An Engineer's Guide to Nuclear Reactor Core Materials.

Format:

Book

Author/Creator:

Griffiths, Malcolm.

Language:

English

Subjects (All):

Nuclear engineering.

Zirconium alloys.

Physical Description:

1 online resource (1136 pages)

Edition:

1st ed.

Place of Publication:

Chantilly : Elsevier Science & Technology, 2025.

Summary:

An Engineer's Guide to Nuclear Reactor Core Materials presents everything a reactor engineer needs to know to work on reactor core structural materials and core internals.This book outlines the impact of radiation damage on materials and provides the necessary tools to perform calculations of atomic displacement and transmutation, especially He.

Contents:

Front Cover

An Engineer's Guide to Nuclear Reactor Core Materials

Copyright Page

Contents

Preface

1 Physical metallurgy of reactor core materials

Preambles

1.1 Crystallography

1.1.1 Crystal structure

1.1.2 Crystallography of shear deformation

1.1.2.1 Body-centred-cubic crystals

1.1.2.2 Face-centred-cubic crystals

1.1.2.3 Hexagonal-close-packed crystals

1.2 Mechanical properties

1.2.1 Uniaxial deformation

1.2.2 Multiaxial deformation

1.3 Design requirements

1.3.1 Tensile strength

1.3.2 Fracture toughness

1.3.2.1 Austenitic alloys

1.3.2.2 Ferritic steels

1.3.2.3 Zr alloys

References

Further reading

2 Radiation damage and transmutation

2.1 Radiation damage production

2.2 Neutron spectra

2.3 Dosimetry

2.3.1 Iron wire dosimetry

2.3.2 Spectral averaging

2.4 Lethargy

2.5 Measures of neutron dose and displacement damage

2.6 Transmutation and gas production

2.7 The 59Ni effect

2.8 Production of freely migrating point defects

2.8.1 Freely migrating defect production without consideration of the 59Ni effect

2.8.2 Freely migrating defect production with consideration of the 59Ni effect

2.9 Gamma damage

2.9.1 The γ-induced displacement mechanism

2.9.2 Comparisons of different assessments of γ-damage

3 Control rod materials

3.1 Boron carbide (B4C)

3.1.1 Nuclear properties

3.1.2 Control rod assemblies containing B4C

3.2 Hafnium (boiling water reactor, pressurized water reactor, VVER)

3.2.1 Nuclear properties

3.2.2 Control rod assemblies containing Hf

3.3 Silver-indium-cadmium (pressurized water reactor)

3.3.1 Nuclear properties

3.3.2 Control rod assemblies containing AgInCd (pressurized water reactors)

3.4 Cadmium (heavy water reactors)

References.

4 Microstructure characterization

4.1 Diffraction principles

4.1.1 X-rays

4.1.2 Electrons

4.1.3 Structure factors

4.1.3.1 Body-centred cubic structure

4.1.3.2 Face-centred cubic structure

4.1.3.3 Hexagonal-close packed structure

4.1.3.4 Face-centred cubic-ordered solid solution (L12 structure)

4.2 Defect analysis in the transmission electron microscopy

4.3 Texture analysis

4.3.1 Pole figures

4.3.1.1 Direct pole figure

4.3.1.2 Inverse pole figure

4.3.1.3 Electron back-scattered diffraction pole figure

4.3.1.4 Crystal orientation distribution function

4.3.2 Kearns' texture parameter

4.4 X-ray diffraction line profile analysis

4.4.1 Line broadening resulting from lattice strain and domain size

4.4.2 Line broadening due to strain

4.4.3 Line broadening due to domain size

4.4.4 Dislocation contrast in transmission electron microscopy and X-ray diffraction analysis

4.4.5 Integral breadth

4.4.6 Warren and Averbach construction

4.5 Case study - characterization of vacancy dislocation loops in electron-irradiated Zr

4.5.1 Characterization of a-type dislocation loops

4.5.2 Characterization of c-component loops

5 Microstructure evolution

5.1 Hardening and dislocation loops

5.1.1 Austenitic alloys

5.1.2 Ferritic steels

5.1.3 Zirconium alloys

5.2 Void/cavity swelling

5.2.1 Thermodynamics of void formation

5.2.2 Thermodynamics of He-stabilized cavity growth

5.2.3 Swelling rates with an evolving microstructure

5.2.4 Austenitic alloys

5.2.5 Ferritic stainless steels

5.2.6 Zr alloys

5.3 Transmutation

5.3.1 Transmutation of alloying elements

5.3.2 He and H gas production from transmutation

5.4 Microchemistry and phase stability

5.4.1 Phase stability in Zr alloys.

5.4.2 Phase stability in stainless steels

5.4.2.1 Inverse Kirkendall diffusion

5.4.2.2 Radiation-enhanced self-diffusion

6 Tensors

6.1 Definition

6.2 The strain tensor

6.3 Transformation of axes

6.3.1 Twinning

6.3.2 Dislocation slip

6.4 The stress tensor

6.5 The Schmid tensor

6.6 Radius normal property

6.7 Tensor analysis: yielding

6.8 Tensor analysis: creep

6.8.1 Isotropic materials

6.8.2 Anisotropic materials

6.8.2.1 Irradiation creep

6.8.2.2 Case study: calandria tube failure

7 Plastic deformation and fracture

7.1 Deformation

7.1.1 Deformation mechanisms

7.1.2 Techniques for assessing deformation mechanisms

7.1.3 Geometrically necessary dislocations

7.1.4 Deformation mechanisms and yield stress

7.2 Fracture

7.2.1 Austenitic alloys

7.2.1.1 Intergranular cracking of austenitic alloys

7.2.1.2 Transgranular cracking of austenitic alloys

7.2.2 Fracture of Zr alloys

7.2.2.1 Impurity effect on fracture of Zr alloys

7.2.2.2 Hydride cracking

7.2.3 Fracture of ferritic and ferritic/martensitic steels

7.3 Case study: ductile-to-brittle transition temperature for hydride cracking in a CANDU reactor

8 Dimensional stability

8.1 Zirconium alloys

8.1.1 Irradiation growth of Zr alloys

8.1.1.1 Effects of microstructure and chemistry

8.1.1.2 Effects of grain structure

8.1.2 Irradiation creep of Zr alloys

8.1.2.1 Mechanisms of irradiation creep

8.1.2.2 Anisotropy of irradiation creep of Zr alloys

8.2 Ferritic and austenitic alloys

8.2.1 Irradiation swelling

8.2.1.1 Austenitic stainless steel alloys

8.2.1.2 Nickel alloys

8.2.1.3 Ferritic and ferritic/martensitic stainless steels.

8.2.2 Irradiation creep in austenitic and ferritic alloys

8.2.2.1 Creep without swelling (B0)

8.2.2.2 Swelling-dependent creep (DṠ)

9 Rate theory

9.1 The balance equations

9.1.1 Recombination

9.1.2 Sink strength for dislocations

9.1.3 Sink strength for cavities

9.1.4 Sink strength for grain boundaries

9.2 Mechanisms and modelling of irradiation growth and creep in Zr alloys

9.2.1 Modelling of irradiation growth in Zr-alloys

9.2.2 Modelling of irradiation creep in Zr-alloys

9.2.3 Modelling of irradiation swelling, creep and embrittlement in cubic metals

9.2.3.1 Irradiation swelling in cubic metals

9.2.3.2 Irradiation creep in cubic metals

9.2.3.3 Embrittlement of cubic metals

9.3 Case study - rate theory applied to He-embrittlement in reactor alloys

10 An engineer's guide to Zr-alloys

10.1 Nuclear properties

10.2 Point defect properties

10.3 Production and properties of Zr

10.3.1 Ore processing

10.3.2 Metal production

10.4 Alloying elements and impurities in nuclear grade Zr

10.4.1 Impurities

10.4.2 Alloying elements

10.4.2.1 Zr-Nb alloys

10.4.2.2 Zr-Sn alloys

10.4.2.3 Zr-Nb-Sn alloys

10.4.3 Phase stability during Irradiation

10.5 Texture

10.5.1 Texture evolution in α-phase alloys

10.5.1.1 α phase alloy plates

10.5.1.2 α phase alloy tubes

10.5.2 Texture evolution in α + β phase alloys

10.5.2.1 α + β phase alloy plates

10.5.2.2 α + β phase alloy tubes

10.6 Physical properties

10.6.1 Elastic properties

10.6.1.1 Elastic moduli

10.6.1.2 Thermal expansion

10.6.1.3 Residual (intergranular) stresses

10.6.2 Mechanical properties

10.6.2.1 Unirradiated material

10.6.2.1.1 Uniaxial tensile properties

10.6.2.1.2 Multiaxial mechanical properties

10.6.2.2 Irradiated material.

10.6.2.2.1 Uniaxial tensile properties

10.6.2.2.2 Multiaxial deformation properties

10.7 Delayed hydride cracking

10.7.1 Unirradiated material

10.7.2 Irradiated material

10.8 Hydrogen pick-up

10.8.1 Hydrogen pickup from corrosion

10.8.2 Hydrogen pickup from dissimilar metal contact

11 A reactor engineer's guide to Ni-alloys

11.1 Composition, physical metallurgy and mechanical properties of Ni-alloys

11.1.1 Alloy compositions

11.1.2 Physical metallurgy

11.1.2.1 Precipitation hardening

11.1.2.2 Thermo-mechanical processing

11.1.3 Mechanical properties

11.2 Irradiation effects

11.2.1 Fast reactors and ion irradiation facilities

11.2.2 Light water reactors

11.2.3 Heavy water reactors

11.2.3.1 Cavities

11.2.3.2 Dislocations

11.2.3.3 Precipitates

11.3 Stress corrosion cracking

11.3.1 Stress corrosion cracking of Ni-alloys

11.3.2 Irradiation-assisted stress corrosion cracking of Ni-alloys

12 An engineer's guide to steels

12.1 Composition and physical metallurgy

12.1.1 Ferritic steels

12.1.2 Austenitic stainless steels

12.2 Mechanical properties

12.3 Swelling

12.4 He embrittlement

12.5 Irradiation-assisted stress corrosion cracking

12.5.1 Phenomenology of stress corrosion cracking

12.5.2 Phenomenology of irradiation-assisted stress corrosion cracking

12.5.2.1 Material

12.5.2.2 Environment

12.5.2.3 Stress

13 An enginee's guide to graphite

13.1 Manufacture and physical properties

13.1.1 Manufacture of graphite

13.1.2 Crystal structure of graphite

13.1.3 Physical properties and microstructure of graphite

13.2 Wigner energy

13.3 Irradiation effects on physical properties

13.4 Dimensional stability of graphite.

13.4.1 Irradiation growth.

Notes:

Description based on publisher supplied metadata and other sources.

Part of the metadata in this record was created by AI, based on the text of the resource.

ISBN:

0-443-30105-0

9780443301056

OCLC:

1557605809