Effect of cladding on thermal behavior of nuclear fuel element with non-uniform heat generation

This paper deals with the numerical prediction of thermal performance characteristics of a rectangular nuclear fuel element dissipating fission heat in its surrounding medium. Assuming uniform heat generation within the fuel element, the steady, two-dimensional heat conduction equation along with appropriate boundary conditions is solved using second order accurate finite difference scheme. On the basis of axial temperature profiles and the variation of Biot number, Bi and heat generation parameter, Q with respect to maximum temperature, θmax within the fuel element, results are presented for a wide range of aspect ratio, Ar, Bi, and Q and discussed in detail. It is found that with the increase in Ar, the effect of axial conduction on the temperature profile diminishes rapidly and ultimately for all values of Ar ≥7, it remains confined to the region very close to the leading edge. It is also concluded that for a set of fixed values of Ar and Q, there exists a lower limiting value of Bi below which it cannot be decreased. Similarly, for a set of fixed values of Ar and Q, there also exists an upper limiting value of Q above which θmax exceeds its allowable limit. Furthermore, it is concluded that the requirement of increase in Bi due to increase in Q, θmax being within its allowable limit, becomes larger and larger as Q becomes higher and higher.

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High-Fidelity Calculation of Effective Thermal Response of Composite Media With Heat Generation Source

ASME 2020 Verification and Validation Symposium ◽

10.1115/vvs2020-8836 ◽

2020 ◽

Author(s):

Kevin Irick ◽

Nima Fathi

Keyword(s):

Heat Transfer ◽

Thermal Conductivity ◽

Thermal Behavior ◽

Nuclear Fuel ◽

Heat Generation ◽

Thermal Response ◽

Thermal Conditions ◽

Unit Cell ◽

High Fidelity ◽

Conductive Heat

Abstract The complexity of conductive heat transfer in a structure increases with heterogeneity (e.g., multi-component solid-phase systems with a source of internal thermal heat generation). Any discontinuity of material property — especially thermal conductivity — would warrant a thorough analysis to evaluate the thermal behavior of the system of interest. Heterogeneous thermal conditions are crucial to heat transfer in nuclear fuel assemblies, because the thermal behavior within the assemblies is governed significantly by the heterogeneous thermal conditions at both the system and component levels. A variety of materials have been used as nuclear fuels, the most conventional of which is uranium dioxide, UO2. UO2 has satisfactory chemical and irradiation tolerances in thermal reactors, whereas the low thermal conductivity of porous UO2 can prove challenging. Therefore, the feasibility of enhancing the thermal conductivity of oxide fuels by adding a high-conductivity secondary solid component is still an important ongoing topic of investigation. Undoubtedly, long-term, stable development of clean nuclear energy would depend on research and development of innovative reactor designs and fuel systems. Having a better understanding of the thermal response of the unit cell of a composite that represents a fuel matrix cell would help to develop the next generation of nuclear fuel and understand potential performance enhancements. The aim of this article is to provide an assessment of a high-fidelity computational model response of heterogeneous materials with heat generation in circular fillers. Two-dimensional, steady-state systems were defined with a circular, heat-generating filler centered in a unit-cell domain. A Fortran-based finite element method (FEM) code was used to solve the heat equation on an unstructured triangular mesh of the systems. This paper presents a study on the effects of a heat-generating filler material’s relative size and thermal conductivity on effective thermal conductance, Geff, within a heterogenous material. Code verification using the method of manufactured solution (MMS) was employed, showing a second-order accurate numerical implementation. Solution verification was performed using a global deviation grid convergence index (GCI) method to assess solution convergence and estimate solution numerical uncertainty, Unum. Trend results are presented, showing variable response in Geff to filler size and thermal conductivity.

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DEVELOPMENT OF AN ACOUSTICAL THERMOMETER FOR A GRAPHITE MATRIX NUCLEAR FUEL ELEMENT.

10.2172/4442672 ◽

1966 ◽

Author(s):

J.B. Roes ◽

D.L. Peat

Keyword(s):

Fuel Element ◽

Nuclear Fuel ◽

Graphite Matrix ◽

Nuclear Fuel Element

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Reliability in Microjoining Processes for Instrumented Nuclear Fuel Element Fabrication

Advanced Materials Research ◽

10.4028/www.scientific.net/amr.1157.31 ◽

2020 ◽

Vol 1157 ◽

pp. 31-37

Author(s):

Călin Truţă ◽

Adrian Amzoi ◽

Dumitru Barbos

Keyword(s):

Fuel Element ◽

Nuclear Fuel ◽

Structural Reliability ◽

Added Value ◽

Impedance Match ◽

Normal Operation ◽

Local Loss ◽

Steel Coil ◽

Nuclear Fuel Element ◽

Prolonged Contact

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.

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