Release Of Gaseous Fission Products Research Articles

Flow visualization experiments of argon injection in a molten salt natural circulation loop

Off-gas systems are implemented in molten salt reactor designs to control the release of gaseous fission products. Two-phase flow in molten salt must be studied to understand how the system will behave in comparison to traditional working fluids like water. Flow visualization experiments and particle image velocimetry measurements were performed for three argon bubble sizes injected into a co-current stream of molten salt in a natural circulation loop facility. Similar bubble sizes were injected in experiments with water to compare the bubble shape, trajectory, and wake flow behavior of the fluids. The bubble region of interest was used to calculate the equivalent diameter and terminal velocity. Results for water showed a wobbling bubble surface and less stable bubble trajectory due to lower surface tension and viscosity compared with molten salt. Particle image velocimetry results demonstrated the increased viscosity of salt dampens turbulent fluctuations for the smaller bubble size. For a cap bubble, turbulent fluctuations were larger and longer lasting than in results for the wake flow of an argon cap bubble in water.

Study on the effect of long-term high temperature irradiation on TRISO fuel

In the core of the WWR-K reactor, a long-term irradiation of tristructural isotopic (TRISO)-coated fuel particles (CFPs) with a UO2 kernel was carried out under high-temperature gas-cooled reactor (HTGR)-like operating conditions. The temperature of this TRISO fuel during irradiation varied in the range of 950–1100 °C. A fission per initial metal atom (FIMA) of uranium burnup of 9.9% was reached. The release of gaseous fission products was measured in-pile. The release-to-birth ratio (R/B) for the fission product isotopes was calculated. Aspects of fuel safety while achieving deep fuel burnup are important and relevant, including maintaining the integrity of the fuel coatings. The main mechanisms of fuel failure are kernel migration, silicon carbide corrosion by palladium, and gas pressure increase inside the CFP. The formation of gaseous fission products and carbon monoxide leads to an increase in the internal pressure in the CFP, which is a dominant failure mechanism of the coatings under this level of burnup. Irradiated fuel compacts were subjected to electric dissociation to isolate the CFPs from the fuel compacts. In addition, nondestructive methods, such as X-ray radiography and gamma spectrometry, were used. The predicted R/B ratio was evaluated using the fission gas release model developed in the high-temperature test reactor (HTTR) project. In the model, both the through-coatings of failed CFPs and as-fabricated uranium contamination were assumed to be sources of the fission gas. The obtained R/B ratio for gaseous fission products allows the finalization and validation of the model for the release of fission products from the CFPs and fuel compacts. The success of the integrity of TRISO fuel irradiated at approximately 9.9% FIMA was demonstrated. A low fuel failure fraction and R/B ratios indicated good performance and reliability of the studied TRISO fuel.

Raman spectroscopy coupled to principal component analysis for studying UO2 nuclear fuels with different grain sizes due to the chromia addition

Current necessities of nuclear power plants have led to increase burn-up of the fuel during operation. In this context, some undesirable processes, such as a higher release of gaseous fission products, may occur. New UO2-based fuels are being developed by adding doping agents. Chromia (Cr2O3) has proved to considerably enhance grain growth during sintering, and thus promote the retention of these fission products. The study of these Cr-doped UO2 fuels at in situ conditions would allow testing the real performance of the fuel in operating conditions. The existence of Raman portable instruments makes this scenario feasible. But first, the measurement protocols need to be developed. Therefore, in this paper, we explore the use of Raman spectroscopy as an approach for the analysis of the effect of Cr2O3 addition in a set of UO2Cr2O3 sintered pellets. The validity of Raman is demonstrated by using Principal Component Analysis (PCA). Three Principal Components describe 98.8% of the total variance of the data, and they are related to the main Raman modes of the samples. In addition, SEM images have shown the presence of bigger precipitates of Cr2O3 not only when the solubility limit of Cr3+ is exceeded, but also when it is not reached. By XRD, the well-known Vegard behavior is observed, and a solubility limit of Cr3+dissolved into the UO2 matrix is found to be (748±16) ppm for the particular sintering conditions used.

The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 2: Validation and Verification

The article presents information on the validation and verification (V&V) of the first version (V1) of the EUCLID integrated code intended for safety analysis of operating or designed liquid metal (sodium, lead, or lead–bismuth) cooled reactors under normal operation and under anticipated operational occurrences by carrying out interconnected neutronics, thermal–mechanical, and thermal–hydraulic calculations. The list of processes and phenomena that have to be modeled in the integral code for correctly describing the above-mentioned operating conditions is given. Based on this list, the most high-quality experimental data are selected for carrying out the validation. It is shown that, for sodium cooled reactors, a significant number of experiments was carried out around the world on studying individual thermal–hydraulic processes and phenomena, which made it possible to perform validation of the thermal–hydraulic module. The validation of the code—as applied to description of processes that take place in fuel rods with oxide or nitride fuel and gas gap—is carried out against the results of post-pile investigations of fuel rods irradiated in fast sodium cooled research and power-generating reactors. The obtained results opened up the possibility to determine the errors of calculating such fuel rod parameters as release of gaseous fission products from the fuel and sizes of pellet and cladding in a limited range of burnup values. To perform validation of the neutronics module as applied to calculation of such parameters as power density distribution over the core and decay heat release, a sufficient number of experiments and benchmarks were selected. The results obtained from experimental operating conditions of a BN-600 reactor and startup conditions of a BN-800 reactor made it possible to estimate how correctly the integral code performs calculations of interconnected thermal–hydraulic and neutronic processes. Only a limited set of experimental investigations is available for heavy liquid metal cooled reactors. In view of this circumstance, programs for obtaining the lacking data are developed. To estimate the quality with which the experiments are modeled by means of the EUCLID/V1 integrated code, a procedure for evaluating the errors of calculation results is developed. In accordance with this procedure, the error of calculating the parameters playing the main role in the reactor safety assessment is evaluated.

Melting and multi-phase flow modelling of nuclear fuel in fast reactor fuel rod

Melting of nuclear fuel in a fast breeder reactor is accompanied by simultaneous release of gaseous fission products. These gases are trapped inside porosities present in the nuclear fuel and escape upon melting. The simultaneous release of fission gas and molten fuel results in a transient multi-phase flow. Initially, the fission gases are highly pressurized and exert hydrodynamic forces on the molten fuel. This in turn causes the molten fuel to undergo displacement. This displacement of molten fuel influences the reactor kinetics and the overall accident outcome. The present work is aimed at developing a mathematical model which simulates the melting and multi-phase flow phenomenon in a fast reactor fuel rod under a slow transient over-power accident scenario. The developed model is validated with the experimental data of the CABRI-E9 test. Thereafter, the behaviour of molten fuel is analysed with a nominal case study and parametric changes. Finally, the influence of molten fuel motion over reactor kinetics is analysed. The results show that under slow transients, molten fuel primarily relocates towards the lower portion of the fuel inner cavity. This behaviour of molten fuel can positively assist in the mitigation of a slow transient over-power accident.

Multiphysics modeling of novel UO2-BeO sandwich fuel performance in a light water reactor

The relatively poor thermal conductivity of the UO2 fuel is a major challenge for optimizing reactor operation and safety performance despite its widespread use in the majority of power reactors. The already low thermal conductivity further degrades with burnup due to accumulation of defects, fission product precipitates and fission gas bubbles. The high thermal stresses cause significant pellet cracking, leading to more pellet expansion and causing pellet cladding interaction and the release of fission product gases. Mitigation of these phenomena is currently accomplished by limiting operating power and ramp rates at high burnup. This study presents a novel design of the fuel pellet. The pellet was a combination UO2 fuel and ceramic BeO, and three types of geometry were proposed to mitigate the thermal stress. The corresponding fuel performances were compared based on the modified version of our previously developed CAMPUS code. The fuel pellets with BeO inner or middle layers in the radial direction were found to greatly decrease fuel centerline temperature, and mitigate the fuel and cladding mechanical interaction by delaying the gap closure time. So, the corresponding fuel performance and reactor safety would be improved. The fabrication cost of these two types of fuel pellets would also be decreased significantly compared to the UO2-BeO composite fuel due to the simplicity of the pellet design.

Multiphysics modeling of UO2-SiC composite fuel performance with enhanced thermal and mechanical properties

The relatively poor thermal conductivity of the UO2 fuel is a major challenge for optimizing reactor operation and safety performance, despite its widespread use in the majority of power reactors. The already low thermal conductivity further degrades with burnup due to accumulation of defects, fission product precipitates and fission gas bubbles. The high thermal stresses cause significant pellet cracking, leading to more pellet expansion and causing pellet cladding interaction and the release of fission product gases. Mitigation of these phenomena is accomplished by limiting operating power and ramp rates at high burnup. This study presents the development of a model for UO2-SiC composite light water reactor fuel with enhanced thermal and mechanical properties to mitigate some of the material performance limitations. The model is constructed using self-defined physics modules which are fully coupled and solved using the COMSOL Multiphysics platform. Key fuel performance phenomena being considered include heat generation and conduction, species diffusion, thermomechanics (thermal expansion, elastic strain, densification, and fission product swelling strain), grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, cladding thermal and irradiation creep and oxidation. All the equations are solved via finite-element method on a 2D axisymmetric geometry of a fuel pellet and cladding. The use of UO2-SiC will enhance thermal conductivity, which in turn will decrease nuclear fuel temperatures. This subsequently will improve fission gas retention and reduce pellet cladding interaction resulting from fuel cracking, relocation, and swelling.

Fully coupled multiphysics modeling of enhanced thermal conductivity UO2–BeO fuel performance in a light water reactor

Commercial light water reactor fuel UO2 has a low thermal conductivity that leads to the development of a large temperature gradient across the fuel pellet, limiting the reactor operational performance due to the effects that include thermal stresses causing pellet cladding interaction and the release of fission product gases. This study presents the development of a modeling and simulation for enhanced thermal conductivity UO2–BeO fuel behavior in a light water reactor, using self-defined multiple physics models fully coupled based on the framework of COMSOL Multiphysics. Almost all the related physical models are considered, including heat generation and conduction, species diffusion, thermomechanics (thermal expansion, elastic strain, densification, and fission product swelling strain), grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, cladding thermal and irradiation creep and oxidation. All the phenomenal models and materials properties are implemented into COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet and cladding. UO2–BeO enhanced thermal conductivity nuclear fuel would decrease fuel temperatures and facilitate a reduction in pellet cladding interaction from our simulation results through lessening thermal stresses that result in fuel cracking, relocation, and swelling, so that the safety of the reactor would be improved.

Kinetics of fission product gas release during grain growth

A theoretical framework for the calculation of fission product gas release by diffusion, irradiation-induced re-solution, grain boundary saturation and grain growth in ceramic nuclear fuel (UO2) under time-varying loads is presented. The diffusion equation is subjected to boundary conditions, which account for the aforementioned effects. Analytical solutions to the equation are obtained. The theory is applied to calculate fission gas release and grain growth as functions of temperature and time. Correlation between grain growth and gas release is calculated and the results are discussed in the light of experimental data.

Development of a model to predict fission product behaviour in spherical fuel elements during water ingress events

At PBMR gaseous fission product releases from spherical fuel elements under normal conditions are calculated by the code NOBLEG. The ability of NOBLEG to calculate noble gas and halogen release under oxidizing conditions during water ingress was developed. Observations made during the water vapour injection tests performed during the irradiation experiment HFR-K6, were used to determine simple relations that can be used to predict gaseous fission product release from spherical fuel elements under oxidizing conditions caused by small water ingress events, for PBMR operational temperatures. A new model was proposed to explain peculiarities observed during the water injection tests.

Simulation of “radiation sealing” of an intergrain boundary and the influence of this effect on the release of gaseous fission products from UO2

A new model is developed for radiation sealing of intergrain boundaries under low-temperature irradiation conditions. The model agrees with the physical concepts concerning the mechanisms of athermal diffusion and radiation re-solution of gas bubbles. The model makes it possible to explain the experimentally observed sharp irradiation temperature dependence of gas release at low temperatures, even though the athermal radiation-stimulated transfer mechanisms predominate. The model was integrated into a mechanistic RTOP code and verified on experimental data, including Halden project data. 2 figures, 25 references.

Irradiation behavior of microspheres of U-Zr alloys

In order to understand the fundamental behavior of metallic fuels, microspheres of U-Zr alloys were prepared and irradiated. The microspheres of about 0.8 mmφ were formed by solidifying fused droplets of U-Zr alloys. Irradiation temperatures were controlled by external electric heating. Burnup reached 1.5 at%. After the irradiation, fractional release of fission-product gases, dimensional change and microstructure were examined. The results were compared with the analysis by a metal-fuel performance code ALFUS.

Transport of volatile fission products in the fuel-to-sheath gap of defective fuel elements during normal and reactor accident conditions

An analytical treatment has been used to model the vapour transport of radioactive fission products released into the fuel-to-sheath gap of defective nuclear fuel elements. The model accounts for both diffusive and bulk-convective transport. Convective transport becomes important as a result of a significant release of gaseous fission products into the gap during a high-temperature reactor accident. However, during normal reactor operation, diffusion is shown to be the dominant process of transport. The model is based on an analysis of several in-reactor tests with operating defective fuel elements, and high-temperature annealing experiments with irradiated fuel specimens.

Modelling of (U, Gd)O 2 fuel behaviour in boiling water reactors

Behaviour of (U, Gd)O 2 fuel during irradiation has been modelled. The model accounts both the thermal-mechanical and the nuclear properties of the Gd-bearing nuclear fuel. The nuclear part of the model has been used to evaluate the radial distributions of Gd isotopes at some specific fuel burnups. The evaluation includes burnable absorber rods with gadolinia contents of 6 and 9 wt%. Gadolinium also affects the thermal properties of the fuel. A thermal conductivity expression for (U, Gd)O 2 has been developed and validated with thermal conductivity measurements, as a function of temperature and Gd 2O 3 concentration. In addition, the effect of Gd atoms on crystal lattice structure and fuel stoichiometry which influence the diffusion of fission product gases generated during irradiation in the fuel has been discussed. An empirical correlation considering this impact has been implemented in the STAV-6 code. The STAV-6 code computations for fission product gas release have been compared with data obtained by post-irradiation examination of a number of Gd-bearing fuel rods irradiated in ABB Atom built boiling water reactors up to a burnup of about 42 MWd/kg U. The results of this comparison have been very satisfactory, thus confirming the validity of the models used to predict (U, Gd)O 2 fuel performance.

Experimental study on the effect of fission product gas release into blockage wake region using a simulated LMFBR fuel subassembly

The objective of this study is to evaluate temperature rise due to gas release in the wake region of LMFBR fuel subassemblies. The experiments were conducted in two sets of grid-spacer-type 37-pin bundles simulating LMFBR fuel subassemblies. In Test section 37GC, the central 24 subchannels were blocked by a stainless steel plate and in Test section 37GE one-half edge part (39 subchannels) of the total flow area was blocked by the same material. The experimental results were compared with data obtained in similar tests using a spacer wire-type pin bundle, designated 37WC. The temperature rises in 37GE and 37WC were nearly identical in value and effect of gas release rate. The marked agreement seems to imply that there is a limit in the content of released gas in the wake region. On the other hand, the temperature rise behind the central blockage in the grid-type bundle, where gas might easily flow out to the core flow region, was far less than in the other geometries.

Gaseous Fission Product Release Predictions in TRISO HTGR Fuel Particles

Structural integrity of SiC layers on TRISO‐coated HTGR fuel particles is a critical property necessary to ensure fission product retention. Characterization and detection of microcracks in SiC layers by high‐pressure (69 MPa) mercury intrusion is discussed. Particular emphasis is placed on relating metallic and gaseous fission‐product release to the level of defective SiC layers. Gaseous fission‐product release during irradiation is shown to be more accurately predicted when the fraction of defective SiC layers as determined by mercury intrusion is taken into account.

A microstructure-dependent model for fission product gas release and swelling in UO 2 fuel

A model for the release of fission gas from irradiated UO 2 fuel is presented. It incorporates the relevant physical processes: fission gas diffusion, bubble and grain boundary movement, intergranular bubble formation and interlinkage. In addition, the model allows estimates of the extent of structural change and fuel swelling. In the latter, contributions of thermal expansion, densification, solid fission products, and gas bubbles are considered. When included in the ELESIM fuel performance code, the model yields predictions which are in good agreement with data from UO 2 fuel elements irradiated over a range of water-cooled reactor conditions: linear power outputs between 40 and 120 kW m −1, burnups between 10 and 300 MW h(kg U) −1, and power histories including constant, high-to-low and low-to-high power periods. The predictions of the model are shown to be most sensitive to fuel power (temperature), the choice of diffusion coefficient for fission gas in UO 2, and burnup. The predictions are less sensitive to variables such as fuel restraint, initial grain size and the rate of grain growth.

The Accident at Three Mile Island

On March 28, 1979, what has been called the worst accident in the history of commercial nuclear power generation occurred at the Three Mile Island no. 2 reactor unit in Pennsylvania. The sequence of events that occurred during the first 16 h of the incident are described. There was extensive damage to the core of the reactor as well as some release of gaseous fission products from the station. However, the radiation doses received by the general public as a result of exposure to this released radioactivity were so small that there will be no detectable adverse health effects. Nevertheless, the accident at Three Mile Island will have a profound effect on commercial nuclear power in the United States and throughout the world.

Neutron damage and fission product gas release in the initial burst region of uranium monocarbides

Thermal neutron damage and fission product gas ( 133 Xe) release in a burst region of uranium monocarbides were studied. After neutron irradiation, the electrical resistivity was measured from room temperature to 800° C. Three recovery stages were revealed in the resistivity of UC irradiated to 4.0 × 10 16 nvt. The lattice parameter of UC with the same irradiation also showed three stages of recovery up to 1050°C. The initial burst of Xe from UC was studied in a dose range between 1.6 × 10 15 and 2.9 × 10 18 nvt. The burst occurred in three steps for lightly irradiated specimens, while there were two steps of the burst in heavily irradiated specimens. The activation energies for each burst step were calculated. From the results obtained here, we concluded that the burst was correlated with the recovery of damage in the neutron-irradiated UC. Les dommages dus aux neutrons thermiques et le dégagement de 133 Xe, produit de fission dans une région de dégagement brutal de monocarbures d'U ont été étudiés. Après irradiation neutronique, la résistivité électrique est mesurée depuis la température ambiante jusqu'à 800°C. Trois stades de restauration apparaissent dans la résistivité de UC irradié à 4 × 10 16 nvt. Le paramètre du réseau de UC présentait pour la même irradiation trois stades de restauration jusqu' à 1050°C. Le dégagement brutal initial de Xe à partir de UC a été étudié pour des doses comprises entre 1,6 × 10 15 et 2,9 × 10 18 nvt. Ce dégagement se produisait en trois stades pour les échantillons peu irradiés et en deux stades pour ceux fortement irradiés. Les énergies d'activation pour chaque stade ont été calculées. Des résultats obtenus, on conclut que le dégagement brutal de gaz était en ralation avec la restauration des dommages provoqués dans UC irradié par les neutrons. Die Strahlenschäden durch thermische Neutronen und die Spaltgasfreisetzung (Xe-133) wurden in einem Ausbruchsbe-reich des Uranmonocarbids untersucht. Nach der Neutronenbestrahlung wurde der elektrische Widerstand zwischen Raum-temperatur und 800° C gemessen. Für den Widerstand des mit 4,0 × 10 16 n/cm 2 bestrahlten UC ergeben sich drei Erholungs-stufen. Die Gitterkonstante des UC hat bis 1050°C ebenfalls drei Erholungsstufen bei derselben Bestrahlung. Die anfängliche Xe-Freisetzung aus dem UC wurde bei einer Dosis im Bereich zwischen 1,6 × 10 15 und 2,9 × 10 18 n/cm 2 untersucht. Die Freisetzung tritt bei gering bestrahlten Proben in drei Stufen auf, während die Freisetzung in zwei Stufen bei stark be-strahlten Proben beobachtet wird. Die Aktivierungsenergie für jeden Freisetzungsschritt wird berechnet. Aus den gewonnenen Ergebnissen wird geschlossen, dass die Freisetzung im UC mit der Erholung der Strahlenschäden durch Neutronen im Zusammenhang steht.

A Simulation Experiment of the Results of Fission Product Gas Release due to Fuel Cladding Failure using Water Circulation System

A Simulation Experiment of the Results of Fission Product Gas Release due to Fuel Cladding Failure using Water Circulation System