Rods In Nuclear Reactors Research Articles

E215 水圧駆動を用いた沸騰水型原子炉の制御棒挿入に関する研究(OS8 軽水路・新型炉・原子力安全(5))

This thesis reports on the hydraulic driving system, powered by an accumulator. This drive system is mainly used for the drive of control rods of nuclear reactors. In case of strong earthquakes, control rods are set in gaps between fuel assemblies to scram nuclear reactors Characteristics of the system have not been analyzed. The analysis of this system is necessary in order to present the designs that are intended to be a variety of situations. So we developed the model of the hydraulic control rod driving system The model that we have created is able to reproduce the actual driving. Also, there is a load on the system by an earthquake. This load is caused by the contact of the deformed fuel assembly and control rod. This load model is obtained by solving the equation of motion of the beam.

Tensile Behavior of Cryorolled Zircaloy-2

Zircaloy-2 is mainly used in nuclear technology, as cladding of fuel rods in nuclear reactors, especially water reactors (BWRs). Hence high strength of Zircaloy-2 is of prime importance. This investigation deals with the effect of cryorolling on Zircaloy-2 by comparing different tensile properties. For this analysis, four samples with various degrees of cryorolling are taken and tensile tests are conducted on these samples. The obtained results are analyzed and the optimum degree of cryorolling of Zircaloy-2 is obtained. The cryorolling improved the mechanical properties of the material as the dislocations are entangled near the grain boundaries and also due to decrease in the grain size. The microstructure of the sample is analysed by optical microscope, before and after cryorolling and the grain structure analysis is done.

지르코늄의 제조(製造)와 재활용기술(再活用技術)

Zirconium is one of the most important material used as cladding of fuel rods in nuclear reactors because of its high dimensional stability, good corrosion resistance and especially low neutron-absorbing cross section. However, Hf free nuclear grade Zr sponge is commercially produced by only three countries including USA, France and Russia. So, Zr has been thoroughly managed as a national strategic material in Korea. Most of the zirconium is used for Korean nuclear industry as nuclear fuel cladding materials manufactured from Hf free Zr alloy raw material. Also, there are some other applications such as alloying element and detonator. In this review, zirconium production and recycling technologies have been reviewed and current industrial status was also analyzed. And recent achievements in innovative reduction technologies such as electrolytic reduction process and molten oxide electrolysis were also introduced.

A Hafnium Material Control Rod Design and Manufacture

Because the metallic hafnium has good combination properties, that makes it the preferred material for control rods of nuclear reactors. To prevent the irradiation deformation and the mechanical property decrease of hafnium plates, and to increase the service life of the control rods, a new type of square-tube control rod is developed. The elemental composition, mechanical property, the corrosion resistance and the physical characteristics of hafnium are researched. Several tests are carried out after the product manufactured, which prove that the design and manufacture of the control rod meet the requirement of the technical conditions.

The Properties Research of Hafnium Plate for Control Rod

Metallic hafnium has good comprehensive performance that makes it the preferred material for control rods of nuclear reactors. The elemental composition, mechanical properties, the corrosion resistance and the physical characteristics of hafnium plate are researched during the development of control rod for reactors. It proved the manufacturing method of hafnium plate is appropriate, characteristic of materials meet the requirement of a control rod.

Design, construction and simulation of a multipurpose system for precision movement of control rods in nuclear reactors

This article presents the design and implementation of a microcontroller-based system for the automatic movement of control rods in nuclear reactors of either power or research types. This system is controlled automatically, is linked to a personal computer system, and has manual controlling ability as well. The important features of this system are: automatic scram of the control rods, activation of alarm in emergency situations, and the ability to tune the control rod movement course both upwards and downwards. In this system, a small tank has been improvised as a coolant reservoir for pool type reactors such as Tehran Research Reactor and its water level is continuously adjusted by special sensors. Also, this system can be applied for controlling various types of control rods such as the regulating rods, safety rods and shim rods; can be connected to all reactor measurement tools and systems such as the period meter, power meter and flux meter; and can receive feedback signals from them. The devised system can be calibrated with these measurement tools by two special potentiometers in the related electronic board. The processes of this system have been simulated by the SIMULINK tool kit of MATLAB software and all responses of the system, including oscillation and transient responses, have been analyzed.

Thermal expansion and heat capacity of dysprosium hafnate

Dysprosium hafnate is a candidate material for as control rods in nuclear reactor because dysprosium (Dy) and hafnium (Hf) have very high absorption cross-sections for neutrons. Dysprosium hafnate (Dy2O3·2HfO2-fluorite phase solid solution) was prepared by solid-state as well as wet chemical routes. The fluorite phase of the compound was characterized by using X-ray diffraction (XRD). Thermal expansion characteristics were studied using high temperature X-ray diffraction (HTXRD) in the temperature range 298–1973 K. Heat capacity measurements of dysprosium hafnate were carried out using differential scanning calorimetry (DSC) in the temperature range 298–800 K. The room temperature lattice parameter and the coefficient of thermal expansion are 0.5194 nm and 7.69 × 10−6 K−1, respectively. The heat capacity value at 298 K is 232 J mol−1 K−1.

Analysis of Pellet-Clad Mechanical Interaction Process of High-Burnup PWR Fuel Rods by RANNS Code in Reactivity-Initiated Accident Conditions

Exploratory analyses were performed by the RANNS code for simulated tests of the reactivity-initiated accident with two high-burnup pressurized water reactor rods in the Nuclear Safety Research Reactor (NSRR). The code performs thermal and finite element mechanical calculations in an axis-symmetrical cylinder geometry. On the basis of the irradiation-induced rod conditions including bonding, the code analyzed a strong pellet-clad mechanical interaction process that would often lead to low-strain split failure. The predicted quantities such as temperature and stress strain of cladding were discussed and compared with the experimental observations. The calculated cladding permanent strain has a reasonable agreement with postirradiation examination data. The process from crack initiation to final split failure was accounted for by the plastic strain occurrence in the cladding.

High Temperature Thermal Insulation System for Millimeter Wave Sintering of B4C

Boron carbide (B4C) is one of advanced materials and is being used in a wide rage of applications. The unique feature of this material is its large neutron-absorbing cross-section. Some of its most prominent applications are controlling rods in nuclear reactors and radiation protection. 24 GHz microwave processing for B4C ceramics were performed under flowing argon gas using the sintering system. Sintering at the high temperature (up to 2200°C) was achieved using thermal insulation system consists of fiber-board, boron nitride powder, and boron nitride case. The sintered samples were achieved 90 % of theoretical.

Characterization of fuel oxidation in rods with clad-holes

Three irradiated reactor rods that had holes in the cladding were examined for oxidation. The techniques used were coulometric titration and optical examination. Coulometric titration has the capability to measure the O/M ratio for small (∼100 mg) samples. Both the longitudinal and radial directions of the rods were examined. Two of the rods had little oxidation. The third rod was oxidized everywhere, but especially at the clad-holes where the O/M values reached a maximum of 2.10. The difference between the three rods is examined. Qualitatively, optical examination revealed similar trends. For the third rod, oxidation proceeded more rapidly along the grain boundaries than within the grains themselves. Also, grain growth and gas bubble formation occurred, which had not been expected because the power levels were too low for that to occur. It seems that the temperatures were increased because of lower thermal conductivities in the higher oxide phases.

An implicit enthalpy formulation for oxidation of nuclear reactor fuel cladding by steam

The oxidation of the zircaloy cladding of fuel rods in nuclear reactors, important during a severe accident, is governed by the diffusion of oxygen in solid zirconium. The latter exists in many phases which depend on the content of oxygen and have different diffusion coefficients. Discontinuous changes in oxygen concentrations occur at the phase boundaries, which also move in the course of diffusion, obeying an equation similar to the Stefan condition in heat transfer problems with phase change. We have devised and implemented a scheme for 1-D problems using the finite-difference approach, which involves setting the diffusion coefficient to zero in ‘forbidden’ concentration ranges. The method is shown to be mathematically equivalent to the enthalpy formulation for the Stefan problem. With a boundary condition at the vapour/solid interface which takes into account steam diffusion and dissociation, the code is applied to analyse the phenomena of steam starvation and the dissolution of the oxide layer, which are relevant to severe accidents. Copyright © 2000 John Wiley & Sons, Ltd.

Transmission computer microtomography for nondestructive monitoring of fuel and control rods in a reactor

A method of x-ray transmission computer microtomography has been developed for solving the problems of monitoring the quality of fuel elements and control rods in nuclear reactors: the geometric resolution of a defect is several microns. The solution of the problem of nondestructive monitoring of such objects has made it possible to perform a wide range of investigations of the technical characteristics of definite x-ray detectors, organize principles of scanning, based on the principles of laser interferometry and the design of data processing systems, that are different from those of conventional systems. The investigations performed have made it possible to implement computer-aided design for problem-oriented computer microtomographs and for the instrumentational implementation of an experimental variant of such a device. Investigations performed on specially fabricated test samples with calibrated defects have demonstrated that the approach to the design of such setups for nondestructive monitoring of objects for nuclear power generation is correct and that this is a promising direction, 8 figures, 9 references.

Irradiation Assisted Mechanical Cracking on Cladding Tubes of Control Rods in Nuclear Reactors.

The tips of the rodlets of rod cluster control assemblies(RCCAs) are exposed to extremely high neutron fluence in fuel assemblies. A high neutron fluence causes degradation of the control rod materials. In a pressurized water reactor(PWR) plant, increases in the outer diameters of the rodlets, induced by absorber swelling, and intergranular cracks at tips of rodlets were observed. First, mechanisms of cracking by material degradation due to neutron irradiation in nuclear reactors were given their definitions because they were previously not clearly defined. Then it was confirmed that the cause of intergranular cracks of RCCA rodlets was irradiation assisted mechanical cracking(IAMC). The cracks were induced by an increase in hoop strain due to the swelling of the absorber and a decrease in elongation due to neutron irradiation. Furthermore, investigations were carried out on the swelling mandrel test, because the structures and environments of the swelling mandrel test were similar to those of RCCA rodlets. As a result, it was found that the mechanism of the swelling mandrel test was not irradiation assisted stress corrosion cracking(IASCC) but IAMC. Also, intergranular cracks on the cladding tubes of RCCA rodlets could be induced at a very low strain rate.

Evaluation of Fretting Damage of Zircaloy Cladding Tubes

The work is part of the search for a technique to study the fretting properties of the cladding of fuel rods in nuclear reactors. The strategy is to compare the fretting scars of specimens tested in a near-full-scale, nonradioactive simulator (so called fuel assembly endurance test) with scars on specimens tested under well controlled conditions in the laboratory. By systematic variations of the laboratory testing parameters it has proved possible to achieve scars with the same type of damage characteristics as those from the fuel assembly endurance test. This makes it possible to estimate, by deduction, the parameters that cause fretting damage under actual service conditions in a reactor.

TRANSURANUS: a fuel rod analysis code ready for use

TRANSURANUS is a computer program for the thermal and mechanical analysis of fuel rods in nuclear reactors and was developed at the European Institute for Transuranium Elements (TUI). The TRANSURANUS code consists of a clearly defined mechanical-mathematical framework into which physical models can easily be incorporated. Besides its flexibility for different fuel rod designs the TRANSURANUS code can deal with very different situations, as given for instance in an experiment, under normal, off-normal and accident conditions. The time scale of the problems to be treated may range from milliseconds to years. The code has a comprehensive material data bank for oxide, mixed oxide, carbide and nitride fuels, Zircaloy and steel claddings and different coolants. During its development great effort was spent on obtaining an extremely flexible tool which is easy to handle, exhibiting very fast running times. The total development effort is approximately 40 man-years. In recent years the interest to use this code grew and the code is in use in several organisations, both research and private industry. The code is now available to all interested parties. The paper outlines the main features and capabilities of the TRANSURANUS code, its validation and treats also some practical aspects.

Prediction of the lifetime of the elements of the safety and control rods of nuclear reactors

The authors construct a mathematical model based on the analytical solution to such parameters as magnetic flux, alignment, sliding friction, erosion, abrasion, corrosion, and irradiation for the purpose of predicting the service life of electromagnetically driven control rods and their drives. The analytical solution was verified experimentally on a bench simulation where it was found that incorrect assembly and alignment not only serve as the largest contributors to shortened service life of the rods and drives but also render the calculations of the model invalid.

Accelerated High-Temperature Tests with Spent PWR and BWR Fuel Rods Under Dry Storage Conditions

Accelerated high-temperature tests on 25 intact pressurized water and boiling water reactor rods were conducted for more than 16 months at 400, 430, and 450°C in a helium gas atmosphere. The pretes...

Behavior of Breached Light Water Reactor Spent Fuel Rods in Air and Inert Atmospheres at 229°C

A whole-rod test was conducted at 229 °C to investigate the long-term stability of light water reactor spent fuel rods with artificially induced defects stored in inert and unlimited air atmospheres. Both boiling water reactor (BWR) and pressurized water reactor (PWR) rods were tested. After 2235 h, visual observations, diametral measurements, and radiographic smears were used to assess the degree of cladding deformation and particulate release. The same examinations plus metallography and x-ray analysis were conducted after 5962 h. Neither of the breached rods tested in inert atmosphere, nor the breached PWR rod tested in unlimited air, showed any measurable change from the pretest condition. The upper defect on the BWR rod tested in unlimited air had a 12.7-mm split after 2235 h and had 10% cladding deformation. The crack grew to 63.5 mm after 5962 h. X-ray analysis indicated that the UO2 had oxidized to U3O8 at this defect. The difference in the behavior of the upper and lower defects is attributed to the accessibility of the air to the fuel due to the positioning of the defect with relation to the pellet/pellet interface.

High Temperature Postirradiation Materials Performance of Spent Pressurized Water Reactor Fuel Rods under Dry Storage Conditions

Postirradiation studies on failure mechanisms of well-characterized pressurized water reactor rods were conducted for up to a year at 482, 510, and 571°C in limited air and inert gas atmospheres. N...

Assay of Light Water Reactor Rods for Their Uranium-235 Content

The assay of a light water reactor rod for its fissile uranium content can be accomplished by irradiating the rod in a water bath with neutrons from an antimony-beryllium source. The different slowing down properties of the moderator for neutrons of different energy cause a much stronger attenuation of the source neutron flux than of the flux originating from the induced fissions. At 25 cm from the source, the fission neutrons already contribute 20% to the signal measured by ordinary thermal-neutron detectors, allowing determination of their intensity as a measure of the fissile content of the rod. Laboratory experiments demonstrated that a 30-mCi source is sufficient to attain 1% precision in 1 min.