Development of a new casting method to fabricate U–Zr alloy containing minor actinides

Metal fuel slugs of U–Zr alloys for a sodium-cooled fast reactor (SFR) have conventionally been fabricated using an injection casting method. However, casting alloys containing volatile radioactive constituents, such as Am, are problematic in a conventional injection casting method. As an alternative fabrication method, low pressure gravity casting has been developed. Casting soundness, microstructural characteristics, alloying composition, density, and fuel losses were evaluated for the following as-cast fuel slugs: U–10 wt% Zr, U–10 wt% Zr–5 wt% RE, and U–10 wt% Zr–5 wt% RE–5 wt% Mn. The U and Zr contents were uniform throughout the matrix, and impurities such as oxyen, carbon, and nitrogen satisfied the specification of total impurities less than 2,000 ppm. The appearance of the fuel slugs was generally sound, and the internal integrity was shown to be satisfactory based on gamma-ray radiography. In a volatile surrogate casting test, the U–Zr–RE–Mn fuel slug showed that nearly all of the manganese was retained when casting was done under an inert atmosphere.     

Fabrication of uranium alloy fuel slug for sodium-cooled fast reactor by injection casting

Metal fuel slugs of U–Zr alloys for a sodium-cooled fast reactor (SFR) have been fabricated using an injection casting method. However, casting alloys containing volatile radioactive constituents such as Am can cause problems in a conventional injection casting method. Therefore, in this study, several injection-casting methods were applied to evaluate the volatility of the metal-fuel elements and control the transport of volatile elements. Mn was selected as a volatile surrogate alloy since it possesses a total vapor pressure equivalent to that of minor actinide-bearing fuels for SFRs. U–10 wt% Zr and U–10 wt% Zr–5 wt% Mn metal fuels were prepared, and the casting processes were evaluated. The casting soundness of the fuel slugs was characterized by gamma-ray radiography and immersion density measurements. Inductively coupled plasma atomic emission spectroscopy was used to determine the chemical composition of fuel slugs. Fuel losses after casting were also evaluated according to the casting conditions.     

Nano-crystalline zirconia: a viable adsorbent for the column chromatographic separation of 58Co from neutron irradiated nickel targets

Owing to the high chemical reactivity of molten uranium alloys, the use of traditional graphite crucibles for casting fuel slugs for a sodium-cooled fast reactor (SFR) is problematic. Moreover, rare earth (RE) elements retained in the fuel slugs for an SFR, which are extracted from the spent fuel by pyro-processing, are more reactive than uranium melt. Therefore, in this study, Y₂O₃ single-layer coatings with thicknesses of approximately 50, 70, and 120 μm and double-layer coatings of TaC/Y₂O₃ and Y₂O₃/TaC were plasma-sprayed onto niobium substrates and tested for thermal shock resistance and compatibility against U–10 wt% Zr and U–10 wt% Zr–5 wt% RE melt. The Y₂O₃ single-layer coating, regardless of coating thickness, and the TaC/Y₂O₃ double-layer coating showed good contact at the interface between the coating and the niobium substrate, with no deterioration after the thermal cycling test. In the interaction studies, the single- and double-layer coatings showed good compatibility with the U–Zr melt. However, the Y₂O₃ coatings with thicknesses of approximately 50 and 70 μm showed severe penetration of the U–Zr–RE melt and reacted with the niobium substrate. The single-layer Y₂O₃ coating with a thickness of 120 μm and the double-layer TaC/Y₂O₃ coating exhibited the most promising performance among the candidate coatings.     

Quantitative and isotopic analysis of released and retained krypton and xenon fission gases from irradiated metallic fuels

We quantitatively measured the amounts and isotopic distributions of the released and retained fission gases (Kr and Xe) from two irradiated metallic fuels (U–10Zr and U–10Zr–5Ce) at approximately 2.9 at.% burnup, using a gas chromatography and a quadrupole mass spectrometer. The obtained Xe/Kr ratios indicate that the released and retained fission gases from the irradiated metallic fuels came primarily from the fission of ²³⁵U, instead of that of heavy isotopes such as ²³⁹Pu and ²⁴¹Pu. The calculated (⁸³Kr + ⁸⁴Kr)/⁸⁶Kr and (¹³¹Xe + ¹³²Xe)/¹³⁴Xe ratios suggest that no fuel rods became defective during the irradiation process.     

Quantification of Nb and Zr in the irradiated pressure tubes from Wolsong CANDU Unit 1 by neutron activation analysis

When the Wolsong 1st CANDU reactor in Korea is reaching the end of its original lifespan, decommissioning has been one of the largest issues faced by the nuclear industry and the government for the estimation of expense. The radioactive sources in the construction stuffs and components of the reactor such as pressure tubes, steam generators must be evaluated for the estimation of the appropriate expense for the decommissioning. Unfortunately we do not have exact information about the initial composition of the pressure tubes. In this study the elemental contents of Nb and Zr in the activated zircaloy pressure tubes was investigated. Nb is one of the important elements that characterize the physical and mechanical properties of the pressure tube and a major source of high radioactivity. The prepared samples were strongly irradiated again in the vertical irradiation channel at the HANARO research reactor in the Korea Atomic Energy Research Institute (KAERI) and then measured by an HPGe detector.     

New approaches to reprocessing of oxide nuclear fuel

Dissolution of UO₂, U₃O₈, and solid solutions of actinides in UO₂ in subacid aqueous solutions (pH 0.9–1.4) of Fe(III) nitrate was studied. Complete dissolution of the oxides is attained at a molar ratio of ferric nitrate to uranium of 1.6. During this process actinides pass into the solution in the form of U(VI), Np(V), Pu(III), and Am(III). In the solutions obtained U(VI) is stable both at room temperature and at elevated temperatures (60 °C), and at high U concentrations (up to 300 mg mL^?1). Behavior of fission products corresponding to spent nuclear fuel of a WWER-1000 reactor in the process of dissolution the simulated spent nuclear fuel in ferric nitrate solutions was studied. Cs, Sr, Ba, Y, La, and Ce together with U pass quantitatively from the fuel into the solution, whereas Mo, Tc, and Ru remain in the resulting insoluble precipitate of basic Fe salt and do not pass into the solution. Nd, Zr, and Pd pass into the solution by approximately 50 %. The recovery of U or jointly U + Pu from the dissolution solution of the oxide nuclear fuel is performed by precipitation of their peroxides, which allows efficient separation of actinides from residues of fission products and iron.     

Fast neutron induced fission neutron spectra below the incident energy

A fast pneumatic transfer system for an instrumental neutron activation analysis and delayed neutron counting system were reconstructed with new designs of a functional improvement at the HANARO research reactor in 2006. The design, conception, operation and control of these systems are described. Also the experimental characteristic parameters by a functional operation test and an irradiation test of these systems, such as the transfer time, the neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc., are reported to provide a user information as well as for the management and safety of the reactor.     

Application of the <Emphasis Type="Italic">k</Emphasis><Subscript>0</Subscript>-NAA method at the HANARO research reactor

The k ₀-standardization method (k ₀-NAA) is known as one of the most remarkable progresses of the NAA with its many advantages. For the application of k ₀-NAA method at the NAA #1 irradiation position where the neutrons are well thermalized in the HANARO research reactor, KAERI, Korea, the determination of the reactor neutron spectrum parameters such as α and f have been carried out. The measured values of α and f using the “Cd-ratio” triple monitor method were 0.127±0.022 and 1010±70, respectively. To evaluate the applicability of k ₀-NAA in our analytical system, the analysis of three kinds of SRMs was executed. The analytical results showed that the relative error of most of the elements was less than 10% and the U-scores were within 2. It is turned out that the procedure of the k ₀-NAA in the HANARO research reactor is available for a practical application in the environmental fields.     

Development of a method for the determination of 93Zr and 94Nb in radioactive waste using TEVA® resin

The behaviour of ⁹⁵Zr and ⁹⁵Nb isotopes on TEVA^® resin column was studied. The influence of hydrofluoric, hydrochloric, oxalic and sulphuric acids on the separation and chemical yield of zirconium and niobium was tested. The method is suitable for the separation of ⁹³Zr and ⁹⁴Nb isotopes from hydrofluoric, hydrochloric and sulphuric acid by using different concentrations of the acid for the load or elutes solutions. Samples from reactor and fuel assemblies such as the control-rod connection rod, shielding/absorber parts of fuel rods, neutron in-core measurement channels, pressure vessel basic construction material and internal cladding, core barrel and reactor protection tube unit were analysed.     

Trace element determination in high-purity aluminum clad samples by k 0-based internal monostandard instrumental neutron activation analysis

The k₀-based internal monostandard instrumental neutron activation analysis (IM-INAA) method was applied for quantification of trace impurities in seven high purity aluminum samples used as fuel cladding in a research reactor. Samples along with BCS CRM 182 (Si–Al alloy) were irradiated in high flux reactor neutrons for 10–15 h. In situ detection efficiency, needed in concentration calculation ratio by IM-INAA, was obtained using gamma rays of activation products produced in the samples. Elemental concentration ratios obtained with respect to Fe (used as internal monostandard) were converted to absolute concentrations by determining concentration of Fe by relative method of NAA. Concentrations of ten trace elements (Sc, Cr, Co, Zn, Ga, La, Ce, Sm, W and Hf) including major element Fe were determined in this work. The method used is non-destructive in nature and does not need multielement standards. Results of IM-INAA were compared with those obtained by relative method of INAA and inductively coupled plasma atomic emission spectrometry (ICP-AES). Details of methodologies and results obtained by all methods are discussed and compared in the paper.     

Use of zirconium and ruthenium isotope correlations in the analysis of spent nuclear fuels

An approach is proposed to use the isotope ratios⁹³Zr/⁹⁶Zr,⁹¹Zr/⁹⁶Zr and¹⁰¹Ru/¹⁰⁴Ru for the determination of plutonium contriubtion to the total burnup of a spent nuclear reactor fuel of low initial enrichment in²³⁵U (3.6%). This approach was verified by determining the total burnup of the fuel from the¹³⁷Cs and¹⁴⁸Nd monitors and subtracting the partial burnup due to²³⁵U from the value obtained.     

Instrumental neutron activation analysis of coal and its combustion residues from a power plant

A growing demand of electrical energy derived from coal combustion led to a significant increase of coal ash as residues. Approximately 70 % of the fly ashes are recycled, while most of the bottom ashes have been land-filled in the ash pond in Korea. In this work, to evaluate the potential impacts of the residues from a coal power plant on the environment, its inorganic elemental components were determined by INAA and PGAA. Coal ash samples were collected from the biggest power plant complex in Korea. These samples were analyzed by using the NAA facilities in the HANARO research reactor of the Korea Atomic Energy Research Institute. A total of 31 elements were analyzed in the samples, and certified reference materials were used for the analytical quality control. The enrichment status of a given metal in fuel coal and ashes was investigated by its concentration ratio. In order to assess the impact of the coal combustion residues on ecosystem, their concentrations determined for each respective type of the samples were compared to both reference data and nearby beach sand samples.     

Implementation of the k 0-standardization method for analysis of geological samples at the Neutron Activation Analysis Laboratory, São Paulo, Brazil

The Neutron Activation Analysis Laboratory (LAN-IPEN) has been analysing geological samples for many years with the INAA comparative method, for geochemical and environmental research. This study presents the results obtained in the implementation of the k ₀-standardization method at LAN-IPEN, for geological samples analysis, by using the program k ₀-IAEA, provided by the International Atomic Energy Agency (IAEA). The thermal to epithermal flux ratio f and the shape factor α of the epithermal flux distribution of the IEA-R1 nuclear reactor of IPEN were determined for the pneumatic irradiation facility and one selected irradiation position, for short and long irradiations, respectively. To obtain these factors, the “bare triple-monitor” method with ¹⁹⁷Au–⁹⁶Zr–⁹⁴Zr was used. In order to validate the methodology, the geological reference materials basalts JB-1 (GSJ) and BE-N (IWG-GIT), andesite AGV-1 (USGS), granite GS-N (ANRT), SOIL-7 (IAEA) and sediment Buffalo River Sediment (NIST–BRS-8704), which represent different geological matrices, were analysed. The concentration results obtained agreed with assigned, with bias <10 % except for Zn in AGV-1. The U-score test showed that all results, except Mg in JB-1, are within 95 % confidence interval. These results indicate excellent possibilities of using this parametric method at the LAN-IPEN for geochemical and environmental studies.     

Characterization of reused U–10Zr heel residue containing rare-earth elements through surface treatment

Journal of Radioanalytical and Nuclear Chemistry - U–Zr–RE heel residue was manufactured after fabricating U–Zr–RE fuel slugs by an injection casting method. In order to...     

Implementation of the <Emphasis Type="Italic">k</Emphasis><Subscript>0</Subscript>-NAA method by using <Emphasis Type="Italic">k</Emphasis><Subscript>0</Subscript>-IAEA software and the NAA#3 irradiation hole at the HANARO research reactor

This study intended to demonstrate the use of k ₀-IAEA software with the NAA#3 irradiation hole in the HANARO research reactor and the gamma-ray spectrometers at the NAA laboratory in the Korea Atomic Energy Research Institute (KAERI). NIST SRM 2586-Soil was chosen as a sample to validate the analytical protocol by using the k ₀-IAEA software. Twenty four elements were determined and the resultant deviations between the experimental results and the certified values were all within 10% except for Dy. In addition, an analysis of three types of synthetic multielement standards (SMELS) prepared by the Institute for Reference Materials and Measurements (IRMM) and Institute of Nuclear Science (INW) in Belgium was executed. The analytical results agreed well with the assigned values and the U-scores for most of the elements were lower than 2.     

EPMA and X-Ray Diffraction of the Degraded Fuel Bundle from the Phebus FPT1 Test

 The degraded fuel rod bundle from the second Phebus test (FPT1) was examined at ITU, Karlsruhe. Metallographic and microprobe analysis of the degraded fuel pieces were carried out. The fuel samples from the upper bundle and cavity edge were porous and had remnants of thick oxidised cladding adhering to them. Electron microprobe analysis line scans across the cladding-fuel interface showed interdiffusion of U and Zr, with U diffusing down the grain boundaries of the oxidised cladding, while point analyses revealed noticeable amounts of Zr (1.5–4.2 wt%) in the UO₂ fuel. EPMA oxygen measurements revealed in the upper part of the bundle a superstoichiometry of x = 0.3–0.4 in UO _2 + x , indicating that fuel fragments in this position had undergone considerable oxidation. X-ray diffraction of the corium pool disclosed a deformed cubic fluorite lattice of UO₂. The lattice parameter of a = 5.2984 ? was considerably reduced compared with pure, stoichiometric UO₂ and was consistent with a lattice containing approximately 45 mol% ZrO₂ that had undergone little oxidation. The corium’s nominal composition of (U_0.5Zr_0.5) O₂ also corresponded to its observed single phase microstructure.     

Metallic Fuels: The EBR-II Legacy and Recent Advances

Experimental Breeder Reactor – II (EBR-II) metallic fuel was qualified for high burnup to approximately 10 atomic per cent. Subsequently, the electrometallurgical treatment of this fuel was demonstrated. Advanced metallic fuels are now investigated for increased performance, including ultra-high burnup and actinide burning. Advances include additives to mitigate the fuel/cladding chemical interaction and uranium alloys that combine Mo, Ti and Zr to improve alloy performance. The impacts of the advances—on fabrication, waste streams, electrorefining, etc.—are found to be minimal and beneficial. Owing to extensive research literature and computational methods, only a modest effort is required to complete their development.     

Chemical and microstructural characterization of remelted Zircaloy by X-ray fluorescence techniques and metallographic analysis

Brazilian nuclear power reactor (PWR type) use, as nuclear fuel, sintered UO₂ pellets with Zircaloy cladding. The cladding material has to present high mechanical strength and corrosion resistance, which are related to the material chemical composition and microstructural characteristics. Zircaloy machining scraps were remelted using VAR process, resulting in a billet, and its elemental composition was determined. The major elements Zr, Sn, Fe, Cr and Ni were determined by EDXRFS; Hf and contaminants were determined by WDXRFS. The chemical analysis did not show changes in the alloy elemental composition, except for Fe and Cr, which their out-off content can be attributed to contamination from steel scraps. The found Cu contamination may be related to the contact of the melted Zircaloy droplets with copper crucible. The billet microstructure was evaluated using optical and scanning electron microscopy and showed the typical zirconium alloy microstructures (needle and plates) resulting from the high cooling rate and thermal gradients, present during the solidification.     

Application of the isotope dilution technique for Zr determination in an irradiated cladding material by multiple collector-inductively coupled plasma mass spectrometry

The determination of ⁹³Zr concentration, a long-lived radionuclide present in spent nuclear fuel and in the structural components of nuclear reactors, is a major issue for nuclear waste disposal purpose and to validate neutronic calculation codes. To measure ⁹³Zr concentration in irradiated cladding material with a high precision, an analytical method based on the use of multiple collector-inductively coupled plasma mass spectrometer (MC-ICPMS) combined to isotope dilution technique was developed. First a radiochemical separation of zirconium from a zircaloy sample (a zirconium alloy used as a cladding material for nuclear fuel elements), has allowed to obtain a very pure zirconium fraction with no potential isobaric interferences for mass spectrometric measurements. Then as the determination of all zirconium isotope ratios in the sample is necessary for the isotope dilution method, a MC-ICPMS procedure was developed to perform these precise measurements. Finally, the determination of ⁹³Zr concentration in the same sample was performed, after preparation and calibration of a ⁹⁶Zr spike solution. The uncertainties obtained on isotope ratios of zirconium by MC-ICPMS were in the order of 0.1%. The final uncertainty obtained on the ⁹³Zr concentration in the nuclear material used and after chemical purification was lower than 0.6%.     

Characterization of corrosion products deposited on fuel rods in a boiling water reactor (BWR-6) plant using hydrogen water chemistry (HWC)

The objective of this paper was to characterize the corrosion products deposited on fuel cladding surfaces under hydrogen water chemistry (HWC) conditions in the boiling water reactor unit of a nuclear power plant. Inductively coupled plasma-optical emission spectroscopy and gamma spectrometry were used to determine the elemental composition and radioactivity along the fuel rods. Scanning electron microscopy/energy dispersive X-ray spectroscopy was used to evaluate crud morphology and X-ray diffraction analysis was employed to determine the crystalline phase. Most of the iron in the crud comprised irregular shapes and flakes, with a number of crystallites (e.g. small quantities of ferrites, MnFe₂O₄). The major constituent of the outer layer was a loosely attached, stable form of hematite (α-Fe₂O₃). The amount of deposition on the fuel rods following the fourth fuel cycle of HWC (~1.0 mg cm^?2) was significantly greater than after the third cycle, due to extended burn-up time. In addition, the radioactivity ratio of ⁵⁸Co to ⁶⁰Co was lower for EOC22 than for EOC21, due to a longer storage period in the spent fuel pool. No apparent changes were observed in the characteristics (composition or oxidizing nature) of crud in the fuel bundles. Finally, no direct negative impact was observed on the fuel cladding surfaces following four fuel cycles of HWC, compared with normal water chemistry conditions.