I. A. Kuznetsov

Engineering methods of analyzing natural-circulation conditions in systems of BN type

The authors report calculations on modes of transition to natural circulation (NC) in systems of BN-600 type on the basis of an engineering method applied to the entire plant. A basic purpose of devising the method was to obtain reliable results on the basis of comparison with experiment. The BN-600 was used in a series of NC experiments by special methods. Initially, experiments were made with only slight core heating. The final stage was direct transition to NC after the emergency protection had operated from a power level of half the nominal value. The model describes a plant nominally having two heat-transfer branches; one has a set of loops in which an accident occurs and the other consists of the other loops, where there are known perturbations in the secondary circuit. The perturbations from the third circuit (steam-water and air one) are represented as piece-wise-linear functions.

Safety of sodium-cooled fast reactors

ConclusionsThe 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.

Calculation of the nonstationary thermoelastic stresses in a plane wall

An approximate method is proposed for calculating the nonstationary thermoelastic stresses in a plane wall for an arbitrary law of variation of fluid temperature with time.

Analyzing the maximum design fault in the core of a fast reactor

This paper addresses the problem of design faults for nuclear power stations with fast reactors, mainly the clogging of the cross section of an individual fuel element, caused by swelling of the element itself, by precipitation of foreign substances from the coolant or by penetration by foreign objects. This leads to a reduction in the flow of coolant through the element and damage, destruction or melting of the fuel element with consequential damage to its immediate surroundings. This paper examines some aspects of such a worst-case fault for reactor type BN-600. The development of the fault is observed at the stages of fuel-element overheating and the failure of its seal by observing the delayed neutrons. If the reactor has not already been shut down, the boiling of the sodium and the melting of the fuel can be registered from the neutron-flux and acoustic noise and also by the system for monitoring the reactivity balance, which initiates an alarm signal.

Studying the interaction of molten fuel with sodium in the active zone of a fast reactor

A r t i c l e s [2-6] p ropose var ious l imi ta t ions on the heat exchange, which a r e de t e rmined in the main by the heat r e s i s t a n c e of the fuel p a r t i c l e s . The approach to the p rob lem of de te rmin ing the effect of the r e s u l tant sodium vapor on the p roces s v a r i e s . It is sugges ted that the vapor envelopes the fuel fully, a f te r which the t r a n s m i s s i o n of heat f rom the fuel to the coolant can be ignored [2], or that a f i lm of sodium r e m a ins on the su r face of the fuel pa r t i c l e s f rom which it a l so evapora te s [3-5]. These models d e s c r i b e the boundary ea ses of the p roce s s of heat t r a n s m i s s i o n . A r t i c l e [6] sugges ts what appea r s to be the most r e a l i s t i c model, which takes into account the poss ib i l i ty of the fo rmat ion of a th in vapor f i lm on the su r face of the fuel p a r t i c les having finite heat r e s i s t a n c e . The fo rmat ion of such a f i lm has been conf i rmed expe r imen ta l ly [7, 8].

SOME RESULTS OF INVESTIGATIONS AND START-UP EXPERIENCE OF AN ATOMIC POWER PLANT WITH THE BN-350 REACTOR

ABSTRACT On July 16, 1973 in the town of Shevtchenko, Kazakh SSR, the power start-up of the first large-scale industrial atomic power plant with a fast reactor was carried out. The start-up of this plant sums up almost 25-year period of work in the USSR on fast breeders. The present paper gives some results of investigations as well as of operation experience during the period of start-up and adjustment works and the first months after the power start-up at power levels up to 30%. These results show that all the equipment of the reactor plant itself and of sodium circuits is efficient, reliable and easily controlled. At the same time, the period of start-up and adjustment works and the first period of operation have revealed a number of difficulties which hamper mastering of the plant.

Dynamics of nonstationary heat-transfer processes in mixing chambers

The transient processes in mixing chambers associated with perturbations of the inlet temperature of the heat-transfer agent are examined considering the heat exchange with the structural elements of the chamber.

The BN-350 and the BOR fast reactors

The reloading system and the construction of the safety control mechanisms of the BN-350 reactor are described and problems of emergency cooling andtransient processes are discussed. The experimental BOR assembly is also discussed, a description is given of the reactor and engineering equipment, and the basic technical characteristics are cited.