Condensation experiments for counter-current flow limitation in an inverted U-tube

In this study, we measured counter-current flow limitation (CCFL) characteristics in an inverted U-tube (18.4 mm diameter and 1.0 m straight-part length) simulating steam generator (SG) U-tubes under conditions of steam condensation at pressures of 0.1–0.14 MPa. Differential pressure ΔP between the top of the inverted U-tube and the lower tank was measured, and the flow patterns wave estimated by comparing the waveforms of ΔP with those in air–water experiments. As a result, we classified the flow patterns under CCFL conditions into CCFL-P, CCFL-L and CCFL-T. The falling water flow rate under CCFL conditions slightly increased as the pressure increased and the cooling water temperature decreased (subcooling of cooling water increased). In the case of CCFL-L, CCFL characteristics in the inverted U-tube were between those in air–water and saturated steam–water experiments at 0.1 MPa. Furthermore, we derived a Wallis type CCFL correlation and its uncertainty from CCFL data, including previously measured data, i.e., J*^1/2_G + 0.88J_L*^1/2 = 0.76 ± 0.05.     

Analysis of LSC phenomena of ATLAS cold leg SBLOCA tests using MARS-KS code

Loop seal clearing (LSC) is an important phenomenon for the safety of a pressurized water reactor (PWR) during a small-break loss-of-coolant accident (SBLOCA). The investigation on an LSC phenomenon of 4″, 6″, and 8.5″ break cold leg SBLOCAs simulated by Advanced thermal–hydraulic Test Loop for Accident Simulation (ATLAS) is performed using a Multi-dimensional Analysis of Reactor Safety-KINS Standard (MARS-KS) code. The LSC triggers earlier for larger break sizes during tests and calculations. LSCs occur during the simultaneous sudden decrease of steam condensation rate and the sudden increase of the break volumetric flow rate while the core volumetric flow rate increases slowly in calculation. The five phases of an SBLOCA transient are blowdown, pressure plateau, LSC, boil-off, and core-recovery phase, which can be identified by observing the volumetric flow rate and the time-dependent pressure variation. Loop seal refilling (LSR) occurs due to the trivial steam flow rate to the crossover leg inlet in calculation. The sensitivity analysis shows that the combination of countercurrent flow limitation (CCFL) model option for hot leg and steam generator (SG) inlet (Kutateladze, c = 1.36, m = 1), crossover legs (Kutateladze, c = 1, m = 1), and SG U-tubes (Wallis, c = 1, m = 1) provide good prediction of the LSC phenomenon and thermal-hydraulics behaviors in an SBLOCA transient by MARS-KS code calculation.     

Experimental Study of Counter-Current Two-Phase Flow in Horizontal Tube Connected to Inclined Riser

In order to determine the Counter-Current Flow Limitation (CCFL) in hot legs of PWRS, CCFL characteristics of air-water and saturated steam-water flow were experimentally investigated in a modeled flow path of a horizontal tube connected to an inclined riser. The ranges of dimensions of experimental tubes were as follows: diameter D 0.026–0.076 m, length of horizontal tube H 0.01–0.4 m, length of inclined riser I 0.038–0.6 m and inclination angle of inclined riser θ 40° or 45°.

Wallis-type correlation (J _g ^*1/2+mJ _i ^*1/2=C) was applicable to the data during a steady separated flow. An analysis based on envelope theory showed that the constant C should be a function of H/D and I. A function of C with those parameters was empirically determined by using data obtained in this study. The developed function correlated well with the results of Richter et al. (D = 0.203 m, H = 1.26 m, I = 0.5 m and θ = 45°). The constant m in the Wallis-type correlation was almost constant 0.75. The problems were discussed, which should be made clear to apply the correlation obtained in this study to an actual PWR hot leg.     

Correlation for countercurrent flow limitation in a PWR hot leg

Numerical simulations were done to evaluate countercurrent flow limitation (CCFL) characteristics in a pressurized water reactor (PWR) hot leg with the diameter of 750 mm by using a volume of fluid (VOF) method implemented in the CFD software, FLUENT6.3.26. The calculated CCFL characteristics agreed well with known values including the UPTF data at 1.5 MPa. Sensitivity analyses for system pressures up to 8 MPa showed that the calculated CCFL characteristics in the Wallis diagram were slightly mitigated from 0.1 MPa to 1.5 MPa with increasing system pressure, but they did not change from 1.5 MPa to 8MPa. Using the CCFLs calculated in this study and values measured under air–water and steam–water conditions, a CCFL correlation and its uncertainty were derived.     

Determination of Multiplication Factors Using Experimental Lattice Parameters of Cluster-Type Fuel Lattices

A new four-factor formula is proposed for deriving a “finite multiplication factor” k*—the ratio between neutron production and absorption rates in a finite medium—from experimentally determined lattice parameters in a slightly-enriched Pu-U lattice.

The “two-group multiplication factor” k* _∞2—related to τ+ L² and equal to (1 τ B² _C). (1 +L²B² _G at criticality—is derived from k*.

The experimental lattice parameters are corrected to account for neutron leakage, from which the “infinite multiplication factor” K∞ is derived.

There are found differences between k*, k*_∞2 and k _∞ amounting to about 1–2% in the Advanced Thermal Reactor (ATR)-type heavy-water lattice, despite the fact that all these three quantities are often indiscriminately called “infinite multiplication factor.”

The proposed four-factor formula for deriving k* was applied to the Deuterium Critical Assembly (DCA) of 1.2%-enriched UO₂ lattices of 28-pin clusters in square arrays spaced at 22.5 cm. The values of k* calculated with a lattice design code METHUSELAHH II were in fairly good agreement wTith those determined by experiment. The calculation tended to somewhat overestimate k*, particularly in lattices of highly voided coolant. A similar tendency was observed in the case of 1.5%-enriched UO₂ lattices of 22.5 cm pitch in the ATR Sumitomo-Two-Region Critical Facility.     

Void distribution and bubble motion in bubbly flows in a 4×4 rod bundle. Part II: numerical simulation

Numerical simulations of bubbly flows in a four by four rod bundle are carried out using a multi-fluid model to examine effects of the numerical treatment of phase distribution and drag model. The transport equations of bubble number density and void fraction are used as the continuity equation of the gas phase. Two drag models are tested: one of them accounts for the bubble deformation (aspect ratio), whereas the other does not. The rod diameter, the rod pitch and the hydraulic diameter of the rod bundle are 10, 12.5 and 9.1 mm, respectively. The gas and liquid volume fluxes are J_G = 0.06 m/s and J_L = 0.9 and 1.5 m/s, respectively. The bubble diameter ranges from 1 to 5 mm. Comparisons between the numerical and measured data show that (1) the restriction on bubble lateral motion due to the presence of rods can be taken into account by using the transport equation of bubble number density, whereas that of the void fraction cannot deal with the restriction and causes large errors in the distribution of void fraction and (2) the reduction in the bubble-relative velocity near the wall is predictable by using the drag model accounting for the bubble deformation effect.     

The FeSe2(cr) solubility determined by solubility experiments of Se co-existing with Fe

To determine the equilibrium constant for ferroselite (FeSe₂(cr)) dissolution reaction, FeSe₂(cr) solubility experiments were performed at 298 ± 1 K from both the over- and under-saturation directions with Fe–Se precipitates that were aged at 348 K. X-ray diffraction (XRD) analysis detected only FeSe₂(cr) as the Se solid phase in the equilibrated precipitates. The E_h values of the equilibrated suspensions ranged from ?188.6 to ?4.9 mV vs. standard hydrogen electrode (SHE) and the pH values ranged from 6.00 to 8.76. Based on the available thermodynamic data, Se₄^2? and Fe²⁺ are thermodynamically stable within this E_h–pH range. Agreement between the solubility data obtained from the over- and under-saturation directions lends credence to the attainment of equilibrium at 298 ± 1 K. The thermodynamic interpretations using the specific ion interaction theory (SIT) model showed that E_h values and the concentrations of Se and Fe are well represented by the 2FeSe₂(cr) solubility reaction (2FeSe₂(cr) ? 2Fe²⁺ + Se₄^2? + 2e^?) with log₁₀K^○ = ?17.09 ± 0.28. The obtained log₁₀K^○ value falls within the uncertainty limits of the log₁₀K^○ value calculated from the available thermodynamic data.     

Trend in criticality calculation results of borated-water-moderation experimental cores relating to 10B(n, α) cross section

ABSTRACT

In connection with the accuracy of the ¹⁰B(n, α) cross section in the thermal- and epithermal-neutron energy regions, criticality calculation results were examined for six benchmark sets of light-water-moderation critical experiments of UO₂ and MOX fuel lattice cores with un-borated and borated water. Two of the benchmark sets were those implemented in the Tank-Type Critical Assembly (TCA). The others were taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP), and the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP). The enrichments of the UO₂ fuel range from 1.9 wt% to 2.6 wt%, and the Pu contents of the MOX fuel do from 2.0 to 6.6 wt%. The boron concentrations in water are up to 1511 ppm. The effective neutron multiplication factors (k_eff ) were taken from the published documents. They were calculated with continuous-energy Monte Carlo calculation codes in combination with JENDL-4.0, and other evaluated nuclear data libraries. It was confirmed that the k_eff values of the critical cores increased with the boron concentrations, which indicates that the ¹⁰B(n, α) cross section in the thermal- and epithermal-neutron energy regions should be larger than those in JENDL-4.0 and other libraries.     

Enrichment of the Uranium Isotopes by Electromigration in a Cation Exchange Membrane

The γ-radiolysis of water subjected to gas bubbling has been studied using a specially desinged gasloop. During the irradiation, N₂ gas was bubbled from the bottom of the irradiation vessel. As the N₂ gas feed rate was raised, the apparent G(H₂) value increased in keeping therewith, from 5 × l0^?3 to 0.26. However in the presence of a sufficient amount of O₂ or H₂O₂, G(H₂) was raised almost to the level of the molecular yield. With reasonable assumptions, it could be concluded that 3～5 × 10^?6 mol/l of H₂O₂ was sufficient to reduce the back reaction of molecular products to less than 10% under the present experimental conditions. It was also found that the G(H₂) value increased with CH₃OH concentration roughly in proportion to log(CH₃OH), and reached 3.1 with 0.1 mol/l CH₃OH.     

Soft X-Ray Emission in the Water Window Region with Nitrogen Filling in a Low Energy Plasma Focus

For operation of the plasma focus in nitrogen, a focus pinch compression temperature range of 74–173 eV (0.86 × 10⁶–2 × 10⁶ K) is found to be suitable for good yield of nitrogen soft X-rays in the water window region. Using this temperature window, numerical experiments using five phase Lee model have been investigated on UNU/ICTP PFF and APF plasma focus devices with nitrogen filling gas. The Lee model was applied to characterize and optimize these two plasma focus devices. The optimum nitrogen soft X-ray yield was found to be Y_sxr = 2.73 J, with the corresponding efficiency of 0.13 % for UNU/ICTP PFF device, while for APF device it was Y_sxr = 4.84 J, with the corresponding efficiency of 0.19 % without changing the capacitor bank, merely by changing the electrode configuration and operating pressure. The Lee model code was also used to run numerical experiments for optimizing soft X-ray yield with reducing L₀, varying z₀ and ‘a’. From these numerical experiments we expect to increase the nitrogen soft X-ray yield of low energy plasma focus devices to maximum value of near 8 J, with the corresponding efficiency of 0.4 %, at an achievable L₀ = 10 nH.     

Properties of Electrodeposited Tungsten Coatings on Graphite Substrates for Plasma Facing Components

Tungsten coating on graphite substrate is considered as one of promising candidate materials of plasma facing components. In this study, tungsten coatings on graphite substrate were successfully prepared by direct current (DC) and pulse current (PC) electrodeposition methods in Na₂WO₄–WO₃ molten salt under the air atmosphere. Pores were found on the surfaces of the tungsten coatings produced by DC electrodeposition method. For the coatings fabricated by PC method, compact and smooth tungsten coatings were successfully obtained. The crystal structure, morphology, density, microhardness, adhesive strength, oxygen content and the thermal conductivity of the coatings fabricated by PC method were investigated. The obtained tungsten coatings had a body centered cubic structure. After electro-deposition for 100 h, the thickness of the tungsten coating reached 810.02 ± 10.40 μm and the oxygen content was 0.03 wt%. The thermal conductivity of the tungsten coating was 134.29 W m^?1 K^?1. The density of the tungsten coating was 18.83 g cm^?3. The hardness of the coating was 492.0 ± 7.8 HV. After deuterium plasma irradiation, the tungsten coatings were prone to blistering.     

Development of lightweight portable neutron survey meter

We have developed a lightweight portable neutron survey meter comprising a proportional gas counter containing a mixed gas of methane gas and nitrogen gas for measuring the ambient neutron dose equivalent H^*(10) up to about 20 MeV neutrons intended for use in nuclear power plants, and accelerator facilities. Since no heavy polyethylene moderator is used, the survey meter is only about 2.2 kg in weight causing a weight reduction of 70% or more compared to that of the conventional moderated-type survey meter. The spectrum-weight function, G(E) is adopted for dose conversion.

In order to evaluate the energy characteristics of the developed survey meter, mono-energetic neutron reference fields and continuous energy neutron reference fields were used. The evaluation was also carried out by the PHITS calculation. Although the neutron energy response to the mono-energetic neutron fields was highly deviated from the ambient dose equivalent H^*(10) in keV energy region, the response to H^*(10) showed very good agreement only within 25% to the continuous energy neutron fields which are close to the actual work-place neutron fields. The neutron energy response was also investigated for high energy quasi-mono-energetic neutron fields and showed much better results compared with those of conventional moderated-type neutron survey meters.     

Behavior of cesium molybdate,Cs2MoO4, in severe accident conditions. (1) Partitioning of Cs and Mo among gaseous species

In order to better understand the behavior of cesium in severe accident of Light Water Reactor (LWR), the high-temperature chemistry of Cs₂MoO₄ in H₂O + H₂ gas was studied. The pseudo–binary system, Cs₂MoO₄–MoO₃, was thermochemically modeled with Redlich–Kister formulation to form a basis to analyze the high-temperature behavior of Cs₂MoO₄. The model prediction was compared with the thermogravimetric measurements of Cs₂MoO₄ in dry and humid argon, which revealed that the mass-loss rate was enhanced in humid atmosphere. The thermochemical model was further applied to predict the partitioning of cesium and molybdenum among gaseous species in the boiling water reactor-core degradation condition typical of short-term station blackout. Effects of the total pressure (3.5–75 bar) as well as the H₂/H₂O ratio (1/4000–2) were examined. CsOH(g) is the predominant cesium species, when the damaged fuel temperature is higher than 2000 K at higher steam pressures, but Cs₂MoO₄(g) would become more important as the steam cools toward the steam dome. The condensation of Cs₂MoO₄ occurs below ～1900 and ～1550 K at 75 and 3.5 bar, respectively. Besides, the ideal mixing of complex component model has also been examined for its simplicity. The latter gave satisfactory prediction as far as the condensed phase composition is concerned.     

Experimental study of gas–liquid two-phase flow through packed bed under natural circulation conditions

Dry-out phenomena in packed beds or porous media may cause a significant digression of cooling/reaction performance in heat transfer/chemical reactor systems. One of the phenomena responsible for the dry-out in packed beds is known as the counter-current flow limitation (CCFL). In order to investigate the CCFL phenomena induced by gas–liquid two-phase flow in packed beds inside a pool, a natural circulation packed bed test facility was designed and constructed. A total of 27 experimental conditions covering various packing media sizes (sphere diameters: 3.0, 6.4 and 9.5 mm), packed bed heights (15, 35 and 50 cm) and water level heights (1.0, 1.5 and 2.0 m) were tested to examine the CCFL criteria with adiabatic air–water two-phase flow under natural circulation conditions. Both CCFL and flow reversal phenomena were observed, and the experimental data including instantaneous and time-averaged void fraction, differential pressure and superficial gas–liquid velocities were collected. The CCFL criteria were determined when periodical oscillations of void fraction and differential pressure appear. In addition, the Wallis correlation for CCFL was utilized for data analysis, and the Wallis coefficient, C, was determined experimentally from the packed bed CCFL tests. Compared to the existing data-sets in literature, the higher C values obtained in the present experiment suggest a possibly higher dry-out heat flux for natural circulation debris systems, which may be due to the water supply from both top and bottom surfaces of the packed beds. Considering the effects of bed height and hydraulic diameter of the packing media, a newly developed model for the Wallis coefficient, C, under natural circulation CCFL is presented. The present model can predict the experimental data with an averaged absolute error of ±7.9%.     

Effect of Gas Atoms on Recovery of Work Hardening in Al-Li Alloys

In the first report of this study, dealing with CCFL and CCFL breakdown phenomena associated with the injection of emergency core cooling spray water into upper plenum during refill-reflood phase of a BWR LOCA, the following tests results were obtained.

The injected water maintained two-phase pool across the top of entire core after CCFL breakdown. The pool level oscillated near spray elevation. The objective of this paper is to clarify the mechanism of these phenomena, evaluating steam and spray flow effects on CCFL breakdown.

It is found that when spray flow rate was slightly larger than the CCFL drainage deter- mined by core steam flow, pool maintained at some constant level near spray elevation, after CCFL breakdown. On the other hand, when spray flow was appreciably larger than CCFL drainage, pool level slowly oscillated. The oscillation was caused by significant changes in steam condensation rate, and the corresponding subcooling penetration into the fuel bundles, when the pool level passed the spray elevation. The TRAC-BD1 analysis of test results suggested the small sector wall effect of test apparatus on CCFL breakdown phenomena.     

Condensing Heat Transfer in Steam-Water Condensing-injector

An experiment on the direct heat transfer process between supersonic steam and subcooled water jet was performed, using a steam-water condensing-injector. Photographic observation provided information on the state of flow, and establishment of a critical separate steam-water flow was confirmed. The temperature and pressure distributions along the flow were measured and the effective coefficients of condensing heat transfer were evaluated from the observed data, based on a model embodying an idealized interface between vapor and liquid. In the vicinity of the water nozzle exit, where the vapor-liquid interface was distinct, the heat transfer coefficients obtained were 14–28 (cal/°C.cm².sec), and some correlation was observed among Nusselt, Reynolds and Jakob numbers, upon adopting the velocity and the physical properties of the steam phase. The relations Nu=6.0.Re ^0.9(Pr=1.04–1.10), and Re=1.8×10⁸.Ja ^3.0, i.e., Nu=1.6×10⁸.Ja ^2.7 were derived as a rough estimation. No clear correlation could be discerned in the corresponding data obtained from observation points further downstream, where a distinct steam-water interface no longer existed. In conclusion, it is proposed that, in deriving the correlations between Nu and Re or Ja, the physical properties of the vapor and the vapor-liquid relative velocity should be adopted, on account of the strong dependence of condensing heat transfer on steam velocity and water subcooling.     

Two-phase Flow Patterns in a Four by Four Rod Bundle

Air-water two-phase flow patterns in a four by four square lattice rod bundle consisting of an acrylic channel box of 68mm in width and transparent rods of 12mm in diameter were observed by utilizing a high speed video camera, FEP (fluorinated ethylene propylene) tubes for rods, and a fiberscope inserted in a rod. The FEP possesses the same refractive index as water, and thereby, whole flow patterns in the bundle and local flow patterns in subchannels were successfully visualized with little optical distortion. The ranges of gas and liquid volume fluxes, {J_G} and {J_L}, in the present experiments were 0:1 ? {J_L} < 2:0 m/s and 0:06 < {J_G} < 8:85 m/s, which covered typical two-phase flow patterns appearing in a fuel bundle of a boiling water nuclear reactor. As a result, the following conclusions were obtained: (1) the region of slug flow in the {J_G}-{J_L} flow pattern diagram is so narrow that it can be regarded as a boundary between bubbly and churn flows, (2) the boundary between bubbly and churn flows is close to the boundary between bubbly and slug flows of the Mishima and Ishii's flow pattern transition model, and (3) the boundary between churn and annular flows is close to the Mishima and Ishii's model.     

Flow structure and bubble characteristics of steam–water two-phase flow in a large-diameter pipe

The flow structure and bubble characteristics of steam–water two-phase upward flow were observed in a vertical pipe 155 mm in inner diameter. Experiments were conducted under volumetric flux conditions of J_G<0.25 m s⁻¹ and J_L<0.6 m s⁻¹, and three different inlet boundary conditions to investigate the developing state of the flow. The radial distributions of flow structure, such as void fraction, bubble chord length and gas velocity, were obtained by horizontally traversing optical dual void probes through the pipe. The spectra of bubble chord length and gas velocity were also obtained to study the characteristics of bubbles in detail. Overall, an empirical database of the multi-dimensional flow structure of two-phase flow in a large-diameter pipe was obtained. The void profiles converged to a so-called core-shaped distribution and the flow reached a quasi-developed state within a relatively short height-to-diameter aspect ratio of about H/D=4 compared to a small-diameter pipe flow. The PDF histogram profiles of bubble chord length and gas velocity could be approximated fairly well by a model function using a gamma distribution and log–normal distribution, respectively. Finally, the correlation of Sauter mean bubble diameter was derived as a function of local void fraction, pressure, surface tension and density. With this correlation, cross sectional averaged bubble diameter was predicted with high accuracy compared to the existing constitutive equation mainly being used in best-estimate codes.     

Requirements for a radiation monitoring system in connection with risk and dose

Conclusions In summary, we have arrived at the seemingly paradoxical conclusion that the requirements for the radiation-monitoring system in region 2, where by definition the average individual dose is less than in region 1, are higher. Conversely, it would seem that if the dose load is smaller, then there is no need to complicate the monitoring system; for example, it is sufficient to establish a monitoring level that is the same as in region 1. The paradox is solved if one takes account of the fact that the lower dose loads in region 2 are associated with the lower content of radionuclides in objects in the environment, i.e., the relatively small useful signals, detected by the given monitoring system. This requires better monitoring systems, since weak signals must be detected and discriminated under conditions of random noise, due to background sources of inoizing radiation and other types of noise. In conclusion, we note that the materials presented in this paper make it possible to distinguish the radioecological environment in two regions quantitatively. Specifically, it must be assumed that regions 1 and 2 have the same radioecological state ifK ₁ ^* <K ₂ ^* ;D ₂<D ₁;N ₁≪N ₂ and ln(K ₁ ^* /K ₂ ^* )<K _thr(K ₂ ^* −K ₁ ^* )/(K ₁ ^* K ₂ ^* ). The latter approximate relation follows from Eq. (3) and the conditiong ₁(K _thr)<g ₂(K _thr), whereK _thr is the threshold value or the control level of the yearly effective collective dose andg(K) is the distribution density of the effective collective dose. Under these conditions, the requirements imposed on the radiation monitoring system in region 2 are more stringent than for an analogous system in region 1. Moscow Scientific and Industrial Association “Radon.” Translated from Atomnaya énergiya, Vol. 88, No. 6, pp. 476–480, June, 2000.     

Effect of steam condensation on countercurrent flow limiting in nearly horizontal two-phase flow

An analytical model that includes the steam condensation effect has been derived and a parametric study has been performed. In addition, a series of experiments were performed and a total of 34 experimental data for the onset of countercurrent flow limiting (CCFL) in nearly horizontal countercurrent two-phase flow have been obtained for various flow rates of water. Comparisons of the present CCFL data with slug formation models show that the agreement between the present as well as the existing model and the data is about the same. However, the deviation between Taitel and Dukler’s model predictions and the data is the largest when j_f<0.04 m s⁻¹. A parametric study of the effect of condensation using the present model shows that, when all local conditions are similar, the model predicted local gas velocities that cause the onset of flooding are slightly lower when condensation occurred. Based on the visual observation and the evaluation of the present work, it has been concluded that the criterion derived for the onset of slug flow can be directly used to predict the onset of inner flooding in nearly horizontal two-phase flow within the experimental ranges of the present work.