Post-irradiation Examination Research Articles

FAST irradiations and initial post irradiation examinations – Part I

The Advanced Fuels Campaign Fission Accelerated Steady-state Test (FAST) at Idaho National Laboratory (INL) completed its first irradiation cycle within the Advanced Test Reactor (ATR). The test focused on the irradiation of alloy fuel forms for use in sodium fast reactors. The first cycle of FAST testing was completed and four rodlets were removed for the initial post irradiation examination (PIE). The rodlet design and irradiation conditions were evaluated using Monte Carlo N-Particle (MCNP) for as-run power history and COMSOL for temperature analysis. These rodlets include a set of low burnups (∼2.5 % fissions per initial metal atoms [%FIMA]), control rodlets, and a helium-bonded annular rodlet (4.7 %FIMA). Non-destructive PIE has been completed and includes visual inspection, neutron radiography and gamma scanning of the FAST capsules and rodlets. Radiography confirmed the integrity of the experiments, revealed that the annulus in the annular fuel was filled at a modest burnup (4.7 %FIMA), and indicated potential slumping of the cooler rodlets at lower burnup. Precision gamma scanning indicated mostly usual fission product behavior, except for cesium in the He-bonded annular fuel. Future destructive PIE will be necessary to fully interpret the effects of accelerated irradiation on U-Zr metallic fuel behavior.

Thermal diffusivity in ion-irradiated single-crystal iron, chromium, vanadium, and tungsten measured using transient grating spectroscopy

The speed-up of radiation science development with the advent of ion-irradiation experiments has, until recently, been omitted in the post-irradiation examination technique. This paper reports the results of transient grating spectroscopy—a rapid, non-destructive, in situ photothermal surface technique—of ion-irradiated single-crystals of iron, chromium, vanadium, and tungsten at room temperature. Thermal diffusivity was used to track damage development throughout irradiation, with 5 MeV self-ion irradiated iron, chromium, and vanadium showing little to no change up to damages of the order of 1 dpa. 5 MeV Si3+-ion irradiated tungsten exhibits a reduction of thermal diffusivity from 0.78(7) to 0.29(2) cm2 s−1 with logarithmically increasing dose over a similar damage range. A comparison to literature of transient grating spectroscopy thermal diffusivity values past and present shows good agreement; radiation-induced change can be clearly distinguished from differences between mono- and poly-crystalline tungsten.

STUDY OF FISSION GAS BUBBLES AND INTERACTION LAYER ON IRRADIATED U3Si2-Al DENSITY OF 4.8 gU/cm3

STUDY OF FISSION GAS BUBBLES AND INTERACTION LAYER ON IRRADIATED U3Si2-Al DENSITY OF 4.8 gU/cm3. Uranium-silicide compound fuel dispersed in aluminium matrix (U3Si2-Al) have been used in a large number of research reactors around the world because of its excellent behavior under irradiation. This fuel also provides high uranium density with typical fuel loading up to 4.8 gU/cm3 to compensate for the reduced fissile amount in LEU. To improve the density of current U3Si2-Al (2.96 gU/cm3) used in Indonesian GA Siwabessy Multipurpose Research Reactor, U3Si2-Al dispersion fuel plate with density of 4.8 gU/cm3 (U235 ∼19.75%) had been irradiated in RSG GAS for 175 days at 15 MW power to burnup level of approximately 40%. The characterization was performed using SEM-EDS and optical microscope to study microstructure of the irradiatted fuel, largely the fission gas bubbles and the interaction layer between U3Si2 fuel and Al matrix. The average diameter of the bubbles with diameter from 0.06 to 0.55 µm was 0.21 µm. The interaction layer was identified as U(Al,Si)2,3 with thickness of approximately 1.5 µm. The relatively small fission gas bubbles and the interaction layer didn’t cause swelling on the fuel and the overall performance of the fuel plate was very good.Keyword: LEU, uranium-silicide, post-irradiation examination, interaction layer, fission gas bubbles.

Experiment Validation Protocol for Flux Wire Measurements in the Advanced Test Reactor

One of the Advanced Test Reactor’s (ATR’s) functions is to irradiate and qualify nuclear fuels and materials. Due to the large number of experiment or test positions, the cost, and the limited number of vessel penetrations for instrumentation, in-core instrumentation for most experiments is not feasible. In such instances, modeling of experiment conditions using high-fidelity neutron transport codes can quantify such conditions as fission power density and fissile material burnup during irradiation. Validation of fissile material burnup can only be performed during post-irradiation examination, which typically occurs months—or even years—following irradiation. In most experiments, fission power density and fissile material burnup are directly proportional to the thermal neutron flux in the ATR. Additionally, fast neutrons are born from fission in the ATR core, affording a validation of power distribution within the reactor’s experiment locations. During each irradiation cycle, flux wires installed throughout the ATR can be used to validate computational models and determine an adjusted neutron flux for many of the experiment positions. The flux wires are installed as requested by the experiment sponsors in several of the ATR flux traps and consist of cobalt-aluminum alloy and nickel wires. Both kinds of wire enable measurements of the thermal and fast neutron flux in each experiment position. This paper presents the protocol for validating computational models for experiments using flux wires installed in the experiment positions, as well as the results for flux wires placed in the ATR safety rod guide tubes. The best estimate is typically referred to as the adjusted neutron flux. The calculated unadjusted neutron flux is referred to as the a priori neutron flux. The methods presented here provide the adjusted neutron flux, given both the measured and a prior fluxes. The adjusted flux is compared to the a priori flux to provide a bias in the calculated results and the adjusted results. Two model types are evaluated; an eigenvalue case and fixed-source case. Both conditions demonstrate relatively good agreement. The uncertainty for the adjusted flux ranges from 5% to 6% for all three energy ranges. For the eigenvalue case, the bias between the a priori and the adjusted neutron flux is within the statistical uncertainty in all but two wire pairs. For the fixed-source model, four wire pairs are outside of the uncertainty of the adjusted flux. The bias between the a priori and adjusted fast neutron flux is outside of the statistical range for four wires in the eigenvalue case and nine wires in the fixed-source model. As the differences are not contained to one flux trap, it can be assumed that the biases in the calculated models are attributed to localized effects in modeling. An additional evaluation was performed for the ATF-1 experiment in the ATR “I” positions. The differences between the adjusted and a priori are more pronounced in two of the test positions, indicating that additional model evaluation is needed, in particular in the region near the boundary of the ATR model. It is also noted that the eigenvalue model provides slightly better results in the flux trap positions. The fixed-source model is more computationally efficient though produces less accurate results; the differences in some cases are negligible. The work documented in this paper provides a methodology that extends the validation protocol established at the ATR for flux measurements to validate computational models with limited measurement capability during a cycle.

Detailed characterization of a PWR fuel rod at high burnup in support of LOCA testing

Experimental investigations of the fuel microstructure and volatile fission products along the pellet of a high burnup specimen (local burnup 76 GWd/tHM) have been conducted to support future transient testing. Detailed microscopy examinations have been carried out at different length scales. Transmission electron microscopy has highlighted a significant amount of damage across the entire radius with the formation of networks of dislocations. The optical and scanning electron microscopy determined the formation of three zones in the pellet with different characteristics. An intermediate region with no grain subdivision, lower porosity than the central zone porosity and high retained fission gas in nanometric bubbles and in the matrix was present between r/r0 ≈ 0.55 and 0.8. The high retention of gas in this region might suggest that the region will be prone to fine fragmentation, in addition to the HBS. An abrupt transition in the structure was observed at mid radius, with a third region developing from the mid radius to the pellet center. In this part of the pellet, metallic and grey phases with size between hundreds of nanometers and a few micrometers have formed at grain boundaries. The majority of fission gas has been released from the grain matrix and the original grains have polygonised, forming sub-grain domains separated by low-angle grain boundaries. No final explanation can be given for the polygonization occurring in the center, but on the basis of the irradiation history and the analysis of all the post irradiation examination (PIE) data, it is postulated that the polygonization within the original grains is an effect of dynamic recovery occurring at high temperature in the fuel center.

BISON fuel performance modeling optimization for experiment X447 and X447A using axial swelling and cladding strain measurements

With the recent need to qualify new reactor designs, such as the Versatile Test Reactor, researchers need fuel performance calculations to determine the safety criteria of proposed designs. In order to validate the fuel performance results obtained by the fuel performance code BISON for new reactor designs, legacy fuel from EBR-II and Fast Flux Test Facility MFF with post-irradiation examination (PIE) data need to be used as validation cases to benchmark models. In this work, BISON has been paired with the Fuels Irradiation & Physics Database (FIPD) and the Integral Fast Reactor Materials Information System (IMIS) to supply PIE data for comparison with simulations of EBR-II experiments X447/X447A. X447/X447A were assessed by implementing models for a fuel-cladding chemical interaction (FCCI) within BISON and optimizing the friction coefficient between the fuel surface and the cladding, the anisotropic swelling factor, and the HT9 first thermal creep scalar (which scales the first term in the HT9 creep equation) to best match the PIE axial fuel swelling height and cladding profilometry for all pins in X447/X447A. The optimal values were found using a generic algorithm developed to select different values for the three parameters until end criteria was met and error couldn’t be reduced further. The BISON-simulated cladding profilometry was evaluated using a standard error of the estimate to account for the profile shape of the cladding profilometry. Optimal values for the friction coefficient, anisotropic fuel swelling factor, and HT9 first thermal creep scalar were found to best fit the BISON simulation results to the PIE measurements found in IMIS and FIPD. Possible improvements to current models to account for the underprediction of fuel swelling at low burnups and the overprediction of fuel swelling at higher burnups that is observed for the axial fuel swelling height are discussed. Although two pins in EBR-II, X447/X447A (DP70 and DP75), were known to fail due to FCCI, none of the pins simulated in BISON reached a cumulative damage fraction above 0.008 with FCCI correlations coupled in the BISON simulations. The error estimate generated for all pins in X447/X447A using optimal values was 209 µm, which is deemed acceptable.

High temperature nanoindentation of iron: Experimental and computational study

Application of reduced activation ferritic/martensitic (RAFM) steels as the structural material in future fusion reactors requires the knowledge of their mechanical properties under relevant operational conditions i.e. temperatures and irradiation by fast neutrons. Execution of the neutron irradiation and post irradiation examination is expensive and lengthy, therefore experimental and computational solutions to ease the characterization of as-irradiated materials are in the scope of interests of nuclear materials scientific community. Moreover, ion irradiation is considered as one possible way to surrogate high flux neutron irradiation damage. The extraction of the mechanical properties after ion irradiation primarily relies on the nanoindentation techniques and its subsequent post-processing to extract engineering relevant information, although some innovative techniques such as compression micropillars and micro-tensile testing also exist. In this work, we have performed nanoindentation on BCC iron, as the basis material for ferritic steels, by using a new Bruker stage developed for high temperature operation. The obtained results were analyzed by means of crystal plasticity finite element method (CPFEM), whereas the constitutive laws of the material were derived and established by using tensile deformation data, thus providing an interconnection of material's behavior under compressive and tensile deformations. The microstructural features such as indentation pile-up formation or dislocation density evolution were obtained by using transmission and scanning electron microscopy, and were compared with the predictions derived by the developed CPFEM model. It is demonstrated that a good agreement between the CPFEM and experimental data set, including tensile and compressive loads as well as associated microstructural changes, is obtained at room temperature and elevated temperature tests.

Irradiation behaviour of alloy D9 clad and wrapper in FBTR

Titanium modified austenitic stainless steel 316 (Alloy D9) is the material of clad and wrapper for the 500 MWe Prototype Fast Breeder Reactor (PFBR) at Kalpakkam, India. Towards evaluating the irradiation performance of the mixed oxide (MOX) fuel design and Alloy D9 structural material, a test fuel subassembly (FSA) made of Alloy D9 wrapper with 37 MOX fuel pins of Alloy D9 cladding was irradiated in Fast Breeder Test Reactor (FBTR) to a peak burnup of 112 GWd/t. The dimensional changes, swelling and mechanical properties of alloy D9 clad and wrapper at displacement damage levels of 40-62 dpa were evaluated through post-irradiation examinations (PIE) and compared with that of AISI 316 clad/wrapper of FBTR. This study has revealed superior performance of alloy D9 compared to AISI 316 counterparts and suggests that MOX fuel subassembly of PFBR with Alloy D9 cladding and wrapper can safely attain the initial target burnup of 100 GWd/t and corresponding displacement damage of 85 dpa.

Metallic Fuel Performance Benchmarks for Versatile Test Reactor Applications

The U.S. Department of Energy Office of Nuclear Energy’s Versatile Test Reactor (VTR) project is designing a new fast-spectrum test reactor. The VTR reference driver fuel design is sodium-bonded U-20Pu-10Zr (wt%) metallic fuel and HT-9 cladding. The BISON fuel performance code is being used to model the VTR driver fuel pin to evaluate the effects of differences between its design and the legacy designs that preceded it. This work summarizes ongoing efforts at Oak Ridge National Laboratory to benchmark BISON for VTR driver fuel analyses, including establishing metallic fuel performance code requirements for VTR applications and benchmarking BISON for VTR driver fuel analyses. Integral fuel pin predictions are compared to legacy calculations and post-irradiation examination data for 261 fuel pins irradiated at Experimental Breeder Reactor II and the Fast Flux Test Facility. The BISON predictions exhibit trends that are generally consistent with the legacy data. Burnup and temperature predictions were found to be more accurate than mechanical predictions such as radial cladding dilation, axial fuel elongation, and plenum pressure. Likely sources of error were identified for evaluation in future work.

Fuel Performance Modeling at High Burn-Up by FEMAXI-6 Code

Abstract The scope of the current research in the field of fuel performance is primarily aimed at an improvement of the operating reliability, safety, and cost-effectiveness of the reactors in operation. The current requirement of nuclear industry is to have fuel suitable for load follow operation. Fission gas release, pellet-cladding mechanical interaction, and stress corrosion cracking are the main phenomena that limit the variability of reactor operation from a safety perspective. To reasonably predict the fuel performance limits, it is necessary to benchmark the computational tools against high-quality experimental data. This work is devoted to the calculation of fuel performance using the code FEMAXI-6 based on the longest irradiation experiment in the Halden reactor. The fuel burn-up was approaching 90 MWd/kgUO2 in three selected rods, which were equipped with the pressure sensors and were subjected to extensive postirradiation examination. During the experiment, the rods were exposed to several periods of power cycling. The rods were manufactured with different fuel grain size and fuel-to-clad gap size.

Effect of water molecules co-intercalation and hydrolysis on the electrochemical deconsolidation of matrix graphite in aqueous nitric acid

Separation of complete TRISO-coated fuel particles from matrix graphite is challenging but important for the post-irradiation examination (PIE) of high-temperature gas-cooled reactor (HTGR) fuel elements. Electrochemical technology is considered one of the most promising and effective approaches in this regard. Herein, to explore the influence of HNO3 concentration on galvanostatic deconsolidation, 4%–68% HNO3 were taken as examples. The matrix graphite anodes gradually exfoliated to fragments in 34% HNO3, while expanding in other concentrations. The characterization results of deconsolidated fragments and galvanostatic curves identified the highest oxygen content was reached in 34% HNO3 due to the combined effect of intercalation, hydrolysis and oxygen evolution. In higher HNO3 concentrations like 68%, the intercalating agent predominates and therefore the intercalation and following hydrolysis reactions are the main causes of matrix graphite deconsolidation, with no oxygen evolution occurring. In contrast, in lower HNO3 concentrations like 4%, the violent oxygen evolution reaction hinders the intercalation and hydrolysis reactions. In short, only the combined interaction of intercalation, hydrolysis and oxygen evolution can deconsolidate the bulk matrix graphite into a more homogeneous powder. This work will provide more insights into the electrochemical deconsolidation mechanism and process of both HTGR fuel elements and other graphite materials.

Understanding fission gas bubble distribution, lanthanide transportation, and thermal conductivity degradation in neutron-irradiated α-U using machine learning

U 10Zr based metallic nuclear fuel is the leading candidate for next-generation sodium-cooled fast reactors in the United States. US research reactors have used and tested this fuel type since the 1960s and accumulated considerable experience and knowledge about the fuel performance. Most of the knowledge, however, remains empirical. The lack of mechanistic understanding of fuel performance puts a large burden on proof through experimental verification for the qualification of U 10Zr fuel for commercial use. This paper proposes an image data-driven machine learning approach, coupled with domain knowledge provided by advanced post irradiation examination, to provide unprecedented quantified insights into the morphology, size, density and the connectivity of fission gas bubbles and their effects on the fission product transportation and thermal conductivity. Specifically, we developed a method to automatically detect, extract statistics, and classify ~19,000 fission gas bubbles into different categories, and quantitatively link the data to lanthanide transportation through connected bubbles and degradation of thermal conductivity along the radial temperature gradient in a neutron irradiated U 10Zr annular fuel. Results indicate the approach can be modified to study other irradiation effects, such as secondary phase redistribution and gaseous fuel swelling in other irradiated nuclear fuels. • This study showcases a novel method utilizing an image data-driven machine learning approach for advanced post irradiation examination. • The features and statistics of 19,000 fission gas bubbles are derived by the decision tree algorithm. • The bubble statistics provide quantified insights into microstructural features of irradiated fuel and how they relate to fuel performance.

Recent progress in the assessment of irradiation effects for in-vessel fusion materials: tungsten and copper alloys

The present contribution highlights results of the recent irradiation campaigns applied to screen mechanical properties of advanced tungsten and copper-based materials—main candidates for the application in the plasma-facing components (PFC) in the European DEMO, which has also been presented at 28th IAEA fusion energy conference. The main challenges in the formulated irradiation programme were linked to: (I) assessment of the ductile-to-brittle transition temperature of newly developed tungsten-based materials; (ii) investigation of an industrial pure tungsten grade under high temperature irradiation, reflecting operational conditions in the high flux divertor region; (iii) assessment of the high temperature strength of CuCrZr-based alloys and composites developed to enable the extension of the operational window for the heat sink materials. The development and choice of the advanced materials is driven naturally by the need to extend the operation temperature/fluence window thereby enlarging the design space for PFCs. The obtained results helped identifying the prospective tungsten and copper-based material grades as well as yielded a number of unexpected results pointing at severe degradation of the mechanical properties due to the irradiation. The results are discussed along with the highlights of the microstructural examination. An outlook for near future investigations involving in-depth post-irradiation examination and further irradiation campaigns is provided.

High intensity proton beam impact at 440 GeV/c on Mo and Cu coated CfC/graphite and SiC/SiC absorbers for beam intercepting devices

Beam Intercepting Devices (BIDs) are essential protection elements for the operation of the Large Hadron Collider (LHC) complex. The LHC internal beam dump (LHC Target Dump Injection or LHC TDI) is the main protection BID of the LHC injection system; its main function is to protect LHC equipment in the event of a malfunction of the injection kicker magnets during beam transfer from the SPS to the LHC. Several issues with the TDI were encountered during LHC operation, most of them due to outgassing from its core components induced by electron cloud effects, which led to limitations of the injector intensity and hence had an impact on LHC availability. The absorbing cores of the TDIs, and of beam intercepting devices in general, need to deal with high thermo-mechanical loads induced by the high intensity particle beams. In addition, devices such as the TDI — where the absorbing materials are installed close to the beam, are important contributors to the accelerator impedance budget. To reduce impedance, the absorbing materials that make up the core must be typically coated with high electrical conductivity metals. Beam impact testing of the coated absorbers is a crucial element of development work to ensure their correct operation.In the work covered by this paper, the behaviour of several metal-coated absorber materials was investigated when exposed to high intensity and high energy proton beams in the HiRadMat facility at CERN. Different coating configurations based on copper and molybdenum, and absorbing materials such as isostatic graphite, Carbon Fibre Composite (CfC) and Silicon Carbide reinforced with Silicon Carbide fibres (SiC-SiC), were tested in the facility to assess the TDI's performance and to extract information for other BIDs using these materials. In addition to beam impact tests and an extensive Post Irradiation Examination (PIE) campaign to assess the performance of the coatings and the structural integrity of the substrates, extensive numerical simulations were carried out.

Validation of New Fission Yield by Analysis of Post-irradiation Examination in LWR

Validation of New Fission Yield by Analysis of Post-irradiation Examination in LWR

МОДЕЛИРОВАНИЕ ТЕРМОМЕХАНИЧЕСКОГО ПОВЕДЕНИЯ ТВЭЛОВ С ВЫГОРАЮЩИМ ПОГЛОТИТЕЛЕМ НА ОСНОВЕ ОКСИДА ГАДОЛИНИЯ В КОДЕ РТОП-СА

The paper presents a description of the RTOP-CA code models designed for simulating the behavior of (U, Gd)O2 fuel rods, presents the examples of program verification using experimental data obtained at the research reactor in Halden and in the post-irradiation examinations of WWER fuel. The experience of operating WWER fuel has shown that differences in the thermomechanical behavior of (U, Gd)O2 fuel elements and UO2 fuel rods during the first - second fuel cycles can be observed. In particular, an accelerated increase in the length of the (U, Gd)O2 fuel rod cladding as compared to UO2 fuel elements is possible. To simulate the behavior of (U, Gd)O2 fuel rods under irradiation in the RTOP-CA code, models that describe the features of uranium-gadolinium fuel were developed. To calculate the evolution of heat release in (U, Gd)O2 fuel rods in the RTOP-CA code, a model of burnup of gadolinium isotopes was implemented. The developed model makes it possible, without using of detailed neutron-physical calculations, to simulate the burnup of gadolinium and the evolution of heat generation along the radius of the pellets. The models take into account the effect of the gadolinium oxide content on the thermophysical and mechanical properties of the fuel. The calculations performed using the RTOP-CA code are in good agreement with the results obtained during the experiments in Halden on the dynamics of the central fuel temperature, gas pressure in the (U, Gd)O2 fuel rod, and the elongation of the fuel column and cladding. To simulate the accelerated elongation of (U, Gd)O2 fuel rods in comparison with standard WWER fuel elements at the initial stages of fuel operation at nuclear power plants in the RTOP-CA code, a model of mechanical interaction of pellets and cladding was developed taking into account the misaligned arrangement of pellets.

Front-end investigations of the coated particles of nuclear fuel samples – ion polishing method

The investigations of the coated-particles of nuclear fuel samples are carried out in three stages: front-end, irradiation in the reactor core, and post-irradiation examination. The front-end stage is the initial analysis of the failures rates of produced samples before they are placed in the reactor core. The purpose of the verification is to prepare the particles for an experiment that will determine the degree of damage to the coated particles at each stage. Before starting experiments with the samples, they must be properly prepared. Polishing the samples in order to uncover the inner layers is an important, initial experimental step. The authors of this paper used a novel way to prepare samples for testing - by applying an ion polisher. Mechanical polishing used frequently for sample preparations generates additional mechanical damages in the studied fuel particle, thus directly affecting the experimental results. The polishing methods were compared for three different coated particles using diagnostic methods such as Raman spectroscopy, scanning electron microscopy, and confocal laser scanning microscopy. Based on the obtained results, it was concluded that the ion polishing method is better because the level of interference with the structures of the individual layers of the tested samples is much lower than with the mechanical method. The same technique is used for the fuel particles undergone ion implantation simulating radiation damage that can occur in the reactor core.

Behavior of metallic fast reactor fuels during an overpower transient

● An extended overpower transient (∼32%) was conducted in EBR-II to test the response of 19 binary- and ternary-design metallic alloy fuel elements. ● Post-irradiation examination demonstrated that the overpower transient does not compromise the integrity of fuel elements. ● Post-transient fuel performance should not be significantly altered by the overpower transient. A slow-ramp (0.1%/s), extended overpower (∼32%) transient was conducted in the Experimental Breeder Reactor-II (EBR-II) on 19 metallic alloy fuel elements, including both binary (U-Zr) and ternary (U-Pu-Zr) fuel designs. The elements were clad in either Type 316 SS or HT9. Before the transient, the elements were pre-irradiated under steady-state or steady-state plus duty-cycle (periodic 15% overpower transient) conditions in various EBR-II sub-assemblies and reached burnups ranging from ∼5 to ∼12 at%. The fuel pins were then consolidated into the X512 sub-assembly for the overpower transient. Cladding integrity was maintained for all fuel pins throughout the entirety of normal operation and the overpower transient test. Post-irradiation examination (PIE) demonstrated that mild restructuring was evident within the fuel phase. However, these changes were not significant enough to expect changes in irradiation performance under normal operating conditions if the fuel elements were returned to service. These results empirically validate that limits on sodium fast reactor operational transients ( e.g. power ramps) could be applied with adequate safety margins to fuel failure or non-reversible damage.

Metallic fuel cladding degradation model development and evaluation for BISON

Fuel cladding chemical interaction (FCCI) and coolant cladding chemical interaction (CCCI), both consume stainless-steel cladding in sodium-cooled fast reactors (SFRs), and effectively degrade cladding mechanical performance through thickness reduction. Therefore, FCCI/CCCI and corresponding cladding degradation models are essential for advanced fuel performance codes, such as BISON, to reliably predict cladding behavior in SFRs. Here, we report the development and implementation efforts for BISON’s FCCI/CCCI correlations and cladding degradation models based on U.S. legacy metallic fuel data. The models were evaluated using Integral Fast Reactor (IFR) program X447/A experiment data supported by the ongoing integration project enabling coordinated application of BISON and EBR-II fuel irradiation and physics database (FIPD). The implemented models were demonstrated to improve BISON’s capabilities of predicting cladding damage and degradation behavior that is more consistent with post-irradiation examination observations. Additionally, some limitations of current BISON modules are identified, which are to be overcome through the BISON-FIPD integration efforts.

On the Limitations of Positron Annihilation Spectroscopy in the Investigation of Ion-Implanted FeCr Samples

New materials for advanced fission/fusion nuclear facilities must inevitably demonstrate resistance to radiation embrittlement. Thermal and radiation ageing accompanied by stress corrosion cracking are dominant effects that limit the operational condition and safe lifetime of the newest nuclear facilities. To study these phenomena and improve the current understanding of various aspects of radiation embrittlement, ion bombardment experiments are widely used as a surrogate for neutron irradiation. While avoiding the induced activity, typical for neutron-irradiated samples, is a clear benefit of the ion implantation, the shallow near-surface region of the modified materials may be a complication to the post-irradiation examination (PIE). However, microstructural defects induced by ion implantation can be effectively investigated using various spectroscopic techniques, including slow-positron beam spectroscopy. This method, typically represented by techniques of positron annihilation lifetime spectroscopy and Doppler broadening spectroscopy, enables a unique depth-profile characterisation of the near-surface region affected by ion bombardment or corrosion degradation. One of the best slow-positron beam facilities is available at the pulsed low-energy positron system (PLEPS), operated at FRM-II reactor in Munich (Germany). Bulk studies (such as high energy ion implantation or neutron irradiation experiments) can be, on the other hand, effectively performed using radioisotope positron sources. In this paper, we outline some basics of the two approaches and provide some recommendations to improve the validity of the positron annihilation spectroscopy (PAS) data obtained on ion-irradiated samples using a conventional 22Na positron source.