Pellet Periphery Research Articles

Putting seed traits into pellets: Using seed mass data to improve seed encapsulation technology for native plant revegetation

Abstract Poor seedling emergence often limits the success of direct seeding in ecological restoration. New techniques for maximising seed use efficiency and seedling emergence are needed to help meet global targets for nature repair in the UN Decade on Restoration. Extruded pellets are widely used in agriculture and represent a promising advancement in seed‐based restoration. However, extruded pellets must be optimised for diverse suites of native species that possess a range of seed sizes and morphotypes. We investigated how seed mass affects the performance of native plant seeds (total % seedling emergence) when encapsulated in extruded pellets designed for revegetation. Two glasshouse trials were undertaken using seeds from 30 native Australian plant species. In Trial 1, we encapsulated seeds in the centre of pellets and determined the relationship between seed mass and emergence. In Trial 2, we encapsulated seeds nearer the periphery of pellets and determined whether the position of seeds (central vs. peripheral) affected emergence for a subset of 10 small‐seeded species. In both trials, emergence from pellets was compared to an optimal, bare‐seeded control to identify any barriers to seed encapsulation under well‐watered conditions. In Trial 1, when seeds were centrally encapsulated, emergence was generally higher for bare‐seeded controls relative to pelleted seeds. However, seed mass predicted emergence when seeds were encapsulated in the pellet centre (R2 = 0.32, p = 0.002), such that larger‐seeded species tended to have higher emergence than smaller‐seeded species. In Trial 2, encapsulating seeds nearer the pellet periphery (relative to the centre) resulted in an average 28‐fold increase in emergence for all 10 small‐seeded species. For half of the small‐seeded species trialled, emergence from the pellet periphery was equivalent to that of bare‐seeded controls. Synthesis and applications: Collectively, our results demonstrate: (1) a positive relationship between seed mass and emergence under central‐encapsulation, and (2) that emergence can be significantly improved for small‐seeded species when seeds are positioned nearer the pellet periphery. Translation of these findings into practice can help optimise emergence outcomes for native species with different sized seeds and nuanced germination requirements.

Spent Nuclear Fuel oxidation under dry storage controlled conditions for studying its radial oxidation behavior

In this work we have studied, for the first time, the oxidation of highly-irradiated Spent Nuclear Fuel (SNF) at in-situ conditions, with a controlled temperature and atmosphere, and measuring in the same experiment the oxidation in both the pellet center and the rim. The oxidation of the SNF matrix is of paramount importance from the nuclear fuel safe storage point of view, mainly because the potential threat that the formation of the oxidized phase U3O8 entails. In this work a measurement protocol for following in-situ the oxidation of SNF at controlled conditions (350 °C, constant air flow), based on the use of Raman spectroscopy as the measuring technique, is presented. The installation of the set-up inside a hot cell has been successfully applied. The radial surface oxidation of the irradiated fuel sample (63 MWd/kgU) is followed. In the center, U3O8 is obtained after 62 h. In the pellet periphery, the first spectrum of U3O8 is not observed until almost 92 h, although even after that not-oxidized spectra are still acquired. These results prove that the higher local burn-up and increase of fission products could enhance the oxidation resistance of the fuel in these conditions. This is the first time that this effect is proved in a temperature-controlled oxidation, single experiment.

High-burnup fuel stress analysis prior to and during a LOCA transient

High-burnup fuel fragmentation and pulverization have been experimentally observed during simulated loss-of-coolant accident (LOCA) transient testing of high-burnup (>62 GWd/tU) nuclear fuel. The cause of this phenomenon is generally accepted to be the formation of stresses arising from the off-normal conditions. However, the primary mechanism(s) contributing to the stress formation is/are still being debated. Leading hypotheses suggest that the concurrence of fission gas bubbles and high burnup structure on the pellet periphery (i.e., the rim) leave the fuel susceptible to pulverization. Other hypotheses suggest that pulverization is artificially inflated during externally heated tests that do not reproduce prototypic fuel stresses and radial temperature profiles as the fuel transitions from full-power nuclear heating to decay heating. This work performs finite element analyses of fuel pellet temperatures and macroscopic stresses prior to and during a LOCA transient to identify (1) operating conditions that may exacerbate the fuel pellet stress state, (2) differences in predicted stresses with nuclear vs. furnace heating, and (3) how fuel cracking impacts the subsequent stress state in the pellet fragments. Analyses predict that macroscopic transient fuel stresses resulting from constrained thermal expansion will remain below the fracture stress of fresh UO 2 once the fragment size reaches ~4 mm or less. Calculated fuel stresses from simulated LOCA tests performed with representative nuclear vs. electrical heating rates are shown to be equal within 5% and essentially unchanged for all times after the first 27 seconds of the transients. Because the fuel stresses do not change once the temperature profile becomes uniform, macroscopic transient stresses would be no different than those following a normal shutdown. Because pulverization of high burnup fuel does not occur prior to a LOCA test, pulverization of high burnup UO 2 cannot be initiated by differential thermal stresses.

Spectroscopic and chemical investigations on volatile fission and activation products within the fuel-cladding interface of irradiated pressurised water reactor fuel rod segments

Abstract. In Germany, the present waste management concept foresees the direct disposal of spent nuclear fuel (SNF) in deep geological repositories for high-level waste available by 2050, at best. Until then, SNF is encapsulated in dual-purpose casks and stored in dry interim storage facilities. Licenses for both casks and facilities will expire after 40 years following loading of the cask and emplacement of the first cask in the storage location. Yet, due to considerable delays in the site selection process and the estimated duration for construction and commissioning of a final repository of at least 2 decades, a prolonged dry interim storage of SNF is inevitable (ESK, 2015). Concerning these considerable timespans, integrity of the cladding is of utmost importance regarding the ultimately conditioning of the fuel assemblies for final disposal. Various processes strain the structural integrity of Zircaloy cladding during reactor operation and beyond such as delayed hydride cracking, fuel-cladding chemical interactions or irradiation damage induced by α-emitters present in the fuel pellet's rim zone (Ewing, 2015). Especially with higher burn-up, the gap between fuel and cladding closes and results in the formation of an interaction layer, in which precipitates of fission and activation products are present, displaying an interface for degradation processes. For chemical analysis and speciation of these agglomerates, Zircaloy-4 and SNF specimens were sampled from fuel rod segments irradiated in commercial pressurised water reactors during the 1980s. Zircaloy-4 specimens were taken from an UOX (50.4 GWdtHM-1) and mixed oxide fuel (MOX) (38.0 GWdtHM-1). In addition, SNF fragments were sampled from the closed gap of both fuel types to examine volatile activation and fission products, which had been segregated from the centre to the pellet periphery during irradiation and thus contribute to the possible chemically assisted cladding degradation effect of the precipitates within the fuel-cladding interface. Spectroscopic analysis of precipitates within the interface layer between fuel and cladding were performed by optical microscopy, X-ray absorption and X-ray photoelectron spectroscopy, as well as by energy-dispersive scanning electron microscopy. Moreover, the radionuclide inventory of the respective Zircaloy-4, fuel and interaction layers was determined using liquid scintillation counting, γ-spectroscopy, gas mass spectrometry, ion chromatography and inductive-coupled plasma mass spectrometry and compared to results received by MCNP/CINDER and webKORIGEN calculations. In this study, we provide results regarding the speciation and chemical composition of previously identified Cs-U-O-Zr-Cl-I bearing compounds found in the interaction layer of irradiated nuclear fuel and inventory analyses of radionuclides present therein, with particular emphasis on Cl-36 and I-129. Furthermore, the agglomerates within the fuel-cladding interface were characterised for the first time utilising synchrotron radiation-based Cl K-edge and I K-edge measurements, resulting in compounds with structural similarities to CsCl and CsI. The outcomes obtained from this study provide further insights into the complex chemistry within the fuel-cladding interface with respect to the aging management and integrity of SNF under the conditions of interim storage. In future studies we will examine whether the different compounds at the fuel-cladding interface have the potential to affect the mechanical properties of Zircaloy cladding.

Investigation of the safety features of advanced PWR assembly using SiC, Zr, FeCrAl and SS-310 as cladding materials

In this work, SiC (Silicon carbide), FeCrAl (ferritic), SS-310 (stainless steel 310) and Zirconium are simulated by MCNPX (Monte Carlo N‐Particle eXtended) code as cladding materials in advanced PWR (Pressurized Water Reactor) assembly. A number of reactor safety parameters are evaluated for the candidate cladding materials as reactivity, cycle length, radial power distribution of fuel pellet, reactivity coefficients, spectral hardening, peaking factor, thermal neutron fraction and delayed neutron fraction. The neutron economy presented by Zr and SiC models is analyzed through the burnup calculations on the unit cell and assembly levels. The study also provided the geometric conditions of all cladding materials under consideration in terms of the relation between fuel enrichment and cladding thickness from the viewpoint to achieve the same discharge burnup as the Zircaloy cladding. It was found that the SiC model participated in extending the life cycle by 2.23% compared to Zr. The materials other than SiC largely decreased discharge burnup in comparison with Zircaloy. Furthermore, the claddings with lower capture cross-sections (SiC and Zr) exhibit higher relative fission power at the pellet periphery. The simulation also showed that using SiC with a thickness of 571.15 μm and 4.83% U-235 can satisfy the EOL irradiation value as Zr. For reactivity coefficient, the higher absorbing materials (SS-310 and FeCrAl) exhibit more negative FTCs, MTCs and VRCs at the BOL But, at the intermediate stages of burnup Zr and SiC have a strong trend of negative reactivity coefficients. Finally, the delayed neutron fraction of SiC and Zr models is the highest among all the four models.

A three-dimensional axial fuel relocation framework with discrete element method to support burnup extension

Under normal operation, oxide nuclear fuel pellets subjected to high thermomechanical loading and irradiation damage can develop complex crack networks, separating the fuel pellet into a pile of fragments. During a Loss of Coolant Accident (LOCA), these pellet fragments, or fine powder especially when originating from the high-burnup fuel pellet periphery, may relocate axially into the ballooned region of the zircaloy cladding. In this work, the axial relocation process of fuel fragments and powder is modeled based on the three-dimensional Discrete Element Method (DEM) in the Abaqus software. A Voronoi tessellation technique is adopted to generate random convex polyhedrons in order to mimic a coarse fragmentation pattern. A parametric study in terms of the mean fragment size from 1.3 mm to 3.0 mm was conducted to investigate its effect on the axial profile of fuel mass fraction and filling ratio after fuel relocation. Then a sensitivity study of the randomness in fragmentation on axial fuel relocation was performed. For high-burnup fuels, in order to characterize the porous High Burnup Structure (HBS), the pellet rim was discretized into separate spherical shape particles. The effect of powder fraction on fuel relocation was demonstrated, and the results indicated that higher powder fraction can significantly increase the fuel mass fraction and filling ratio level in the ballooned region, which showed qualitative agreement with experimental observations. Finally, the verification against a one-dimensional theoretical relocation model was conducted to verify the robustness and high fidelity features of the DEM model in simulating axial fuel relocation under LOCA. The modeling approach serves as a mechanistic framework to support future burnup extensions for the oxide nuclear fuel.

Development of the MFPR/R code for characterization of the rim zone and high burnup structure evolution in UO2 fuel pellets

The MFPR/R code designed for mechanistic modelling of irradiated UO2 fuel, was further developed for characterization of the rim zone and high burnup structure (HBS) evolution in UO2 fuel pellets. The application of the MFPR/R semi-empirical criterion for fuel restructuring, related to the dislocation density threshold, allows calculating the penetration depth of HBS as a function of burnup, in a reasonable agreement with observations. For modelling of the porosity evolution in HBS at the UO2 fuel pellet periphery, the MFPR/R model for coalescence of randomly distributed immobile pores owing to their growth and impingement was further developed and validated. The advanced model considering multiple collisions provides a reasonable consent with experimental data, thus demonstrating the validity of the proposed pore coarsening mechanism in HBS.

Assembly-level analyses of accident-tolerant cladding concepts for a long-life civil marine SMR core using micro-heterogeneous duplex fuel

In this reactor physics study, we examine the neutronic performance of accident-tolerant fuel (ATF) claddings – austenitic type 310 stainless steel (310SS), ferritic Fe-20Cr-5Al (FeCrAl), advanced powder metallurgic ferritic (APMT), and silicon carbide (SiC)-based materials – as alternative cladding materials compared with Zircaloy-4 (Zr) cladding. The cores considered use 18% 235U enriched micro-heterogeneous ThO2-UO2 duplex fuel and, for purposes of comparison, 15% 235U enriched homogeneously mixed all-UO2 fuel, loaded into 13×13 pin arrays. A constant cladding coating thickness of 655 μm is assumed. We use the WIMS reactor physics code to analyse the associated reactivity, achievable discharge burnup, spectral variations, rim effect and reactivity feedback parameters for the candidate cladding materials at the assembly level.The results show that candidate fuels with 310SS cladding exhibit a ∼13% discharge burnup penalty compared to Zr due to the presence of a very high nickel (Ni) concentration. The high neutron absorption cross-sections of iron (Fe) in the FeCrAl and APMT claddings also lead to a ∼10% discharge burnup penalty. The fuels with SiC cladding can achieve a ∼1% higher discharge burnup compared to Zr due to the low thermal neutron absorption cross-sections of its constituents and the softer neutron spectrum. The claddings with lower capture cross-sections (SiC and Zr) exhibit higher relative fission power at the pellet periphery. For both candidate fuels, the end-of-life 239Pu (for UO2 fuel) and 233U (for duplex fuel) inventories are higher for the claddings (Fe-based: FeCrAl, APMT and steel-based: 310SS) with higher thermal capture cross-sections, unlike for SiC and Zr, where SiC provides higher end-of-life 239Pu and 233U inventories despite having lower capture cross-section than that of the Zr. Reactivity feedback parameter values (moderator and fuel temperature coefficients) are more negative for the duplex fuel than the UO2 fuel for all the candidate claddings, with claddings with harder spectra exhibiting more negative values. The duplex fuel yields a softer spectrum than the UO2 fuel with the candidate claddings, which improves neutron economy and thus discharge burnup.

Neutronic analysis for accident tolerant cladding candidates in CANDU-6 reactors

The neutronic penalty and advantages are quantified for potential accident tolerant claddings in a CANDU-6 (Canada Deuterium Uranium) reactor. Ferritic-based alloy (FeCrAl and APMT), steel-based alloy (304SS and 310SS) and silicon carbide (SiC) claddings are compared with Zircaloy-II. High thermal capture in nickel-59 and iron-56 imply a neutronic penalty in iron and steel-based bundles. A minimum enrichment of 1.0% and 1.1% is required for ferritic and steel-based claddings, respectively, to achieve the CANDU-6 burnup average criticality. An average increase of 0.35% in reactivity is introduced when SiC is considered, while keeping a natural enrichment condition. Average thermal neutron absorption rate is found to be 203, 8 and 6 times higher in UO2pellets than in silica, iron and steel-based claddings, respectively. A spectral hardening is observed in fuel, cladding and coolant for all enriched cells. Neutron flux is 30.5% higher at 0.0253 eV in SiC and Zr bundles. For FeCrAl, APMT, 304SS and 310SS, less 239Pu is produced during all fuel residence time and a minimum of 50% more 135Xe poison is being produced at equilibrium. As well, a minimum of 60% more 235U is left at end-of-life. For bundles cladded with SiC and Zr, 7.6% higher fission rates are found on the pellet periphery. Decreasing the cladding thickness by 200 μm made it possible to satisfy the criticality requirements for all claddings with an enrichment below 1%. Moderator and coolant temperature coefficients are found higher in SiC and Zr bundles. At mid-burnup, the Doppler effect and the voiding effect are higher in FeCrAl, APMT, 304SS and 310SS cladded cells.

Modelling of pore coarsening in the high burn-up structure of UO2 fuel

The model for coalescence of randomly distributed immobile pores owing to their growth and impingement, applied by the authors earlier to consideration of the porosity evolution in the high burn-up structure (HBS) at the UO2 fuel pellet periphery (rim zone), was further developed and validated. Predictions of the original model, taking into consideration only binary impingements of growing immobile pores, qualitatively correctly describe the decrease of the pore number density with the increase of the fractional porosity, however notably underestimate the coalescence rate at high burn-ups attained in the outmost region of the rim zone. In order to overcome this discrepancy, the next approximation of the model taking into consideration triple impingements of growing pores was developed. The advanced model provides a reasonable consent with experimental data, thus demonstrating the validity of the proposed pore coarsening mechanism in the HBS.

On the condition of UO2 nuclear fuel irradiated in a PWR to a burn-up in excess of 110 MWd/kgHM

Post-irradiation examination results are presented for UO2 fuel from a PWR fuel rod that had been irradiated to an average burn-up of 105 MWd/kgHM and showed high fission gas release of 42%. The radial distribution of xenon and the partitioning of fission gas between bubbles and the fuel matrix was investigated using laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS) and electron probe microanalysis. It is concluded that release from the fuel at intermediate radial positions was mainly responsible for the high fission gas release. In this region thermal release had occurred from the high burn-up structure (HBS) at some point after the sixth irradiation cycle. The LA-ICP-MS results indicate that gas release had also occurred from the HBS in the vicinity of the pellet periphery. It is shown that the gas pressure in the HBS pores is well below the pressure that the fuel can sustain.

Effect of Fibroblast Growth Factor 2 on Equine Synovial Fluid Chondroprogenitor Expansion and Chondrogenesis.

Mesenchymal stem cells have been identified in the synovial fluid of several species. This study was conducted to characterize chondroprogenitor (CP) cells in equine synovial fluid (SF) and to determine the effect of fibroblast growth factor 2 (FGF-2) on SF-CP monolayer proliferation and subsequent chondrogenesis. We hypothesized that FGF-2 would stimulate SF-CP proliferation and postexpansion chondrogenesis. SF aspirates were collected from adult equine joints. Colony-forming unit (CFU) assays were performed during primary cultures. At first passage, SF-cells were seeded at low density, with or without FGF-2. Following monolayer expansion and serial immunophenotyping, cells were transferred to chondrogenic pellet cultures. Pellets were analyzed for chondrogenic mRNA expression and cartilage matrix secretion. There was a mean of 59.2 CFU/mL of SF. FGF-2 increased the number of population doublings during two monolayer passages and halved the population doubling times. FGF-2 did not alter the immunophenotype of SF-CPs during monolayer expansion, nor did FGF-2 compromise chondrogenesis. Hypertrophic phenotypic markers were not expressed in control or FGF-2 groups. FGF-2 did prevent the development of a “fibroblastic” cell layer around pellet periphery. FGF-2 significantly accelerates in vitro SF-CP expansion, the major hurdle to clinical application of this cell population, without detrimentally affecting subsequent chondrogenic capacity.

An empirical formulation to describe the evolution of the high burnup structure

In the present work the behavior of fuel pellets for LWR power reactors in the high burnup range (average burnup higher than about 45MWd/kgU) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup. Gradually, a new microstructure develops in that ring, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behavior.It is generally accepted that the evolution of porosity in the high burnup structure (HBS) is determinant of the retention capacity of the fission gases rejected from the fuel matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity.Although the mechanisms governing the microstructural transformation have not been completely elucidated yet, some empirical expressions can be given, and this is the intention of the present work, for representing the main physical parameters. Starting from several works published in the open literature, some mathematical expressions were developed to describe the behavior and progress of porosity at local burnup values ranging from 60 to 300MWd/kgU. The analysis includes the interactions of different orders between pores, the growth of the pore radius by capturing vacancies, the evolution of porosity, pore number density and overpressure within the closed pores, the inventory of fission gas dissolved in the matrix and retained in the pores. The model is mathematically expressed by a system of non-linear differential equations.In the present work, results of this calculation scheme are compared with experimental data available in the open literature and with simulations performed by other authors. The results of these separate tests are quite satisfactory so, the next step will be the incorporation of this model as a new subroutine of the DIONISIO code, to expand the application range of this general fuel behavior simulation tool.

Coupled thermochemical, isotopic evolution and heat transfer simulations in highly irradiated UO2 nuclear fuel

Predictive capabilities for simulating irradiated nuclear fuel behavior are enhanced in the current work by coupling thermochemistry, isotopic evolution and heat transfer. Thermodynamic models that are incorporated into this framework not only predict the departure from stoichiometry of UO2, but also consider dissolved fission and activation products in the fluorite oxide phase, noble metal inclusions, secondary oxides including uranates, zirconates, molybdates and the gas phase. Thermochemical computations utilize the spatial and temporal evolution of the fission and activation product inventory in the pellet, which is typically neglected in nuclear fuel performance simulations. Isotopic computations encompass the depletion, decay and transmutation of more than 2000 isotopes that are calculated at every point in space and time. These computations take into consideration neutron flux depression and the increased production of fissile plutonium near the fuel pellet periphery (i.e., the so-called “rim effect”). Thermochemical and isotopic predictions are in very good agreement with reported experimental measurements of highly irradiated UO2 fuel with an average burnup of 102GWdt(U)−1. Simulation results demonstrate that predictions are considerably enhanced when coupling thermochemical and isotopic computations in comparison to empirical correlations.Notice: This manuscript has been authored by UT-Battelle, LLC, under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes.

Classical molecular dynamics investigation of microstructure evolution and grain boundary diffusion in nano-polycrystalline UO2

The High Burnup Structure (HBS) observed at pellet periphery in conventional Light Water Reactor nuclear fuels and around spots presenting high plutonium content in mixed (U, Pu) oxide fuel – MOX fuel – consists of a restructuration of the original grains into smaller ones. The process is often postulated to occur because of the accumulation of irradiation damage and the retention of fission products in the matrix. The computing power nowadays available enables for simulating larger systems at the atomic scale up to the point that nano-polycrystalline material can now be investigated by empirical potential molecular dynamics. Simulations of nano-polycrystalline UO2 structures have been carried out at various temperatures to investigate atom mobility close to grain boundaries. The variation of Arrhénius parameters for the diffusion coefficient of oxygen, uranium and xenon as a function of the distance from a grain boundary was studied, leading to the distinction of three zones: the grain boundary layers (up to 1nm depth) presenting enhanced diffusion, an intermediate zone (1 to roughly 2nm depth) with intermediate diffusion values and the bulk of the grains. The following Arrhénius relations for grain boundary diffusion were derived:Oxygendiffusion:DO(T)=5.4×10-4exp[(0.74eV)/(kBT)]cm2/sUraniumdiffusion:DU(T)=1.7×10-6exp[(0.50eV)/(kBT)]cm2/sXenondiffusion:DXe(T)=1.4×10-5exp[(0.50eV)/(kBT)]cm2/swhere kB stands for Boltzmann’s constant and T for temperature. Besides, changes in the microstructure could be “observed” over the 1ns duration of the simulations, such as the formation of small gas clusters.

Dynamic leaching studies of 48 MWd/kgU UO 2 commercial spent nuclear fuel under oxic conditions

The leaching of a high-burn-up spent nuclear fuel (48 MWd/KgU) has been studied in a carbonate-containing solution and under oxic conditions using a Continuously Stirred Tank Flow-Through Reactor (CSTR). Two samples of the fuel, one prepared from the centre of the pellet (labelled CORE) and another one from the fuel pellet periphery, enriched with the so-called High Burn-Up Structure (HBS, labelled OUT) have been used. For uranium and actinides, the results showed that U, Np, Am and Cm gave very similar normalized dissolution rates, while Pu showed slower dissolution rates for both samples. In addition, dissolution rates were consistently two to four times lower for OUT sample compared to CORE sample. Considering the fission products release the main results are that Y, Tc, La and Nd dissolved very similar to uranium; while Cs, Sr, Mo and Rb have up to 10 times higher dissolution rates. Rh, Ru and Zr seemed to have lower dissolution rates than uranium. The lowest dissolution rates were found for OUT sample. Three different contributions were detected on uranium release, modelled and attributed to oxidation layer, fines and matrix release.

CERAMOGRAPHY ANALYSIS OF MOX FUEL RODS AFTER AN IRRADIATION TEST

KAERI (Korea Atomic Energy Research Institute) fabricated MOX (Mixed Oxide) fuel pellets as a cooperation project with PSI (Paul Scherrer Institut) for an irradiation test in the Halden reactor. The MOX pellets were fitted into fuel rods that included instrumentation for measurement in IFE (Institutt for Energiteknikk). The fuel rods were assembled into the test rig and irradiated in the Halden reactor up to 50 MWd/kgHM. The irradiated fuel rods were transported to the IFE, where ceramography was carried out. The fuel rods were cut transversely at the relatively higher burn-up locations and then the radial cross sections were observed. Micrographs were analyzed using an image analysis program and grain sizes along the radial direction were measured by the linear intercept method. Radial cracks in the irradiated MOX were observed that were generally circumferentially closed at the pellet periphery and open in the hot central region. A circumferential crack was formed along the boundary between the dark central and the outer regions. The inner surface of the cladding was covered with an oxide layer. Pu-rich spots were observed in the outer region of the fuel pellets. The spots were surrounded by many small pores and contained some big pores inside. Metallic fission product precipitates were observed mainly in the central region and in the inside of the Pu spots. The average areal fractions of the metallic precipitates at the radial cross section were 0.41% for rod 6 and 0.32% for rod 3. In the periphery, pore density smaller than 2 μm was higher than that of the other regions. The grain growth occurred from 10 μm to 12 μm in the central region of rod 6 during irradiation.

Thermal diffusivity of homogeneous SBR MOX fuel with a burn-up of 35MWd/kgHM

The effect of burn-up on the thermal conductivity of homogeneous SBR MOX fuel is investigated and compared with standard UO2 LWR fuel. New thermal diffusivity results obtained on SBR MOX fuel with a pellet burn-up of 35 MWd/kgHM are reported. The thermal diffusivity measurements were carried out at three radial positions using a shielded “laser-flash” device and show that the thermal diffusivity increases from the pellet periphery to the centre. The fuel thermal conductivity was found to be in the same range as for UO2 of similar burn-up. The annealing behaviour was characterized in order to identify the degradation due to the out-of-pile auto-irradiation.

LOCAL BURNUP CHARACTERISTICS OF PWR SPENT NUCLEAR FUELS DISCHARGED FROM YEONGGWANG-2 NUCLEAR POWER PLANT

Spent UO₂ nuclear fuel discharged from a nuclear power plant (NPP) contains fission products, U, Pu, and other actinides. Due to neutron capture by 238 U in the rim region and a temperature gradient between the center and the rim of a fuel pellet, a considerable increase in the concentration of fission products, Pu, and other actinides are expected in the pellet periphery of high burnup fuel. The characterization of the radial profiles of the various isotopic concentrations is our main concern. For an analysis, spent nuclear fuels originating from the Yeonggwang-2 pressurized water reactor (PWR) were chosen as the test specimens. In this work, the distributions of some actinide isotopes were measured from center to rim of the spent fuel specimens by a radiation shielded laser ablation inductively coupled plasma mass spectrometer (LA-ICP-MS) system. Sampling was performed along the diameter of the specimen by reducing the sampling intervals from 500 μm in the center to 100 μm in the pellet periphery region. It was observed that the isotopic concentration ratios for minor actinides in the center of the specimen remain almost constant and increase near the pellet periphery due to the rim effect apart from the 236 U to 235 U ratio, which remains approximately constant. In addition, the distributions of local burnup were derived from the measured isotope ratios by applying the relationship between burnup and isotopic ratio for plutonium and minor actinides calculated by the ORIGEN2 code.

Identification of Radial Position of Fission Gas Release in High-Burnup Fuel Pellets under RIA Conditions

The radial positions of fission gas release (FGR) in high-burnup fuel pellets were examined after pulseirradiations that simulated reactivity-initiated accident (RIA) conditions in the Nuclear Safety Research Reactor (NSRR). The molar ratio of xenon (Xe) to krypton (Kr) (Xe/Kr ratio) in the released gas showed that fission gas was released from the entire region of the pellets of the examined PWR fuels during the pulse irradiations. On the other hand, most fission gas was released from the center and/or intermediate regions of the examined BWR fuel pellets. The analyses of the thermal stress distribution in fuel pellets during the pulse irradiations were carried out by means of a computer code, and the results supported the FGR positions that were estimated from the measured Xe/Kr ratios. Consequently, it is likely that fission gas is not released selectively from the rim structure at the pellet periphery under RIA conditions.