Hydraulic simulation of local blockage in a LMFBR fuel subassembly

Experimental studies of local flow blockage in a LMFBR fuel subassembly have been carried out using a simulating model in a water test loop. The studies verified the numerical results of an analytical code. The experiments were conducted in a 61-pin bundle containing a planar blockage without leakage flow. Central and edge blockage were used. The wake flow behind the blockage was visualized with dye or air bubble injection to grasp the flow characteristics. The velocity, static pressure and residence time were measured in the wake flow region.     

Pressure pulse on a subassembly wall due to gas release from a failed fuel pin

This paper presents a mathematical model to predict the pressure pulse on a subassembly of fuel pins due to rapid release of gas from a failed pin into liquid coolant between the pins. The subassembly is simulated by a rigid circular tube, and liquid flow inside the tube is assumed incompressible, inviscid, and irrotational. A gas bubble along the centerline of the subassembly is considered to be formed as a result of the gas release from the plenum, and a pressure pulse on the subassembly wall is a consequence of the liquid being accelerated by the gas bubble. It is assumed that the gas bubble grows spherically until it touches the subassembly wall, and then expands as a cylinder with hemispherical ends. This analysis is particularly applicable to the EBR-II reactor.     

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 $\text{4.0 × 10}{}^{16}$ 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 $\text{1.6 × 10}{}^{15}$ and $\text{2.9 × 10}{}^{18}$ 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.     

Rapid aqueous release of fission products from high burn-up LWR fuel: Experimental results and correlations with fission gas release

Studies of the rapid aqueous release of fission products from UO₂ and MOX fuel are of interest for the assessment of the safety of geological disposal of spent fuel, because of the associated potential contribution to dose in radiological safety assessment. Studies have shown that correlations between fission gas release (FGR) and the fraction rapidly leached of various long-lived fission products can provide a useful method to obtain some of this information. Previously, these studies have been limited largely to fuel with burn-up values below 50 MWd/kg U. Collaborative studies involving SKB, Studsvik, Nagra and PSI have provided new data on short-term release of ¹³⁷Cs and ¹²⁹I for a number of fuels irradiated to burn-ups of 50–75 MWd/kgU. In addition a method for analysis of leaching solutions for ⁷⁹Se was developed. The results of the studies show that the fractional release of ¹³⁷Cs is usually much lower than the FGR covering the entire range of burn-ups studied. Fractional ¹²⁹I releases are somewhat larger, but only in cases in which the fuel was forcibly extracted from the cladding. Despite the expected high degree of segregation of fission gas (and by association ¹³⁷Cs and ¹²⁹I) in the high burn-up rim, no evidence was found for a significant contribution to release from the rim region. The method for ⁷⁹Se analysis developed did not permit its detection. Nonetheless, based on the detection limit, the results suggest that ⁷⁹Se is not preferentially leached from spent fuel.     

Modeling and parametric studies of the effect of inhomogeneity on fission gas release in LWR MOX fuel

To analyze the effect of an inhomogeneous mixture of an PuO₂ powder on fission gas release in MOX fuel, a model has been developed using the assumption that gas release mechanism in Pu-rich particles is identical with that in UO₂ fuel. A parametric study was performed to see the respective effect of the number density, size and fraction of Pu retained in the Pu-rich particles on gas release in MOX fuel. The model shows that, for the condition of all the other remaining parameters being fixed, more gas is released in a MOX fuel for lower number density of, smaller size of, and larger fraction of Pu retained in, the Pu-rich particles. However, there exists some condition or combination of parameters for which the effect of inhomogeneity on gas release is negligible depending on the characteristics of MOX fuel. Comparison with measured data for OCOM MOX fuel shows that the present model can predict the level of gas release in MOX fuel once the release mechanism in the Pu-rich particles is known.     

The effect of irradiation-induced re-solution on fission gas release

A finite-difference technique is used to compute exact solutions to the diffusion equation describing fission gas release from UO₂ nuclear fuel during steady reactor operation. The resolution of gas atoms from grain-boundary bubbles is treated in two alternative ways, and the results of the parallel calculations compared. Predictions of gas release using simple analytical models are compared with the numerical results and are found in general to describe the process very accurately.     

Effects of interaction between molten zircaloy and irradiated MOX fuel on the fission product release behavior

As a first step for obtaining experimental data on the effects of high-temperature chemical interaction on fission product release behavior, we focused on the dissolution of irradiated uranium plutonium mixed oxide (MOX) fuel by molten zircaloy (Zry) and carried out a heating test under the reducing atmosphere. Pieces of an irradiated MOX fuel pellet and cladding were subjected to the heating test at 2373 K for five minutes. The fractional release rate of cesium (specifically ¹³⁷Cs) was monitored during the test and its release behavior was evaluated. The observation of microstructures and measurements of elemental distribution in the heated specimen were also performed. We demonstrated experimentally that the fuel dissolution by molten Zry accelerated the release of Cs from the fuel pellets.     

A methodology to predict a fission gas release ratio of MOX fuel with heterogeneous microstructure

The objective of this study is to formulate a methodology to predict a fission gas release ratio of MIMAS MOX. An irradiated MIMAS MOX fuel with plutonium rich agglomerates was subjected to elemental analyses by electron probe micro analysis and secondary ion mass spectrometry in order to investigate xenon distribution. The results of the elemental analyses showed that the plutonium rich agglomerates at the periphery of the fuel pellet sample retained a high concentration of xenon as gas bubbles. Then, the results were used as reference data for modification of models in a fuel rod analysis code, FEMAXI-7. Using the modified FEMAXI-7, we applied an approach to prediction of fission gas release ratio of MOX fuel with plutonium rich agglomerates. In the approach, two separated analyses using FEMAXI-7 were performed for the plutonium rich agglomerates and the matrix. Fission gas release ratios obtained from the two analyses were processed through weighted-average with burnup ratios of the plutonium rich agglomerates and the matrix. Finally, the fission gas release ratios were compared with results of rod puncture tests. As a result of the comparison, it was confirmed that the proposed approach could well predict fission gas release ratio of MOX fuel with plutonium rich agglomerates.     

Effect of non-equilibrium fission gas and fuel creep on swelling and release in irradiated carbide fuels

The code UCSWELL was developed to simulate fission gas behavior in carbide fuels. In the present work, one of the limiting assumptions in UCSWELL - that matrix gas bubbles are in equilibrium with gas atom concentration - is removed and non-equilibrium matrix fission gas bubbles are allowed, but with relaxation to equilibrium by means of vacancy diffusion and thermal and radiation-induced creep of the fuel. For a given grain size, the difference in swelling between equilibrium and non-equilibrium with relaxation bubble fission gas treatment increases with decreasing irradiation temperature. At a given temperature, the non-equilibrium effect is more pronounced for larger grain fuel. This is to be expected because the creep rate (and hence the rate at which bubbles grow to an equilibrium size) decreases as temperature decreases and/or as grain size increases. At temperatures, where the creep rate is grain size insensitive, grain size remains important to the equilibrium process in so far as the grain boundary is a source of vacancies to the non-equilibrium bubbles. While the difference in these quantities is at the most on the order of 20% for the steady operating conditions considered, it is anticipated that the non-equilibrium effects become more pronounced during reactor overpower and undercooling transients.     

Radiation damage and simulated fission product effects on the properties of inert matrix materials

Candidate inert matrix materials for actinide transmutation (MgAl₂O₄, CeO₂) or immobilization (ZrSiO₄) containing ²⁴¹Am were characterized. The currently most considered material, ZrO₂, was produced, with La₂O₃ as stand-in for Am, and with and without simulated fission products to investigate burnup effects. The oxygen potential was measured using an EMF cell. The accumulation of radiation damage due to Am decay was investigated by periodically measuring lattice parameters and hardness. Sequential leaching tests in deionized water, aimed at correlating the leaching behaviour of Am and of the matrix with radiation damage, showed significant release of Am and of some matrix components.     

A model for fission gas release and gaseous swelling of the uranium dioxide fuel coupled with the FALCON code

A comprehensive model GRSW-A was developed to analyse the processes of fission gas release, gaseous swelling and microstructural evolutions in the uranium dioxide fuel during base irradiation and under transient conditions. The GRSW-A analysis incorporates a number of models published in open literature, as well as some original models that were already published by the authors elsewhere. Consequently, only the most prominent aspects of GRSW-A and its coupling with the FALCON fuel behaviour analysis and licensing code are described in this paper. The analysis of fuel behaviour in the REGATE experiment is presented, which includes the base irradiation of the fuel segment in a PWR to a burn-up of about 50 MWd/kgU, which was followed by a power ramp in the SILOE research reactor. Besides, the generalized data on fission gas release (FGR) in PWR fuel during the base irradiation up to a burn-up of about 70 MWd/kgU is interpreted using coupled FALCON and GRSW-A. Moreover, a mechanistic interpretation of the published data for pellet swelling during the base irradiation up to a burn-up of 100 MWd/kgU is put forward. In all the cases, the coupled FALCON/GRSW-A analysis has shown the improved prediction capability compared to the original FALCON MOD01, which is achieved due to the account for the mutual effect of thermal and, in particular, high-burn-up-assisted mechanisms of fission gas release and swelling under steady-state and transient conditions.     

On the proliferation issues of a fusion fission fuel factory using a molten salt

The fusion fission fuel factory (FFFF) is a hybrid fusion fission reactor using a neutron source, which is in this case taken similar to the source of the Power Plant Conceptual Study - Water Cooled Lithium Lead (PPCS-A) design, for fissile material production instead of tritium self-sufficiency. As breeding blanket the first wall of the ITER design is attached to a molten salt zone, in which ThF₄ and UF₄ solute salts are transported by a LiF-BeF₂ solvent salt. For this blanket design, the fissile material is assessed in quantity and quality for both the Th-U and the U-Pu fuel cycle.The transport of the initial D-T fusion neutrons and the reaction rates in this breeding blanket are simulated with the Monte Carlo code MCNP4c2. The isotopic evolution of the actinides is calculated with the burn-up code ORIGEN-S.For the Th-U cycle the bred material output remains below 10 g/h with a ²³²U impurity level of 30 ppm, while for the U-Pu cycle supergrade material is produced at a rate up to 100 g/h.     

The development of a mechanistic code on fission product behavior in the polycrystalline UO2 fuel

The current status of a mechanistic code (RTOP) on fission product behavior in the polycrystalline UO₂ fuel is described. Outline of the code and implemented physical models is presented. The general approach to the code validation is discussed. It is exemplified by the results of validation of the models of oxidation and grain growth. The different models of intragranular and intergranular gas bubbles behavior have been tested and the sensitivity of the code in the framework of these models has been analyzed. An analysis of available models of the resolution of grain face bubbles is also presented.     

Experimental study on void fraction in a simulated BWR fuel assembly (evaluation of cross-sectional averaged void fraction)

Void fraction measurement of a vertical (4 x 4) rod bundle has been conducted in a steam-water two phase flow, using an advanced X-ray CT scanner. A large amount of rod bundle data was obtained. It was found from the results that the cross-sectional averaged void fraction data for a rod bundle can be correlated by the Drift-Flux model and that the Zuber-Findlay correlation underestimates the data in a void fraction area of 80% or more. This is because the data range over which this correlation was developed, does not cover this experimental range. Therefore, a modified correlation was developed based on the authors' data.     

A probabilistic study of the effect of retardation factor uncertainty using a compartment model for radionuclide release into the biosphere

The effect of retardation factor uncertainty with reference to the results of a compartment model, representing radionuclide release into the biosphere from a disposal facility, is studied through the application of polynomial chaos theory. We review the derivation of the working equations and apply the polynomial chaos expansion to these equations. The uncertainty in the retardation factor typically covers several orders of magnitude and stems from uncertainties in the distribution coefficient, these large uncertainties make any quantitative analysis of radionuclide release highly problematic. In this paper, we assume that the uncertainty in the retardation factor is fully described by a log-uniform distribution. We compare results using polynomial chaos against a semi-analytical solution for a short decay chain and present numerical results from polynomial chaos applied to a longer generic decay chain.     

Model development for the estimation of fission product release under normal and accident conditions in a HTGR

This article has attempted to estimate the radioactivity release from fuel materials during normal and transient conditions by coupling the TRISO fracture and the fission product (FP) diffusion. Two calculation models, named TRISO Fracture Analyzer (TRIFA) and DIFfusion Analyzer (DIFA), are developed. TRIFA is initially used to calculate the fraction of fractured fuel particles, thus determining the amount of fission gas release. The obtained particle fracture function is then used as input for the diffusion rate calculation. DIFA simulates with a single spherical fuel element, a pebble, irradiated under normal and accident conditions. It describes the diffusive transport of fission products by numerically solving the diffusion equation. The finite difference method is applied to obtain fission product release rates from a pebble to coolant. The model comparisons show that the new developed models are reliable, fast, and correspond with previous results of other models. As for HTR-10, the coupled models, TRIFA and DIFA, are applied to calculate the level of fission product release after accidents. The following conclusions can be drawn. First, the mitigation should be carried out until the maximum fuel temperature reaches under transient. Second, the mitigation should be intensively considered if the burn-up exceeds 5%FIMA (∼48 GWd/MTU) when transient happens. Additionally, it is found that there is the threshold burn-up where the rapid FP release occurs due to the numerous TRISOs fractured. Further investigations are needed to extend the use of the method developed in this work to the safety assessments for high-temperature gas-cooled reactors (HTGRs). This article will hopefully serve as a platform for designing the advanced TRISO that can minimize the activity release, and providing the rationale of development of the intensive accident mitigation system in future.     

Experimental study on the mechanism of flow blockage formation in fast reactor

Various sources of solid particles might exist in the coolant flow of a liquid metal cooled fast reactor (e.g., through chemical interaction between the coolant and impurities, air, or water, through corrosion of structural materials, or from damaged/molten fuel). Such particles may cause flow blockage accidents in a fuel assembly, resulting in a reduction in coolant flow,which potentially causes a local temperature rise in the fuel cladding, cladding failure, and fuel melt. To understand the bl...     

On the effects of fission product noble metal inclusions on the kinetics of radiation induced dissolution of spent nuclear fuel

Radiation induced oxidative dissolution of UO₂ is a key process for the safety assessment of future geological repositories for spent nuclear fuel. This process is expected to govern the rate of radionuclide release to the biosphere. In this work, we have studied the catalytic effects of fission product noble metal inclusions on the kinetics of radiation induced dissolution of spent nuclear fuel. The experimental studies were performed using UO₂ pellets containing 0%, 0.1%, 1% and 3% Pd as a model for spent nuclear fuel. H₂O₂ was used as a model for radiolytical oxidants (previous studies have shown that H₂O₂ is the most important oxidant in such systems). The pellets were immersed in aqueous solution containing H₂O₂ and and the consumption of H₂O₂ and the dissolution of uranium were analyzed as a function of H₂ pressure (0–40 bar). The noble metal inclusions were found to catalyze oxidation of UO₂ as well as reduction of surface bound oxidized UO₂ by H₂. In both cases the rate of the process increases with increasing Pd content. The reduction process was found to be close to diffusion controlled. This process can fully account for the inhibiting effect of H₂ observed in several studies on spent nuclear fuel dissolution.