Fuel Element Cans Research Articles

Characteristics of radiation porosity formed upon irradiation in a BN-600 reactor in the fuel-element cans of cold-deformed steel EK-164 (06Kh16N20M2G2BTFR)-ID c.d.

At present, it is the austenitic cold-deformed steel EK164 (06Kh16N20M2G2BTFR)-ID that is considered as a promising material for the achievement of a maximum damage (no less than 110 dpa) and maximum burnup (≥15%). In this work, we have determined the characteristics of porosity formed upon irradiation in a BN-600 reactor to the maximum damaging dose of 77 dpa in the materials of fuel-element cans made of cold-deformed steel EK164-ID c.d. A comparison has been made with analogous characteristics obtained earlier using the standard material, i.e., the cold-deformed steel ChS68 (06Kh16N 15M2G2TFR)-ID c.d.

Dynamics and wear of fragments of a lower part of a fuel assembly with free fuel elements

A procedure and findings of investigations of the dynamic characteristics and the wear of fuel elements in the framework of the development of a perspective design of fuel elements without fastening in a lower support grid are presented. At the first phase of the work, natural vibration frequencies and the dynamic rigidity of fuel elements are determined in various design variants of lower spans. Resonance frequencies of spans of fuel elements are revealed by minimums of the dynamic rigidity. At the second phase, investigations of vibrational wear within contact of fuel elements with cells of spacer grids are performed on models of a lower part of a fuel assembly. Four embodiments of the lower part of the fuel assembly are investigated. As a result of the application of high vibrational loads, the wear of fuel element cans and grid cells is observed in more than a half of contact assembles of fuel elements with spacer grids. The most and least wear-resistant designs of the lower part of the fuel assembly are determined by results of tests.

Structural steels for cores of fast neutron reactors

The directions of development of stainless steels for fuel rod claddings and fuel element cans of fast neutron reactors are considered. The principles of complex alloying and treatment of steels ensuring low swelling at irradiation dozses of 80–90 dpa (dislocations per atom) are described. Ways for further development of functional characteristics of steels of austenitic and martensitic classes are outlined.

Tritium migration in nuclear desalination plants

Tritium transport, as one of important items of radiation safety assessment, should be taken into consideration before construction of a Nuclear Desalination Plant (NDP). The influence of tritium internal exposition to the human body is very dangerous because of 3H associations with water molecules. The problem of tritium in nuclear engineering is connected to its high penetration ability (through fuel element cans and other construction materials of a reactor), with the difficulty of extracting tritium from process liquids and gases. Sources of tritium generation in NDP are: nuclear fuel, boron in control rods, and deuterium in heat carrier. Tritium passes easily through the walls of a reactor vessel, intermediate heat exchangers, steam generators and other technological equipment, through the walls of heat carrier pipelines. The release of tritium and its transport could be assessed, using mathematical models, based on the assumption that steady state equilibrium has been attained between the sources of tritium, produced water and release to the environment. Analysis of the model shows the tritium concentration dependence in potable water on design features of NDP. The calculations obtained and analysis results for NDP with BN-350 reactor give good convergence. According to the available data, tritium concentration in potable water is less than the statutory maximum concentration limit. The design of a NDP requires elaboration of technical solutions, capable of minimising the release of tritium to potable water produced.

Theoretical and experimental thermal-hydraulic analysis of a new sodium cooled fast breeder reactor core

This paper presents the experimental and theoretical results of the thermal-hydraulic design of a new fast breeder reactor core concept. The main feature of this concept is the omission of fuel element cans. The hydraulic function of these fuel element cans is substituted by a winding flow path through the radial blanket and a ring chamber without tubes. A computer code based on the quasi-continuum-theory and especially adapted to the features of the new core concept is developed for theoretical investigations. The pressure drop of the rod bundles is specified by a resistance tensor. The experimental investigations are realized in a test facility, where sodium is simulated by water. Pressures and velocities are measured. Theoretical and experimental results show good agreement. The aim of flattening of the coolant outlet temperature distribution can be reached with satisfying accuracy.

Study of the radiation-induced thermal decomposition of cesium iodide and the corrosion cracking of water-moderated-water-cooled power-reactor fuel-element cans under gamma radiation

The authors studied the effect of temperature (20-380/sup 0/C) and pressure (0.2-15.5 MPa) on the ..gamma.. radiation of CsI in helium-filled Zr-1% Nb alloy cans. It was shown that the decomposition of CsI attains its maximum velocity at a temperature of 380/sup 0/C and a helium pressure of 15.5 MPa (G = 6.8 +/- 1.0 molecules per 100 eV), while lowering the temperature or the pressure diminishes the decomposition velocity of the CsI. The action of ..gamma.. radiation on CsI leads to the failure of helium-filled zirconium cans through corrosion.

Crack resistance of cold-worked 09Kh16N15M3B steel sheets

During the operation of nuclear reactors, microcracks can form in the material of the fuel-element cans because of the mechanical action of the fuel and the chemical interaction of the fission products with the can material. The efficiency of the fuel elements is determined by the crack resistance of the material which depends on the size of the existing defects, the magnitude of the load, and the properties of the material. The authors studied the room temperature crack resistance of the 09Kh16N15M3B stainless steel in sheet form and its modifications used for producing fuel-element cans. The observed ductile fracture of the test specimens of both the steels with significant plastic deformation requires the application of the methods of nonlinear fracture mechanics. The strength properties as well as the crack resistance of the 09Kh16N15M3B steel and its modification increase with increasing cold deformation up to 20%.

Operating experience with the VV�R-1000 reactor fuel-element assemblies of the fifth unit of the Novovoronezh nuclear power station

The initial operation of the fifth unit was characterized by an increased number of transitional regimes, the number of which did not exceed the design values. No effect of the considerable number of power variations on the working efficiency of the fuel elements was observed. An analysis of the primary circuit coolant activity with respect to fission products shows that operation of the fifth unit is characterized by a low degree of depressurization of the fuel element cans. The number of fuel elements with microdefects is significantly lower than the limiting value, which confirms the high reliability of the fuel elements. After attaining nominal power level, no marked depressurization of the fuel elements develops.

Effect of contact-layer flaws on serviceability of fuel-element cans

Effect of contact-layer flaws on serviceability of fuel-element cans

Methods of measuring the distortion of fuel-element cans during irradiation

Methods of measuring the distortion of fuel-element cans during irradiation

Yield of gaseous fission products from defective fuel elements in the BOR-60 reactor

1. In the BOR-60 reactor the formation of defects in the fuel-element cans is accompanied by the escape of radioactive gaseous fission products. The extent of this escape for a burnup of over 5% equals the proportionate yield from the fuel, and amounts to (5.3±1.0)% for133Xe. The yield of the remaining gaseous fission products is approximately proportional to the square root of the halflife, in agreement with the laws of the diffusion model describing the escape of gases from under the can. The yield of solid fission products directly after the formation of a defect is negligible. 2. Owing to the low yield of gaseous fission products at the instant of defect formation and their considerably lower yield on subsequent irradiation, the radioactivity in the gas system of the reactor does not limit the permissible number of defective fuel elements in the active zone. In developing fuel elements which allow the escape of gaseous fission products into the coolant, it is essential to ensure the retention of the solid fission products (131I,137Cs,95Zr−95Nb,140Ba−140La); the period of holding the gases under the can may be made very short. 3. The can hermeticity-monitoring system based on delayed neutrons is not capable of detecting the appearance of defects in the fuel elements, but its readings may serve as an indication of any substantial opening of the can. Constant γ-spectrometric monitoring of the gas radioactivity enables us to estimate the number of defective fuel elements in the active zone of the reactor at any moment of time.

Irradiation-induced embrittlement of magnox cans

Abstract The creep ductility of Magnox AL80 (Mg/0.8 wt per cent A1/0.005 wt per cent Be) fuel element cans is known to decrease with increasing irradiation. A possible cause of this embrittlement is void nucleation at inert gas bubbles formed on grain boundaries. The necessary size and spacing of these bubbles are calculated together with the total number of gas atoms which they must contain. The numbers of inert gas atoms arising from the following processes are also calculated: (a) the lithium (n, α) reaction, (b) the injection of fission fragments into the can, (c) fission of plutonium which has diffused into the can. Each mechanism can produce more than the theoretical minimum quantity of gas to account for the observed embrittlement. However, the appropriate grain boundary distribution of gas is unlikely to arise from (b) and (c) and the quantity from (a) exceeds the minimum only slightly, even if all the gas is in bubble nuclei. The need for continued control over the lithium content of the canning ...

Sintered 13% Cr steel for fast reactor fuel element cans

Sintered 13% Cr steel for fast reactor fuel element cans

Sintered 13% Cr steel for fast reactor fuel element cans

A study was made of the effects of various metallurgical factors on the structure and strength and creep characteristics of sintered 13% Cr ferritic steel. The beneficial effect of alloying with molybdenum on these properties is noted and discussed, and the influence exerted by various plastic working conditions is examined. Data yielded by determination of various properties of sintered ferritic steels show that the changes observed are attributable to differences in internal oxidation, diffusion, and carbide formation processes, which play an important part in the evolution and variation of properties during the production of the steels and prolonged testing under load. While internal oxidation and diffusion are of importance from the viewpoint of stability of properties in the initial condition, carbide formation is responsible for embrittlement during creep.

Monitoring gas-tightness of fuel element cans in gas cooled reactors

Monitoring gas-tightness of fuel element cans in gas cooled reactors

Study of the development of leakages in fuel element cans by means of Kr88

Study of the development of leakages in fuel element cans by means of Kr88

Measurement of the rate of diffusion of plutonium in irradiated Mg/0.5 wt % Zr fuel element cans by controlled abrasion and alpha-counting

Previous work has shown that plutonium can permeate the matrix of a Mg/0.5 wt% Zr (Magnox ZR55) fuel element can irradiated in contact with uranium and that the concentrations produced at an irradiation level of 0.17 at% burn-up range from ≈ 1×10 −3 at% near the can inner surface to ≈ 1 × 10 −5 at% near the can outer surface. By alpha-counting either material removed by controlled abrasion or the abraded surface itself, it has now been established that this solute plutonium segregates to a free surface by diffusion from the adjacent matrix. The amount accumulating at a surface as a result of a given annealing treatment is proportional to the initial concentration of plutonium in the underlying layers. A study of the rates of accumulation of plutonium at a surface at different temperatures indicated that the activation energy of the diffusion process is 31 kcal/mole. Measurement of the concentration gradients developed in the matrix, adjacent to a surface at which plutonium was segregating, enabled values of the diffusion coefficient D to be obtained for temperatures of 400 °C and 450 °C. The value of D 0 derived from the experimental results is in the range 0.095–0.225 cm 2/s (mean value 0.15 cm 2/s).

Precipitation of plutonium compounds in irradiated magnesium alloy fuel element cans

Thin foils which contain the inner surface of irradiated fuel element cans have been cut with an ultramicrotome. transmission electron microscopy revealed numerous fine precipitates within 25 μm of the inner surface of irradiated AL80 (Mg/0.8 wt% Al) cans but very few in irradiated ZR55 (Mg/0.5 wt% Zr) cans. Alpha-counting of foils showed that plutonium had penetrated into the can wall to differing extents in the two materials; it was confined to the region containing the precipitate in an AL80 can irradiated at 405 °C but had diffused about 1 mm into a ZR55 can irradiated at 300 °C for half the time. some of the precipitate particles in AL80 have been identified as PuAl 3 by electron diffraction. It is concluded that the precipitation of plutonium-aluminium compounds at the inner surface prevents plutonium diffusing through the can wall of AL80 clad fuel elements during irradiation.

Plutonium and fission product distribution in some irradiated experimental magnox fuel element cans

An experimental study has been made of the distributions of plutonium and fission products across the wall thickness of some magnesium alloy cans after irradiation at ≈ 400 °C for 20–37 months in contact with uranium fuel. Some cans were experimental, of Magnox ZR55 (Mg/0.5 wt% Zr); some were standard, of Magnox AL80 (Mg/0.8 wt% Al). Plutonium was determined by methods involving controlled abrasion and alpha-counting; fission product activities were measured by photometry of contact autoradiographs. The ZR55 cans contained > 100 ppm Pu near the can inner surface and < 4 ppm Pu near the outer surface; plutonium had also segregated on the outer surface to the extent of 0.27 μg/cm 2 in the sample examined. In the AL80 cans, irradiated at a similar temperature for twice the time, plutonium was confined to within 30–150 μm of the inner surface and over 60% of it was probably on the inner surface itself or in the oxide layer. Fission products had penetrated in substantial quantities to the outer surface of the ZR55 cans, with some segregation at the surface itself.

Experimental study on the vibrations of various fuel rod models in parallel flow

The parallel flow vibrations in different types of fuel element cans were studied by means of tests in air and in water. The vibration amplitudes of an isolated rod are represented in a correlation incorporating the main geometric and hydrodynamic parameters. The additional effect of flow asymmetry and of some technological parameters (pumps, grids, play in the attachments) is demonstrated. The experimental results are analyzed and compared with those of other experiments. The tests described are part of a technological study on the risks of wear and hammering entailed by the vibrations of fuel elements in an ORGEL-type channel.