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.     

Fuel for advanced sodium-cooled fast reactors: current status and plans

Different fuel compositions have been irradiated in the BR-10, BOR-60, BN-350, and BN-600 reactors and investigated: PuO₂, UC, UN, UPuN, UO₂, UPuO₂, metallic doped and undoped alloys, and fuel compositions with inert matrices. Studies of fuel elements with UPuN as well as with fuel compositions based in MgO and ZrN irradiated in the BOR-60 reactor were completed in 2009. The main results of the investigations of different fuel compositions are presented and the problems of ensuring the serviceability of BN-800 and -1200 fuel elements are discussed.     

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.     

Core and fuel cycle for an advanced sodium-cooled fast reactor

The development of BN-1200 is based on the greatest possible use of tested and scientifically validated and developed technical solutions implemented in BN-350, -600, and the BN-800 design as well as new technical solutions that increase facility cost-effectiveness and safety. The BN-1200 design must permit the reactor to operate with different cores, including with denser fuel. The main fuel variant considered is oxide fuel and for the nearest term nitride fuel, for which the production technology involves the same steps as the oxide technology. The main approaches for choosing the parameters of the BN-1200 core as well as the results of computational studies are presented.     

Safety properties of bn-1200 with mixed oxide and nitride fuel: a comparison

Along with mixed oxide fuel, the possibility of using in BN-1200 dense nitride fuel, making it possible to attain higher technical-economic performance, is also studied. However, safety analysis will determine the choice of fuel type. In this connection, it is important to perform a comparative analysis of the inherent safety properties for variants of the BN-1200 core with mixed uranium-plutonium and nitride fuel for the most serious unanticipated loss-of-power accident with failure of all emergency protection organs of the reactor simultaneously. A two-dimensional version of the COREMELT computer code was used in the calculations. The computational analysis showed that the inherent safety of BN-1200 is much greater with nitride than with mixed uranium-plutonium fuel.     

Power effect of reactivity in fast power reactor with allowance for behavior of fuel under irradiation

Conclusions The proposed computational approaches for determining the thermal conditions of the fuel elements with allowance for their operating history have made it possible to bring closer together the calculated and experimental values of the power effect of reactivity and its dependence on the power level in the BN-350.Analysis of the causes of the divergence of the experimental and calculated values of the power effect of reactivity showed that these divergences, as well as changes in the effect during operation, are due mainly to the thermal conditions of the fuel elements. The principal indeterminacy in the calculation of the thermal conditions of the fuel elements is due to the contact thermal conductivity and its variation with the operating conditions of the fuel elements in the reactor. In order to eliminate this indeterminacy it is necessary to carry out experimental investigations to ascertain the contact thermal conductivity as well as to refine the physicomechanical properties of the fuel (thermal conductivity, ultimate strength etc.) as a function of the burn-up and temperature.Translated from Atomnaya Énergiya, Vol. 47, No. 3, pp. 157–161, September, 1979.     

Layout and structural solutions for the reloading system of an advanced fast reactor

The particulars of fuel reloading in fast reactors are described. General data on a BN-1200 power-generating unit, the principles of the construction of its reloading system, the purpose and construction of individual types of reloading equipment, and required experimental work are presented. The comparative characteristics of the BN-800 and BN-1200 reloading equipment are presented.     

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.     

Demonstration Experiment of 3 BN-600 MOX Vibropac FAs Irradiation for the Excess Weapons Plutonium Disposal

The paper presents results of a demonstration experiment on conversion of 50 kg of weapon-grade plutonium in the form of metal ingots into granulated MOX-fuel to be used for manufacturing fuel pins and 3 fuel assemblies (FAs) for the fast power-generating reactor BN-600, irradiation parameters of these FAs and the data from post-irradiation examinations. It can be concluded from the PIE results that the 3FAs were successfully irradiated in BN-600 without any fuel pin failures. Therefore, disposition of weapongrade plutonium with a weight of about 20 kg was successfully done. This represents the first disposition of Russian surplus weapon-grade plutonium as an international cooperation (this experiment was performed in collaboration between RIAR and JNC). The possibility of using MOX vipac fuel as a method for weapon plutonium disposition is clearly shown.     

Cobalt content and vacancy swelling of ChS-68 steel

Analysis of the state of fuel-element claddings before and after operation in a fast reactor revealed a statistically significant factor correlating with the vacancy swelling of 06Kh16N15M2G2TFR (ChS-68) austenitic steel. This factor, which depends on the quantitative composition of the alloying and impurity elements in steel melts, explains the reason for the large (more than a factor of 2) differences in the size changes of spent fuel elements in BN-600 reactor fuel assemblies and makes it possible to predict the swelling resistance of different batches of casing tubes made of ChS-68 steel. __________ Translated from Atomnaya énergiya, Vol. 102, No. 3, pp. 163–168, March, 2007.     

Optimization of the core composition for a sodium-cooled fast reactor with oxide fuel and fuel-utilization improvement by adopting a closed fuel cycle

The transition to a closed fuel cycle after several years of operation of the BN-800 with oxide uranium fuel in an open fuel cycle is examined. It is shown that there is an advantage to using new fuel assemblies with 91 fuel elements with diameter 8.6 mm in a regime with four refuelings. On the basis of new fuel assemblies with mixed uranium-plutonium oxide fuel, transitional recycling to a closed fuel cycle without separating uranium and plutonium and without external plutonium makeup is examined. It is confirmed that a negative sodium void effect of reactivity is achieved with admissible values of the linear power density of a fuel element. It is shown that a regime with four refuelings can be obtained by adding uranium with enrichment no higher than 15% to replace the poison which is removed. __________ Translated from Atomnaya énergiya, Vol. 104, No. 2, pp. 94–99, February, 2008.     

高温气冷堆抗氧化燃料元件氧化行为分析

通过复杂化学平衡计算,对高温气冷堆抗氧化燃料元件在正常运行和极端事故工况下的氧化行为进行了理论分析。结果表明高温气冷堆在出现大量空气或者水进入堆芯的极端事故工况下,抗氧化燃料元件能够起抗氧化保护作用,保持燃料元件的完整性。然而在正常运行工况下,抗氧化燃料元件可能会发生活化氧化,抗氧化涂层的完整性可能会被破坏。通过计算提出适当提高He中CO浓度来防止抗氧化燃料元件被活化氧化。     

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.     

BN-800 as a New Stage in the Development of Fast Sodium-Cooled Reactors

The role of fast reactors in a strategy for developing nuclear power in Russia because of the inevitable exhaustion of natural uranium deposits in the foreseeable future is discussed. The BN-800 reactor, which is under construction and incorporates unique solutions – greatly enhancing the safety of the reactor – to technical and constructional problems, is examined. Cost assessments taking account of the complete life cycle show that fast reactors could be no more expensive than the most widely reactors in the world – water-moderated water-cooled reactors.Closing the BN-800 nuclear fuel cycle will make it possible to solve the problem of utilizing plutonium and actinides. This makes fast reactors safer for the environment.     

Monitoring the distribution of energy release in BN-600 by γ scanning fuel assemblies

A complex of computational and experimental measures for monitoring the distribution of energy release in the core has been perfected over the 25-year history of BN-600 operation in the Beloyarskaya nuclear power plant. Continual monitoring is conducted together with calculations based on three-dimensional multigroup calculations and periodic γ scanning of regular BN-600 fuel assemblies with upgrading of the core of this reactor. At the present time, in the course of switching BN-600 to a new core with four refuelings and maximum fuel burnup increased to 11.1% h.a., the experimental procedure has been upgraded and optimized taking account of the experience gained, three series of such measurements have been completed, and new experimental data on the character of the radial and axial neutron-field distribution have been obtained. Translated from Atomnaya énergiya, Vol. 105, No. 6, pp. 339–344, December, 2008.     

Change in shape and size of fuel assembly duct in BN-350 reactor

The paper presents analytical data obtained by directly measuring samples taken at different levels of a hexagonal fuel assembly duct used in the BN-350 fast reactor. The difference between the length of the deflection arc and the width of an intact duct face displays the accumulated irradiation-creep deformation during the operation period. The deflection of the duct faces induced by the irradiation creep was analyzed, and irradiation-creep rate versus distance from the core center was analyzed for the assembly duct in the BN-350 fast reactor.     

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...     

Neutronic and safety aspects of inert matrix fuel utilization in fast reactors for plutonium and minor actinides transmutation

The solving of ecological problems of future nuclear power is connected with the solving of long-lived radioactive waste utilization problems. It concerns primarily plutonium and minor actinides (MAs), accumulated in the spent fuel of nuclear reactors. One of the ways this can be solved is to use a fast reactor with uranium-free or inert matrix fuel (IMF). The physics of this type of reactor was widely investigated during last year for BN-800 reactors. The solution of the most important problems was: a decrease in non-uniformity of power distribution and an increase of the Doppler effect. The next stage of such core investigations is an evaluation of self-protection to beyond design accidents. Preliminary results show a high safety level of BN-800 reactors with IMF in the event of unprotected loss of coolant flow (ULOF) accident.     

Plutonium use in nuclear power in Russia

Conclusions The following concept of plutonium utilization based on the evolutionary development of the traditional technology in our country arises: The main problem of any short-term program of dealing with plutonium must be solved — reliable and safe storage of separated energy plutonium and freed weapons plutonium before utilization in reactors. Plutonium (mainly energy plutonium) is utilized primarily in BN-800 fast reactors and the development of technology using weapons plutonium in BN-600 reactors starts. In the future attention should be focused on nuclear-power centers patterned after the Industrial Association “Mayak” (RT-1 plant, Complex 300, BN-800) with reliable nonproliferation of weapons plutonium. It is extremely important to speed up work on the completion of Complex 300: This work must be completed before BN-800 is ready. In the future efforts must be concentrated on the following: development and implementation, in BN-800, of an economically more efficient plutonium-burning core; the possibility of building light-water reactors with the required degree of safety for effective plutonium utilization must be justified (including a “cold” core based on cermet fuel); and, development and implementation of technology for a safe and an ecologically acceptable closed nuclear fuel cycle based on plutonium and²³³U with burnout of Am, Np, and Cm. Ministry of Atomic Energy of the Russian Federation. Institute of Physics and Power Engineering. A. A. Bochvar All-Union Scientific-Research Institute of Standardization in Machine Building. Special Design Office for Machines. Translated from Atomnaya énergiya, Vol. 76, No. 4, pp. 326–332, April, 1994.     

Intrinsic Safety of Future BN-800 Based Nuclear Technology

More than 40 years of experience in performing research and development work and operation of fast sodium-cooled reactors is analyzed. It is shown that such reactors possess a system of intrinsic safety properties, making possible long-time reliable operation and reducing to a minimum the consequences of an accident.A BN-800 unit under construction with the core switched to nitride fuel can serve as a basis for nuclear technology with intrinsic safety in accordance with the requirements of the strategy for the development of nuclear power in Russia in the first half of the 21st century.