Natural UO Research Articles

A non destructive technique for determination of U 233 and Pu content in MOX fuel pins for fast reactors

MOX fuel pins containing both U 233O 2 and PuO 2 have been fabricated for making an experimental subassembly for irradiation in Fast Breeder Test reactor (FBTR) at Kalpakkam, India. This unique composition of the fuel pin is chosen to simulate the thermo-mechanical conditions of the upcoming Prototype Fast Breeder Reactor (PFBR) in the existing Fast Breeder Test Reactor. Since the fertile matrix is natural UO 2, it was difficult to monitor the percentage of U 233O 2 through chemical methods and neutron assay methods. During the fabrication of MOX fuel pins at Advanced Fuel Fabrication Facility; Bhabha Atomic Research Centre, Tarapur, Passive Gamma Scanning (PGS) was employed as one of the characterisation tools for verifying the fuel composition. PGS was found to be effective in estimating the percentage composition of both U 233O 2 and PuO 2 and also in ensuring the uniform distribution of the fissile material in MOX fuel pins. PGS is also found effective in monitoring the correct loading of natural UO 2 insulation pellets and MOX fuel pellets in welded MOX pins.

O and H diffusion in uraninite: Implications for fluid–uraninite interactions, nuclear waste disposal, and nuclear forensics

Diffusion coefficients for oxygen and hydrogen were determined from a series of natural uraninite–H 2O experiments between 50 and 700 °C. Under hydrous conditions there are two diffusion mechanisms: (1) an initial extremely fast-path diffusion mechanism that overprinted the oxygen isotopic composition of the entire crystals regardless of temperature and (2) a slower volume-diffusive mechanism dominated by defect clusters that displace or eject nearest neighbor oxygen atoms to form two interstitial sites and two partial vacancies, and by vacancy migration. Using the volume diffusion coefficients in the temperature range of 400–600 °C, diffusion coefficients for oxygen can be represented by D = 1.90e −5 exp (−123,382 J/RT) cm 2/s and for temperatures between 100 and 300 °C the diffusion coefficients can be represented by D = 1.95e −10 exp (−62484 J/RT) cm 2/s, where the activation energies for uraninite are 123.4 and 62.5 kJ/mol, respectively. Hydrogen diffusion in uraninite appears to be controlled by similar mechanisms as oxygen. Using the volume diffusion coefficients for temperatures between 50 and 700 °C, diffusion coefficients for hydrogen can be represented by D = 9.28e −6 exp (−156,528 J/RT) cm 2/s for temperatures between 450 and 700 °C and D = 1.39e −14 exp (−34518 J/RT) cm 2/s for temperatures between 50 and 400 °C, where the activation energies for uraninite are 156.5 and 34.5 kJ/mol, respectively. Results from these new experiments have implications for isotopic exchange during natural UO 2–water interactions. The exceptionally low δ 18O values of natural uraninites (i.e. −32‰ to −19.5‰) from unconformity-type uranium deposits in Saskatchewan, in conjunction with theoretical and experimental uraninite–water and UO 3–water fractionation factors, suggest that primary uranium mineralization is not in oxygen isotopic equilibrium with coeval clay and silicate minerals. The low δ 18O values have been interpreted as resulting from the low temperature overprinting of primary uranium mineralization in the presence of relatively modern meteoric fluids having δ 18O values of ca. −18‰, despite petrographic and U–Pb isotope data that indicate limited alteration. Our data show that the anomalously low oxygen isotopic composition of the uraninite from the Athabasca Basin can be due to meteoric water overprinting under reducing conditions, and meteoric water or groundwater can significantly affect the oxygen isotopic composition of spent nuclear fuel in a geologic repository, with minimal change to the chemical composition or texture. Moreover, the rather fast oxygen and hydrogen diffusion coefficients for uraninite, especially at low temperatures, suggest that oxygen and hydrogen diffusion may impart characteristic isotopic signals that can be used to track the route of fissile material.

Burn up extension with slightly enriched uranium and plutonium fuel in CANDU reactors

A study of fuel burn-up and concentrations of uranium and plutonium isotopes for the three fuel cycles of a CANDU reactor are carried out in the present work. The infinite and effective multiplication factors are calculated as a function of fuel burn up for the natural UO 2 fuel, 1.2% enriched UO 2 fuel and for the 0.45% PuO 2–UO 2 fuel. The amount of 235U and 238U consumed and 239Pu, 240Pu and 241Pu produced in the three fuel cycles are also calculated and compared.

Post irradiation examination of thermal reactor fuels

The post irradiation examination (PIE) facility at the Bhabha Atomic Research Centre (BARC) has been in operation for more than three decades. Over these years this facility has been utilized for examination of experimental fuel pins and fuels from commercial power reactors operating in India. In a program to assess the performance of (U,Pu)O 2 MOX fuel prior to its introduction in commercial reactors, three experimental MOX fuel clusters irradiated in the pressurized water loop (PWL) of CIRUS up to burnup of 16 000 MWd/tU were examined. Fission gas release from these pins was measured by puncture test. Some of these fuel pins in the cluster contained controlled porosity pellets, low temperature sintered (LTS) pellets, large grain size pellets and annular pellets. PIE has also been carried out on natural UO 2 fuel bundles from Indian PHWRs, which included two high burnup (∼15 000 MWd/tU) bundles. Salient investigations carried out consisted of visual examination, leak testing, axial gamma scanning, fission gas analysis, microstructural examination of fuel and cladding, β, γ autoradiography of the fuel cross-section and fuel central temperature estimation from restructuring. A ThO 2 fuel bundle irradiated in Kakrapar Atomic Power Station (KAPS) up to a nominal fuel burnup of ∼11 000 MWd/tTh was also examined to evaluate its in-pile performance. The performance of the BWR fuel pins of Tarapur Atomic Power Stations (TAPS) was earlier assessed by carrying out PIE on 18 fuel elements selected from eight fuel assemblies irradiated in the two reactors. The burnup of these fuel elements varied from 5000 to 29 000 MWd/tU. This paper provides a brief review of some of the fuels examined and the results obtained on the performance of natural UO 2, enriched UO 2, MOX, and ThO 2 fuels.

Alteration of UO 2+ x under oxidizing conditions, Marshall Pass, Colorado, USA

As a natural analogue of the processes and products of spent nuclear fuel (SNF) alteration, we have examined the sequence of phases that form during the alteration of natural UO 2+ x in a U-deposit at Marshall Pass, Colorado. We have determined the paragenesis of U(VI)-phases including the fate of trace elements: W, Mo, As, Sb, Cu, Ba, Ce, Y, Pb and Th. Enrichment of trace elements, especially W and Mo, in this system resulted in a unique alteration sequence: uraninite → amorphous U-oxyhydrate gel → schoepite(I)/vandendriesscheite/compreignacite → uranophane → schoepite(II)/“dehydrated” schoepite(I) → Ba–Mo–W–U phase/U-arsenates/U–Sb phase → “dehydrated schoepite” (II) → soddyite/swamboite. In this sequence, the Ba–Mo–W uranyl phase and U–Sb phase are newly characterized phases. These results suggest that the UO 2+ x alteration, involving higher concentrations of certain radionuclides and metallic compounds, may lead to a different paragenesis of U(VI)-phases, as compared with the expected alteration sequence of UO 2 interacting with a typical groundwaters. This was also noted in a previous study of the alteration of Pb-rich uraninite [R.J. Finch, R.C. Ewing, J. Nucl. Mater. 190 (1992) 133–156]. Some trace elements, such as CaO 2.08 wt.%, PbO 1.69 wt.%, WO 3 1.39 wt.%, As 2O 3 0.50 wt.% and MoO 3 0.41 wt.%, can locally concentrate, but still form uranyl phases. As a consequence, the mobility of U and radionuclides is governed by the stability of these metal-uranyl phases, such as Pb-oxide hydrates, Ba-uranyl molybdates/ tungstates and U-antimonate.

High power ramping in commercial PHWR fuel at extended burnup

Argentinean Atucha and Embalse NPP, both PHWR with on line refuelling with approximately 68,000 fuel elements, irradiated at standard burnup, during the refuelling the fuel changes the power close to a mathematical ideal step. Recently, Atucha core started to operate fully with slightly enriched uranium (SEU, 0.85% of enrichment) increasing the burnup more than 50%, thus producing the first irradiated data base of on line refuelling at extended burnup using commercial fuel elements. The BACO code was extensively used in Argentina to model the physical behaviour of both fuel elements, natural UO 2 and SEU. The hoop stress predicted by BACO at the inner surface of the cladding correlate very well with the fuel failure probability over a wide range of tests. Using the data base of present commercial extended burnup fuel, a simple criterion of fuel failure taking into account probabilistic and parametric analysis correlate very well with the present irradiation experience, in agreement with previous BACO experience which confirm the expected behaviour and the SEU fuel element design.

Power flattening in the fuel bundle of a CANDU reactor

The strong non-uniformity of the fission power production density in the CANDU fuel bundle could have been mitigated to a great degree. A satisfactory power flattening has been achieved through an appropriately evaluated method by varying the composition of the LWR spent fuel/ThO 2 mixture in a CANDU fuel bundle in radial direction and keeping fuel rod dimensions unchanged. This will help also to greatly simplify fuel rod fabrication and allow a higher degree of quality assurance standardization. Three different bundle fuel charges are investigated: (1) the reference case, uniformly fueled with natural UO 2, (2) a bundle uniformly fueled with LWR spent fuel, and (3) a bundle fueled with variable mixed fuel composition in radial direction leading to a flat power profile (100% LWR spent fuel in the central rod, 80% LWR + 20% ThO 2 in the second row, 60% LWR + 40% ThO 2 in the third row and finally 40% LWR + 60% ThO 2 in the peripheral fourth row). Burn-up grades for these three different bundle types are calculated as ∼7700, ∼27,300, and 10,000 MW.D/MT until reaching a lowest bundle criticality limit of k ∞ = 1.06. The corresponding plant operation periods are 170, 660, and 240 days, respectively.

Theoretical calculations for the production of 99Mo using natural uranium in Iran

In the field of medical diagnostics, 99mTc with a half life of 6 h is widely used for medical purposes and produced from the decay of 99Mo ( t 1/2=66 h). The need for this isotope in the Iranian medical centres is about 20 Ci/week and is currently being imported. In this paper, the theoretical calculations related to the production of 99Mo, using natural UO 2 and irradiating it within the Tehran Research Reactor are presented. The calculations are carried out by the use of the MCNP-4A and ORIGEN-2 codes. It is expected that following chemical separations, approximately 20 Ci/week of 99Mo may be obtained by irradiating 100 g of natural UO 2 for a period of 7 days inside the reactor. This amount may be produced every other week due to the limitations on the reactor operation. Detailed calculations have been presented to the proper regulatory bodies for approval. Once authorized, the production is expected to begin in the first half of 2003.

Investigation of the effects of the resonance absorption in a fusion breeder blanket

In generic neutronic studies of a fusion breeder, resonance absorption is mostly neglected. In this work, the effects of resonances on the neutronic parameters a fast and moderated fusion blanket are investigated. In the present case, a ( D, T) fusion reactor acts as an external high energetic (14.1 MeV) neutron source. The fissile fuel zone, containing 10 rows in radial direction, covers the cylindrical fusion plasma chamber. The fissile fuel is natural UO 2. Fissile zone is cooled (a) with pressurised helium gas for the fast blanket and (b) with light water for the moderated, each of them with a volume ratio of V coolant/ V fuel=2 in the fissile zone. The study has shown that careful resonance self-shielding calculations are indispensable for neutronic studies of a moderated fusion blanket. On the other hand the omission of the resonance self-shielding in generic studies of a fast fusion blanket can be tolerated to some degree. Furthermore, a correct description of the fusion neutron source spectrum has great importance on neutronic parameters, along with the resonance self-shielding calculations.

Simfuel Leaching Experiments in Presence of Gamma External Source (60CO)

One of the factors considered within the studies of performance assessment on spent fuel under final repository conditions is the effect of the radiation on its leaching behavior. Radiation from spent fuel can modify some properties of both solid phase and leachant and therefore it would alter the chemical behavior of the near field. Particularizing in the effect of the radiation on the leachant, it will cause generation of radiolytic species that could change the redox potential of the environment and therefore may bring on variations in the leaching process. In this work, the chemical analogue utilized was SIMFUEL (natural UO 2 doped with non-radioactive elements simulating fission products) and the leachants selected were saline and granite bentonite waters both under initial anoxic conditions. To emulate γ radiation field of a spent fuel, leaching experiments with external 60 Co sources in a irradiation facility (Nayade) were performed. Initial dose rate used was 0.014 Gy/s. Preliminary results indicate that radiation produces an increase of the uranium dissolution rate, being the concentrations measured close to those obtained in oxic atmosphere without radiation field. In addition the solubility solid phases from experimental conditions were calculated, for both granite bentonite water and 5 m NaCl media. On the other hand, a tentative approach to model the role of γ radiolysis in these SIMFUEL tests has been carried out as well.

Radiation damage in nuclear fuel materials: the “rim” effect in UO 2 and damage in inert matrices for transmutation of actinides

Polygonization or grain subdivision at high damage levels (often referred to as “rim effect”) is gaining increasing importance for UO 2 which is the fuel of nuclear electricity producing power stations. Also, recently, the concept of replacing U-238, and hence replacing natural UO 2, by “inert matrices” for fissile U-235 or Pu-239, or also for transmutation of higher actinides (Np, Am, Cm) is being studied. High energy ion implantation was used to simulate fission damage in both UO 2 and in such matrices (e.g. Al 2O 3, MgAl 2O 4 etc.). The energy of implanted ions was varied between 40 keV and 2.6 GeV. Emphasis is given to the search for possible dose rate effects. Results on damage formation in UO 2 exist now for dose rate variations by more than a factor of 10 10. The implications for the technological application of inert matrix materials are briefly discussed.

Electrochemical experiments on UO 2 corrosion in aqueous solutions

Electrodes of natural UO 2 have been electrochemically tested in various aerated and deaerated 3% Na 2CO 3 and other solutions. UO 2 rapidly takes up dissolved oxygen to form UO 2+χ of corresponding rest potential. UO 2 oxidizes in several stages; the rate of oxidation rises with increasing (anodic) potentials. The attack and redeposition of corrosion products on the electrode appears to be localized.

X-ray powder diffraction study of annealed uraninite

Natural UO 2+ x (uraninite) samples were annealed in a reducing atmosphere up to 1200°C. The initial unit cell parameters of 0.5385, 0.5439 and 0.5458 nm increased after annealing to 0.5469, 0.5447 and 0.5462 nm, respectively. In one sample the unit cell parameter decreased from 0.5463 nm to 0.5456 nm. Migration of point defects in the oxygen sub-lattice occurs at 300°C. In the interval of 500 to 700°C, uranium vacancies migrate and cluster. The formation of uranium vacancies is enhanced by the presence of radiogenic Pb. Above 750°C, the unit cell parameter increases as a result of reduction U 6+ to U 4+. Uraninite incongruently decomposes at 1100°C. Density increased after annealing except for one sample in which a large number of bubbles (1–2 μm in size) formed. The complex annealing behavior of uraninite is due to the presence of impurities (Pb, Ca, Th, Zr, REE), mixed oxidation states of uranium, and α-decay event radiation damage.

Characterization of HWR fuel pellets fabricated using UO 2 powders from different conversion processes

Heavy-water-reactor (HWR) fuel manufacturing involves the fabrication of high density natural UO 2 pellets with controlled microstructure. It is necessary to investigate the correlation between powder characteristics, pellet properties and process parameters, especially when the UO 2 powders are from different conversion processes. To establish the UO 2 pelletizing technology better to obtain such tightly quality controlled pellets, it is found necessary, apart from the quality requirements described in various specifications, to determine the particle morphology, the pore size distribution in particular, with regard to the compaction behaviour, which strongly affects the densification during sintering. Pore structures of different UO 2 powders converted via the AUC and ADU processes and the evolution of pore size distribution in the compacted pellets from these different UO 2 powders are analysed and compared by mercury porosimetry and microstructure observations of compacted pellets by scanning electron microscopy. Further, the connection of the evolution of pores in compacted pellets to the attainable densities of sintered pellets is discussed for the optimization of process parameters and as a proposal for quality control requirements and methods.

Alteration of Natural UO2 under Oxidizing Conditions from Shinkolobwe, Katanga, Zaire: A Natural Analogue for the Corrosion of Spent Fuel

Alteration of Natural UO₂ under Oxidizing Conditions from Shinkolobwe, Katanga, Zaire: A Natural Analogue for the Corrosion of Spent Fuel

Neutron Spectrum Dependence of Natural UO2 Doppler Effect Measured in FCA

Neutron Spectrum Dependence of Natural UO₂ Doppler Effect Measured in FCA

Finite element analysis of local overheating within plutonium enriched UO2 fuel rods caused by PuO2 islands

Abstract Within natural UO 2 fuel elements enriched with plutonium, this last material should form PuO 2 solid solutions inside the UO 2 pellets, in a wide range of concentrations. If the solutions are obtained by mechanical mixing of the oxides, PuO 2 islands are formed in the UO 2 matrix. These islands may be the source of several problems in the fuel behaviour, the most important being the overheating of the matrix in the neighbourhood of the particles. It is caused by the large fission cross section of plutonium compared with that of uranium. A detailed study of the thermal effects produced by PuO 2 particles in the UO 2 matrix and the cladding is then important for the specification of their permissible size. A portion of the fuel rods with spherical particles in the most significant places was studied. In order to obtain the dimensionless overheating of the fuel and cladding produced by the presence of those particles, the spacial distribution of temperature was calculated, solving the stationary and linear bidimensional equation of heat conducting using a finite element code. Several geometrical variables and material properties have been taken as dimensionless parameters. A satisfactory convergence of the numerical results to an asymptotic limit with a wellknown exact solution, has been obtained.

Distributions of 134Cs and 137Cs in the axial UO 2 blankets of irradiated (U,Pu)O 2 fuel pins

The axial distributions of 134Cs and 137Cs in the axial blanket column of uranium-plutonium oxide fuel pins showed that the migration properties of the precursors of these two Cs isotopes were significantly different. The isotopic distributions were directly related to the half-lives of the two principal precursors of each cesium isotope, 133I and 133Xe for 134Cs, and 137I and 137Xe for 137Cs. Both cesium isotopes preferentially collected on natural UO 2 blanket pellets having high oxygen content.

Isotopic Analysis of Natural UO2Fuel Irradiated to 22000 MWd/MTU. Theory vs Experiment

The uranium and plutonium isotopic distributions of 45 irradiated fuel rods of natural uranium dioxide are compared to theoretical predictions made using three-dimensional P-1 neutron diffusion techniques. The calculations are different in that normalization to experimental results is made only by use of the total core energy output and measured critical rod-bank heights. This is in contrast to normalizing each individual fuel-rod burnup to the experimental value and then investigating resultant isotopic distributions in the rod. The comparison indicates good agreement but identifies the need for a spatial spectrum variation of the 238U epithermal resonance absorption cross section and improved time -dependence of the 238U and 239Pu cross sections.

The Reactivity of Natural UO2Irradiated to 6 × 1021n/cm2

Natural uranium dioxide specimens of Shippingport PWR-l blanket-rod geometry are exposed in the Materials Testing Reactor (flux 2 × 1014 n/cm2−sec) and discharged periodically (every three weeks) f...