LWR Fuel Research Articles

Thermophysical characterization of UFe3B2 and USiNi: An experimental study

In the development for advanced nuclear fuels for LWRs, the focus has traditionally centered on a limited array of materials like doped-UO2, UN, and U3Si2, leaving the vast majority of potential uranium compounds unexplored. To expand the search and expedite the discovery of advanced LWR fuels, we developed a machine learning (ML) classification model dedicated to identifying uranium compounds with excellent thermal conductivity. However, the model's predictive accuracy and the feasibility of such an ML-assisted approach to advanced nuclear fuel discovery have yet to be empirically validated. Therefore, this study examines the thermal properties of two uranium compounds, UFe3B2 and USiNi, that are predicted to outperform UO2 at high temperatures. Our findings confirm that the model precisely predicted the thermal conductivities of both compounds, thereby adding credibility to its reliability. However, the experimental process unveiled several challenges, such as the fabrication of single-phase USiNi and microcrack-free UFe3B2 samples, which our ML model, focused primarily on thermal conductivity, could not anticipate. These challenges underscore the need for a more comprehensive ML framework that considers the intricacies of sample fabrication. This advancement is crucial for the more efficient development of practical, accident-tolerant, advanced LWR fuels.

STEM Analysis of High Burnup Structure in LWR Fuels.

STEM Analysis of High Burnup Structure in LWR Fuels.

Assessment of Rans Turbulence Model Performance in Tight Lattice Lwr Fuel Subchannels

Assessment of Rans Turbulence Model Performance in Tight Lattice Lwr Fuel Subchannels

SUPERFACT: A Model Fuel for Studying the Evolution of the Microstructure of Spent Nuclear Fuel during Storage/Disposal.

The transmutation of minor actinides (in particular, Np and Am), which are among the main contributors to spent fuel α-radiotoxicity, was studied in the SUPERFACT irradiation. Several types of transmutation UO2-based fuels were produced, differing by their minor actinide content (241Am, 237Np, Pu), and irradiated in the Phénix fast reactor. Due to the high content in rather short-lived alpha-decaying actinides, both the archive, but also the irradiated fuels, cumulated an alpha dose during a laboratory time scale, which is comparable to that of standard LWR fuels during centuries/millenaries of storage. Transmission Electron Microscopy was performed to assess the evolution of the microstructure of the SUPERFACT archive and irradiated fuel. This was compared to conventional irradiated spent fuel (i.e., after years of storage) and to other 238Pu-doped UO2 for which the equivalent storage time would span over centuries. It could be shown that the microstructure of these fluorites does not degrade significantly from low to very high alpha-damage doses, and that helium bubbles precipitate.

Evolution of the uranium concentration in dissolution experiments with Cr-(Pu) doped UO2 in reducing conditions at SCK CEN

Cr-doped UO2-based model materials were prepared at SCK CEN, mimicking modern LWR fuels, to understand the influence of Cr doping on the spent fuel dissolution behaviour in geological repository conditions. Tests were carried out with four model materials: depleted UO2, Cr-doped depleted UO2, Pu-doped UO2 and Pu-Cr-doped UO2. Static dissolution experiments have been performed up to 4 months in autoclaves under 10 bar H2 pressure with a Pt/Pd catalyst in media at pH 13.5 and at pH 9. The Cr-doping appeared to reduce the U concentrations by a factor 6 at pH 13.5, but it had no or not much effect at pH 9.Graphic abstract

Dispersion of Surrogate LWR Fuel Experiments Under LOCA Conditions

In this study a series of experiments were performed subjecting surrogate nuclear fuel rods to high-pressure transients to induce fuel dispersion representative of the expected conditions of a fuel rod during a hypothetical loss-of-coolant accident. Experiments were conducted on like-for-like pressurized water reactor geometries in both a single-rod and rod-bundle configuration. In the rod-bundle configuration, a matched index of refraction techniques was employed to provide optical access to the bundle internals and to view the surrogate fuel dispersion event. Both configurations used small lead pellets as a surrogate fuel and were observed with a high-speed camera to capture the transient on a resolved timescale. For the single-rod experiments, the test rod was subjected to pressure transients at 4.0, 8.0, and 12.0 MPa multiple times, and for the rod-bundle experiments, the rod was subjected to 8.0 MPa transients in order to compare mechanical behavior against the single-rod test at 8.0 MPa. For both configurations, the results showed highly variable behavior in both the quantity of fuel dispersed and the mean displacement relative to the burst rod origin, likely due to statistical variations in the internal fuel stack orientation. Measurements of the rod plenum internal pressure showed no discernible difference in depressurization rates at a given pressure, indicating the likelihood that the mass flow rate is limited by the valve orifice in the current experimental configuration. The bundle tests also showed that a 5 × 5 array appears to be too small to capture the full spatial distribution of dispersed fuel, thus future tests will employ a larger bundle size and particle collection technique.

Analytical study of SPERT-CDC test 859 using fuel performance codes FEMAXI-8 and RANNS

The fuel-failure-limit data obtained in the simulated reactivity-initiated-accident experiment SPERT-CDC 859 (SPERT859) has entailed a lot of discussions if it represents fuel-failure behavior of typical commercial LWRs for its specific pre-irradiation condition and fuel state. The fuel-rod conditions before and during SPERT859 were thus assessed by the fuel-performance codes FEMAXI-8 and RANNS with focusing on cladding corrosion and its effect on the failure limit of the test rod. The analysis showed that the fuel cladding was probably excessively corroded even when the influential calculation conditions such as fuel-swelling and creep models were determined so that the lowest limit of the cladding oxide layer thickness was captured. Such assumption of excessive cladding corrosion during pre-irradiation well explains not only the test-rod state before pulse irradiation but also the fuel-failure limit observed. Such understanding undermines anew the representativeness of the test data as a direct basis of safety evaluation for LWR fuels.

Study of Neptunium, Americium and Protactinium Addition for 300MWth GFR with Uranium Carbide Fuel

A study of Neptunium, Americium, and Protactinium addition for GFR 300MWth with Uranium Carbide fuel has been performed. The purpose of this study was to determine the characteristics of addition Neptunium, Americium, and Protactinium in a 300MWth Gas-Cooled Fast Reactor. Neutronics calculation was design by using Standard Reactor Analysis Code (SRAC) version 2006 with data nuclides from JENDL-4.0. Neutronics calculations were initiated by calculating the fuel cell calculation (PIJ calculation) and continued with the reactor core calculation (CITATION calculation). The reactor core calculation used two-reactor core configurations, namely the homogeneous core configuration and heterogeneous core configuration. The Neptunium, Americium, and Protactinium additions were performed after obtaining the optimal condition from heterogeneous core configuration. The addition of Neptunium and Americium which are Spent Nuclear Fuel (SNF) from LWR fuels, aims to reduce the amount of Neptunium and Americium in the world and also to reduce the effective multiplication factor (k-eff) value from the reactor. The results obtained that the addition of Neptunium and Americium causes the k-eff value was decreased at the beginning of burn-up time, but increase at the end of burn-up time. It was because Neptunium and Americium absorb neutrons at the beginning of burn-up time and turns into fissile material at the end of burn-up time. The addition of protactinium in the reactor causes the k-eff value to be decreased both at the beginning of the burn-up time and at the end of the burn-up time. It happens because Protactinium absorbs neutrons both at the beginning of the burn-up time and at the end of the burn-up time. Therefore protactinium is often called a burnable poison.

Thresholds for failure of high-burnup LWR fuels by Pellet Cladding mechanical interaction under reactivity-initiated accident conditions

ABSTRACT To contribute to the future updating on the Japanese safety criteria for pellet/cladding mechanical interaction (PCMI) failure of light water reactor fuels under reactivity-initiated accident (RIA) conditions, this paper summarizes the recent important outcomes from research programs with the Nuclear Safety Research Reactor (NSRR). Applicability of current criteria, which are defined as a function of fuel burnup and possibility of introducing another parameter for new criteria were evaluated based on the results of the RIA-simulated pulse irradiation tests, post-test examinations, and supporting analytical work, such as the reevaluation of fuel enthalpies in earlier NSRR experiments. Failure-threshold curves based on cladding hydrogen content as a primary measure of fuel degradation have been proposed as a possible alternative that can be used to judge the occurrence of PCMI failure to ensure conservativeness in a more pertinent manner.

A new fuel modeling capability, CTFFuel, with a case study on the fuel thermal conductivity degradation

Abstract A new fuel modeling capability, CTFFuel, is developed from the subchannel code, CTF. This code is a standalone interface to the CTF fuel rod models, allowing for fuel rod simulations to be run independently from the fluid. This paper provides an overview of the code with a case study on the thermal conductivity degradation of LWR fuels to demonstrate its capabilities. The modeling of fuel thermal conductivity degradation in the code is improved through the addition of new modeling options to account for the irradiation effects via globally defined parameters. After the initial implementation, a variety of order-of-accuracy tests and code comparisons are performed to test software quality. A controlled analysis is allowed by CTFFuel to verify the numerical scheme of CTF’s conduction solution and to benchmark its fuel temperature predictions against FRAPCON-4.0’s. Overall, the software quality and verification procedure ensures that the new model is coded correctly, that it properly interacts with the rest of the code.

Using the Time-Correlated Induced Fission Method to Simultaneously Measure the 235U Content and the Burnable Poison Content in LWR Fuel Assemblies

The uranium neutron coincidence collar uses thermal neutron interrogation to verify the 235U mass in low-enriched uranium (LEU) fuel assemblies in fuel fabrication facilities. Burnable poisons are commonly added to nuclear fuel to increase the lifetime of the fuel. The high thermal neutron absorption by these poisons reduces the active neutron signal produced by the fuel. Burnable poison correction factors or fast-mode runs with Cd liners can help compensate for this effect, but the correction factors rely on operator declarations of burnable poison content, and fast-mode runs are time-consuming. This paper describes a new analysis method to measure the 235U mass and burnable poison content in LEU nuclear fuel simultaneously in a timely manner, without requiring additional hardware.

LWR fuels with enhanced safety and performance

LWR fuels with enhanced safety and performance

A Collision Probability and MOC-Based Lattice and Burnup Code for Analysis of LWR Fuel Assemblies

The BOXER3 code was developed in the Bhabha Atomic Research Centre during the 1980s as a three-dimensional code for the analysis of a pressurized heavy water reactor supercell containing fuel, moderator, and a reactivity device inserted perpendicular to the fuel channel, with options for carrying out calculations in a general two-dimensional geometry (infinite and homogeneous in one direction) and a one-dimensional plane geometry. Taking into account the computing resources available then, the code was run in few groups after obtaining condensed group cross sections for various materials from a one-dimensional multigroup calculation.In this paper, we describe various developments carried out recently for enabling its use as an assembly-level lattice-burnup code. In addition to the collision probability method originally available, the method of characteristics for solving the multigroup transport equation has been added. This development permits the treatment of anisotropic scattering wherever necessary and available in cross-section libraries. Other developments include coupling of the code to the WIMS 69/172-group library, a method for the evaluation of the pin-dependent Dancoff factor, and the introduction of burnup. The transport equation in the collision probability method is cast in a form more suitable for iterations as well as for the method of renormalization of collision probabilities used in the work. The analysis of several benchmark problems has been carried out and the results obtained using the new code are presented.

Mechanical performance of SiC three-layer cladding in PWRs

The silicon carbide cladding concept is currently under investigation with regard to increasing the accident tolerance and economic performance of light-water reactor fuels. In this work, the stress fields in the multi-layered silicon carbide cladding for LWR fuels are calculated using the commercial finite element analysis software ADINA. The material properties under irradiation are implemented as a function of temperature. The cladding is studied under operating and accident conditions, specifically for the loss-of-coolant accident (LOCA). During the LOCA, the blowdown and the reflood phases are modeled, including the quench waterfront. The calculated stresses along the cladding thickness show a high sensitivity to the assumptions regarding material properties. The resulting stresses are compared with experimental data and the probability of failure is calculated considering a Weibull model.

Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial (R-Z) or plane radial-circumferential (R-θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

Thermochemical Compatibility and Oxidation Resistance of Advanced LWR Fuel Cladding

The thermochemical compatibility of potential replacement fuel rod cladding materials for zirconium alloys in light water reactors was assessed. Considered were the family of FeCrAl alloys, Nb-1%Zr (similar to PWC-11), and a hybrid SiC-composite with a metallic barrier layer. The niobium alloy was also seen as requiring an oxidation protective layer, and a diffusion silicide was investigated. Metallic barrier layers for the SiC-composite reviewed included an FeCrAl alloy, Nb-1%Zr, and chromium. Thermochemical calculations were performed to determine oxidation behavior of the materials in steam and for hybrid SiC-composites to determine possible interactions between the metallic layer and SiC. In addition, experimental exposures of SiC-FeCrAl alloy reaction couples at 673, 1073, and 1273 K for 168 h in an inert atmosphere were made, and microanalysis was performed. Whereas all materials were determined to oxidize under higher oxygen partial pressures in the steam environment, these varied by material. The computed and experimental results indicate the formation of liquid phase eutectic in the FeCrAl-SiC system at the higher temperatures.

원형 핀과 구멍의 접촉에서 헤르츠 응력장 가정을 위한 조건

LWR Fuel Development Division, Korea Atomic Energy Research Institute(Receive August 15, 2015; Revised September 20, 2015; Accepted September 21, 2015)Abstract − This paper focuses on the conformal contact problem. A typical example of conformal contact is thecontact between a pin and hole. In particular, this paper focuses on the condition for assuming a contact stressfield to be a Hertzian pressure profile by using well-known classical solutions associated with Hertzian contact.Persson first developed the conformal contact analysis method around half a century ago, but there have beenno significant improvements since then. The present research also adopted this method, but developed new solu-tions from the viewpoint of application to structural design. The analysis began with a comparison between Per-sson°Os conformal contact stress and the Hertzian stress fields. The next step was to check the differences inthe normalized stress values of both. This study used the tolerance for the difference in the peak stresses of Per-sson°Os solution and the Hertz solution to validate the Hertzian assumption. This gave the range for the dif-ference in radii of the pin and hole when the contact force and mechanical properties of the material are specified.The results showed that, at a tolerance of 5%, the Hertzian assumption is valid if half of the contact angle is lessthan 35°ý. In addition, the Hertzian assumption holds even for a relatively long contact length, in contrast to thegeneral incomplete contact problem. This paper discusses these results along with other aspects of the applicationto the design.Keywords − conformal contact(등각 접촉), contact angle(접촉 각), contact length(접촉 길이), Hertzian stressassumption(헤르츠 응력장 가정), contact mechanics(접촉역학)

Feedback on neutron capture cross sections of 238Pu and 241Am from analyses of measured isotopic compositions of irradiated LWR fuels and MOX core physics experiments

The atomic fractions of 238Pu and 241Am in MOX fuels recycled in light water reactors are 1% to 2% and not significant compared with those of major Pu isotopes. On the other hand, recent evaluated nuclear data libraries, such as JENDL-4.0 and JEFF-3.2, give noticeably different thermal and epithermal neutron capture cross sections for 238Pu and 241Am. The thermal neutron capture cross sections of 238Pu and 241Am in JEFF-3.2 are 31% and 9% larger than those of JENDL-4.0, respectively. This paper shows the effect of the differences in the neutron cross sections on analysis results of two different integral experiments. The first is the isotopic compositions of 238Pu on UO2 and MOX fuels irradiated in BWR and PWR, and the second is the critical experiments of the water moderated cores fully loaded with MOX fuels. The former was analyzed by using the continuous energy Monte Carlo burnup calculation code MVP-BURN and the latter by the continuous energy Monte Carlo calculation code MVP. The comparisons between the calculated and measured results indicate that the most likely thermal and epithermal neutron capture cross sections of 238Pu and 241Am should be around at the middle between those of JEFF-3.2 and JENDL-4.0.

Towards the inclusion of open fabrication porosity in a fission gas release model

A model is proposed for fission product release in oxide fuels that takes into account the open porosity in a mechanistic manner. Its mathematical framework, assumptions and limitations are presented. It is based on the model for open porosity in the sintering process of crystalline solids. More precisely, a grain is represented by a tetrakaidecahedron and the open porosity is represented by a continuous cylinder along the grain edges. It has been integrated in the TRANSURANUS fuel performance code and applied to the first case of the first FUMEX project as well as to neptunium and americium containing pins irradiated during the SUPERFACT experiment and in the JOYO reactor. The results for LWR and FBR fuels are consistent with the experimental data and the predictions of previous empirical models when the thermal mechanisms are the main drivers of the release, even without using a fitting parameter. They also show a different but somewhat expected behaviour when very high porosity fuels are irradiated at a very low burn-up and at low temperature.

Void Reactivity Coefficient Analysis during Void Fraction Changes in Innovative BWR Assemblies

The study of the void reactivity variation in innovative BWR fuel assemblies is presented in this paper. The innovative assemblies are loaded with high enrichment fresh UO2and MOX fuels. UO2fuel enrichment is increased above existing design limitations for LWR fuels (>5%). MOX fuel enrichment with fissile Pu content is established to achieve the same burnup level as that of high enrichment UO2fuel. For the numerical analysis, the TRITON functional module of SCALE 6.1 code with the 238-group ENDF/B-VI cross section data library was applied. The investigation of the void reactivity feedback is performed in the entire 0–100% void fraction range. Higher values of void reactivity coefficient for assembly loaded with MOX fuel are found in comparison with values for assembly loaded with UO2fuel. Moreover, coefficient values for MOX fuel are positive over 75% void fraction. The variation of the void reactivity coefficient is explained by the results of the decomposition analysis based on four-factor formula and neutron absorption reactions for main isotopes. Additionally, the impact of the moderation enhancement on the void reactivity coefficient was investigated for the innovative assembly with MOX fuel.