The next generation of fast reactors

The results of computational and design studies of a 1200 MW, lead-cooled pool-type fast reactor with U---Pu nitride fuel based on the same principles as the previously considered BREST-300 design (Adamov, E.O., Orlov, V.V., Filin et al. Proc. Int. Topical Meeting on Advanced Reactors Safety, ARS'94, Pittsburgh, USA, 1994, pp. 509–516.) are presented. In connection with a capacity increase and to ensure full implementation of the LCFR concept merits in the BREST-1200 design, a number of new solutions have been accepted compared with the more conservative initial design.     

Fluorex reprocessing system for the thermal reactors cycle and future thermal/fast reactors (coexistence) cycle

A new reprocessing technology, FLUOREX was proposed for thermal reactors cycle and future thermal/fast reactors (coexistence) cycle. The proposed system is a hybrid system that combines fluoride volatility and solvent extraction methods. Spent fuel will be sheared and cladding material will be removed by dry oxidation/reduction method such as AIROX process. Fluorination and purification of most uranium can be easily achieved by fluoride volatility method with compact facility. About 10% residues including plutonium can be treated in well-established PUREX method, which means this facility load will be about 1/10 of the conventional PUREX facility with same capacity. Between fluorination process and PUREX process, there is a pyrohydrolysis process where the fluoride compounds from fluorination process are converted to the oxides. Pure mixture of Pu and U can be obtained by solvent extraction method without separating Pu and U, which is suitable for conventional MOX fuel fabrication. The system can recover pure U and MOX with the decontamination factor of over 10⁷ and can drastically reduce the cost and waste generation compared with the conventional one.

Semi engineering scale experiments for the fluorination, pyrohydrolysis, and dissolution of Pu containing materials were carried out. From those experimental results, key elemental processes were fundamentally proofed.     

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.     

Electromagnetic flowmeters for fast reactors

This paper is in four parts. Section 1 explains the theory of the induced-voltage electromagnetic flowmeter and then considers various types which have been used. For the primary circuit of fast reactors both flow-through type and probe type have been proposed, although obtaining magnets which operate satisfactorily at high temperatures has been a problem. In the secondary circuit the high magnetic Reynolds numbers cause the field to be swept out of the magnet gap and this has led to the use of the long saddle-coil flowmeter.In Section 2 flux-distortion flowmeters are described. These have been proposed mainly for monitoring the primary circuit flow and again both flow-through and probe types have been tested. Sections 3 and 4 continue the discussion of the flux-distortion flowmeter by introducing two methods of analysing its performance. The first is a finite difference method which solves the non-linear problem by using a time marching method. It is shown that a linear approximation is adequate for the likely levels of flow encountered in the fast reactor and consequently two linearised solutions are used. The first method is a finite difference one and allows the instantaneous response of a step change in velocity to be observed as well as the effect of bubbles.In Part 4 the second linearized method uses current rings to divide up the conducting material. By considering the interaction of all the rings, it is possible to obtain the current distribution and hence the magnetic field. In conclusion it is suggested that further development would be useful of the devices which are most suited to the liquid metal fast breeder reactor.     

快中子堆启堆试验

以大、中型池型试范快堆为例,介绍有关快堆启堆的两个主要试验阶段的经验:(1)装料前预备工作、充钠、堆整体试验;(2)装料、逼近临界及提升功率试验。     

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.     

89Sr production in fast reactors

Conclusions In summary, we have proposed a new method for producing⁸⁹Sr for medical purposes from natural yttrium according to the reaction (n, p) in fast-neutron reactors. Investigations confirm the computational parameters of the production: from 2 to 15 mCi⁸⁹Sr per gram of the starting yttrium. We have shown that⁹⁰Sr can be extracted and the final product with the required radionuclide purity can be obtained. Commercial production of⁸⁹Sr in BR-10 and BOR-60 has now started. Translated from Atomnaya énergiya, Vol. 82, No. 5, 396–399, May, 1997.