Fuel Performance Analysis Code Research Articles

Preliminary Verification of Key Physical Models in Oxide Fuel Performance Analysis Code FEATURE Based on COMSOL

Nuclear performance analysis is of great significance for fuel design optimization and the safe operation of a reactor. Considering the design requirements of the China initiative Accelerator Driven System (CiADS) project, it is necessary to develop relevant fuel performance analysis tools to predict fuel performance evolution during service. The Fuel Element Analysis Tool for Undercritical Reactor Engineering (FEATURE) code, based on the advanced multiphysics coupling platform COMSOL software, is under development in this context, with a current focus on oxide fuel for fast reactors. This paper mainly conducts preliminary comparative verification of the oxygen redistribution, plutonium redistribution, and fission gas release modules of the FEATURE code through model examples, code-to-code evaluation, and experimental data. The results indicate that the FEATURE code exhibits good accuracy and reliability in simulating oxygen and plutonium redistribution, as well as fission gas release, making it suitable for integrated coupling calculations.

Transient fuel behavior analysis of UN fuel with a two-layered SiC cladding based on multiphysics method

Recently, UN fuel has emerged as a promising alternative to UO2 fuel due to its exceptional thermophysical attributes, including heightened thermal conductivity, elevated melting point, and increased fissile density. Meanwhile, SiC is recognized as a member of the next generation of accident-tolerant cladding materials owing to its impressive strength at high temperatures, resistance to creep, minimal thermal expansion, and improved irradiation characteristics. In this study, utilizing our developed fuel performance analysis code CAMPUS, we integrate a multiphysics model featuring UO2 fuel paired with a two-layered SiC cladding into the CAMPUS framework. Subsequently, we conduct simulations and analyses of the fuel performance for combinations such as UO2-two-layered SiC, UN-two-layered SiC, UN-Zircaloy, and UO2-Zircaloy. These evaluations encompass both standard operation scenarios and potential accident conditions within a pressurized water reactor (PWR). Notably, the incorporation of UN fuel demonstrates a significant reduction in maximum fuel temperature during normal operation, as well as under Loss-of-Coolant Accident (LOCA) and Reactivity Initiated Accident (RIA) conditions. However, upon introducing the two-layered SiC cladding, it becomes apparent that its temperature performance is somewhat inferior to that of Zircaloy under normal operation and RIA scenarios. Conversely, when subjected to a LOCA scenario, the thermal and mechanical characteristics of the two-layered SiC cladding outperform those of the Zircaloy cladding.

Coupled neutronics, thermal-hydraulics, and fuel performance analysis of dispersion plate-type fuel assembly in a cohesive way

The nuclear reactor is a multi-physics and multiscale coupling system. The conditions of high temperature and high neutron flux in the reactor pose a significant challenge to the thermodynamic performance of the fuel. To evaluate the plate-type fuel's performance, a coupling calculation of neutron physics code OpenMC, fuel performance analysis code BEEs-Plates, and nuclear reactor system analysis code NUSAC has been implemented based on the open-source MOOSE platform. A novel and efficient method for mesh grids generation and data mapping between external codes and MOOSE was proposed to ensure accurate data exchange. This method can simplify the mesh grids generation process of external codes and accurately transfer data across different mesh grids. Finally, the coupling calculation of the JRR-3 M research reactor fuel assembly was carried out based on the above coupling code, and the neutronics parameters, thermal–hydraulic parameters, and fuel performance-related parameters of the fuel assembly were analyzed. The calculation results show that the maximum fuel burnup will reach 16.91% FIMA after 240 days. With the increase of fuel burnup, the peak power of the fuel assembly is significantly shifted towards both ends of the height direction. The VonMises stress, volume stress, creep strain, and displacement of the fuel will be considerable. Finally, the impact of the coupling calculation on the accuracy of the computational efficiency is preliminarily investigated.

Upgradation of fission gas release model in the nuclear fuel analysis code FuSPAC

The evaluation of fuel performances in the whole life cycle is important to the fuel integrity. In our previous work, a fuel performance analysis code FuSPAC was developed and validated against the in-pile test data and FRAPCON code. In the current work, this code was upgraded by replacing the modified Forsberg-Massih for fission gas release (FGR) with a mechanism model. The microscopic behaviors related to the FGR, including the gas atom diffusion, trapping, re-solution and bubble nucleation inside the granular particles, the vacancy diffusion at the inter-granular boundaries, the grain growth and the grain recrystallization were modeled. The upgraded FuSPAC code was validated against the in-pile test data for FGR and internal void volume. The sensitivity analysis on the sub-models shown that the intra-granular fission gas diffusion, the re-solution, and the inter-granular vacancy diffusion had significant effects on the FGR. The effects from rod power can be well reflected by the upgraded code compared to the original code.

BISON analysis of FeCrAl and Zircaloy cladding deformation during simulated BWR cyclic dryout conditions

The BISON fuel performance analysis code has been used to study FeCrAl and Zircaloy cladding performance to better understand cladding deformation behavior in boiling water reactors during conditions caused by anticipated operational occurrences (AOOs) or by anticipated transients without scram (ATWS). BISON models of novel separate effects tests simulating ATWS events predicted mechanical responses to thermal cycling boundary conditions. We assessed FeCrAl and Zircaloy deformation and burst models to determine those best suited for cyclic dryout event analysis. In addition, we compared diametral strain profiles of the post-experiment claddings to simulation results. The capability of the BISON code for modeling BWR AOO or ATWS events is promising.

Development and evaluation of fuel performance analysis code FuSPAC

The nuclear fuel rod should maintain self-sustainable nuclear fission, as the first safety barrier to prevent the leakage of radioactive products. The fuel rod performance analyses are essential to ensure reliable long-term operation of fuel rod. In this work, the fuel performance analysis code FuSPAC is developed based on the commonly used 1.5-dimensional method for UO2 ceramic fuel with Zirconium cladding. The physical and numerical models used in FuSPAC code are described. By comparing the predicted values from FuSPAC code with the available in-reactor fuel rod experimental data and the numerical results from the well-recognized FRAPCON code, the capacity of FuSPAC code on the prediction of fuel pin and cladding under long term irradiation conditions are validated from five aspects, including fission gas release behaviour, thermal behaviour, rod internal void volume, cladding corrosion and cladding hoop strain, showing good accuracy in the behaviour analysis of the irradiated UO2 fuel rod.

Investigation on radial fuel relocation and its influence on heat split phenomenon of dual-cooled annular fuel element

Radial fuel relocation due to fuel fragmentation can significantly change the heat split, further affecting the thermal performance and safety of dual-cooled annular fuels. Considering the rarity of in-pile irradiation data for annular fuels, three different fuel relocation and its recovery models were developed based on modifications of relocation models for solid fuel rods or the mechanistic relocation simulation of pre-fragmented, interacting and deformable fuel fragments. The three relocation models were implanted into the fuel performance analysis code FROBA-ANNULAR, and thermo-mechanical simulation was carried out to investigate the influence of fuel relocation on gap size variation and heat split variation in an annular fuel. The results indicate that different relocation models lead to significant difference on the prediction of gap size and heat split. Relocation models developed through mechanistic simulation of pre-fragmented fragments seems to be more reasonable, compared with the other two models developed based on unreasonable and extreme assumptions. Mechanistic simulation results indicate that radial relocation of fragmented annular fuels can be bidirectional even at the initial burnup stages, resulting in the reduction of gap sizes of both internal and external fuel-cladding gaps. In addition, it was found that the gap size variation due to radial fuel relocation significantly changed the heat split and matching-rate between heat split and flow flux distribution, exerting an influence on fuel temperature and MDNBR.

A new method to simulate dispersion plate-type fuel assembly in a multi-physics coupled way

The plate-type fuel has an excellent thermal–mechanical behavior as a type of nuclear fuel. To simulate and analyze the behavior of plate-type fuel assembly under irradiation conditions, a fuel performance analysis code BEEs-Plates was developed based on MOOSE (Multiphysics Object-oriented Simulation Environment). The properties, mechanical behavior, and irradiation effects of dispersion fuel meat were simulated by using the equivalent property models, and a one-dimensional Petrov-Galerkin stability model was used to solve the governing equations of coolant. The coupling between the three-dimensional solid domain and the one-dimensional fluid domain was also realized. The JRR-3 research reactor fuel assembly model was then used to perform the coupling simulation, and the assembly’s temperature, stress, strain distribution, and coolant temperature, the size of the channels, flow distribution were analyzed. The results under the boundary conditions and input parameters applied herein show that the burnup effect has limited effects on the thermohydraulic performance of the assembly, with the variation of maximum temperature below 1 K and the change of channel width below 3.9%. However, its mechanical behavior is greatly affected by burnup, with the maximum stress of 396.2 MPa at the meat-cladding interface. The uneven power distribution causes the assembly to bend to one side, but its effect on the coolant flow distribution is negligible.

Optimum coating layer thicknesses of a TRISO having an 800-μm UO2 kernel under normal operation conditions of a 10-MWth block-type HTGR

The study evaluates the optimal thicknesses of the coating layers of an 800-μm UO2 tri-structural isotropic coated fuel particle (TRISO) for a 10-MWth block-type micro modular high temperature gas-cooled reactor (HTGR) using a response surface methodology (RSM) and an HTGR fuel performance analysis code, COPA. The objective function to be minimized in the optimal design of a TRISO consists of the product of the packing fraction of TRISO particles in a compact, the fractional releases of major metallic fission products from a TRISO and the failure probability of the silicon carbide (SiC) layers. A constraint on the sum of the coating layer thicknesses was made using a maximum packing fraction. Two fuel performance analyses such as a failure fraction analysis and a fission product release analysis are applied to the optimal designs of a TRISO that the RSM produced. An 800-μm UO2 TRISO having buffer, inner high-density pyrocarbon (IPyC), SiC, outer high-density pyrocarbon (OPyC), each with a thickness of 110, 40, 57, 35 μm, or 120, 35, 60, 27 μm is suitable for the operation of a 10-MWth block-type micro modular HTGR within 7600 days when both the SiC and through-coatings failure fractions do not occur and the fractional release of silver, cesium, strontium and krypton are below 1 × 10−9. The packing fraction of the optimal TRISOs in a compact is 40%.

Development of fuel performance analysis code, BISON for MOX, named Okami: Analyses of pore migration behavior to affect the MA-bearing MOX fuel restructuring

• 2-D analyses with modelling for MOX fuel restructuring in use of fast reactor. • Effects of O/M ratio and pellet-cladding gap width on pore migration behavior. • Irradiation behavior evaluation for the MA-bearing MOX irradiation test in Joyo. • Influence of pellet eccentricity on central void formation in 2-D analyses. Fuel composition, i.e. Pu content, minor actinide (MA) content, oxygen-to-metal (O/M) ratio, etc., affects the irradiation behavior of MA-bearing mixed oxide (MOX) for fast reactor. To evaluate the O/M dependence of pore migration regarding fuel restructuring at the beginning of irradiation, we are developing BISON for MOX in cooperation with Idaho National Laboratory (INL) and have installed a pore migration model considering vapor pressure of vapor species and thermal conductivity for MOX. The O/M dependence of fuel restructuring observed in MA-bearing MOX irradiation in the experimental fast reactor Joyo was evaluated by a two-dimensional (2-D) analysis. To solve partial differential equations for energy conservation and advection diffusion equations to calculate temperature and pore migration, respectively, a correlation of thermal conductivity as a function of porosity, temperature, deviation from stoichiometry, Am content and Np content was used. In addition, the pore migration velocity was calculated by using a correlation depending on vapor pressure applying Rand-Markin's model. Four MA-bearing MOX test pins with different O/M ratio and pellet/cladding gap size were irradiated in B14 test in Joyo to study the effects of O/M ratio and gap size on irradiation behavior at the beginning of irradiation. In post-irradiation examinations (PIE), a remarkable restructuring was observed in stoichiometric fuels, and a discrepancy in central void diameter among various O/M ratios. This restructuring behavior could be evaluated by considering the influence of O/M ratio on vapor pressure. Also, a central void is assumed to offset from the geometric center to the larger-gap side due to the pellet eccentricity, but the PIE results were inconsistent in this regard. It is expected that the irradiated pellet might displace to another side during sample preparation in PIE from a reverse offset location during irradiation by comparing the result of 2-D analysis and radial profile of fuel density.

Validation of Serpent-SUBCHANFLOW-TRANSURANUS pin-by-pin burnup calculations using experimental data from the Temelín II VVER-1000 reactor

This work deals with the validation of a high-fidelity multiphysics system coupling the Serpent 2 Monte Carlo neutron transport code with SUBCHANFLOW, a subchannel thermalhydraulics code, and TRANSURANUS, a fuel-performance analysis code. The results for a full-core pin-by-pin burnup calculation for the ninth operating cycle of the Temelín II VVER-1000 plant, which starts from a fresh core, are presented and assessed using experimental data. A good agreement is found comparing the critical boron concentration and a set of pin-level neutron flux profiles against measurements. In addition, the calculated axial and radial power distributions match closely the values reported by the core monitoring system. To demonstrate the modeling capabilities of the three-code coupling, pin-level neutronic, thermalhydraulic and thermomechanic results are shown as well. These studies are encompassed in the final phase of the EU Horizon 2020 McSAFE project, during which the Serpent-SUBCHANFLOW-TRANSURANUS system was developed.

Multiphysics analysis of thorium-based fuel performance under reactor normal operating and accident conditions in a light water reactor

The ThO2 fuel has higher thermal conductivity and melting boiling point than the UO2 fuel, which is beneficial to the fast removal of heat and the improvement of fuel melt margin. In this paper, the material properties and thermodynamic behaviors of thorium-based fuel were firstly reviewed, and then the thermal physical properties and the fuel behavior models of Th0.923U0.077O2 fuel and Th0.923Pu0.077O2 fuel have been implemented in fuel performance analysis code FRAPCON and FRAPTRAN. Finally, the performances of Th0.923U0.077O2 fuel, Th0.923Pu0.077O2 fuel and UO2 fuel under both normal operating conditions and transient conditions (both RIA and LOCA) are analyzed and compared. The thorium-based fuels are found to have lower fuel centerline temperature than that of the UO2 fuel under normal operating condition, and a delayed pellet-cladding mechanical interaction (PCMI) under both normal operating and transient conditions. Furthermore, the fission gas release, cladding strain and internal fuel energy under transient conditions are found to be lower too. Lastly, the cladding displacement and temperature under transient conditions are also compared. The thorium-based fuel was found to have a higher safety margin and accident resistance than the UO2 fuel under both normal operating condition and transient accident conditions.

A comparative study of in-pile behaviors of FeCrAl cladding under normal and accident conditions with updated FROBA-ATF code

Accident tolerant fuel (ATF) concepts are at present still in the engineering design stage. FeCrAl, which has great oxidation resistance in high temperature steam environment, is an ATF potential cladding material with higher research priority. However, considering its features, including high thermal expansion, low creep rate and strong absorption of neutron, further evaluation on its in-pile behaviors is still necessary for future industrial application. To address this issue, the self-developed fuel performance analysis code, FROBA, was further developed into a new version named FROBA-ATF. A transient heat transfer model was added into the code’s thermo-hydraulic module for transient simulation, the rigid mechanical model was rewritten into unrigid for ATF cladding and material properties of FeCrAl cladding were also added. These updated modules were introduced and the updated code was assessed in this paper. On this basis, a comparative study on in-pile behaviors of UO2-FeCrAl fuel rod and UO2-Zr4 fuel rod under normal and accident conditions were performed in this paper. Even through the performance analysis revealed that FeCrAl cladding could greatly reduce the oxidation thickness, the thermal performance of UO2-FeCrAl fuel rod was unsatisfactory due to its higher thermal expansion and lower creep rate, especially at beginning of life. With geometry optimization, the steady-state performance of UO2-FeCrAl fuel rod was enhanced. What’s more, the thermal and mechanical performance of the improved UO2-FeCrAl was evaluated and compared under Loss of coolant accident (LOCA) and Reactivity Initiated Accident (RIA) conditions. The results showed that the optimized ATF had better thermal performance, lower cladding hoop stress, and could provide more coping time under accident conditions.

A new relocation recovery model for LWR fuel

The pellet cracking and relocation recovery is an important physical phenomenon for LWR fuel and also is an indispensable model for the fuel performance analysis code. During rise to power, there exists a large temperature gradient within the pellet from the centerline to the cooler periphery and it is enough to induce cracking to release the thermal stress. Simultaneously, the pellet diameter increases rapidly. Once the fuel rod gap is closed, the cladding constraint produced by pellet-cladding interaction plays an important role in the process of relocation recovery. In this paper, a new relocation recovery model as a function of power, contact pressure and time is established. The new model has been successfully implemented in the fuel rod performance analysis code JASMINE. Based on comparisons between code predictions and experimental measurements, the results indicate that the newly developed relocation recovery model has an excellent predictive ability.

The development of TRIAC-BATAN: a triso fuel performance analysis code

A computer code for analyzing the safety feature of triso coated particles for high temperature reactor has been developed based on PANAMA code. This python based code were separated into several modules to perform the main TRIAC-BATAN and linear interpolation calculation, to read the input data file, and to control the sequence of all TRIAC-BATAN calculation. In this initial development phase, TRIAC-BATAN can produce similar pattern as PANAMA calculation in the case of failure fraction of triso particles, either for Depressurized Loss Of Forced Cooling (DLOFC), 100°C higher than DLOFC temperature and constant accident temperature at 1600°C. By using the euclidean distance, TRIAC-BATAN calculation and PANAMA for DLOFC condition are separated in around 3.23 ⋅ 10−7.

ABAQUS analysis of the SiC cladding fuel rod behavior under PWR normal operation conditions

Silicon Carbide (SiC) cladding is anticipated to be superior to zirconium-based cladding with regard to tolerating light water reactor (LWR) accidents, but it is also crucial that the cladding performance under normal operation conditions be well understood. This paper focuses on the modeling of thermomechanical behavior of a two-layer SiC cladding in Pressurized Water Reactor (PWR) fuel rod under normal operation conditions. For this purpose, a set of ABAQUS subroutines has been written to incorporate material properties and behavior models of Chemical Vapor Deposited (CVD) monolithic SiC, Chemical Vapor Infiltration (CVI) SiCf/SiC Ceramic Matrix Composite (CMC) and UO2. Especially, apart from UO2 irradiation behaviors and fuel rod phenomena models (gas pressure, gap heat transfer, etc.) which have been extensively studied, an alternative swelling model of SiC dedicated for transient simulation and a pseudo-ductile behavior model of CVI SiCf/SiC have been implemented. Both models are important for conducting best estimate analysis of the fuel rod in-pile behavior. Then after benchmarking with the BISON fuel performance analysis code, the ABAQUS model is applied to a downscaled PWR fuel rod with two-layer SiC cladding design, considering different cladding thicknesses and covering the conditions of simple constant power levels, power ramps, power operation mode, reactor hot standby and cold shutdown. The effect of swelling differential between CVD SiC and CVI SiCf/SiC is also discussed. With the help of finite element analysis (FEA), the tightly coupled thermomechanical behavior of SiC cladding fuel rod can be well studied.

Gas releases and rod internal pressure of BN-added UO2 fuel

Gadolinium (Gd), in the form of gadolinia (Gd2O3), has been used as integral burnable absorber to UO2 fuel by different companies. Recently, alternative burnable absorber materials are under investigation. In this paper, the results of the study on the use of boron mixed in UO2 as an integral burnable absorber are presented. Boron nitride was selected as the form of B to be added to UO2 fuel based on the stability of B during sintering. Fuel performance analysis code FRAPCON-UNI, which is based on FRAPCON code and has the modifications for boron and helium generation and gasses release models, was used for the calculation of fuel performance. Helium generation was calculated using a model and the result was verified against ORIGEN calculation results. Gas release calculation was based on Forsberg Massih model and different characteristics of the gasses were considered. Power histories of maximum burnup rods (B rod, no B rod, and Gd rod) were selected from reactor physics data. The results indicated significant increases in rod internal pressure by 500 ppm boron addition at the end of life with minimum influence of a grain boundary boron fraction due to high diffusivity of helium and nitrogen.

Sensitivity analysis of fuel pin failure performance under slow-ramp type transient overpower condition by using a fuel performance analysis code FEMAXI-FBR

The FEMAXI-FBR is a fuel performance analysis code and has been developed as one module of core disruptive evaluation system, the ASTERIA-FBR. The FEMAXI-FBR has reproduced the failure pin behavior during slow transient overpower. The axial location of pin failure affects the power and reactivity behavior during core disruptive accident, and failure model of which pin failure occurs at upper part of pin is used by reflecting the results of the CABRI-2 test. By using the FEMAXI-FBR, sensitivity analysis of uncertainty of design parameters such as irradiation conditions and fuel fabrication tolerances was performed to clarify the effect on axial location of pin failure during slow transient overpower. The sensitivity analysis showed that the uncertainty of design parameters does not affect the failure location. It suggests that the failure model with which locations of failure occur at upper part of pin can be adopted for core disruptive calculation by taking into consideration of design uncertainties.

Oxide thickness-dependent transient cladding hoop stress

Abstract A few thrice-burned Zry-4 fuel assemblies which were loaded in one of the PWRs operating in Korea were found to be failed due to PCI during a power ramp following a rector trip, while thrice-burned Zr–Nb fuel assemblies and twice-burned Zry-4 ones were intact. To investigate the effect of fuel rod oxide thickness on power ramp-induced cladding hoop stress, three intact fuel rods were selected, which include an intact twice-burned Zry-4 fuel rod, an intact thrice-burned Zr-4 fuel rod and an intact thrice-burned Zr–Nb fuel rod. With the use of a fuel performance analysis code, burnup-dependent steady-state cladding stress and ramp power-dependent cladding stresses at the power-ramped burnup were predicted for the three intact fuel rods. It was found that the cladding oxide thickness has a considerable effect on the ramp power-dependent cladding hoop stresses. In addition, the cladding maximum stress of the thrice-burned Zry-4 fuel rod with 125 μm oxide thickness exceeded an ultimate cladding tensile strength of the Zry-4 cladding when the pellet–clad friction coefficient-dependent cladding stress concentration ratio was considered. However, the thrice-burned Zr–Nb fuel rod with 50 μm oxide thickness was evaluated to have a considerable margin against the power ramp-induced PCI failure.

A fuel performance analysis for a 450 MW th deep burn-high temperature reactor

A performance analysis for a 450 MW th deep burn-high temperature reactor (DB-HTR) fuel was performed using COPA, a fuel performance analysis code of Korea Atomic Energy Research Institute (KAERI). The code computes gas pressure buildup in the void volume of a tri-isotropic coated fuel particle (TRISO), temperature distribution in a DB-HTR fuel, thermo-mechanical stress in a coated fuel particle (CFP), failure fractions of a batch of CFPs, and fission product (FP) releases into the coolant. The 350 μm DB-HTR kernel is composed of 30% UO 2 + 70% (5% NpO 2 + 95% PuO 1.8) mixed with 0.6 moles of silicon carbide (SiC) per mole of heavy metal. The DB-HTR is operated at the constant temperature and power of 858 °C and 39.02 mW per CFP for 1395 effective full power days (EFPD) and is subjected to a core heat-up event for 250 h during which the maximum coolant temperature reaches 1548.70 °C. Within the normal operating temperature, the fuel showed good thermal and mechanical integrity. At elevated temperatures of the accident event, the failure fraction of CFPs resulted from the mechanical failure (MF) and the thermal decomposition (TD) of the SiC barrier is 3.30 × 10 −3.