Rod Temperature Research Articles

Rod bundle cooling by a spray system in spent fuel pool accident conditions

The knowledge of the phenomenology of loss of coolant/cooling scenarios in spent fuel pools is of prime interest since a huge amount of fission product can be released in accident conditions. In this context, the ASPIC facility is dedicated to the study of the heat transfers that occur at the scale of one rod bundle stored in a spent fuel pool cell. Investigations have been carried out on a full high assembly composed of 17 × 17 electrically heated rods. The water level in the cell is controlled and set, in this test series, so that about one meter of the assembly is uncovered. Therefore, the upper part of the assembly gets overheated unless some counter measure is undertaken. In this contribution, the ability of a spray system to limit the temperature rise of the uncovered part of the rod bundle is assessed. It has been observed that the water spray reduces the temperature level on the whole cross section of the assembly and lowers significantly the risk of degradation of the rod bundle. Besides, for a bundle heat power of 20kW, a spray mass flow rate of 10 gs−1 or higher leads to rod temperatures lower than 480°C.

MC/sub-channel coupling for steady state and transient simulation of Xi’an Pulsed Reactor

The Monte Carlo (MC) code RMC and sub-channel code CTF are coupled for Xi’an Pulsed Reactor (XAPR) pulse and depletion simulation. A python script is developed to handle data exchange through files. $3.45 pulse and $2 pulse are simulated with the pulsed state core. The pulse peak power and the Full Width at Half-Maximum (FWHM) results are compared with experiments as a validation and good agreement is achieved. Detailed 3-D power and temperature distributions are also obtained. Results show that the core power peak is close to the pulse rod where the reactivity is introduced. The radial power peak of each rod is at the boundary because of the self-shielding effect. The rod temperature distribution follows the same trend with the power, and the coolant temperature is not changed during the pulse period of about 0.12 s, which suggests that the heat transfer plays a negligible role in such a short time. For the steady state core, depletion simulation is carried out for the lifetime of the first fuel cycle of 120 Effective Full Power Days (EFPD). Validation is done by calculating the temperature distribution and the differential worth of the regulating rod at 0 EFPD, which both agree well with experiments. Results show that the power distribution is almost unchanged over time, only slightly more flat, indicating the material is not greatly depleted. The temperature distribution of the core generally agrees with the power distribution, except the radial temperature peak of each rod is at the center because heat is conducted outwards at steady state. Detailed coolant temperature distribution is obtained thanks to the utilization of CTF. Temperatures at the assembly boundary and near the central water chamber is noticeably lower than the other part, which is not shown in the parallel multi-channel models.

COOLING OF CONTROL RODS AND SPECIFICATION OF CRITICAL TEMPERATURES TO JUSTIFY SAFE OPERATION

The paper analyzes the cooling conditions of the control rods in the mixed core of the WWER-1000 reactor containing fuel assemblies of different vendors (TVSA, WFA/RWFA). The coolant heat-up in the guide thimbles due to own energy release in control rod materials and control rod temperatures required to justify safe operation were calculated for the most conservative case, when the core consists only of TVSAs with a standardized guide thimble inlet plug of NCCP. The energy release was obtained using the calculation method for the boron carbide absorber (B4C), dysprosium titanate (Dy2O3·TiO2), control rod cladding (42CrNiMo), water gap between the cladding and the guide thimble and in the guide thimble (zirconium alloy), fully immersed and 70% immersed RCCA 10th control group. The highest calculated energy release will be observed in fully immersed control rods with boron carbide material at the height position corresponding to the zone of maximum neutron fluence and will be 43.8 W/cm3. The maximum energy release of the 70% immersed (from the bottom of the core) 10th control group is 29.4 W/cm3 for fresh B4C. For the condition of the reactor facility at which the temperatures in FA guide thimbles are maximum, for a "fresh" control rod characterized by maximum energy release in the neutron-absorbing material, the maximum coolant heat-up in the guide thimbles does not exceed 28.9 °C, with the maximum coolant heat-up in FAs being 44 °C, and the maximum control rod cladding temperature not exceeding 335.6 °C at a coolant saturation temperature of 345.8 °C, which indicates that there will be no coolant boiling in the guide thimbles under operation. The maximum temperature typical for the central part of the control rod absorber and for the most conservative calculation for a "fresh" absorber is 529.6 °C, which is significantly lower the melting point of boron carbide, which is ~2350 °C. The maximum calculated temperature in the contact zone of the B4C absorber with control rod cladding is 348 °C. The performed calculations and further analysis of the results confirm that during operation of the RCCA control rods with both “fresh” and “burned-up” B4C, there will be no boiling of the coolant and melting of the absorber in the central part of the control rod. This means that the VVER-1000 core with TVSAs and WFAs/RWFAs and the RCCA rods are reliably cooled.

Optimized Design of Downhole Heater Seal for Oil Shale In-Situ Heat Injection

Sealing is an important prerequisite for downhole heater work. This paper proposes a combination of soft and hard and welding sealing programs, which were analyzed using theoretical calculations, numerical simulation, and in-situ testing. The results show that 316 stainless steel can meet the stuffing seal requirements; The first stuffing compresses and gradually reduces, while the second stuffing essentially does not deform. Stuffing deformation to fill the gap in the sealing hole, resulting in a sealing layer. The compression rate is 0.43%, 8.45%, and 12.64%, indicating that the locking stress should be more than 2000 N; The temperature at the weld is heated by heat conduction and distributed in a concentric circle. Thermal stress will influence the 50mm barrier, but the 100mm boundary will be mostly unaffected. Actually, the thermal stress that destroys the weld seal may be reduced by adjusting the heater's output or raising the gas injection rate. During the beginning of the in-situ heat injection, the temperature of the heating rods rises simultaneously with the outlet temperature. As consequently, the two show opposite tendencies. The heat generated by the heating rods will cause the injected gas to be preheated in advance.

Flow mixing and heat transfer deterioration investigation of supercritical carbon dioxide in wire-wrapped bundles

As one selection of the fourth-generation reactors, the gas cooled reactor has received enormous attention such as high power conversion efficiency and simple cycle layout. In this paper, the flow mixing and heat transfer characteristics of supercritical carbon dioxide within a seven-pin wire-wrapped bundle were investigated. Due to the lack of more experiments data of velocity field, two experiments were selected to validate the uncertainty of different turbulence models from the view of temperature field. The analysis of mass exchange between different subchannels revealed that coolant flows consistently in a certain direction in edge and corner subchannels, whereas in center subchannels, the cross flow changes direction periodically. The mixing coefficients at three types of subchannel interfaces were fitted by a sine function with the R2 values are all greater than 0.96. For different mass flux and pressures, the root mean square error between correlations and simulations was less than 1 × 10-3, except when Re equals 15000. Additionally, the heat transfer deterioration of SCO2 was observed. While wire spacers can mitigate the extent of heat transfer deterioration, they cannot completely eliminate it. The Performance Evaluation Criterion (PEC) was introduced to assess the effectiveness of wire spacers in enhancing heat transfer. The results indicate that the PEC exceed unity, meaning that the enhanced heat transfer capability outweighs the additional resistance loss caused by wire spacers. With increasing pressure, the PEC decreases but remains greater than unity. The mixing of coolant between different subchannel induced by wire spacers can reduce the wall temperature of rods at the edge of the bundle. However, the seven-pin wire-wrapped bundle was investigated in this work, and the results may not accurately reflect the flow and heat transfer inside a real reactor rod bundle. In the future, a larger scale rod bundle should be conducted to obtain more universally applicable and realistic flow and heat transfer phenomenon for reactor assembly design.

Understanding post-irradiation creep behavior of M5Framatome zirconium alloy

After in reactor use, the nuclear fuel assembly of pressurized water reactors is stored in water and is eventually transported, in a dry cask, to a long-term storage facility or to a reprocessing plant. During transportation in dry environment the fuel rod temperature and internal pressure both increase. This leads to simultaneous creep deformation and annealing of the cladding tubes made of M5Framatome11M5 and M5Framatome are trademarks or registered trademarks of Framatome or its affiliates, in the USA or other countries. zirconium alloy. This study shows that post-irradiation creep, or heat treatment, leads to the recovery of the radiation induced hardening and also to the recovery of the uniform elongation because of <a>-loop annealing. On the other hand, it is shown that cladding tubes made of M5Framatome alloy exhibit low post-irradiation creep rates, especially at low applied stress. This strain rate is significantly lower than for the unirradiated material and similar to the as-irradiated material, despite the significant <a> loop annealing. The low creep rates are mainly attributed to the Nb-rich nano-precipitates that act as dispersed strengthening particles. These nano-precipitates appear under irradiation and are stable at temperatures up to 450 °C explaining their strengthening effect even after significant annealing.

Uncertainty calculation of thermal hydraulic simulation of DEC-A accident scenario at VVER-1000 NPP using CIAU method

All thermal hydraulic best estimate simulations of transient scenarios at Nuclear Power Plants (NPPs) are subject to uncertainties. One method to evaluate those uncertainties is implemented by the ”Code with the capability of Internal Assessment of Uncertainty” (CIAU-Code), which calculates the uncertainty for the parameters primary system mass, primary system pressure and hot rod temperature by accuracy extrapolation.The present work now presents an extended use of the CIAU method, by deriving the uncertainty of the parameter iodine-131 release to the environment. For the analysis, a steam generator hot header break with the assumption of a stuck open safety relief valve (BRU-A) at the first opening at the VVER-1000 reactor was selected. Such transient constitutes a design extension condition A (DEC-A), characterized by multiple failures of safety systems but with the reactor core remaining intact. The thermal hydraulic best estimate simulation was conducted with RELAP5-3D. Additionally, the Iodine Spiking (IS) phenomenon for the examined accident scenario was simulated using two different empirical IS models (NRC IS Model and BOKU University IS Model). To allow the application of the CIAU method for the IS calculation a theoretical derivation is provided.The results show that for the selected DEC-A scenario a high uncertainty range has to be expected. The analysis allows a conservative estimation of the risk expected in the event of the examined transient sequence.

Fuel fragmentation and relocation (FFR) model in SPACE code PART 1: Development and verification

Ballooning of cladding, which can happen during a loss of coolant accident (LOCA), causes relocation of fragmented fuel. Recently, consideration of fuel fragmentation, relocation, and dispersal (FFRD) phenomena for safety analyses has been requested by a regulatory body. Therefore, new models related to the fuel fragmentation, relocation, and dispersal (FFRD) in SPACE code have been developed. In this study, the FFR model, excluding dispersal, will be discussed. The fuel fragmentation and relocation models, which were developed by Quantum Technology (QT), are modified and developed in SPACE code. In particular, the fuel relocation limitation threshold parameter and the fuel power distribution during relocation are newly developed. The FFR models are also verified with simple known problems. The newly developed and implemented FFR models in SPACE are validated with calculations using Halden IFA-650 test data. The newly developed FFR model in SPACE can reasonably predict the fuel relocation and axial rod temperature distribution due to fuel mass redistribution. However, intensive validation tests including various conditions are necessary to validate the new FFR model in SPACE code.

Research on deformation and temperature characteristics of micro negative Poisson's ratio anchor rod

This study aims to investigate the global deformation, temperature characteristics, and their correlation during the quasi-static tensile process of micro-NPR anchor rods. The LEW-500 static tensile testing system was used to carry out the uniaxial tensile testing with three deformation rates, and the deformation and temperature of the anchor rods were studied through the DIC system and IRT system. The mechanical results indicate that the micro-NPR anchor rods have the characteristics of high elongation, high constant resistance, no yielding platform, and no obvious necking. The DIC results show that the deformation of anchor rods can be divided into the initial and accelerated differentiation stages. In the initial differentiation stage, the statistical dispersion of deformation is low and increases slowly with time. In the accelerated differentiation stage, a relatively concentrated local deformation appears, shown by the sharp rise and decline of statistical dispersion. The IRT results indicate that the surface temperature of the anchor rods also increases with deformation, and the upward trend exhibits distinct characteristics in two differentiation stages. The faster the deformation rate, the higher the temperature of the anchor rods’ surface. The correlation coefficient of axial strain and temperature difference in the anchor rods is calculated, and the correlation coefficient is maintained at a high level in the accelerated differentiation stage. The linear fitting of the standard deviation of axial strain and the temperature difference is perfect. This study provides a foundation and a new idea for future deformation patterns and monitoring analysis in on-site applications of micro-NPR anchor rods through the joint analysis of two non-contact measurement methods.

Numerical simulation of transient boiling and void cross flow in non-uniformly heated 5 × 5 rod bundle

A multi-dimensional numerical simulation has been conducted for a transient boiling experiment by applying rapid and non-uniform power to fluid in a 5×5 bundle channel simulated a BWR fuel assembly in atmospheric pressure. The experiment is to investigate thermal hydraulic behavior in applying transient power to subcooled reactor coolant in a stand-by mode of a nuclear power plant such as reactivity initiated accident (RIA). The numerical simulation has been conducted by TRACE (version 5.0 / patch 5) which is a nuclear system analysis code to solve thermal hydraulics of boiling two phase flow by a two fluid model. The bundle channel in the upward vertical direction has been simulated by a three dimensional component with the Cartesian coordinate system attaching heat structures to simulate heating rods for thermal conduction and heat transfer calculation. It has been revealed that the numerical result is possible to simulate the void cross flow through the sub-channel in the bundle channel qualitatively as observed in the experiment. However, there was room for improvement for a wall heat transfer model in TRACE because the numerical result is larger than the experimental result in terms of the rising rate of the rod temperature during the rapid initial increasing of applied power, and it has been inferred that the evaluation model for an onset of nucleate boiling in TRACE also has to be improved because the void initiation time in the numerical result was earlier than the experimental result. Therefore, TRACE has been optimized by modifying the two physical models described above. The numerical results with the modified TRACE have been in better agreement with the experimental results than those with the original TRACE. Using the modified TRACE, a numerical experiment has been conducted putting an inlet velocity and subcooling of fluid and applied power as parameters in addition to the experimental conditions. The results have shown that there is a square law correlation between a void cross flow and applied power to the fluid. The experiments and analyses in this study have elucidated the behavior of transient boiling two phase flow in a fuel assembly of a nuclear power plant at the transient event such as RIA.

Performance analysis of irradiated thermo-mechanical coupled petal-shaped fuel rod under different power and heat transfer conditions

As a new type of fuel rod, petal-shaped fuel rod (PSFR) has the advantages of lowering the core temperature and increasing the power density, and thus has a broad application prospect. Based on the thermal elastic–plastic theory, this paper establishes an analytical model for irradiated thermal–mechanical coupling performance of PSFR by user subroutine. Through numerical simulations of the full-cycle operation of PSFR under different power and heat transfer conditions, the performance parameters of fuel rods, such as fuel rod temperature and stress, are obtained, and their spatial distribution and the variation pattern with the operation time are analyzed. The results show that increasing line power leads to a simultaneous increase in fuel rod temperature, stress, maximum deformation, and creep. However, an appropriate reduction in heat transfer coefficient decreases the cladding stresses. The above research results will provide an essential theoretical basis for determining new PSFR’s deformation and operation risk.

Experimental analysis of a power-to-heat storage with high-temperature phase change materials to increase flexibility and sector coupling

To boost the flexibility, sector coupling and manageability of renewable energy systems, a unique power-to-heat storage (electric charging, thermal discharging) is proposed. The hybrid thermal energy storage system, including phase change materials, is built using flat pillow-plates and heating rods. Experimental testing is conducted to assess the prototype's electrical and thermal performance. In addition, a parametric study involving several charging and discharging control strategies is proposed in this context.The suggested hybrid thermal storage system provides a total storage capacity of 4.87kWh using nitrate salts as phase-change material (eutectic mixture of KNO3 and NaNO3). The charging efficiency ranges from 65 to 90%, depending on the charging/discharging strategy, and the discharging period can be shortened by more than 1 h. Additionally, active management of the heating rod temperature reduces the charging period by half, making the storage system highly flexible. This demonstrates the ability of this system to aid in modularity and flexibility for effective energy management in the context of renewable energy and industry thanks to an increased storage capacity.

Extended Development of a Fission Gas Release Behavior Model Inside Spherical Fuel Grains for LWR Reactors

Fission gas plays a significant role in fuel rod performance following accidents. The amount of fission gas increases dramatically under accidental conditions. This leads to a subsequent rise in the fuel rod internal pressure and temperature due to aggravated gap conductance between the fuel pellet and cladding. As a result, fuel rod performance degrades. Therefore, studying fission gas behavior is crucial for accident assessment and evaluating fuel rod performance. Minimizing the Impact of fission gas on fuel rods is essential for maintaining their integrity and safety within nuclear reactors. One important aspect of ensuring safety is predicting fission gas release (FGR). In this study, we presented an extended model to be used in light water reactors (LWRs). The FGR can be modeled using COMSOL Multiphysics with the finite element method. This modeling approach considers both normal and abnormal conditions, with the latter categorized as Class-II type incidents. The model assumes that the gas diffusion inside a spherical grain varies over time. By examining perfect sinks with gas production, perfect sinks without gas production, and imperfect sinks under steady-state conditions, different initial and boundary conditions are set. To validate the accuracy and universality of expressions used in the model, input parameters from other models and experiments are utilized. By comparing the model’s results with these inputs, the accuracy and applicability of the expressions can be confirmed. This validation process ensures that the model provides reliable predictions for fission gas behavior in fuel rods under both normal and abnormal operating conditions. Based on our findings, it is evident that the FGR fraction displays an upward trend as diffusion coefficients and temperatures rise. Conversely, larger grain sizes and higher linear heat generation rates are associated with a reduction in the FGR fraction. Notably, enhanced resolution leads to a postponed onset of FGR. Furthermore, the influence of the diffusion coefficient on the FGR fraction primarily stems from the interconnected effects of temperature and linear heat generation rate.

Investigations of Multiphysics Models on a Megawatt-Level Heat Pipe Nuclear Reactor Based on High-Fidelity Approaches

This work describes the research of high-fidelity multiphysics models for the MegaPower nuclear reactor, a megawatt-level heat pipe reactor. Combining the Monte Carlo neutronics model, the heat pipe analysis model, the fuel analysis model, and the thermoelasticity model produces the Multi-Physics Coupling code for Heat pipe nuclear reactors (MPCH) code platform. Using the heat pipe analysis model, a database of heat pipes is generated to save computing costs. Comparison is made among four calculating modes with differing degrees of coupling. It was discovered that the thermal expansion effect reduces core reactivity by 537 ± 11 pcm and the temperature feedback coefficient by 61%. With the incorporation of the heat pipe module, a temperature difference arises between the wall of heat pipes, which can reach a maximum value of 80 K at steady state. Simultaneously, the global fuel rod temperature difference increases from 34 K (under the assumption of uniform heat pipe wall temperature) to 93 K, and the monolith temperature variance increases from 34 to 108 K. At the periphery of the monolith, the increased temperature variation causes a monolithic stress of 188.6 MPa. To further investigate the safety of the reactor, three-heat-pipe-failure scenarios are evaluated. The heat pipe analysis model reveals that a single heat pipe failure results in a monolith peak temperature of 1046 K, giving a maximum monolith stress of 237 MPa. The maximum monolith stresses and temperatures for the two-heat-pipe-failure scenario and the three-heat pipe-failure scenario are 330 MPa/1128 K and 471 MPa/1233 K, respectively. In steady-state operation, the stresses exceed the yield tensile strength (131MPa) whereas those generated by the failure of three heat pipes exceed the ultimate tensile strength (345 MPa) in high temperature. These results illustrate the necessity of including coupled multiphysics models into the design and safety evaluation of innovative nuclear reactors.

Instability Behaviors and Suppression of the Unsteady Autoignited Turbulent Jet Flame in Hot Coflow

ABSTRACT Autoignition widely exists in combustion devices that feature hot air. The unsteady autoignited flames caused by autoignition are potential hazards to the stable operation of combustion systems, often causing flameout or damage to the combustion devices. In this paper, we present our recent work in understanding the key factors determining the instability of autoignited turbulent jet flames and a novel method to suppress flame instability. Firstly, we studied the flame using a jet-in-hot-coflow burner. The instability behaviors were characterized using a pressure probe and a high-speed camera. When the fuel mole fraction was below a critical value, the amplitude spectrums of the unsteady autoignited flame had low-frequency and high-frequency peaks. The experimental results show that the low-frequency peak is related to the autoignition intermittency and can be eliminated by enhancing chemical reactions. The high-frequency peak is related to the autoignition frequency, and the peak amplitude can be reduced by improving reactants mixing. Considering that the heating rod can enhance the reaction as a heat source and improve mixing as a bluff body, we inserted an electric heating rod into the autoignition spatial region to suppress the flame instability. The suppression of flame instability by heating rod at different temperatures and locations were experimentally studied. The heating rod insertion can significantly reduce low-frequency and high-frequency instability. Moreover, increasing the rod temperature can also effectively reduce the amplitude of high-frequency pressure pulsation. The results show that the electric heating rod is an efficient method to suppress the instability of autoignited flame.

Temperature field simulation of HTHP diamond synthesis cavity stacked with sheet alloy catalyst and graphite

In this study, the thermodynamics and heat conduction of the diamond synthesis cavity of flake catalyst were theoretically analyzed and mathematically deduced to discuss the temperature field and the distribution of pyrophyllite synthesis cavity in flake catalyst-graphite's high-pressure and high-temperature (HTHP) synthesis. The thermal flux and temperature distribution of the pyrophyllite synthesis cavity and the temperature field of the flake catalyst-graphite synthesis rod were simulated and calculated. The calculations revealed that the synthetic rod's temperature rapidly increased in the initial 100 s of heating; When the heating time increased from 100 s to 150 s, the heating rate of the synthetic rod started to slow down. The temperature of the synthetic rod was stable after 170 s of heating, and the whole synthetic rod system reached the thermal equilibrium state; the temperature field of the synthetic rod was also stable. In a steady state, the maximum radial temperature gradient of the sheet catalyst graphite synthetic rod is 0.54 °C/mm, and the axial temperature gradient is about 2.93 °C/mm. The axial temperature gradient inside the synthetic rod was significantly higher than the radial temperature gradient. Further calculations showed that the heat flux of the pyrophyllite synthetic block on the inner wall of the pyrophyllite synthetic cavity was 2734 W in the heating direction, which was greater than that in the non-heating direction around 1312 W. The heat flux of the conductive steel cap in the heating direction was 2436 W, accounting for 60.1% of the total heat flux; This is the main reason for the large axial temperature gradient of the lamellar catalyst-graphite synthesis rod in the pyrophyllite cavity. A comparison of the temperature field of the cavity under different powers revealed that the temperature inside the cavity produces excessive diamond growth wasteland when the power is too high or too low.

The Long-Term Interfacial Evolution and Prediction of Carbon- and Glass-Fiber-Reinforced Epoxy Hybrid Rods under a Hygrothermal Environment.

In order to promote the engineering applications of carbon- and glass-fiber-reinforced epoxy hybrid rods, it is necessary to fully understand its long-term hygrothermal durability. In the present study, the water absorption behaviors of a hybrid rod in a water immersion environment are studied experimentally, the degradation rules of the mechanical properties are obtained, and establishing a life prediction model is attempted. The water absorption of the hybrid rod confirms to the classical Fick's diffusion model, and the water absorption concentration is determined by radial position, immersion temperature, and immersion time. In addition, the radial position of water molecules diffused into the rod is positively correlated with the diffusion concentration. The short-beam shear strength of the hybrid rod decreased significantly after 360 days of exposure; this is because water molecules interact with the polymer through hydrogen bonds to produce bound water during the immersion process, leading to resin matrix hydrolysis and plasticization, as well as interfacial debonding. In addition, the ingression of water molecules caused degradation in the viscoelastic behavior of the resin matrix in hybrid rods. The glass transition temperature of hybrid rods decreased by 17.4% after exposure at 80 °C for 360 days. The Arrhenius equation was used calculate the long-term life of short-beam shear strength in the actual service temperature based on the time-temperature equivalence theory. The stable strength retention for SBSS was found to be 69.38%, which is a useful durability design parameter for hybrid rods in civil engineering structures.

CFD analysis of porous flow blockage in a gas-lift enhanced LBE-cooled fuel assembly

Porous blockage is likely to occur in wire-wrapped fuel assemblies and cause severe consequences such as overheating and mass flow decrease. In this study, flow behavior upstream, inside and downstream of the blockage is obtained for different porosities to analyse the overheating and pressure change. Additionally, the effect of reactivity feedback considering the void fraction and radial and axial power distributions on LBE-cooled porous blockage are given. It is concluded that flow separation upstream and recirculation downstream is the main reason for the overheat phenomenon due to poor convective heat transfer. Static pressure has a local rise both upstream with small porosities and in the flow recovery region downstream, while the gradient of pressure drops inside the blockage is large and begins forwards for small porosities because of flow resistance. The study of reactivity insertion reveals that although the effect of the void fraction on the reactivity and heat power is straightforward, the volume-averaged coolant and rod temperatures are virtually unaffected within 0.02 s. The wall and bulk temperature of central blockage C1 is generally higher than that of edge blockage E1.

Safety analysis of VVER fuel rod modelling in reactivity initiated accident

The paper briefly reviews the behaviour of WWER type fuel rods studied under conditions of a high rate of power rating of fuel; the investigations were carried out to validate the licensing criteria with the aim of assessing design basis reactivity initiated accidents (RIA).The procedure used to calculate thermomechanical and corrosion behaviour of fuel rods in accidents is described.The results acquired from the analysis of the thermomechanical condition of WWER-1000 fuel rods in a design basis reactivity initiated accident caused by an ejection of a regulation element of the control and protection system (RE CPS) are discussed.The aim of the analysis is to validate the fulfillment of specific safety requirements imposed on fuel rod condition in design basis accidents.Consideration is given to accidents attended with a rise of fuel rod (fuel and/or cladding) temperature in comparison to the state under normal operating conditions. This may proceed from deterioration of heat transfer from fuels, e.g., in accidents due to loss of the coolant by the primary circuit or an increase in power rating of fuel in RIA.Design procedures have been worked out to simulate fuel rod behaviour in accidents, which were implemented in the RAPTA-5 fuel performance code. Using the latter, design analysis of fuel rod behaviour under design RIA conditions is implemented; its results serve to draw conclusions whether acceptance criteria are fulfilled.The results of the RAPTA-5 code design analysis of the thermomechanical state of WWER-1000 fuel rods in that accident have demonstrated the fulfillment of all the acceptance criteria.

Modeling and program development of supercritical CO2 sub-channel based on Modelica language

The application of supercritical carbon dioxide (S-CO2) in small nuclear reactors is a new research direction with promising prospects. Most of the research on S-CO2 cooled reactors focuses on the design and optimization of core parameters, but few on detailed thermal and hydraulic analysis of the reactor interior using the sub-channel method. To supplement the lack of research in S-CO2 small nuclear reactors, this paper developed an S-CO2 cooled reactor two-dimensional sub-channel analysis model based on Modelica language. To verify the correctness of the model and program, the calculated results of the sub-channel program are compared with the results of KAIST MMR in Korea, the power deviation is less than 4%, the coolant temperature deviation is less than 7%, and the fuel rod temperature deviation is less than 10%. Based on the developed sub-channel model, the transient analysis is carried out, including reactor power increase, reactivity insertion, and coolant flow loss accidents, and then the safety performance of the reactor is evaluated. The transient response of the reactor under accident conditions is obtained, and the self-stability and self-adjustment characteristics are proved. This paper relies on theoretical modeling and program development methods, studying the sub-channel thermal-hydraulic characteristics of steady-state and transient conditions, which provides theoretical reference and program support for the study of the S-CO2 small nuclear reactor.