Separation of Precipitate-forming Elements during High Level Waste Treatment Using Acidic Zirconium Salt of Dibutyl Phosphoric Acid

Iron extraction and the limits of iron dibutyl phosphate precipitation were investigated for use in TPE/RE recovery and partitioning with the use of the acidic Zr salt of dibutyl phosphoric acid (ZS HDBP, Zr:HDBP = 1:9), dissolved in 30% TBP with Isopar-L. The presence of Mo and the solvent loading with RE affected the Fe extraction in the opposite ways. Slow kinetics and process irreversibility were found for the Fe extraction with the ZS HDBP solution. To increase acceptable Fe concentration in a HLW, reduction of Fe(III) by ascorbic acid (AA) was studied to be carried out continuously in the feed flow, just before entering the head contactor of the partitioning extraction cycle. The Mo extraction is of the same order as that of TPE and RE; so its selective stripping prior to the TPE/RE separation is required. This can be done using the DTPA or H₂O₂ solution in diluted nitric acid. The effectiveness of the process was verified by laboratory scale trials on the centrifugal contactor rig using simulated HLW. The Zr-to-HDBP ratio of 1:6 was found to be useful to decrease the Fe and Mo extractability for their better backwashing with complexants.     

Behavior of molybdenum in pyrochemical reprocessing: A spectroscopic study of the chlorination of molybdenum and its oxides in chloride melts

The high temperature reactions of molybdenum and its oxides with chlorine and hydrogen chloride in molten alkali metal chlorides were investigated between 400 and 700 °C. The melts studied were LiCl-KCl, NaCl-CsCl and NaCl-KCl and the reactions were followed by in situ electronic absorption spectroscopy measurements. In these melts Mo reacts with Cl₂ and initially produces MoCl₆²⁻ and then a mixture of Mo(III) and Mo(V) chlorocomplexes, the final proportion depending on the reaction conditions. The Mo(V) content can be removed as MoCl₅ from the melt under vacuum or be reduced to Mo(III) by Mo metal. The reaction of Mo when HCl gas is bubbled into alkali chloride melts yields only MoCl₆³⁻. MoO₂ reacts in these melts with chlorine to form soluble MoOCl₅²⁻ and volatile MoO₂Cl₂. MoO₃ is soluble in chloride melts and then decomposes into the oxychloride MoO₂Cl₂, which sublimes or can be sparged from the melt, and molybdate. Pyrochemical reprocessing can thus be employed for molybdenum since, after various intermediates, the end-products are chloride melts containing chloro and oxychloro anions of molybdenum plus molybdate, and volatile chlorides and oxychlorides that can be readily separated off. The reactions were fastest in the NaCl-KCl melt. The X-ray diffraction pattern of MoO₂Cl₂ is reported for the first time.     

Corrosion behavior of Mo–Re based alloys in liquid Li

The corrosion behavior was investigated in liquid Li at 1473 K for 1.8 Ms using a capsule test for two Mo–Re based alloys, Mo–15mol%Re–0.1mol%Zr and Mo–15mol%Re–0.1mol%Zr–0.1mol%Ti. They showed mass gains due to the corrosion. The amount of mass gains increased with increasing corrosion time. Two types of corrosion products were formed on the surface of them. One was Re₂Zr and the other was ZrN according to the analyses using X-ray diffraction method and Auger electron spectroscopy. A deposition mechanism of these compounds was proposed on the basis of the experimental results. However, despite the appearance of these compounds, there were not any cracks on the surface even after the corrosion test for 1.8 Ms. From these results, these alloys were found to exhibit superior corrosion resistance against liquid Li.     

Liquid Metal Extraction for Removal of Molybdenum from Molten Glass Containing Simulated Nuclear Waste Elements

Laboratory-scale experiments for removing Mo and MoO3 from molten borosilicate glass were performed using liquid Cu as an extractant. Removal of Mo from the simulated HLW glass containing oxides of Nd, Fe, Zr, Mo, Sn, Ni, Sr, Cd, Ru, and Se was also performed, and the fractions of these elements transferred into Cu were examined. Mixtures of Cu anda ternary SiO₂-B₂O₃-Na₂O glass containing metallic Mo or MoO₃ were heated in an alumina crucible at 1,673K in an Ar environment. The amounts of Mo and MoO₃ added to 10 g of the ternary glass were fixed at 0.1 and 0.15 g, respectively. As for the glass containing metallic Mo, more than 90% of Mo was extracted into liquid Cu. Spherical Cu metal buttons containing Mo formed on the bottom of the crucible when Cu was added at more than 10 times that of Mo on a mass basis. Removal of Mo from the glass containing MoO3 was also achieved by the addition of Si as a reducing agent for the reduction from MoO₃ to Mo. The fraction of Mo extracted into liquid Cu depended on the molar ratio of Si to Cu added to the glass. The fraction increased up to 84% with an increase in the molar ratio of Si/Cu. However, the excess addition of Si may enhance the chemical interaction between the metal phase and the glass phase, and some of the metal phase containing Mo remained in the glass phase without forming a metal button. The optimum molar ratio of Si/Cu that produces the highest removal fraction was found to be approximately 0.5. Almost the same removal fraction of 88% was obtained from the simulated HLW glass under the condition of Si/Cu = 0.5. Nearly 100% of Ru was extracted into Cu with Mo, while Sr, Zr, and Nd were hardly extracted and remained in the glass.     

Thermophysical properties of U2Mo intermetallic

The current paper has investigated specific heat capacity, coefficient of thermal expansion and phase transition temperatures of U₂Mo intermetallic. The transformation of U₂Mo phase to bcc (γ-uranium) phase occurred at 853 K. The coefficient of thermal expansion has been determined in the temperature range 423–873 K to be 14.65 × 10^?6 K^?1. The specific heat data showed a smooth curve up to 773 K and above it a change in trend to follow a λ-shape was observed. The specific heat values of U₂Mo have been found to be lower than that calculated from the additivity rule. This study provides first time data on U₂Mo. The data obtained from this investigation were compared to available literature on U–Mo alloys.     

Impact of Material Configuration Behind First Wall on Helium Formation in In-Vessel Components of Fusion Reactor

The results of neutron transport calculations of the He formation based on the JENDL gas-production cross section file are discussed for some metals and alloys, namely ²⁷A1, Ti, ⁵¹V, Cr, ⁵⁵Mn, Fe, Ni, Zr, Mo, austenitic stainless steel (Ti modified 316 SS: PCA), Ni-base alloy (Inconel 625), ferritic steel (Fe-11Cr-1Mo: HT-9), Ti-base alloy (Ti-6A1-4V) and V-base alloy (V-5Cr-5Ti). Impacts of the two shields having the steel-rich and the H₂O-rich compositions and the two blankets having the Li₂O/Be-base and the liquid Li/Be-base compositions on the He formation rate in the above-mentioned metals and alloys are discussed. The relation between the He formation rate and the fast neutron flux (14.1 MeV>E>0.1 MeV) is investigated. The decrease of He formation at any distance Δ from the first wall more than Δ_as, the distance where the shape of neutron spectrum reaches its asymptotic form, is modelled by the simple formula based on the exponential dependence, as those reported so far for the fast neutron flux and the displacement damage rate.     

Creep-fatigue evaluation of normalized and tempered modified 9Cr---1Mo

Creep-fatigue is a fatal failure mode of the high temperature structural materials of liquid metal fast breeder reactors (LMFBRs). In this report, two important issues are discussed for creep-fatigue evaluation of normalized and tempered modified 9Cr---1Mo (modified 9Cr---1Mo(NT)) steel which is a promising structural material for the steam generator of large-scale LMFBRs in Japan. Several evaluation methods based on the ductility exhaustion concept are discussed for the prediction of tension strain hold creep-fatigue damage of this material. A time-fraction type of linear damage summation concept based on a new ductility exhaustion theory is proposed from the point of view of its appropriate conservatism for time extrapolation and its simplicity.Also, a life reduction mechanism of low cycle fatigue with strain hold at the compression side is discussed, based on the data observed by a scanning type electron microscope. Creep damage or the tension mean stress caused by compression strain hold hardly reduce the low cycle fatigue life of this material. A new concept based on the location of oxidation on the test specimen surface can explain the reduction in low cycle fatigue life of modified 9Cr-1Mo(NT) steel.     

Structural role of molybdenum in nuclear glasses: an EXAFS study

The Mo environment has been investigated in inactive nuclear glasses using extended X-ray absorption spectroscopy (XAS). Mo is present in a tetrahedron coordinated to oxygen in the form of molybdate groups [MoO₄]²⁻ (d(Mo-O)=1.78 Å). This surrounding is not affected by the presence of noble metal phases in the nuclear glass. Relying on the XAS results, on the bond-valence model and on molecular dynamics simulations of a simplified borosilicate model glass, we show that these groups are not directly linked to the borosilicate network but rather located within alkali and alkaline-earth rich domains in the glass. This specific location in the glass network is a way to understand the low solubility of Mo in glasses melted under oxidizing conditions. It also explains the possible phase separation of a yellow phase enriched in alkali molybdates in molten nuclear glasses or the nucleation of calcium molybdates during thermal aging of these glasses. Boron coordination changes in the molten and the glassy states may explain the difference in the composition of the crystalline molybdates, as they exert a direct influence on the activity of alkalis in borosilicate glasses and melts.     

Influence of chemical speciation in reactor cooling system on pH of suppression pool during BWR severe accident

The influences of chemical speciation for Cs–I–Te–Mo–Sn–B–C–O–H system, simulating a state in the reactor cooling system (RCS) of BWR, on pH of the suppression chamber (S/C) water pool were analytically investigated with PHREEQC code. Major conditions were chosen on the basis of the outputs from a BWR severe accident analysis by THALES2 code and chemical thermodynamic analysis with VICTORIA2.0 code. The chemical thermodynamic analysis showed that the chemical speciation of important volatile FPs, Cs and I was strongly influenced by Mo and B₄C control material. As a consequence, pH of the S/C water pool was predicted to range from approximately 6 to 10, depending on the fraction of volatile FPs transported from the RCS to the S/C water pool and the H₂/H₂O ratio associated with the oxygen potential. It was implied that the formation of volatile I species such as I₂ in the S/C water pool was larger by three orders at the lowest pH than that at the highest pH.     

Composition and structure of fission product precipitates in irradiated oxide fuels: Correlation with phase studies in the Mo-Ru-Rh-Pd and BaO-UO2-ZrO2-MoO2 Systems

Composition and crystal structure of fission product precipitates in irradiated oxide fuels were studied by X-ray microanalysis and X-ray diffraction using instruments shielded for α-contamination and β-γ-radiation. Pin cross-sections, fuel micro-samples from 300 μm hollow drillings and residues from the dissolution of irradiated material in HNO₃ were investigated. The metallic phases found are hcp ?-Ru(Mo,Tc,Rh,Pd) solid solutions with broad variations in concentration of the components, bcc β-Mo(Tc,Ru) and fcc α-Pd(Ru,Rh). The dominating ceramic precipitate is composed of ($\text{Ba}{}_{\mathrm{1-x-y}}\text{Sr}{}_{\mathrm{x}}\text{Cs}{}_{\mathrm{y}}$)(U,Pu,RE,Zr,Mo)O₃ which crystallizes in the cubic perovskite type. The Mo fraction of these phases is related to the local oxygen potential of the fuels. The in-pile observed results agree well with phase studies in the quaternary Mo-Ru-Rh-Pd system where complete solid solubility exists between hcp Ru and the hexagonally stabilized Mo-Rh and Mo-Pd phases. Agreement is further attained with phase studies in the pseudoquaternary BaO-UO₂-ZrO₂-MoO₂ system which is characterized by a cubic perovskite phase Ba(U,Zr,Mo)O₃.     

Study on mo steel for high temperature gas-cooled reactor material characterization of forged low Si mo steel

The Japan Atomic Energy Research Institute (JAERI) has carried out a series of research and development work related to the high temperature gas-cooled reactor (HTGR) and, accordingly the high temperature engineering test reactor (HTTR) will be constructed in the near future. As the reactor pressure vessel (RPV) material, Mo steel will be used. Material characterization tests have been carried out to evaluate the applicability of the Mo steel for the RPV and to prepare for the licensing. The present paper summarizes the fracture toughness behavior including K_Id and K_Ia, irradiation embrittlement susceptibility and degradation of steel due to the long term aging at high temperature of the forged low Mo steel. These tests reveal good fracture toughness which well meets the requirements of the ASME Code, low neutron irradiation embrittlement susceptibility, little embrittlement by long term aging and so on. The present test results demonstrate good applicability of forged low Mo steel to the RPV of HTGR.     

The credit analysis of recycling beryllium and uranium in BeO-UO2 nuclear fuel

This study quantifies the credits of beryllium and uranium which are used as the raw materials for BeO-UO₂ nuclear fuel by analyzing the influence of their credits on the nuclear fuel cycle cost was analyzed, where the credit was defined as the value of raw materials recovered from spent fuel and the raw materials that were re-cycled. The credits of beryllium and uranium at 60 MWD/kg burn-up were –0.22 Mills/kWh and –0.14 Mills/kWh, respectively. These findings were based on the assumption that the optimal mixing proportion of beryllium in the BeO-UO₂ nuclear fuel is 4.8 wt%. In sum, the present study verified that the credits of beryllium and uranium in relation to BeO-UO₂ nuclear fuel are significant cost drivers in the cost of the nuclear fuel cycle and in estimating the nuclear fuel cycle of the reprocessing option for spent nuclear fuels.     

RPV material investigations of the former VVER-440 Greifswald NPP

The real toughness response of RPV material can only be determined after the final shut down of the NPP. Such a chance is given now by investigating material from the former Greifswald NPP (VVER-440/230).In the first part the paper deals with fast neutron fluence calculations and retrospective dosimetry based on Niobium. Unfortunately, a second neutron reaction besides ⁹³Nb(n,n’) leading to ^93mNb-activity is the reaction ⁹²Mo(n,γ)⁹³Mo. Based on the found Nb and Mo contents in the RPV material, it turned out that the ^93mNb generation on the Mo path mostly dominates over the fast neutron induced generation from Nb.The comparison between the calculated and the measured ^93mNb activities typically resulted in deviations of 50%. Possible reasons for the observed differences are discussed.In the second part first results of fracture mechanic investigations are reported. SE(B) specimens from three thickness positions were tested and evaluated according to the test standard ASTM E1921-05. Cleavage fracture toughness values, K_Jc, were determined and Master Curve based reference temperatures (T₀) were evaluated. The T₀ measured at the inner surface of the RPV did not represent the conservative condition. The T₀ of disc 1-1.3 located between the surface and 1/4 thickness is about 40K higher compared with those of the surface.The measured K_Jc values are not enveloped by the 5% fractile indexed with T₀ according to the Master Curve concept. However, the 5% fractile indexed with the VERLIFE reference temperature RTT_o that includes an additional margin envelops the measured K_Jc values. Therefore the VERLIFE lower bound curve conservatively describes the fracture toughness of the investigated weld metal.     

Wetting by sodium at high temperatures in pure vapour atmosphere

The wetting angle of sodium on different pure metals (Ni, Ta, Mb, Mo, W), alloys (stainless steel 304L, Inconel 600, RTG 36, TZM) and oxides (Feldmühle E37, Degussa AL23, sapphire, ZrO₂, Y₂O₃) was measured in the temperature range from 520 to 720°C. A new measurement method was applied consisting of a combination of the classical sessile drop with a gas controlled heat pipe. The method is especially suited for high temperatures where the previous methods are impracticable because of excessive sodium evaporation. It has furthermore the advantage that the measurements are made under pure sodium vapour. In the temperature range mentioned all investigated materials are well wetted. The maximum measured wetting angle is 7.5°. The wetting angles are nearly independent from temperature, but they show a time-dependency consisting in general of a rapid initial decrease and a reaching of the equilibrium value after about 1 hour.     

Characterization of the reaction layer in U-7wt%Mo/Al diffusion couples

The reaction layer in chemical diffusion couples U-7wt%Mo/Al was investigated using optical and scanning electron microscopy, electron probe microanalysis and X-ray diffraction (XRD) techniques. When the U-7wt%Mo alloy was previously homogenized and the γ(U, Mo) phase was retained, the formation of (U, Mo)Al₃ and (U, Mo)Al₄ was observed at 580 °C. Also a very thin band was detected close to the Al side, the structure of the ternary compound Al₂₀UMo₂ might be assigned to it. When the decomposition of the γ(U, Mo) took place, a drastic change in the diffusion behavior was observed. In this case, XRD indicated the presence of phases with the structures of (U, Mo)Al₃, Al₄₃U₆Mo₄, γ(U, Mo) and α(U) in the reaction layer.     

Characterization of the interaction layer grown by interdiffusion between U-7wt%Mo and Al A356 alloy at 550 and 340 °C

U(Mo) alloys are under study to get a low-enriched U fuel for research and test reactors. Qualification experiments of dispersion fuel elements have shown that the interaction layer between the U(Mo) particles and the Al matrix behaves unsatisfactorily. The addition of Si to Al seems to be a good solution. The goal of this work is to identify the phases constituting the interaction layer for out-of-pile interdiffusion couples U(Mo)/Al(Si). Samples γU-7wt%Mo/Al A356 alloy (7.1 wt%Si) made by Friction Stir Welding were annealed at 550 and 340 °C. Results from metallography, microanalysis and X-ray diffraction, indicate that the interaction layer at 550 °C is formed by the phases U(Al,Si)₃, U₃Si₅ and Al₂₀Mo₂ U, while at 340 °C it is formed by U(Al,Si)₃ and U₃Si₅. X-ray diffraction with synchrotron radiation showed that the Si-rich phase, previously reported in the interaction layer at 550 °C near U(Mo) alloy, is U₃Si₅.     

The effects of tungsten addition on the microstructural stability of 9Cr–Mo Steels

The effects of tungsten addition on the microstructure and high-temperature tensile strength of 9Cr–Mo steels have been investigated by using three different steels: M10 (9Cr–1Mo), W18 (9Cr–0.5Mo–1.8W), and W27 (9Cr–0.1Mo–2.7W) steels. The tungsten-added 9Cr steels have revealed better high-temperature tensile strength. Microchemical analysis for (Cr,Fe)₂ (C,N) revealed that the tungsten addition increased the Cr/Fe ratio, which resulted in the lattice expansion of (Cr,Fe)₂ (C,N), and then the enhanced pinning effect on the glide of dislocation. In addition, in M10 steel, the M₂₃C₆ carbides quickly grew and agglomerated, while the tungsten-added 9Cr steels revealed a fine and uniform distribution of M₂₃C₆ carbides. Dislocation recovery during tempering treatments was delayed in tungsten-added 9Cr steels, which was correlated with the stabilized precipitates and the decreased self-diffusivity of iron. It is, thus, believed that the excellent high-temperature tensile strength of tungsten-added 9Cr steels is attributed to the stabilized M₂X carbo-nitrides and M₂₃C₆ carbides and the decreased self-diffusivity of iron.     

Releases of Cesium and Poorly Volatile Elements from UO2 and MOX Fuels under Severe Accident Conditions

Radionuclide release from fuel under severe accident conditions has been investigated in the VEGA program at the Japan Atomic Energy Agency. In this program, three types of fuel, two UO₂ fuels irradiated at PWR and BWR and a MOX fuel irradiated at the ATR Fugen, were heated up to about 3130K in helium atmosphere at 0.1 MPa. Comparison of experimental data and evaluation with computer code analyses showed that Cs release is essentially identical among the three fuels. The Cs release from fuel may differ below about 1770K due to a difference in migration to grain boundaries during irradiation. The difference was not also observed for releases of poorly volatile elements, namely, U, Pu, Sr and Mo between UO₂ and MOX fuels. The release rate of Pu became slightly higher than that of U at 3130 K. The release rate of Sr increased at 3130 K, while that of Mo was quite low at temperatures above 2310 K.     

Investigation of alternative layouts for the supercritical carbon dioxide Brayton cycle for a sodium-cooled fast reactor

Analyses of supercritical carbon dioxide (S-CO₂) Brayton cycle performance have largely settled on the recompression supercritical cycle (or Feher cycle) incorporating a flow split between the main compressor downstream of heat rejection, a recompressing compressor providing direct compression without heat rejection, and high and low temperature recuperators to raise the effectiveness of recuperation and the cycle efficiency. Alternative cycle layouts have been previously examined by Angelino (Politecnico, Milan), by MIT (Dostal, Hejzlar, and Driscoll), and possibly others but not for sodium-cooled fast reactors (SFRs) operating at relatively low core outlet temperature. Thus, the present authors could not be sure that the recompression cycle is an optimal arrangement for application to the SFR. To ensure that an advantageous alternative layout has not been overlooked, several alternative cycle layouts have been investigated for a S-CO₂ Brayton cycle coupled to the Advanced Burner Test Reactor (ABTR) SFR preconceptual design having a 510 °C core outlet temperature and a 470 °C turbine inlet temperature to determine if they provide any benefit in cycle performance (e.g., enhanced cycle efficiency). No such benefits were identified, consistent with the previous examinations, such that attention was devoted to optimizing the recompression supercritical cycle. The effects of optimizing the cycle minimum temperature and pressure are investigated including minimum temperatures and/or pressures below the critical values. It is found that improvements in the cycle efficiency of 1% or greater relative to previous analyses which arbitrarily fixed the minimum temperature and pressure can be realized through an optimal choice of the combination of the minimum cycle temperature and pressure (e.g., for a fixed minimum temperature there is an optimal minimum pressure). However, this leads to a requirement for a larger cooler for heat rejection which may impact the tradeoff between efficiency and capital cost. In addition, for minimum temperatures below the critical temperature, a lower heat sink temperature is required the availability of which is dependent upon the climate at the specific plant site.     

Structure of UMo4O14 oxide

The structure and composition of UMo4O14 synthesized by solid state reaction method have been investigated by High Resolution Transmission Electron Microscopy (HRTEM), High Angle Annular Dark Field Scanning Transmission Electron Microscopy (HAADF-STEM), Selected Area Electron Diffraction SAED, and energy Dispersive X-ray microanalysis (EDX). The unit cell is orthorhombic, Cccm (66), a = 0.732, b = 3.247, c = 0.81, α = 90.0, β = 90.0, γ = 90.0. It comprises 2 layers of U and Mo cations. Each following layer is shifted relatively to the previous one by vector a/2. The infinite layers of composition Mo₂O₇ running perpendicular to b direction can be easily distinguished in the image. These double MoO₃ octahedra wide slabs are typical for U–Mo oxides. They are linked to each other by another MoO₃ octahedra and infinite one octahedral wide slab, forming a two rows of hexagonal coordinated bi-pyramids occupied by …U–O–Mo–O… strings. One of characteristic features of the UMo₄O₁₄ phase is the incommensurate modulation observed in some crystals. The goal of this paper is to reveal three-dimensional atomic structure of U–Mo oxide and to clarify in what way it depends on U/Mo composition.