Study on Pb-Bi-water direct contact boiling two-phase flow and heat transfer

For the development of 45w%Pb-55w%Bi cooled direct contact boiling water small fast reactor (PBWFR), Pb-Bi-water direct contact boiling two-phase flow loop has been fabricated and operated. The loop consists of a Pb-Bi flow loop (four heater pin bundle, a chimney, an upper plenum, a level meter tank, an air-water cooler, and an electromagnetic flow meter) and a water-steam flow loop (a pump, a preheated, an injection nozzle, the chimney, the upper plenum with mist separators and dryers, a condenser, a buffer tank, and an air-water cooler). At the rated operating condition system pressure is 7 MPa. The sub-cooled water was injected into a Pb-Bi flow in the chimney. A power of the heater pin bundle was controlled to obtain the inlet and outlet temperatures of the heater bundle. The Pb-Bi and steam flows were simulated analytically using one-dimensional models of frictional and form losses and a drag force. The Pb-Bi-steam two-phase frictional pressure loss was calculated by means of the two-phase flow multiplication factor of Lockhart-Martinelli model. It was found that Pb-Bi temperature decreased quickly in the chimney due to high heat transfer rate of Pb-Bi-water direct contact boiling. The volumetric overall heat transfer coefficient was 60–310 kW/m³K, and decreased with the superheat.     

Safety design of Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR)

In Pb–Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb–Bi coolant above the core, and Pb–Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb–Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits.     

Design study on reactor structure of Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR)

Pb–Bi-cooled direct contact boiling water small fast reactor (PBWFR) can produce steam by direct contact of feedwater with primary Pb–Bi coolant above the core, and circulate Pb–Bi coolant by means of buoyancy of steam bubbles. The PBWFR is capable of eliminating components of the cooling system such as primary pumps and steam generators, and thereby making the reactor system simple and compact. The specifications of the PBWFR are as follows: the fuel is Pu–U nitride; the core height is 75 cm; the core diameter is 278 cm; the average burnup is 80 GWd/t; the refueling interval is 10 years; the rated electric power is 150 MWe; the rated thermal power is 450 MWt; the core outlet/inlet temperatures are 460 °C/310 °C, respectively; and the operating steam pressure is 7 MPa. The reactor structure design has been formulated, where reactor vessel sizes are 4200 mm (ID) × 19,750 mm (H), the guard vessel is a closed type, the upper structure is made of chimneys, and the core support structure is hung up. An ultrasonic flow meter is installed inside the vessel. The seismic evaluation, design of refueling procedure and cost evaluation have been performed.     

Code development and safety analyses for Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR)

Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR) can produce steam from the direct contact of feed-water and lead bismuth eutectic (LBE) in the chimney of 3 m height, which eliminates the bulky and flimsy steam generators. Moreover, as the coolant LBE is driven by the buoyancy of steam bubbles, the primary pump is not necessary in the reactor. The conceptual design makes the reactor simple, compact and economical. Owing to the large thermal expansion coefficient of LBE and good performance of steam lift pump, the reactor is expected to have good passive safety. A new computer code is developed to investigate the thermal–hydraulic behaviors and safety performance of PBWFR in the present work. Unprotected rod run-out transient over power (UTOP) and unprotected loss of flow (ULOF)/unprotected loss of heat sink (ULOHS) are simulated to test and verify its safety. The results show that PBWFR has very good inherent safety due to the satisfactory neutron and thermal–physical properties of LBE. Cladding materials turn to be the key factor to restrict its safety performance and UTOP is more dangerous for PBWFR. It's suggested that it should appropriately reduce the maximum value of the control rods to mitigate the consequence of UTOP due to good reactivity feedbacks in the core.     

Characterization of solid particles sampled from condensates in boiling water reactor

The growth, activation and deposition of corrosion products are the primary sources of radiation buildup on the surface of out-of-core piping in nuclear power plants. The buildup of radiation can have negative effects on the performance of the facility and cause harm to staff during maintenance outages for refueling. This paper reports on the crystalline and amorphous structures of corrosion products sampled in the boiling water reactors in nuclear power plants of Kuo-Sheng and identified using an acid dissolving technique. X-ray diffraction, scanning electron microprobe and inductively coupled plasmaatomic emission spectroscopy were used to analyze the samples. The results indicate that the quantity of amorphous iron oxide at inlet of the condensate demineralizer in Unit 2 is higher than that in Unit 1. The proportion of crystalline to amorphous corrosion products can affect the efficiency of removal. Thus, these results can be used to explain the difference in removal efficiency of condensate demineralizers in different units. Moreover, the iron oxide structures with various properties were observed in different operational periods. It is probable that the higher proportion of amorphous structures with a smaller particle size would reduce efficiency in the removal of condensate demineralization in Unit 2.     

Experimental study on convective boiling heat transfer in narrow-gap annulus tubes

Since convective boiling or highly subcooled single-phase forced convection in micro-channels is an effective cooling mechanism with a wide range of applications, more experimental and theoretical studies are required to explain and verify the forced convection heat transfer phenomenon in narrow channels. In this experimental study, we model the convective boiling behavior of water with low latent heat substance Freon 113 (R-113), with the purpose of saving power consumption and visualizing experiments. Both heat transfer and pressure drop characteristics were measured in subcooled and saturated concentric narrow gap forced convection boiling. Data were obtained to qualitatively identify the effects of gap size, pressure, flow rate and wall superheat on boiling regimes and the transition between various regimes. Some significant differences from unconfined forced convection boiling were found, and also, the flow patterns in narrow vertical annulus tubes have been studied quantitatively.     

Nonlinear analysis of nuclear coupled density wave instability in time domain for a boiling water reactor core undergoing core-wide and regional modes of oscillations

The objective of the paper is to develop a nuclear coupled thermal-hydraulic model in order to simulate core-wide (in-phase) and regional (out-of-phase) stability analysis in time domain within the limitation of desktop research facility for a boiling water reactor subjected to operational transients. The integrated numerical tool, which is a combination of thermal-hydraulic, neutronic and fuel heat conduction models, is used to analyze a complete boiling water reactor core taking into account the strong nonlinear coupling between the core neutron dynamics and primary circuit thermal-hydraulics via the void-temperature reactivity feedback effects. The integrated model is validated against standard benchmark and published results. Finally, the model is used for various parametric studies and a number of numerical simulations are carried out to investigate core-wide and regional instabilities of the boiling water reactor core with and without the neutronic feedback effects. Results show that the inclusion of neutronic feedback effects has an adverse effect on boiling water reactor core by augmenting the instability at lower power for same inlet subcooling during core-wide mode of oscillations, whereas the instability is being suppressed during regional mode of oscillations in presence of the neutronic feedback. Dominance of core-wide instability over regional mode of oscillations is established for the present case of simulations which indicates that the preclusion of the former will automatically prevent the latter at the existing working condition.     

Experimentaland numerical investigation of liquid metal boiling in fuel subassemblies under natural circulation conditions

Boiling of Na-K eutectic alloy (22% Na; 78% K) in parallel channels under the natural circulation condition has been investigated at the AR-1 facility of IPPE. Hydrodynamics and heat transfer data have been obtained at this facility for various experiment set-ups of both single and parallel sections. A thermohydraulic code system based on the subchannel analysis code SABENA-3D has been developed and used to simulate these experiments. Comparisons show that the code system can reproduce the boiling phenomena in the test section with sufficient accuracy, correctly predicting the heat transfer conditions prior to and during the boiling.     

Study on Pb-Bi natural circulation phenomena

For the development of 45w%Pb-55w%Bi cooled direct contact boiling water small fast reactor (PBWFR), experimental study on Pb-Bi-water direct contact boiling two-phase flow has been performed using Pb-Bi-water direct contact boiling two-phase flow loop. For stable start-up of the boiling flow operation, Pb-Bi single-phase natural circulation must be realized in a Pb-Bi flow system of the loop before water injection into Pb-Bi. The Pb-Bi flow system consists of a four-heater-pin bundle, a chimney, an upper plenum, a level meter tank, a cooler, and an electromagnetic flow meter. A stable Pb-Bi single-phase natural circulation was realized in the range of flow rate from 1.5 l/min to 4.8 l/min by heating Pb-Bi in the heater-pin bundle with a power up to 7.7 kW. The inlet and outlet temperatures of the heater bundle were in the ranges from 243°C to 278°C, and from 251°C to 278°C, respectively. The natural circulation flow was simulated analytically using one-dimensional flow model including frictional, form and drag forces. Total hydraulic head through the loop were calculated from Pb-Bi densities at measured Pb-Bi temperatures in the loop. It was found that the calculated flow rate agreed well with the measured ones, which indicated the validity of the analytical models.     

5MW THR模拟启动实验研究

介绍了5MW THR 启动过程的实验研究结果,描述了两相流稳定性对低压自然循环反应堆启动的影响,提出了沸水启动要经过压水启动和压水向沸水转换两个过程来实现。     

Thermal-hydraulic and stress analysis of AP1000 reactor containment during LOCA in dry cooling mode

Some kinds of break in the reactor coolant system may cause the coolant to exit rapidly from the failure site,which leads to the loss of coolant accident (LOCA).In this paper,a stress analysis of an AP1000 reactor containment is performed in an LOCA,with the passive containment cooling system (PCCS) being available and not available for cooling the wall's containment.The variations in the mechanical properties of the wall's containment,including elastic modulus,strength,and stress,are analyzed using the ABAQUS code.A general two-phase model is applied for modeling thermal-hydraulic behavior inside the containment.Obtained pressure and temperature from thermal-hydraulic models are considered as boundary conditions of the ABAQUS code to obtain distributions of temperature and stress across steel shell of the containment in the accident.The results indicate that if the PCCS fails,the peak pressure inside the containment exceeds the design value.However,the stress would still be lower than the yield stress value,and no risk would threaten the integrity of the containment.     

铅铋冷却氮化物燃料小型模块化快中子反应堆堆芯物理特性分析

国际原子能机构(International Atomic Energy Agency,IAEA)认为小型模块化反应堆具有很好提高核能安全性、经济性和防止核扩散的能力,是未来核能最具发展前景的堆型之一。为适应未来核能发展的需求,提出了一种铅铋冷却氮化物燃料小型模块化反应堆(Small Modular Pb-Bi Cooled Reactor with Nitride Nuclear Fuel,SMPBN)设计方案,并利用PIJ组件计算程序和CITATION堆芯计算程序对SMPBN的物理特性和安全特性,包括反应性系数及其随燃耗变化、卸料燃耗、功率峰因子、燃料转换比和停堆余量等进行了深入分析。通过分析,认为SMPBN在20年寿期内,具有很好的燃料转换能力,不需要换料,反应性波动很小,反应性系数均为负值,具有固有安全性,符合国际上第四代反应堆的要求。     

New research programme of JAEA/CLADS to reduce the knowledge gaps revealed after an accident at Fukushima-1: introduction of boiling water reactor mock-up assembly degradation test programme

ABSTRACT

The new R&D programme of JAEA/CLADS tests complements the previous investigations related to BWR severe accidents. A series of tests aim at closing the gaps in understanding of the Fukushima Dai-Ichi degradation sequence at each unit. The paper emphasises the problem of control blade degradation, which influences the accident progression at an early stage and shows the approach for thorough investigation of this problem.     

Study on small long-life LBE cooled fast reactor with CANDLE burn-up – Part I: Steady state research

Small long-life reactor is required for some local areas. CANDLE small long-life fast reactor which does not require control rods, mining, enrichment and reprocessing plants can satisfy this demand. In a CANDLE reactor, the shapes of neutron flux, nuclide number densities and power density distributions remain constant and only shift in axial direction.

The core with 1.0 m radius, 2.0 m length can realize CANDLE burn-up with nitride (enriched N-15) natural uranium as fresh fuel. Lead–Bismuth is used as coolant.

From steady state analysis, we obtained the burn-up velocity, output power distribution, core temperature distribution, etc. The burn-up velocity is less than 1.0 cm/year that enables a long-life design easily. The core averaged discharged fuel burn-up is about 40%.     

Corrosion resistance of ceramics SiC and Si3N4 in flowing lead-bismuth eutectic

A corrosion test was conducted to investigate the corrosion characteristics of SiC and Si₃N₄ in a flowing lead-bismuth eutectic (Pb-Bi) with a fluid temperature of 550 °C, a velocity of 1 m/s, and an oxygen concentration of 1 × 10⁻⁶ wt% for an exposure time of 2000 h. The weight losses of the SiC and Si₃N₄ specimens were lower than those of corrosion-resistant steels tested under the same condition. The specimens showed excellent resistance against an element dissolution and an oxidation in the Pb-Bi flow. The surface was slightly damaged due to some stresses of the Pb-Bi flow and/or those generated by adhered Pb-Bi in the test procedure.     

Features of a control blade degradation observed in situ during severe accident conditions in boiling water reactors

In the present paper, new results using in situ video are presented regarding boiling water reactor (BWR) control blade degradation up to 1750 K at the beginning of a nuclear severe accident. Energy-dispersive X-ray spectrometry (EDS) mapping indicated stratification of the absorber blade melt with formation of a chromium and boride-enriched layer. High-content-B- and C-containing material with increased melting temperature acted like a shielding and was found to prevent further relocation of control blade claddings. The interacted layers around the B₄C-granules prevented direct steam attack of residual B₄C. The results provide new insights for understanding of the absorber blade degradation mechanism under reducing conditions specific to Fukushima Dai-Ichi Unit 2 resulting from prolonged steam starvation.     

Study on lead–bismuth droplets'' generation and their removal by an electrostatic precipitator

The applicability of the electrostatic precipitator for the removal of lead–bismuth droplets generated in the direct-contact boiling lead–bismuth cooled fast reactor is investigated. A small apparatus in which argon gas bubbles through the pool of lead–bismuth and an electrode mounted in the test section is used. The ESP operating voltage was 1000 V. It was found that the removal efficiency of the electrostatic precipitator increases with time up to 96.5%. It appears that the probability of droplet removal is almost independent of the droplet size. There is a small increase in this probability for larger droplets, which is caused likely by the fact that the larger droplets travel at lower velocities. Otherwise the effect of velocity on the removal efficiency is negligible. The electrostatic precipitator current was decreasing during the experiment, which is probably caused by the reduction of the number of droplets in the test section as the electrostatic precipitator was getting more efficient. The electrostatic precipitator current was on the order of 7 μA. The experiment demonstrated the applicability of the electrostatic precipitator for removal of lead–bismuth droplets.     

Mass flow rate sensitivity and uncertainty analysis in natural circulation boiling water reactor core from Monte Carlo simulations

Our aim was to evaluate the sensitivity and uncertainty of mass flow rate in the core on the performance of natural circulation boiling water reactor (NCBWR). This analysis was carried out through Monte Carlo simulations of sizes up to 40,000, and the size, i.e., repetition of 25,000 was considered as valid for routine applications. A simplified boiling water reactor (SBWR) was used as an application example of Monte Carlo method. The numerical code to simulate the SBWR performance considers a one-dimensional thermo-hydraulics model along with non-equilibrium thermodynamics and non-homogeneous flow approximation, one-dimensional fuel rod heat transfer. The neutron processes were simulated with a point reactor kinetics model with six groups of delayed neutrons. The sensitivity was evaluated in terms of 99% confidence intervals of the mean to understand the range of mean values that may represent the entire statistical population of performance variables. The regression analysis with mass flow rate as the predictor variable showed statistically valid linear correlations for both neutron flux and fuel temperature and quadratic relationship for the void fraction. No statistically valid correlation was observed for the total heat flux as a function of the mass flow rate although heat flux at individual nodes was positively correlated with this variable. These correlations are useful for the study, analysis and design of any NCBWR. The uncertainties were propagated as follows: for 10% change in the mass flow rate in the core, the responses for neutron power, total heat flux, average fuel temperature and average void fraction changed by 8.74%, 7.77%, 2.74% and 0.58%, respectively.     

Water level in boiling water reactors — Measurement, modelling, diagnostic

The paper deals with the hydrostatic water level measurement in connection with the application of knowledge-based and model-based methods of signal processing using Fuzzy Set Theory, and under utilisation of internal gamma radiation as well as application of Artificial Neural Networks (ANN). The utilisation of Fuzzy-Set Theory and ANN's is explained. The principle of different measuring methods are described and placed underneath with applications, like Hybrid Observers and diverse measuring system for boiling water reactors.     

Experimental study of vapor local characteristics in upward low pressure boiling tube

Radial distribution of vapor local parameters, including local void fraction, interfacial velocity, bubble size, bubble frequency and interfacial area concentration, are investigated through the measurement in an upward boiling tube using dual-sensor optical probe. In addition, a new local parameter -"local bubble number concentration" is developed on the basis of bubble frequency. The analysis shows that this parameter can reflect bubble number density in space, and has clear physical meaning.