Introduction to Nuclear Reactor Physics

The course is not on the list Without time-table
Code Completion Credits Range Language
17ZAF Z,ZK 6 4+2 Czech
Department of Nuclear Reactors

Lectures begin with description of fundamentals of microstructure of matter up to the level of electrons, protons and neutrons. It is followed by nuclear reactions description with special focus on interaction of neutrons with nuclei. Probability of these reactions is expressed by cross-sections depending on neutron energy. Fissioning of heavy nuclei is physical principal of fission reactor operation. Students get acquainted with conditions for fission chain reaction realization, yield of fission products and energy released in reaction. Next, the most important types of nuclear reactors are described including the complete schematic drawing of a nuclear power plant with light-water reactor.

Calculation of a homogeneous nuclear reactor and diffusing medium is based on diffusion equation application. The equation was derived based on Fick?s law for diffusion in gases. Students learn to determine spatial distribution of neutron flux in diffusion matter, in homogeneous reactors with and without reflector. There are also summarized differences between homogeneous and heterogeneous reactors.

In the last part of the lectures, the gathered knowledge is extended by fundamentals of short-term and long-term nuclear reactor kinetics. Relation between determined neutron flux and reactor power distribution is stressed. Students are also taught the basics of dosimetry and radiation protection.



Syllabus of lectures:

1. Nuclear and sub-nuclear physics

Time range: 3lectures

Description of nucleus structure, binding energy and interaction among nucleons. Energy levels of nuclei, excited states. Determination of atomic density.

2. Neutron interactions

Time range: 3lectures

List of possible reactions of neutrons with nuclei. Microscopic and macroscopic cross-sections definitions. Examples of dependence of cross-section for individual reactions on neutron energy for particular isotopes. Distinction of fissile and fissionable heavy isotopes. Fissioning of heavy isotopes, fission yields and energetic balance. Neutron flux definition.

3. Nuclear reactors

Time range: 2lectures

Definition of basic terms important for nuclear reactor operation. Examples of individual nuclear reactors based on used fissile material, moderator and coolant. Design of fuel assemblies for pressurized water reactors of western (PWR) and eastern type (VVER). Complete schematic drawing of a nuclear power plant with light-water reactor.

4. Neutron diffusion and moderation

Time range: 6lectures

Diffusion equation derivation based on Fick?s law for diffusion of gases. Boundary conditions for diffusion equation solution definition. Neutron balance in the defined system. Diffusive medium characterisation based on its properties. Spatial neutron flux distribution in infinite and finite diffusive matters with point, linear and planar neutron sources.

5. Bare homogeneous reactors, reflected reactors and heterogeneous reactors

Time range: 6 lectures

Definition of multiplication factor and reactivity in infinite and finite system with fissile material. Definition of bare homogeneous reactor without moderator. Derivation of 1group critical equation. Bare homogeneous thermal reactor and derivation of 2group critical equation. Physical meaning of neutron age and diffusion length. Derivation of purpose of reflector for crtical dimensions and power flattening in a bare reactor. Differences between heterogeneous and homogeneous reactors.

6. Time depended reactor

Time range: 2 lectures

Definition of short-term, middle-term and long-term kinetics of a nuclear reactor. Fuel burnup, accumulation of fission products buildup. Specific influence on reactivity of 135Xe and 149Sm. Point kinetics equations. Influence and importance of delayed neutrons.

7. Heat generation in nuclear reactors

Time range: 2 lectures

Correlation between neutron flux and power generation in a nuclear reactor. Axial and radial temperature profile in a fuel assembly. Transfer of heat from a reactor.

8. Dosimetry and radiation protection

Time range: 2 lectures

Definition of basic dosimetry quantities. Description of potential health threats connected with ionizing radiation utilisation. Weighting factors for type of radiation and tissue for calculation of equivalent and effective dose. Radiation protection.

Syllabus of tutorials:

Content of exercises supports lectures with concrete calculations.

1. Calculation of atomic density

Time range: 1 ex.

2. Neutron interactions

Time range: 1 ex.

Calculation of microscopic and macroscopic cross-sections, reaction rate, neutron utilisation coefficient, reproduction factor and neutron flux.

3. Neutron diffusion

Time range: 4 ex.

Calculation of spatial neutron flux distribution in a diffusive media. Calculations for point, planar and linear neutron source. Application of boundary conditions for an infinite and finite diffusive medium. Calculation for interface boundary conditions.

4. Calculation of bare homogeneous reactors and reflected reactors

Time range: 5 ex.

Spatial distribution of neutron flux in homogeneous bare reactor for 1group and 2group diffusion equation. Application of 1group and modified 1group critical equation for calculation of critical dimensions and masses of spherical, rectangular and cylindrical reactors. Reflector saving and critical mass reduction for bare reactor and reflected reactor.

5. Point kinetics of a nuclear reactor

Time range: 1 ex.

Study Objective:

Students are aware of nucleus structure, nature and types of nuclear reactions. They learn characteristics of diffusive media and fissile and fissionable materials. They get acquainted with stationary solution of neutron flux distribution, reactor power and kinetics of its long-term and short-term changes. Students learn fundamentals of dosimetry and radiation protection.

After the subject completion, students are prepared to determine material composition (specifically to calculate atomic densities) important for all analysis conducted in reactor physics. They are also able to independently calculate spatial neutron flux distribution in simple geometries utilizing the diffusion equation in 1group and even 2group approximation. Students are also prepared to extend their knowledge to advanced tasks. This represents especially further methods for neutron flux calculation and calculation of kinetics of nuclear reactor power changes.

Study materials:

Key references:

John R. Lamarsh, Anthony J. Baratta, Introduction to Nuclear Engineering, Prentice Hall, New Jersey, 2001

Recommended references:

Paul Reuss, Neutron Physics, EDP Sciences, Les Ulix Cedex A, Francie, 2008

James, J. Duderstadt, Louis J. Hamilton, Nuclear Reactor Analysis, John Wiley & Sons, USA, 1976

Further information:
No time-table has been prepared for this course
The course is a part of the following study plans:
Data valid to 2019-06-26
For updated information see http://bilakniha.cvut.cz/en/predmet1182406.html