Blockage Accident Research Articles

Thermal Hydraulic and Neutronics Coupling Analysis for Plate Type Fuel in Nuclear Reactor Core

The thermal hydraulic and neutronics coupling analysis is an important part of the high-fidelity simulation for nuclear reactor core. In this paper, a thermal hydraulic and neutronics coupling method was proposed for the plate type fuel reactor core based on the Fluent and Monte Carlo code. The coupling interface module was developed using the User Defined Function (UDF) in Fluent. The three-dimensional thermal hydraulic model and reactor core physics model were established using Fluent and Monte Carlo code for a typical plate type fuel assembly, respectively. Then, the thermal hydraulic and neutronics coupling analysis was performed using the developed coupling code. The simulation results with coupling and noncoupling analysis methods were compared to demonstrate the feasibility of coupling code, and it shows that the accuracy of the proposed coupling method is higher than that of the traditional method. Finally, the fuel assembly blockage accident was studied based on the coupling code. Under the inlet 30% blocked conditions, the maximum coolant temperature would increase around 20°C, while the maximum fuel temperature rises about 30°C. The developed coupling method provides an effective way for the plate type fuel reactor core high-fidelity analysis.

Estimating Wax Plug Transportation Force in Crude Oil Pipeline Pigging

Wax plug growth in pipeline pigging significantly impacts the crude oil transportation, even resulting in a wax blockage accident of pipeline. A pigging facility was constructed to conduct the pigg...

Numerical investigation of flow blockage accident in SFR fuel assembly

Sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declare. Due to the good thermo-physical properties of sodium, SFR has drawn a lot of attentions and there are considerable design experiences related to SFR. In the design of SFR, the pitch to diameter ratio (P/D) is kept small. Debris may accumulate in the flow channels and result in the flow blockage accident due to the corrosion of structural material. If single or several sub-channels are partially or totally occupied by debris, flow area will be suddenly reduced and a recirculation region can be expected downstream the blockage. Coolant temperature in this region can be greatly increased which may threaten the integrity of fuel assembly. In this paper, RANS method is employed to evaluate the impaired heat transfer process governed by blockages formed in a sodium-cooled 61-pin wire-wrapped rod bundle. The simulation results were firstly validated against the experimental data obtained from the literature. A good agreement demonstrated the capability of the employed approach to evaluate turbulent properties of sodium flow as well as the impaired heat transfer characteristics caused by blockages. With the validated method, the influence from blockages are systematically evaluated in which blockage size and its location are varied. Simulation results clearly indicate that a sharp increase of cladding and coolant temperature near the blockage can be expected even if the blockage area is not quite large. The local effect from blockage is also identified in the simulation. The variance of cladding and coolant temperature tends to vanish downstream the blockage and it is not feasible to detect the existence of blockage based on cladding temperature near the outlet even if the blockage area is sufficiently large.

Transient thermal-hydraulic analysis for fuel assembly blockage accident by coupling CFD code with the neutron kinetics model

Lead-based reactor coolants have many excellent characteristics but the coolants may seriously erode the fuel cladding and the structural materials and the flow area will be reduced by the corrosion products. This phenomenon is called blockage accident.In this paper, the blockage accident of single subassembly for a lead-based reactor has been analyzed by applying the commercial computational fluid dynamics (CFD) software coupled with the point reactor kinetics model. Firstly, the neutronics/thermal-hydraulics coupling model was established by user defined function (UDF) of FLUENT software based on the point reactor kinetics function with one group delayed neutron. Then the neutronics/thermal-hydraulics coupling model was validated with a single fuel pin model by comparing the fuel power, the cladding temperature and the coolant temperature with the calculation results of the RELAP5 code. Secondly, after considering mesh independence, turbulence models and other factors, the steady state of a single subassembly with 61 fuel pins at full power for a lead-based reactor designed by INEST/FDS Team was simulated in the normal operating conditions. Finally, a single subassembly for the center blockages of the flow channel was simulated by the neutronics/thermal-hydraulics coupling model. The temperature distributions and the maximum temperature were compared with that of the normal operating conditions and the effect of axial position for blockage was analyzed. It shows that the core outlet temperature and the highest temperature of cladding are increasing after blockage accident happening. And the highest temperature of cladding is lower calculated by neutronics/thermal-hydraulics coupling model than the condition without considering reactivity feedback. And the maximum cladding temperature for the 500 mm axially location blockage accident is higher.

Computational fluid dynamics and subchannel analysis of lead–bismuth eutectic-cooled fuel assembly under various blockage conditions

In recent years, extensive research has been carried out for liquid metal-cooled reactors. One of the design challenges is keeping the cladding temperature below the design limit, especially under local blockage. Thus, accurate prediction of coolant and fuel cladding temperatures is necessary for thermal-hydraulics analysis of LBE-cooled fuel assemblies. In this study, CFD and subchannel analysis were carried out to investigate the effect of various blockage accidents on the thermal-hydraulic performance of a single fuel assembly. Based on actual geometry and material properties, modelling of a single fuel assembly was established, and five blockage accidents were numerically simulated by the CFD software STAR-CCM+. Temperature variations along the axial direction of the cladding inner surface and the subchannel centre were obtained and compared. The influence of blockage parameters on the coolant reverse flow was investigated and detailed information on flow fields was obtained. The maximum temperature increase occurred inside the blockage, which influenced flow and heat transfer behaviour. Finally, a comparison of key parameters between STAR-CCM+ and COBRA-YT was conducted to show the applicability of the two codes. The results indicated that the cladding temperature during a blockage accident should be evaluated to avoid exceeding the design criteria.

Fragmentation characteristics of molten materials jet dropped into liquid sodium pool

The nuclear reactor is locally melted to form molten corium under instantaneous total blockage accidents in fuel subassembly of Sodium-cooled Fast Reactor (SFR). Molten Fuel-Coolant Interaction (MFCI) will result in violent interphase heat transfer and intensive fragmentation of molten core materials, which significantly affects the formation and cooling of core debris bed. Due to approximate ambient Weber Number (Wea) with molten metallic uranium, kilogram quantities of molten copper are used as fuel simulant to drop into liquid sodium pool to study the fragmentation characteristics during MFCI. Considering the influence of instantaneous solidification at the contact interface on fragmentation, initial temperature of molten copper is decreased to obtain the instantaneous contact interface temperature (Ti) less than the melting point (Tmp) of copper. Furthermore, hydrodynamic fragmentation is independently examined through experiments with molten aluminum jet under different initial temperatures of molten aluminum jet. The Ti between molten aluminum jet and liquid sodium is varied within the range of Tmp of aluminum and boiling point (Tbp) of liquid sodium. According to the present results, nucleate boiling of liquid sodium is proved to occur during molten copper-liquid sodium interaction and thermodynamic effects are much more predominant than hydrodynamic effects to cause finer fragmentation. The fragmentation of same molten material generally increases consistently with Ti. Due to considerable energy release capability of molten copper, molten state can be potentially maintained at the contact interface to cause adequate fragmentation under the same thermal conditions. Instantaneous solidification originating from contact interface greatly increases the criteria of fragmentation. Due to poorer thermal attributes, molten UO2 is provided with smaller variation rate of Ti under the approximate Ti, which suggests longer duration for violent sodium boiling to stimulate fragmentation. As the temperature of liquid sodium increases, interfacial instability is enhanced to fracture the solid crust with reduced thickness, contributing to fragmentation. The fragmentation caused by transition boiling seems to be more significant than that of nucleate boiling. Furthermore, the fragmentation mechanism of molten materials jet discharged into sodium pool is further discussed in the present study.

Visualization of phase distribution in lead–bismuth eutectic during one-dimensional solidification process

Solidification process of lead–bismuth eutectic (LBE) is one of the key phenomena to prevent flow channel blockage accident in an LBE-cooled accelerator-driven system. However, the solidification of liquid metal cannot be observed optically and it is difficult to detect noninvasively. In this study, the one-dimensional solidification process of the LBE was visualized by pulsed neutron transmission imaging. Neutrons have higher transmittivity to the LBE than X-ray and neutron transmission spectrum of the LBE sample can be obtained by pulsed neutron imaging technique. The solid and liquid phases of the LBE were identified during the solidification process by the presence or absence of Bragg edge in the measured neutron transmission spectrum, and the transient behavior of the solid–liquid interface could be visualized. In addition, the characteristic spatial distribution of the crystalline structure was found in Bragg-edge transmission image after the solidification.

Study of wax deposition law by cold finger device

Wax deposits in oil pipelines can reduce the effective inner diameter of such pipelines, which may lead to pipeline blockage accidents. In this study, a cold finger experimental device was constructed, and eight groups of experimental scenarios under cold flow conditions were performed in accordance with field conditions. The effects of deposition time and cold and hot bath temperatures on wax deposition were investigated. Results show that when the cold bath temperature remains unchanged, and only the hot bath temperature increases, the deposition quality and rate of mixed waxy crude oil will initially decrease and then increase. During this period, the temperature that corresponds to the maximum deposition mass is the starting point of the wax precipitation peak.

A new hydrate deposition prediction model considering hydrate shedding and decomposition in horizontal gas-dominated pipelines

Hydrate formation and blockage in gas-dominated pipelines has always been an important issue in research of pipeline flow assurance. In this work, a novel hydrate deposition prediction model considering hydrate shedding and decomposition for gas-dominated pipeline is proposed and solved by the finite difference method and iterative technique. Finally, the model has been applied in the existing gas-dominated pipeline experiments. The results show that: (a) the predicted results are in good agreement with the experimental data. (b) when the maximum degree of pipeline blockage reaches the blockage limit, the hydrate layer will shear off, and the layer that has been sheared tends to fall off twice. (c) when the pressure is the same, the higher the negative subcooling (N-subcooling), the faster the hydrate decomposition rate, and when the N-subcooling is similar to the subcooling, the decomposition with the same hydrate volume is faster than the formation. (d) increasing the hydrate shedding amount or increasing the hydrate decomposition rate can be used to deal with hydrate deposition and blockage accidents in natural gas pipelines.

CFD analysis of flow blockage phenomena in a LBE-cooled 19-pin wire-wrapped rod bundle

Flow blockage accident is a severe accident. When fuel assembly is partially or totally blocked, the heat transfer between coolant and cladding is significant impaired due to a rapid decrease of coolant flow, which leads to a local increase of cladding and coolant temperature and possibly results in a release of fission products. A recirculation region is formed downstream the blockage and a temperature peak can be found in this region. In this paper, CFD modelling and simulation for a LBE-cooled 19-pin wire-wrapped bundle were carried out to consider the influence from the flow blockage on the rod surface temperature and cross flow. Three-dimensional flow and temperature distributions of LBE were obtained using commercial CFD software. The influence from the blockage area and its location was investigated as well as the axial length of blockages. It was found that flow blockages in the center of bundle result in a higher surface temperature while comparing with blockages in the side sub-channels. The cross flow between sub-channels through gaps has been evaluated and their dependency on the blockage area is also evaluated based on CFD simulation results.

A Theoretical Model for the Rake Blockage Mitigation in Deep Cone Thickener: A Case Study of Lead-Zinc Mine in China

Deep cone thickener (DCT) is key equipment in cemented paste backfill (CPB) technology. However, rake blockage occurs frequently in DCT during the dewatering process of the unclassified tailings being thickened from dilute slurry to thickened tailings or paste. Rake blockage has disastrous effects on the CPB operation. In order to investigate the influencing factors of rake blockage in DCT, a mathematical model of rake power in DCT was developed. In addition, stacking mud bed (made of thickened tailings) from the DCT in Huize lead-zinc mine (HLZM) in different rake blockage accidents was sampled and tested to investigate the effect of tailings characters on rake blockage. Results indicated that the concentration of the mud bed and the friction between the mud bed and the cone wall contributed to the rake blockage. The concentration and friction were influenced by the high content of coarse particles in the mud bed. Moreover, activating devices for bed mud, as the corrective and preventive action, were developed to prevent the rake blockage, which was valid in HLZM.

Development and Validation of SAS4A Code and Its Application to Analyses on Severe Flow Blockage Accidents in a Sodium-Cooled Fast Reactor

Local subassembly faults (LFs) have been considered to be of greater importance in safety evaluation in sodium-cooled fast reactors (SFRs) because fuel elements were generally densely arranged in the subassemblies (SAs) in this type of reactors, and because power densities were higher compared with those in light water reactors. A hypothetical total instantaneous flow blockage (HTIB) at the coolant inlet of an SA gives most severe consequences among a variety of LFs. Although an evaluation on the consequences of HTIB using SAS4A code was performed in the past study, SAS4A code was further developed by implementing analytical model of power control system in this study. An evaluation on the consequences of HTIB in an SFR by this developed SAS4A code was also performed in this study. It was clarified by the analyses considering power control system that the reactor would be safely shut down by the reactor protection system triggered by either of 116% over power or delayed neutron detector (DND) trip signals. Therefore, the conclusion in the past study that the consequences of HTIB would be much less severe than that of unprotected-loss-of-flow (ULOF) was strongly supported by this study. Furthermore, SAS4A code was newly validated using four in-pile experiments which simulated HTIB events. The validity of SAS4A application to safety evaluation on the consequence of HTIB was further enhanced in this study. Thus, the methodology of HTIB evaluation was established in this study together with the past study and is applicable to HTIB evaluations in other SFRs.

Analysis of core blockage scenarios during pump shutdown accidents for small size lead-cooled fast reactor using RELAP5-HD

Two models of a designed small size lead-cooled fast reactor (LFR) were built by RELAP5-HD to simulate the behavior of the reactor system during pump shutdown accidents, along with different levels of core blockage. Two typical pump shutdown accidents of pump rotor seizure accident and loss-of-pump-power accident were selected to do the simulation and analysis of the behavior of the nuclear power reactor during pump shutdown accidents, when all the safety systems were assumed to be unavailable. A one dimensional vessel-one dimensional core model was built to simulate the long term cooling condition of the pump shutdown accidents under the reactor shutdown or operation conditions. The simulation results would help to analyze whether pump accidents could cause core damage. In order to show the heat removing capability of the coolant in the reactor system, a more detailed one dimensional vessel-three dimensional core model was built to simulate the long term cooling condition of the pump shutdown accidents along with different levels of core blockage accidents.

Transient Thermal-Hydraulic Analysis of Single Flow Channel Blockage in the JRR-3M 20-MW Research Reactor

A flow blockage analysis model of a single channel is established using the best-estimate code RELAP5/MOD3.4. The reactor core is divided into seven hot channels, one average channel, one bypass channel, and corresponding fuel plates to take into account the interaction between the obstructed channel and adjacent channels. The coolant system is also modeled in detail to perform a better estimation. As a typical pool-type research reactor, JRR-3M is chosen for the analysis. The results indicate that the model can effectively simulate a single-channel blockage accident using the RELAP5/MOD3.4 code. Also, the thermal-hydraulic parameters in the blocked channel would be significantly affected if bubbles are generated as the blockage ratio continues to increase, which may damage the integrity of the fuel plate. Meanwhile, as for flow blockage of a single channel, the effect on adjacent channels is limited, even under high-blockage-ratio conditions.

Sub-channel analysis for Pb-Bi-cooled direct contact boiling water fast reactor

Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR) featured with a direct-contact heat exchanger between lead-bismuth eutectic coolant and water could significantly simplify the primary system and enhance the natural circulation capability, meeting the potential needs for small modular reactor design. It is of great importance to conduct thermal-hydraulic analysis of the PBWFR core in detail. In this paper, a self-developed SUB-channel AnalysiS code SUBAS is adopted to study the thermal hydraulic characteristics of the PBWFR core. The fidelity and the reliability of the code have been preliminarily benchmarked. With SUBAS, the space grid is studied to figure out its impact on the temperature and flow distributions in each sub-channel. Besides, the application of space grids would increase the pressure drops and decreases the cross flow between adjacent sub-channels. To study the transient performance of the PBWFR core, the power transient and the inlet blockage accident are calculated by SUBAS. The results of the power transient show the cross-flow effect would be weakened in the sub-channel which has higher coolant temperature and larger mass flow rate. For the inlet blockage accident, the results indicate the influence of the small area blockage is relatively weak on the overall performance of the assembly but is significant on the local parameters. With consideration of time and space, the blockage influence only exists in a certain area. This research may provide contribution to the design of PBWFR.

Partial flow blockage analysis of the hottest fuel assembly in SNCLFR-100 reactor core

In this paper, we perform an unprotected partial flow blockage analysis of the hottest fuel assembly in the core of the SNCLFR-100 reactor, a 100 MWth modular natural circulation lead-cooled fast reactor, developed by University of Science and Technology of China. The flow blockage shall cause a degradation of the heat transfer between the fuel assembly and the coolant potentially, which can eventually result in the clad fusion. An analysis of core blockage accidents in a single assembly is of great significance for LFR. Such scenarios are investigated by using the best estimation code RELAP5. Reactivity feedback and axial power profile are considered. The cross-sectional fraction of blockage, axial position of blockage, and blockage-developing time are discussed. The cladding material failure shall be the biggest challenge and shall be a considerable threat for integrity of the fuel assembly if the cross-sectional fraction of blockage is over 94%. The blockage-developing time only affects the accident progress. The consequence will be more serious if the axial position of a sudden blockage is closer to the core outlet. The method of analysis procedure can also be applied to analyze similar transient behaviors of other fuel-type reactors.

Analysis of flow blockage accidents in rectangular fuel assembly based on CFD methodology

Flow blockage accidents in rectangular fuel assembly were investigated by three-dimensional CFD method in detail. Totally six coolant channels of the fuel assembly were modeled. Meanwhile, in order to make understand of the thermal hydraulic characteristics of the flow blockage accidents comprehensively, three coolant channels which included the obstructed channel were investigated and analyzed thoroughly based on simulation results. RELAP5 calculation was performed here to compare with CFD simulation results under non-blocked condition, and the comparison results indicated approximate agreement of such two types of results. On the basis of the CFD simulation, velocity and temperature profiles were discussed for some typical blockage cases, and conclusion was drawn that the redistribution of the mass flow rates occurred after the formation of the blockage, and due to the formation of obstruction, temperature of the coolant and the fuel increased rapidly which caused higher peak temperature in the blockage channel. Simultaneously, the increasing flow resistance would lead to the existence of jet-flow and reverse flow in the obstructed channel. In addition, DNBR calculation indicated that heat flux on the cladding surface would not exceed critical heat flux, thus DNB would not occur under the investigated situation.

Safety studies for the MYRRHA critical core with the SIMMER-III code

The presented studies are carried out within the European 7th framework project MAXSIMA, in which the MYRRHA reactor, which stands for Multi-purpose hYbrid Research Reactor for High-tech Applications, developed at SCK-CEN (Belgian Nuclear Research Centre), is investigated. The SIMMER code is employed for severe accident investigations of the reactor at KIT and SCK-CEN in both critical and ADS subcritical modes. In this paper only studies for the critical core are presented. The SIMMER-III model has been set up and assessed first for the neutronic feedback coefficients. Its calculated fuel, coolant and structure feedbacks agree well with the results evaluated by means of the European Reactor ANalysis Optimized System (ERANOS). For benchmarking of the SIMMER-III coupled neutronics and fluid-dynamics model, several Unprotected Transients due to Over Power (UTOP) have been calculated and compared with results of transient system codes. Very good agreement is demonstrated. In case of the largest and quickest reactivity insertion under hypothetical accident conditions, the reactor is assumed to turn for a short time into a slightly prompt supercritical state, but a quite mild power excursion takes place. Blockage accidents are studied in detail with SIMMER only. In total three scenarios have been investigated, namely the blockages of a single fuel assembly (FA), the protected core blockage scenario and the damage propagation of defective pin failures. Our studies demonstrated no core damage propagation can possibly occur under the different blockage scenarios.

Removal and tribological behaviors of waxy deposition layer in cleaning process

PurposeThe wax deposition in oil wells and pipelines is very viciously negative to the petroleum extraction and crude oil transportation, and it even causes severe blockage accident. This study aims to describe cleaning experiments performed on wax deposition of different deposition layer and experimental conditions to investigate the removal and tribological properties and chip formation.Design/methodology/approachAn optical arrangement was used to visually record the cleaning process, whereas the friction forces were measured by a custom-built tribometer. Various measurements were performed with tool rake angles of 45° and −30° and cleaning depths from 1 to 5 mm.FindingsResults from experiments and modeling suggest that the transition of chip was dependent on rake angle, wax performance and cleaning depth. While the cleaning depth increased, the friction and cleaning resistant force also increased. With the increase of cleaning depth, the wax layer cleaning quantity increased and the chip strengthened; hence, the curvature radius of chip was enhanced to form platy chip. The chip of wax–oil mixture was discontinuous units, and it was easy to adhere on the rake face with the increasing depth of cut. With an increase in cleaning depth, the friction and cleaning-resistant force also increased.Originality/valueIt is concluded that for effectively cleaning and stabilizing of pipeline cleaning machine, different cleaning parameters should be applied to accommodate wax layer or wax–oil mixture.

Failure analysis of multiphase flow corrosion–erosion with three-way injecting water pipe

Abstract Water injection is an essential component in a reaction effluent air cooler (REAC) system because its primary function is to dissolve the generated ammonium salt, which leads to deposition or blockage accidents. A damage incident in a three-way pipe made of carbon steel under the multiphase flow field was investigated. The failure analysis was performed by means of scanning electron microscope (SEM) inspection and computational fluid dynamics (CFD) simulation. CFD results show that a large velocity gradient exists near the area of 5–12d at the bottom of the main pipe. This gradient results in a region of low flow velocity, high wall shear stress, and high turbulent kinetic energy. The flow state becomes very chaotic, and the non-uniformity coefficient of velocity is high. The corrosive medium (NH4Cl or H2S) dissolving in water increases the causticity of fluid medium and aggravates the flow corrosion. The high risk area from the CFD simulation coincides with the breakage area of the three-way pipe on the spot. This failure incident is attributed to the flow corrosion–erosion.