledge of fuel-coolant interactions, the film boiling of water is caused immediately after the first contact of high-temperature melt and water, but the vapor film is locally collapsed by some reasons and the direct contact is extensively propagating in some cases which may produce the explosive vapor generation. In the melt-injection mode previously investigated by numerous researchers, the triggering of explosive interactions is considered as a local rewetting caused by instabilities of the vapor film as the melt temperature decrease. In the coolant-injection mode discussed by the present study, on the other hand, the water temperature poured into bulk melt continues to rise for penetration, in general, that should be effective to stabilize the film boiling. The present experiments showed, however, that the explosive boiling occurred in a condition that both water and melt initial temperatures were high enough for maintaining stable film boiling on the melt-water interface that is clearly different manner of the melt injection mode. Such unstable phenomena are observed when the instantaneous interfacial contact temperature exceeds the homogeneous nucleation temperature of water and the amount of saturated water is accumulated in a melt pool.

Flow-Induced Void Fraction Transition Phenomenon in Two-Phase Flow

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.Copyright

RELAP5/MOD3 Code Verification Through PWR Pressure Vessel Small Break LOCA Tests in OECD/NEA ROSA Project

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.Copyright

Primary-Side Two-Phase Flow and Heat Transfer Characteristics of a Horizontal-Tube PCCS Condenser

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 wet-well, 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. (authors)

RELAP5 analyses on the influence of multi-dimensional flow in the core on core cooling during LSTF cold-leg intermediate break LOCA experiments in the OECD/NEA ROSA-2 Project

Test-2 and Test-7 of the second Organization for Economic Co-operation and Development / Nuclear Energy Agency (OECD/NEA) Rig-of-Safety Assessment (ROSA-2) Project were performed with the Large Scale Test Facility (LSTF), which simulated thermal-hydraulic responses during a pressurized water reactor cold-leg intermediate break loss-of-coolant accident (IBLOCA). Test-2 simulated 17% cold-leg break with single failure of an emergency core cooling system (ECCS). The core liquid level decreased to the bottom at loop seal clearing (LSC), causing high cladding temperature excursion. Test-7 simulated 13% cold-leg break with full injection of the ECCS. Compared to Test-2, the cladding surface temperature in Test-7 was quite low due to greater liquid level recovery after the LSC. To well understand the observed phenomena and to improve the best-estimate code predictive capability, RELAP5 post-test analyses were performed. The RELAP5 analyses employed two core models: one is a single-channel core model that simulates the whole core with one channel of a vertical stack of nine equal-height volumes, and the other is a multiple-channel core model that is composed of three channels in which adjacent vertically stacked volumes are horizontally connected with cross-flow junctions. The analyses with the multi-channel core model predicted better than with the single-channel core model for such parameters as core-collapsed liquid level and cladding surface temperature for both Test-2 and Test-7, by more realistically representing multi-dimensional flow in the core. Such a practical method for better representation of multi-dimensional flows turned out to be important to improve the IBLOCA analysis.

Gas-Liquid Bubbly Turbulent Upward Flow in Square Duct

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.Copyright