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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Qualification of JEFF3.1.1 library for high conversion reactor calculations using the ERASME/R experiment

With its low CO{sub 2} production, Nuclear Energy appears to be an efficient solution to the global warming due to green-house effect. However, current LWR reactors are poor uranium users and, pending the development of Fast Neutron Reactors, alternative concepts of PWR with higher conversion ratio (HCPWR) are being studied again at CEA, first studies dating from the middle 80's. In these French designs, low moderation ratio has been performed by tightening the lattice pitch, achieving a conversion ratio of 0.8-0.9 with a MOX fuel coming from PWR UOX recycling. Theses HCPWRs are characterized by a harder neutron spectrum and the calculation uncertainties on the fundamental neutronics parameters are increased by a factor 3 regarding a standard PWR lattice, due to the major contribution of the Plutonium isotopes and of the epithermal energy range to the reaction rates. In order to reduce these uncertainties, a 3-year experimental validation program called ERASME has been performed by CEA from 1984 to 1986 in the EOLE reactor. Monte Carlo analysis of the ERASME/R experiments with the Monte Carlo code TRIPOLI4 allowed the qualification of the recommended JEFF.3.1.1 library for major neutronics parameters. K{sub eff} of the MOX under-moderated lattice is over-predicted by 440 {+-} 830 pcm (2{sigma}); the conversion ratio, indicator of the good use of uranium, is also slightly over-predicted: 2 % {+-} 4 % (2{sigma}) and the same for B4C absorber rods worth and soluble boron worth, over-predicted by 2 %, both in the 2 standard deviations range. The radial fission maps of heterogeneities (water-holes, B4C and fertile rods) are well reproduced: maximal (C-E)/E dispersion is 1.3 %, maximal power peak error is 2.7 %. The void reactivity worth is the only parameter poorly calculated with an overprediction of +12.4% {+-} 1.5%. ERASME/R analysis of MOX reactivity, void effect and spectral indexes will contribute to the reevaluation of {sup 241}Am and Plutonium isotopes nuclear data for the next library JEFF3.2. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Accident tolerant, non-oxidizable MOX fuel fabrication

Thorium dioxide (ThO{sub 2}) has a set of physical-chemical properties that make it well suited to serve as a nuclear fuel matrix: it has a very high melting point, it cannot be further oxidized, it has low leachability in water, reasonable thermal conductivity (better than UO{sub 2}), and of course it is fertile. Collectively, these properties allow for a fuel ceramic with increased in / out of core stability. As such, Thor Energy and a consortium of entities that include Westinghouse, KAERI, Fortum, NNL, ITU and IFE have recently fabricated (Th,Pu)O{sub 2} pellet specimens in Norway. The Th-MOX production methodology was developed over the course of 2 years of ceramic production trials involving Ce-Th oxide mixtures. It was understood that parameters developed for CeO{sub 2} material might need to have been modified if trial results with PuO{sub 2} led to sub-optimal pellets. Among observations made in this work, the most important production parameter was the preparation of constituent / mixed powders. Correct powder morphology was key to achieving high fuel densities. The process parameters developed for (Th,Ce)O{sub 2} pellets were adopted with the expectation that these would serve as a sound basis for obtaining dense, homogeneous (Th,Pu)O{sub 2} fuel ceramic. The Th-MOX fuel manufacture campaign was successful in producing enough (Th,Pu)O{sub 2} pellets to assemble two 30 cm long fuel pins for testing in the Halden test reactor. These fuel rods will be fitted with various instruments to enable temperature, fission gas release and dimensional changes to be measured while operating at power in conditions similar to those of a BWR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Host-guest prospects of neodymium and gadolinium ultraphosphate frameworks for nuclear waste storage: Multi-temperature topological analysis of nanoporous cages in RP{sub 5}O{sub 14}

Highlights: • The temperature-dependent cage topologies of RP{sub 5}O{sub 14} (R = Nd or Gd) were examined. • A topological analysis revealed similar cage structure for RP{sub 5}O{sub 14} (R = Nd or Gd). • One exception: the 8/4 cage topology of NdP{sub 5}O{sub 14} lacks structural cross-linking. • Volume (UO{sub 2} or PuO{sub 2}) ≫ 8/4 topology in RP{sub 5}O{sub 14} (R = Nd or Gd). • RP{sub 5}O{sub 14} (R = Nd or Gd) are good prospective host media for nuclear waste storage. - Abstract: Rare-earth ultraphosphate (RP{sub 5}O{sub 14}) framework materials are potential host media for nuclear waste storage, since cages within their nanoporous structures have volumes that match well those of prospective guests such as uranium or plutonium ions. Good volume matches of host cages and guest compounds are nonetheless not the only structural requirement for ensuring viable nuclear waste storage. Host structures also need to be stable enough to withstand the typical environmental conditions of long-term storage of spent nuclear fuel. To this end, the nanoporous cage topologies of neodymium and gadolinium ultraphosphate, NdP{sub 5}O{sub 14} and GdP{sub 5}O{sub 14}, are investigated as a function of temperature. Topological analysis shows that, while both compounds are essentially isostructural, thus displaying the same type of cage structures, the cage volumes of NdP{sub 5}O{sub 14} are significantly larger than those of GdP{sub 5}O{sub 14}, with one stark exception. This exception concerns the cage of NdP{sub 5}O{sub 14} whose 8/4 topology lacks structural cross-linking that would otherwise give it much more strength. This 'squashed' cage appears to owe its origins to the specific nature of crystallographic twinning in NdP{sub 5}O{sub 14}, which causes strain that needs alleviating; squashing the 8/4 cage in NdP{sub 5}O{sub 14} would be the most susceptible option towards this end, since its cage manifests the weakest structural construct. GdP{sub 5}O{sub 14} succumbs to a gradual contraction of its 8/4 cage only with increasing temperature above 300 K; this is well below its second-order monoclinic-to-orthorhombic phase transition at 420–430 K; which is 420 K for NdP{sub 5}O{sub 14}. Fortunately, the volumes of heavy metal ions that arise from spent nuclear fuels do not match the size requirements of cages with 8/4 topology in RP{sub 5}O{sub 14} hosts needed for encapsulation; otherwise, radiation leakage of such containment would present a risk. Both NdP{sub 5}O{sub 14} and GdP{sub 5}O{sub 14} would therefore seem to offer good prospects as host media for nuclear waste storage.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Improved PARE (Pressurized Airborne Release Equipment) for Damage Ratio, Airborne Release Fraction and Respirable Fraction Testing (Version 4)

This paper describes an improved version of the PARE (Pressurized Airborne Release Equipment) system (Sutter 1983a and Sutter 1983b and Ballinger et al 1987) for powder venting testing. This effort seeks to improve the accuracy of risk estimation for the venting of pressurized powders and the venting of gases through pressurized powders. This includes experimental measurement for estimating DR (damage ratio), ARF (airborne release fraction) and RF (respirable fraction) quantities. The present study was performed to address previously indicated problematic factors (i.e. noted gaps) (US DOE 1994).

42 ENGINEERING↗
DOE OSTI2015

DARWIN2 Low burnup validation for burnup credit using SFCOMPO PIE database - Verification of ICF values - 14433

The DARWIN package, developed by the CEA and its French partners (AREVA and EDF), provides the parameters required for fuel cycle applications, especially fuel inventory. It is widely used in the French Burnup Credit (BUC) calculation route, based on the connection of the depletion code DARWIN and the Criticality-Safety Package CRISTAL. The experimental validation process of BUC nuclides provided by DARWIN2.3 package consists in the comparison between JEFF-3.1.1-based calculated values of isotopic concentrations and experimental values obtained from the chemical analysis of PWR fuel rod cuts. As a complement to the extensive validation using the French PIE database, specific PIE data chosen for their quality were selected and analyzed in the SFCOMPO database of NEA. A set of 56 samples chosen in PWR-UO{sub 2} experiments was selected, with initial {sup 235}U enrichment between 2.7% and 4.1% and local burnup estimated from {sup 148}Nd measurement varying between 7 and 35 GWd/tHM. This paper provides an overview of calculation - experiment comparison obtained in Takahama-3 17x17, Trino-Vercellese 15x15 and Obrigheim 14x1 for uranium, plutonium isotopes, as well as available fission products contributing to the assembly reactivity. Consistent biases compared to SCALE5 results were obtained. From these results, an analysis is driven using 15 selected samples with low burnup, selected among asymptotic rods of irradiated fuels at 1 or 2 cycles. This paper shows the consistency of the calculation - experiment biases obtained with DARWIN2.3 towards previously defined penalized biases used for the Isotopic Correction Factors (ICF) in BUC applications. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Microextraction–TQ–ICP–MS for the Direct Analysis of U and Pu from Cotton Swipes

Here, the microextraction sampling technique was integrated with triple quadrupole–inductively coupled plasma–mass spectrometry (TQ–ICP–MS) to directly sample and measure the isotopic compositions of uranium (U) and plutonium (Pu) from cotton swipes. Once extracted, the U/Pu were directed into the TQ–ICP–MS instrument for isotopic determination. Carbon dioxide (CO 2 ) and helium (He) gases were delivered to a collision reaction cell within the ICP–MS system for ion separation. The CO 2 reacts with the U + forming UO + which is ultimately separated from the Pu + ions of interest in the third quadrupole. This study demonstrates direct liquid extraction of U/Pu from a solid surface and subsequent measurement by TQ–ICP–MS in <60 s. Flow rates were optimized (0.3 mL min –1 CO 2 and 5 mL min –1 He) in the reaction cell of the ICP–MS system to maximize the Pu signal while minimizing U interferences (i.e., 238 U + tail and 238 UH + ) at m/z 239. Low levels of Pu (~2 pg) were deposited on a cotton swipe along with U at concentrations ranging from 20 to 200 ng. The 240 Pu/ 239 Pu ratio was measured with <7% relative difference from the certified value at all U concentrations. Major and minor U isotope ratios were also measured with <4% relative difference. This highlights that the microextraction–TQ–ICP–MS method can extract a mixed U/Pu sample directly from a cotton swipe and measure both isotopic systems without chemical separation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2013

Americium characterization by X-ray fluorescence and absorption spectroscopy in plutonium uranium mixed oxide

Plutonium uranium mixed oxide (MOX) fuels are currently used in nuclear reactors. The actinides in these fuels need to be analyzed after irradiation for assessing their behaviour with regard to their environment and the coolant. In this work the study of the atomic structure and next-neighbour environment of Am in the (Pu,U)O₂ lattice in an irradiated (60 MW d kg⁻¹) MOX sample was performed employing micro-X-ray fluorescence (µ-XRF) and micro-X-ray absorption fine structure (µ-XAFS) spectroscopy. The chemical bonds, valences and stoichiometry of Am (~0.66 wt%) are determined from the experimental data gained for the irradiated fuel material examined in its peripheral zone (rim) of the fuel. In the irradiated sample Am builds up as Am³⁺ species within an [AmO₈]¹³⁻ coordination environment (e.g. >90%) and no (<10%) Am(IV) or (V) can be detected in the rim zone. The occurrence of americium dioxide is avoided by the redox buffering activity of the uranium dioxide matrix. - Graphical abstract: Americium LIII XAFS spectra recorded for the irradiated MOX sub-sample in the rim zone for a 300 μm×300 μm beam size area investigated over six scans of 4 h. The records remain constant during multi-scan. The analysis of the XAFS signal shows that Am is found as trivalent in the UO₂ matrix. This analytical work shall open the door of very challenging analysis (speciation of fission product and actinides) in irradiated nuclear fuels. - Highlights: • Americium was characterized by microX-ray absorption spectroscopy in irradiated MOX fuel. • The americium redox state as determined from XAS data of irradiated fuel material was Am(III). • In the sample, the Am³⁺ face an AmO₈¹³⁻coordination environment in the (Pu,U)O₂ matrix. • The americium dioxide is reduced by the uranium dioxide matrix.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2012

Application of wavelet scaling function expansion continuous-energy resonance calculation method to MOX fuel problem

More and more MOX fuels are used in all over the world in the past several decades. Compared with UO{sub 2} fuel, it contains some new features. For example, the neutron spectrum is harder and more resonance interference effects within the resonance energy range are introduced because of more resonant nuclides contained in the MOX fuel. In this paper, the wavelets scaling function expansion method is applied to study the resonance behavior of plutonium isotopes within MOX fuel. Wavelets scaling function expansion continuous-energy self-shielding method is developed recently. It has been validated and verified by comparison to Monte Carlo calculations. In this method, the continuous-energy cross-sections are utilized within resonance energy, which means that it's capable to solve problems with serious resonance interference effects without iteration calculations. Therefore, this method adapts to treat the MOX fuel resonance calculation problem natively. Furthermore, plutonium isotopes have fierce oscillations of total cross-section within thermal energy range, especially for {sup 240}Pu and {sup 242}Pu. To take thermal resonance effect of plutonium isotopes into consideration the wavelet scaling function expansion continuous-energy resonance calculation code WAVERESON is enhanced by applying the free gas scattering kernel to obtain the continuous-energy scattering source within thermal energy range (2.1 eV to 4.0 eV) contrasting against the resonance energy range in which the elastic scattering kernel is utilized. Finally, all of the calculation results of WAVERESON are compared with MCNP calculation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Dissolution of metal oxides and separation of uranium from lanthanides and actinides in supercritical carbon dioxide

This paper investigates the feasibility of extracting and separating uranium from lanthanides and other actinides by using supercritical fluid carbon dioxide (sc-CO{sub 2}) as a solvent modified with tri-n-butylphosphate (TBP) for the development of a counter current stripping technique, which would be a more efficient and environmentally benign technology for spent nuclear fuel reprocessing compared to traditional solvent extraction. Several actinides (U, Pu, and Np) and europium were extracted in sc-CO{sub 2} modified with TBP over a range of nitric acid concentrations and then the actinides were exposed to reducing and complexing agents to suppress their extractability. According to this study, uranium/europium and uranium/plutonium extraction and separation in sc-CO{sub 2} modified with TBP is successful at nitric acid concentrations of less than 6 M and at nitric acid concentrations of less than 3 M with acetohydroxamic acid or oxalic acid, respectively. A scheme for recycling uranium from spent nuclear fuel by using sc-CO{sub 2} and counter current stripping columns is presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Thermodynamic calculations of oxygen self-diffusion in mixed-oxide nuclear fuels

Mixed-oxide fuels containing uranium with thorium and/or plutonium may play an important part in future nuclear fuel cycles. There are, however, significantly less data available for these materials than conventional uranium dioxide fuel. In the present study, we employ molecular dynamics calculations to simulate the elastic properties and thermal expansivity of a range of mixed oxide compositions. These are then used to support equations of state and oxygen self-diffusion models to provide a self-consistent prediction of the behaviour of these mixed oxide fuels at arbitrary compositions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

The features of neutronic calculations for fast reactors with hybrid cores on the basis of BFS-62-3A critical assembly experiments

The different (U-Pu) fuel compositions are considered for next generation of sodium fast breeder reactors. The considerable discrepancies in axial and radial neutron spectra for hybrid reactor systems compared to the cores with UO{sub 2} fuel cause increasing uncertainty of generating the group nuclear constants in those reactor systems. The calculation results of BFS-62-3A critical assembly which is considered as full-scale model of BN-600 hybrid core with steel reflector specify quite different spectra in local areas. For those systems the MCNP 5 calculations demonstrate significant sensitivity of effective multiplication factor K{sub eff} and spectral indices to nuclear data libraries. For {sup 235}U, {sup 238}U, {sup 239}Pu the results of calculated radial fission rate distributions against the reconstructed ones are analyzed. Comparative analysis of spectral indices, neutron spectra and radial fission rate distributions are performed using the different versions of ENDF/B, JENDL-3.3, JENDL-4, JEFF-3.1.1 libraries and BROND-3 for Fe, Cr isotopes. For analyzing the fission rate sensitivity to the plutonium presence in the fuel {sup 239}Pu is substituted for {sup 235}U (enrichment 90%) in the FA areas containing the plutonium. For {sup 235}U, {sup 238}U, {sup 239}Pu radial fission rate distributions the explanation of pick values discrepancies is based on the group fission constants analyses and possible underestimation of some features at the experimental data recovery method (Westcott factors, temperature dependence). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Application of fully ceramic microencapsulated fuels in light water reactors

This study performs a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in light water reactors (LWRs). In particular, pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor (PWR). Using uranium-based fuel and Pu/Np-based fuel in TRistructural isotropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher fissile material density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design with 19.75% enrichment would need roughly 12% additional fissile material in comparison to that of a standard UO{sub 2} rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a 'color-set' array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the Pu/Np-based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, the feasibility of core designs fully loaded with Pu/Np FCM lattices was demonstrated using the NESTLE three-dimensional core simulator. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Americium Transmutation Feasibility When Used as Burnable Absorbers - 12392

The use of plutonium in Mixed Oxide (MOX) fuel in traditional Pressurized Water Reactor (PWR) assemblies leads to greater americium production which is not addressed in MOX recycling. The transuranic nuclides (TRU) contribute the most to the radiotoxicity of nuclear waste and a reduction of the TRU stockpile would greatly reduce the overall radiotoxicity of what must be managed. Am-241 is a TRU of particular concern because it is the dominant contributor of total radiotoxicity for the first 1000 years in a repository. This research explored the feasibility of transmuting Am-241 by using varying amounts in MOX rods being used in place of burnable absorbers and evaluated with respect to the impact on incineration and transmutation of transuranics in MOX fuel as well as the impact on safety. This research concludes that the addition of americium to a non-uniform fuel assembly is a viable method of transmuting Am-241, holding down excess reactivity in the core while serving as a burnable poison, as well as reducing the radiotoxicity of high level waste that must be managed. The use of Am/MOX hybrid fuel assemblies to transmute americium was researched using multiple computer codes. Am-241 was shown in this study to be able to hold down excess reactivity at the beginning of cycle and shape the power distribution in the core with assemblies of varying americium content loaded in a pattern similar to the traditional use of assemblies with varying amounts of burnable absorbers. The feasibility, safety, and utility of using americium to create an Am/MOX hybrid non-uniform core were also evaluated. The core remained critical to a burnup of 22,000 MWD/MTM. The power coefficient of reactivity as well as the temperature and power defects were sufficiently negative to provide a prompt feedback mechanism in case of a transient and prevent a power excursion, thus ensuring inherent safety and protection of the core. As shown here as well as many other studies, this non-uniform assembly type successfully addresses the concerns of reduced control rod worth within advanced MOX assemblies because the Am/MOX hybrid fuel rods are on the periphery of the assembly. The radiotoxicity of the high level waste that must be managed from a single use of an Am/MOX hybrid batch is reduced for the first 1000 years in comparison to a regular UO{sub 2} batch. However, there is no heat-load advantage without multi-recycling the Am/MOX hybrid assemblies. As shown through numerous studies, the heterogeneous assembly model used as the basis for this research is capable of multiple recycles while still maintaining negative temperature and power coefficients of reactivity. Multi-recycling would provide an even greater reduction to the radiotoxicity of the high level waste as well as provide a heat-load advantage compared to a once through UO{sub 2} fuel cycle. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Reactor physics behavior of transuranic-bearing TRISO-particle fuel in a pressurized water reactor

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) - only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is retained. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Proliferation resistance of small modular reactors fuels

In this paper the proliferation resistance of different types of Small Modular Reactors (SMRs) has been examined and classified with criteria available in the literature. In the first part of the study, the level of proliferation attractiveness of traditional low-enriched UO{sub 2} and MOX fuels to be used in SMRs based on pressurized water technology has been analyzed. On the basis of numerical simulations both cores show significant proliferation risks. Although the MOX core is less proliferation prone in comparison to the UO{sub 2} core, it still can be highly attractive for diversion or undeclared production of nuclear material. In the second part of the paper, calculations to assess the proliferation attractiveness of fuel in typical small sodium cooled fast reactor show that proliferation risks from spent fuel cannot be neglected. The core contains a highly attractive plutonium composition during the whole life cycle. Despite some aspects of the design like the sealed core that enables easy detection of unauthorized withdrawal of fissile material and enhances proliferation resistance, in case of open Non-Proliferation Treaty break-out, weapon-grade plutonium in sufficient quantities could be extracted from the reactor core.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Oxygen potential measurement and point defect chemistry of UO{sub 2}

Many studies on the oxygen potential of UO{sub 2} have been carried out so far. However, the oxygen potential data for UO{sub 2} near the stoichiometric composition in the high temperature region (1673-1873 K) are limited. Recently, Kato's group had investigated oxygen potential of (U, Pu)O{sub 2} as functions of Pu content, oxygen-to-metal (O/M) ratio and temperature. They evaluated the data based on defect chemistry. The UO{sub 2} data are fundamental to evaluate oxygen potential in UO{sub 2}-PuO{sub 2} solid solution. In this work, the oxygen potential data of UO{sub 2+x} were extended to the high temperature range of 1673-1873 K by gas equilibrium method. The measured data were analyzed based on a defect chemistry model. Kroeger-Vink's diagram were obtained in UO{sub 2+x}. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Effect of carbon impurity during the sintering of MOX fuels

UO{sub 2} and PuO{sub 2} powders are currently used as raw materials for the industrial fabrication of mixed uranium plutonium oxide (MOX) for light-water reactors. The mixture is pelletized before being sintered at 1700 Celsius degrees under a reducing atmosphere of Ar/4%H{sub 2} with 1200 vpm H{sub 2}O. At the end of the oxide ceramic sintering process, a swelling phenomenon sometimes occurs and could be responsible for the fuel non-compliance with safety specifications. This phenomenon is often caused by the presence of carbon impurities in the raw materials. In this paper, it will be shown how carbon impurities (naturally present or deliberately introduced) in the raw materials affect MOX fuel densification by sintering. For that purpose, MOX densification has been monitored by dilatometric measurements and the gases released during sintering have been continuously qualitatively and quantitatively analyzed. Gases released or trapped in the fuel during densification have been studied, as well as the sintered fuel characteristics (density, microstructure, U/Pu distribution). Results confirm that carbon impurities can play a major role in the fuel swelling: the oxides react with carbon and form gaseous CO molecules that could be trapped in the closed pores during sintering. This leads to a pressure increase within pores responsible for an irreversible swelling by creep deformation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗