Safety of sodium-cooled fast reactors

Conclusions The accumulated experience in the operation of NPP, including those with fast reactors, shows that during normal operation, with due regard for possible operational difficulties and accidents, they ensure a significantly lower level of risk for personnel and the surrounding population than is present in industrial regions and those prone to natural disasters. Therefore, the dangers connected with the widespread development of nuclear power arise not so much from a real risk as from a risk which in principle can be realized in very improbable accidents. From this point of view sodium-cooled fast reactors have certain advantages. The probability of the maximum accident of the rupture of pipelines in high-pressure reactors must be considerably higher. Here a single event, and one difficult to detect, such as the failure to detect a flaw in manufacture, is enough to initiate the very dangerous first step of an accident. The rupture of equipment in the primary loop of a fast reactor at practically atmospheric pressure is considerably less probable, and the integral assembly is quite safe. All the other chains of development of maximum accidents in a fast reactor require the simultaneous realization of several events in systems and devices which are constantly being monitored (SS and power supply systems, etc.). The above considerations together with such important properties of sodium as the large reserve before the boiling point and the practically inertialess transport of heat from the reactor to structural elements and heat-transfer devices under natural circulation conditions gives one confidence that the level of risk for future industrial NPP with fast reactors will be at least no higher than that for NPP with thermal reactors.Translated from Atomnaya Énergiya, Vol. 43, No. 6, pp. 464–472, December, 1977.     

Particulars of neutron-physical calculations of sodium-cooled fast reactors with mixed oxide fuel

The neutron-physical characteristics of reactor systems with a fast spectrum, sodium coolant, and uraniumplutonium fuel load have been analyzed on the basis of computational studies of the BFS-62-3A critical assembly and a BN-600 hybrid core with mixed oxide fuel. The large differences in the spectra in an expanded thermal range to 1 keV for the central and peripheral regions with uranium oxide and mixed oxide fuel show that spatially differentiated fission and absorption cross sections must be used for the main uranium and plutonium isotopes in neutron-physical calculations.     

Unanticipated accidents in sodium-cooled fast reactors

The definitions and requirements of normative documents for unanticipated accidents at nuclear power plants with fast reactors are analyzed. Definitions are constructed between one another and with a collection of scenarios which can lead to unanticipated accidents, likewise determined by normative documents independently of the probability of these accidents actually happening. It is concluded that the normative approaches to fast-reactor safety must be refined with respect to strengthening the probabilistic criteria as a tool limiting the list of required unanticipated accidents for validating reactor safety. Special attention is devoted to the need to strengthen the motivation of designers to make the maximum possible use of passively triggered safety systems.     

Inservice inspection of the reactor block of sodium-cooled fast breeder reactors

By using sodium as coolant special boundary conditions result for the inservice inspection (ISI) of fast breeder reactors. For that reason in general it is not successful applying the methods and equipment proved for the 151 of light water reactors.This report presents inspection methods and equipment developed for the ISI of the reactor block of sodium cooled fast breeder reactors. The survey takes into account the state of the art as well as some R&D-work at home and abroad. Entering into particulars the methods and equipment used for leak monitoring, the inspection of the reactor vessel wall, the inspection' of reactor internals above and below the sodium level, monitoring of structure home noise and the measurement of the gap between the reactor vessel and the guard vessel are described.     

Escape of fission products from defective fuel rods of light water reactors

The escape behaviour of various fission product isotopes from defective fuel rods in PWRs and BWRs is analyzed.Diffusion in the UO₂ is the rate controlling step for the release of noble gases from defective fuel rods. The escape of fission iodine from defective fuel rods is controlled by a mechanism which includes migration and additional delay steps, probably in the nature of a chemical reaction.The inferred effective diffusion constants for fission gases are noticeably higher for defective fuel rods than for intact fuel rods. The difference is about two orders of magnitude. The enhancement of diffusion in defective fuel rods is believed to be due to the increase in the -ratio of the UO₂ in the defective fuel rods.     

Modeling the diffusion yield of radioactive fission products from uranium dioxide fuel

The numerical aspects of a systematic solution of the problem of diffusion and yield of radioactive fission products from a homogeneous sphere, simulating a uranium dioxide fuel kernel, with realistic boundary conditions are discussed. The numerical scheme is based on a one-group method of calculating the production and radioactive mutual transmutations of fission products in combination with the standard expansion of the radial dependence of the concentration in terms of the eigenfunctions of the Laplace operator. It is demonstrated on illustrative examples that the approach is highly economical.     

Acoustic resonances as a means of blockage detection in sodium-cooled fast reactors

This paper outlines the basis of a system giving early warning of the presence and build up of a local blockage in sodium-cooled fast reactor sub-assemblies. The method uses acoustic resonances and reports on the analysis of signals recorded during special experiments on the Dounreay Fast Reactor. These tests proved the detectability of the standing waves and showed theoretical predictions of their frequency and shift in the presence of a blockage to be in good agreement with measurements.     

Statistical error in calculating the nuclide composition of low-burnup fuel

The error arising in the change of the ²³⁵U and ²³⁹Pu concentrations as a result of the statistical error in the microscopic cross sections during a computational fuel-run simulation with the MCU and MCNP programs is investigated. The analysis is limited to the thermal neutron spectrum and low fuel burnup. A simplified model simulating a fuel-run calculation using MCU and MCNP type statistical programs is constructed. This model is used to analyze for a commercial uranium-graphite reactor the effect of the rate of recalculation of and the statistical error in the microscopic cross sections over a run on the calculation of the ²³⁵U and ²³⁹Pu concentrations. The results show that the influence of the statistical error on the computed ²³⁵U and ²³⁹Pu concentration is negligible even with 10⁵ neutron histories in the statistical computational sample over a run.__________Translated from Atomnaya Énergiya, Vol. 98, No. 2, pp. 91–97, February, 2005.     

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.     

History of the engineering and scientific-technical contribution of BR-5 and BR-10 reactors to the development of fast sodium-cooled reactors

The first experimental sodium-cooled reactor BR-5 in Europe was built and put into operation in a record short time – three years (1956–1959). The main goal of building such a reactor was to master the use of sodium coolant and sodium equipment under radiation-hazardous operating conditions. It is shown that the reactor made it possible to solve other problems also. The first sodium apparatus, systems for monitoring and purifying the coolant, and methods and facilities which were later used in the development of fast power reactors were mastered on BR-5. The components of successful multiyear (43 years) operation and the results of investigations are presented. The article focuses primarily on the experiential aspects which were later used to develop and operate subsequent fast reactors. Translated from Atomnaya énergiya, Vol. 106, No. 3, pp. 134–140, March, 2009.