Molten fuel-coolant interaction phenomena with application to carbide fuel safety

This paper reviews the major phases occurring during an energetic molten fuel/coolant interaction (MFCI), the categories of interaction and modes of contact between molten fuel and liquid coolant, the film boiling destabilization and collapse mechanisms, and the important fragmentation mechanisms of the melt. Two major models that describe the processes involved in an MFCI event are discussed: the spontaneous nucleation model and the pressure detonation model. Finally, the MFCI experiments involving carbide fuel and liquid sodium are reviewed and the potential for an energetic interaction between molten carbide fuel and liquid sodium is discussed. Recommendations are given for future work on MFCI phenomena relative to the carbide fuel/sodium system.     

The result of a wall failure in-pile experiment under the EAGLE project

The WF (wall failure) test of the EAGLE program, in which 2 kg of uranium dioxide fuel-pins were melted by nuclear heating, was successfully conducted in the IGR (Impulse Graphite Reactor) of NNC/Kazakhstan. In this test, a 3 mm-thick stainless steel (SS) wall structure was placed between fuel pins and a 10 mm-thick sodium-filled channel (sodium gap). During the transient, fuel pins were heated, which led to the formation of a fuel-steel mixture pool. Under the transient nuclear heating condition, the SS wall was strongly heated by the molten pool, leading to wall failure. The time needed for fuel penetration into the sodium-filled gap was very short (less than 1 s after the pool formation). The result suggests that molten core materials formed in hypothetical LMFBR core disruptive accidents have a certain potential to destroy SS-wall boundaries early in the accident phase, thereby providing fuel escape paths from the core region. The early establishment of such fuel escape paths is regarded as a favorable characteristic in eliminating the possibility of severe re-criticality events. A preliminary interpretation on the WF test results is presented in this paper.     

Reactor noise measurements in the safety and regulating systems of CANDU stations

Reactor noise measurements of safety and regulating system intrumentation are performed in the CANDU nuclear power stations of Ontario Power Generation (OPG) and Bruce Power. Station signals included in the noise measurements are in-core flux detectors (ICFD), ion chambers (I/C), flow transmitters, pressure transmitters, and resistance temperature detectors (RTD). Their frequency dependent noise signatures are regularly measured during steady-state operation, and are used for parameter estimation and anomaly detection.

The specific applications include the following areas:

Flux noise measurements to detect and characterize (a) anomalies of in-core flux detectors, ion chambers and their electronics, (b) mechanical vibration of fuel channels and in-core detector tubes induced by coolant/moderator flow.

Pressure and flow noise measurements to estimate the in-situ response times of flow/pressure transmitters and their sensing lines installed in the reactor's coolant loops.

Temperature noise measurements to estimate the in-situ response times of thermal-well or strap-on type RTDs installed in the reactor's coolant and moderator loops.

Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium-cooled fast reactors

The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt-through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior, including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop an evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct contact and thermal interaction of molten materials with coolant. The fragmented core materials form a sediment debris bed in the lower plenum. It is necessary to remove decay heat safely from this debris bed to achieve IVR. A simulation code to analyze the behavior of debris bed with decay heat was developed based on SIMMER-III code by implementing physical models, which simulate the interaction among solid particles in the bed. The code was validated by several experiments on the fluidization of particle bed by two-phase flow. These evaluation methodologies will serve as a basis for advanced safety assessment technology of SFRs in the future.     

Assessment of Fuel Coolant Interactions (FCIs) in the FBR Core Disruptive Accident (CDA)

Application of general behavior principles (GBPs) and consideration of relevant contact modes suggest that only incoherent small-scale fuel coolant interactions (FCIs) with negligible damage potential appear possible with the molten oxide fuel-liquid sodium system as the fuel disperses away from the core into a coolable non-critical array.

In contrast to the SPERT-1, BORAX-1 and SL-1 nuclear transients that ultimately led to energetic vapor or steam explosions, the presence of molten fuel and liquid sodium in the FBR core always requires the presence of solid cladding which separates the fuel and coolant and, hence prevents energetic FCIs prior to coolant escape.

Furthermore, unlike the CORECT-II experiments which examined dynamic re-entry of liquid sodium on molten fuel pools that resulted in unstable interfaces leading to significant sodium entrapment and relatively energetic FCIs, the prevailing contact mode in the FBR core disruptive accident (CDA) scenario is displacement of the lighter and less viscous liquid sodium by the heavier and more viscous molten fuel resulting in stable interfaces with no significant sodium entrapment and FCIs. Dynamic re-entry of liquid sodium into the core is not possible with the two-component steel vapor-liquid sodium system, since the interface contact temperature upon steel vapor condensation is well in excess of the sodium boiling temperature. A pressure reduction in the steel vapor region due to condensation is immediately compensated for by an equivalent pressure increase due to sodium evaporation.

Finally, considering that the molten oxide fuel-liquid sodium interface contact temperature is well below the sodium homogeneous nucleation temperature which in turn is well below the fuel melting temperature, not only eliminates the potential for large-scale vapor explosions as molten fuel streams are injected into liquid sodium pools, but also implies that small scale superheat explosions are possible which are consistent with the usually observed incoherent sharp pressurization events (amplitudes up to the order of 10 MPa and duration of the order of 1 ms). These general behavior characteristics are also consistent with complete fuel fragmentation with fragment sizes ranging from 100 to 1,000 μm, and the absence of significant or damaging FCIs.     

Effects of cooling channel blockage on fuel plate temperature in Tehran Research Reactor

In this study, the variation of the temperature distribution of the fuel plate in Tehran Research Reactor core was studied in case of coolant channels blockage. While the experimental method is not possible, both the analytical and simulation methods were used to obtain the more reliable data. The results show that one channel blockage will increase the fuel temperature to about 100%, but it does not lead to clad melt down still. With further calculation and simulation it is understood that if the coolant velocity drops to 90% of its nominal value, it may causes the clad melt-ing down. At least two channels with complete blockage even at the positions far from the core center can also melt down the clad.     

Experimental studies on the upward fuel discharge for elimination of severe recriticality during core-disruptive accidents in sodium-cooled fast reactors

In order to eliminate the energetic potential in the case of postulated core-disruptive accidents (CDAs) of sodium-cooled fast reactors, introduction of a fuel subassembly with an inner-duct structure (FAIDUS) has been considered. Recently, a design option of FAIDUS which leads molten fuel to upward discharge has been considered as the reference core design of the Japan Sodium-Cooled Fast Reactor (JSFR). In this study, a series of experiments which consisted of three out-of-pile tests and one in-pile test were conducted to obtain experimental knowledge of the upward discharge of molten fuel. Experimental data which showed a sequence of upward fuel discharge and effects of initial pressure conditions on upward discharge were obtained through the out-of-pile and in-pile test. Preliminary extrapolation of the present results to the supposed condition in the early phase of the CDA in the JSFR design suggests that the sufficient upward flow rate of molten fuel is expected to prevent the core melting from progressing beyond the fuel subassembly scale and that the upward discharge option will be effective in eliminating the energetic potential.     

Studies on Sodium Boiling Accidents in a Large Fast Breeder Reactor, (II) Radial Propagation of Sodium Boiling by Reactivity Effect

A computer code BORE was developed, with which analyses were performed on channel plugging accidents that would occur on a 1,500 MWe LMFBR. The BORE code calculated the dynamic characteristics of coolant boiling and fuel failure propagation radially in the core, and the requirements of core instrumentation systems were also analyzed. The results show that coolant boiling and/or fuel failure in a channel plugging accident are propagated only to a limited number of adjacent channels when sensors are installed that detect anomalies in channel flow, channel outlet temperature, boiling or reactivity. It is also concluded that the coolant void effect is not serious from the standpoint of safety when the time required for boiling propagation to adjacent channels can be made longer than 0.15 sec.     

Analysis of the MOX critical experiments BASALA by pin-by-pin core analysis method using three-dimensional direct response matrix

An analysis of the MOX critical experiments BASALA was performed to verify the pin-by-pin core analysis method using a three-dimensional direct response matrix. The BASALA experiments simulate full MOX BWR cores, and they were carried out in the EOLE critical facility of the French Atomic Energy Commission (CEA) by the Nuclear Power Engineering Corporation (NUPEC) in collaboration with CEA. The BASALA experimental cores are very heterogeneous because their size is much smaller than that of commercial power plants. The main features of the pin-by-pin core analysis method using the three-dimensional direct response matrix are that the response matrix can reflect the intra-assembly heterogeneous effect, the diffusion approximation is not involved, and the fuel rod fission rate can be directly evaluated. The maximum difference of the critical k-effective values among all nine cores analyzed was about 0.4% Δk. The root mean square differences between the calculated and measured radial fuel rod fission rate distributions in the test assembly of all cores were within 1.8% and nearly comparable to measurement error. The calculated results of the reactivity worth agreed with the measured results within 9%. These good agreements mean that the pin-by-pin core analysis method using the three-dimensional direct response matrix accurately reflects the effects of the intra- and inter-assembly heterogeneities in heterogeneous systems like the BASALA experimental cores.     

Integral boundary layer heat transfer modeling for Tehran Research Reactor

In the present work the validity of applying the Boussinesq approximation in the analysis of natural convection heat transfer along nuclear fuel plates with large coolant channel aspect ratios is evaluated. The Boussinesq approximation is introduced into the integral boundary layer equations governing the system to describe the velocity and temperature distributions of the coolant in the cooling channels. The fuel plate temperature is related to the adjacent coolant fluid temperature by a fundamental law in conduction heat transfer. Air and water are considered as fluids. The coolant flow is assumed to be fully developed which is a convenient assumption for coolant channels having large aspect ratios. Obtained results indicate that the Boussinesq approximation is merely applicable over a limited range of coolant channel outlet fluid temperatures. The use of this approximation produces conservative estimation of the critical plate power for air flow and non-conservative estimation of the critical plate power for water flow.     

Rewetting front propagation under anticipated operational occurrences for boiling water reactors – development of two-dimensional analytical model

An analytical model to predict a rewetting velocity applicable to high pressure and high flow rate condition during anticipated operational occurrences (AOOs) is developed by applying Wiener–Hopf technique coupled with appropriate kernel substitutions. The model considers the effects of enhanced cooling in the vicinity to liquid film front termed “precursory cooling” and heat input from fuel pellets on back side of wall as boundary conditions of a heat conduction equation. A simplified two-dimensional model neglecting an effect of axial heat conduction is also proposed. It is found through the comparison among the models and experimental data that the contribution of the heat conduction in the wall-depth direction is essential in the prediction of the rewetting velocity at the thermal-hydraulic condition simulating AOOs and the axial heat conduction has little influence when an enhanced heat transfer coefficient in the dried-out region is appropriately given as a function of distance from the liquid film front.     

Laminar flow induced deflections of stacked plates

Some research and power reactors such as the Engineering Test Reactor (ETR), the Materials Test Reactor (MTR) and the Shippingport Reactor have core designs which consist of parallel, flat or curved plate fuel assemblies. The fuel is contained in the thin plates which are separated by narrow channels through which coolant flows to remove heat generated within the plates. Since the plates are flexible, the coolant flowing through the channels causes the plates to deflect. At high coolant velocities large deflections have been observed causing the plates to deform plastically leading to structural failure or plate collapse. This work examines a single plate bounded by two channels and determines the static plate deflection as a function of plate, channel and flow parameters. The deflection is due to differences in pressure and flow velocity in the channels bounding the plate and also due to different channel dimensions caused by tolerance effects. The classical thin plate equations are used with a nonlinear hydrodynamic loading function expressing the external fluid forces on the plate surfaces.     

Visual study of ex-pin phenomena for SFR with metal fuel under initial phase of severe accidents by using simulants

In the present Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metal fuel within a subchannel is suggested as an inherent safety strategy in the initiating phase of a hypothetical core disruptive accident (HCDA). This safety strategy provides a negative reactivity driven by the melt dispersion; therefore, it could reduce the possibility of occurrence of a severe recriticality event. In the initiating phase, the melt could be injected into the subchannel horizontally by the internal pressure of the fuel pin. Complex phenomena occur during intermixing of the melt with the coolant after the horizontal injection of the melt. It is rather difficult to understand the several combined mechanisms that occur that are related to the dispersion and fragmentation of the melt. Thus, it seems worthwhile to study the horizontal injection of melt at lower temperatures, which could help to observe the dispersion phenomenon and understand the fragmentation mechanism. In this work, for a parametric study, tests were performed under structural conditions, coolant void conditions, and boiling conditions. As a result, in some cases, the injected molten materials were stuck around the injection hole. On the other hand, the molten materials were dispersed upward sufficiently well under the boiling condition when R123 was used as the coolant. The built-up vapor pressure was found to be one of the driving forces for the upward dispersion of the molten materials.     

钠冷快堆中熔融池模型的建立与验证

快堆在超设计基准事故下运行时,会导致钠沸腾和干涸,如果不能及时停堆,接着就会产生燃料元件的熔化坍塌,在组件盒下部形成熔融池.为了对熔融池给出合理的安全分析,采用机理建模的方法,建立了完整的熔融池模型,并在法国的SCARABEE系列实验中的BF1三种功率的实验上进行了验证,和实验吻合较好,通过和所验证过的GEYSER及BF2等实验模型进行比较,得出了有关熔融池机理的相关结论.通过排热和温升等相关数据的比较,对熔融池向外的排热形式给出了合理分析,并得出了相关结论.     

Analysis of likelihood of lower head failure and ex-vessel fuel coolant interaction energetics for AP1000

A one-dimensional model is formulated to assess the thermal response of the Westinghouse Advanced Plant (AP1000) lower head based on two bounding melt configurations. Melt Configuration I involves a stratified light metallic layer on top of a molten ceramic pool, and melt Configuration II represents the conditions that an additional heavy metal layer forms below the ceramic pool. The approach consists of the specification of initial conditions; determination of the mode, the size and the location of lower head failure based on heat transfer analyses; computer simulation of the fuel coolant interaction processes; and finally, an examination of the impact of the uncertainties in the initial conditions and the model parameters on the fuel coolant interaction energetics through a series of sensitivity calculations. The results of the calculations for melt Configuration I show that the heat flux remains below critical heat flux (CHF) in the molten oxide pool, but the heat flux in the light metal layer could exceed CHF because of the focusing effect associated with presence of the thin metallic layers. The thin metallic layers are associated with smaller quantities of the molten oxide in the lower plenum following the initial relocation into the lower head. The calculations show that the lower head failure probability due to the focusing effect of the stratified metal layer ranges from 0.04 to 0.30. On the other hand, the thermal failure of the lower head at the bottom location for melt Configuration II is assessed to be highly unlikely. Based on the in-vessel retention analysis, the base case for the ex-vessel fuel coolant interaction (FCI) is assumed to involve a side failure of the vessel involving a metallic pour into the cavity water. The FCI sensitivity calculations intended to assess the implications of the uncertainties in initial conditions and the FCI modeling parameters show that the FCI loads range from a few MPa to upward of 1000 MPa (maximum pool pressure) with corresponding impulse loads ranging from a few kPa s to a few hundred kPa s.     

Experimental investigation of solid–liquid mixtures freezing behavior in flow channels

The penetration and freezing of hot-core material mixtures through flow channels during core disruptive accidents (CDAs) within a sodium-cooled fast reactor is one of the major concerns confronting safety designers of the next-generation reactors. The main objective of this study is to investigate those fundamental characteristics of penetration and solidification involved in channeling molten metal and solid particle mixtures over cold structures. In this study, a low-melting-point alloy (viz., Bi–Sn–In alloy) and mixtures with solid particles (of copper and bronze) were used as a simulant melt, while L-shape metal (of stainless steel and brass) and stainless steel fuel pin bundle were used as cooling structures. Two series of basic experiments were performed to study the effect solid particles have on penetration and cooling behavior under various thermal conditions of melt by varying solid particle volume fraction, structure temperature and structural geometry. Melt flows and distributions were recorded using a digital video camera and subsequently analyzed. The melt penetration length into the flow channel and the proportion of melt adhesion on structural surfaces were measured. Results indicate that penetration length becomes shorter for molten-metal/solid particle mixtures (mixed melts) compared with pure molten metal (pure melt) as well as decreases with increasing solid particles volume fraction of the melt. The present study will contribute to a better understanding of the basic thermal-hydraulic phenomena of melt freezing in the presence of solid particles and to provide an experimental database for validation and improvement of the models of fast reactor safety analysis codes.     

Experimental and computational study on jet final depth under coolant jet-melt interaction

Pouring coolant into molten material provides an efficient method for cooling molten core debris in light water reactor. This coolant jet-melt interaction mode needs to be studied for proposed application and safety concern. The jet breakup pattern and its final depth are crucial factors for efficient cooling. In the present study, the hydraulic penetration behavior of coolant jet is investigated using experimental and numerical approaches. A series of visual experiments are conducted using low-density gasoline as coolant jet and high-density water as molten fuel. The images of jet behaviors and the data of gasoline jet penetration depth are obtained and analyzed. Based on FLUENT15.0 a 3D axisymmetric model is built and Volume of Fluid (VOF) method is used. The hydraulic penetration behaviors of jet and final penetration depth are correctly simulated and analyzed. The fluctuating phenomenon of penetration depth and the effects of various parameters are discussed. Jet velocity and density ratio are key factors to final penetration depth. The conclusions are helpful to understand the parameter influence and the fluctuation mechanism of final penetration depth and substantiate the understanding of the coolant jet hydraulic penetration behavior during coolant jet-melt interaction.     

熔盐球床堆堆芯燃料管理优化初步分析

核热泉高温堆是具有满功率自然循环特性的熔盐球床堆,其特有的冷却剂横向流动特性需要基于径向分区换料的燃料循环方案,以及相应的燃料管理优化分析方法。局部搜索算法和模拟退火算法是普遍应用于轻水堆换料优化问题的随机性优化方法。本文应用这两种方法对核热泉高温堆的换料优化问题进行了分析。分析结果表明,相比于局部搜索算法,模拟退火算法能以较高的概率跳出局部最优陷阱,获得全局最优的优化结果,且优化质量独立于初始解的选取。最终的优化结果给出了较好的堆芯功率展平与压制燃料中心温度的效果。针对核热泉高温堆的一维分区换料方式,模拟退火算法是非常合适的优化分析方法。     

An on-line fuel management method using in-core detector readings for CANDU reactors

An on-line fuel management method for a CANDU reactor has been developed. In the method, the in-core detector readings are used for channel power generation for refueling channel selection. The in-core detector readings are converted to measured mesh readings, and the Kalman filtering technique is applied to reduce calculation and measurement errors of the mesh readings. Then, the estimated channel powers are fed into the refueling channel selection process, in which the channels are refueled so that the difference of zone power from the reference one is minimized. The performance of the method has been demonstrated against the operating data of CANDU 6 reactor. Also, it is found that the core tracking fuel management could be implemented, so that the proposed method would contribute to economic and safe operation of the reactor.     

A parametric study of multichannel molten cladding motion under unprotected loss-of-flow accident conditions in LMFBR's

One of the key assumptions of the present multichannel clad motion model was that the total pressure drop over the voided channel could be supplied as a boundary condition. The incoherency effect on cladding motion can be significant for a full-scale subassembly, and therefore parametric studies of the total pressure drop and oscillatory pressure effect due to sodium chugging were examined using the multichannel model.There is an axial blanket region in demonstration plant or commercial-power-plant designs instead of a reflector in FFTF design above the top of fuel. It was shown that due to the difference in the thermal conductivities between the blanket material and reflector, significant changes in the timings of various events of the cladding relocation might occur. It is also noted that depending on the effect of the sodium voiding on the reactivity, the fuel may become molten when the molten cladding is still around. The possibility of the occurrence of this situation is studied by increasing the power in the present model.