Chemical forms of uranium evaluated by thermodynamic calculation associated with distribution of core materials in the damaged reactor pressure vessel

ABSTRACT

To suggest efficient process of the fuel debris treatment after the retrieval from the Fukushima Daiichi Nuclear Power Plant (1F), thorough investigation is indispensable on potential source of U in the fuel debris. The present study aims to estimate chemical forms of U in the in-vessel fuel debris, especially in the minor phases such as metallic phases, by performing the thermodynamic calculation under various conditions considering material relocation and changing environment during the accident progression in the 1F Unit 2. Input conditions for the thermodynamic calculation such as composition, temperature, and oxygen amount were set mainly based on the transient change in the core material distribution which had been calculated with severe accident analysis codes such as MAAP and MELCOR. The result showed that chemical form of U varied depending on the local amount of Fe and O. In regions of low steel content, the U-containing metallic phase was dominated by α-(Zr,U)(O), while regions of high steel content were dominated by Fe₂(Zr,U) (Laves phase). A few percent of U transferred to the metallic phases were highly expected under reducing conditions. Therefore, those metallic phases should be one of the potential sources of U.     

Leaching of actinide elements from simulated fuel debris into seawater

For the prediction of the leaching behavior of actinide elements contained in the fuel debris that has arisen from the severe accident in Fukushima Daiichi Nuclear Power Station (NPS), a simulated fuel debris consisting of UO₂–ZrO₂ solid solution doped with ¹³⁷Cs, ²³⁷Np, ²³⁶Pu and ²⁴¹Am tracers was synthesized, and agitated leaching tests were conducted for the simulated fuel debris in seawater. The synthesized simulated fuel debris was immersed and shaken in natural seawater collected at a coast 11 km away from Fukushima Daiichi NPS. The brief leaching test conditions were T = 25 °C and solid–liquid ratio = 4 g/l, and the test duration was up to 31 days. The ratio of tracers leached into seawater from the simulated fuel debris by the agitated leaching test for 4 days was evaluated to be 0.09% for U, 0.01% for Np, 0.01% for Pu, 0.01% for Am and 35.39% for Cs by the α or γ spectrometry of the soluble fraction. The leaching of actinides from the real fuel debris in reactor units 1–3 in Fukushima Daiichi NPS is expected to be suppressed in comparison with that from normal light water reactor spent fuel.     

Calculation of gamma and neutron emission characteristics emitted from fuel debris of Fukushima Daiichi Nuclear Power Station

ABSTRACT

Determination of fuel debris location and distribution inside the primary containment vessel of Fukushima Daiichi Nuclear Power Station is important to decide further decommissioning step and strategy. We calculate neutron and photon spectra including the contribution of secondary particles in the primary containment vessel of Fukushima Daiichi Nuclear Power Station. The Calculated Neutron and photon spectra can be used as the base for determination suitable spectrometer system or detector for searching, localizing and treatment of fuel debris.     

Methodological aspects for finite element modelling of lid systems for Type B(U) transport packages

Abstract

The regulatory compliance of the containment system is of essential importance for the assessment process of Type B(U) transport packages. The requirements of the International Atomic Energy Agency safety standards for transport conditions imply high loading on the containment system. The integrity of the containment system has to be ensured in mechanical and thermal tests. The containment system of German spent nuclear fuel and high level waste transport packages usually includes bolted lids with metal gaskets. The finite element (FE) method is recommended for the analysis of lid systems according to the guideline BAM-GGR 012 for the assessment of bolted lid and trunnion systems. The FE analyses provide more accurate and detailed information about loading and deformation of such kind of structures. The results allow the strength assessment of the lid and bolts as well as the evaluation of relative displacements between the lid and the cask body in the area of the gasket groove. This paper discusses aspects concerning FE simulation of lid systems for type B(U) packages for the transport of spent nuclear fuel and high level waste. The work is based on the experiences of the BAM Federal Institute for Materials Research and Testing as the German competent authority for the mechanical design assessment of such kind of packages. The issues considered include modelling strategies, analysis techniques and interpretation of results. A particular focus of this paper is on the evaluation of the results with regard to FE accuracy, influence of the FE contact formulation and FE modelling techniques to take the metallic gasket into account.     

The influences of Pu and Zr on the melting temperatures of the UO2–PuO2–ZrO2 pseudo-ternary system

Specimens of (U, Pu, Zr)O₂ were prepared as simulated corium debris that were assumed like debris generated in the severe accident of the Fukushima Daiichi Nuclear Power Plant and their melting temperatures were measured by the thermal arrest technique in order to evaluate the influence of plutonium and zirconium content on the melting temperature of the corium debris. From the evaluation, it was found that the influence of zirconium on the melting temperatures of both (U, Pu, Zr)O₂ and (U, Zr)O₂ was similar and that the melting temperature of (U, Pu, Zr)O₂ had a local maximum value in the Pu-content between 0 and 20 mol%. The UO₂–PuO₂–ZrO₂ pseudo-ternary phase diagram at 2900 and 3000 K was evaluated from the present experimental results and previously reported results.     

Description of the German Concept of Final Storage Cask

Abstract

The German storage concept for the direct final storage of spent fuel assemblies from LWR reactors is described. The final storage concept is designed in such a way that it encompasses the whole spectrum of fuel elements to be stored from German reactors, Le. U fuel assemblies and MOX fuel assemblies with a mean bumup of 55 GW.d.t^?1 heavy metal were considered. The further design requirements are defined in such a way that the cask concept satisfies the conditions for type B(U) transport, interim storage and fmal storage. The safe long-term containment of the activity is guaranteed by an inner cask welded leak-tight; the sufficient shielding and the transport packaging are ensured by a shielding cask.     

Effect of quenching on molten core-concrete interaction product

ABSTRACT

Characterization of fuel debris is required to develop fuel debris removal tools for decommissioning Fukushima Daiichi nuclear power plant (1F). Especially, knowledge about the characteristics of molten core-concrete interaction (MCCI) product is needed because of the limited information available at present. Samples from a large-scale MCCI test performed under quenching conditions, VULCANO VW-U1 were analyzed to evaluate the characteristics of the surface of MCCI product. Four samples were selected from test sections at different locations. As a result, the characteristics of the samples were found to be similar. Several corium phases, such as cubic-(U,Zr)O₂ and tetragonal ZrO₂, were detected by X-ray diffraction (XRD), but concrete-based phases, such as the crystalline SiO₂ phase, were not detected by XRD because the quantity of the SiO₂ phase was too small to be measured. The Vickers hardness of each phase in these samples was higher than that of previously analyzed samples in another VULCANO test campaign, VBS-U4. Based on a comparison between MCCI product generated under quenching condition, such as VW-U1, and gently cooled MCCI product, such as VBS-U4, the MCCI product generated under quenching condition is more homogeneous, and its hardness is higher than that of the gently cooled MCCI product.     

A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

An engineering code to predict the irradiation behavior of U–Zr and U–Pu–Zr metallic alloy fuel pins and UO₂–PuO₂ mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel–clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios.FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal–fuel version is called FEAST-METAL, and is described in this paper. The oxide–fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel–clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors.FEAST-METAL was benchmarked against the open-literature EBR-II database for steady state and furnace tests (transients). The results show that the code is able to predict important phenomena such as clad strain, fission gas release, clad wastage, clad failure time, axial fuel slug deformation and fuel constituent redistribution, satisfactorily.     

Corrosion of high burn-up structured UO2 fuel in presence of dissolved H2

The influence of high burn-up structured material on UO₂ corrosion has been studied in an autoclave experiment. The experiment was conducted on spent fuel fragments with an average burn-up of 67 GWd/tHM. They were corroded in a simplified groundwater containing 33 mM dissolved H₂ for 502 days. All redox sensitive elements were reduced. The reduction continued until a steady-state concentration was reached in the leachate for U at 1.5 × 10⁻¹⁰ M and for Pu at 7 × 10⁻¹¹ M. The instant release of Cs during the first 7 days was determined to 3.4% of the total inventory. However, the Cs release stopped after release of 3.5%. It was shown that the high burn-up structure did not enhance fuel corrosion.     

Removal of zirconium from spent fuel solution by alginate gel polymer

The uptake behaviour of zirconium (Zr) in alginate gel polymer and removal of Zr from spent fuel solution have been studied by the batch and column methods. As a first step, alginate gel polymer was synthesized and conditioned. The uptakes of Zr were examined in several concentrations of HNO₃ solution (from 0.01 M to 9 M HNO₃) by a batch method. Stronger affinity of Zr was shown than strontium (Sr), cobalt (Co), uranium (U) and iron (Fe) in 1 M HNO₃. It has been reported that cation binding was stronger with ions of higher charge in the alginate gel polymer. In contrast, the free aqueous ion, Zr⁴⁺, is the dominant form of the Zr species in very acidic solution. Therefore, the strong affinity of Zr is explained. The uptake rate of Zr was also evaluated in 2.6 M HNO₃ solution, which was close to the HNO₃ concentration in actual HLLW from fuel reprocessing. The uptake of Zr reached equilibrium within 2 h. For the column experiment, fission products (FPs) solution containing rare earth elements (REEs), platinum group metals, alkaline metals, alkaline earth metals and the other elements was prepared from actual spent fuel and the concentration of HNO₃ was adjusted to 2.6 M. In the column experiment, the alginate gel polymer was packed into a column, and then a chromatographic experiment was performed using the FPs solution prepared from actual spent fuel. As a result, over 95% of the Zr was removed from the FPs solution. Molybdenum (Mo), technetium (Tc), yttrium (Y), palladium (Pd), tellurium (Te), cesium (Cs) and REEs were eluted by the successive use of H₂O, and 1 M and 3 M HNO₃.     

Host phases for actinides in simulated metallic waste forms

Argonne National Laboratory has developed an electrometallurgical process for conditioning spent sodium-bonded metallic reactor fuel prior to disposal. A waste stream from this process consists primarily of stainless steel cladding hulls containing undissolved metal fission products and a low concentration of actinide elements. This waste will be immobilized in a metallic waste form whose baseline composition is stainless steel alloyed with 15 wt% Zr (SS-15Zr). This paper presents transmission electron microscope, energy-dispersive X-ray spectroscopy, and electron diffraction observations of SS-15Zr alloys containing 2-11 wt% U, Np, or Pu. The major U- and Pu-bearing materials are Cr-Fe-Ni-Zr intermetallics with structures similar to that of the C15 polymorph of Fe₂Zr, significant variability in chemical compositions, and 0-20 at.% actinides. A U-bearing material similar to the C36 polymorph of Fe₂Zr had more restricted chemical variability and 0-5 at.% U. Uranium concentrations between 0 and 5 at.% were observed in materials with the Fe₂₃Zr₆ structure.     

Source term evaluation for the upgraded LEU Pakistan Research Reactor-1 under severe accidents

Evaluation of source term has been carried out for the upgraded LEU PARR-I system taken as a typical material test reactor (MTR). The modeling and simulation of release of radioactivity has been carried out by developing a Matlab based computer program which uses the ORIGEN2 code for core inventory calculations. For post 180 full-power days continuous operation, various accident scenarios, with instantaneous release of radioactivity to containment, have been considered including the startup, fuel loading, and loss-of-coolant accidents. For noble gases, iodine and for aerosols, the release rate studies have been carried out for the normal, emergency and for the isolation states of containment. The values of source term as well as that of containment retention factor show rapid increase followed by an approach towards saturation values as the exhaust rate values are increased. The isotope-dependency of the containment retention factor has been studied and the results indicate strong sensitivity for ⁸⁵Kr, ¹³⁷Xe, ¹³⁸Xe and ¹³⁸Cs towards exhaust rate values.     

Burnup studies of spent fuels of varying types and enrichment

This paper describes the results of fuel burnup measurements, made over a period of several years on discharged fuel from nuclear power plant and research reactor. The measured and calculated burnup of different spent fuel types, viz.: Natural uranium CANDU fuel bundles; 10.5% enriched booster rods; 20% enriched MTR fuel elements have been presented. High-resolution gamma spectrometry, using ¹³⁷Cs and ¹³⁴Cs burnup monitors was employed in different reactors to estimate the amount of ²³⁵U depletion in the respective fuel. The experimental data was compared with those of calculations to optimize fuel-scheduling programme. The burnup data is useful for assessment of fuel performance in the core and resolving design issues related to long-term storage facilities. It has been observed that the gamma spectrometry is very effective in identifying exact position of individual booster bundles inside the discharged booster assemblies, which is useful in safeguard applications. It is concluded that the distribution of measured isotopic activity ratios of ¹³⁴Cs/¹³⁷Cs along the height of the spent fuel gives accurate estimate of the axial neutron flux profiles in the core. The activity ratio technique therefore provides a useful method to determine flux peaking factors at the particular locations of the fuel assemblies in the reactor.     

Burnup Determination of High Burnup and Dry Processed Fuels Based on Isotope Dilution Mass Spectrometric Measurements

Destructive methods were used for the burnup determination of a PWR nuclear fuel irradiated to a high burnup in power reactors, and of a dry processed fuel fabricated from a spent PWR fuel and irradiated in the Hanaro research reactor. The total burnup was determined from a measurement of the Nd and Cs isotope burnup monitors. The methods included U, Pu, ¹⁴⁸Nd, ¹⁴⁵Nd+¹⁴⁶Nd, total of the Nd isotopes, ¹³³Cs and ¹³⁷Cs determinations by the isotope dilution mass spectrometric method (IDMS) by using quadrupole spikes (²³³U, ²⁴²Pu, ¹⁵⁰Nd, and ¹³³Cs). The methods involved two sequential anion exchange resin (AG 1X8 and 1X4) separation procedures and a Cs purification with a cation exchange resin (AG 50WX4) separation procedure. The results obtained by the Nd and Cs isotopes from the mass spectrometric measurement were compared with those by the ORIGEN code.     

Zr-rich layers electrodeposited onto stainless steel cladding during the electrorefining of EBR-II fuel

Argonne National Laboratory is developing an electrometallurgical treatment for spent nuclear fuels. The initial demonstration of this process is being conducted on U–Zr alloy fuel elements irradiated in the Experimental Breeder Reactor II (EBR-II). We report the first metallographic characterization of cladding hull remains for the electrometallurgical treatment of spent metallic fuel. During the electrorefining process, Zr-rich layers, with some U, deposit on all exposed surfaces of irradiated cladding segments (hulls) that originally contained the fuel alloy that was being treated. In some cases, not only was residual Zr (and U) found inside the cladding hulls, but a Zr-rind was also observed near the interior cladding hull surface. The Zr-rind was originally formed during the fuel casting process on the fuel slug. The observation of Zr deposits on all exposed cladding surfaces is explained with thermodynamic principles, when two conditions are met. These conditions are partial oxidation of Zr and the presence of residual uranium in the hulls when the electrorefining experiment is terminated. Comparisons are made between the structure of the initial irradiated fuel before electrorefining and the morphology of the material remaining in the cladding hulls after electrorefining.     

First-principles calculation of mechanical properties of simulated debris ZrxU1−xO2

ABSTRACT

To elucidate the mechanical properties of fuel debris inside the Fukushima Daiichi Nuclear Power Plant, we use first-principles calculations to evaluate mechanical properties of cubic Zr_xU_1?xO₂, which is a main component of the fuel debris. We focus on the dependence of mechanical properties on the fraction x of zirconium and compare our results with recent experiment of simulated debris, in which dependences of elastic moduli and fracture toughness on the ZrO₂ content showed deviation from a simple linear relation. We show that elastic moduli drop at around x = 0.25 and increase again for larger values of x, as has been observed in experiments. The reason of the drop is a softening owing to disordered atomistic structures induced by the solute zirconium atoms. We also find that stress–strain curves for the x = 0.125 case show marked hysteresis owing to the existence of many meta-stable states. We show that this hysteresis leads to slightly increased fracture toughness, but it is not enough to account for the significant increase of fracture toughness observed in experiments.     

Leaching behavior of gamma-emitting fission products and Np from neutron-irradiated UO2–ZrO2 solid solutions in non-filtered surface seawater

The gamma ray radionuclides Cs-137, Ba-140, I-131, Ce-141, Ru-103, Zr-95, and Np-239 were produced by neutron irradiation of UO₂–ZrO₂ solid solutions that were synthesized as simulated fuel debris under reducing and oxidizing conditions. The leaching ratio of radionuclides was investigated under atmospheric conditions at 25 °C for non-filtered natural surface seawater, as well as deionized water after filtration with a membrane of 0.45-µm pore size or that of nominal molecular weight limit of 3 kDa. The uranium molar concentration was affected by the oxidation state in the solid solution samples. The congruent dissolution of Cs, I, and Ba with the hexavalent uranium of U₃O₈ was facilitated in the seawater samples, whereas a lower leaching ratio of nuclides was observed in the deionized water samples. Neptunium-239, originally produced from uranium-238 in U₃O₈, showed behavior that was similar to that of Cs, I, and Ba. However, the dissolution of Np (as a parent nuclide of Pu-239) in the debris of UO₂ and UO₂–ZrO₂ was suppressed in the same manner as Zr(IV) and Ce(IV). The concentration exhibited no filtration dependence after 15 d, which shows that most of the leached nuclides can exist in their ionic form in seawater.     

Leaching behavior of gamma-emitting fission products,calcium, and uranium from simulated MCCI debris in water

ABSTRACT

Gamma-ray radionuclides (Cs-137, Ba-140, I-131, Ru-103, and Zr-95) were produced by neutron irradiation of simulated molten core–concrete interaction (MCCI) debris, which was synthesized by the heat treatment of a mixture of UO₂ with concrete components at a relatively low temperature of 1473 K under reducing and oxidizing conditions. The major uranium solid phases were unreacted UO₂ and calcium uranate. The leaching ratio of the radionuclides in the powdered sample of the simulated MCCI debris was investigated under atmospheric conditions at 298 K in 0.1 mol/dm³ NaClO₄ after filtration of the leachate through a 0.45-µm pore membrane. The uranium molar concentration in the filtrate was affected by the oxidation state in the solids. In the present study, however, the effect of various solid phase conditions on the leaching ratio normalized to that of uranium matrix could not be clarified. It was found that the leaching ratio of various fission products (R_M) was proportional to that of Cs (R_Cs), and this trend did not depend on the oxidation state of uranium, the type of uranium complex (including a colloidal species), or the presence of Ca, Si, cement, or Zr.     

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.     

Development of thermodynamic databases in the system U–Zr–Ce–Cs–Fe–B–C–I–O–H for application to simulating phase equilibria in severe nuclear accidents

In case of severe nuclear accidents involving melt down of nuclear fuels at high temperatures, it is of considerable importance to accurately evaluate the highly-volatizing behavior of fission products (FPs) over multicomponent debris. Particularly, cesium (Cs)- and iodine (I)- bearing chemical species are regarded as notable FPs. In the present work, the authors have generated original thermodynamic databases for the system U–Zr–Ce–Cs–Fe–B–C–I–O–H featuring Cs- as well as I-bearing subsystems, which are contained in oxide, iodide, and metal (including borides and carbides) sub-databases. It has been confirmed that the phase diagrams calculated by the present set of the databases reproduce the corresponding literature data well in various kinds of subsystems of the above multicomponent system. The present set of databases has subsequently been applied to simulate phase equilibria and volatizing behavior of Cs- and I-including species, respectively, in multicomponent debris under specific temperature and atmospheric conditions corresponding to severe nuclear accidents.