A Review of the Effects of 

Radiation on Microstructure 

and Properties of Concretes 

Used in Nuclear Power Plants

Office of Nuclear Regulatory Research

NUREG/CR-7171

ORNL/TM-2013/263
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A Review of the Effects of 

Radiation on Microstructure 

and Properties of Concretes 

Used in Nuclear Power Plants

Manuscript Completed: August 2013

Date Published: November 2013

Prepared by: 

Kaspar William

University of Houston

Department of Civil and Environmental

 Engineering

Houston, TX 77204-4003

Yunping Xi

University of Colorado

Department of Civil, Environmental

 and Architectural Engineering

Boulder, CO 80309-0428

Dan Naus

Oak Ridge National Laboratory

Materials Science and Technology Division

Oak Ridge, TN 37831-6069

Herman L. Graves, III, NRC Technical Monitor

NRC Job Code N6978

Prepared for:

Division of Engineering

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

NUREG/CR-7171

ORNL/TM-2013/263

iii

ABSTRACT

The current understanding of the effects of radiation on concrete materials is summarized. Much 

of the research related to this topic was conducted from the 1960s into the 1970s in support of 

development of prestressed concrete reactor vessels for high-temperature reactors and 

radioactive waste storage facilities. From the 1970s until 2012, the efforts addressed 

development of computational models of radiation damage to concrete and assessment of the 

effects of radiation on concrete durability. Concrete structures used as biological shields and as 

support for the reactor pressure vessel in current light-water reactor plants are subjected to two 

types of radiation, gamma and neutron. Gamma and neutron radiation are described, and their 

interactions with concrete constituents are noted. A summary of the effects of neutron and 

gamma radiation on the mechanical and physical properties of concrete is provided. Information 

is presented on thermal effects due to gamma heating of the concrete. Coupling effects of 

mechanical loading, thermal effect, moisture content, and radiation are described, and an 

example provided using a computational model to indicate radiation damage in a shield for a 

next-generation nuclear facility. Nondestructive testing methods are identified for assessment of 

the condition of reinforced concrete structures. Radiation limits for neutron and gamma 

exposure are listed. Finally, recommendations are provided to (1) expand the very limited 

database on the effect of neutron and gamma radiation on concrete’s microstructure and 

mechanical and physical properties, (2) conduct calculations to establish the neutron and 

gamma fields that develop in concrete shields and support structures after various operating 

periods, and (3) quantify the interaction of temperature and irradiation effects.

v

FOREWORD

Commercial nuclear power plants (NPPs) in the United States contain concrete structures 

whose performance and function are necessary for the protection and safety of plant operating 

personnel, the general public, and the environment. Typical safety-related concrete structures 

contained in light-water reactor (LWR) plants are generally grouped into four categories: primary 

containments, containment internal structures, secondary containments/reactor buildings, and 

other structures. These safety-related concrete and other structures must be capable of 

maintaining structural integrity for the operating life of the plant.

Since 2008 the Office of Nuclear Regulatory Research (RES) staff has had discussions with 

other NRC staff, U.S. industry and academic representatives, and Japanese researchers on the 

topic of the effects of radiation fluences on concrete. Based on these discussions, RES planned 

and managed a research program to provide information on the effect of radiation on concrete 

structures, particularly in operating environments for existing LWRs and for some new reactor 

designs. 

The objective of this research is threefold: (1) to determine whether current methods used to 

evaluate the effects of gamma and neutron radiation fluences on concrete structures are 

sufficient in determining the concrete properties for new reactor designs, (2) to determine the 

changes in the dimensional or physical properties of concrete that will be brought about by the

elevated temperatures and the radiation fluences anticipated in new, high-temperature reactor 

designs, and (3) how these changes may also affect long-term NPP operation.

This NUREG/CR provides a review of concrete as a nuclear shielding material and the effects of 

nuclear radiation on the mechanical properties of concrete. Also discussed are computational 

models for radiation damage of concrete, assessment techniques to evaluate radiation damage,

and radiation exposure limits for concrete NPP structures. The information in this report,

coupled with the applicable code provisions discussed, could form the basis for new regulatory 

guidance on the review of radiation effects on concrete used in NPP structures.

vii

CONTENTS

Section Page

ABSTRACT................................................................................................................................iii

FOREWORD ..............................................................................................................................v

LIST OF FIGURES ....................................................................................................................ix

LIST OF TABLES.....................................................................................................................xiii

EXECUTIVE SUMMARY...........................................................................................xv

ACKNOWLEDGMENTS ...........................................................................................xvii

ABBREVIATED TERMS .......................................................................................................... xix

1. INTRODUCTION ................................................................................................................ 1

2. GAMMA RAY AND NEUTRON ATTENUATION ................................................................. 3

2.1 GAMMA RADIATION................................................................................................. 4

2.2 NEUTRON RADIATION............................................................................................. 5

2.3 RADIATION SHIELDING EFFECTS.......................................................................... 7

3. CONCRETE AS A NUCLEAR SHIELDING MATERIAL ...................................................... 9

3.1 AGGREGATES USED IN CONCRETE FOR SHIELDING STRUCTURES................ 9

3.2 IMPACT OF IRRADIATION ON THE CRYSTALLINE STRUCTURE OF 

AGGREGATES.........................................................................................................11

3.3 CEMENTS AND INTERNAL STRUCTURE OF CEMENT PASTE ............................12

3.4 RADIATION SHIELDING PROPERTIES OF CONCRETE........................................18

3.5 THE STATES OF WATER IN CONCRETE AND THE FUNCTION OF WATER 

FOR RADIATION SHIELDING..................................................................................20

3.6 THE EFFECT OF POLYMER FIBERS......................................................................22

3.7 DESIGN OF CONCRETE RADIATION SHIELDING.................................................23

4. HEAT OF RADIATION AND THE THERMAL STRESS DUE TO RADIATION ...................25

4.1 HEAT GENERATED BY GAMMA AND NEUTRON IRRADIATION ..........................25

4.2 TEMPERATURE VARIATION DUE TO THE HEAT OF RADIATION........................27

4.3 MACROSCOPIC THERMAL STRESS IN SHIELDING STRUCTURES DUE 

TO RADIATION ........................................................................................................30

4.4 MICROSCOPIC THERMAL STRESS IN CONCRETE DUE TO RADIATION ...........30

4.5 THERMAL EXPANSION AND THERMAL CONDUCTIVITY .....................................34

5. THE EFFECTS OF NUCLEAR RADIATION ON PROPERTIES OF CONCRETE..............39

5.1 INTERACTION BETWEEN NUCLEAR RADIATION AND THE INTERNAL 

STRUCTURE OF CONCRETE.................................................................................39

5.2 COMPRESSIVE STRENGTH...................................................................................45

5.3 TENSILE STRENGTH ..............................................................................................53

5.4 MODULUS OF ELASTICITY.....................................................................................55

5.5 CREEP, SHRINKAGE, AND VOLUMETRIC VARIATION.........................................56
viii

5.6 EFFECT OF HIGH TEMPERATURE ON THE NUCLEAR SHIELDING 

CAPACITIES OF CONCRETE..................................................................................59

5.7 EFFECT OF LOW TEMPERATURE ON THE NUCLEAR SHIELDING 

CAPACITIES OF CONCRETE..................................................................................61

6. INTERACTION OF NUCLEAR RADIATION, DURABILITY, AND SHIELDING 

EFFECTIVENESS..............................................................................................................63

6.1 ASR AND THE COUPLING EFFECT BETWEEN ASR AND RADIATION ................63

6.1.1 ASR of Concrete ...........................................................................................63

6.1.2 ASR and Radiation of Concrete.....................................................................64

6.2 CARBONATION .......................................................................................................66

6.3 RADIOLYSIS AND EVAPORATION OF WATER......................................................67

6.4 THE EFFECT OF CONCRETE CRACKING/SPALLING ON SHIELDING.................70

7. COMPUTATIONAL MODELS FOR RADIATION DAMAGE OF CONCRETE.....................73

7.1 INTRODUCTION ......................................................................................................73

7.2 NUMERICAL MODELS FOR RADIATION DAMAGE EVOLUTION ..........................73

7.3 DAMAGE-PLASTICITY MODELS FOR CONCRETE................................................80

8. DETECTION AND EVALUATION OF RADIATION DAMAGE IN NPP CONCRETE 

STRUCTURES ..................................................................................................................83

8.1 NONDESTRUCTIVE TEST METHODS FOR CONCRETE AND 

REINFORCED CONCRETE STRUCTURES............................................................83

8.2 SPECIFIC CONDITIONS RELATED TO THE EFFECT OF RADIATION ON 

CONCRETE..............................................................................................................87

8.2.1 Temperature effects ......................................................................................88

8.2.2 ASR potential of existing concrete and detection...........................................88

8.2.3 Carbonation...................................................................................................89

9. RADIATION EXPOSURE LIMITS FOR NPP CONCRETE.................................................91

10. SUMMARY AND CONCLUSIONS .....................................................................................93

11. REFERENCES ..................................................................................................................97
ix

LIST OF FIGURES

Figure Page

Figure 2.1. The photoelectric effect, Compton scattering, and pair production of 

gamma radiation. ................................................................................................ 4

Figure 2.2. Relationship among gamma ray energy, atomic number, and interaction 

mechanisms........................................................................................................ 5

Figure 2.3. Inelastic and elastic scattering processes of neutron radiation............................ 6

Figure 2.4. Experimental setups for measuring mass attenuation coefficients. ..................... 7

Figure 3.1. Multiple-scale internal structure of concrete (schematic)..................................... 9

Figure 3.2. Density of quartz as a function of fast neutron fluence. ......................................12

Figure 3.3. Powers’ model for the internal layered structure of C-S-H..................................13

Figure 3.4. Feldman and Sereda’s model for the internal layered structure of C-S-H...........14

Figure 3.5. (a) Internal structure of 1.4 nm tobermorite for C-S-H; (b) Detailed 

molecular structure of C-S-H..............................................................................14

Figure 3.6. SEM images of cement paste irradiated with gamma radiation. .........................16

Figure 3.7. SEM images of cement paste that was made with a silica fume additive 

and irradiated with gamma radiation. .................................................................17

Figure 3.8. SEM images of cement paste that was made with a fly ash additive and 

irradiated with gamma radiation. ........................................................................18

Figure 3.9. Relationship between equilibrium water content of hardened cement 

paste and relative vapor pressure. .....................................................................20

Figure 3.10. Changes in the linear attenuation coefficients of concrete versus the 

changes in w/c (a) for a broad beam of gamma radiation from a Cs-137 

source and (b) for a broad beam of neutron radiation from an Am–Be 

source................................................................................................................22

Figure 3.11. Changes in strength of concrete versus the changes in w/c...............................22

Figure 3.12. (a) Compressive strength data with various radiation doses and (b) 

modulus of elasticity test data with various radiation doses................................23

Figure 4.1. Measured and calculated (a) heat generation due to the total thermal 

neutrons and (b) the calculated heat generation due to total thermal and 

reactor thermal neutrons in ordinary concrete. ...................................................27

Figure 4.2. Temperature rise in ordinary concrete because of gamma rays due to 

capture of thermal neutrons. ..............................................................................29

Figure 4.3. A model for concrete with aggregates of different sizes for analyzing the 

microscopic stress at the aggregate–cement paste interface due to the 

heat of radiation (the internal structure of each aggregate element is 

explained in Figure 4.4)......................................................................................32

Figure 4.4. The four phases in each element in Figure 4.3. .................................................33
x

Figure 4.5. The interface pressure p induced by the expansion of aggregate under 

radiation.............................................................................................................33

Figure 4.6. Relative expansion of concrete with chromite filler as aggregate under 

different temperatures: (a) first heating; (b) second heating; (1) irradiated; 

(2) heated; and (3) unheated. ............................................................................35

Figure 4.7. Relative expansion of concrete with sandstone and river sand aggregate 

under different temperatures: (a) first heating; (b) second heating; 1, 2, 

and 3 indicate samples irradiated to fluences of 3 × 1020, 1.2 to 1.45 × 

1020, and 4 to 6 × 1019 n/cm2

, respectively; ---- Control samples kept at 

20°C; ____ control samples kept at 20°C to 220°C............................................36

Figure 4.8. Thermal conductivity of concrete after neutron irradiation (Kc) relative to 

thermal conductivity of untreated concrete (Kco).................................................37

Figure 5.1. The specific pore surface area vs dose of gamma radiation...............................40

Figure 5.2. The porosity of concrete vs dose of gamma radiation. .......................................41

Figure 5.3. The intensity of CaCO3 vs dose of gamma radiation..........................................41

Figure 5.4. The pore size distributions of irradiated (1MGy) and unirradiated concrete 

specimens in the surface of concrete. Filled circles are unirradiated 

samples. Open squares are irradiated samples. ................................................42

Figure 5.5. The pore size distributions of irradiated (1MGy) and unirradiated concrete 

specimens in the center of concrete. Open circles are unirradiated 

samples. Open squares are irradiated samples. ................................................43

Figure 5.6. Effect of neutron radiation on the residual compressive strength of 

concrete.............................................................................................................46

Figure 5.7. Effect of gamma radiation on the residual compressive strength of 

concrete.............................................................................................................46

Figure 5.8. Effect of gamma radiation on the compressive strength of concrete and 

mortar: (a) test data by Vodák et al. (2005) and (b) text data by Soo and 

Milian (2001)......................................................................................................47

Figure 5.9. Compressive strength of concrete under different levels of neutron 

radiation.............................................................................................................49

Figure 5.10. The problems in the test data collected by Hilsdorf et al. (1978) on 

compressive strength of concrete. .....................................................................49

Figure 5.11. Comparisons of the test data collected. .............................................................51

Figure 5.12. Effect of neutron radiation on tensile strength of concrete..................................54

Figure 5.13. Effect of gamma radiation on splitting tensile strength of concrete. ....................54

Figure 5.14. Effect of gamma radiation on flexural tensile strength of concrete......................55

Figure 5.15. The effect of neutron radiation on the modulus of elasticity of concrete: Ec

= after neutron radiation and Eco = untreated.....................................................55

Figure 5.16. Static modulus of elasticity of concrete after different levels of neutron 

radiation.............................................................................................................56

Figure 5.17. The effect of gamma radiation on creep and shrinkage of concrete. ..................57
xi

Figure 5.18. The effect of neutron radiation on the volume of concrete..................................57

Figure 5.19. Creep of concrete made with North Notts coarse aggregate stored under 

dry conditions (T = 20oC –22oC, relative humidity = 65%) and wet 

conditions (under water at T = 20 oC to 22°C) (creep in m/m). ...........................58

Figure 5.20. Shrinkage of concrete made with North Notts coarse aggregate stored 

under dry conditions (T = 20oC –22oC, relative humidity = 65%) and wet 

conditions (under water at T = 20 oC to 22oC) (shrinkage in m/m)......................59

Figure 5.21. (a) Gamma-ray attenuation coefficients vs temperature and 

(b) attenuation index of neutron vs temperature for the four concretes: D 

mix (dolomite crushed coarse aggregate), F mix (river natural coarse 

aggregate), H mix (hematite crushed coarse aggregate), and S mix 

(serpentine crushed coarse aggregate)..............................................................60

Figure 5.22. (a) Linear attenuation coefficient μ of A-type concretes as a function of 

freeze-thaw cycles; and (b) Linear attenuation coefficient μ of B-type 

concretes as a function of freeze-thaw cycles. ...................................................62

Figure 6.1. (a) The height of terrace emerged at the boundary of irradiated and 

unirradiated areas of crystalline quartz after being immersed in 1 mol/dm3

NaOH solution at 353 K for 4 h. (b) Evolution of a terrace on 1 × 1015

Ar/cm2 irradiated crystalline quartz in 1 mol/dm3 NaOH solution at 

different temperatures........................................................................................65

Figure 6.2. (a) Leaching of Sr from samples with w/c = 0.3. (b) Leaching of Cs from 

samples with w/c = 0.3.......................................................................................67

Figure 6.3. Gamma ray dose vs water retained in concrete. ................................................68

Figure 6.4. (a) Test configurations of gamma irradiation tests; and (b) arrangement of 

containers. .........................................................................................................68

Figure 6.5. (a) Temperature in the room and in the specimens and (b) water released 

vs irradiation duration.........................................................................................69

Figure 6.6. Generation of hydrogen vs irradiation period......................................................69

Figure 6.7. The interface pressure and the crack pattern due to rebar corrosion..................70

Figure 6.8. Crack growth around steel bars due to the rust formation. .................................70

Figure 6.9. Schematic of the positioning system and the concrete sample...........................71

Figure 6.10. Experimental setup of the testing system for cracks in concrete. .......................71

Figure 6.11. The relationships between crack width, material thickness, and shielding 

performance.......................................................................................................72

Figure 7.1. Modulus of elasticity of concrete after neutron irradiation (Ec) related to 

modulus of elasticity of untreated concrete (Eco)...............................................74

Figure 7.2. Contour maps of relative humidity and temperature after 1 h, 1 week, and 

6 months............................................................................................................76

Figure 7.3. Contours after the fifth irradiation cycle and evolution of temperature at 

three points of the target bunker. .......................................................................76
xii

Figure 7.4. Radiation damage progression with depth for ordinary Portland cement 

concrete under radiation fluence for up to 50 years: (a) fast neutrons, 

(b) thermal neutrons...........................................................................................78

Figure 7.5. Radiation damage progression with depth for ordinary Portland cement 

and heavyweight concretes under radiation fluence up to 1 year: (a) fast 

neutrons, (b) thermal neutrons...........................................................................79

Figure 7.6. Comparison of Damage and Plasticity Models in Uniaxial Compression 

and Tension.......................................................................................................81

Figure 7.7. Comparison of Shear Response Predictions of Damage and Plasticity 

Models under Simple Shear...............................................................................82
xiii

LIST OF TABLES

Table Page

Table 3.1. Examples of aggregates based on density.........................................................10

Table 3.2. Properties of aggregates for radiation attenuation..............................................11

Table 3.3. Measured linear attenuation coefficients μ (1/cm) and half-value 

thicknesses (X1/2 in cm) of 12 concrete mixtures at different energies ..............19

Table 5.1. The interactions of gamma rays with cement paste and aggregate....................44

Table 5.2. The interactions of neutrons with cement paste and aggregate..........................44

Table 5.3. Deterioration of concrete due to neutron radiation..............................................44

Table 5.4. The neutron fluence level in each step in the test...............................................48

Table 5.5. The experimental conditions found in papers referenced. ..................................50

Table 5.6. Strength and stiffness of concrete samples........................................................52

Table 5.7. Attenuation coefficient percentages of tested concrete at different 

temperatures......................................................................................................61

Table 6.1. Swelling capacity and swelling pressure of ASR gel. .........................................66

Table 8.1. Nondestructive test methods to determine structural properties and 

assess conditions of concrete ............................................................................84

Table 8.2. Nondestructive test methods for determining material properties of 

hardened concrete in existing structures............................................................85

Table 8.3. Listing of possible nondestructive approaches for use in assessment of 

fire-damaged concrete .......................................................................................88

xv

EXECUTIVE SUMMARY

This document summarizes the current understanding of the effects of irradiation on concrete 

materials. Much of the research related to this topic was conducted from the 1960s into 1970s in 

support of development of prestressed concrete reactor vessels for high-temperature gascooled reactors and radioactive waste storage facilities. Concrete structures used as biological 

shields and as support for the reactor pressure vessel in current LWR plants are subjected to 

two types of radiation, gamma and neutron.

Gamma radiation refers to electromagnetic radiation of high frequency; high energy per photon. 

Gamma rays typically have energies above 100 keV and wavelengths less than 10 picometers. 

Gamma rays have an insignificant effect on solid materials composed of ionic and metallic 

bonds, but they can result in the destruction of anisotropic chemical bonds such as covalent 

bonds. Water in the concrete can be decomposed by gamma rays by a process called radiolysis 

and can be converted to hydrogen, oxygen, and hydrogen peroxide. Water can also be removed 

from the concrete by evaporation due to heat generated by gamma radiation. Because most of 

the water in concrete is contained in the cement paste, gamma radiation has a greater effect on 

the cement paste than it has on the aggregate materials.

In the instance of neutron irradiation, the neutrons do not interact with the electrons but with the 

nuclei of atoms. Because neutrons interact with the nuclei of atoms, the lattice spacing within 

the material may change after the collisions. Therefore, the neutrons have a more significant 

effect on dense and well-crystallized materials (e.g., aggregates) than on randomly structured

materials with high porosity (e.g., cement paste). Interactions between nuclear radiation and the 

internal structure of concrete produce geometric changes of the structure resulting from 

displacements of atoms from their lattice sites and phase transformations of the concrete 

constituents resulting in a reduction in porosity and/or formation of microcracks. These 

interactions result in changes in the various concrete properties.

A summary of the effects of neutron and gamma radiation on the mechanical and physical 

properties of concrete is provided. Information is presented on thermal effects due to gamma 

heating of the concrete. Coupling effects of mechanical loading, thermal effect, moisture 

content, and radiation are described, and an example provided using a computational model to 

indicate radiation damage in a shield for a next-generation nuclear facility. Nondestructive 

testing methods are identified for assessment of the condition of reinforced concrete structures.

Radiation limits for neutron and gamma exposure are listed. Finally, potential gaps are identified 

to (1) expand the very limited database on the effect of neutron and gamma radiation on 

concrete’s microstructure and mechanical and physical properties, (2) conduct calculations to 

establish the neutron and gamma fields that develop in concrete shields and support structures 

after various operating periods, and (3) quantify the interaction of temperature and irradiation 

effects.

xvii

ACKNOWLEDGMENTS

The authors would like to acknowledge the continuing support and guidance throughout the 

program provided by U.S. Nuclear Regulatory Commission Technical Monitor, Mr. Herman L. 

Graves, III.

The authors also would like to acknowledge Mr. Walter Koncinski of Oak Ridge National 

Laboratory for his detailed review and many contributions to the report.

The authors wish to take this opportunity to recognize the fundamental contributions of our 

former colleague and friend Professor Hubert Hilsdorf towards understanding radiation effects 

on properties of concrete.

xix

ABBREVIATED TERMS

AASHTO American Association of State Highway and Transportation Officials

AC alternating current

ACI American Concrete Institute

ASME American Society of Mechanical Engineers

ANSI American National Standards Institute

ASR alkali-silica reaction

ASTM American Society for Testing and Materials

BS British Standard

BWR boiling-water reactor

CDP combined damage plasticity

CH calcium hydroxide

C-S-H calcium silicate hydrate

DP Drucker-Prager

EPRI Electric Power Research Institute

ESEM environmental scanning electron microscopy

HCP hardened cement paste

HVT half-value thickness

INFN Italian Institute for Nuclear Physics

LET linear energy transfer

LWR light-water reactor

NDT nondestructive test

NPP nuclear power plant

OPC ordinary Portland cement

PP polypropylene

PWR pressurized-water reactor

SEM scanning electron microscopy

SHRP Strategic Highway Research Program

SPES Selective Production of Exotic Species

w/c water-to-cement ratio

1

1. INTRODUCTION

Compared with other construction materials, concrete has a high shielding capacity against 

nuclear radiation, good long-term durability, and good high-temperature resistance, all at 

relatively low construction costs. Therefore, concrete is widely used for primary and secondary 

containment structures of nuclear power plants (NPPs) exposed to radiation and elevated 

temperatures. Besides being a basic infrastructure material, concrete is used to build shielding 

structures and interim storage facilities for spent nuclear fuel. Radiation and temperature effects 

on the properties of concrete depend on a range of variables, such as the intensity of the 

radiation field, the temperature level, and the period of exposure. In addition to the separate 

effects of radiation and temperature, there is also a one-way coupling effect between the two 

(i.e., a radiation-induced thermal effect).

There are two types of radiation in NPPs, neutron radiation and gamma rays. Neutron radiation 

is the emission of neutrons; gamma rays are a special form of photon radiation. The two types 

of radiation interact with concrete in different ways. A variety of concrete materials, with different 

types of cement, aggregates, and additives, have been used in NPPs. Some concretes are 

more susceptible to gamma radiation, and others are more susceptible to neutron radiation,

depending on the constituent materials used in the mix.

Elevated temperatures have two different effects on concrete. One involves phase 

transformations of the concrete constituents under high temperatures; the other refers to 

mechanical damage in the form of spalling. Both effects may result in significant deterioration of 

the stiffness, strength, and ductility of concrete and concomitant loss of load-bearing capacity.

Phase transformations depend primarily on the temperature level, but not on the rate of 

temperature changes. Spalling damage occurs under high heating or cooling rates. A moderate 

elevation in temperature can occur in concrete exposed to gamma radiation.

This report summarizes the results of research on the effects of radiation and temperature on 

the material properties of concrete [i.e., mechanical properties, such as strength, stiffness, and 

ductility (energy absorption capacity), and durability properties, such as time- and ratedependent creep and shrinkage and environmental aging]. Basic information on neutron and 

gamma radiation is introduced first. That is followed by a description of the internal structure of 

concrete and by detailed discussions on heating due to radiation and its effects on the 

properties of concrete. The effect of radiation on the properties of concrete and the concomitant 

coupling effect of radiation and temperature are further examined, in addition to testing methods

of the effects of radiation and high temperatures on concrete structures. Radiation limits for 

neutron and gamma exposure of NPP structures are identified. Areas for additional research are 

noted. 

3

2. GAMMA RAY AND NEUTRON ATTENUATION

Nuclear reactions are driven by the transformation of atoms; typically, atoms of one element are 

converted into atoms of another element. This process is different from chemical reactions. 

Chemical reactions usually involve electrons being shared, or transferred, to form compounds.

Instead of the electrons taking part in the reaction, nuclear reactions are characterized by the 

atom’s nuclei undergoing changes. Nuclear reactions yield large changes in energy and 

measurable changes in mass (Silberberg 2007). In addition, when an atom transforms, it emits 

radiation. The magnitude of the released energy is equivalent to the energy difference between 

the atom’s initial state and the atom’s final state. 

Radioactivity of nuclei is associated with unstable nuclides, which transform to a more stable 

form by emitting radiation. The stability of a nucleus is related to the ratio of N/Z, where N is the 

neutron number and Z is the proton number. The sum of N and Z equals the mass number A (A 

= N + Z). The N/Z ratio of stable light nuclei (A < 20) is close to 1.0; the ratio for heavy nuclei 

such as uranium-235 can reach 1.554 (N = 143, Z = 92, and A = 235). The N/Z ratio of a stable 

nuclide can be changed into an unstable nuclide artificially by bombarding stable nuclides with 

neutrons or high-energy particles produced by nuclear reactors.

Irradiation by charged particles such as electrons and protons can lead directly to a gain or a 

loss of charge of an atom, which is called direct ionizing radiation (ionization is the process of an 

atom gaining or losing electrons). Irradiation by noncharged particles such as neutrons and 

photons does not lead directly to a change in charge of an atom, and thus is called indirect 

ionizing radiation. Although uncharged particles do not cause direct ionization of atoms, they do 

release charged particles (electrons or protons) as a consequence of their interaction with 

matter and will subsequently cause direct ionization of atoms. Therefore, both direct and indirect 

ionizing radiation can cause ionization.

The literature distinguishes between radiation that is electrically charged and radiation that is 

electrically neutral. The electrically charged radiation includes alpha and beta particles. Alpha 

rays are high-speed, high-energy particles expelled from unstable nuclei; each alpha particle is

the nucleus of a helium atom, which is made up of two protons and two neutrons. Alpha 

radiation is a weak form of direct ionizing radiation that is readily stopped by a sheet of paper.

Beta particles, which are much lighter, are given off by the nucleus of an unstable isotope to 

restore an energy balance. The beta particles are high-energy, high-speed electrons or 

positrons (a positron has an electric charge of +1e, spin of 1⁄2, and has the same mass as an 

electron; an electron has an electric charge of -1e and a spin of 1⁄2). Beta radiation is also a 

direct ionizing radiation and is stronger than alpha radiation. It can be halted by a thin lead plate.

Gamma rays and neutrons are electrically neutral. Gamma rays are energetic photons or light 

waves in the same electromagnetic family as light and x rays. Photons are considered gamma

radiation when they possess an energy level that is higher than 0.1 MeV. Gamma radiation, 

which is much stronger than alpha or beta radiation, is an indirect ionizing radiation and is 

biologically hazardous. X rays are also electromagnetic radiation, but their energy level is lower 

than that of gamma rays. X rays are also distinguished from gamma rays in that x rays are 

emitted by electrons whereas gamma rays are emitted by the atomic nucleus. Neutrons are also 

electrically neutral particles, and thus, when emitted by an atomic nucleus, are also classified as 

an indirect form of ionizing radiation.

In direct ionizing irradiation, the charged particles interact strongly with the materials in a 

radiation shield and are attenuated by the shield. In the case of alpha ray irradiation, a positively 
4

charged helium atom travels through a shielding material, gains two negatively charged 

electrons, and becomes a neutral helium atom. In indirect ionizing irradiation, noncharged 

particles such as neutrons and gamma photons interact less strongly with shielding materials. In 

general, the indirect ionizing radiation is more hazardous than the direct ionizing radiation. In the 

following two sections, the interactions of photons and neutrons with shielding materials will be 

discussed.

2.1 GAMMA RADIATION

Gamma radiation refers to electromagnetic radiation of high frequency; high energy per photon. 

Gamma rays typically have energies above 100 keV and wave lengths less than 10 picometers. 

The absorbed dose in a small volume element is equal to the energy imparted by ionizing 

radiation to the material in a volume element divided by the mass of the material in the volume 

element. The unit of absorbed dose is a gray having units of joule kg-1

: 1 Gy = 100 rad. Photons 

interact with shielding materials in three different ways: photoelectric effect, Compton scattering, 

and pair production. Each involves the ejection of electrons from atoms, as shown in Figure 2.1.

For the photoelectric effect, a gamma ray is absorbed, resulting in the ejection of an orbital 

electron. For Compton scattering, the gamma ray is not absorbed but deflected (scattered), and 

an electron is also ejected. For pair production, the gamma ray is absorbed, and a pair 

(consisting of a positron and an electron) is ejected.

Figure 2.1. The photoelectric effect, Compton scattering, and pair production of 

gamma radiation. (Kontani et al. 2010)

From a radiation-shielding point of view, the relative importance of the three interactions 

between photons and the shielding material depends on both photon energy and the atomic 

number of the shielding material. According to Kaplan’s review (1989), when the atomic number 

is 20, the photoelectric effect is dominant if the photon energy is from 0.01 to 0.1 MeV, and 

Compton scattering is dominant if the photon energy is from 0.1 to 100 MeV. When the atomic 

number is 60, the photoelectric effect is dominant if the photon energy is from 0.01 to 0.5 MeV, 

Compton scattering is dominant if the photon energy is from 0.5 to 5 MeV, and pair production is 

dominant if the photon energy varies from 5 to 100 MeV. According to Kontani et al. (2010), the 

energy range of gamma rays produced by nuclear reactors varies from 100 KeV to 10 MeV.

Compton scattering is the dominant interaction mechanism in that energy range. The dominant 

ranges of the three types of interactions are illustrated in Figure 2.2. In general, with shielding 

materials that have a high atomic number, the dominant interaction mechanism for photon 

radiation depends on the photon energy level. High photon energy results in pair production, low 

energy causes photoelectric effect, and Compton scattering is dominant in between.
5

Figure 2.2. Relationship among gamma ray energy, atomic 

number, and interaction mechanisms. (Kontani et 

al. 2010)

Compton scattering reduces the energy of a gamma ray by transferring a portion of its energy to 

the ejection of an electron and deflects the gamma photon from its original path. Because they 

are deflected and not totally absorbed during the Compton scattering process, the gamma rays 

can be scattered numerous times before completely passing through the material, and thus a 

buildup of gamma radiation may occur within the shielding material. This effect can actually 

increase the total amount of radiation passing through the shielding and needs to be considered 

when designing a radiation shield.

Because gamma rays are attenuated by interactions with electrons, concrete to be used as a 

gamma-shielding material should be designed to maximize the density of electrons within it. 

This can be done by (1) using a type of cement with higher number of electrons per atom 

content, (2) increasing the density of concrete by better compaction, and (3) using heavier 

aggregate (Callan 1953). The first two methods are difficult to apply in practice because there 

are only a few types of Portland cement available. Also, a concrete mixture can be compacted 

to a point where the coarse aggregate in the mix may become segregated. A more practical 

approach would be to use heavier aggregates, such as iron, iron ore aggregates, and various 

other high-density aggregates in the concrete mixture (Callan 1953; Kaplan1989).

Gamma rays have an insignificant effect on the solid materials composed of ionic and metallic 

bonds, but they can break anisotropic chemical bonds such as a covalent bond. Siliceous 

materials may be degraded by gamma rays because the Si-O bond is a covalent bond (Kontani 

et al. 2010). See the discussion of the alkali-silica reaction (ASR) in Section 6.1.

Gamma rays can reduce the water content in a shielding material such as concrete by two 

mechanisms. The gamma rays can decompose water through a process called radiolysis which 

converts the water to hydrogen, oxygen, and hydrogen peroxide (Bouniol and Aspart, 1998).

Water can also be driven from concrete by evaporation caused by the heat generated by 

gamma-ray irradiation. In concrete, most of the water is contained in the cement paste, not in 

the aggregate, and thus the effect of gamma-ray irradiation on cement paste is more significant 

than its effect on aggregate.

2.2 NEUTRON RADIATION

Neutron radiation interacts with the nuclei of atoms; however, the interaction depends mainly on 

the kinetic energy of the neutrons. The neutrons are classified by their energy into three types: 

thermal neutrons, epithermal neutrons, and fast neutrons. The energy level of thermal neutrons 
6

is less than 1 eV, epithermal neutron energy levels are between 1 eV and 0.1 MeV, and the 

energy level of fast neutrons is greater than 0.1 MeV (Hilsdorf et al. 1978). The epithermal 

neutrons are also referred to as intermediate or resonance neutrons. The amount of neutron 

radiation absorbed is typically expressed by the fluence (Shultis and Faw 2000). Neutron 

fluence is the measure of the number of neutrons n that penetrate a unit area, typically

expressed in terms of neutrons per square centimeter. The rate of neutron radiation is 

measured in units of flux, typically expressed in terms of neutrons per square centimeter per 

second.

When a high-energy (i.e., fast) neutron collides with a nucleus, the neutron loses some energy,

and the nucleus generates scattered gamma rays and ejects a neutron. The ejected neutron 

(compared with the incoming neutron) and the nucleus both change their directions. The 

process is called inelastic scattering. This process may repeat many times until the fast neutron 

becomes an intermediate neutron. When an intermediate neutron collides with a nucleus, the 

nucleus ejects a neutron, and the ejected neutron (compared with the incoming neutron) and 

the nucleus both change their directions. The intermediate neutron does not have enough 

energy to cause the nucleus to release gamma rays. This process is called elastic scattering;

however, the intermediate neutron will lose some energy during the process. The elastic 

scattering process may repeat until an intermediate neutron becomes a thermal neutron. When 

a thermal neutron collides with a nucleus, no further neutron ejection occurs, no gamma rays 

are released, and all the energy is absorbed in the collision. The inelastic and elastic scattering 

processes of neutron radiation are illustrated in Figure 2.3.

Figure 2.3. Inelastic and elastic scattering processes of neutron radiation.

(Kontani et al. 2010)

Increasing the density of a material is desirable for neutron attenuation because this will 

increase the number of atoms per unit volume as well as the probability of the collision between 

neutrons and nuclei. 

Because neutrons interact with nuclei of atoms, the lattice spacing within the material may change 

after the collision of neutrons with nuclei. Therefore, neutrons have a more significant effect on 

dense and well-crystallized materials than on randomly structured materials with high porosity. In 

concrete, aggregates (coarse and fine aggregates) are in a crystallized phase, and cement paste 

is an amorphous phase, and thus, neutron radiation causes more distortion and damage to the 

internal structure of aggregates than to the structure of cement paste. The neutrons collide with 

nuclei in cement paste, but the nuclei do not accumulate enough damage to alter the properties of 
7

the cement paste because of the high porosity and randomly layered internal structure of cement 

paste, which will be described in detail in Chapter 3. 

2.3 RADIATION SHIELDING EFFECTS

For radiation shielding, a higher transfer of energy corresponds to a more effective radiation 

shield, but the radiation also has an effect on the shielding material. Therefore, if a concrete 

biological shield attenuates gamma rays and neutrons effectively, it will be affected to a greater 

extent by the irradiation than a less effective radiation shielding material. This greater absorption 

should be observed as a change in the mechanical and physical properties of the shield 

material.

The attenuation mechanisms of concrete are different with photons and neutrons. The radiation 

shielding for photons can be calculated relatively accurately based on three processes for 

photons: photoelectric effect, Compton scattering, and pair production. The radiation shielding 

for neutrons can be calculated less accurately than that for photons. Kaplan (1989) has an 

extensive review on radiation attenuation of photons and neutrons by concrete. 

The intensity of photons that pass through a shield can be experimentally measured in different 

ways, depending on the experimental setup (see Figure 2.4): narrow-beam attenuation (e.g., 

with photons passing through a collimator) and broad-beam attenuation (i.e., without a 

collimator). The radiation intensities of photons after travelling through a shielding material are 

given by

x I I e−μ = 0 (2.1)

x I I Be−μ = 0 (2.2)

where I0 is the photon intensity at the source, B is the broad-beam buildup factor, μ is the linear 

attenuation coefficient in cm-1

, and x is the depth in centimeters in the shielding structure from 

the surface of the shielding material. The difference of Eq. (2.1) and Eq. (2.2) is the factor B.

Equation (2.2) is for broad-beam attenuation whereas Eq. (2.1) is for narrow-beam attenuation.

In both equations, the attenuation coefficient μ determines the rate of attenuation. It depends on 

the energy of the photons and the radiation resistance of the shielding material.

Figure 2.4. Experimental setups for measuring mass attenuation coefficients. (Yilmaz 

et al. 2011 on the left and Kharita et al. 2011 on the right.)

For neutron irradiation, determination of the reduction in the radiation intensity is more 

complicated than for photon irradiation. This is because there are primary and secondary 
8

neutrons that are produced in the interaction process, and photons are produced from inelastic 

scattering. Therefore, attenuation of both neutrons and photons needs to be considered. In 

general, neutron attenuation can be calculated by the transport theory used for calculation of 

photon attenuation, but the solutions are more complicated and less accurate. The attenuation 

of fast neutrons can be calculated approximately by Eq. (2.3):

x R I I e−Σ = 0 (2.3)

where ∑R is the macroscopic removal cross section in cm-1

. The absorption for thermal neutrons 

is given by Eq. (2.4):

LN x I I e/

0

− = (2.4)

where LN is the diffusion length (in centimeters), which depends on neutron radiation resistance 

of the material.
9

3. CONCRETE AS A NUCLEAR SHIELDING MATERIAL

The internal structure of concrete can be considered as a multiple-scale structure such as 

shown in Figure 3.1. At the mesoscale, concrete can be considered as a two-phase composite 

material with aggregate as inclusions and cement paste as a matrix. At the microscale, the 

internal structure of hydrated cement paste can be divided into several hydration products at 

different scale levels. 

Figure 3.1. Multiple-scale internal structure of concrete (schematic).

The internal structure of aggregates is more or less of a crystalline nature, but the internal 

structure of hydrated cement paste is not. This is important because the response of a 

crystalline material to gamma rays and neutron irradiation can be different from that of a 

noncrystalline material.

3.1 AGGREGATES USED IN CONCRETE FOR SHIELDING STRUCTURES 

Aggregates can be classified in different ways. Aggregates can be divided into coarse and fine 

aggregate, in which the sizes of coarse aggregate particles are larger than 4.75 mm, and the 

sizes of fine aggregate particles are smaller than 4.74 mm. In accordance with mineralogical 

compositions, aggregates can be divided into rock types, such as siliceous aggregates and 

calcareous aggregates, or by density (e.g., normal weight, lightweight, and heavyweight).

Table 3.1 presents examples of aggregates and concrete unit weights for each of the density 

classifications.
10

Table 3.1. Examples of aggregates based on density (Kaplan 1989)

Type 

aggregate

Density

(kg/m3

) Examples

Concrete density 

(kg/m3)

Normal weight 1520–1680 Limestone, granite, sandstone 2400

Light weight < 1120 Volcanic pumice, blast furnace slag, cinder < 1850

Heavy weight > 2080 Magnetite, barytes, ferrophosphorus > 4000

Typically, normal-weight aggregates have been used in the biological shields of pressurized 

water reactors, and heavyweight aggregates have been used in the shields and support 

structures of a few boiling water reactors (BWRs).

Several different types of aggregates have been studied for use in shielding structures, 

including the following (Kaplan 1989).

• Hydrous aggregates. Hydrous aggregates are good materials for radiation shielding 

because of their high hydrogen content, which is beneficial in the attenuation of neutrons. 

The hydrogen in the aggregate comes mainly from chemically bound water. Commonly used 

hydrous aggregates are serpentine, limonite, goethite, and bauxite.

• Heavy aggregates. The density of the shielding material is important for the attenuation of 

photons. Therefore, high-density (heavy) aggregates are used to increase the density of 

concrete in radiation shields and thus reduce the section-thickness requirements to achieve 

a given level of shielding. Typical heavy aggregates can be naturally occurring ores such as 

magnetite, witherite, barytes, ferrophosphorus or synthetics such as iron and steel. In one 

case, where a 3.05 m thickness of ordinary concrete would be required, a thickness of only 

2.4 m of magnetite concrete gave equivalent shielding efficiency (Callan 1953).

• Boron-containing aggregates. Boron-containing aggregates can be used for radiation 

shields because of their large neutron cross section. Neutron attenuation often leads to 

production of secondary “hard” gamma radiation. Elements with a large neutron cross

section will produce “soft” gamma radiation instead, which is less penetrating into the 

shielding structure. Boron aggregates include colemanite, borocalcite, and ferroboron.

Table 3.2 summarizes radiation attenuation characteristics for several of the aggregates 

considered for use in shielding structures. Guidance is available for assessing aggregates for 

specific use in radiation-shielding concretes (Lee et al. 2013).
11

Table 3.2. Properties of aggregates for radiation attenuation (Kaplan, 1989)*

ρ (g/cm3) μ (1/cm) ΣR (1/cm) LN (cm)

Barytes 3.50 0.0945 0.0926 4.72

Magnetite 3.55 0.0959 0.1060 2.67

Limonite-steel 4.54 0.1226 0.1210 1.96

Ferrophosphorus 4.68 0.1264 0.1250 1.93

* ρ is density of aggregate, μ is linear attenuation coefficient, ΣR is effective removal cross section, 

and LN is diffusion length.

3.2 IMPACT OF IRRADIATION ON THE CRYSTALLINE STRUCTURE OF AGGREGATES

Under nuclear irradiation, the atomic structure of some aggregates can be converted from 

crystalline structure to distorted amorphous structure with a decrease in specific gravity and an 

increase in volume (Kontani et al. 2010). Two negative impacts are associated with the 

conversion. One is that the distorted internal structure leads to higher reactivity of the converted 

aggregates, which may make the aggregates more reactive to certain aggressive chemicals. A 

typical example is the ASR of concrete, which is discussed in detail in Section 6. The other 

negative impact is that the increase in volume can cause microcracking of concrete.

The effect of nuclear irradiation on quartz was studied by several researchers (Ichikawa et al.,

2002). Crystalline quartz, or α-quartz, with specific gravity of about 2.65 is converted to distorted 

amorphous quartz with specific gravity of 2.27 under a fluence of 1 × 1020 n/cm2 for fast 

neutrons with an energy level > 0.1 MeV, and under a dose of 1 × 1012 Gy for beta and gamma 

rays. Figure 3.2 shows the density of quartz as a function of fast neutron fluence. Regular 

amorphous quartz can be converted into distorted amorphous quartz by a fluence of 0.2 × 1020

n/cm2 for fast neutrons and a dose of 0.1 × 1012 Gy for beta and gamma rays, which are both 

lower than the values for crystalline quartz. The volume expansion (i.e., density reduction) 

associated with the conversion is quite large (a 14% difference between the two specific 

gravities). The conversion of aggregate usually starts at its surface layer.
12

Figure 3.2. Density of quartz as a function 

of fast neutron fluence.(Bouniol 

2011.)

3.3 CEMENTS AND INTERNAL STRUCTURE OF CEMENT PASTE

The chemical composition of cement paste depends on the type of cement used in construction.

Portland cement is the most commonly used cement for concrete shielding structures. Other 

cements with different chemical compositions are available (Mehta and Monteiro 1993) but have 

not been used in construction of NPP concrete structures.

Mixing water and Portland cement makes cement paste. It is composed of several hydration 

products of Portland cement. For example, the volume of calcium hydroxide (CH) particles is 

about 25% of the total volume of cement paste. CH particles are needle- or plate-shaped 

crystals and have a significant effect on long-term properties of concrete. Calcium silicate 

hydrate (C3H2S3, or simplified as C-S-H) has a nonstoichiometric amorphous structure. The 

crystalline structure of C-S-H in cement paste has not been fully resolved. The volume of C-S-H 

is about 50% of the total volume of cement paste. Ettringite (C6AS3H32) particles are needleshaped crystals. Their formation is associated with large volume expansion. The volume of 

ettringite is about 15% of the total volume of cement paste. The remainder of the cement paste 

volume consists of capillary water and empty capillary pores.

Calcium hydroxide and ettringite are crystalline materials, but C-S-H is not. The atomic structure 

of C-S-H is important because C-S-H is the material that binds all components together and 

produces the strength of concrete. The atomic structure of C-S-H has been studied extensively 

and theoretically (Jennings and Xi 1992). However, until recently, the experimental techniques 

could not provide direct visualization of the detailed structure of C-S-H at the nanometer level.

Light microscopy, scanning electron microscopy (SEM), and environmental SEM provide 

resolution at the micrometer level. Transmission electron microscopy, x-ray diffraction (XRD), 

and atomic force microscopy methods provide more-detailed resolution. Using indirect test 

methods such as adsorption of nitrogen or water, and mercury intrusion porosimetry, average 

structural information of C-S-H has been obtained. Because these properties give no 
13

information as to how the pores are arranged in space, the conclusions about the pore structure 

of C-S-H were drawn by inference. For example, the internal surface areas measured for C-S-H 

are around 200 m2/g based on adsorption isotherms, which is very high. Because of the very 

high internal surface areas measured, some basic nanostructure models were proposed for 

C-S-H. They are layered structures with plate-like solid layers or randomly shaped flexible 

layers. One of them is shown in Figure 3.3, which was proposed by Powers and Brownyard 

(1948); another model, shown in Figure 3.4, was proposed by Feldman and Sereda (1968).

The Powers Model (Powers and Brownyard 1948) is supported by water adsorption test data. 

The model considers that the colloidal particles of C-S-H are made up of two or three layers of 

thin sheets that are bonded together. Bond forces are mainly surface forces. The water within 

the particles or between the layers is interlayer water, and the water on the surface of the 

particles is adsorbed water (see Figure 3.3). The maximum and minimum average distances 

between the layers are 30Å and 4Å, respectively, which can be obtained by analysis of surface 

area based on water adsorption test data. If the water between the layers is removed, it will not 

reenter. This irreversible interlayer water loss corresponds to irreversible shrinkage of cement 

paste. Feldman and Sereda (Feldman and Sereda 1968) suggested that C-S-H is a completely 

irregular array of single layers (see Figure 3.4). Bonding between layers is considered to be 

through solid-solid contacts. The interlayer water can move reversibly in and out of the interlayer 

space. 

The basic layer 

is two or three 

layers of 

particles 

formed plates

Adsorbed water

Interlayer 

water

Figure 3.3. Powers’ model for the internal layered 

structure of C-S-H. (Powers and Brownyard 

1948)
14

The basic layer is one layer of particles 

formed irregular arrays

Figure 3.4. Feldman and Sereda’s model for the internal layered 

structure of C-S-H. (Feldman and Sereda 1968)

Taylor (Taylor 1986, Taylor 1997) reported a model for C-S-H at the molecular level. It 

rationalizes results from thermal gravimetric analysis, density, selected-area electron diffraction, 

and analysis of composition by analytical electron microscopy. The proposed structure is an 

intimate mixture of imperfect Jennite and 1.4 nm tobermorite formed into an imperfect array of 

layers [see Figure 3.5(a)], and the prefix refers to the layer thickness. This structure explains 

many of the chemical characteristics of C-S-H.

(a) (b)

Figure 3.5. (a) Internal structure of 1.4 nm tobermorite for C-S-H

(Taylor 1986, 1997); (b) Detailed molecular structure of CS-H (Pellenq et al. 2009). Yellow and red lines are for Si-O 

bond, green and gray spheres are calcium, and blue and white 

spheres are oxygen and hydrogen atoms of water molecules.
15

One important feature related to the present topic is that when being heated at 55oC, the 1.4 nm 

tobermorite loses interlayer water and undergoes unidimensional lattice shrinkage, giving 

1.1 nm tobermorite (C5S6H5 approximately). Under radiation heating, 55oC is a temperature 

frequently reached. Thus the loss of interlayer water may lead to significant shrinkage of cement 

paste. 

Jennings (1986) proposed that the basic element of C-S-H, called “basic building block,” be

considered to be an 11 nm diam sphere with a specific surface area of 210 m2

/g. The internal 

structure of the element is layered of the type proposed by the Taylor model. The elements are 

assembled into one of the three types of structures, each with its own structure of specific 

porosity. Various material properties can be predicted by varying the proportion of each of the 

three structures formed in the cement paste, which depend on material parameters such as the 

ratio of water to cement. The movement of water is reversible, but associated changes in 

structure are irreversible.

Figure 3.5(b) shows molecular dynamic simulation of C-S-H, including the position and binding 

of water molecules in the interstitial pores (Pellenq et al. 2009). The model can predict the 

changes in structure at different calcium contents, the location and binding energy of impurities, 

and the amount and location of water under various ambient conditions (Jennings and Bullard 

2011).

These structural models were proposed to explain experimental results of adsorption isotherms 

(pore size distribution and surface area), creep, shrinkage, and some transport properties of 

concrete, such as chemical diffusivity, mainly because those properties are related to the 

nanostructure of concrete. On the other hand, strength and stiffness of concrete are mainly 

controlled by microcracks, packing density of solid particles, and the interface bond between 

cement paste and aggregate, not by the nanostructural features. 

Lowinska-Kluge and Piszora (2008) conducted a study on the microstructure of cement paste 

under various doses of gamma radiation. The microscopic images of the cement paste tested 

shown in Figure 3.6 indicate that pseudomorphoses had taken place after a dose of 130 MGy 

and that the microstructure of cement paste was changed significantly by the gamma radiation.

Figure 3.7 shows SEM images of irradiated cement paste with silica fume as an additive.

Pseudomorphs were observed at a dose of 290 MGy. With increasing dose, microcracks 

appeared in the sample. At a dose of 1188 MGy, a completely amorphous cement matrix and 

microcracking was produced, as shown in Figure 3.7(4). Figure 3.8 shows SEM images of 

irradiated cement paste with fly ash as an additive. Pseudomorphs appeared at a dose of 

290 MGy, which is similar to the results for the cement paste made with a silica fume additive.

Bubbles observed in Figure 3.8(3) are the effect of separation of chemically bonded water. With 

increasing dose of radiation, microcracks appeared, and the paste was completely destroyed at 

a dose of 1188 MGy, as shown in Figure 3.8(4).

These studies indicate that radiation affects the microstructure of cement paste. The changes in 

microstructure can impact the effectiveness of the concrete as a shielding material. Additional 

research is recommended to increase the understanding of the effect of neutron and gamma 

irradiation on the nanostructure of concrete.
16

Figure 3.6. SEM images of cement paste irradiated with gamma radiation. 

(1) 0 MGy, (2) 130 MGy, (3) 290 MGy, and (4) 836 MGy. (Lowinska-Kluge and 

Piszora 2008). Note the change in scale between (1) and (2), (3), and (4).
17

Figure 3.7. SEM images of cement paste that was made with a 

silica fume additive and irradiated with gamma 

radiation. (1) 0 MGy, (2) 290 MGy, (3) 466 MGy, and 

(4) 1188 MGy. (Lowinska-Kluge and Piszora 2008). Note 

the change in scale between (1) and (2), (3), and (4).
18

Figure 3.8. SEM images of cement paste that was made with a fly 

ash additive and irradiated with gamma radiation. 

(1) 0 MGy, (2) 290 MGy, (3) 466 MGy, and (4) 1188 MGy. (LowinskaKluge and Piszora 2008). Note the change in scale 

between (1) and (2), (3), and (4).

3.4 RADIATION SHIELDING PROPERTIES OF CONCRETE

Concrete is a composite material with aggregates and a cement paste binder as components.

Shielding properties of concrete depend on the shielding properties of its constituents. The linear 

attenuation coefficient for gamma rays in Eq. (2.1) can be written independently of mass density:

( ) x ( )d I I e I e − μ ρ ρ − μ ρ = 0 = 0 (3.1)

where ρ is the density of the shielding material (g/cm3

) and d is density times distance traveled

(g/cm2

). The mass attenuation coefficient, μ/ρ, of a composite material can be written in terms of 

its components:

∑ ( ) =

=

N

i

wi i

1

μ ρ μ ρ (3.2)

where

i is the ith constituent,

N is the total number of constituents,

wi is the weight fraction of the ith constituent, 

(μ/ρ)i is the mass attenuation coefficient of the ith constituent.
19

Another parameter that is related to the linear attenuation coefficient μ, called the half-value 

thickness, HVT, is the depth of shielding where the intensity of the primary photon beam is 

reduced by 50%. [HVT is different from half-life, which is defined as a time interval during which 

activity of the radionuclide decreases to one half of the initial value (Hala and Navratil 2003)]:

μ

ln 2 HVT = X1/ 2 = (3.3)

Research on the radiation-shielding properties of concrete has been extensive. Table 3.3 lists 

some of the test data on linear attenuation coefficients (Abdo et al. 2002, Kharita et al. 2008,

Yilmaz et al. 2011, Sahin et al. 2011). Theoretical values for the mass attenuation coefficients 

(μ/ρ) of concretes can also be calculated by analytical models (Schmidt 1969, Bashter 1997) 

and numerical programs, such as WinXCom (Gerward et al. 2004) and the NXcom (El-Khayatt 

2011). One can see from Table 3.3 that the linear attenuation coefficients decrease through an 

increase in energy and that the concrete mix affects the HVT.

Similarly, the macroscopic removal cross section ∑R for neutron radiation can also be related to 

the properties of the constituents:

∑ ( ) =

Σ = Σ

N

i

R wi R i

1

ρ ρ (3.4)

Because of the interactions between concrete and neutrons during the operation of a nuclear 

reactor, many elements in shielding concrete become radioactive. The activity of the affected 

shielding concrete should be measured after decommissioning of the reactor (Alhajali et al. 2009).

Table 3.3. Measured linear attenuation coefficients μ (1/cm) and half-value thicknesses 

(X1/2 in cm) of 12 concrete mixtures at different energies (Yilmaz et al. 2011)

Concrete

Linear attenuation coefficients μ (cm-1) and HTV (cm)

59.5 keV 661 keV

μExp.

(χ1/2)Exp.

μExp.

(χ1/2)Exp.

1st method 2nd method 1st method 2nd method

M0 0.593 1.168±0.110 1.053±0.022 0.175 3.97±0.372 3.174±0.076

MS2.5 0.450 1.540±0.147 1.045±0.024 0.198 3.50±3.343 3.647±0.080

MS5 0.389 1.781±0.178 1.237±0.027 0.202 3.430±0.346 3.678±0.088

MS10 0.383 1.809±0.179 1.301±0.031 0.179 3.871±0.375 3.673±0.095

MF5 0.380 1.823±0.178 1.322±0.033 0.182 3.807±0.350 3.795±0.087

MF10 0.328 2.112±0.215 1.430±0.031 0.171 4.052±0.409 3.749±0.101

MF20 0.296 0.341±0.243 1.573±0.040 0.169 4.100±0.422 4.285±0.102

MF30 0.272 2.547±2.241 1.64±0.045 0.161 4.304±0.421 4.125±0.094

MB5 0.302 2.294±0.222 1.643±0.044 0.157 4.414±0.428 4.293±0.107

MB10 0.298 2.325±0.239 1.613±0.041 0.172 4.029±0.382 4.208±0.088

MB20 0.319 2.172±0.204 1.488±0.034 0.173 4.005±0.408 3.250±0.084

MB30 0.275 2.520±0.252 1.593±0.039 0.171 4.052±0.393 3.611±0.090
20

Several studies were conducted to calculate the shielding capacity of concrete slabs made of 

different aggregates. Tables were prepared for nuclear engineering applications. The tables 

give the dose equivalent per unit incident fluence as a function of the mass thickness of the 

concrete slabs (Wyckoff and Chilton 1973, Shultis and Faw 2000).

3.5 THE STATES OF WATER IN CONCRETE AND THE FUNCTION OF WATER FOR 

RADIATION SHIELDING

The water content in cement paste is important to the effect of radiation on the properties of 

concrete. States of water in cement paste can be divided into free water, adsorbed water, and 

chemically combined water. Free water is water available to react with the cement and is found 

in air voids and capillary pores. Free water can evaporate with its rate dependent on the relative

humidity, temperature, and airflow over the concrete surface. When represented by adsorption 

isotherms, free water is also called capillary water, as shown in Figure 3.9. Adsorbed water 

consists of water molecules that are physically adsorbed onto the surface of solids in the 

hydrated cement paste (e.g., gel pores and interlayer space). The gel water is lost at relatively 

low relative humidity levels (e.g., 30%) and causes shrinkage of cement paste whereas water 

associated with the interlayers (see Figure 3.3 and Figure 3.4) is evaporable only under high 

vacuum. They are shown as gel water in Figure 3.9.

Chemically combined water is the water reacted with Portland cement to form CH, C-S-H, and 

ettringite. This type of water, shown as “nonevaporable water” in Figure 3.9, is not evaporable in

dry air and or under high vacuum (i.e., it is unaffected by relative humidity). It can be released at 

high temperature, which is called dehydration.

Figure 3.9. Relationship between equilibrium water 

content of hardened cement paste and 

relative vapor pressure. (Hilsdorf 1967)
21

The change of water content in concrete is important for radiation shielding because the loss of 

adsorbed water and interlayer water, by change of temperature and/or relative humidity, is one 

of the causes for drying shrinkage, which can result in shrinkage cracking. The loss of adsorbed 

water by a change of loading is one of the causes for creep; the loss of chemically combined 

water through dehydration reduces both the strength and stiffness of concrete.

The energy transfer during the collision of a neutron and a nucleus is proportional to the mass of 

the neutron and the mass of the nucleus. If the mass of the nucleus is similar to the mass of the 

neutron, larger energy transfer will occur. Therefore, elements with nuclei that have a mass

closer to the neutron mass will attenuate the neutron radiation more effectively. Hydrogen and 

oxygen have nucleus masses that are similar to neutron mass and are therefore desirable in 

terms of neutron radiation shielding. Hydrogen and oxygen are the elements found in water, so 

a greater proportion of water in the concrete mixture will improve its ability to shield neutron 

radiation. Specifically, the mass of a neutron is 1.008 u whereas the mass of a hydrogen 

nucleus consisting of one proton is 1.007 u (where u is the atomic mass unit, u = 1.66 × 10–27

kg) (Kaplan 1989). The mass of one Oxygen-16 atom is 15.995 u.

Most of the hydrogen in concrete comes from water. To be conservative, it was assumed that 

only the chemically bound water is considered to remain in the concrete. The total amount of 

chemically bound water can be estimated by 

Wb = 0.24 m C (3.5)

where Wb is the weight of bound water, C is the total weight of cement, and m is a maturity 

factor. The maturity factor is determined by the conditions (temperature and relative humidity) 

under which the concrete was cured. If the conditions are reasonable, the maturity factor is 0.8 

(ANSI/ANS 2006a). If the conditions are less than optimal, a lower maturity factor is used.

Hilsdorf (1967) developed a model to calculate the water content in hardened concrete made 

from Portland cement cured at different levels of relative humidity.

The description above implies that more water in cured concrete can provide better radiation 

shielding. In order to find out whether this is true, Kharita et al. (2010) conducted a systematic 

experimental study on 75 concrete samples with various water/cement ratios (w/c), by weight, 

and found that w/c has no significant effect on shielding properties of concrete under roomtemperature conditions. The test data are shown in Figure 3.10. However, using w/c as an 

experimental parameter resulted in other major changes in the concretes. The concrete strength

was reduced by increasing w/c, as shown in Figure 3.11. Therefore, the effect of water in 

concrete on radiation shielding properties is actually quite complicated. On one hand, increase 

in w/c reduces the strength of concrete because the porosity of concrete increases. As a result, 

the shielding properties of concrete with high w/c should have decreased. On the other hand, 

the increased amount of water helped to improve the shielding properties. The combined effect 

of the two opposing trends may result in the insignificant variation of shielding properties of 

concrete over a large range of w/c. Thus the actual water content of the concrete, not the w/c,

will be a better experimental parameter for future studies of the effect of water on radiation 

shielding properties of concrete.
22

(a) (b)

Figure 3.10. Changes in the linear attenuation coefficients of concrete versus the 

changes in w/c (a) for a broad beam of gamma radiation from a Cs-137 

source and (b) for a broad beam of neutron radiation from an Am–Be 

source. (Kharita et al. 2010)

Figure 3.11. Changes in strength of concrete 

versus the changes in w/c.

(Kharita et al. 2010)

3.6 THE EFFECT OF POLYMER FIBERS

It is well known that fibers can improve mechanical properties of concrete, including 

compressive and tensile strength and post-peak ductility, but very few studies have investigated 

the effect of nuclear radiation on fiber-reinforced concrete. A recent experimental study by 

Martinez-Barrera et al. (2011) showed indirectly the effect of gamma radiation on the 

mechanical properties of concrete reinforced with polypropylene (PP) fiber. The concrete 

specimens were exposed to different gamma radiation doses (5, 10, 50, 100, and 150 kGy) in 

air at room temperature with a dose rate of 3.5 kGy/h. Then, strength and stiffness tests were 

performed on the irradiated concrete samples.

As shown in Figure 3.12(a) and (b), gamma irradiation can impact the compressive strength and 

the compressive modulus of PP-fiber-reinforced concrete. Depending on the gamma dose,
23

(a) (b)

Figure 3.12. (a) Compressive strength data with various radiation doses and (b) modulus 

of elasticity test data with various radiation doses. (Martinez-Barrera et al. 

2011.)

gamma radiation can increase or decrease mechanical properties of concrete that has been 

reinforced with polymer fibers. Based on the above test data, the mechanical properties of PPfiber-reinforced concrete with fiber contents of < 1.5%, by volume, improve with increasing fiber 

content and increasing dose up to 50 kGy. The improvement is considered to be due to the 

ionizing energy generating more surface contacts between the fibers and the hydrated cement 

phase. However, at higher doses (100 and 150 kGy), the mechanical properties are decreased.

3.7 DESIGN OF CONCRETE RADIATION SHIELDING

Design of concrete radiation shielding for NPP structures (e.g., nuclear reactors, hot cells, 

radiochemical plants, nuclear fuel fabrication facilities) is an involved process. Basic design 

procedures have been noted to involve several basic steps (Kaplan 1989):

1. Determine the type and intensity of the radiation source to be shielded. This type of 

information is normally provided by the designers or users of the nuclear installation for 

which the shielding is required.

2. Decide the maximum level of radiation that can be permitted at the external surface of the 

shield. This is generally based on recommendations of the regulatory authority (e.g., the 

International Commission on Radiological Protection).

3. Determine the required attenuation or reduction factor resulting from (1) and (2).

4. Consider the potential choices and proportioning of materials that could be used in the 

concrete mix to achieve the desired results.

5. Determine the nuclear shielding properties of various concretes under consideration (e.g., 

estimate the thickness required to achieve the required photon and neutron attenuation).
24

6. Determine the temperature distribution in the concrete shield so that thermal stresses may 

be calculated to establish whether they are within acceptable limits.

7. Decide, based on the above, the type of concrete that would be most suitable for the shield 

and develop a detailed design.

A more-detailed discussion related to each of these basic steps is provided in the source 

reference (Kaplan 1989). Additional information on concrete radiation shields for NPPs related 

to specifications, analysis, design, and testing is available (e.g., ANSI/ANS 1987, 1999, 2006a, 

2006b; Jaeger 1968, 1975; Shultis and Faw 2000; USNRC 2009).
25

4. HEAT OF RADIATION AND THE THERMAL STRESS DUE TO 

RADIATION

Irradiation of concrete by neutron and gamma radiation can lead to an increase in temperature 

of up 250°C (Fillmore 2004). Such a large increase in temperature can have significant impacts 

on the mechanical and radiation-shielding properties of concrete. Granata and Montagnint 

(1972) and Acevedo and Serrato (2010) reported that concrete samples were heavily damaged 

when heated to 280°C (a temperature commonly reached in heavy-water reactors) and at a 

fluence of 1020 n/cm2

.

In general, high temperature causes two forms of damage in concrete. One is the deterioration 

in mechanical properties of concrete, such as physico-chemical changes of the cement paste 

and aggregate (so-called phase transformations), and thermal incompatibility between the 

aggregate and the cement paste. The other one is spalling of concrete. Spalling can result in 

severe reduction of cross-sectional area, leading to the exposure of the interior concrete to 

excessive temperatures, which in turn causes further spalling damage, called progressive 

spalling damage. There are several theories explaining the spalling mechanisms of concrete 

under high temperature, which will not be discussed here (See Willam et al., 2009).

Experimental results have shown that spalling damage occurs under a high heating rate in the 

temperature range of 200°C to 350°C. In some of the tunnel fires in Europe, the heating rate 

reached 100°C/min to 200°C/min, and the maximum temperature could have reached above

1000°C in a few minutes (CEMENBUREAU, 2004). Spalling of concrete typically occurs near 

the concrete surface, but may occur at depths up to 50% its cross section thickness. Concrete 

spalling is not expected to occur as a result of heat generated by radiation.

The heat of radiation affects properties of concrete at two levels: the structural level and the 

microstructural level. At the structural level, the thermal gradient due to the heat of radiation 

results in thermal stress, which may be high enough to create damage in concrete. At the 

microstructural level, the mismatch of thermal strains in cement paste and in aggregate 

responding to the heat of radiation may cause a large stress at the interface between the 

aggregate and cement paste, which may cause microcracking in the cement paste. 

4.1 HEAT GENERATED BY GAMMA AND NEUTRON IRRADIATION

Radiation energy is converted to heat when absorbed by a shield. The heat generated by 

irradiation can reach quite high temperatures, depending on the shielding material used and the 

configuration of the shielding structure. Hilsdorf et al. (1978) noted a temperature as high as 

250°C was developed in some of the investigations he summarized, which is high enough to 

generate considerable damage in shielding materials. Therefore, the heat generated by gamma 

and neutron radiation is very important in that it can impact the functional and performance 

requirements of the shield.

In the case of photon radiation, kinetic energy is transferred from the photons to electrons in the 

shielding material. This energy is then transformed to heat. Equation (4.1) can be used to 

calculate the heat generated by photons:

( ) ( ) ( ) P P P P P H r = c E E ,r Φ E ,r dE ∫ μ (4.1)

where

H(r) is the rate of heat generation at spatial position r in W cm-3

, 
26

EP is photon energy in MeV, r is the position coordinate in cm, 

μp(EP, r) is the energy deposition coefficient for energy deposition in cm-1

, 

Φ(EP, r) is the photon flux spectrum at energy EP in photons cm2 s-1 MeV-1

, 

c is the conversion factor, c = 1.6 × 10-13 W s MeV-1

.

At a given point in space, Eq. (4.1) is an integration over the entire photon energy spectrum. For 

gamma irradiation, Eq. (4.2) can be used to calculate the approximate volumetric heating rate 

due to gamma photons at any point x:

( ) ( ) γ Eγ μγ γ Eγ H x ≅ × Φ −13 1.6 10 (4.2)

where 

Hγ(x) is the heating rate at point x in W cm-3

, 

x is the coordinate or depth in a shielding material in cm, 

Eγ is the gamma ray energy in MeV, 

μγ is the linear energy-absorption coefficient for gamma rays of energy Eγ in cm-1

, 

Φγ is the gamma ray flux density at energy Eγ in photons cm-2 s-1

, 

Eγ is the gamma ray energy in MeV.

Unlike Eq. (4.1), no integration is involved in Eq. (4.2). Equation (4.2) can be further simplified 

Hγ ≅ × Eave Φγin

−13 1.6 10 (4.3)

where

Hγ is the heating rate per unit area (W cm-2

), 

Eave is the average energy of the photons in MeV, 

Φγin is the incident photon flux density in photons cm-3 s-1

.

In the case of neutron irradiation, the heat generated by the process can be divided into direct 

and indirect heat. For elastic scattering, no gamma ray is generated, and the heat generated by 

the interaction of neutrons and nuclei is deposited at the site of interaction. For inelastic 

scattering, the heat generated by the interaction is deposited on site (direct heat), but the 

gamma rays produced by the interaction deposit their energies at various distances from the 

site of reaction (indirect heat). The approximate volumetric heating rate at a point x due to 

neutrons may be conservatively estimated by

( ) ( ) ( ) N c N N EB H x ≅ × Σ E Φ x −13 1.6 10 (4.4)

where 

HN is the approximate volumetric heating rate at point x in W cm-3

, 

Σc(EN) is the macroscopic capture cross section for neutrons of energy EN in cm-1

,

ΦN(x) is the neutron flux density at a point x in neutrons cm-2 s-1

,

EB is the binding energy for capture reaction in MeV.

Equation (4.4) can be simplified to provide an estimate of the approximate heating per unit area 

for neutrons:

HN ≅ × EKB ΦNin

−13 1.6 10 (4.5)

where

HN is the approximate heating per unit area for neutrons in W cm-2

, 
27

EKB is the sum of average kinetic energy associated with a neutron and the average binding 

energy associated with neutron capture in MeV, 

ΦNin is the incident neutron flux density in neutrons cm-2 s-1

.

Abdo and Amin (2001) conducted an experimental study to compare the estimations by above 

equations to measured data. Figure 4.1(a) is a comparison of the measured and calculated heat 

generated in concrete due to total thermal neutrons, in which the calculation was based on an 

equation similar to Eq. (4.4) (Yevick 1966; Jaeger 1968; ANSI/ANS 2006a). The basic trend is 

that both total thermal neutron flux and heat generation are linear functions of the shielding

depth. The heat generation due to capture of total thermal neutrons, which can reach about 6.6

× 10-3 mW cm-3 at a thickness of 50 cm, generates a temperature rise of 1.7°C. 

(a) (b)

Figure 4.1. Measured and calculated (a) heat generation due to the total 

thermal neutrons and (b) the calculated heat generation due to 

total thermal and reactor thermal neutrons in ordinary concrete

(Abdo and Amin 2001).

Figure 4.1(b) shows the heat generation due to both the total thermal neutrons and the reactor 

thermal neutrons along the beam direction (Z-direction) and normal to it (R-direction). The figure 

shows a marked depression in the heat generation due to the reactor thermal neutrons in both 

Z- and R-directions as compared with that due to the total thermal neutrons. This depression in 

the heat generation reaches about 30% at a shield thickness of 10 cm in both Z- and Rdirections. The depression in the heat generation reaches about 60% at 40 and 10 cm in the Rand Z-directions, respectively. The results show that the heat generation due to total thermal 

neutrons provides the main contribution to the heat generation, especially in the first layers of 

the shield compared with reactor thermal neutrons. However, at deep penetrations the 

contribution of the total thermal neutrons to the heat generation is comparatively small (Abdo 

and Amin 2001).

4.2 TEMPERATURE VARIATION DUE TO THE HEAT OF RADIATION

After the heat generated by the irradiation is determined, the temperature field in a shielding 

structure can be evaluated by considering the heat of radiation as a heat source in a thermal 
28

analysis for the shielding structure. Because the irradiation effect is a long-term effect and 

shielding structures are usually very large (such as a thick wall of a biological shield), a onedimensional steady-state thermal conduction equation with a heat source can be used for 

evaluating temperature distribution due to irradiation by gamma rays. For example,

x λ

P E I e

x

T k − = − ∂

∂

2 0

2

(4.6)

where

T is temperature distribution in °C, 

k is the thermal conductivity of concrete in W cm-1 °C-1

, 

x is the depth into the shielding structure in cm, 

Ι 0 and EP have the same meaning as in Eq. (2.1) and Eq. (4.1),

λ is the relaxation length in cm, and is equal to

μ

λ 1 = for gamma radiation or ΣR

= 1 λ for neutron radiation (4.7)

where 

μ is defined in Eq. (2.1)

ΣR is defined in Eq. (2.2).

The right-hand side of Eq. (4.6) is the heat source due to the heat produced by photon radiation.

A solution of temperature distribution can be obtained by integrating Eq. (4.6) twice:

( ) e Cx D

k

E I T x P x = − + +

λ 0 − λ

2

(4.8)

where C and D are two integration constants which can be determined based on boundary 

conditions of the shield structure such as the inner and outer surface temperatures of a concrete 

wall of a biological shield. For radiation attenuation purposes, the thickness of a shield structure 

is usually much larger than the relaxation length λ. When the radiation is applied only on one 

side of the structure (at x = 0), the temperature cannot vary linearly with the depth because the 

temperature rise due to the radiation in the deep part of the shield must be zero, and thus C

must be 0 in Eq. (4.8), and Eq. (4.8) becomes 

( ) e D

k

E I T x P x = − +

λ 0 − λ

2

(4.9)

Using the boundary condition at the surface x = 0, T(0) = 0 (no temperature rise on the surface 

of the shielding structure due to the radiation), we have D = λ2

EPI0/k, and then

( ) ( ) λ P x λ

e

k

E I T x − = 1− 0

2

(4.10)

Because of the boundary condition used in Eq. (4.9), T(0) = 0, T(x) in Eq. (4.10) should be the 

temperature rise due to radiation, ΔT(x):

( ) ( ) λ P x λ

e

k

E I T x − ∆ = 1− 0

2

(4.11)

The thermal stress in the shielding structure depends on the temperature gradient and, in 

absence of kinematic restraints, it does not depend on uniform temperature rise in the material.

The temperature gradient can be obtained by taking a derivative of Eq. (4.10) with respect to x:
29

λ P 0 x / λ

e

k

E I

dx

dT − = (4.12)

It is obvious that at the surface x = 0, the temperature gradient reaches the maximum, which 

means that the thermal stress due to the radiation reaches the maximum on the surface of the 

shielding structure.

Abdo and Amin (2001) conducted an experimental study to compare the temperature estimates 

with the measured temperature data. Figure 4.2 shows a comparison of the measured and 

calculated temperature rise generated by gamma rays, in which the temperature calculation was 

based on Eq. (4.11) (Yevick 1966, Jaeger 1975, Kaplan 1989). Figure 4.2 shows the 

temperature rise in ordinary concrete because of gamma rays due to capture of the total thermal 

neutrons (measured and calculated) and that due to reactor thermal neutrons. The figure shows 

that the temperature rise has its maximum values in the first layer of the shield thickness and 

that it decreases rapidly at deep depths in the shield. The figure shows also that the maximum 

evaluated temperature rise is about 1.7 oC, which has no effect on dehydration of the concrete.

Figure 4.2. Temperature rise in ordinary concrete because of 

gamma rays due to capture of thermal neutrons

(Abdo and Amin 2001).
30

4.3 MACROSCOPIC THERMAL STRESS IN SHIELDING STRUCTURES DUE TO 

RADIATION 

Thermal stresses occur in a shielding material due to the heat of radiation at two different scale 

levels. One, which is due to the temperature gradient as noted in Eq. (4.12),may be called 

macroscopic thermal stress.. The other one, which is due to the differential thermal expansions 

of the constituent phases in the shielding material that occur even under a uniform heating of 

radiation, may be called microscopic thermal stress. The macroscopic thermal stress due to the 

gamma and neutron irradiation is discussed in this section. Microscopic thermal stress is

discussed in the Section 4.4.

With the thermal gradient of Eq. (4.11) and given properties of a shielding wall such as 

coefficient of thermal expansion, modulus of elasticity, and Poisson’s ratio, thermal stress in a

shielding structure can be evaluated. For example, the thermal stress in a large shielding wall 

subjected to irradiation on one side of the wall can be calculated as follows (Kaplan 1989):

( )

ν

α

σ σ −

∆ = +

1

E T

x ave (4.13)

where

σave is the average stress due to radiation heating, 

α is the linear coefficient of thermal expansion of the shielding material,

E is the modulus of elasticity of the shielding material,

ν is Poisson’s ratio of the shielding material,

ΔT is the temperature rise at point x given by Eq. (4.11).

The average stress is given by

( ) 



 



 



 



 − − − − = − λ λ − λ

ν

α λ

σ σ /

2

0

0 1 2 1 2

1

P L

ave e

k L L

E E I

(4.14)

where σ0 is the stress before radiation heating, L is the thickness of shielding wall in cm, and

other parameters having the same meaning as noted previously. Compressive and tensile 

stresses are indicated by positive and negative values, respectively.

The thermal stress calculated this way is the thermal stress induced by the heat of radiation.

The effect of irradiation on transport and mechanical properties of shielding materials must also 

be considered. For example, gamma and neutron radiation can have an impact on the thermal 

conductivity, coefficient of thermal expansion, modulus of elasticity, and Poisson’s ratio. These 

effects will be discussed in the following chapters.

4.4 MICROSCOPIC THERMAL STRESS IN CONCRETE DUE TO RADIATION 

When exposed to neutron irradiation, concrete increases in volume, with the increase mainly 

due to a change in volume of the aggregate in the concrete (Kontani et al. 2010). When 

subjected to fast neutron irradiation, the lattice structure of the minerals in aggregates is 

disturbed, leading to an increase in volume. The crystalline lattice structure present in the 

aggregates is much more affected by neutrons than the more-amorphous lattice structure of the 

cement paste. On the other hand, the volume of cement paste decreases, exhibiting shrinkage 

that is due to the evaporation of pore water under the heat of radiation and due to the phase 

change of water under irradiation. The mismatch of expansion in aggregate and shrinkage of 
31

cement paste may cause damage in the interface between the two phases. A method to deal 

with this type of damage development is described in this section; the studies on the expansion 

and shrinkage mechanisms resulting from irradiation will be described in Chapter 5.

Microscopic thermal stress analysis is more complicated than macroscopic thermal stress 

analysis. The damage development depends on both the size and shape of the aggregates. An 

assumption of spherical aggregate shape is made to simplify the problem. For the effect of 

aggregate size, two different situations may be considered. The first is to consider a special 

situation in which all aggregates are of the same size. Based on the result of this special 

situation, the second situation is to consider a general case in which there is a continuous 

distribution of particle sizes. Only the model for the first situation is briefly described in this 

report. This model was originally developed to simulate the expansion of ice formed in spherical 

pores and the damage development near the pores (Eskandari-Ghadi et al. 2013). The model 

can be used for the damage development in the interface of aggregate and cement paste under 

the heat of radiation.

To evaluate the onset of the damage and the effect of damage due to aggregate expansion on 

the properties of concrete under irradiation, concrete is considered as a two phase material with 

cement paste as matrix and many spherical particles of size rs (the radius of spheres) as 

inclusions. Figure 4.3 is a sketch for a concrete with many different sizes of aggregates. This 

sketch can also be used for concrete with the same size of aggregate, in which all inclusions are 

spheres of the same radius.

Taking any one element out of the concrete will yield Figure 4.4, in which Phase 1 is the 

aggregate, Phase 2 is the damaged cement paste near the aggregate, Phase 3 is the original 

cement paste, and Phase 4 is the effective porous medium (i.e., concrete). Because of the 

volumetric increase of the aggregate due to the heat of radiation and the expansion of lattice 

structure, there is a pressure p generated by the aggregate on the surrounding cement paste, 

as shown in Figure 4.5. The pressure results in a tensile stress in the surrounding cement 

paste, which may cause damage in the cement paste in the vicinity of the aggregate. 

In the case of a spherical aggregate, the distribution of stresses in the cement paste matrix is 

determined as follows (Timoshenko and Goodier 1970):

3

3

R

r

p s σ R = − , 3

3

2R

r

p s σ θ = σ φ = 

(other stresses are zeros) (4.15)

where σR is the radial stress, and σθ and σφ are the two tangential stresses at any point in the 

concrete due to the pressure p on the surface of the spheres. R in Eq. 4.15 and Figure 4.5 is the 

radial coordinate in spherical coordinate system when the center of the aggregate is the origin 

of the coordinate system, σR is a compressive stress, and σθ and σφ are tensile stresses. The 

tensile stresses are equal to p/2 at R = rs, and they decrease with the cube of the distance from 

the center of the sphere. 
32

Figure 4.3. A model for concrete with aggregates of different 

sizes for analyzing the microscopic stress at the 

aggregate–cement paste interface due to the heat 

of radiation (the internal structure of each 

aggregate element is explained in Figure 4.4).
33

Phase 1: Spherical pores

Phase 2: Damaged solid

Phase 3:Original solid

Phase 4: Effective porous medium

Figure 4.4. The four phases in each element in 

Figure 4.3.

Figure 4.5. The interface pressure p induced by the 

expansion of aggregate under radiation.
34

Drucker-Prager (DP) plasticity may be applied to the cement paste in concrete when the 

damage generated by the interface pressure due to the heat of radiation is being considered.

DP plasticity is used here to facilitate the discussion, although any other constitutive theory

suitable for the material under consideration could be applied. With DP plasticity, the following 

equation can be obtained to evaluate plastic behavior of the concrete:

I + J = k α 1 2 , 

c t

c t

f f

f f

+

− = 3

2

α , and 

c t

c t

f f

f f k

+ = 2

3

2 (4.16)

where 

Ι 1 is the first invariant of stress tensor, 

J2 is the second invariant of the deviatoric stress tensor,

fc is the compressive strength of the material,

ft is the tensile strength of the material.

For the stress components given by Eq. (4.16), 1 I is zero and 3 ( 2) 3 3 J 2 = prs R . The 

maximum value of 2 J is p 3 2 at s R = r . Thus if p equals [ ( )] c t c t 4 f f 3 f + f , then there will 

be damage due to plastic behavior at s R = r . If p is larger than [ ( )] c t c t 4 f f 3 f + f , then the 

damaged part will start to grow in the radial direction. Assuming a slow variation of pressure 

with respect to time, which is true for most of the radiation processes, the damage growth is 

usually a small value with respect to the radius of aggregate and it can be evaluated in terms of 

the change of pressure (Eskandari-Ghadi et al. 2013).

The pressure increase is related to the volume expansion of aggregates, which depends on the 

radiation intensity and the duration of concrete irradiation. The pressure increase is also related 

to the shrinkage of cement paste due to a loss of water. Both the expansion of aggregate and 

the shrinkage of cement paste will be discussed in Section 4.5.

4.5 THERMAL EXPANSION AND THERMAL CONDUCTIVITY

Hilsdorf et al. (1978) summarized several experimental studies and concluded that for a neutron 

fluence up to 5 × 1019 n/cm2 there is no significant difference between the coefficient of thermal 

expansion of temperature-exposed samples and that of neutron-irradiated samples. These test 

data imply that the neutron radiation does not generate extra expansion above the thermal 

expansion. Dubrovskii et al. (1966) conducted a test on the expansion of heat-resistant concrete 

with chromite filler as aggregate under a high level of neutron irradiation. The neutron fluence

was 2.0 to 2.4 × 1021 n/cm2

, and temperatures were up to 550oC. Their results are shown in 

Figure 4.6.
35

Figure 4.6. Relative expansion of concrete with chromite filler as aggregate 

under different temperatures: (a) first heating; (b) second heating; 

(1) irradiated; (2) heated; and (3) unheated. (Dubrovskii et al. 1966)

The slopes of the curves in Figure 4.6 represent the coefficient of thermal expansion of 

concrete. The difference between Figure 4.6(a) and Figure 4.6(b) is due to drying shrinkage 

during the first heating. One can see that there are no great differences between the expansions 

of the irradiated and heated concrete samples. The concrete samples used in the tests were 

made of Portland cement and chromite fillers, which is a heat-resistant concrete (or refractory 

concrete). In a follow-up study by Dubrovskii et al. (1967) on concrete samples made with 

Portland cement and sandstone and river sand, they found that the nature of thermal expansion 

for irradiated concrete was changed. As shown in Figure 4.7, the expansion of irradiated 

samples was increased significantly. Some of the samples expanded so much that they were

damaged in the steel container, and the tests had to be stopped, as shown by the short curves 

in Figure 4.7. Again, the differences between (a) the first heating and (b) the second heating 

were due to drying shrinkage of cement paste.
36

Figure 4.7. Relative expansion of concrete with sandstone and river sand aggregate under 

different temperatures: (a) first heating; (b) second heating; 1, 2, and 3 indicate 

samples irradiated to fluences of 3 × 1020, 1.2 to 1.45 × 1020, and 4 to 6 × 1019 n/cm

respectively;---- Control samples kept at 20°C;____ control samples kept at 20°C 

220°C. (Dubrovskii et al. 1967)

The large expansion of concrete was due to the expansion of the aggregate used in the 

concrete. The aggregate was in the form of crystalline quartz. It is known that crystalline quartz 

becomes amorphous during irradiation (as shown in Figure 3.2), a process that is accompanied 

by anisotropic expansion and a density reduction of 15% (Bouniol 2011).

Available test data showed that nuclear radiation has a significant effect on thermal conductivity.

The conductivity of irradiated concrete can be a fraction of the value of unirradiated concrete 

(Gray 1972, Dubrovskii et al. 1967; also see Hilsdorf et al. 1978) with the reduction in the range 

of 30%, as shown in Figure 4.8. However, it is not clear that the variation in thermal conductivity 

is due to thermal effect, radiation effect, or the combined effect of the two. Furthermore, one can 

see from Figure 4.8 that the reduction of thermal conductivity for normal weight concrete starts 

from the level of neutron radiation < 1 × 1019 n/cm2 and the level of reduction remains the same 

up to 4 × 1019 n/cm2

. This indicates that the reduction mechanism is not heavily dependent on 

the level of radiation.
37

Figure 4.8. Thermal conductivity of concrete after neutron irradiation (Kc) relative 

to thermal conductivity of untreated concrete (Kco). (Gray 1972)

39

5. THE EFFECTS OF NUCLEAR RADIATION ON PROPERTIES OF 

CONCRETE

5.1 INTERACTION BETWEEN NUCLEAR RADIATION AND THE INTERNAL STRUCTURE 

OF CONCRETE

There are two types of interactions between nuclear radiation and the internal structure of 

concrete. One is related to geometric changes of the structure resulting from displacements of 

atoms from their lattice sites. The other is related to phase transformations of the constituents in 

concrete, resulting in reduction in porosity and/or formation of microcracks. As a result, the 

geometric changes and phase transformations can produce changes in properties of concrete.

The geometric changes to the internal structure of concrete will be discussed first, followed by 

the phase transformations.

A crystalline structure can be visualized as a three-dimensional lattice or grid formed by an 

orderly arrangement of atoms. During gamma and neutron irradiation of concrete, the atoms in 

the lattice structure of some constituents in the concrete are displaced from their lattice sites.

This displacement results in a new formation that is less stable than the previous one. This type 

of defect is known as a dislocation, or a lattice defect, and has an effect on the mechanical 

properties of materials, including strength and ductility. Dislocations can also contribute to 

macroscopic swelling, or geometric expansion, leading to an increase in internal stresses and 

strains.

Minerals with covalent bonds are more likely to be affected by radiation than those with ionic 

bonds (Kontani et al. 2010). Some minerals display significant changes in volume whereas

others seem to be unaffected by radiation. It is difficult to distinguish the volumetric change due 

to radiation from the volumetric change due to the heat of radiation because radiation is typically 

accompanied by a temperature increase.

Bouniol and Aspart (1998) investigated the disappearance of oxygen in concrete under 

irradiation and the role of peroxide in radiolysis eventually resulting in carbonation. They noted 

that the species theoretically obtained is peroxide octahydrate that can be easily produced in 

aqueous solution in presence of calcium and H2O:

Ca(OH)2 aq + H2O2 aq + 6H2O → CaO2.8H2O ↓

The resulting hydrate is slightly soluble, but is capable of losing its water of crystallization in air 

and then decomposing with simultaneous carbonation:

CaO2.8H2O → CaO2 + 8H2O

CaO2 + H2O → Ca(OH)2 + 1/2O2

Ca(OH)2 + CO2 → CaCO3 + H2O

The calcite crystals fill in pores and the porosity of concrete is reduced. The authors conducted 

experimental studies to verify the above (i.e., hydrated Portland cement paste was directly 

attacked by an aqueous solution containing 30% (by weight) H2O2). 

Vodák et al. (2005) conducted an experimental study on the effect of gamma radiation on 

porosity and calcite content of concrete. Figure 5.1 shows that the specific pore surface area is 

reduced with increasing dose of gamma radiation, Figure 5.2 shows that the porosity of 
40

concrete is reduced with increasing dose of gamma radiation, and Figure 5.3 shows that the 

relative intensity (i.e., content) of calcite is increased with an increasing dose of gamma 

radiation. Figure 5.3 presents results of diffraction analysis of calcite content in concrete for 

different values of dose. Results have been normalized to intensity of Si diffraction line that 

exhibits same value for all samples under consideration. These experimental results are in 

support of Bouniol and Aspart’s proposed phase transformations under nuclear irradiation. If an 

excessive amount of calcite is formed during the phase transformation process, then 

microcracks may occur near the pores, which may contribute to the strength reduction of 

concrete.

The phase transformation noted above is similar to the natural carbonation reaction in concrete, 

a reaction of portlandite with CO2 to form calcite and water. The difference is that the natural 

carbonation reaction takes place at the surface layer of concrete, where CO2 comes from the 

environment and penetrates into the concrete through diffusion, whereas the radiation-induced 

carbonation occurs in the concrete that has been irradiated and thus is not limited to the surface 

of a concrete structure. Another major difference between the two types of carbonation reaction 

is that the natural carbonation reaction can produce an increase in concrete strength whereas

the radiation-induced carbonation produces a decrease in both the size of the pore space and 

the strength of the concrete as the result of a succession of chemical reactions, starting with 

radiolysis of water and terminating in the formation of calcite crystals. 

Vodák et al. (2011) conducted another experimental study to provide a more-detailed 

investigation on the porosity variation in cement paste under gamma irradiation as well as under 

natural carbonation. Figure 5.4 shows a comparison of pore size distributions in the surface 

layer of concrete with irradiation and without. Figure 5.5 shows a comparison of pore size 

distributions in the center of concrete with irradiation and without. Both figures share a similarity 

in that the peak in the pore size distributions around 10 to 20 nm disappeared.

Figure 5.1. The specific pore surface area vs dose of gamma radiation. 

(Vodák et al. 2005)
41

Figure 5.2. The porosity of concrete vs dose of gamma radiation. (Vodák et al. 

2005.)

Figure 5.3. The intensity of CaCO3 vs dose of gamma radiation. (Vodák et al. 

2005)
42

Figure 5.4. The pore size distributions of irradiated (1 MGy) and unirradiated 

concrete specimens in the surface of concrete. Filled circles are 

unirradiated samples. Open squares are irradiated samples. (Vodák et 

al. 2011)
43

Figure 5.5. The pore size distributions of irradiated (1 MGy) and unirradiated 

concrete specimens in the center of concrete. Open circles are 

unirradiated samples. Open squares are irradiated samples (Vodák et 

al. 2011).

Kontani et al. (2010) provided a very good summary of the interactions of gamma and neutron 

radiation with cement paste and aggregates and of the deterioration mechanisms of concrete 

under nuclear irradiation. Table 5.1 summarizes the interactions of gamma rays with cement 

paste and aggregate. Table 5.2 summarizes the interactions of neutrons with cement paste and 

aggregate.

The effects of radiation on the components of concrete (aggregate and cement paste) translate 

to effects on properties of concrete itself, such as compressive strength, tensile strength, 

modulus of elasticity, and thermal expansion. Table 5.3 summarizes the deterioration 

mechanisms of concrete due to neutron radiation.
44

Table 5.1. The interactions of gamma rays with cement paste and aggregate (Kontani et 

al. 2010)

Water phase Solid phase

Cement 

paste

• Water may be decomposed by 

radiolysis to generate hydrogen and 

hydrogen peroxide, which in turn 

decomposes into water and oxygen.

• Gamma heating may cause 

additional hydration of unhydrated 

cement and transformation to 

hydrated cement.

• Hydrogen peroxide generated in 

radiolysis process may react with 

cement paste.

• SiO bond of 

calcium silicate 

hydrate in cement 

paste may be 

slightly 

decomposed due 

to the covalent 

nature of the bond.

• Electrons are 

ejected by 

scattering of 

gamma rays and 

collide with the 

solid phases of 

cement paste or 

aggregates.

Aggregate • Small amount of water may be 

released by the radiolysis in the form 

of hydrogen and oxygen gases as 

well as by gamma heating in the 

form of vapor.

• Siliceous 

aggregate may be 

slightly 

decomposed.

Table 5.2. The interactions of neutrons with cement paste and aggregate (Kontani et al. 

2010)

Water phase Solid phase

Cement 

paste

Molecular products from water 

may be the same as those for 

gamma rays, but the yields 

are different due to the 

different LET*

Although dislocation of atoms in solid cement particles 

may take place, the lattice defect by the distortion may 

not be accumulated, and the deformation of the paste 

may be negligible due to the porosity of the paste and 

the fineness of the cement particles

Aggregate Interaction with water is the 

same as that in cement paste

Lattice constants are increased due to the dislocation 

of atoms, and lattice defects are accumulated

*

LET = Linear Energy Transfer, which is the amount of radiation energy deposited in matter per unit particle path 

length in Joules per meter or keV/μm (Hala and Navratil 2003).

Table 5.3. Deterioration of concrete due to neutron radiation (Kontani et al. 2010)

Water phase Solid phase

• Drying of cement paste reduces its 

stiffness. Drying of cement paste 

produces microcracks to reduce its 

strength or increases surface energy of 

cement paste to increase strength.

• Additional hydration may increase the 

stiffness and the strength of concrete

• Hydrogen peroxide generated in the 

radiolysis process may react with 

cement paste. 

• Neutrons have some effects on water 

phase of cement paste. 

• Neutrons may have some effects on solid phases of 

cement paste, but the details are not understood 

very well.

• Volume expansions of aggregates may introduce 

microcracking in cement paste and decrease 

stiffness and strength of concrete, but the details are 

not understood very well.

• Concrete containing siliceous aggregates seems to 

crack more severely than those made with 

nonsiliceous aggregates, even if the expansions of 

aggregate alone are the same, but the mechanisms 

causing the difference is not understood.
45

5.2 COMPRESSIVE STRENGTH

A general assessment is that compressive strength of concrete decreases under gamma and 

neutron radiation exposure. Experimental studies in the literature have focused on specific 

strength reduction levels at specific radiation intensity (Hilsdorf et al. 1978). The purpose is to 

find the critical radiation intensities for gamma and neutron irradiation beyond which severe 

strength reduction will occur.

Many compressive strength tests were performed on concrete cubes of various compositions to 

study the effects of thermal-neutron radiation (Kaplan 1989). The tests showed insignificant 

changes in the compressive strength of irradiated concrete compared with that of heated 

unirradiated concrete. The results indicate that most of the strength degradation comes from the 

increased temperature and not from the thermal-neutron radiation itself at those levels.

Compressive strength tests have been performed on samples subjected to both neutron and 

gamma radiation, but few focus on gamma radiation alone (e.g., neutron radiation frequently 

occurs with gamma radiation).

In assessing the reduction of concrete strength under nuclear irradiation, 1 × 1019 n/cm2 for fast 

neutrons and 2 × 1010 rad (2 × 108 Gy) for gamma rays have been used as the critical levels of 

nuclear radiation. Concrete structures have been regarded to be sound as long as the levels of 

radiation do not exceed the critical levels, even if the NPPs were operated for more than 

30 years. The critical levels were based on experimental data collected by Hilsdorf et al. (1978).

Figure 5.6 presents a plot, prepared by Kontani et al. (2010) of the test data collected by 

Hilsdorf et al. (1978), showing the effect of neutron radiation on the residual compressive 

strength of concrete (i.e., residual equals ratio of radiated to unirradiated strength). The change 

in compressive strength is insignificant for neutron fluence below 2 × 1019 n/cm2

. Above that 

limit, the compressive strength of concrete appears to decline as the radiation level increases.

The threshold fluence value proposed by Hilsdorf et al. (1978) was 1 × 1019 n/cm2

, not 1 × 1020

n/cm2 as shown in Figure 5.6.

Figure 5.7 presents a plot, prepared by Kontani et al. (2010) of the test data collected by 

Hilsdorf et al. (1978), showing the effect of gamma rays on the residual compressive and tensile 

strengths of concrete. In the figure, the threshold value is shown as 2.0 × 1010 rad (2.0 × 108

Gy). In an Electric Power Research Institute (EPRI) report (EPRI 2012), the threshold value was 

considered as 1010 rad. In the original paper of Hilsdorf et al. (1978), no threshold value was 

proposed for gamma rays (only the 1 × 1019 n/cm2 was proposed for neutrons).

Most experimental studies of radiation effects on concrete were carried out from the 1960s into 

the 1970s in support of development of prestressed concrete reactor vessels for hightemperature gas-cooled reactors and radioactive waste storage facilities. More recently, Vodák 

et al. (2005) conducted an experimental study of the effect of gamma radiation on the strength 

of concrete. The concrete specimens were exposed to various levels of gamma irradiation. 

Based on their results, presented in Figure 5.8(a), the authors indicated that the compressive 

strength of concrete exposed to a gamma radiation of 5 × 105 Gy was reduced by about 10% 

compared to unirradiated samples. In fact, they noted that the strength reduction started from 

3 × 105 Gy. As stated by the authors, their results were not consistent with others, such as the 

commonly accepted reference value of 2 × 108 Gy. One possible problem of the test was that no 

method of shielding neutron radiation was mentioned, and therefore, the reduction in strength 

may be due to the combined effect of neutron and gamma radiation instead of only gamma
46

radiation. Also, the data scatter was relatively large and the data set relatively small so that 

statistically valid conclusions could not be derived.

Figure 5.6. Effect of neutron radiation on the residual compressive strength of 

concrete. (Kontani et al. 2010)

Figure 5.7. Effect of gamma radiation on the residual compressive strength of 

concrete. (Kontani et al. 2010)
47

(a)

(b)

Figure 5.8. Effect of gamma radiation on the compressive strength of 

concrete and mortar: (a) test data by Vodák et al. (2005) 

and (b) text data by Soo and Milian (2001).

The data shown in Figure 5.8(b) are from tests by Soo and Milian (2001) on Portland cement 

mortars. The data are similar to the test data from Vodák et al. (2005). Soo and Milian found 

that loss of compressive strength could occur at gamma doses that are much less than the 

threshold dose of 2 x 108 Gy (2 x 1010 rad) noted in Kontani et al. (2010). They postulated that 

the loss of strength could be connected with the radiolysis of the water of hydration in the 

cement as well as pore water. A loss of hydrogen and oxygen radiolytic species during 

irradiation would decrease the level of cement hydration and thus the strength of the cement. In 

Figure 5.8(b), the strength of the materials irradiated at the low dose rate (Co-60 gamma flux of 

31Gy/hr) falls rapidly once the dose exceeds about 105 Gy, but the strength for the higher dose 
48

rate (Co-60 gamma flux of 3.8 x 103 Gy/hr) does not decrease until the dose exceeds about 

107 Gy (Soo and Milian 2001). There was no explanation provided for this phenomenon.

Fujiwara et al. (2009) conducted carefully designed compressive strength tests on neutronirradiated concrete samples and found no significant reduction of the compressive strength due 

to neutron radiation. In the study, the irradiation temperature was kept lower than 65oC, which is 

the limiting value stipulated in the design standard for nuclear power stations in Japan (and in 

the United States). The specimens were 50 mm diam × 100 mm long cylinders. Table 5.4 

presents neutron fluence levels for each step in the tests as well as the number of irradiation 

cycles. 

Table 5.4. The neutron fluence level in each step in the test 

(Fujiwara et al. 2009)

Step Fast neutron fluence, 

n/cm2 (E > 0.1 MeV)

Irradiation cycles

1 0.5 × 1018 1

2 3.0 × 1018 6

3 2.0 × 1018 4

The maximum fast neutron fluence used in the study was 12.0×1018 n/cm2 (E > 0.1MeV), which 

sufficiently exceeds the total fast neutron fluence levels for the concrete located at the exterior 

of a reactor pressure vessel for a typical BWR after 60 years of operation. Test results 

presented in Figure 5.9 show that compressive strength of the irradiated concrete specimens 

was roughly equivalent to that of concrete specimens cured for the same duration under the 

basic environment of 20oC and 60% relative humidity. Therefore, their test data indicated that 

the radiation exposure at levels below 12 × 1018 n/cm2 did not significantly affect the 

compressive strength of concrete. 

Because Professor Hilsdorf’s paper was very important in determining the critical radiation 

values, Fujiwara et al. (2009) analyzed in detail the test data collected by Hilsdorf et al. (1978). 

They studied the original papers, compared the test specimens and conditions, and found the 

following:

• The temperatures in some of the specimens during irradiation reached 100oC or higher 

(approximately 140oC to 550oC). Thus the strength reduction could very well be induced by 

the elevated temperatures and not by the radiation, as shown in Figure 5.10 by red circles.

• The sizes of some specimens were very small (some were as small as 8 to 15 mm), which 

did not characterize the compressive strength of concrete, as shown in Figure 5.10 by green 

circles.

• Some specimens were tested for bending strength but not for compressive strength.

Specimens tested for bending strength are shown as blue circles in Figure 5.10.
49

Figure 5.9. Compressive strength of concrete under different 

levels of neutron radiation. (Fujiwara et al. 2009) 

Figure 5.10. The problems in the test data collected by Hilsdorf et al. (1978) on 

compressive strength of concrete. (Fujiwara et al. 2009)

Similar problems with the test data collected by Hilsdorf et al. were also pointed out by Kontani 

et al. (2010), as shown in Figs. 5.6 and 5.7 and summarized in Table 5.5. With these 

observations, Fujiwara et al. (2009) designed their test specimens and testing conditions 

carefully to avoid those problems. As noted in Figure 5.9, the Fujiwara et al. data indicated that 

fast neutron fluences up to 12 x 1018 did not significantly affect concrete strength.
50

Fujiwara et al. (2009) also compared their test data with those collected by Hilsdorf et al. (1978).

They took out the test data shown in Figure 5.10 in the red, green, and blue circles because of 

the reasons listed above (temperature, specimen size, and noncompressive strength). The 

comparisons are shown in Figure 5.11, in which the purple circles are Fujiwara’s test data and 

the blue circles are selected compressive strength results using data screened from that 

presented by Hilsdorf et al.

Table 5.5. The experimental conditions found in papers referenced. (Hilsdorf et al. 1978, 

Kontani et al. 2010) (screened) from Hilsdorf’s collection. One can see that a clear 

decrease in concrete 

Author*

Property 

Measured

Cement 

Type†

Specimen 

Size (mm)

Irradiation Conditions

Temperature 

(°C)

Neutron Energy Level & 

Fluence (n/cm2

)

Alexander[8] Compress. OPC 51 × 51 × 51 20 to 100 Thermal neutron 

0.29 × 1019 to 2.3 × 1019

Batten[9] Compress. OPC 51 × 51× 203 Not available No information

5.0 × 1019

Dubrovskii[2] Compress. Liquid glass φ15 × 15 

cylinder

200 to 550 No information

2.0 × 1021 to 2.4 × 1021

Elleuch[3] Compress. Aluminous 

cement

25 × 25 × 50 210 Fast neutron

2.6 × 1020 to 11.1 × 1020

Houben[4]

Schaaf[5]

Compress. OPC 8 × 8 × 70 15 to 200 Fast neutron

3.0 × 1019 to 8 × 1019

Price[6] Bending OPC 51 × 51× 203 50 Thermal neutron

0.5 × 1019 to 7 × 1019

Stoces[10] Compress. OPC φ50 × 70 

cylinder

<80 Fast neutron

4.2 × 1018

* The numbers shown in [ ] in the first column are the reference numbers in Kontani et al. 2010.

† OPC: ordinary Portland cement
51

Figure 5.11. Comparisons of the test data collected. [Hilsdorf et al. 

1978 (shown as “Data screened from Hilsdorf”) and 

Fujiwara et al. 2009 (shown as “Present data”). 

So far, the highest neutron radiation level used in the experimental studies was from 2.0 to 

2.4 × 1021 n/cm2

, which was used by Dubrovskii et al. (1966) as shown in Figure 5.6, Figure 

5.10, and Table 5.5. Their test results are very important because the test data are the only data 

set that showed the trend of strength reduction with increasing neutron radiation beyond 

1.0 × 1020 n/cm2

. Therefore, the experimental conditions used by Dubrovskii et al. (1966) have 

been carefully reexamined and the following four important findings resulted: 

• Two binders were used in the study, as shown in Table 5.6, one was Portland cement 

and the other was liquid water glass. This is different from the information shown in 

Table 5.5 by Kontani et al. (2010), which only lists the liquid water glass information

noted by Hilsdorf et al. The samples made of Portland cement were relevant to studying 

the effect of nuclear radiation on compressive strength of concrete.

• The samples were 15 mm diam cylinders with 15 mm height. Only fine aggregate 

(chromite filler) was used. Although the size of sample was small as indicated in Figures

5.6 and 5.10, it is an adequate size for studying the strength of Portland cement mortar 

(which is related to Portland cement concrete) under nuclear irradiation.

• The mean neutron energy level used in the study was 0.23 MeV; 30% of neutrons had 

an energy greater than 0.7 MeV. Thus the neutron energy level is in the energy level for 

fast neutrons [the energy level of fast neutrons is greater than 0.1 MeV according to 

Hilsdorf et al. (1978)].

• The radiation tests were conducted under high temperatures, in the range between 

200 oC and 550oC. However, the temperature effect was considered by using hot 

(heated) samples that underwent the same temperature history as irradiated samples 

and cold (unheated) samples that were maintained under standard laboratory conditions.

The differences between the hot and cold samples accounted for the temperature effect, 
52

and the difference between the hot and irradiated (irradiated and heated) samples 

showed the radiation effect. As marked in Table 5.6, about 26% of the fall in strength 

must be attributed to the radiation, and 34% must be attributed to the thermal effect 

(Dubrovskii et al. 1966). Therefore, assuming there was no synergistic effect, in the 

papers by Hilsdorf et al. (1978), Fujiwara et al. (2009), and Kontani et al. (2010, 2011), 

the test data from Dubrovskii et al. (1966) should be shown with the 26% reduction, not 

with the 60% reduction. The 60% reduction was due to the combined effect of high 

temperature and neutron radiation.

Dubrovskii et al. conducted more studies on the effect of nuclear radiation on concrete

(Dubrovskii et al. 1967, 1968). In the studies, “ordinary concrete” samples were used, which 

was defined as concrete made with Portland cement and aggregates of nonmetallic materials 

such as limestones, sandstones, granites, and serpentine. In their previous study (Dubrovskii et 

al. 1966), chromite filler was used as the aggregate. The test data from the studies in 1967 and 

1968 were not included in the collection of test data by Hilsdorf et al. (1978). 

Table 5.6. Strength and stiffness of concrete samples (Dubrovskii et al. 1966)*

Samples of chromite concrete 

using Portland cement binder

Samples of chromite concrete 

using liquid glass binder

Irradiated Hot Cold Irradiated Hot Cold

56.6 / 0.64 110.4 / 0.77 124.6 / 0.817 82.1 / 0.615 90.6 / 0.682 209.5 / —

62.3 / 0.68 95.1 / 0.725 126.3 / 0.885 71.3 / 0.545 134.8 / — 262.7 / 0.885

41.3 / 0.68 81.5 / 0.725 94 / 0.68 74.7 / 0.58 141.6 / 0.853 286.5 / 0.885

53.8 / 0.68 78.1 / 0.634 126.3 / 0.885 — 141.6 / 0.767 207.2 / 0.885

37.4 / 0.612 35.1 / — 126.3 / 0.817 — — —

37.4 / 0.68 101.9 / — 78.1 / — — — —

— 51 / 0.578 163.1 / — — — —

Average value

78.1 / 0.665 79 / 0.686 119.8 / 0.816 76.1 / 0.58 127.9 / 0.767 241.5 / 0.885

Relative value %

40 / 82 66 / 84 100 / 100 32 / 66 53 / 87 100 / 100

*Numerator of the fraction gives the strength of the sample in kg/cm2; denominator gives the slope of the 

strain curve on the compression diagram.

In one study by Dubrovskii et al. (1967), the concrete samples were cylinders measuring 40 mm 

in height and diameter. The temperature variation during irradiation was from 20°C to 220°C.

The test data showed that for radiation to a neutron fluence less than 5 × 1019 n/cm2

, the 

changes in strength and stiffness of concrete were insignificant. When the neutron fluence was 

increased to 1.45 × 1020 n/cm2

, noticeable changes were observed. Compressive strength of the 

concrete was reduced by 10% to 20%. However, the level of fluence used in this study did not 

reach 2.0 to 2.4 × 1021 n/cm2

, which was used for the study on the concrete with chromite filler 

(Dubrovskii et al. 1967). Part of the reason for this was that concrete samples expanded 

considerably during the irradiation process up to a neutron fluence level of 1.45 × 1020 n/cm2

. At 

higher levels, the samples were damaged in the steel container due to the excessive expansion.
53

The expansion of concrete came from the expansion of the aggregates that were 80% to 95% 

silicon oxide in the form of crystalline quartz; the quartz becomes amorphous during neutron 

irradiation with associated expansion.

In another study by Dubrovskii et al. (1968), the properties of serpentine concrete under neutron 

irradiation were studied. Finely ground serpentine was used in the concrete mixture. Large 

changes were observed when the neutron fluence was increased to 1.3 to 1.7 × 1021 n/cm2

.

Compressive strength of the concrete was reduced by 60%, and linear expansion reached 1.3% 

to 1.7%, with these changes considered to be due to microcracks that formed in concrete.

Based on results presented in this section, it is very difficult to deduce from these studies the 

extent of reduction of compressive strength due to nuclear irradiation and the mechanism of 

concrete deterioration by nuclear radiation. The experimental conditions of some of the data 

collected by Hilsdorf et al. (1978) were very different from the testing conditions of Fujiwara et 

al. (2009) and Vodák et al. (2005). They were also not representative of the operating conditions 

of NPPs (for example, operation temperature is no higher than 65oC). More importantly, there 

has been a lack of experimental studies under high levels of irradiation, and the studies 

conducted by Dubrovskii et al. (1966, 1967, 1968) were each for one type of concrete.

Therefore, Kontani et al. (2010, 2011) noted that there is a pressing need to conduct new and 

more systematic radiation tests to study the critical levels of gamma and neutron radiation 

corresponding to noticeable strength reduction. Systematic experimental studies are needed for 

determining appropriate reference levels of nuclear radiation for assessing the integrity and 

soundness of NPP concrete structures. Interactions between radiation effects and temperatures 

need to be investigated as well as the effects of neutron and gamma ray radiation on cement 

paste and aggregates in concrete.

5.3 TENSILE STRENGTH

Tensile strength of concrete can be tested by different methods, such as splitting-tension,

flexural-tension, and direct-tension testing. The tensile strength measured by the flexural 

tension test is usually higher than the value obtained from the splitting-tension test because of 

nonuniform stress distribution in the concrete beam. As a result, the tensile strength test data 

exhibit large scattering, especially when tensile strengths have been obtained by different 

testing methods and compared together. This is apparent in the paper by Hilsdorf et al. (1978) 

that summarized tensile test data for concrete subjected to irradiation. The results indicate that 

for a neutron fluence of 5 × 1019 n/cm2

, the tensile strength of irradiated specimens can be 20% 

to 80% less than the strength of unirradiated and unheated specimens, as shown in 

Figure 5.12. Thus, depending on the test conditions and the radiation level, the tensile strength 

of concrete can be reduced significantly.
54

Figure 5.12. Effect of neutron radiation on tensile 

strength of concrete. (Hilsdorf et al. 1978)

For evaluation of the effect of gamma radiation, Vodák et al. (2005) performed tensile-strength 

tests on samples similar to those described in Section 5.2 for compressive strength of concrete.

Their results are shown in Figures 5.13 and 5.14 for splittingtensile strength and flexural tensile 

strength, respectively. Although the authors indicated that there was a trend in tensile-strength 

reduction due to increasing gamma radiation similar to that in Figure 5.8(a) for the compressive 

strength, statistically valid conclusions cannot be derived because the data scatter was 

relatively large and the data set relatively small. It was noted that a reduction in tensile strength 

for some specimens had started to occur by the time the dose to the specimens reached 

approximately 3 × 105 Gy (i.e., an amount corresponding to approximately 40 years of operation 

at the Temelin NPP).

Figure 5.13. Effect of gamma radiation on splitting tensile 

strength of concrete. (Vodák et al. 2005)
55

Figure 5.14. Effect of gamma radiation on flexural tensile strength of 

concrete. (Vodák et al. 2005) 

Because the tensile strength test data are very limited and available results exhibit significant 

data scatter, it is difficult to obtain a definite reference level of radiation that leads to a significant 

reduction of the tensile strength of concrete. Kaplan (1989) summarized that in general neutron 

fluences greater than 1019 n/cm2 will lead to a reduced tensile strength.

5.4 MODULUS OF ELASTICITY

Similar to the strengths of concrete, the stiffness, or modulus of elasticity, of concrete is reduced 

under nuclear radiation, as shown in Figure 5.15 (Hilsdorf et al. 1978). In the figure, “Ec/Eco” is 

the ratio of the modulus of elasticity after irradiation and heating to the modulus of elasticity 

before irradiation and heating. A neutron fluence in the range of 1 × 1019 n/cm2 leads to a slight 

decrease of the modulus of elasticity. When the fluence was larger than 5 × 1019 n/cm2

, the 

reduction can be as high as 30% of the original stiffness.

Figure 5.15. The effect of neutron radiation on the 

modulus of elasticity of concrete: Ec = 

after neutron radiation and Eco = 

untreated. (Hilsdorf et al. 1978)
56

Fujiwara et al. (2009) reported test data in which the static modulus of elasticity of irradiated 

specimens was not significantly different from that of nonirradiated specimens at the levels of 

fast neutron fluence expected on the reactor pressure vessel exterior of a typical BWR over a 

60 year period [3.0 × 1018 n/cm2 (E > 0.1 MeV)]. There was a tendency for the values to 

decrease slightly at levels above 6.0 × 1018 n/cm2

, but the decrease was not significant, as 

shown in Figure 5.16.

Figure 5.16. Static modulus of elasticity of 

concrete after different levels of 

neutron radiation. (Fujiwara et al. 

2009) 

The test data presented in this section showed the combined effect of neutron radiation and 

thermal effect. Available test data are therefore not sufficient to separate the effects of radiation 

and temperature, which is one of the main goals of the ongoing study by Kontani et al. (2010, 

2011).

5.5 CREEP, SHRINKAGE, AND VOLUMETRIC VARIATION

The effect of gamma radiation on the creep and shrinkage of concrete was investigated using a 

limestone aggregate concrete representative of that used in the concrete pressure vessel of the 

Oldbury Nuclear Power Station (McDowall 1971). Cylindrical specimens, 105 mm in diameter by 

305 mm long were subjected to a load of 10 N/mm2 at an average sample temperature of 30°C.

Figure 5.17 presents shrinkage and creep results obtained from specimens subjected to an 

average gamma dose rate of 11.4 × 103 rad/h as well as those for unirradiated control 

specimens. In the figure, the creep of untreated (i.e., unirradiated and unheated) concrete is 

shown to be higher than that of irradiated concrete. As indicated earlier, the modulus of 

elasticity of irradiated concrete is expected to be lower than that of regular concrete, and thus 

the creep of irradiated concrete would be expected to be higher. However, the results in Figure 

5.17 show an opposite trend. One theory for creep being opposite of what was expected was 

that results for the irradiated concrete at any point in time were obtained from specimens that 

were drier than the control specimens.

Figure 5.18 shows the effect of neutron radiation on volume expansion of concrete (Gray 1972,

Kelly et al. 1968; also see Hilsdorf et al. 1978). The cement paste undergoes shrinkage while

the aggregate expands under neutron radiation. Thus, the amount of shrinkage of cement paste 
57

and the amount of expansion of aggregate determine the overall volumetric variation of the 

concrete. It is clearly shown in Figure 5.18 that the concrete made of flint aggregate expands 

much more than the concrete made of limestone aggregate; the cement paste in the two types 

of specimens would be expected to have similar shrinkage.

Figure 5.17. The effect of gamma radiation on creep and shrinkage of concrete. 

(McDowall 1971) (Note: rad = 0.01 Gy).

Figure 5.18. The effect of neutron radiation on the volume of concrete. 

(Gray 1972, Kelly et al. 1968)
58

Hilsdorf et al. (1978) pointed out that, for the volumetric expansion shown in Figure 5.18, 

concrete specimens that were only exposed to elevated temperatures but not irradiated did not 

show such a volume increase; rather, they showed the expected shrinkage. The shrinkage was 

of the cement paste, resulting from loss of water. This means that the volume expansion was 

mainly from the volume expansion of aggregates due to the neutron irradiation, and the volume 

expansion of aggregate under neutron irradiation can be very significant. Hilsdorf et al. (1978) 

compared the test data shown in Figure 5.18 for volumetric expansion with tensile strength test 

data given in the research reports (Gray 1972, Kelly et al. 1968) and further concluded that

tensile strength of concrete is lower when the volume increase during neutron irradiation is 

larger.

With increasing age of nuclear reactors in service in the United States and around the world, 

deformation properties of concrete, such as creep and shrinkage, have increased importance 

for long-term performance evaluation of the concrete structures. Brooks (2005) conducted a 

long-term study (i.e., 30 years) to test the creep and shrinkage of various concretes. These tests 

were conducted without nuclear irradiation, but the results are very important for long-term 

evaluation of concrete structures in NPPs.

The creep test data in Figure 5.19 indicate that (1) the creep rates (the slopes of the curves) of 

dry-stored concrete samples decreased at about one year after loading, the rates remained the 

same all the way to 30 years, and there was no sign for the asymptotic trend of the ultimate 

creep; and (2) the rates of wet-stored concrete samples started to increase after about 10 years.

The range of the measured 30 year creep coefficient was from 1.2 to 9.2, depending on the 

water/cement ratio (or strength) and aggregate type. For those test data, specific creep and 

creep compliance were underestimated by all methods of prediction.

Figure 5.19. Creep of concrete made with North Notts coarse aggregate stored 

under dry conditions (T = 20oC – 22oC, relative humidity = 65%) and 

wet conditions (under water at T = 20 oC - 22°C) (creep in 

m/m)(Brooks 2005)
59

The shrinkage test data in Figure 5.20 indicate that (1) the drying shrinkage of concrete started

to level off after 1 year because the internal moisture state of the concrete reached equilibrium,

and (2) the swelling appeared to increase after about 1 year; however, the total amount of 

swelling was quite small, 200 to 300 microstrain.

Figure 5.20. Shrinkage of concrete made with North Notts coarse 

aggregate stored under dry conditions (T = 20oC –22oC, 

relative humidity = 65%) and wet conditions (under 

water at T = 20 oC to 22oC) (shrinkage in m/m). (Brooks 

2005)

5.6 EFFECT OF HIGH TEMPERATURE ON THE NUCLEAR SHIELDING CAPACITIES OF 

CONCRETE

Yousef et al. (2008) conducted an experimental study related to the effect of high temperature

on the strength and gamma-ray and neutron shielding properties of four types of concrete. The 

samples were exposed to temperatures between 20oC and 800oC. Only the temperature effects 

on shielding properties are discussed here. Figure 5.21(a) shows the test data on attenuation 

coefficients of gamma rays. Results indicate small losses in attenuation coefficients for gamma 

rays at temperatures up to 100oC, with no significant loss in attenuation resulting for 

temperatures between 100oC and 400oC. From 400oC and 500oC, there also was no significant 

loss in gamma rays attenuation except for one of the concretes (D-mix), which suffered a higher

loss in gamma-ray attenuation as the temperature was increased from 350oC to 500oC. The F 

and D mixes lost their integrity between 500oC and 550oC. The effect of temperature on neutron 

attenuation index, noted in Figure 5.21(b), exhibited a similar trend to that for the gamma 

attenuation coefficient at temperatures to 500oC. Although the hematite coarse-aggregate 

concrete performed best, the shielding capacities of all the concrete mixes exhibited only small 

reductions at temperatures almost up to 400oC. 
60

(a) (b)

Figure 5.21. (a) Gamma-ray attenuation coefficients vs temperature and 

(b) attenuation index of neutron vs temperature for the four concretes: 

D mix (dolomite crushed coarse aggregate), F mix (river natural coarse 

aggregate), H mix (hematite crushed coarse aggregate), and S mix 

(serpentine crushed coarse aggregate). (Yousef et al. 2008)

Sakr and El-Hakim (2005) conducted an experimental study of the effect of high temperature on 

attenuation capacities of concretes made with different types of aggregates. Three types of 

aggregates were used in the study: gravel, baryte, and ilmenite. The effect of different durations 

(1, 2, and 3 h) of high temperatures (250oC, 500oC, 750oC and 950oC) on the physical, 

mechanical, and shielding properties of heavyweight concretes were studied. Some of the test 

data are shown in Table 5.7. The table presents the attenuation coefficient percentage (μ%) of 

Co60 and Cs137 gamma sources for different types of concrete heated for 2 h to different 

temperatures. The table shows that the attenuation coefficient depends on temperature and that it 

decreases by about 16% for gravel concrete, 17% for baryte concrete, and 16% for ilmenite 

concrete at 750oC, compared with the respective attenuation coefficients at room temperature 

(25oC) for different energy levels (1330, 1170, and 660 keV). The attenuation coefficient of 

ilmenite concrete was higher than that of gravel concrete for Co60 by 31% and for Cs137 by 34%, 

respectively, at 750oC. Similar results were obtained for the baryte concrete. This can be 

attributed to the fact that ilmenite concrete had a higher density than either the gravel or baryte 

concretes.
61

Table 5.7. Attenuation coefficient percentages of tested concrete at different 

temperatures (Sakr and El-Hakim 2005)

Type of 

concrete

Temperature 

(°C)

Attenuation coefficient percent (μ%)

Co60 Cs137

1330 keV 1170 keV (660 keV)

Gravel 25

250

500

750

950

12.72

11.53

11.01

10.69

9.55

12.21

11.16

10.55

10.31

9.18

7.33

6.67

6.28

6.15

5.50

Baryte 25

250

500

750

950

15.79

14.53

13.89

13.25

12.61

15.54

14.14

13.52

12.91

12.27

9.73

8.94

8.54

8.15

7.76

Ilmenite 25

250

500

750

950

16.59

14.61

14.27

13.93

13.61

16.22

14.27

13.95

13.62

13.31

9.81

8.66

8.46

8.26

8.1

5.7 EFFECT OF LOW TEMPERATURE ON THE NUCLEAR SHIELDING CAPACITIES OF 

CONCRETE

Under a low temperature, ice may form in concrete and the volume expansion associated with 

the ice formation may cause damage in concrete. The damage induced by low temperatures will 

have an impact on the shielding capacity of concrete. Basyigit et al. (2006) conducted an 

experimental study on the shielding capacity of two types of concretes after the concretes were 

subjected to freeze-thaw cycles (see Figure 5.22). When normal weight coarse aggregate was 

used, the concrete was called A, if the coarse aggregate was barite, it was called B. The indices 

of 2, 3, and 4 represent w/c ratios of 0.65, 0.51, and 0.43, respectively. Figure 5.22(a) and 

Figure 5.22(b) present linear attenuation coefficient results for the A- and B-type concretes, 

respectively, and indicate that the linear attenuation coefficient of both concretes decreased with 

an increasing number of freeze-thaw cycles.

Comparing the two figures, one can see that the attenuation coefficient of B-type concretes 

(with barite aggregates) is much higher than that of the A-type, which is due to the contribution 

of barite aggregate (heavyweight aggregate). For A-type concrete, the effect of w/c was not as 

significant as it was for the B-type concrete. Although the biological shield of an NPP in all 

likelihood will not be subjected to freeze-thaw cycles, the results are important from the 

perspective of indicating the potential impact of damage (i.e., cracking) on the attenuation of 

shielding concretes.
62

(a) (b)

Figure 5.22. (a) Linear attenuation coefficient μ of A-type concretes as a function of 

freeze-thaw cycles; and (b) Linear attenuation coefficient μ of B-type 

concretes as a function of freeze-thaw cycles. (Basyigit et al. 2006)
63

6. INTERACTION OF NUCLEAR RADIATION, DURABILITY, AND 

SHIELDING EFFECTIVENESS

In theory, radiation can have significant effects on durability of concrete. In practice, limited 

research has been conducted on the durability of concrete under the influence of nuclear 

radiation. In this chapter, the interaction of nuclear radiation, durability, and shielding 

effectiveness will be addressed in four areas: (1) alkali-silica reaction, (2) carbonation,

(3) radiolysis and evaporation of water, and (4) concrete cracking/spalling.

6.1 ASR AND THE COUPLING EFFECT BETWEEN ASR AND RADIATION

Ichikawa and Koizumi (2002) and Ichikawa and Kimura (2007) showed that ASR nonreactive 

aggregates can become reactive (or ASR-sensitive) under irradiation, depending on the 

intensity of the radiation, the reactivity of the aggregate, and the alkali content of the cement.

Therefore, ASR and the possible coupling effect between ASR and nuclear radiation have 

become important research topics.

6.1.1 ASR of Concrete

The alkali-silica reaction (ASR) in concrete can be described as a chemical reaction involving 

alkali cations and hydroxyl ions from concrete pore solutions (Hobbs 1987). ASR will produce a 

gel that swells with the absorption of moisture. The amount of ASR gel and the swelling 

pressure vary depending on the reaction temperature, type and proportions of the reacting 

materials, gel composition, gradation of aggregates, and other factors. Sometimes the pressure 

generated by ASR gel in concrete is sufficient to generate and propagate microcracks in the 

concrete. The chemical reactions occurring in ASR-affected concrete structures are slow and 

complex. In a simplified manner, it can be described as a two-step reaction between alkalis 

(sodium and potassium) in cement and reactive silica aggregates. The first step is the chemical 

reaction between the reactive silica in the aggregate with the alkali present in cement paste to 

produce alkali-silica gel:

Reactive silica in aggregate + Alkali in cement paste → Alkali-silica gel (6.1) [xSiO2] [yNa(K)OH] [Na(K)ySixOz aqueous]

The second step is the expansion due to the alkali-silica gel interacting with sufficient moisture:

Alkali-silica gel + Moisture → Expanded alkali-silica gel (6.2) [Na(K)ySixOz aqueous] [H2O] [Na(K)ySixOz wH2O]

The silica content and the internal structure of aggregate are very important. In a more 

sophisticated description for ASR, it has been considered as four fundamental reactions (Kurtis 

et al. 2002). These reactions are the dissolution of silica by hydroxyl attack, the reaction of 

surface silanol groups with OH-

, the binding of cations to the silicate, and the reaction to form 

calcium silicate hydrates. It should be pointed out that for understanding the exact nature of 

ASR, detailed information on the reactions and kinetics at microscale and nanoscale are 

needed, for which there are some experimental results available in the literature (Jennings 1986,

Glasser 1992, Zanni et al. 1994, Monteiro et al. 1997).
64

From the abovementioned chemical equations, one can see two important conditions for the 

expansive ASR to occur in a concrete structure. One is the high concentration of alkalis, which 

results in an equally high concentration of OH- ions (to maintain charge equilibrium). It is this 

high OH- concentration, and thus high pH, that leads to the initial breakdown of reactive silica 

components in the aggregates, as shown in Eq. (6.1). The other is the presence of moisture in 

the concrete structure. Usually, the structures that have a high risk of ASR are the ones 

exposed to wet environments because, as shown in Eq. (6.2), one of requirements for 

expansive ASR to occur is moisture. As such, many concrete dams have experienced ASR 

problems (Saouma and Xi 2004). ASR has also been identified in one NPP in the United States 

(i.e., the Seabrook NPP), where the intrusion of moisture into sections of walls in certain belowgrade structures resulted in indications of ASR degradation as evidenced by pattern cracking 

(see http://www.nrc.gov/info-finder/reactor/seabrook/concrete-degradation.html). After cracking 

occurs in a concrete structure, the cracking can expose the interior of the concrete to 

aggressive agents such as chlorides or sulfates. This can then lead to further deterioration 

through other mechanisms, such as corrosion of steel reinforcement. 

Many different types of aggregates that have been used in concrete have caused ASR. All of 

the reactive aggregates contain silicon dioxide (SiO2). Some are crystalline, such as tridymite, 

cristobalite, and strained quartz; others, such as opal, chert, and glassy volcanic materials, are 

glassy. Other types of reactive aggregates are composites having both crystalline and glassy 

components, including graywackes and phyllites. 

6.1.2 ASR and Radiation of Concrete

When an aggregate is subjected to neutron irradiation, the degree of expansion of the 

aggregate strongly depends on the type of aggregate. The expansion at a fluence of 

5 × 1019 n/cm2 is about 1% for limestone and flint, and about 0.1% for serpentine. However, 

when these aggregates were used to make concrete, the resulting concrete exhibited 

expansions differing from the estimated expansion based on expansion of the aggregates. For 

example, limestone and flint showed the same degree of expansion under neutron radiation.

The expansion of concrete with flint is about five times larger than that with limestone; 

moreover, the concrete with quartzite sand showed severe expansion, but the crystalline quartz 

itself showed only small expansion. These results suggest that the aggregates may react with 

surrounding cement paste in concrete to generate greater expansion due to neutron radiation 

than that experienced by the aggregate alone. Ichikawa and Koizumi (2002) considered this 

additional expansion as possibly occurring as a result of a coupling effect between ASR and 

radiation. 

The experimental study by Ichikawa and Koizumi (2002) showed that the reactivity of silica-rich 

quartz to alkali may be significantly increased by nuclear radiation. Samples used for their 

experiment were 2 × 2 cm crystalline and amorphous quartz plates. A part of the plates was 

irradiated with 200 keV Ar ions to different dosages, and the plates were then immersed in an 

aqueous solution of NaOH (to simulate the high-alkali pore solution in concrete). The height of 

the terraces generated at the boundary of the irradiated and unirradiated areas was measured, 

which is an indication of the increased reactivity of quartz to the alkali. Irradiation of the plate to

less than 2 × 1014 Ar/cm2 does not generate a terrace, which means that the reactivity to NaOH 

was not changed, however, when increased to 6 × 1014 Ar/cm2

, the reactivity was significantly 

increased, with a plateau forming as shown in Figure 6.1(a). Figure 6.1(b) shows the evolution 

of the plateau under three different temperatures, the higher the temperature, the higher the 

height of the terrace. Because the irradiation used in the study, 6 × 1014 Ar/cm2

, was close to 
65

that for amorphization of crystalline quartz, it can be concluded that the amorphization of 

crystalline quartz significantly increases the reactivity of crystalline quartz to the alkali.

(a) (b)

Figure 6.1. (a) The height of terrace emerged at the boundary of 

irradiated and unirradiated areas of crystalline quartz after 

being immersed in 1 mol/dm3 NaOH solution at 353 K for 4 

h. (b) Evolution of a terrace on 1 × 1015 Ar/cm2 irradiated 

crystalline quartz in 1 mol/dm3 NaOH solution at different 

temperatures. (Ichikawa and Koizumi 2002)

Such an increase in the reactivity does not necessarily induce ASR; the rate of ASR 

development depends not only on the reactivity of aggregate but also on the chemical 

composition of Portland cement. When the combination of aggregate and cement is quite close 

to the critical condition for a detrimental ASR, then the increase in aggregate reactivity due to 

nuclear radiation could become a very important factor, making a nonreactive aggregate ASRsensitive. 

The experimental study by Ichikawa and Kimura (2007) showed that the reactivity of 

plagioclase, one of the major minerals composing volcanic rocks, can be increased by nuclear 

radiation. Irradiation of plagioclase with a 30 keV electron beam at a dose of more than 

0.9 × 108 Gy converts a crystalline plagioclase to an amorphous form that is 35 times more 

reactive to alkali than the crystalline form. The high reactivity of the irradiated plagioclase

indicates that irradiation can induce ASR in aggregate materials that are not normally 

susceptible to ASR.

In addition to the amount of its free expansion, the stiffness of ASR gel is also very important in

determining the extent of damage due to ASR. When the ASR gel is soft and its stiffness is low, 

it is able to permeate into the surrounding porous cement paste, and thus the swelling pressure 

is not high and the cracking potential is low. In that case, a large amount of ASR gel formation 

may not create severe damage and cracking in concrete. On the other hand, when the ASR gel 

is stiff, a small amount of ASR can generate significant damage. The stiffness of ASR gel 

depends mainly on its chemical composition, such as the ratio of Na2O to SiO2. Struble and

Diamond (1981a, 1981b) conducted a study to investigate the swelling pressure of ASR in 

terms of ASR chemical composition. Table 6.1 presents examples of maximum swelling 

capacity and pressure for different ASR gel chemical compositions.
66

Table 6.1. Swelling capacity and swelling pressure of ASR gel. (Struble 

and Diamond 1981a)

Mole Ratios Total Free 

Swelling, %

Maximum Swelling 

Na2O CaO SiO2 H2O Pressure, MPa

1 0.27 1.00 1.93 81.6 0.40

2 0.27 0.17 1.00 1.16 8.0 0.30

3 0.28 1.00 1.32 3.6 0.14

4 0.30 1.00 1.77 2.5 0.27

5 0.33 1.00 1.98 0.5 0.06

6 0.34 1.00 2.24 1.3 10.89

7 0.35 0.18 1.00 1.57 3.1 0.51

8 0.42 1.00 1.99 63.2 3.09

9 0.46 1.00 1.84 2.0 0.09

10 0.53 1.00 2.1 69.8 2.22

From Table 6.1, it is evident that sample No. 6 has the highest swelling pressure (10.89 MPa) 

and only 1.3% free-swelling. Thus the stiffness of ASR gel in sample No. 6 must be high. On the 

other hand, sample No. 9 has a low swelling pressure (0.09 MPa) and higher free-swelling 

capacity (2%). Therefore, the stiffness of ASR gel in sample No. 9 must be low. 

The test data of Ichikawa and Koizumi (2002) showed that nuclear irradiation may change the 

ASR expansion potential of aggregate. A similar question may be asked about whether the 

stiffness of ASR gel can change under nuclear irradiation. If it can, then ASR coupled with

nuclear irradiation can generate more severe damage than the ASR alone.

6.2 CARBONATION

Carbonation is the chemical reaction of portlandite, Ca(OH)2, in the cement matrix with carbon 

dioxide gas leading to calcite (CaCO3). Radiation-induced carbonation was discussed in 

Section 5.1, which is similar to the natural carbonation reaction in concrete. One of the main 

consequences of the carbonation reaction is the reduction of the pH value in the pore solution of 

concrete from an initial range of 12.5 and 13.5 to a value below 9 in the carbonated zones (Van 

Gerven et al. 2004). In the presence of radiation, the carbonation reaction is accelerated as 

shown by Vodák et al. (2005, 2011).

Bar-Nes et al. (2008) conducted an experimental study on the combined effect of carbonation 

and gamma irradiation on concrete. Carbonation tests were performed on (a) samples that were 

irradiated at atmospheric conditions to a dose of 107 Gy during a 6 month period, (b) reference 

samples that were nonirradiated and left in free atmospheric conditions during the same period,

and (c) sealed samples that were not subjected to either free atmospheric conditions or 

irradiation during the whole period. All samples were prepared with a constant w/c ratio of 0.45.

No carbonation was detected for the samples in group (c). The reference samples, exposed to 

free atmospheric conditions but not irradiated, showed an average carbonation depth of 
67

approximately 0.5 to 0.6 mm. For the irradiated samples, a much larger carbonation depth was 

detected (average of around 3 mm), reaching, in some cases, depths of up to 6 mm. 

The effects of carbonation on the properties of concrete are quite complicated. Consumption of 

portlandite during the reaction increases the porosity while deposition of calcite decreases the 

porosity. As a result, the strength and transport properties of concrete depend more on other 

parameters than on radiation-induced carbonation. For instance, the leaching rate of heavy 

metals such as Sr and Cs in nuclear waste was studied by Bar-Nes et al. (2008). Although the 

carbonation depth was increased by the nuclear irradiation, the diffusion coefficients of Sr and 

Cs were affected differently as noted below.

Figure 6.2(a) shows the leaching test data of Sr from cement paste (w/c = 0.3), in which the 

cumulative fraction of Sr leached from the irradiated sample is lower than that from the 

nonirradiated sample. For the Sr ions, irradiation reduces the diffusion coefficient by a factor of 

28 early in the experiment for the w/c = 0.3 pastes, probably due to increased carbonation. For 

the w/c = 0.45 pastes, the apparent diffusion coefficient has been reduced due to irradiation by 

a factor of 9 (not shown). With a reduced diffusion coefficient, the leaching rate of Sr is lower.

(a) (b)

Figure 6.2. (a) Leaching of Sr from samples with w/c = 0.3. (b) Leaching of Cs 

from samples with w/c = 0.3. (Bar-Nes et al. 2008)

For Cs ions, the diffusion coefficient of the irradiated sample increased slightly, which is different 

from the effect of radiation on the diffusion coefficient of Sr. As shown in Figure 6.2(b), the 

cumulative fraction of Cs leached from the irradiated sample is higher than that from the 

nonirradiated sample. This may be caused by the high solubility of Cs in the paste. Cs ions do 

not precipitate as a carbonate phase, and the carbonation has no effect on the Cs ion diffusivity. 

The small increase in Cs diffusivity due to irradiation might originate from the formation of 

microcracks in the product due to the radiolysis process.

6.3 RADIOLYSIS AND EVAPORATION OF WATER

The decomposition of water under radiation is called radiolysis. It is different from vaporization 

of water under elevated temperature, as shown in Figure 6.3. Both interstitial water and water in 

the pore structure of concrete may participate in the radiolysis process. One of the products of 

radiolysis is hydrogen gas, H2. In addition to the effects of radiolysis resulting from radiation 

noted in Chapter 2, water radiolysis is important for dry storage facilities and for radioactive 

waste conditioning. For long-lived wastes, concrete materials are used as an immobilization 

matrix undergoing mixed irradiation (i.e., a combination of alpha, beta, and gamma radiation)
68

(Bouniol and Aspart 1998, Bouniol 2011). The two main outcomes resulting from water 

radiolysis are (1) buildup of internal overpressure that may lead to cracking of the waste form 

and (2) the accumulation of hydrogen gas in a storage area, hence, the buildup of an explosive 

gas mixture when mixed with atmospheric oxygen (Bar-Nes et al. 2008). Therefore, radiolysis of 

water is more of a safety problem related to those applications than to the damage of concrete.

Figure 6.3. Gamma ray dose vs water 

retained in concrete. (Kontani et al. 

2011)

Kontani et al. (2011) conducted a carefully designed test to study the water released by 

concrete during irradiation. Figure 6.4(a) and (b) show the experimental setup. The radiation

level was controlled by the distance from the concrete samples to the radiation source.

Figure 6.5(a) shows temperature changes of containers at the first and second rows and

changes of irradiation room temperatures. The temperatures of containers in the second row 

were 15oC higher than the room temperatures, and the temperatures of the containers in the 

first row were 30oC higher than the room temperatures. Figure 6.5(b) shows the water released 

vs irradiation periods due to the heat of radiation. After 40 to 50 days of irradiation, release of 

water vapor due to the heat of gamma rays reached the maximum. The amount of water 

released was greater in the first row than in the second row because of differences in 

temperatures.

(a) (b)

Figure 6.4. (a) Test configurations of gamma irradiation tests; and 

(b) arrangement of containers. (Kontani et al. 2011)
69

(a) (b)

Figure 6.5. (a) Temperature in the room and in the specimens and (b) water 

released vs irradiation duration. (Kontani et al. 2011)

Figure 6.6 shows the rates of hydrogen generation from each container. Those from containers 

in the first row showed maximum values of 8.5 x 10-5 mol/h just after the start of irradiation and 

then started to reduce rapidly. Those from containers in the second row showed maximum 

values of 3.0 x 10-5 mol/h 2 days after the start of irradiation and then reduced very slowly. After 

60 days of irradiation, the generation rates of hydrogen were greater in the second row than in 

the first row although the temperatures in the first row were higher. This means that the 

radiolysis of water in the concrete at the second row could occur more easily than that in the 

concrete in the first row, which is unexpected.

Figure 6.6. Generation of hydrogen vs irradiation period. (Kontani et al. 

2011)
70

6.4 THE EFFECT OF CONCRETE CRACKING/SPALLING ON SHIELDING 

The rust formation during rebar corrosion generates a high interface pressure between the rebar 

and the surrounding concrete. The level of pressure depends on many factors, such as the 

volume ratio of rust to steel and the thickness of the concrete cover. If the pressure is high 

enough, cracks may form in the concrete and the concrete cover may spall. Because applied 

loads to shielding structures are relatively low, the formation of cracks may not be a problem for 

the load-carrying capacity of the structure, but may have an impact on the radiation shielding 

capacity of the concrete.

Lee et al. (2008) conducted a numerical study on the effect of corrosion cracks on shielding 

capacity of a concrete wall. The interface pressure and the crack pattern due to rebar corrosion 

are shown schematically in Figure 6.7. The results of finite-element simulation of the crack 

growth are shown in Figure 6.8. Assuming that the wall thickness is 100 cm and the cover 

thickness is 10 cm, and that the spalling of the cover will occur due to the rebar corrosion, it is 

estimated that the radiation safety standards will be maintained under any conditions related to 

the effect of cracking. The analysis can also be used for other geometries of the wall and cover 

thickness. Results indicated that as long as the wall thickness minus cover thickness is more 

than 90 cm, radiation safety standards will be maintained.

Figure 6.7. The interface pressure and the crack pattern due to rebar corrosion. (Lee et 

al. 2008)

Figure 6.8. Crack growth around steel bars due to the rust formation. (Lee 

et al. 2008)
71

Lee et al. (2007) proposed a testing system for quantifying the effect of cracks on shielding 

performance in a concrete body. The correlation between crack width and shielding 

performance is deduced through experiments. The measuring system has a collimator, and it 

can be mounted on a concrete sample with collinear cracking. Schematics of the experimental 

setup and the concrete sample are shown in Figure 6.9. A photograph of the experimental setup 

is shown in Figure 6.10. Figure 6.11 presents the relationship between crack width, concrete 

thickness, and the relative intensity of radiation.

Lee et al. (2007) noted that if the thickness of the shielding material exceeds 20 cm and the 

crack width is < 0.4 mm, the estimated effect of the crack on shielding performance will not be 

greater than 10%. They also noted that the cracks were straight; in an actual structure the crack 

formation would be irregular, and the irregularity would have an additional effect.

Figure 6.9. Schematic of the positioning system and the concrete 

sample. (Lee et al. 2007)

Figure 6.10. Experimental setup of the testing 

system for cracks in concrete.

(Lee et al. 2007)
72

Figure 6.11. The relationships between crack width, 

material thickness, and shielding 

performance. (Lee et al. 2007)
73

7. COMPUTATIONAL MODELS FOR RADIATION DAMAGE OF 

CONCRETE

There are very few computational publications in the open literature addressing radiation 

damage to concrete (Hora and Patzak 2007). Most papers consider the shielding properties of 

concrete and its constituents under gamma and fast neutron irradiation as described in the 

previous sections. However an integrated approach to a computational assessment of thermal 

and irradiation damage of structural concrete in NPPs is still missing. 

An understanding of the transport of thermonuclides in reinforced and prestressed concrete 

members and structures is within reach if we consider recent computational advances in 

hygrothermal analysis of concrete spalling under rapid heating and drying (Xotta et al. 2013a) 

and the multiphysics area of durability of aging dry cask structures under sustained creep and 

shrinkage and transient environmental conditions. In fact, decomposing structural concrete into 

plain concrete and steel members and further decomposing plain concrete into swelling 

aggregate and shrinking cement paste at the mesoscale of observations are scenarios that 

have been and are still being pursued by several research teams at different scales of 

observation. However integration into a consistent multiscale analysis procedure for the 

thermonuclide transport and for the associated temperature and moisture transport and the 

resulting mechanical damage is still a lofty goal when full-scale concrete dry casks, biological 

shields, reactor pressure vessel support structures, and containment structures in NPPs are to 

be considered. For illustration, the recent papers by Pomaro et al. (2011a, 2011b) that address 

many of the underlying issues in a fairly pragmatic way will be reviewed.

7.1 INTRODUCTION

Irradiation in the form of fast and thermal neutrons, primary gamma rays, or gamma rays 

produced as a result of neutron capture affects concrete. Changes in the properties of concrete 

depend primarily on the behavior of concrete aggregates that can undergo a volume change 

when exposed to radiation. Radiation damage in concrete aggregates is caused by changes in 

the lattice structure of the minerals in the aggregates. Fast neutrons are mainly responsible for 

the growth caused by atomic displacements; such growth has been measured in certain 

aggregates. Quartz aggregates, made of crystals with covalent bonding, seem to be more 

affected by radiation than calcareous aggregates, which contain a weaker ionic bonding. 

Neutron fluences on the order of 1 × 1019 n/cm2 and gamma doses of 1010 rad have been noted 

in the literature as critical values for concrete strength degradation.

7.2 NUMERICAL MODELS FOR RADIATION DAMAGE EVOLUTION

An example of a computational model for evaluating radiation damage in concrete is the recent 

study by Pomaro et al. (Pomaro et al. 2011a, 2011b). Their research originated from a 

partnership between the Department of Civil, Environmental and Architectural Engineering of 

the University of Padova and the National Institute of Nuclear Physics (INFN) of Legnaro 

(Padova), aimed at the design of the Selective Production of Exotic Species (SPES) project. 

The work was related to construction of a next-generation nuclear facility dedicated to the 

production of special kinds of radioactive ion beams called “exotic species.” The overall 

objective of that activity was to investigate the behavior of concrete when exposed to nuclear 

radiation and to evaluate the consequent effects on its response by means of a proper definition 

of radiation damage in relation to the mechanical properties of the concrete.
74

Specifically, the subject of Pomaro’s study was the concrete shielding surrounding a nextgeneration nuclear facility; the purpose was to develop a full understanding of the physical 

aspects involved in the radiation exposure of concrete to nuclear radiation generated by the 

fission reactions due to a primary proton beam impinging a uranium-carbide target.

In that work, the available finite-element research code NEWCON3D (Salomoni et al. 2009) for 

fully coupled hygro-thermo-mechanical analyses of cementitious materials was upgraded to 

take into account the effects of nuclear radiation on the material via the introduction of a new 

damage variable. In particular, the required damage law was based on the experimental 

evidence of stiffness degradation (Hilsdorf 1978), as shown in Figure 7.1. The lower bound to

data in Figure 7.1 was used to indicate the impact of neutron fluence on modulus. The elastic

degradation acts in conjunction with the already-implemented chemo-thermo-mechanical 

damage formulation.

Figure 7.1. Modulus of elasticity of concrete after neutron 

irradiation (Ec) related to modulus of elasticity of 

untreated concrete (Eco) (See Figure 5.15 for key to 

data points). (Hilsdorf et al. 1978)

The effect of the mechanical and thermo-chemical damage is taken into account in a 

multiplicative way: 

D = 1 − (1 − Dm) (1 – Dtc) (7.1)

where Dm is the mechanical damage parameter, and Dtc is the thermo-chemical damage 

parameter that describes thermo-chemical material degradation at elevated temperatures, 

mainly due to microcracking and cement dehydration.

The extended damage model was developed by assuming that the nuclear radiation damage 

can activate a damage process that may be combined with the mechanical and thermochemical 
75

ones. Hence the multiplicative damage relationship is maintained when the total damage is 

redefined as, 

D = 1 – (1 – Dm) (1 − Dtc) (1 − Dr ) (7.2)

where Dr accounts for radiation damage, whose evaluation is empirically based on the stiffness 

degradation shown in Figure 7.1.

The implementation within the numerical code NEWCON3D results in a coupled system of finite 

element equations of linear momentum, mass, and energy balance:



















 +

=









































+









































.

. T

TG

HG

f c

T

h

u

0 0 TR

0 Q 0

0 0 0

T

h

u

0 TH TS

L I TP

K HU TU 







(7.3)

where 

u, h, and T bar are the nodal values of the basic variables (displacement, relative humidity,

and temperature;

HU and TU account for shrinkage and thermal dilation;

L and TP are coupling matrices representing the influence of the mechanical and thermal 

field on the hygral field;

Q is the diffusivity matrix;

TH the coupling matrix between the hygral and thermal field;

TS is the matrix of heat capacity;

TR is the matrix of thermal conductivity;

c accounts for creep;

HG represents the humidity variation load due to drying;

TG accounts for heat fluxes.

Additional analyses were performed with the Monte Carlo code FLUKA (Fassò et al. 2005), 

developed at CERN and INFN of Milan. FLUKA was used to describe the radiation and the 

temperature field for the design specifics of the SPES facility. Use of the Monte Carlo technique 

and the finite-element code were upgraded to take into account the radiation exposure effects

on concrete and the thermal aspect (i.e., the temperature rise in the shielding due to the 

deposition of radiation energy).

The results of the 3D simulations are summarized in Figure 7.2, which illustrates the change of 

humidity and temperatures during drying and cooling. Figure 7.3 depicts the evolution of 

temperatures after five irradiation cycles at three locations of the shielding structure.
76

Figure 7.2. Contour maps of relative humidity and temperature after 

1 h, 1 week, and 6 months. (Pomaro et al. 2011b)

Figure 7.3. Contours after the fifth irradiation cycle and evolution of 

temperature at three points of the target bunker. (Pomaro 

et al. 2011b).
77

Numerical analyses were also conducted to investigate radiation damage (i.e., the impact on 

concrete modulus of elasticity) and its progression in the concrete structure at ages up to 

50 years, assuming a neutron flux density of 1012 n cm-2 s-1

, which is nearly twice that expected 

at SPES (Pomaro et al. 2011a). Two extreme situations were considered for neutron energies: 

the case in which the entire integrated flux is prescribed by fast neutrons and the one that 

develops from thermal neutrons. This enabled the minimum and maximum threshold for 

radiation damage to be predicted under realistic neutron spectrum.

The sample analyzed was a 3.5 m long prism with a square cross section of 1 m2

. Both ordinary 

Portland cement concrete and heavyweight concretes were considered in the analyses. Results 

were developed in terms of damage parameter versus distance, with damage considered as 

being totally due to radiation. Following the diffusion theory for thermal neutrons and the twogroup theory for fast neutrons, the resulting damage behavior evolves as a function of depth for 

ordinary Portland cement concrete as presented in Figure 7.4 for times to 50 years. These 

results indicate that when fast neutrons are impinging, the damaged thickness is more 

pronounced and the level of damage induced by fast neutrons exceeds the one induced by 

thermal neutrons. Figure 7.5 summarizes damage versus thickness results obtained for 

heavyweight and ordinary Portland cement concretes for a fast and thermal neutron flux of 

1012 n cm-2 s-1 integrated over a 1 year exposure. The heavyweight concretes exhibited 

improved shielding properties relative to the ordinary (normal weight) Portland cement concrete.
78

Figure 7.4. Radiation damage progression with depth for ordinary Portland 

cement concrete under radiation fluence for up to 50 years: (a) 

fast neutrons, (b) thermal neutrons. (Pomaro 2011a)
79

Figure 7.5. Radiation damage progression with depth 

for ordinary Portland cement and 

heavyweight concretes under radiation 

fluence up to 1 year: (a) fast neutrons, (b) 

thermal neutrons. (Pomaro, 2011a)
80

7.3 DAMAGE-PLASTICITY MODELS FOR CONCRETE 

The Pomaro continuum damage model (Pomaro et al. 2011) raises the question of whether the 

mechanical characterization properly captures the main material features of concrete. In the 

earlier discussion it was observed that both radiation and temperature affect the aggregates of 

the composite material. In other terms, the granular nature of concrete is an important feature 

that is the basis of the so-called “Reynolds effect of volume dilatancy” under shearing. A brief

review of the limitations of continuum damage models follows. 

Surveys of recent concrete literature indicate an increasing number of proposals to combine 

plasticity and continuum damage mechanics to characterize the mechanical behavior of 

concrete materials (Lee and Fenves 1998; Carol et al. 2001; Grassl et al. 2002; Salari et al. 

2004; Tao and Phillips 2005; Grassl and Jirasek 2006; Voyiadjis et al. 2008). Thereby, it is 

understood that both formulations are dissipative in nature, even though they start from very 

different constitutive concepts. Irrespective of the thermodynamic setting, continuum damage 

mechanics resorts to reduction of the initial elastic stiffness properties in the form of a Cauchytype elastic secant relationship. In contrast, plasticity is understood in terms of strength and with 

reference to a description of an evolving yield surface separating elastic behavior from plastic 

behavior. Before examining the main shortcomings and limitations of the two dissipation models 

for concrete materials, let us briefly report the main relations for the different constitutive 

models:

1. Scalar Damage Models (Mazars 1984, Mazars and Hamon 2013):

(7.4)

2. Plasticity Models (Drucker and Prager 1952, 

(7.5)

3. Scalar Damage-Plasticity Models (Lee and Fenves 1998; Grassl and Jirasek 2006; Vojiadjis

2009):

(7.6)

In the case of scalar damage, the entire isotropic stiffness tensor is reduced by the scalar 

valued (1-D) factor, which retains the fundamental construct of the undamaged reference (e.g., 

linear isotropic scalar-valued). Consequently, the main deficiency of scalar damage models is

the initial reference stiffness, which does not change during progressive damage when frictional 

materials such as concrete exhibit strongly dilatant behavior near the maximum load resistance 

in uniaxial compression. Therefore, the concrete behavior exhibits a transition from compaction 

to dilatation under increasing axial compression. This transition of volume change is missing in 

scalar-valued damage models when the initial elastic Poisson’s ratio remains fixed. The other 

deficiency of continuum damage models is the reversibility of deformations under load-unload 

cycles, which does not account for permanent deformations. In the case of elastoplasticity, it is 

the question of the flow rule (and hence the plastic potential) whether the elastic reference 

stiffness is sufficiently modified by plastic dilatancy( i.e., the volumetric plastic strains). In fact, 

most pressure-sensitive plasticity formulations for granular materials resort to nonassociated 

flow rules to monitor the plastic dilatation of concrete. Clearly, the main attribute of plasticity is 

the irreversible nature of the plastic deformations under load cycles. Therefore, one may 
81

conclude that the main shortcoming of scalar damage models may be corrected by their 

combination with plasticity.

Figures 7.6 and 7.7 illustrate the results of three simple benchmark tests comparing the triaxial

response of three material models, (i) the classic elastic damage model of Mazars in its latest 

manifestation (2013), (ii) the associated plasticity model of Drucker and Prager (1952), and (iii) 

the so-called damage plasticity model of Lee and Fenves (1998) that has been implemented in 

Abaqus Unified FEA (DASSALT SYSTEMS 2013). Starting from identical elastic stiffness and 

strength properties in uniaxial tension and compression, and assuming linear softening of the 

tensile strength, the uniaxial calibration response in tension and compression are compared in 

Figure 7.6 together with the ensuing prediction of the volumetric behavior in simple shear. The 

results of the uniaxial tension test in the right of Figure 7.6 show similar behavior for all three 

material models; there is a slight difference in the postpeak branch, which may be attributed to 

the exponential rather than the linear softening assumption used in the damage model.

Figure 7.6. Comparison of Damage and Plasticity Models in Uniaxial 

Compression and Tension. (Xotta et al. 2013b)

Comparing Mazars’ Damage Model with the associated DP Model and the Damage-Plasticity 

Model in uniaxial compression results in the response diagrams shown on the left side of Figure 

7.6. The post- peak dilatancy of the two plasticity-based models is in stark contrast to the 

volumetric compaction of Mazars’ Damage model. In short, the scalar damage model does not 

capture the well-known transition of volumetric concrete behavior from compaction to expansion 

under axial compression. Moreover, the slightly different dilatations of the DP model and 

Combined-Damage-Plasticity (CDP) models may be explained by the different plastic flow rules 

used: the associated DP flow rule is used in the DP model; nonassociated flow is used in the 

“Barcelona” model proposed by Lubliner (1989) for the CDP model in Abaqus. 

For the case of simple shear, the three predictions are shown in the left part of Figure 7.7. There 

is a fundamental difference between Mazars damage model and the two plasticity models; 

specifically the right part of Figure 7.7 shows no volume change whatsoever for Mazars’ 

damage model, but the two plasticity models exhibit significant volume dilatation in shear. 

Moreover a slight difference between DP and CDP models due to the different yield functions 

used in the DP and the CDP models may be noticed in the peak strength and the postpeak 

behavior.
82

Figure 7.7. Comparison of Shear Response Predictions of Damage and 

Plasticity Models under Simple Shear. (Xotta et al. 2013b)

In conclusion, the brief study illustrates the drastic difference between scalar damage and 

pressure-sensitive plasticity models when the shear dilatancy is considered. In fact, the 

comparison demonstrates the need to combine damage with pressure-sensitive plasticity when 

dilatancy effects of frictional materials are to be captured. In summary, the lack of dilatancy of 

scalar damage models may be compared with the lack of dilatancy in pressure–insensitive von 

Mises plasticity.
83

8. DETECTION AND EVALUATION OF RADIATION DAMAGE IN NPP 

CONCRETE STRUCTURES

All commercial NPPs in the United States contain concrete structures whose performance and 

function are necessary for protection of the safety of plant operating personnel and the general 

public as well as the environment. A myriad of concrete-based structures are contained as a 

part of a light-water reactor (LWR) plant to provide foundation, support, shielding, and 

containment functions. Typical safety-related concrete structures contained in LWR plants may 

be grouped into four general categories: primary containments, containment internal structures, 

secondary containments/reactor buildings, and other structures.

The primary containment is a vital engineered safety feature of an NPP that is subjected to 

various operating and environmental stressors (e.g., ambient pressure fluctuations, temperature 

variations, and earthquakes). Concrete containments are metal-lined, reinforced concrete 

pressure-retaining structures that in some cases may be post-tensioned. The concrete 

containment includes the concrete shell and shell components, shell metallic liners, and 

penetration liners that extend the containment liner through the surrounding shell concrete. The 

reinforced concrete shell, which generally consists of a cylindrical wall with a hemispherical or 

ellipsoidal dome and flat base slab, provides the necessary structural support and resistance to 

pressure-induced forces. Leak tightness is provided by a steel liner fabricated from relatively 

thin plate material (e.g., 6 mm thick) that is anchored to the concrete shell by studs, structural 

steel shapes, or other steel products. In addition to the containment, a number of other concrete 

structures are contained as part of a light-water plant to support and protect safety-related 

systems and components. The other structures are primary, secondary, and biological shield 

walls as well as floors and supporting structures in the containments, reactor buildings, auxiliary 

(or intermediate) buildings, diesel generator buildings, intake structures, and service-water 

pump houses. The exterior walls and roofs, shield walls and buildings, interior floors that 

support heavy equipment and piping, and foundation mats are constructed of reinforced 

concrete. Some of the interior walls of these structures are constructed of (reinforced or 

unreinforced) masonry blocks. Beams and columns that support the floors are either of 

structural steel or reinforced concrete. Of these concrete structures, the structures most 

susceptible to the potential effects of radiation exposure would be those that provide biological 

shielding and support for the reactor pressure vessel. The presence of concrete degradation in 

these structures resulting from nuclear radiation could be determined through visual 

examination for signs of distress (e.g., cracking, volume change, and misalignment), application 

of nondestructive test methods, or removal and evaluation of concrete samples. However, prior 

to any assessments of concrete that has been subjected to nuclear radiation, precautions need 

to be taken to ensure that exposure of the personnel performing the assessments poses no risk 

to their health and safety.

8.1 NONDESTRUCTIVE TEST METHODS FOR CONCRETE AND REINFORCED 

CONCRETE STRUCTURES

Nondestructive testing (NDT) is a branch of materials science that utilizes noninvasive 

techniques to determine the integrity of a material, component, or structure or to quantitatively 

measure some characteristic of an object. Objectives of NDT are to:

• determine material properties; 

• detect, characterize, locate, and size discontinuities/defects; 
84

• determine the quality of manufacturing or fabrication of a component or structure; and 

• check for deterioration after a period of service for a component or structure.

With respect to application of NDT methods to concrete, they can indicate strength, density and 

quality; locate and characterize voids or cracks; locate steel reinforcement and indicate depth of 

concrete cover; and indicate corrosion of reinforcing materials. Tables 8.1 and 8.2 present NDT

methods to determine structural properties and to assess conditions of concrete and to 

determine material properties of hardened concrete in existing structures, respectively.

Table 8.1. Nondestructive test methods to determine structural properties and assess 

conditions of concrete (ACI 2013)

Property

Methods

Primary Secondary Comment

Reinforcement location Covermeter; Groundpenetrating radar (GPR) 

(ASTM D4748)

X-ray and γ-ray 

radiography

Steel location and 

distribution; concrete 

cover

Concrete component 

thickness

Intrusive probing impactecho (I-E) (ASTM 

C1383); GPR (ASTM 

D4748)

Verify thickness of 

concrete; provide more 

certainty in structural 

capacity calculations; I-E 

requires knowledge of 

wave speed, and GPR of 

dielectric constant in 

place of specific 

calibration to known 

thickness on site

Steel area reduction Intrusive probing; 

ultrasonic thickness gage 

(requires direct contact 

with steel)

Radiography Observe and measure 

rust and area reduction in 

steel; observe corrosion 

of embedded posttensioning components; 

verify location and extent 

of deterioration; provide 

more certainty in 

structural capacity 

calculations

Local or global strength 

and behavior

Load test, deflection or 

strain measurements

Acceleration, strain, and 

displacement 

measurements

Ascertain acceptability 

without repair or

strengthening;

deterrmine accurate load 

rating

Corrosion potentials Half-cell potential (ASTM 

C876)

Identification of likelyhood of corrosion activity

Corrosion rate Linear polarization 

(SHRP-S-324 and 

SHRP-S-330)

Corrosion rate of embedded steel; rate influenced 

by environmental

conditions
85

Table 8.1. (continued)

Property

Methods

Primary Secondary Comment

Location of 

delaminations, voids, and 

other hidden defects

Impact-echo; infrared 

thermography (ASTM 

D4788); impulseresponse; radiography; 

GPR (ASTM D6087); 

sounding (ASTM D4580)

Pulse-echo; UPV;

intrusive drilling, and 

borescope

Assessment of reduced 

structural properties; 

extent and location of 

internal damage and 

defects; sounding 

limited to shallow 

delaminations

ACI = American Concrete Institute

ASTM = American Society for Testing and Materials

SHRP = Strategic Highway Research Program

Table 8.2. Nondestructive test methods for determining material properties of 

hardened concrete in existing structures (ACI 2013)

Property

Possible Methods

Primary Secondary Comment

Compressive strength

(ACI 228.1R)

Cores for compression 

testing (ASTM 

C42/C42M and 

C39/C39M); pullout 

testing (post-installed) 

(ASTM C900)

Penetration 

resistance (ASTM 

C803/C803M)

Strength of in-place 

concrete; comparison of 

strength in different 

locations.

Relative compressive 

strength

Rebound number (ASTM 

C805/C805M); UPV

(ASTM C597)

In-place pulloff test 

(ACI 503.1R: BS 

1881-207) with 

appropriate 

calibration

Rebound number 

influenced by near surface 

properties; UPV gives 

average result through the 

thickness

Tensile strength Splitting-tensile strength 

of core (ASTM 

C496/C496M)

In-place pulloff test 

(ACI 503.1R; ACI 

228.1R; BS 1881-

207)

Assess tensile strength of 

concrete

Density Relative density (specific 

gravity) of samples 

(ASTM C642)

Nuclear gage

(ASTM 

C1040/C1040M)

Moisture content In-place moisture probes 

(ASTM F2170)

Preinstalled fiber optic 

relative humidity sensors

Nuclear gage

(ASTM 

C1040/C1040M); 

moisture content by 

drying (ASTM C642)

Static modulus of 

elasticity

Compression test of 

cores (ASTM 

C469/C469M)
86

Table 8.2. (continued)

Property

Possible Methods

Primary Secondary Comment

Dynamic modulus of 

elasticity

Resonant frequency 

testing of sawed 

specimens (ASTM 

C215)

Ultrasonic pulse 

velocity (ASTM 

C597); impactecho; spectral 

analysis of 

surface waves 

Requires knowledge of density 

and Poisson’s ratio (except ASTM 

C215); dynamic elastic modulus is 

typically greater than static elastic 

modulus

Shrinkage/expansion Length change of drilled 

or sawed specimens 

(ASTM C341/C341M)

Measure of incremental potential 

length change

Resistance to 

chloride penetration

90-day ponding test

(AASHTO T259); ASTM 

C1543) permit ion 

migration test

Electrical 

indication of 

concrete’s ability 

to resist chloride 

ion penetration 

(ASTM C1202)

Establishes relative susceptibility 

of concrete to chloride ion 

intrusion; assess effectiveness of 

chemical sealers, membranes, 

and overlays

Air content; cement 

content; and 

aggregate properties 

(scaling, alkali-silica 

reactivity, freezing 

and thawing

susceptibility)

Petrographic 

examination of concrete 

samples removed from 

structure (ASTM C856 

and ASTM 

C457/C457M); cement 

content (ASTM C1084)

Petrographic 

examination of 

aggregates 

(ASTM C294, 

ASTM 

C295/C295M)

Assist in determination of 

cause(s) of distress; degree of 

damage; quality of concrete when 

originally cast and current

Alkali-silica reactivity Cornell/SHRP rapid test 

(SHRP-C-315); 

petrography

Establish in field if observed 

deterioration is due to ASR

Carbonation, pH Phenolphthalein 

(qualitative indication); 

pH meter; Fiber optic pH 

sensors embedded in 

concrete

Other pH 

indicators (e.g., 

litmus paper)

Assess corrosion protection value 

of concrete with depth and 

susceptibility of steel reinforcement 

to corrosion; depth of carbonation

Fire damage Petrography; rebound 

number (ASTM 

C805/C805M)

SASW; UPV; 

impact-echo; 

impulseresponse

Rebound number permits 

demarcation of damaged surface

Freezing-andthawing damage (inplace)

Petrography SASW; Impulseresponse; UPV

Chloride ion content Acid-soluble (ASTM 

C1152/C1152M) and 

water soluble (ASTM 

C1218/C1218M)

Specific ion 

probe (SHRP-S328)

Chloride ingress increases 

susceptibility of steel 

reinforcement to corrosion

Air permeability SHRP surface airflow 

method (SHRP-S-329);

Torrent (1992) and 

autoclam airpermeability tests

Measure in-place permeability 

index of the near-surface 

concrete: within 0.6 in. (15 mm)
87

Table 8.2. (continued)

Property

Possible Methods

Primary Secondary Comment

Electrical resistance 

of concrete

AC resistance using 

four-probe resistance 

meter (FSTM 5-578)

SHRP surface 

resistance test 

(SHRP-S-327)

AC resistance useful for 

evaluating effectiveness of 

admixtures and cementitious 

additions; SHRP method useful 

for evaluating effectiveness of 

sealers

Water absorprion 

(sorptivity)

ISAT, Figg, covercrete, 

or autoclam sorptivity 

tests

In-place moisture content 

considered for interpreting the 

data

AASHTO = American Association of State Highway and Transportation Officials

AC = alternating current

ACI = American Concrete Institute

ASTM = American Society for Testing and Materials

BS = British Standard

SHRP = Strategic Highway Research Program

Visual and NDT methods are effective in identifying areas of concrete exhibiting distress but 

often cannot quantify the extent or nature of the distress. Quantifying and determining the 

nature of distress in concrete are generally accomplished through removal of cores or other 

samples using an established procedure. When core samples are removed from areas 

exhibiting distress, a great deal can be learned about the cause and extent of deterioration 

through property determinations and petrographic studies. Additional uses of concrete core 

samples include calibration of NDT devices, conduct of chemical analyses, visual examinations, 

determination of steel reinforcement corrosion, and detection of the presence of voids or cracks.

Information on methods for examining hardened concrete to characterize materials and to 

identify and quantify degradation is available (Walker et al. 2006). Qualitative or quantitative 

characterization of the microstructure can be done using techniques such as nanoindentation, 

mercury intrusion porosimetry, XRD, thermal gravimetric analysis, nuclear magnetic resonance, 

SEM, and energy-dispersive spectroscopy.

8.2 SPECIFIC CONDITIONS RELATED TO THE EFFECT OF RADIATION ON CONCRETE 

NPP concrete structures, particularly those related to the biological shield or support for the 

reactor pressure vessel, can be subjected to neutron and gamma radiation during their service 

lives. As the service lifetimes of these plants are being extended, the impact of radiation on the 

concrete structures may have increased significance that can affect their functional and 

performance requirements. Irradiation of concrete has the potential to impact the concrete’s 

microstructure as well as its mechanical and physical properties. An assessment of the impact 

of radiation on the concrete microstructure and properties would require removal of concrete 

samples from the areas of interest and application of techniques such as identified in Tables 8.1

and 8.2.

In addition to the impacts of radiation on the microstructure and properties of concrete, radiation 

can contribute to elevated temperatures in the concrete as well as increased potential for ASR

and concrete carbonation to occur. Techniques associated with identification of concrete that 

has experienced or is experiencing elevated temperatures, ASR, or carbonation may also 

provide indications that the concrete has been affected by radiation. 
88

8.2.1 Temperature effects

Gamma and neutron radiation can produce elevated temperatures and thermal gradients in 

concrete structures. Thermal loading of reinforced concrete can result in damage ranging from 

cosmetic blemishes to more serious effects (e.g., misalignment and distortions). Heating of 

concrete may result in a variety of structural changes such as cracking, spalling, debonding of 

aggregate, expansion, and mineralogical or chemical changes such as discoloration, 

dehydration, and dissociation. With respect to the cement paste, evaporation may occur as well 

as dissolution, dehydration, and dissociation of ettringite, gypsum, calcium hydroxide, calcium 

carbonate, and other phases, such as calcium silicate hydrates.

Information related to thermal effects on concrete structures generally is obtained through a 

condition assessment that involves a general visual inspection of the structure for signs of 

distress (e.g., cracking, spalling, deflections, distortions, misalignment, and exposed steel).

Nondestructive approaches identified for application to fire-damaged concrete can also be 

utilized as part of the assessment process. Table 8.3 presents a listing of possible NDT

approaches. Examples of approaches that have been used in the field include: rebound 

hammer, impact echo, indirect ultrasonic pulse measurements based on refraction of 

longitudinal waves to identify depth of damage, and drilling resistance based on measurement 

of either the thrust exerted to drill the material at constant feed rate or the work to drill a unit 

deep hole (Colombo and Felicetti 2007). Detailed information is obtained by removal of cores 

and conduct of laboratory tests to provide information on the mechanical properties or internal 

condition of the concrete. 

Table 8.3. Listing of possible nondestructive approaches for use in assessment of firedamaged concrete (Columbo and Felicetti 2007)

Average response of 

cover concrete

Point-by-point response of small 

samples

Special interpretation 

techniques

Hammer tapping

Schmidt rebound hammer

Windsor probe

Capo test

Building Research 

Establishment internal 

fracture

Ultrasonic pulse velocity

Small-scale mechanical testing

Differential thermal analysis 

Thermogravimetric analysis

Dilatometry (thermomechanical 

analysis)

Thermoluminescence

Porosimetry

Microcrack density analysis

Colorimetry

Petrographic analysis

Chemical analysis

Ultrasonic pulse velocity method

Impact echo

Sonic tomography

Modal analysis of surface waves

Ground-penetrating radar

Electric resistivity

8.2.2 ASR potential of existing concrete and detection

ASR has been identified at one U.S. NPP (i.e., Seabrook). It was noted in Section 6.1 of this 

report that there may be a coupling effect between radiation and ASR that can potentially 

accelerate ASR activity or cause ASR to occur with aggregates that are not normally reactive.

As plants age, the potential of ASR to occur in structures forming the biological shield or support 

for the reactor pressure vessel may increase as these structures are located in areas in which 

they are subjected to moderate elevated temperature in combination with radiation.
89

The primary method utilized for detection of ASR is through visual examinations indicating 

evidence of expansion, relative movements between structural elements, and cracking. ASR 

capable of being detected visually, however, is probably in a fairly advanced stage of 

development. Removal, examination, and testing of concrete samples from suspect areas can 

be used to confirm the presence of ASR. Petrographic examinations of thin sections of 

aggregate materials, and tests developed for identification of ASR reactivity products (e.g., use 

of sodium cobalt nitrite solution to detect potassium and uranyl acetate to detect sodium) are 

examples of examination and testing methods (Walker et al. 2006).

Testing methods are available for evaluating the potential for new concrete to undergo ASR 

expansion; however, there is no standard testing method for existing concrete. Therefore it is 

desirable to develop a reliable testing protocol to evaluate the ASR potential and its occurrence 

for concrete in the existing structures. 

Since ASR requires the participation of both alkali in cement paste and reactive silica in 

aggregate in the presence of moisture, tests need to be performed to examine the remaining 

reaction potentials of the two components. This is necessary because alkali in cement paste 

may be exhausted at a certain age of a structure, but at a later time more alkali may be supplied 

from the environment. In that case, the ASR will resume and continue if there is sufficient 

reactive silica in the aggregate.

Two specific tests can be conducted to examine the remaining reaction potential. One is to 

measure the expansion of concrete specimens from an existing structure in 100% relative 

humidity (e.g., at 40°C); the other is to measure the expansion of the concrete specimens 

submerged in NaOH solutions (e.g., 1M concentration, at 40°C). Comparing the difference in 

the two measured expansions will show the remaining ASR expansion potential of the existing 

concrete. In these two tests, the first one (the 100% relative-humidity test) is to determine 

whether the concrete still has sufficient alkali left to form ASR gel; the second one (the NaOH 

test) investigates the remaining reaction potential of the aggregate used in the concrete. Both 

tests would require the removal of samples from the concrete structures of interest.

8.2.3 Carbonation

In Section 5.1, it was indicated that radiation can produce carbonation of concrete that is 

internal to the concrete (i.e., it does not start at exposed concrete surface). The extent of 

concrete carbonation can be easily determined either in situ or in a laboratory by treating the 

surface of a freshly removed concrete sample with phenolphthalein. The carbonated portion will 

be uncolored. Incrementally drilled concrete powder samples also can be extracted and sprayed 

with phenolphthalein. In addition, acid etching or microscopic analysis of thin sections cut from 

the concrete can assess the presence of carbonation.

91

9. RADIATION EXPOSURE LIMITS FOR NPP CONCRETE

In assessing the reduction of concrete strength under nuclear radiation, several threshold 

values were used for neutron and gamma radiation. For instance, 1 × 1020 n/cm2 for fast 

neutrons and 2 × 1010 rad (2 × 108 Gy) for gamma rays have been used as the reference levels 

of nuclear radiation in concrete [Section III, Division 2 of the American Society of Mechanical 

Engineers (ASME) Boiler and Pressure Vessel Code (ASME 2010), also see Fillmore 2004].

Concrete structures have been regarded to be sound as long as the levels of radiation do not 

exceed the reference levels. IAEA-TECDOC-1025 (IAEA 1998) lists 1 × 1019 n cm-2 for fast 

neutrons and 1010 rad (108 Gy) for gamma rays.

According to the available data (Mirhosseini 2010, Fillmore 2004), primary shields may 

experience neutron radiation fluence beyond 5 × 1019 n cm-2

. Concrete reactor vessels may be 

exposed to thermal neutron, fast neutron, and gamma radiation with the estimated values of 

6 × 1019 n cm-2

, 2-3 × 1018 n cm-2 and 1 × 1011 rad (1 × 109 Gy), respectively, after 30 years. The 

tests results from the British Experimental Pile “0” Reactor by United Kingdom Atomic Energy 

Authority show a constant flux around 3 × 1011 n cm-2 s-1

, which is equivalent to a fluence of 

3.78 ×1020 n cm-2 after 40 years (Alexander 1963).

Thresholds were set by various national standards for radiation exposure in 1960s, 1970s, and 

1980s to reduce risk of shielding concrete property degradation due to radiation.

• The ASME Pressure Vessel and Piping Code (Section III, Division 2) (ASME 2010) provides 

a limit for neutron exposure of 1 × 1020 n cm-2

.

• The British Specification for Prestressed Concrete Pressure Vessels for Nuclear Reactors

(British Standards Institute 1973) notes that although it was not possible to quantify 

precisely the effects of neutron radiation on the properties of concrete, it was noted that for 

radiation fluence less than 5 × 1019 n cm-2 the effects of radiation on concrete properties is 

not considered to be significant.

• According to American National Standards Institute (ANSI/ANS 2006a), compressive 

strength and modulus of elasticity are degraded for neutron exposure > 1 × 1019 n cm-2 or to 

a gamma radiation dose exceeding 1010 rad (108 Gy).

• In the Engineering Compendium on Radiation Shielding (Jaeger 1975) it is stated that 

damage to concrete by nuclear heating is more important than the effects of radiation 

damage. It was recommended that the incident neutron flux density should be limited to 

5 × 1019 n cm-2 s-1 and for 1 MeV gamma rays it should be limited to 4 × 1010 gamma 

photons cm-1 s-1

.

In Japan, Ichikawa and Kimura (2007) considered that γ-rays do not affect the nature of 

concrete up to a dose of 1 × 1010 Gy. Irradiation by fast neutrons at a fluence of more than 

5 × 1019 n cm-2 causes the deterioration of concrete due to the expansion of aggregates and the 

shrinkage of cement paste.

In an Electric Power Research Institute (EPRI) report (EPRI 2012), the threshold value for 

gamma radiation was considered as 1010 rad (108 Gy).

93

10.SUMMARY AND CONCLUSIONS

The current understanding of the effects of irradiation on concrete materials is summarized in 

this document. Much of the research related to this topic was conducted from the 1960s into

1970s in support of development of prestressed concrete reactor vessels for high-temperature 

gas-cooled reactors and radioactive waste storage facilities. Concrete structures used as 

biological shields and as support for the reactor pressure vessel in current LWR plants are 

subjected to two types of radiation, gamma and neutron.

Gamma radiation refers to electromagnetic radiation of high frequency; high energy per photon. 

Gamma rays typically have energies above 100 keV and wavelengths less than 10 picometers. 

The absorbed dose in a small volume element is equal to the energy imparted by ionizing 

radiation to the material in a volume element divided by the mass of the material in the volume 

element. The unit of absorbed dose is a gray having units of joule kg-1

: 1 Gy = 100 rad. The 

photons interact with the shielding concrete in different ways: photoelastic effect, Compton 

scattering, and pair production. Gamma rays have an insignificant effect on solid materials 

composed of ionic and metallic bonds, but they can result in the destruction of anisotropic 

chemical bonds such as covalent bonds. Water in the concrete can be decomposed by gamma 

rays by a process called radiolysis and can be converted to hydrogen, oxygen, and hydrogen 

peroxide. Water can also be removed from the concrete by evaporation due to heat generated 

by gamma radiation. Because most of the water in concrete is contained in the cement paste, 

gamma radiation has a greater effect on the cement paste than it has on the aggregate 

materials.

In the instance of neutron irradiation, the neutrons do not interact with the electrons but with the 

nuclei of atoms. The amount of neutron radiation absorbed is typically expressed by fluence, 

which is a measure of the total number of neutrons that penetrate a unit area expressed as 

n cm-2

. Because neutrons interact with the nuclei of atoms, the lattice spacing within the material 

may change after the collisions. Therefore, the neutrons have a more significant effect on dense 

and well-crystallized materials (e.g., aggregates) than on randomly structured materials with 

high porosity (e.g., cement paste). Under nuclear radiation, the internal structure of some 

aggregates can be converted from an ordered crystalline structure to a disordered structure, 

resulting in a decrease in specific gravity and an increase in volume.

Interactions between nuclear radiation and the internal structure of concrete produce geometric 

changes of the structure resulting from displacements of atoms from their lattice sites and phase 

transformations of the concrete constituents resulting in a reduction in porosity and/or formation 

of microcracks. These interactions result in changes in the various concrete properties. As the 

gamma radiation dose increases, the specific pore surface and the porosity of the Portland 

cement paste have been shown to decrease. It also has been shown that as the gamma 

radiation dose increases the concentration of calcite increases resulting in carbonation of the 

concrete. Experimental studies have shown that the reactivity of silica-rich quartz to alkali can 

be significantly increased by exposure to nuclear radiation. As a result, irradiation can produce 

ASR in concretes containing aggregates that are typically nonreactive.

Results indicating the effect of irradiation on the mechanical and physical properties of concrete 

are very limited. The impact of neutron and gamma radiation on concrete properties is 

dependent on the concrete mix proportions, type of cement, and type of aggregate. Tests 

presenting the effect of gamma dose on the compressive strength of mortar and concrete 

indicated that the compressive strength of concrete that had received a dose of 5 × 105 Gy was 

approximately 10% less than that of unirradiated specimens. However, the data scatter 
94

exhibited by these results was relatively large and the data set relatively small so that 

statistically valid conclusions could not be derived. More recent results indicate that the 

compressive strength of concrete is relatively unaffected by fast neutron irradiation at exposures 

up to fluences of 2 × 1019 n cm-2

. Flexural- and splitting-tensile strength results obtained from 

specimens that had been subjected to gamma radiation exhibited a trend for the tensile strength 

to decrease with increasing dose, with the reduction on the order of 10% or less at a gamma 

dose of 5 × 105 Gy relative to results obtained from unirradiated specimens. However, the data 

scatter exhibited by these results was relatively large and the data set relatively small so that 

statistically valid conclusions could not be derived. It was noted by one researcher that, in 

general, neutron fluences greater than 1 × 1019 n cm-2 will lead to a reduction in the tensile 

strength of concrete. The static modulus of elasticity has been shown to start decreasing slightly 

at a fast neutron fluence of 0.6 × 1019 n cm-2

; however, at 1.2 × 1019 n cm-2 it was still within 20% 

of the modulus values obtained by testing unirradiated specimens. Relatively short term (i.e., 

approximately 300 day) creep (load = 10 N/mm2

) and shrinkage test results obtained from a 

limestone aggregate concrete representative of that used for the Oldbury Power Station 

concrete pressure vessel subjected to an average gamma dose rate of 11.4 × 103 rad hr-1

indicated that the creep of the irradiated concrete was less and that the shrinkage was more 

than results obtained from the unirradiated specimens. The effect of radiation on the coefficients 

of thermal expansion and thermal conductivity of concrete was reported in the literature to be 

similar to that obtained from unirradiated temperature-exposed concrete. Results indicated that 

the thermal conductivity of concrete subjected to neutron radiation decreased to about 60% of 

the value for unirradiated concrete and that it remained relatively constant at neutron radiation 

fluences up to about 3.7 × 1019 n cm-2

.

The combination of mechanical loading, thermal effect, moisture content, radiation effect, and

time determines the response behavior of concrete in nuclear environments. However, the 

coupling effects among the physical and chemical driving forces need to be studied. Although 

limited, theoretical methods are available that can be used to model the effect of nuclear 

radiation on concrete. A few multiscale modeling methods exist that can address the 

deterioration mechanisms taking place, ranging from the micrometer-level for the internal 

structure of concrete to the meter-level for concrete containment structures. A computational 

model for evaluating radiation damage in concrete has been used related to the design of a 

next-generation nuclear facility for the Selective Production of Exotic Species Project in Italy. A 

finite-element code for fully coupled hygro-thermo-mechanical analyses of cementitious 

materials was upgraded in this study to include the effects of nuclear radiation through a 

thermochemical damage parameter (i.e., empirically based on resulting stiffness degradation).

The resulting damage law accounts for the coupling of moisture, heat transfer, and the 

mechanical field in concrete that was treated as a fully coupled porous medium. This enabled 

the development of the damage front in the concrete shielding wall of the facility under neutron 

irradiation and provided results within the wall thickness for long-term radiation spans and 

several concrete mixtures to evaluate shielding properties. Contour maps of damage relative to 

1 month, 1 year, 5 years, and 50 years of irradiation were developed.

In-service testing and inspection programs for U.S NPP concrete structures have the primary 

goal of ensuring that the structures have sufficient structural margins to continue to perform in a 

reliable and safe manner. A secondary goal is to identify environmental stressors or aging 

factors before they reach sufficient intensity to potentially degrade structural components. 

Nondestructive examination methods having application to reinforced-concrete structures are 

identified. Candidate inspection and testing methods having potential application to detection of 

the effects of radiation on concrete materials are noted (e.g., temperature, ASR, and 

carbonation).
95

Limitations on radiation exposure contained in several codes are identified. The limitations are 

placed on exposure to ensure that the concrete properties, notably strength, do not deteriorate 

as a result of exposure to neutron and gamma radiation. Reference levels provided in the ASME 

Boiler and Pressure Vessel Code of 1 × 1020 n cm-2 for fast neutrons and 2 × 1010 rad (2 × 108

Gy) were based primarily on a paper published by Professor Hilsdorf et al. summarizing 

available data. Revisiting the data used by Professor Hilsdorf et al. indicates that all of the data 

at the higher exposure levels (i.e., fast neutron fluence > 1 × 1019 n cm-2

) were obtained using 

materials or test conditions that were not representative of LWRs.

Results presented in this report indicate that data related to the effect of gamma and neutron 

radiation on concrete mechanical and physical properties are very limited, particularly data 

representative of conditions associated with LWR plants. There is a need to improve and 

expand the database on irradiated concrete degradation and to develop an improved 

understanding of the basic mechanisms causing radiation damage. Expansion of the database 

can occur through laboratory studies as well as through testing concrete samples obtained from 

aged NPPs. In parallel with this, calculations should be performed to identify the neutron and 

gamma fields that develop in the concrete structures of interest, biological shield and support for 

the reactor pressure vessel, after various periods of operation (e.g., 40, 60, 80, and 100 years).

The estimated exposure levels for the various plant designs can then be compared to exposure 

limits for neutron and gamma radiation that have been developed from data in the database to 

determine whether exposure levels are high enough to degrade the concrete, or at what period 

of operation an exposure limit will be exceeded. 

97

11. REFERENCES

1. Abdo, A.E., and Amin, E. (2001) “Distribution of Temperature Rise in Biological Shield due 

to Thermal Neutrons”, Annals of Nuclear Energy, 28, 275-283.

2. Abdo, A.E., Kansouh, W.A., and Megahid, R.M. (2002) “Investigation of Radiation 

Attenuation Properties for Baryte Concrete”, Jpn. J. Appl. Phys., 41, 7512-7517.

3. Acevedo C.E., and Serrato, M.G. (2010) “Determining the Effects of Radiation on Aging 

Concrete Structures of Nuclear Reactors – 10243”, WM2010 Conference Proc., March 7-

11, Phoenix, AZ.

4. ACI (12013) “Report on Nondestructive Test Methods for Evaluating Concrete in 

Structures, ACI 228.2R-13, American Concrete Institute”, Farmington Hills, MI. 

5. Alexander, S. C. (1963) “Effects of irradiation on concrete: Final results”, United Kingdom 

Atomic Energy Authority.

6. Alhajali, S., Kharita, M.H., Naoom, B., Yousef, S., and AlNassar, M. (2009) “Estimation of 

the Activation of Local Reactor Shielding Concretes”, Progress in Nuclear Energy, 51, 

374-377. 

7. ANSI/ANS (1987) “Program for Testing Radiation Shields in Light Water Reactors 

(LWRS)”, ANSI/ANS-6.3.1-1987, American Nuclear Society, La Grange Park, Illinois.

8. ANSI/ANS (1999) “Neutron and Gamma-Ray Cross Sections for Nuclear Radiation 

Protection Calculations for Nuclear Power Plants”, ANSI/ANS-6.1.2-1999, American 

Nuclear Society, La Grange Park, Illinois.

9. ANSI/ANS (2006a) “Nuclear Analysis and Design of Concrete Radiation Shielding for 

Nuclear Power Plants”, ANSI/ANS-6.4-2006, American Nuclear Society, La Grange Park, 

Illinois.

10. ANSI/ANS (2006b) “Specification for Radiation Shielding Materials”, ANSI/ANS-6.4.2-

2006, American Nuclear Society, La Grange Park, Illinois. 

11. ASME (2010) “ASME Boiler and Pressure Vessel Code”, American Society of Mechanical 

Engineers, New York.

12. Bar-Nes, G., Katz, A., Peled, Y., and Zeiri, Y. (2008) “The Combined Effect of Radiation

and Carbonation on the Mobilization of Sr and Cs Ions in Cementitious Pastes”, Materials 

and Structures, 41, 1563-1570.

13. Bashter, I.I. (1997) “Calculation of Radiation Attenuation Coefficients for Shielding 

Concretes”, Ann. Nucl. Energy, 24(7), 1389-1401.

14. Basyigit, C., Akkurt, I., Altindag, R., Kilincarslan, S., Akkurt, A., Mavi, B., and Karaguzel, R. 

(2006) “The Effect of Freezing-Thawing (F-T) Cycles on the Radiation Shielding Properties 

of Concretes”, Building and Environment, 41, 1070-1073.
98

15. Bouniol, P., and Aspart, A. (1998) “Disappearance of Oxygen in Concrete under 

Irradiation: The Role of Peroxides in Radiolysis”, Cement and Concrete Research, 28(11), 

1669-1681.

16. Bouniol, P. (2011) “Radiation Effects in Concrete”, ppt presentation, SFEN L’irradiation 

dans les matériaux des réacteurs nucléaires, 22 Novembre 2011.

17. Brooks, J.J. (2005) “30-year Creep and Shrinkage of Concrete”, Magazine of Concrete 

Research, 57(9), 545-556.

18. Callan, E. J. (1953) “Concrete for Radiation Shielding”, Journal of the American Concrete 

Institute , Title No. 50-2, 17-44.

19. Carol, I., Rizzi, E., and Willam, K. (2001) “On the Formulation of Anisotropic Elastic 

Degradation. Part I: Theory based on a Pseudo-logarithmic Damage Tensor Rate”, Int. J. 

Solids and Struct., 38, 491–518.

20. CEMENBUREAU (2004) “Improving Fire Safety in Tunnels: The Concrete Pavement 

Solution,” Brussels, Belgium 

(http://www.efca.info/downloads/Improving_Fire_Safety_In_Tunnels.pdf).

21. Colombo. M. and Felicetti (2007) “New NDT Techniques for the Assessment of FireDamaged Concrete Structures,” Fire Safety Journal, 42, 461-472.

22. DASSALT SYSTEMS (2013) “Abaqus Unified FEA,” Waltham, MA 

(http://3ds.com/products/simulia/portfolio/abaqus/overview)

23. Dubrovskii, V.B., Ibragimov, S.S., Ladygin, A.Y., and Pergamenshchlk, B.K. (1966) “Effect 

of Neutron Irradiation on Some Properties of Heat-Resistant Concretes”, Atomnaya 

Energiya, 21(2), 108-112.

24. Dubrovskii, V.B., Ibragimov, S.S., Kulakovskii, M.Y., Ladygin, A.Y., and Pergamenshchlk, 

B.K. (1967) “Irradiation Damage in Ordinary Concretes”, Atomnaya Energiya, 23(4), 310-

316.

25. Dubrovskii, V.B., Ibragimov, S.S., M.Y., Ladygin, A.Y., Kulakovskii, M.Y., and 

Pergamenshchlk, B.K. (1968) “Radiation Stability of Serpentine Concrete”, Atomnaya 

Energiya, 25(6), 515.

26. Drucker, D.C., and Prager, W. (1952) “Soil Mechanics and Plastic Analysis for Limit 

Design”, Quarterly of Applied Mathematics, 10(2), 157–165.

27. El-Khayatt, A.M. (2011) “NXcom - A Program for Calculating Attenuation Coefficients of 

Fast Neutrons and Gamma-rays”, Ann. Nucl. Energy, 38 (1), 128–132.

28. EPRI (2012) “Effect of Radiation on Concrete – A Literature Survey and Path Forward”, 

Final Report 1025584, Electric Power Research Institute, Palo Alto, CA.

29. Eskandari-Ghadi, M., Zhang, W.P., Xi, Y., and Sture, S. (2013) “Modeling of Moisture 

Diffusivity of Concrete at Low Temperatures”, Journal of Engineering Mechanics, ASCE, 

139(7), 1-13.
99

30. Fassò A., Ferrari A., Ranft J., Sala P.R. (2005) “FLUKA: A Multi-particle Transport Code”, 

CERN-2005- 10, INFN/TC_05/11, SLAC-R-773. 

31. Feldman, R.F., and Sereda, P.J. (1968) "A Model for Hydrated Portland Cement Paste as 

Deduced from Sorption-Length Change and Mechanical Properties", Materials and 

Structures, RILEM, Paris, 1(6), 509-519; (1969), 2(8), 155-162.

32. Fillmore, D.L. (2004) “Literature Review of the Effects of Radiation and Temperature on 

the Aging of Concrete”, INEEL/EXT-04-02319, Idaho National Engineering and 

Environmental Laboratory, Bechtel BWXT Idaho, LLC.

33. Fujiwara, K., Ito, M., Sasanuma, M., Tanaka, H., Hirotani, K., Onizawa, K., Suzuki, M., and 

Amezawa, H. (2009) “Experimental Study of the Effect of Radiation Exposure to 

Concrete”, Proc. of 20th Int. Conf. on Stru. Mech. In Reactor Tech. (SMiRT 20), Espoo, 

Finland, Aug. 9-14, SMiRT 20 – Division I, Paper 1891.

34. Gerward, L., Guilbert, N., Jensen, K.B., Levring, H. (2004) “WinXCom – A Program for 

Calculating X-ray Attenuation Coefficients”, Radiat. Phys. Chem. 71, 653–654.

35. Glasser, F.P. (1992) “Chemistry of the Alkali-Aggregate Reaction”, The Alkali-silica 

Reaction in Concrete. Ed. R.N. Swamy, Van Nostrand Reinhold, New York, 30-54.

36. Grassl, P., Lundgren, K., and Gylltoft, K. (2002) “Concrete in Compression: a Plasticity 

Theory with a Novel Hardening Law”, International Journal of Solids and Structures, 39, 

5205-5223.

37. Grassl, P., and Jirasek, M., (2006) “Damage-plastic Model for Concrete Failure”,

International Journal of Solids and Structures, 43(22-23), 7166-7196.

38. Gray, B.S. (1972) “The Effect of Reactor Radiation on Cements and Concrete”, 

Commission of the European Communities, Luxembourg, 17-39.

39. Granata, S. and Montagnint, A. (1972) “Studies on Behavior of Concrete under 

Irradiation”, ACI Special Publication SP-34: Concrete for Nuclear Reactors, Vol. II, 1163-

1172.

40. Hala, J., and Navratil, J.D. (2003) “Radioactivity, Ionizing Radiation, and Nuclear Energy”, 

Konvoj, Spol. S.R.O. (Ltd.), Brno, Czech Republic.

41. Hilsdorf, H.K. (1967) “A Method to Estimate the Water Content of Concrete Shields”, 

Nuclear Engineering and Design, 6, 251-263.

42. Hilsdorf, H. K., Kropp, J., and Koch, H. J. (1978) “The Effects of Nuclear Radiation on the 

Mechanicals Properties of Concrete”, American Concrete Institute Special Publication SP55, 223-251.

43. Hobbs, D.W. (1987) “Deleterious Expansion of Concrete due to ASR,” Magazine of 

Concrete Research, 38(137), 191-205.

44. Hora, Z., and Patzak, B. (2007) “Analysis of Long-Term Behavior of Nuclear Reactor 

Containment”, Nuclear Engineering and Design, 237, 253-259. 
100

45. IAEA (1998) “Assessment and Management of Ageing of Major Nuclear Power Plant 

Components Important to Safety: Concrete Containment Buidings”, IAEA-TECDOC-1025, 

International Atomic Energy Agency, p. 162,

46. Ichikawa, T., and Koizumi, H. (2002) “Possibility of Radiation-Induced Degradation of 

Concrete by Alkali-Silica Reaction of Aggregates”, Journal of Nuclear Science and 

Technology, 39(8), 880-884. 

47. Ichikawa, T., and Kimura, T. (2007) “Effect of Nuclear Radiation on Alkali-Silica Reaction 

of Concrete”, Journal of Nuclear Science and Technology, 44(10), 1281-1284. 

48. Jaeger, R.G. (Ed.) (1968) “Engineering Compendium on Radiation Shielding”, Vol. I, Heat 

Generation by Neutrons, Springer-Verlag, New York.

49. Jaeger, R.G. (Ed.) (1975) “Engineering Compendium on Radiation Shielding, Vol. II, 

Shielding Materials, Springer-Verlag, Berlin.

50. Jennings, H.M., and Xi, Y. (1992) "Relationships Between Microstructure and Creep and 

Shrinkage of Cement Paste", Material Science of Concrete, III, Ed. J. Skalny, The Amer. 

Cer. Soc., Westerville, OH, 37-69.

51. Jennings, H.M., (1986) "Aqueous Solubility Relationships for Two Types of Calcium 

Silicate Hydrate," Journal of American Ceramic Society, 69, 614-618.

52. Jennings, H.M., and Bullard, J.W. (2011) “From Electrons to Infrastructure: Engineering 

Concrete from the Bottom Up”, Cement and Concrete Research, 41, 727-735,

53. Kaplan, M. F. (1989) Concrete Radiation Shielding: Nuclear Physics, Concrete Properties, 

Design and Construction. Essex: Longman Scientific & Technical.

54. Kelly, B.T., Brocklehurst, J.E., Mottershead, D., McNearney, S., and Davidson, I. (1968) 

“The Effect of Reactor Radiation on Concrete”, United Kingdom Atomic Energy Authority, 

G.R.D., Rislay, Nr. Warrington, Lancs, Nr. 8, 237-265.

55. Kharita, M.H., Takeyeddin, M., Alnassar, M., and Yousef, S. (2008) “Review on the 

Addition of Boron Compounds to Radiation Shielding Concrete”, Progress in Nuclear 

Energy, 53, 207-211.

56. Kharita, M.H., Yousef, S., and AlNassar, M. (2010) “The Effect of the Initial Water to 

Cement Ratio on Shielding Properties of Ordinary Concrete”, Progress in Nuclear Energy, 

52, 491-493.

57. Kharita, M.H., Yousef, S., and AlNassar, M. (2011) “Development of Special Radiation 

Shielding Concretes using Natural Local Materials and Evaluation of their Shielding 

Characteristics”, Progress in Nuclear Energy, 50, 33-36.

58. Kontani, O., Ichikawa, Y., Ishizawa, A., Takizawa, M., & Sato, O. (2010) “Irradiation Effects 

on Concrete Structures”, Proc. of International Symposium on the Ageing Management & 

Maintenance of Nuclear Power Plants, 173-182.
101

59. Kontani, O., Ichikawa, Y., and Ishizawa, A. (2011) “Irradiation Effects on Concrete 

Durability of Nuclear Power Plants”, Proceedings of ICAPP 2011, Nice, France, May 2-5, 

Paper 11361.

60. Kurtis, K.E., Collins, C.L., Monteiro, P.J.M. (2002) “The Surface Chemistry of the AlkaliSilica Reaction: A Critical Evaluation and X-ray Microscopy”, Concrete Science and 

Engineering, 4.

61. Lee, J., and Fenves, G.L. (1998) “Plastic-Damage Model for Cyclic Loading of Concrete 

Structures”, Journal of Engineering Mechanics, ASCE, 124(8), 892–900.

62. Lee, C.M., Lee, K.J., and Cho C.H. (2008) “Spalling Effect on Gamma-ray Shielding 

Performance in MACSTOR-400 (Korean CANDU spent fuel storage module), Prog. Nucl. 

Energy, doi:10.1016/j.pnucene.2008.02.007.

63. Lee, C.M., Lee, Y.H., and Lee, K.J. (2007) “Cracking Effect on Gamma-ray Shielding 

Performance in Concrete Structures”, Progress in Nuclear Energy, 49, 303-312.

64. Lee. S.Y., Daugherty, A.M. and Broton, D.J. (2013) “Assessing Aggregates for RadiationShielding Concrete,” Concrete International, 35(5), 31-38.

65. Lowinska-Kluge, A., and Piszora, P. (2008) “Effect of Gamma Irradiation on Cement 

Composites Observed with XRD and SEM Methods in the Range of Radiation Dose 0 –

1409 MGy”, Acta Physica Polonica A, 114, 399 – 411.

66. Lubliner, J., Oliver, J., Oller, S., and Oñate, E. (1989) “A Plastic-Damage Model for 

Concrete”, International Journal of Solids and Structures, 25, 299–329.

67. Martinez-Barrera, G., Urena-Nunez, F., Gencel, O., and Brostow, W. (2011) “Mechanical 

Properties of Polypropylene-fiber Reinforced Concrete after Gamma Irradiation”, 

Composite: Part A, 42, 567-572.

68. Mazars, J. (1984) “Application de la mécanique de l’endommagement au comportement 

non linéaire et à la rupture du béton de structure”, Thèse de Doctorat d’Etat, Université 

Paris VI, France.

69. Mazars, J. and Hamon, F. (2013) “A New Model to Forecast the Response of Concrete 

Structures under Severe Loadings: the μ Damage Model”, FramCOS-8 International 

Conference on Fracture Mechanics of Concrete and Concrete Structures, March 10-14, 

Toledo, Spain.

70. McDowall, D.C. (1971) “The Effect of Gamma Irradiation on the Creep Properties of 

Cncrete”, Commission of the European Communities, Luxembourg, 55-69.

71. Mehta, P.K., and Monteiro, P.J.M. (1993) “Concrete, Structure, Properties, and Materials”, 

Prentice Hall, Englewood Clifs, New Jersey.

72. Mirhosseini, S.S. (2010) “The Effect of Nuclear Radiation on Aging Reinforced Concrete 

Structures in Nuclear Power Plants”, MS Thesis, University of Waterloo, Ontario, Canada.
102

73. Monteiro, P.J.M.; Wang, K.; Sposito, G.; dos Santos, M.C.; and de Andrade, W.P. (1997) 

“Influence of Mineral Admixtures on the Alkali-Aggregate Reaction”, Cement and Concrete 

Research, 27(12), 1899.

74. Pellenq, R.J.M., Kushima, A., Shahsavari, R., Van Vliet, K.J., Buehler, M.J., Yip, S., and Ulm, 

F,-J. (2009) “A Realistic Molecular Model of Cement Hydrates”, Proc. National Academy of 

Science, U.S.A., 106, 16102-16107. 

75. Powers, T.C., and Brownyard, T.L. (1948) “Studies of the Physical Properties of Hardened 

Portland Cement Paste”, PCA Res. Bull. 22, Chicago.

76. Pomaro, B., Salomoni, V.A., Gramegna, F., Prete, G., and Majorana, C.E. (2011a) 

“Radiation Damage Evaluation on Concrete within a Facility for Selective Production of 

Exotic Species (SPES Project), Italy”, J. of Hazardous Materials, 194, 169-177.

77. Pomaro, B., Salomoni, V.A., Gramegna, F., Prete, G., and Majorana, C.E. (2011b) 

“Radiation Damage Evaluation on Concrete Shielding for Nuclear Physics Experiments”, 

Ann. Solid Struct. Mech., 2, 123-142.

78. Sahin, R., Polat, R., Icelli, O., and Celik C. (2011) “Determination of Transmission Factors 

of Concretes with Different Water/Cement Ratio, Curing Condition, and Dosage of Cement 

and Air Entraining Agent”, Annals of Nuclear Energy, 38. 1505-1511.

79. Sakr, K., and El-Hakim, E. (2005) “Effect of High Temperature or Fire on Heavy Weight 

Concrete Properties”, Cement and Concrete Research, 35, 590-596.

80. Salari, M.R., Saeb, S., Willam, K.J., Patchet, S.J., and Carrasco, R.C. (2004) “A coupled 

Elastoplastic Damage Model for Geomaterials”, Comput. Methods Appl. Mech. Engrg., 

193 (27-29), 2625-2643.

81. Salomoni, V., Majorana, C., Mazzucco, G. Xotta, G. and Khoury, G.A. (2009) “Multiscale 

Modelling of Concrete as a Fully Coupled Porous Medium,” Ch. 3 in Concrete Materials: 

Properties, Performance and Applications, NOVA Science Publishers Inc., ISBN: 978-1-

60741-250-2, 171-231, 2009. 

82. Saouma, V., and Xi, Y. (2004) “Literature Review of Alkali Aggregate Reactions in

Concrete Dams”, Project report to Swiss Federal Office for Water and Geology FOWG, 

78p.

83. Schmidt, F.A.R. (1969) “Analytical Radiation Shielding Calculations for Concrete –

Formulas and Parameters”, Nuclear Engineering and Design, 10, 308-324.

84. Shultis, J.K., and Faw, R.E. (2000) “Radiation Shielding”, American Nuclear Society, Inc., 

La Grange Park, IL.

85. Silberberg, M.S. (2007) “Principles of General Chemistry”, McGraw-Hill, New York, New 

York.

86. Soo, P., and Milian, L.M. (2001) “The Effect of Gamma Radiation on the Strength of 

Portland Cement Mortars”, Journal of Materials Science Letters 20, 1345-1348.
103

87. Struble, L., and Diamond, S. (1981a) “Swelling Properties of Synthetic Alkali Silica Gels”, 

Journal of the American Ceramic Society, 64(11), 652-655.

88. Struble, L., and Diamond, S. (1981b) “Unstable Swelling Behavior of Alkali Silica Gels”, 

Cement and Concrete Research, 11, 611-617.

89. Tao, X. Y., and Phillips, D. V. (2005) "A Simplified Isotropic Damage Model for Concrete 

under Bi-axial Stress States," Cement & Concrete Composites, 27(6), 716-726.

90. Taylor, H.F.W. (1986) "Proposed Structure of Calcium Silicate Hydrate Gel", J. Amer. Cerem. 

soc., 69(6), 464-467.

91. Taylor, H.F.W. (1997) “Cement Chemistry”, 2nd Edition, Tomas Telford.

92. Timoshenko, S. and Goodier, J. N. (1970) “Theory of Elasticity”, New York, McGraw-Hill.

93. USNRC (2009) “Concrete Radiation Shields and Generic Shield Testing for Nuclear Power 

Plants”, Regulatory Guide 1.69, U.S. Nuclear Regulatory Commission, Washington, D.C.

94. Van Gerven, T, Van Baelen, D., Dutre, V., Van de casteele, C. (2004) “Influence of 

Carbonation and Carbonation Methods on Leaching of Metals from Mortars”, Cement 

Concrete Research, 34, 149–156.

95. Vodák, F., Trtík, K., Sopko, V., Kapičková, O., and Demo, P. (2005) “Effect of γ-Irradiation 

on Strength of Concrete for Nuclear Safety Structures”, Cement and Concrete Research, 

35, 1447-1451.

96. Vodák, F., Vydra, V., Trtík, K., and Kapičková, O. (2011) “Effect of γ-irradiation on 

Properties of Harderned Cement Paste”, Materials and Structures, 44, 101-107.

97. Voyiadjis, G.Z., Taqieddin, Z.N., and Kattan, P.I. (2008) “Anisotropic Damage-plasticity 

Model for Concrete”, Int. J. Plasticity, 24, 1946–1965.

98. Voyiadjis, G.Z., Taqieddin, Z.N., and Kattan, P.I. (2009) “Theoretical Formulation of a 

Coupled Elastic-plastic Anisotropic Damage Model for Concrete Using the Strain Energy 

Equivalence Concept”, Int. J. Damage Mech., 18(7), 603- 638. 

99. Walker, H.N., Lane, D.S., and Stutzman, P.E. (2006) “Petrographic Methods of Examining 

Concrete: A Petrographic Manual”, FHWA-HRT-04-150, U.S. Department of 

Transportation, Federal Highway Administration, McLean, Virginia.

100. Willam, K.J., and Warnke, E.P. (1974) “Constitutive Model for the Triaxial Behaviour of 

concrete”, Proc. Intl. Assoc. Bridge Structl. Engrs, Report 19, Section III, Zurich, p. 30.

101. Willam, K.J., Xi, Y., Lee, L., and Kim, B. (2009) “Response of Reinforced Concrete 

Structures in Nuclear Power Plants”, SESM No. 02-2009, Department of Civil, 

Environmental, and Architectural Engineering, University of Colorado, Boulder.

102. Wyckoff, J.M., and Chilton, A.B. (1973) “Dose Due to Practical Neutron Energy 

Distributions Incident on Concrete Shielding Slabs”, Proc. of the 3rd Int. Congress IRPA, p. 

694, Washington, D.C. 
104

103. Xotta G, Salomoni, V.A. and Majorana, C.E. (2013a) “Thermo-hygro-mechanical Mesoscale Analysis of Concrete as a Viscoelastic-damaged Material”, Engineering 

Computations, Volume 30, Issue 5.

104. Xotta G, Beizaee, S., Mousavi, R. and Willam, K. (2013b) “Comparisons of Damage and 

Plasticity Models in Uniaxial Tension, Compression and Simple Shear”, Internal Report -

University of Houston, Texas.

105. Yevick, J.G. (Ed.) (1966) “Fast Reactor Technology: Plant Design, Shielding”, The M.I.T. 

Press, UK.

106. Yılmaz, E., Baltas, H., Kiris, E., Ustabas, I., Cevik, U., and El-Khayatt, A.M. (2011) 

“Gamma Ray and Neutron Shielding Properties of Some Concrete Materials”, Ann. Nucl. 

Energy, doi:10.1016/j.anucene.2011.06.011.

107. Yousef, S., AlNassar, M., Naoom, B., Alhajali, S., and Kharita, M.H. (2008) “Heat Effect on 

the Shielding and Strength Properties of Some Local Concrete”, Progress in Nuclear 

Energy, 50, 22-26.

108. Zanni, H., Nieto, P., Fernandez, L., Couty, R., Barret, P., Nonat, A., and Bertrandie, D. 

(1994) “Sol-gel Transition in Silica Alkaline Solutions: A NMR Study”, J. Chim. Phys., 91, 

901-908.



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NUREG/CR-7171 A Review of the Effects of Radiation on Microstructure and Properties

of Concretes Used in Nuclear Power Plants