Development and Testing of Fuel Elements with Vibrationally Compacted Oxide Fuel for Fast Reactors

The main results of a series of scientific-research and technological studies performed at the State Science Center of the Russian Federation – Scientific-Research Institute of Nuclear Reactors to substantiate the use of fuel elements with vibrationally compacted oxide fuel in fast reactors are presented. In the course of this work, the physical-mechanical and technological characteristics of granular UO₂ and UPuO₂ fuel were studied; radiation tests and materials-engineering investigations of experimental and test fuel elements were performed in BOR-60, BN-350, and -600 reactors. More than 30,000 fuel elements were fabricated. Maximum burnup 30% heavy atoms was attained in BOR-60 using fuel assemblies with the standard construction and 32.3% heavy atoms was obtained using experimental fuel elements with a collapsible fuel assembly. In testing fuel elements with vibrationally compacted UPuO₂ in BN-600, maximum burnup of 9.6% (10.8% heavy atoms for individual fuel elements) was achieved. Postreactor investigations showed that the service life of the fuel elements is determined only by the choice of the cladding material. In accordance with the concept developed at the Ministry of Atomic Energy of Russia for the utilization of weapons plutonium, the Institute set about to implement in practice a technology for converting the metallic weapons-grade plutonium into mixed uranium–plutonium oxide fuel on the basis of pyroelectrochemistry and vibrational compaction.     

Prospective Power-Reactor Fuel Cycles Based on Water-Free Reprocessing of Spent Fuel

A pyroelectrochemical process for reprocessing spent fuel and fabricating granular oxides UO₂, PuO₂ or (U, Pu)O₂ from chloride melts has been developed at the Scientific-Research Institute of Nuclear Reactors for a prospective nuclear fuel cycle. The basic equipment has been developed. The basic results of a comprehensive study of fuel elements with vibrationally compacted (U, Pu)O₂ fuel for fast reactors are presented. The performance of the reactors remains high up to 30% burnup in standard BOR-60 reactor fuel assemblies and 32% burnup in experimental fuel elements. An assessment is made of the effectiveness of the pyroelectrochemical methods and vibrational compaction technology for plutonium utilization.     

Application of the magnetic saturation method for eddy-current inspection of spent fuel elements from fast reactors

Experimental results illustrating the effectiveness of the magnetic saturation method in eddy-current inspection of spent fuel elements with cladding made of austenitic steel (ChS-68) and ferritic-martensitic steels (éP-450, éP-823) are presented. It is shown that decreasing and stabilizing the magnetic permeability of the cladding material greatly improve the quality of nondestructive inspections of standard and experimental fuel elements of fast reactors which are currently operating (BOR-60, BN-600) or being designed (BREST OD-300). __________ Translated from Atomnaya énergiya, Vol. 102, No. 2, pp. 116–120, February, 2007.     

Post-Reactor Studies of Fuel Rods with Uranium-Plutonium Nitride Fuel with Gas and Liquid-Metal Sublayer

Atomic Energy - ETVS fuel assemblies containing nitride fuel rods with different structural implementation and cladding made of different materials are tested in the cores of the BN-600 and BOR-60...     

Investigation of 85Kr Emission and Its Effect on the Radiation Conditions during the Preparation of BOR-60 Fuel for Recovery

The main stages of the BOR-60 closed fuel cycle, implemented on the experimental base at the Scientific-Research Institute of Nuclear Reactors, are examined. The ⁸⁵Kr emission at the stages of preparation of the spent BOR-60 fuel assemblies for recovery is determined experimentally. It is shown that the maximum ⁸⁵Kr emission as a result of destruction of fuel element cladding with oxide uranium fuel is 68%; its contribution to the irradiation dose to the public as a result of mechanical disassembly of the fuel elements in a single BOR-60 fuel assembly with 10% burnup and a 10-yr holding time does not exceed 1·10^–4% of the dose limit (1 mSv/yr).     

Results of U<Subscript>0.55</Subscript>Pu<Subscript>0.45</Subscript>N and U<Subscript>0.4</Subscript>Pu<Subscript>0.6</Subscript>N mixed mononitride fuel tests in a bor-60 reactor to burnup 12% h.a.

The results of post-reactor studies of U_0.55Pu_0.45N and U_0.4Pu_0.6N mixed mononitride fuel elements (density 85% of the theoretical value) and a helium sublayer are presented. The fuel elements are irradiated in a BOR-60 reactor to burnup 9.4 and 12.1% h.a., respectively, with power density 430 and 540 W/cm. All fuel elements remained hermetic; the ChS-68 steel cladding (20% cold deformation) retained excess plasticity. The maximum zone of interaction between the cladding and the fuel and fission products did not exceed 15 μm. The swelling rate of U_0.4Pu_0.6N and U_0.55Pu_0.45N fuel was 1.1 and 0.68%/% burnup, respectively. The gas release did not exceed 19.3 and 19%. The steel damage dose was 43 dpa. The character of the porosity distribution in the fuel affects the swelling and gas release.     

Decladding of Irradiated Fuel Elements

The results of investigations of the leading operation of a nonaqueous technology for reprocessing fuel elements from nuclear reactors — dissolution of fuel claddings in a zinc-based melt — are presented. Data obtained in experiments on simulators and samples of irradiated fuel elements in standard BOR-60 and SM-2 packages with different burnup and holding time are presented. In the experiments, the metallic melt was separated from the fuel by filtering through a mesh and regenerated by vacuum distillation for reuse. The uranium and plutonium extraction was 99.99%. The behavior of individual radionuclides is described. __________ Translated from Atomnaya Energiya, Vol. 99, No. 4, pp. 273–276, October, 2005.     

Radiation Characteristics of Europium-Containing Control Rods in a SM-2 Reactor after Long-Term Operation

Data are presented on the activity of europium isotopes accumulated in the absorbing elements of the control units of propulsion reactors (7.4·10¹⁸ Bq), BN-600 and BOR-60 fast-neutron reactors (2.6·10¹⁷ Bq), and the SM-2 research reactor (4.8·10¹⁵ Bq). A universal computational procedure was developed for investigating the radiation characteristics of absorbing elements. Features of this procedure are demonstrated for the example simulating the isotopic composition of SM-2 absorbing elements. It is shown that the computed and experimental values of the parameters of the height distribution of the activity of europium isotopes agree within the limits of computational and measurement errors, and the computed total activity over the volume of an absorption element is 15% higher than the experimental value.     

Fabrication of Mixed Urianium–Plutonium Fuel on the Paket Setup and Waste Handling Methods

The results of fabrication of fuel elements with mixed uranium–plutonium oxide fuel are presented. The experimental fuel assemblies assembled from the fuel elements were tested in BN-350 and -600 reactors. Postreactor investigations of the fuel elements showed that there was no substantial difference in the behavior of the fuel cores consisting of the mixed fuel as compared with UO₂ fuel. Solid and liquid radioactive wastes are produced during the fuel fabrication process. A classification of the wastes and methods for handling them is given. It is shown that the off-grade sintered pellets should be pulverized and returned to the beginning of the mixed-fuel fabrication process.     

Study of the properties of ammonium polyuranate precipitates and UO<Subscript>2</Subscript> powders and pellets obtained with different technologies

The results of a differential-thermal analysis are used to compare the properties of ammonia polyuranate precipitates, UO₂ powders and pellets, obtained by different methods as well as metallic uranium. It is found that the phase NH₃·3UO₃·5H₂O forms in regular precipitation of ammonium polyuranate. When using nanotechnology, the phases NH₃·2UO₃·3H₂O and 4NH₃·6UO₃·8H₂O are also present in the precipitate. UO₂ powder prepared from such precipitate has high activity, since all phase transformations in it occur at a lower temperature. Modified fuel pellets of uranium dioxide, which are obtained by means of nanotechnology or mechanical addition of ammonia-containing reagents to powder, differ from the standard powders by a lower rate and more complex mechanism of oxidation, similar to metallic uranium. Modified UO₂ fuel pellets fabricated at the Physics and Power-Engineering Institute, are now undergoing tests in the BOR-60 reactor. After tests on the irradiated new modified fuel have been completed, it will be possible to judge its reliability.     

Behavior of vibrationally compacted mixed fuel in BN-800

The modified computer code MACROS was used to study the behavior of fuel elements with vibrationally compacted mixed fuel under irradiation in BN-800. Verification calculations were performed for fuel elements irradiated in BN-600 under close to nominal conditions. Attention was focused mainly on the factors playing an important role in their serviceability: swelling, creep, corrosion behavior of cladding, temperature distribution, gas-release, and degree of structural reformation of the fuel core.     

A microstructure-dependent model for fission product gas release and swelling in UO2 fuel

A model for the release of fission gas from irradiated UO₂ fuel is presented. It incorporates the relevant physical processes: fission gas diffusion, bubble and grain boundary movement, intergranular bubble formation and interlinkage. In addition, the model allows estimates of the extent of structural change and fuel swelling. In the latter, contributions of thermal expansion, densification, solid fission products, and gas bubbles are considered. When included in the ELESIM fuel performance code, the model yields predictions which are in good agreement with data from UO₂ fuel elements irradiated over a range of water-cooled reactor conditions: linear power outputs between 40 and 120 kW m⁻¹, burnups between 10 and 300 MW h(kg U)⁻¹, and power histories including constant, high-to-low and low-to-high power periods.The predictions of the model are shown to be most sensitive to fuel power (temperature), the choice of diffusion coefficient for fission gas in UO₂, and burnup. The predictions are less sensitive to variables such as fuel restraint, initial grain size and the rate of grain growth.     

Tests of Models of BREST-OD-300 Reactor Fuel Elements in an Autonomous Lead-Cooled Channel of a BOR-60 Reactor

A special feature of the BREST-OD-300 reactor that is now being designed is that it employs a container-type heat-conducting fuel element with mixed uranium–plutonium mononitride fuel, a lead sublayer, and an expansion volume at the top to collect gaseous products. The fuel elements are arranged in a square array with a wide spacing and are spaced by laminated spacing lattices.The substantiation of the technical solutions adopted for the construction of the reactor fuel elements and fuel assemblies, specifically, the combined effect of the coolant and heat loads on the fuel-element cladding and the spacing lattices, led to the choice of the BOR-60 sodium-cooled fast research reactor as an experimental base and required the development and construction of an autonomous lead-cooled channel loaded into a cell through a passage in the rotatable plugs of the reactor. The channel was tested for two microruns with the BOR-60 reactor operating at 45 MW. The lead temperature at the fuel assembly entrance was 595°C, the working temperature of the cladding was 658°C, the damaging dose was 6.5 displacements/atom, and the fuel burnup was 0.44% h.a. Analysis of the activity of the gas and the lead showed that the fuel elements are sealed. Post-reactor studies have been conducted since August 2002.     

Methods for protecting fuel-element cladding from corrosion in lead coolant and the heat-transfer sublayer

The results of investigations of the corrosion of commercial and experimental steels in lead and the possibilities of corrosion protection are presented. The effect of lead coolant and the lead heat-transfer sublayer on fuel-element cladding are examined. Methods based on thermodynamic calculations and experimental data are proposed for protecting fuel element cladding in a lead-cooled reactor from the corrosive effect of the coolant by creating a new corrosion resistant chromium steel and from the corrosive effect of the heat-transfer sublayer by alloying with the components of steel. The results of this work have been implemented in the experimental fuel elements for the BREST-OD-300 reactor which were irradiated in a BOR-60 reactor. __________ Translated from Atomnaya énergiya, Vol. 104, No. 2, pp. 88–94, February, 2008.     

Recovery of Uranium and Plutonium from Spent Fuel Elements of Nuclear Reactors

The results of investigations of the preliminary removal of the products of radioactive decomposition from irradiated nuclear fuel to obtain uranium and plutonium which are suitable for reuse in fuel fabrication are presented. Nitrate-alkali melts are used for the operation. The experiments are performed on simulators and irradiated samples of BOR-60 fuel in remote-controlled hot boxes. The coefficients of removal of fission products are presented. A technological scheme, which will shorten the fuel cycle, for purifying hot nuclear fuel is recommended. __________ Translated from Atomnaya Energiya, Vol. 99, No. 5, pp. 387–392, November, 2005.     

Prospects for Using Mixed Oxide Fuel in Fast Sodium-Cooled Reactors

When designing new fast reactors, it is desirable to increase as much as possible the breeding occurring in the core in order to ensure the minimum excess reactivity for burnup on the one hand and a closed fuel cycle without replenishment with external plutonium and without separating plutonium from uranium during chemical reprocessing of irradiated fuel on the other. The latter requirement greatly decreases the risk of plutonium proliferation in such a fuel cycle. This requires a core breeding ratio 1.05–1.08. Such values can be achieved by using technologically perfected and tested oxide fuel with its volume fraction in the core increased to 55–60%. The results of computational-theoretical studies on the selection and optimization of cores with high fuel fractions for BN-1600 and BN-800 reactors are presented in this article. It is shown that such cores can be built in principle.     

Radiation Conditions during the Preparation of Irradiated BOR-60 Fuel Assemblies for Reprocessing

The basic stages in the preparation of irradiated BOR-60 reactor fuel for reprocessing are examined. It is determined that during the separation of the fuel part of the fuel elements the coefficient of transfer of ¹³⁷Cs from the fuel into aerosol is 5·10^–6 and for fragmentation the value is 3·10^–5. It is found that the real catching efficiency for aerosol particles caught by a V-05 filter ranges from 42 to 99%. The specific entry of radioactive aerosols into the ventillation center after the first stage of air purification was 0.3 MBq for -emitters and 7.7 MBq for and emitters per 1 kg of reprocessed fuel. The total collective dose formed at the stages of preparation of a large batch of irradiated fuel (four spent fuel asemblies with average burnup 11.4% and a 10.5 to 23.7 yr holding period) for reprocessing was 11.5·10^–3 persons·Sv.     

Production and investigation of thorium-containing fuel compositions

A technology has been developed for obtaining fuel tablets with the compositions (U, Th)O₂, (U, Th, Ca)O₂, and (U, Th)O₂+MgO by combined precipitation of uranium, thorium, magnesium, or calcium components from inert solutions, followed by heat treatment of the powders, compression into pellets, and sintering of the pellets. Work on optimizing the technological processes for obtaining fuel pellets so as to obtain good pellet quality was performed. The effect of the properties of the precipitates and powders, fabricated using different technological regimes on the properties of the finished objects was studied. The work includes detailed investigations of powders (x-ray phase analysis, electron-microscopic investigation) and sintered fuel tablets (change in the geometric dimensions as a result of sintering, determination of the density, and study of the microstructure). The behavior of fuel compositions (U, Th)O₂ and (U, Th)O₂+MgO in contact, with coolants under conditions where the fuel elements become unsealed was studied: with water at 300°C and sodium at 700°C. 3 figures, 3 tables, 6 references. State Science Center of the Russian Federation-A. I. Leipunskii Physics and Power-Engineering Institute. Translated from Atomnaya énergiya, Vol. 88, No. 5, pp. 346–353, May, 2000.     

Investigation of the cause of peculiar irradiation behavior of 9Cr-ODS steel in BOR-60 irradiation tests

Four experimental fuel assemblies (EFAs) containing 9Cr-ODS steel cladding fuel pins were previously irradiated in the BOR-60 to demonstrate the in-reactor performance of 9Cr-ODS steel for use as fuel cladding tubes. One of the EFAs achieved the best data, a peak burn-up of 11.9at% and a neutron dose of 51 dpa, without any microstructure instability or any fuel pin rupture. On the other hand, in another EFA (peak burn-up, 10.5at%; peak neutron dose, 44 dpa), peculiar irradiation behaviors, such as microstructure instability and fuel pin rupture, occurred. Investigations of the cause of these peculiar irradiation behaviors were carried out. The detection sensitivity in an ultrasonic inspection test was shown to be low for the metallic Cr and metallic Fe inclusions. The peculiar microstructure change reappeared with high-temperature thermal-aging of the 9Cr-ODS steel containing metallic Cr inclusions. The strength and ductility of the defective part containing metallic Cr inclusions were appreciably lower than those of a standard part without the inclusions. The combined effects of matrix Cr heterogeneity (presence of metallic Cr inclusions) and high-temperature irradiation were concluded to be the main cause of the peculiar microstructure change in 9Cr-ODS steel cladding tubes in the BOR-60 irradiation tests. They contributed to the fuel pin rupture.     

Thermal-fluid characteristics of the transuranics fuel in a deep-burn HTR core

The Deep Burn Project is evaluating the feasibility of the DB-HTR (Deep Burn High Temperature Reactor) to achieve a very high utilization of transuranics (TRU) derived from the recycle of LWR spent fuel. This study intends to evaluate the thermal-fluid and safety characteristics of TRU fuel in a DB-HTR core using GAMMA+ code. The key design characteristics of the DB-HTR core are more fuel rings (five fuel-rings), less central reflectors (three rings) and decay power curves due to the TRU fuel compositions that are different from the UO₂ fuel. This study considered three types of TRU kernel compositions such as 100%(PuO₂ + NpO₂ + Am), 99.8%(PuO_1.8, NpO₂) + 0.2%UO₂ + 0.6 mole SiC getter, and 70%(PuO_1.8, NpO₂) + 30%UO₂ + 0.6 mole SiC getter. The first fuel type of TRU kernel produces higher decay power than the UO₂ kernel. For the second and the third fuel types, removing the initial Am isotopes and reducing the volumetric packing fraction of TRISO particles will reduce the decay power. The flow distribution, core temperature and TRISO temperature profiles at the steady state were examined. As a safety performance, this study mainly evaluated the peak fuel temperature during LPCC (low pressure conduction cooling) event with considering the impact of decay power, the annealing effect of the irradiated thermal conductivity of graphite, and the impact of the FB (fuel block) end-flux-peaking. For the 600 MW_th DB-HTR core, the peak fuel temperature of 100%(PuO₂ + NpO₂ + Am) TRU was found to be much higher than the transient fuel design limit of 1600 °C due to the lack of heat absorber volume in the central reflector as well as to the increased decay power of the TRU fuel compositions. For a 0.2%UO₂ mixed or a 30%UO₂ mixed TRU, the peak fuel temperature was decreased due to the reduced decay power, however, it was still higher than 1600 °C due to the lack of heat absorber volume in the central reflector.