Steam-superheating fuel elements of the reactors in the I. V. Kurchatov Beloyarsk nuclear power station

Samoilov, A. G.; Pozdnyakova, A. V.; Volkov, V. S.

doi:10.1007/bf01118888

Cited by 5 publications

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“…The prototype of the second type of channel structure is a nominal steam-superheating channel used in the reactors in the first phase of the Beloyarskaya nuclear power plant [9]. This structure was found to be reliable for prolonged opera- tion with temperatures close to those of CR-SCP.…”

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Direct-Flow Channel Reactor with Supercritical Coolant Pressure

et al. 2005

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The concept of a direct-flow channel reactor with supercritical-pressure water (CR-SCP) is presented. Neutron-physics, thermohydraulic, and strength calculations are used to substantiate the fundamental core design with a heavy-metal moderator which at supercritical pressure is competitive with other modern reactor designs with respect to fuel-cycle indicators. Two types of fuel-element and fuel-channel structures are examined. It is shown that fuel elements based on micropellets and a metal matrix are highly reliable and have higher operating characteristics (burnup, service life, geometric stability, and so on) than fuel elements with uranium-dioxide fuel. A CR-SCP design and the technological scheme of a power-generating unit are presented, and the systems which are required to ensure normal operation and safety are determined. The main technical-economic indicators of a power-generating unit with installed electric power 850 MW are estimated.The interest in supercritical coolant pressure in reactors is due to the fact the new reactors must be competitive. In the last few years a surge of interest has been observed in Japan, USA, Germany, France, Canada, and other countries [1]. Reactors with supercritical-pressure water are being considered as one direction in the international program for the development of fourth-generation reactors.Some countries are undertaking efforts to develop direct-flow water-moderated water-cooled vessel reactors with supercritical-coolant pressure. The core of these reactors can be designed to operate on thermal and fast neutrons with almost the same thermal layouts of the power-generating units and the same efficiency. It is important to note that a fast-neutron reactor makes it possible, aside from appreciably decreasing the capital costs, to improve fuel utilization [2], since it makes it possible to increase the breeding ratio of nuclear fuel up to 1 and aim at a closed fuel cycle. However, such reactors are characterized by a change in water density by a factor of 10 from the core entrance to the core exit. This is accompanied by a change in the neutron spectrum over the height of the core and makes it difficult to smooth and stabilize the energy-release fields not only

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Direct-Flow Channel Reactor with Supercritical Coolant Pressure

et al. 2005

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“…The characteristics of pressurized pipes made of the alloy Zr-2.5%Nb and δ hydride zirconium at 150°C are as follows: σ 0.2 = 630 MPa, ƒ T = 0.62, K 1H = 6 MPa·m 1/2 [15,16], K α 1C = 40 MPa·m 1/2 , E = 87290 MPa [2,3], µ = 0.35 [10,17], σ f = 637 MPa [18], and σ * or = 150 MPa [19]. A calculation using expressions (9), (12), and (13) gives the maximum critical hydride length l max = 22.5 µm and thickness t = 1.5 µm.…”

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Critical length and thickness of hydride plates with delayed hydride cracking in zirconium alloys

Shmakov

2004

At Energy

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A method for calculating the critical length and thickness of hydrides with delayed cracking in commercial zirconium alloys is developed. The critical characteristics of hydride precipitates are estimated for pipes made of the alloy Zr-2.5%Nb. It is shown that the theoretical results agree with the experimental data.Delayed hydride cracking, which is one mechanism for fracture of hydrogenated alloys of zirconium, niobium, titanium, and vanadium, is based on diffusion accumulation of hydrogen along a concentrator of tensile stresses, formation of hydride plates oriented perpendicular to the applied load, growth of the plates up to a critical size, and fracture of a region embrittled by hydrides. Cyclic repetition of these processes results in gradual fracture of the object. The criterion for destruction is the relation K 1 ≥ K 1H , where K 1 is the stress intensity coefficient and K 1H is the threshold stress intensity coefficient for delayed hydride cracking (Fig. 1). It is believed [1] that on section 1 the cracking velocity is sensitive to the stress intensity coefficient, while on section 2 it is almost constant up to values K 1C characterizing the strain crack toughness (cracking resistance) of a material.The first cases of fracturing of zirconium objects by the mechanism of delayed hydride cracking occurred in the mid-1970s and were due to rupture of pressurized pipes made of the alloy Zr-2.5%Nb in CANDU reactors (Canada). Theoretical models for predicting the basic characteristics of the process -the cracking velocity and the threshold coefficients K 1H -have been developed only over the last 10 yr in connection with the importance of the operational safety of power reactors [2-10]. A general drawback of the well-known models is that they neglect the dependence of the critical parameters of hydride precipitates on the temperature and stress intensity coefficient. This assumption limits the predictive power of the models and makes the theoretical results less reliable. In the present paper a method is proposed for estimating the critical length and thickness of hydride plates for delayed hydride cracking in commercial zirconium alloys.The distribution of tensile stresses σ y along the line continuing a sharp crack with normal break-away (type I) is shown in Fig. 2. The size of the plastic flow zone of the material r p and the stress level σ p = σ y (r p ) are determined for planar deformation from the well-known expressions

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