Centerline Temperature Research Articles

In-pile thermal conductivity of uranium dioxide at low burnup

In-pile thermal conductivity of uranium dioxide (UO2) was investigated at low burnup levels (<0.2 kWd/kg-UO2) to elucidate the concurrent effects of the fission damage and the thermal recovery. In-pile experiments were performed in the Halden reactor. The fuel centerline and the cladding temperatures were measured during the experiments, and the power level of specially designed test rodlets were also monitored. The uncertainty of the pellet-cladding gap's thermal resistivity was minimized with a liquid-metal bond, while comparable measurements were made in reference test rodlets without liquid metal bonding. The experimental data were analyzed using the inverse heat transfer approach. The heat conduction equation was solved by applying the measured temperature of the cladding as a boundary condition, and the calculated fuel centerline temperature was compared with the measured fuel centerline temperature to determine the relative matrix resistivity. Results showed that the in-pile matrix resistivity, the phonon-lattice interaction terms of the thermal conductivity, was ∼1.5–2.5 times higher than its unirradiated value due to fission-induced damage. Furthermore, the in-pile annealing caused a significant recovery of the matrix resistivity, and the amount of recovery increased from ∼20 to ∼33%, while the annealing temperature increased from 700 to 1,000 °C.

The development of TM2-RIA code for TRIGA type research reactors

This work presents the development of indigenous code TM2-RIA for steady-state and transient analyses of TRIGA type research reactor at Istanbul Technical University Energy Institute. The 2-dimensional code adapts piecewise constant function for solving point kinetics equation linked with thermal–hydraulic equations for cylindrical fuel type. TM2-RIA provides flexibility of implementing different delayed neutron parameters, it also offers different thermal-hydraulics structure options (number of channels and control volumes). The code is tested against the experimental data and calculations of EUREKA-2/RR code. Considering many experimental uncertainties, the TM2-RIA 5-channels model agrees acceptably with the steady-state experimental results. In the transient analyses, three pulsing scenarios are investigated, based on the results, the code has performed differently with different neutronic sets and thermal-hydraulics structures. That indicates how the reactor kinetics is being influenced by the differences in the recorded thermal history, and it shouldn’t be neglected even for small deviation. Besides, the constant overestimation of the fuel temperatures implies the need for modelling the heat sink induced by central Zirconium rod in fuel element and investigating its effect. That being said, the code agrees fairly well in many cases corresponding to those changes, when it predicts the peak power and fuel centerline temperature. Moreover, the gap between the experimental and the numerical results for the fuel temperature gets smaller with the increase of the pulsing reactivity. Another promising aspect is that the adapted point kinetics approach shows similar performance, in comparison with the common Cohen’s approach used by EUREKA-2/RR, while providing relative simplicity in developing and implementing the TM2-RIA code.

Coupled Monte Carlo-burnup and CFD analysis of coated UN and UC fuels in an HPLWR fuel assembly

Ceramic nuclear fuels may have potential benefits to be considered as nuclear fuels in future Supercritical Water Reactors (SCWR), they offer the advantage of high thermal conductivity and high-density values compared to uranium dioxide. In recent years, uranium nitride (UN) and uranium carbide (UC) are being considered as viable options for alternative nuclear fuels in SCWR, but an important issue is related to their chemical reactivity with water and nickel, which forces the use of coatings for fuel pellets, reducing heavy metal volume and reducing their thermal and neutronic performance. The main purpose of the study is the analysis of the burnup dependent performance of coated ceramic fuels in the SCWR compared with conventional UO2 fuel. Particularly, it focuses on the study of the fuel centerline temperature and the cladding surface temperature. Furthermore, the evolution of the neutron multiplication factor, the fissile inventory ratios and the main heavy isotopes composition along burnup are analyzed and compared to asses proliferation resistance issues. The use of UN fuel coated with Zirconium Carbide (ZrC) and UC fuel coated with Titanium Nitride (TiN) in the fuel assembly of the High-Performance Light Water Reactor is analyzed, and the results are compared with those of the conventional UO2 fuel. The comparative study was conducted for a hypothetic operation time of 540 days. Due to the large coolant density variation along with the active height of the fuel assembly, coupled calculations are performed using the Monte Carlo MCNP6 code and the CFD code ANSYS-CFX 19, taking into account the capabilities of CFD codes to describe the heat transfer mechanisms at supercritical conditions. The main results show significantly lower values of the centerline temperature distribution in UN_ZrC and UC_TiN fuels compared to UO2 fuel. The fuel burnup values obtained at the end of the 540 days are practically the same, a slightly higher burnup is obtained for UC_TiN fuel because of its lower amount of heavy metals in the initial composition. The evolution of the fissile inventory is practically the same for all the analyzed fuel options, around 65% of fissile inventory is still present in the core at EoC. Total plutonium production shows a decrease of 1.85% (UN–ZrC) and 4.53% (UC–TiN) with respect to UO2 fuel. Coated ceramic fuels show a slightly improved proliferation resistance compared with UO2 fuel, since a slightly smaller fraction of 239Pu is obtained in their compositions.

Numerical study of hydrogen/methane buoyant fires using FireFoam

Hydrogen/methane buoyant fires with various hydrogen volume fractions ranging from 0% to 20% were numerically studied in this paper. The modified eddy dissipation concept combustion model for multi-fuels in the large eddy simulation (LES) framework was employed for combustion, and especially the infinitely fast rate based on “global” concept was improved. Combined with the weighted sum of gray gas model for emission/absorption coefficient, the finite volume discrete ordinates model was used to compute the radiative heat transfer. The predicted centerline temperature, velocity, and flame height are in good consistence with the measured data. Furthermore, the detailed analysis was conducted on the dependency of the parameters such as centerline temperature and velocity, flame height, and soot volume fraction on hydrogen volume concentration.

Effect of strip feeding into mold on fluid flow and heat transfer in continuous casting process

Steel strip feeding into the mold during continuous casting is known as an innovative technology. The newly applied technology is designed to further improve the slab quality. To analyze the complex phase change processes, molten sodium thiosulphate (Na2S2O3·5H2O) was used in the experimental investigation as a transparent analog for metallic alloys. Then, a numerical model incorporating fluid flow, heat transfer and phase change during strip feeding into the mold process was developed. The generalized enthalpy-based method was applied to describe the phase change behavior, and the porous media theory was used to model the blockage of fluid flow by the dendrites in the mushy zone between the strip and melt as well as the solidified shell and melt. The validated model was then used for the simulation of the real strip feeding into the mold process in an industrial scale. The whole shape of the strip under the effect of jet flow from the submerged entry nozzle (SEN) was presented. Results show that the strip will reach a pseudo-steady state after experiencing steel sheath formation, steel sheath melting and strip melting processes. When using the feeding method that is the strip narrow side toward the SEN in the present condition, the strip immersion length can reach 4.5 m below the meniscus and the slab centerline temperature can be decreased by 21 K to a maximum. When the strip feeding speed increased from 0.3 to 0.5 m/s, the minimum temperature of the centerline could be lowered by 4 K or so.

Preliminary neutronic/thermal-hydraulic evaluation and safety system optimization of PWR loaded with fully ceramic microencapsulated fuel

As one of the Accident Tolerant Fuels (ATFs), Fully Ceramic Microencapsulated (FCM) presents advantageous safety performance in the severe accidents compared with UO2 owing to its high melting point, superior heat transfer ability and proliferation resistance capability. In this study, preliminary neutronic and thermal-hydraulic analysis of Chinese improved three-loop Pressurized Water Reactor (PWR) with FCM pellet and SiC cladding (FCM-SiC) were conducted to demonstrate the fuel performance of ATFs. First, a three dimensional full core model loaded with FCM-SiC was developed to acquire the cycle length, power distribution and isotopic inventories. And low reactor power and high fuel enrichment were used to satisfy the annual reload requirement. Then, the reactor thermal-hydraulic safety evaluation was carried out by the modified subchannel and system analysis codes, which depicts lower Fuel Centerline Temperature (FCT) and Peak Cladding Temperature (PCT) in FCM-SiC reactor compared to the reactor with traditional fuels. The Minimum Departure from Nucleate Boiling Ratio (MDNBR) of cladding is much higher than the criterion of 1.3 as well. Furthermore, the reactor safety system was simplified and optimized by taking full advantage of the preponderance of ATFs, and the results show that the reactor of FCM-SiC can operate safely with only 10% initial flowrate of Low Pressure Safety Injection (LPSI) or without Accumulators (ACCs) during Large Break Loss of Coolant Accident (LBLOCA), while the conventional zircaloy is failed under the same conditions. This work could provide guidelines for the optimized design of the reactor safety system in the future.

Spatial temperature distribution of rectangular n-heptane pool fires with different aspect ratios and heat fluxes received by adjacent horizontal targets

Rectangular or line pool fires present some different behaviors compared to circular and square pool fires. A series of experiments of rectangular n-heptane pool fires have been conducted in open space. A set of fuel pans with the same width of 8 cm but different aspect ratios of length to width (n = l/w:1, 2, 3, 4, 6, 8, 10, 12 and 14) have been used. The results have showed that when n ≤ 6, the temperature rise ΔT at the same height is nearly unchanged from the pool center to the short pool edge, so the centerline temperature distribution of flame can be used to predict the vertical temperature distributions at different positions along the long pool edge. However, when n ≥ 8, an attenuation of the temperature occurs near the short pool edge. For the vertical distribution of centerline temperature rise, it is believed that the rectangular pool fires of n ≤ 2 have the similar patterns as point fire source and square fire source; for rectangular pool fires of n ≥ 3, one new correlation of flame centerline temperature rise has been proposed. Meanwhile, the heat fluxes received by adjacent horizontal targets also decrease when the horizontal distance increases.

Measurements and modeling of CO 4th positive (A–X) radiation

We present absolute intensity measurements of CO 4th positive emission from a CO2/Ar plasma jet produced using an atmospheric pressure plasma torch facility. The centerline temperature of the plasma jet is approximately 6300 K. A VUV emission spectroscopy system was adapted to the plasma torch facility to measure spectrally resolved emission down to 140 nm. The emission from the plasma in the VUV, UV and near IR is consistent with thermochemical equilibrium. Carbon lines in the UV and VUV are found to be optically thick and their amplitude is therefore substantially affected by the line broadening parameters used for the calculation. For the CO 4th positive emission, these measurements test the performance of various models for the electronic transition moment function (ETMF). These ETMFs are used to calculate sets of Einstein coefficients that are used by SPECAIR to calculate the spectrally resolved emission of the plasma for comparison with experiments. We find that the EMTF of Kirby and Cooper and Spielfiedel accurately reproduce the measured emission from 150–215 nm.

Computational simulation of round thermal jets in an ambient cross flow using a large-scale hydrodynamic model

This paper presents the numerical simulation of single, circular, turbulent, thermal jets discharged into an ambient fluid body with a uniform cross flow. The study utilizes a 3D hydrodynamic model to predict the dynamics of the evolving jets, with the model simulations calibrated against benchmark laboratory experimental datasets. Within the numerical-experimental model comparisons, the mean centreline temperature and velocity fields of the evolving jets are investigated in order to understand and predict the jet diffusion characteristics within the flowing ambient fluid body. Direct comparison between the numerical model predictions and laboratory datasets reveals that, with appropriate parameterization of the mixing processes and the selection of an appropriate numerical grid resolution, the large-scale hydrodynamic model can simulate both the near- and far-field thermal jet behaviour with good overall agreement, thus revealing a valid modelling tool used by environmental regulators for assessing the conformity of water quality of marine wastewater discharges.

The OFFBEAT multi-dimensional fuel behavior solver

The OpenFOAM Fuel BEhavior Analysis Tool, or OFFBEAT, is a multi-dimensional fuel performance code under collaborative development at the Laboratory for Reactor Physics and Systems Behaviour (LRS) of the EPFL and at the PSI. The code can be used both for studying complex 2D or 3D local effects and for more traditional 1.5D base irradiation analyses. OFFBEAT is based on the open-source C++ library OpenFOAM®, thus the governing equations are discretized with modern finite volume techniques using unstructured meshes, and the solver is fully parallelized via geometrical domain decomposition. This paper presents the numerical framework, governing equations and solution strategy behind OFFBEAT and it gives a brief overview of the material properties and behavioral models already implemented. Finally, a set of experimental rods is chosen to provide the first validation results for the thermal behavior of low burnup axisymmetric rods, by comparing the fuel centerline temperatures with experimental data.

Surrogate modeling of advanced computer simulations using deep Gaussian processes

The continuous advancements in computer power and computational modeling through high-fidelity and multiphysics simulations add more challenges on assessing their predictive capability. In this work, metamodeling or surrogate modeling through deep Gaussian processes (DeepGP) is performed to construct surrogates of advanced computer simulations drawn from the nuclear engineering area. This work is centered around three major ideas: (1) surrogate modeling through deep Gaussian processes (DeepGP), (2) simulation assessment through surrogate-based uncertainty quantification (UQ) methodologies, and (3) drawing conclusions regarding the underlying uncertainty of the four simulations considered in this paper. First, DeepGP models are trained, optimized, and validated to yield variety of features: (1) achieving high accuracy (small error metrics) on the validation set, (2) automatically capturing the surrogate model uncertainty (i.e. interpolation errors), (3) fitting multiple outputs with different scales simultaneously, (4) handling high dimensional input spaces, and (5) learning from small data amounts. Second, the validated DeepGP surrogates are utilized to efficiently perform UQ tasks such as uncertainty propagation (through Monte Carlo sampling), parameter screening (through Morris screening), and variance decomposition (through Sobol Indices) to investigate the selected simulations. Third, the thermal-hydraulics (fluid flow) results demonstrate the importance of inlet temperature uncertainty in void fraction predictions. For the reactor physics application (fuel depletion/consumption), DeepGP accurately captures the uncertainty in criticality calculations, which is about 0.6% (i.e. a considerable value for this application). For the application of kinetic parameters (nuclear data), DeepGP successfully explains 95% or more of the variance in all 12 outputs. Finally, DeepGP-based UQ analysis of the fuel performance application (materials science) shows the importance of the clad surface temperature, fuel porosity, and linear heat rate in explaining the variance of the maximum fuel centerline and surface temperatures.

Investigations on a typical small modular PWR using coupled neutronic-thermal-mechanical evaluations to achieve long-life cycle-length

The main purpose of this study was to design small modular PWR with power of 150 MWth for long life operation, based on the coupled neutronic and thermal-mechanical evaluations. In this regard, pitch and fuel rod diameter were changed from 0.9 cm to 2.5 cm and 0.6 cm–0.96 cm, respectively; and desirable pairs (pitch - fuel rod diameter) were obtained. Neutronic parameters such as excess reactivity, maximum to average ratio of the radial neutron flux and fuel burnup were calculated using MCNP code. Also, thermal-mechanical parameters of the fuel rod including fuel centerline temperature, cladding hoop stress and fuel rod internal gas pressure were computed using FRAPCON code. Independency of results from number of axial nodes in FRAPCON code was analyzed. The number of axial nodes was considered 40 to ensure the accuracy of the FRAPCON results. Neutronic and thermal-mechanical cost values for each pair were calculated and Pareto front of the desirable pairs were obtained. Based on the results, eight desirable pairs were selected. By assuming equal importance for neutronic and thermal-mechanical parameters, total cost value was defined and pair of (1.1, 0.64) was selected as the most desirable one. Finally, fuel cycle lifetime was assessed for desirable pair considering neutronic (criticality and burnup) and thermal-mechanical (cladding strain, oxidation and hydriding as well as fuel rod internal gas pressure) limits. It was observed that fuel cycle lifetime was enhanced to 1750 days (40 GWD/ton).

Analysis of forced convection of Phan–Thien–Tanner fluid in slits and tubes of constant wall temperature with viscous dissipation

In the present work, analytical solutions are presented for thermal convection of the linear Phan–Thien–Tanner fluid (LPTT) in slits and tubes of constant wall temperature by taking account of the viscous dissipation term. Unlike the similar previous studies in which the advection term was neglected in the heat transfer equation, it is considered in this investigation. A continuous relation between the Nusselt number and the Brinkman number is obtained. Expressions for the temperature distribution are derived in closed form and in terms of a Frobenius series for the slit and tube flows, respectively. Based on these solutions, the effects of fluid elasticity and Brinkman number on thermal convection of LPTT fluid flows are studied in detail. It is shown that at negative Brinkman numbers (fluid cooling), increasing the Deborah number leads to a decrease in the Nusselt number, but an increase in the centerline temperature. Nonetheless, this trend is opposite for positive Brinkman numbers (fluid heating), i.e., an increase in the Nusselt number and a decrease in the centerline temperature. Also, there is a Brinkman number beyond which the Nusselt number is smaller than zero, meaning that there is weak heat convection in the flow. Also, the results confirm that the extensibility parameter affects the temperature profile in the same way as the Deborah number.

Natural Convection Heat and Mass Transfer in the Vertical Cylindrical Porous Channel Under the Effects of Time-Periodic Boundary Condition

Abstract Heat and mass transfer are investigated numerically with steady-state laminar natural convection through a vertical cylindrical enclosure filled with a liquid-saturated porous medium. The vertical wall is under a constant magnetic field and various durations of periodic heating boundary condition; the top and bottom surfaces are kept at a constant cold temperature. Continuity, momentum, and energy equations are transformed to dimensionless equations. The finite difference approach with the line successive over-relaxation (LSOR) method is used to obtain the computational results. This study covers the heat transfer, the temperature distribution, and the velocity field in the domain under the variation of different parameters. The code used is validated by modifying it to analyze the Nusselt number in the existing experimental literature of Izadpanah et al. (1998, “Experimental and Theoretical Studies of Convective Heat Transfer in a Cylindrical Porous Medium,” Int. J. Heat Fluid Flow, 19(6), pp. 629–635). This work shows that Nusselt number decreases (with varying gradient) as the aspect ratio increases, and that it increases as the Rayleigh number increases. The centerline temperature has a proportional relationship with the heating amplitude and the heating period (as the system receives more heat) and is inversely proportional with Rayleigh number. Increasing the Rayleigh number causes increased convective velocity, which affects the position of the hot region, and causes a decrease in the temperature field. Increasing the aspect ratio results in a warm stream at the center of the cylinder, and when the time period of the heating increases, the circulation becomes faster and the intensity of the temperature contour layers decreases. In this work, a correlation for Nu as a function of the mentioned parameters is developed.

Thermal hydraulic analysis of the PWR with high uranium density accident tolerant fuels under accident transients with and without reactivity

High uranium density accident tolerant fuel (ATF) with high economy and safety seems to be one of the most promising candidates to replace UO2 in future reactor. The thermal hydraulic analysis of the PWR loaded with high uranium density ATF fuels needs to be carried out to demonstrate this fuel performance. The thermal hydraulic performance of UN and UN+U3Si2 composites were evaluated in the PWR under unprotected complete loss of coolant flow accident (CLOFA) and reactivity initiated accident (RIA). The processes of CLOFA and RIA transients with and without reactivity were carried out using system code Relap5/Mod3.4, and the thermal hydraulic behaviors of UN and UN+U3Si2 composite under CLOFA and RIA transients were assessed by sub-channel code COBRA-EN. The results show that UN and UN+U3Si2 fuels had more intensive negative reactivity feedback to further mitigate reactor power resulting in lower fuel temperature during CLOFA and RIA transients. The UN and UN+U3Si2 composites demonstrated better safety relative to UO2 because of higher minimal departure from nucleate boiling ratio (MDNBR) and lower peak fuel centerline temperature (PFCT). UN+U3Si2 composite performed better thermal hydraulic performance when compared to UN, but the safety margin may be lower due to low melting temperature of U3Si2.

Multiphysics modeling of thorium-based fuel performance with a two-layer SiC cladding in a light water reactor

The performance of thorium-based mixed oxide fuel ((Th,U)O2 fuel and (Th,Pu)O2 fuel) with a two-layer SiC cladding in a light water reactor have been investigated by finite element multiphysics modeling method. The material properties of specific thorium-based Th0.923U0.077O2 fuel, Th0.923Pu0.077O2 fuel and SiC cladding were firstly reviewed and implemented into the multiphysics model. Then the performance of Th0.923U0.077O2 fuel, Th0.923Pu0.077O2 fuel and UO2 fuel separately combined with Zircaloy cladding and two-layer SiC cladding have been investigated and compared under PWR normal operation conditions. Finally, the thorium-based fuel with Zircaloy cladding was found to decrease the fuel centerline temperature, especially for the Th0.923U0.077O2 fuel, but with a much earlier gap closure time. And the two-layer SiC cladding was found to effectively mitigate the pellet-cladding mechanical interaction (PCMI) but also greatly increase the fuel centerline temperature. So the combination of the thorium-based fuel with two-layer SiC cladding is expected to improve the reactor safety by keeping a moderate fuel centerline temperature and meanwhile greatly delaying the PCMI time.

Fuel performance optimization of U3Si2-SiC design during normal, power ramp and RIA conditions

Benefited from its higher mechanical strength and much-improved oxidation resistance at very high temperatures (>1200 °C), SiC cladding is one of the promising accident tolerant fuel (ATF) concepts. Considering the thermal conductivity degradation of SiC cladding after irradiation and its negligible creep rate, adopting conventional UO2 with SiC cladding can lead to aggravated fission gas release and elevated fuel centerline temperatures. In this work, U3Si2 fuel with its high thermal conductivity is considered in place of UO2, when adopting SiC cladding. First, the preliminary full core simulation for U3Si2-SiC was conducted to show its holistic fuel performance compared to the current UO2-Zircaloy fuel. Then the most limiting fuel rod was selected based on the calculated chemically vapor deposited (CVD) SiC failure risk from the full core simulation. For the limiting rod, fuel-clad gap optimization was conducted to minimize CVD SiC failure risk. Subsequently, power ramp history was appended to the normal power history and full core simulation analysis for updated U3Si2-SiC design was re-performed to ensure the effectiveness of the chosen gap width. For the limiting rod, reactivity-initiated accident (RIA) analysis at hot zero power (HZP) was analyzed to evaluate its performance and infer its potential failure modes. The results from normal operation and power ramp simulations implied that CVD SiC failure risk was mainly induced by the strong pellet-cladding mechanical interaction (PCMI) and the failure probability of CVD SiC can be effectively minimized to 2.0×10-7 through enlarging the fuel-clad as-fabricated gap width. The leading failure mode for U3Si2-SiC fuel system during RIA changed from fuel melting to Ceramic Matrix Composite (CMC) failure with increasing burnup.

Large Eddy Simulation of soot evolution in turbulent reacting flows: Presumed subfilter PDF model for soot–turbulence–chemistry interactions

In turbulent reacting flows, soot evolution is strongly influenced by small-scale soot–turbulence–chemistry interactions. Specifically, soot is formed during combustion of fuel-rich mixtures and, in non-smoking flames, is rapidly oxidized at slightly fuel-rich mixtures before being transported by turbulence into fuel-lean mixtures. Furthermore, different soot evolution mechanisms are dominant over distinct regions of mixture fraction. For these reasons, a new subfilter PDF is proposed to account for this distribution of soot in mixture fraction space. In this model, the sooting mode of a soot subfilter PDF is locally activated only at fuel-rich mixture fractions where surface growth is locally faster than oxidation. This model is first validated a priori against a recent DNS database of a turbulent nonpremixed jet flame. Compared to a subfilter PDF model without any mixture fraction dependence, a significant decrease in the oxidation source term is observed, while the surface growth source term remains hardly affected. This new model is implemented within a Large Eddy Simulation (LES) framework, applied to two laboratory-scale turbulent nonpremixed sooting jet flames, and validated via comparisons with experimental measurements of time-averaged centerline temperature and soot volume fraction. Both the new model and the previous model predict the temperature fairly well. However, compared to the previous model, the soot volume fraction predicted by the new model is significantly increased and in much better agreement with the experimental measurements. In addition, the sensitivity to the subfilter PDF is significantly greater than the sensitivity to grid resolution and is greater or at least comparable to the sensitivity to chemical mechanism.

A Revised Capsule Design for the Accelerated Testing of Advanced Reactor Fuels

The Advanced Test Reactor (ATR) has been used successfully for the testing of fast reactor fuel for nearly two decades. These successes have been in spite of numerous challenges for testing fast reactor fuel in the ATR (a thermal spectrum reactor), but the solutions to those challenges have resulted in excessively long irradiation times (~10 years) for high-burnup targets as well as experiments that are highly sensitive to fabrication tolerances and eccentricities. This paper presents a solution to the problems of extended irradiation times and fabrication sensitivities. Thermal and neutronic analyses were performed to show that a reduced-diameter fuel pin with an equivalent linear heat generation rate can provide a prototypic thermal profile (peak centerline and inner clad temperature) along with a near-prototypic power profile within the ATR thermal spectrum. This allows the experiment to reach a high burnup in an expeditious timeframe compared to traditional ATR fast fuel irradiations. In addition, problems with fabrication sensitivities were addressed by introducing a double-encapsulated experiment that pushes the high heat flux helium gap farther away from the fuel pin. Fuel pin position eccentricities are also mitigated by using a large sodium bond between the pin and capsule fuel. The advantages and potential pitfalls of this revised design are discussed, including the effect of length scales on fuel system behavior.

Modification and update of FROBA-ROD code and its applications in fuel rod behavior analysis for PWRs

To analyze the behavior of PWRs fuel rods under steady operation conditions, some models such as modified Forsberg-Massih model and TUBRNP model were applied to FROBA-ROD code, which was developed by Xi'an Jiaotong University in 2012. Then more significant fuel rod behaviors could be simulated by FROBA-ROD code, which facilitated the safety analysis of fuel elements especially under accident conditions. To conduct transient thermal calculations and to improve the simulation accuracy, existing thermal models in FROBA were modified with the new fuel rod thermal conductivity model and burnup distribution model. The new models were validated by comparing it with the results of authorized commercial codes FRAPCON-4.0 and ABQUS and by comparing it with IFA-432 Rod 1 experimental benchmark data conducted by Oak Ridge National Laboratory. The relative calculation errors of centerline temperature calculated by modified FROBA-ROD, original FROBA-ROD and FRAPCON were 5.44%, 11.68% and 9.47% separately, which indicated that the modification and update of FROBA-ROD were valid. Modified FROBA-ROD code was applied to perform the behavior analysis of the AP1000 fuel rod. The simulation results indicated that there were enough safety margins for fuel rod behavior of AP1000 at rated operation conditions. The maximum oxidation thickness of cladding appeared at the upper part of the cladding tube in AP1000, which indicated that this area suffered a higher risk of cladding failure. Modified FROBA-ROD code is expected to deal with fuel rod scenarios under accident conditions after some additions of transient models.