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Lecture

Multigroup theory: Two-group method and group structures

In course

PHYS-443: Physics of nuclear reactors

In this course, one acquires an understanding of the basic neutronics interactions occurring in a nuclear fission reactor as well as the conditions for establishing and controlling a nuclear chain reaction. ...

Description

This lecture covers the multi-group method, focusing on the two-group method, group structures, multi-group constants, main equations, and boundary conditions. It explains the heuristic derivation of multi-group diffusion equations and the determination of group constants.

Instructors (3)

Carlo Fiorina

Andreas Pautz

Mathieu Hursin

Official source

Ontological neighbourhood

Energy engineering

Nuclear power: Topics in nuclear power

Related concepts (32)

Neutron cross section

In nuclear physics, the concept of a neutron cross section is used to express the likelihood of interaction between an incident neutron and a target nucleus. The neutron cross section σ can be defined as the area in cm2 for which the number of neutron-nuclei reactions taking place is equal to the product of the number of incident neutrons that would pass through the area and the number of target nuclei. In conjunction with the neutron flux, it enables the calculation of the reaction rate, for example to derive the thermal power of a nuclear power plant.

Neutron temperature

The neutron detection temperature, also called the neutron energy, indicates a free neutron's kinetic energy, usually given in electron volts. The term temperature is used, since hot, thermal and cold neutrons are moderated in a medium with a certain temperature. The neutron energy distribution is then adapted to the Maxwell distribution known for thermal motion. Qualitatively, the higher the temperature, the higher the kinetic energy of the free neutrons. The momentum and wavelength of the neutron are related through the de Broglie relation.

Nuclear reactor physics

Nuclear reactor physics is the field of physics that studies and deals with the applied study and engineering applications of chain reaction to induce a controlled rate of fission in a nuclear reactor for the production of energy. Most nuclear reactors use a chain reaction to induce a controlled rate of nuclear fission in fissile material, releasing both energy and free neutrons.

Fast-neutron reactor

A fast-neutron reactor (FNR) or fast-spectrum reactor or simply a fast reactor is a category of nuclear reactor in which the fission chain reaction is sustained by fast neutrons (carrying energies above 1 MeV or greater, on average), as opposed to slow thermal neutrons used in thermal-neutron reactors. Such a fast reactor needs no neutron moderator, but requires fuel that is relatively rich in fissile material when compared to that required for a thermal-neutron reactor.

Neutron moderator

In nuclear engineering, a neutron moderator is a medium that reduces the speed of fast neutrons, ideally without capturing any, leaving them as thermal neutrons with only minimal (thermal) kinetic energy. These thermal neutrons are immensely more susceptible than fast neutrons to propagate a nuclear chain reaction of uranium-235 or other fissile isotope by colliding with their atomic nucleus. Water (sometimes called "light water" in this context) is the most commonly used moderator (roughly 75% of the world's reactors).

Related lectures (36)

Neutron Transport Theory: Fundamentals and Applications

Explores neutron transport theory, scalar neutron flux, and Fick's law in reactor physics.

The Diffusion of Neutrons

Covers the transport of neutrons, scalar neutron flux, neutron currents, and more, along with fuel cycles and types of nuclear reactors.

Introduction and Review of Nuclear Physics

Covers the basics of nuclear physics, including reactor physics, neutron interactions, fission, fusion, and radioactivity.

Interactions of Neutrons with Matter

Explores the interaction of neutrons with matter through cross sections, scattering, absorption, and energy dependencies.

Multigroup theory