Numerical Solution to the Contact Interaction Problem of Nuclear Fuel Element Components Using the Mortar Method and the Domain Decomposition Method

The paper presents algorithms for solving axisymmetric contact interaction problems for several thermoelastic bodies using unmatched meshes. We employed the finite element method to obtain numerical solutions to problems of thermal conductivity and the theory of elasticity. We took contact interaction into account by applying the mortar method and the method of domain decomposition. The mortar method requires solving an ill-conditioned system of linear algebraic equations with a zero block at the main diagonal. To solve it numerically, we used a modified method of successive over-relaxation (MSSOR), which makes it possible to reduce solving the system of equations for all contacting bodies to sequentially solving systems of equations for each body separately. We showcase our algorithm results by solving an example problem simulating thermomechanical processes in a nuclear fuel element. We analyse the features of the stress-strain state in the structure and compare the results obtained using the mortar method and the domain decomposition method. The computational domain in the problem considered consisted of 10 nuclear fuel pellets and a cladding section. The analysis results showed that the quantitative stress-strain state properties in a system of bodies obtained by the two methods are quite close to each other. This confirms the fact that these algorithms may be correctly applied to solving similar problems

Application and Development Progress of Cr-Based Surface Coatings in Nuclear Fuel Element: I. Selection, Preparation, and Characteristics of Coating Materials

To cope with the shortcomings of nuclear fuel design exposed during the Fukushima Nuclear Accident, researchers around the world have been directing their studies towards accident-tolerant fuel (ATF), which can improve the safety of fuel elements. Among the several ATF cladding concepts, surface coatings comprise the most promising strategy to be specifically applied in engineering applications in a short period. This review presents a comprehensive introduction to the latest progress in the development of Cr-based surface coatings based on zirconium alloys. Part I of the review is a retrospective look at the application status of zirconium alloy cladding, as well as the development of ATF cladding. Following this, the review focuses on the selection process of ATF coating materials, along with the advantages and disadvantages of the current mainstream preparation methods of Cr-based coatings worldwide. Finally, the characteristics of the coatings obtained through each method are summarized according to some conventional performance evaluations or investigations of the claddings. Overall, this review can help assist readers in getting a thorough understanding of the selection principle of ATF coating materials and their preparation processes.

Reliability in Microjoining Processes for Instrumented Nuclear Fuel Element Fabrication

The paper presents the assembling flux of thermocouple-instrumented nuclear fuel element for research reactor, from the point of view of the welding / brazing engineer, considering nuclear quality and safety requirements: fuel element structural reliability (no radioactive leaks through joints) and temperature signal reliability (thermocouple sheath integrity), this signal being an essential parameter for reactor normal operation and emergency shut-down. The paper is a real case study for an experimental instrumented element recently developed at INR-Pitesti describing technology choices as balance between fabrication complexity and risk of failure in joining processes, especially in later stages when added value increases. All joints (welded or brazed) fall into microjoining category, and it is shown how some special provisions may influence reliability. Focus is put on brazing thin-walled Inconel sheathed thermocouples, where erosion and local loss of ductility are known issues, leading to sheath rupture. Choosing as filler the less aggressive BNi-9 helped too little. A simple but efficient technique has been developed to address this matter adequate to narrow spaces inside a nuclear fuel element, where no room is available for solutions described in literature e.g. distal preplacing of filler. The solution prevents sheath from having prolonged contact with large volume of molten filler by using locally a miniature barrier (thin stainless-steel coil or sleeve) which only allows capillary wetting, being also a perfect real-time visual indicator of brazing progress and completion. As proved in the present paper, this method along with using filler formulation with lower Carbon content (without organic binder) enhances significantly, 8 times at least, resistance to bending fatigue. A particular vacuum brazing chamber design is employed: narrow quartz tube with external induction coil and top fitting letting outside the long thermocouples attached, reducing much the chamber volume and degassing. Careful impedance match is therefore required to overcome induction power loss due to the larger coil-to-workpiece gap. Additional joining problems are discussed e.g. inherent differential expansion of long parts during induction heating which afterwards may put tension upon braze during solidification and determine delayed cracking, this being avoided through wise order of operations. Another concern is the final precision weld between instrumentation segment having attached the hard-to-handle long thermocouples bunch and nuclear segment with the heavy Uranium pellets. The result of this research is successful assembling of first Romanian prototype with joints exhibiting He leak rate bellow 1E-09 std.cc/sec and overall reliability proved during reactor irradiation testing.

Dose Rates From the Accidental Withdrawal of a Fuel Element From an Open Pool-Type Reactor

This paper studies the radiological consequences resulting from withdrawal of nuclear fuel element (FE) from a core of open pool type reactor during normal operation. The FE withdrawal accident may be occurred due to human error during routine transport of spent FE process or from failure of FE clamp during reactor normal operation. For both cases, the FE will move vertically upward toward pool water surface. In case of accidental failure of fuel clamp, a negative reactivity insertion in the core after FE withdrawal and the reactor will be shutdown. MCNP5 code was used in this study to calculate the radiation dose rate levels in the reactor hall and inside the control room during FE withdrawal. The results show that the operator in control room will receive dose rate lower than permissible dose rate limit when FE reaches at depths more than to 211 cm for vertical FE and at depths more than to 246 cm for horizontal FE.

The Evaluation Method of Uranium Hexafluoride Leakage Accident Release Source Term

Uranium hexafluoride is the intermediate material of uranium fuel enrichment process, which is widely used in uranium conversion plant, uranium enrichment plant and nuclear fuel element plant[1]. Because of its active chemical properties and its radioactive and chemical toxicity, great importance should be attached to the uranium hexafluoride release accident. This paper describes the possible leakage scenarios for uranium hexafluoride accident. And the general step of the evaluation for uranium hexafluoride leakage accident release source term is given, as well as an application example for the feed facility of a gaseous diffusion plant.