Calculations of thermodynamic properties of PuO2 by the first-principles and lattice vibration

Plutonium dioxide (PuO₂) is a key compound of mixed oxide fuel (MOX fuel). To predict the thermal properties of PuO₂ at high temperature, it is important to understand the properties of MOX fuel. In this study, thermodynamic properties of PuO₂ were evaluated by coupling of first-principles and lattice dynamics calculation. Cohesive energy was estimated from first-principles calculations, and the contribution of lattice vibration to total energy was evaluated by phonon calculations. Thermodynamic properties such as volume thermal expansion, bulk modulus and specific heat of PuO₂ were investigated up to 1500 K.     

Applicability of CeO2 as a surrogate for PuO2 in a MOX fuel development

The applicability of cerium oxide, as a surrogate for plutonium oxide, was evaluated for the fabrication process of a MOX (mixed oxide) fuel pellet. Sintering behavior, pore former effect and thermal properties of the Ce–MOX were compared with those of Pu–MOX. Compacting parameters of the Pu–MOX powder were optimized by a simulation using Ce–MOX powder. Sintering behavior of Ce–MOX was very similar to that of Pu–MOX, in particular for the oxidative sintering process. The sintered density of both pellets was decreased with the same slope with an increasing DA (dicarbon amide) content. Both the Ce–MOX and Pu–MOX pellets which were fabricated by an admixing of 0.05 wt% DA and sintering in a CO₂ atmosphere had the same average grain size of 11 μm and a density of 95%T.D. The thermal conductivity of the Pu–MOX was a little higher than that of the Ce–MOX at a lower temperature but both conductivities became closer to each other above 900 K. Cerium oxide was found to be a useful surrogate to simulate the Pu behavior in the MOX fuel fabrication.     

Separation of uranium from (U, Th)O2 and (U, Pu)O2 by solid state reactions route

Solid state reactions of UO₂, ThO₂, PuO₂ and their mixed oxides (U, Th)O₂ and (U, Pu)O₂ were carried out with sodium nitrate upto 900 °C, to study the formation of various phases at different temperatures, which are amenable for easy dissolution and separation of the actinide elements in dilute acid. Products formed by reacting unsintered as well as sintered UO₂ with NaNO₃ above 500 °C were readily soluble in 2 M HNO₃, whereas ThO₂ and PuO₂ did not react with NaNO₃ to form any soluble products. Thus reactions of mixed oxides (U, Th)O₂ and (U, Pu)O₂ with NaNO₃ were carried out to study the quantitative separation of U from (U, Th)O₂ and (U, Pu)O₂. X-ray diffraction, X-ray fluorescence, thermal analysis and chemical analysis techniques were used for the characterization of the products formed during the reactions.     

A study on the burn-up characteristics of inert matrix fuels

Proper disposal of minor actinides (MA), long-lived fission products (LLFPs), and transuranium element (TRU) plays a key role in the sustainable development of fission nuclear power. Adoption of inert matrix fuels (IMFs) can effectively reduce the amount of ²³⁷Np and Np element in the spent fuel of present-day commercial power reactors. In order to study the burn-up characteristics of IMFs caused by the unique composition, burn-up calculations and MA accumulation of two typical IMFs, PuO₂ + ZrO₂ + MgO and PuO₂ + ThO₂, are performed in this paper. Results indicate that k_inf at beginning of life (BOL) and reactivity drop with burn-up for PuO₂ + ZrO₂ + MgO are much larger than those of PuO₂ + ThO₂ IMF. The yields of ²³⁷Np and Np element in IMFs are two orders smaller than those of UO₂ and mixed oxide (MOX) fuels. For the same PuO₂ volume fraction and a certain burn-up, the masses of ²³⁷Np, Np element, and ²⁴¹Am for PuO₂ + ZrO₂ + MgO are smaller than those of PuO₂ + ThO₂; however, the mass of total MA is larger. IMF has high destruction efficiencies of TRU and plutonium (Pu). The results and conclusion provide basic data for the ongoing IMF design and application study.     

Criticality investigations for the fixed bed nuclear reactor using thorium fuel mixed with plutonium or minor actinides

Prospective fuels for a new reactor type, the so called fixed bed nuclear reactor (FBNR) are investigated with respect to reactor criticality. These are ① low enriched uranium (LEU); ② weapon grade plutonium + ThO₂; ③ reactor grade plutonium + ThO₂; and ④ minor actinides in the spent fuel of light water reactors (LWRs) + ThO₂. Reactor grade plutonium and minor actinides are considered as highly radio-active and radio-toxic nuclear waste products so that one can expect that they will have negative fuel costs.The criticality calculations are conducted with SCALE5.1 using S₈–P₃ approximation in 238 neutron energy groups with 90 groups in thermal energy region. The study has shown that the reactor criticality has lower values with uranium fuel and increases passing to minor actinides, reactor grade plutonium and weapon grade plutonium.Using LEU, an enrichment grade of 9% has resulted with k_eff = 1.2744. Mixed fuel with weapon grade plutonium made of 20% PuO₂ + 80% ThO₂ yields k_eff = 1.2864. Whereas a mixed fuel with reactor grade plutonium made of 35% PuO₂ + 65% ThO₂ brings it to k_eff = 1.267. Even the very hazardous nuclear waste of LWRs, namely minor actinides turn out to be high quality nuclear fuel due to the excellent neutron economy of FBNR. A relatively high reactor criticality of k_eff = 1.2673 is achieved by 50% MAO₂ + 50% ThO₂.The hazardous actinide nuclear waste products can be transmuted and utilized as fuel in situ. A further output of the study is the possibility of using thorium as breeding material in combination with these new alternative fuels.     

Low cost, proliferation resistant, uranium–thorium dioxide fuels for light water reactors

Our objective is to develop a fuel for the existing light water reactors (LWRs) that, (a) is less expensive to fabricate than the current uranium-dioxide (UO₂) fuel; (b) allows longer refueling cycles and higher sustainable plant capacity factors; (c) is very resistant to nuclear weapon-material proliferation; (d) results in a more stable and insoluble waste form; and (e) generates less high level waste. This paper presents the results of our initial investigation of a LWR fuel consisting of mixed thorium dioxide and uranium dioxide (ThO₂–UO₂). Our calculations using the SCALE 4.4 and MOCUP code systems indicate that the mixed ThO₂–UO₂ fuel, with about 6 wt.% of the total heavy metal U-235, could be burned to 72 MW day kg⁻¹ (megawatt thermal days per kilogram) using 30 wt.% UO₂ and the balance ThO₂. The ThO₂–UO₂ cores can also be burned to about 87 MW day kg⁻¹ using 35 wt.% UO₂ and 65% ThO₂with an initial enrichment of about 7 wt.% of the total heavy metal fissile material. Economic analyses indicate that the ThO₂–UO₂ fuel will require less separative work and less total heavy metal (thorium and uranium) feedstock. At reasonable future costs for raw materials and separative work, the cost of the ThO₂–UO₂ fuel is about 9% less than uranium fuel burned to 72 MW day kg⁻¹. Because ThO₂–UO₂ fuel will operate somewhat cooler, and retain within the fuel more of the fission products, especially the gasses, ThO₂–UO₂ fuel can probably be operated successfully to higher burnups than UO₂ fuel. This will allow for longer refueling cycles and better plant capacity factors. The uranium in our calculations remained below 20 wt.% total fissile fraction throughout the cycle, making it unusable for weapons. Total plutonium production per MW day was a factor of 3.2 less in the ThO₂–UO₂ fuel than in the conventional UO₂ fuel burned to 45 MW day kg⁻¹. Pu-239 production per MW day was a factor of about 4 less in the ThO₂–UO₂ fuel than in the conventional fuel. The plutonium produced was high in Pu-238, leading to a decay heat about three times greater than that from plutonium derived from conventional fuel burned to 45 MW day kg⁻¹ and 20 times greater than weapons grade plutonium. This will make fabrication of a weapon more difficult. Spontaneous neutron production from the plutonium in the ThO₂–UO₂ fuel was about 50% greater than that from conventional fuel and ten times greater than that from weapons grade plutonium. High spontaneous neutron production drastically limits the probable yield of a crude weapon. Because ThO₂ is the highest oxide of thorium while UO₂ can be oxidized further to U₃O₈ or UO₃, ThO₂–UO₂ fuel appears to be a superior waste form if the spent fuel is to be exposed ever to air or oxygenated water. And, finally, use of higher burnup fuel will result in proportionally fewer spent fuel bundles to handle, store, ship, and permanently dispose of.     

Non destructive determination of PuO2 content in MOX fuel pins for fast reactors using Passive Gamma Scanning

Use of Passive Gamma Scanning for non destructive evaluation of PuO₂ content in mixed oxide (MOX) fuels for fast reactors is demonstrated. Experiments have been carried out on MOX fuel pins for the hybrid core of Fast Breeder Test Reactor having nominal PuO₂ content of 44% and MOX pins having nominal PuO₂ content of 21% for the Prototype Fast Breeder Reactor. A comparison of results obtained using a conventional NaI(Tl) detector and that using a through well shaped detector is also presented.     

Evaluation of high plutonia (44% PuO2) MOX as a fuel for fast breeder test reactor

Uranium plutonium mixed oxide (MOX) containing up to 30% plutonia is the conventional fuel for liquid metal cooled fast breeder reactor (LMFBR). Use of high plutonia (>30%) MOX fuel in LMFBR had been of interest but not pursued. Of late, it has regained importance for faster disposition of plutonium and also for making compact fast reactors. Some of the issues of high plutonia MOX fuels which are of concern are its chemical compatibility with liquid sodium coolant, dimensional stability and low thermal conductivity. Available literature information for MOX fuel is limited to a plutonium content of 30%. Thermodynamic assessment of mixed oxide fuels indicate that with increasing plutonia oxygen potential of the fuel increases and the fuel become more prone to chemical attack by liquid sodium coolant in case of a clad breach. In the present investigation, some of these issues of MOX fuel have been studied to evaluate this fuel for its use in fast reactor. Extensive work on the out-of-pile thermo-physical properties and fuel-coolant chemical compatibility under different simulated reactor conditions has been carried out. Results of these studies were compared with the available literature information on low plutonia MOX fuel and critically analyzed to predict in reactor behaviour of this fuel containing 44% PuO₂. The results of these out-of-pile studies have been very encouraging and helped in arriving at a suitable and achievable fuel specification for utilization of this fuel in fast breeder test reactor (FBTR). As a first step of test pin irradiation programme in FBTR, eight subassemblies of the MOX fuel are undergoing irradiation in FBTR.     

MOX Fuel Performance Experiment Specified for Japanese PWR Utilisation in the HBWR

In order to obtain high burn-up MOX fuel irradiation performance data, SBR and MIMAS MOX fuel rods with Pufissile enrichment of about 6 wt% have been irradiated in the HBWR. In-pile performance data of MOX have been obtained, and the peak burn-up of MOX pellet have reached to 66 GWd/tM as of October 2004. MOX fuel temperature is confirmed to have no significant difference compared to UO₂, if taking into account adequately for thermal conductivity degradation due to PuO₂ addition and burn-up development, and measured fuel temperature agrees well with HB-FINE code calculation up to high burn-up region. Fission gas release of MOX is possibly larger than UO₂ based on temperature and pressure assessment. No significant difference is confirmed between SBR and MIMAS MOX on FGR behaviour. MOX fuel swelling rate agrees well with solid swelling rate. Cladding elongation data shows onset of PCMI in high power region. Ramp test data from other experiment programs with various types of MOX fabrication route confirms superior PCI resistance of MOX compared to UO₂, due to enhanced creep rate of MOX. The irradiation is expected to continue until achieving of 70 GWd/tM (MOX pellet peak).     

ADAGIO technique: From UO2 fuels to MOX fuels

The amount of gas at the grain boundaries plays an important role in the fuel transient behaviour during accident conditions, such as a loss-of-coolant accident (LOCA) or a reactivity-initiated accident (RIA). Direct experimental determination of the grain boundary gas inventory has been performed for MOX fuel irradiated in an EDF pressurised water reactor (PWR) using the ADAGIO technique (ADAGIO is a French acronym meaning ‘Discriminatory Analysis of Accumulated Inter-granular and Occluded Gas’). The ADAGIO protocol applied to a MOX MIMAS fuel produced inter-granular gas fraction results that were consistent with those reached with other methods of evaluation i.e. electron probe microanalysis (EPMA). Furthermore, a new methodology for the numerical treatment of ⁸⁵Kr release kinetics which was developed for UO₂ was applied to MOX fuels. The corresponding results evidenced two types of release kinetics. These kinetics were attributed to the inter-granular bubbles of the UO₂ matrix and the bubbles located in the restructured zones, i.e. Pu agglomerates.     

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 UO₂ fuel, 1.2% enriched UO₂ fuel and for the 0.45% PuO₂-UO₂ fuel. The amount of ²³⁵U and ²³⁸U consumed and ²³⁹Pu, ²⁴⁰Pu and ²⁴¹Pu produced in the three fuel cycles are also calculated and compared.     

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

Within natural UO₂ fuel elements enriched with plutonium, this last material should form PuO₂ solid solutions inside the UO₂ pellets, in a wide range of concentrations. If the solutions are obtained by mechanical mixing of the oxides, PuO₂ islands are formed in the UO₂ 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 PuO₂ particles in the UO₂ 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.     

Fast Reactor Operating on Uranium–Plutonium Oxide Fuel in a Self-Regulatable Regime

Mathematical simulation is used to show that it is possible to develop a fast reactor operating on uranium–plutonium oxide fuel (UO₂)_1–x (PuO₂) _x , the same for all fuel elements in the core, and with uranium carbide in breeding elements with heavy coolant (PbBi eutectic). A self-regulatable regime is obtained in the reactor. This enhances safety while minimizing control. Tailings uranium with 0.1% ²³⁵U and a mixture of plutonium isotopes, which is obtained from spent fuel, making it possible to conduct operation in an actinide-closed fuel cycle, is used in the fuel and uranium carbide. ²³⁸U is actually consumed in the reactor, but most fission products are produced from ²³⁹Pu.     

A theoretical study of electronic and valence properties of U/PuO2

We consider the basic electronic and thermodynamic properties of the U/PuO₂ phase, paying special attention to the contrasts between the behaviour of this material and that of UO₂. We also report calculated ionisation potentials for Pu which play an important rôle in our analysis.     

Corrosion of irradiated MOX fuel in presence of dissolved H2

The corrosion behaviour of irradiated MOX fuel (47 GWd/tHM) has been studied in an autoclave experiment simulating repository conditions. Fuel fragments were corroded at room temperature in a 10 mM NaCl/2 mM NaHCO₃ solution in presence of dissolved H₂ for 2100 days. The results show that dissolved H₂ in concentration 1 mM and higher inhibits oxidation and dissolution of the fragments. Stable U and Pu concentrations were measured at 7 × 10⁻¹⁰ and 5 × 10⁻¹¹ M, respectively. Caesium was only released during the first two years of the experiment. The results indicate that the UO₂ matrix of a spent MOX fuel is the main contributor to the measured dissolution, while the corrosion of the high burn-up Pu-rich islands appears negligible.     

Helium behavior in oxide nuclear fuels: First principles modeling

UO₂ and (U, Pu)O₂ solid solutions (the so-called MOX) nowadays are used as commercial nuclear fuels in many countries. One of the safety issues during the storage of these fuels is related to their self-irradiation that produces and accumulates point defects and helium therein.We present density functional theory (DFT) calculations for UO₂, PuO₂ and MOX containing He atoms in octahedral interstitial positions. In particular, we calculated basic MOX properties and He incorporation energies as functions of Pu concentration within the spin-polarized, generalized gradient approximation (GGA) DFT calculations. We also included the on-site electron correlation corrections using the Hubbard model (in the framework of the so-called DFT + U approach). We found that PuO₂ remains semiconducting with He in the octahedral position while UO₂ requires a specific lattice distortion. Both materials reveal a positive energy for He incorporation, which, therefore, is an exothermic process. The He incorporation energy increases with the Pu concentration in the MOX fuel.     

Interpretation of the VENUS VIP-BWR program with APOLLO2.8 reference and optimized MOC schemes for 8 × 8 UO2 and MOX BWR assemblies

The interpretation of the VIP-BWR program conducted in the CEN·SCK Mol VENUS critical facility (Belgium), has been performed with the new APOLLO2.8 product and its CEA2005V4.1 library based on the JEFF3.1.1 file. Both reference SHEM-MOC (281groups without equivalence) and Optimized BWR 26G (26 groups with equivalence) schemes are used for UO₂ and MOX BWR assembly calculations. The VIP-BWR program was aimed to provide an experimental database for BWR neutronics tools in mixed Gd poisoned configurations with 8 × 8 UO₂ and MOX assemblies. The experimental conditions are relatively representative of actual industrial BWR core characteristics, at least in terms of void fraction. Measured pin-by-pin power distributions enable to exact valuable information at various interfaces. For fresh (UO₂/UO₂–Gd) and recycled UO₂ (UO₂ only) cores loadings, the information is given through the “UO₂” core. In the case of partial MOX loadings (UO₂/MOX interface), the power distributions are available through the “T-MOX” core. All critical sizes are predicted within 1 with SHEM-MOC reference calculation scheme. For UO₂ core, the (C–E) on k_eff are (95 ± 266) pcm and (203 ± 266) pcm for SHEM-MOC and Optimized scheme respectively. For MOX core, the results are (87 ± 214) pcm and (283 ± 214) pcm. The uncertainties take into account both measurement uncertainties and technological uncertainties such as enrichment, clad thicknesses, grid pitch or fuel densities.     

Radiotoxicity hazard of U-free PuO2+ZrO2 and PuO2+ThO2 spent fuels in LWR

The radiotoxicity hazard of U-free Rock-like oxide: ROX (PuO₂+ZrO₂) and Thorium oxide: TOX (PuO₂+ThO₂) LWR spent fuels is investigated and radiotoxicity hazard of MOX spent fuel is considered as a reference case. The long-term ingestion radiotoxicity hazard of ROX spent fuel is one third and nearly one fourth of that of TOX and MOX spent fuels, respectively. This is because the discharged Pu and long lived Np in ROX fuel is less than that of TOX and MOX fuels. In TOX fuel, discharged Pu and MA are lower than that of MOX fuel but the long-term radiotoxicity hazard of spent fuel is nearly the same as MOX spent fuel. At the cooling 10⁵ years, the radiotoxicity hazard of TOX spent fuel is approximately ten and three times higher than that of ROX and MOX spent fuels, respectively due to higher toxic contribution of ²²⁹Th in TOX spent fuel.     

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

MOX fuel pins containing both U²³³O₂ and PuO₂ 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 U²³³O₂ 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 U²³³O₂ and PuO₂ 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 UO₂ insulation pellets and MOX fuel pellets in welded MOX pins.     

Thermal diffusivity of homogeneous SBR MOX fuel with a burn-up of 35 MWd/kgHM

The effect of burn-up on the thermal conductivity of homogeneous SBR MOX fuel is investigated and compared with standard UO₂ LWR fuel. New thermal diffusivity results obtained on SBR MOX fuel with a pellet burn-up of 35 MWd/kgHM are reported. The thermal diffusivity measurements were carried out at three radial positions using a shielded “laser-flash” device and show that the thermal diffusivity increases from the pellet periphery to the centre. The fuel thermal conductivity was found to be in the same range as for UO₂ of similar burn-up. The annealing behaviour was characterized in order to identify the degradation due to the out-of-pile auto-irradiation.