Concrete is an important structural material used in nuclear power plant (NPP) design. Due to relatively high amount of hydrogen as well as the presence of heavier elements, it also acts as a biological shielding. One of the important tasks for prolongation of operational life time is the determination of concrete components' condition after long-term irradiation. The paper aims to present the current activities in the CV Řež institute (Research Centre Řež—CVR) regarding the investigation of ionizing radiation effects on concrete properties. In its first part, the paper deals with experimental identification of the character of mixed neutron and gamma spectra in the concrete part of the VVER-1000 Mock-Up. Using the knowledge, the radiation field character can be scaled up to the commercial power plants with VVER-1000 light water reactor. It also provides justification for usage of the 60Co source for performed irradiation experiments with concrete. The second part of the article describes the experimental studies of the properties of gamma-irradiated concrete samples by strong 60Co source. This irradiation experiment can be understood as the first step in characterizing concrete degradation as gamma flux in biological shielding is significantly higher than that of neutron flux. In order to better understand the concrete properties and the behavior under irradiation, nondestructive as well as destructive testing methods were applied. We found that after 48 days of irradiation by the 60Co source the sample obtained dose from gamma corresponding to approximately 1% of the total during the NPP lifetime operation. Concrete microstructure degraded and the modulus of elasticity slightly decreased within 5%. Conversely, destruction tests prove significant flexural strength decrease by 27% in case of normal test and by 63% at the loss of coolant accident (LOCA) test.
In nuclear power plants (NPPs), concrete acts as a structural material and also shielding barrier and has to withstand the unique environmental conditions typical for nuclear power plants and bear its function for the whole operational life-time period, typically 60–80 years. Concrete inside NPPs is exposed to accelerated aging due to neutron and gamma radiation and the accompanying heat. Generally, deterioration of the concrete properties is expected at a dose of 1 × 1019 neutrons/cm2 or 108 Gy for gamma radiation, which is manifested as significant decrease of compressive strength [1,2]. However, it is not yet concluded whether thermal neutrons cause less damage in concrete than fast and therefore the question of the energy cutoff and the thermal-to-fast neutron ratio in the spectrum is very important . The direct impacts of radiation on concrete structure are the displacement of atoms in the lattice structure and phase transformations of the constituents. Neutron radiation affects mainly the concrete filler, i.e., aggregate, whereas gamma radiation has an adverse impact on the binder, i.e., the cement paste [4–6].
An important factor in the aging process is the impact of generated heat inside the concrete structure due to radiation. According to Fillmore , the temperature rise in concrete due to irradiation alone can reach 250 °C, whereas at 180 °C, the free water is expelled and the dehydration process, i.e., the loss of nonevaporable water, and the breakdown of cement gel starts . For concrete with Portland cement as a binder, the limit temperature is 400 °C as at this stage the hydration product portlandite (Ca(OH)2) decomposes to quicklime and water. The generation of lime due to high temperatures then causes secondary internal stresses. The physicochemical changes in the cement paste and aggregate and the thermal incompatibility between these two phases are responsible for volume changes and thus cracking.
In case of severe accident, the impact of the mechanisms of deterioration is quickly amplified together with the rise of pressure inside the containment. In Research Centre Řež (CVR), the conditions of loss of coolant accident (LOCA) are simulated with an experimental installation able to generate a rise of temperature to 250 °C and pressure up to 1 MPa within few minutes.
The utilization of nondestructive testing (NDT) methods in modern civil engineering is a well-established and commonly used way of structural and material diagnostics. However, a fairly significant inaccuracy of the obtained results in the fully NDT of the quasi-brittle materials (concrete, ceramics, and stone masonry) is a significant disadvantage, and therefore it cannot be used for direct determination of a measured quantity, but only for preliminary estimation of the material degradation, finding of weak spots, mutual comparison of different materials' behavior, or for laboratory determination of relative values measured with respect to a reference set of materials. Rebound methods for determination of concrete strength or the common ultrasonic pulse velocity measurement of Young modulus of elasticity of concrete are typical examples. The latter method belongs to the group of linear acoustic methods. The sensitivity of linear acoustic methods to small defects is very low and the measured values frequently show high variance. Therefore, in the case of structural analysis, which serves for expertise or reconstruction projects, partly destructive methods have to be engaged. Nevertheless, there are cases in which any in situ inspection of the structure for sampling (e.g., core drilling) is very limited or impermissible, though the monitoring of the structure state is extremely essential in the means of safety precautions. Nowadays, this problem becomes very topical in the field of nuclear energy industry regarding the current need to prolong the lifetime of NPPs. A number of European nuclear power plants were built in the 80 s and 90 s of the 20th century with planned lifetime of 30 years, the limit of serviceability of NPPs in the U.S. and Great Britain has been established to be 40 years , yet it is being found out that these are in such a good shape to be serviceable for the next 50 years. However, in order to meet very strict safety criteria, monitoring and periodical inspection of their structures is required. This example presents only one of the reasons why there is such a high demand for the development of precise nondestructive diagnostics with high accuracy and reliability. Therefore, a new approach to NDT including nonlinear ultrasonic defectoscopy becomes a very perspective and requested way of solution.
2 Studies With Concrete in the Research Centre Řež
Structural members made of concrete constitute an essential part of the containment structure, since next to their very good load-bearing capacity, thermal resistance, and stability, they at the same time provide an effective shielding protecting the environment from radiation. However, the most exposed parts situated around the pressure vessel, called biological shielding concrete, suffer from the radiation damage. The effects of radiation on concrete structure and properties are still not thoroughly known and thus it is needed to conduct as much experiments as possible by both technical and economical means. Some of that means can be provided by the full scale mock-up in the LR-0 reactor in Řež, which enables measurements of the neutron–gamma mixed field parameters occurring in the large energy-producing reactors. Because of the mixed radiation field presence, there is a question of the validity of experiments realized with the artificial gamma source represented by few energy lines. To support the validation of the realized structural experiments, a series of neutron and gamma flux density measurements were carried out in a concrete simulator of biological shielding at the LR-0 reactor. Because concrete contains water, it acts as moderator of the fast neutron flux behind the pressure vessel and thus it increases thermal neutron flux and back-scattering at the outer boundary of the pressure vessel. Additionally, in some cases, high energy gamma radiation can origin at the thermal neutron capture in structural materials. Therefore, the as exact as possible knowledge about concrete composition and structure is essential.
Based on literature review, it is believed that neutron radiation (however, no consensus on the cutoff energy exists) affects predominantly crystalline phases, i.e., aggregate, while gamma rays have impact rather on the hydrated phases of concrete, i.e., cement paste . Therefore, presented experiments focused on measurement of the neutron to gamma share, spectrum of the field and further on gamma irradiation of cement paste samples, or mortars containing fine aggregate to study the potential damage of the interfacial transition zone between the grain of aggregate and the surrounding paste, as it is the weakest part of any concrete composite.
3 Instruments and Methods
3.1 Radiation Field Behind the Power Plant Simulator (VVER-1000 Mock-Up).
Thermal neutrons behind the reactor vessel are originating by collisions of fast neutrons in concrete biological shielding. Only the fast neutrons are penetrating the reactor vessel.
Thermal neutron flux in the concrete shielding was measured by 3He proportional detectors. Additionally, fast neutron spectrum measurements using method of recoiled protons using organic scintillator was performed. Scintillator spectrometer was used for gamma transport measurement as well. Measurement device is described, for example, in Ref. . The testing was carried out at an experimental reactor LR-0 with full-scale reactor core, vessel, and biological shielding simulator. Because LR-0 is a low-flux, so-called zero-power reactor, no real damage or impact of the flux on the concrete properties can be observed, only the pure neutron and photon transport.
The core in the Mock-Up is driven by 32 shortened VVER-1000-like fuel assemblies. Closer description of the configuration can be found in Ref. . The major parts of the VVER-1000 Mock-Up are the radial full-scale section of the reactor baffle, barrel, reactor pressure vessel (RPV), and concrete biological shielding (see Fig. 1). Light water moderator with boric acid concentration of 4.6 ± 0.1 g/kg is used to compensate reactivity excess.
The water layer density is decreased by displacer, which reduces the water density in the outer reactor core boundary (downcomer) to the value near to the VVER-1000 operating conditions. The structure of the pressure vessel and biological shielding simulator is modularized, thus enabling the neutron flux and spectrum measurement in different depths of the steel and concrete. Measurements of neutron and gamma flux presented in this work were carried out in the point on the pressure vessel outer surface (Pt-7) and in the tube in shielding model (Pt-8).
Radiation transport simulations were carried out in the MCNP6 Monte Carlo code . Nuclear data for the described model were taken from ENDF/B-VII.0  library, the transport in water uses adequate thermal scattering treatment, transport in steel deals with free gas treatment . Neutron source was described as an assemblywise fixed source based on pin-power profile calculated in critical calculation . The source term can be understood as reliable, because pinwise fission profile was validated during previous experiments [17–20], the absolute determination of flux was carried out by means of activation foils .
3.2 Gamma Irradiator.
To induce the radiation effect on concrete, much stronger source than the LR-0 reactor has to be used. For such purposes, gamma irradiator has been employed. The irradiator is using quasi-point cobalt 60Co source with the nominal activity 200 TBq. Typical distance between the sample and this source in the irradiator is 80 and more millimeters. The temperature inside the irradiation chamber depends strongly on the outdoor temperature. For this experiment, the nominal temperature was 24 ± 3 °C. It was distributed according to the distance from the 60Co source. The maximum acted in the middle of the length of the samples, i.e., in the place where the tensile strength was determined. The same situation happened with the gamma dose, which had the maxima in the mid of the tested samples.
3.3 Experimental Description of Gamma Irradiation Followed by Loss of Coolant Accident Test.
Samples of mortars with dimensions 40 × 40 × 160 mm were manufactured by standard procedure ČSN EN 206-1 . Smaller pieces of dimensions 20 × 20 × 40 or 0.1 × 20 × 40 mm were prepared for microscopic analyses. The used binder component was Portland mixed cement CEM II 32.5 R and common siliceous fine aggregate (grain size range 0–2.5 mm) was selected as filler (Table 1). The samples were left in laboratory conditions for setting and hardening and afterwards placed in gamma radiation cell with a gamma source of 172 TBq (60Co). The irradiation dose rate was 0.5 to 4.5 kGy per hour (Fig. 2) and the temperature within the cell was held at 24 ± 3 °C. Samples were irradiated for 21 days, then they were tested by nondestructive measurements and afterward were put into the cell for another 27 days of irradiation. The overall exposition dose was from 1.6 to 1.8 × 106 Gy (1.6 to 1.8 × 108 rad) corresponding to approximately 1% of the expected photon dose during the power plant lifetime according to Maruyama et al. . After the end of irradiation, the samples were tested both nondestructively and destructively (Fig. 3).
Part of samples was also subjected to LOCA test (see the experimental device in Fig. 4), which stands for loss of coolant accident, a Severe Accident to which all concrete structures in nuclear power plants have to be designed. In such case, the concrete is exposed to rapid rise of temperature and pressure creating gradients within its mass and causes damage of its internal structure. The experiment simulated rise of temperature and pressure up to 250 °C and 1 MPa, respectively, in a few minutes by a steam preheated in a steam generator. The record of the pressure and temperature evolution is in Fig. 5. Moreover, post-LOCA test was also conducted, which is a complex of processes that occurred immediately after LOCA accident. This was simulated by a three-days slow cooling of concrete samples by a shower of boric acid. The samples were afterward tested destructively and compared with data obtained on irradiated samples without LOCA.
Nondestructive testing includes ultrasonic and resonance methods to determine the dynamic modulus of elasticity Ec (MPa) according to the Czech standards [27,28]. Destructive tests were three-point bending test to assess the flexural strength σt (MPa) and conventional loading procedure to assess compressive strength σc (MPa) following the standards .
Microscopic techniques, namely, scanning electron microscopy (SEM) and light optical microscopy, were used for chemical analysis and mineralogical phase determination.
During the reactor operation, the structural materials are exposed to the mixed field of neutrons and gammas, where according to the theory, both types of particles can induce themselves. That means, high-energy neutrons produced from fission reactions are losing energy by scattering to energies corresponding to thermal equilibrium. However, these low-energy neutrons can be absorbed on structural materials (water, steel) and produce high energy photons. Thus, in case of photons, the broad spectrum originates from fission, but the high-energy photons are produced mainly from (n,g) reactions as stated above. Threshold for neutron production by photon (g,n) is very high in structural materials of light water reactor, higher than the energies of photons appearing in the fission reactor; therefore, this channel for neutron production in the biological shielding can be neglected. Important fact is that the thermal neutrons and low energy gammas produced in reactor core are not able to penetrate the layers of reactor internals, water, and the RPV boundary to reach the concrete shielding; thus, the knowledge of the fast neutron field behind the RPV is substantial.
4.1 Fast Neutron Flux in Biological Shielding of VVER-1000 Mock-Up.
The integral fast neutron flux is responsible for damage and for origin of gamma and low-energy neutrons behind the reactor pressure vessel. The neutron flux and spectrum behind the RPV and in biological shielding were measured by stilbene scintillator and compared with calculations. Comparison shows quite good agreement, especially for the case behind the RPV (Pt-7), see Fig. 6. The measured neutron spectrum shape corresponds to calculations. However, the narrow peaks in spectrum cannot be recognized by stilbene because of its resolution.
4.2 Gamma Flux in Biological Shielding of VVER-1000 Mock-Up.
Measurements with scintillation spectrometer show that the measured shape of gamma spectrum in energies below 5 MeV roughly corresponds with calculations in both measurement points. Significant discrepancy appears above this threshold (see Fig. 7). The comparison of the photon spectra in different parts outside the reactor vessel can be found in Table 2. In the concrete, the spectrum is slightly shifted to higher energies, which is the reason why the average energy of the gamma spectrum increases.
Quantity important from the point of view of material studies is the ratio of induced gammas to neutrons (gamma-to-neutron ratio). This ratio for particle energies above 1 MeV is presented in Table 3, where large discrepancy, around 50%, occurs in calculation in case of measurement in hollow for ionizing chamber (Pt-8, see Fig. 1). Experimental values in broad energy groups can be found in Table 4, where the rapid decrease in the higher energies of the fast neutron flux can be apparently observed.
4.3 Thermal and Epithermal Neutron Transport.
The thermal neutrons in concrete shielding appear when the fast neutrons are slowed down by scattering on the light nuclei contained in concrete (mainly on hydrogen and silicon). Spatial distribution of thermal and epithermal neutrons was measured with ionizing chamber simulator segment (Pt-8) replaced by solid concrete blocks thus creating layers of concrete with different thickness. The measurements were carried out with and without Cd cover of the detector to obtain Cd ratio (ratio of thermal and epithermal neutrons). More can be found in Ref.  (Table 5).
Results of 3He detector measurement show increasing discrepancy of calculation and measurement with increasing concrete layer thickness. This discrepancy reaches up to 40% in case of thermal neutrons and up to 20% in case of epithermal neutrons. Results of Cd ratio comparison show even larger discrepancy. See graphs in Figs. 8 and 9 for results.
4.4 Experimental Study of the Effect of Gamma Irradiation.
Study of the damage induced by gamma radiation was simulated by the 60Co irradiation in irradiator. In real operation, not only 60Co peaks (1.17 and 1.33 MeV) are present behind the RPV boundary, but broad spectrum of energies, including much higher, are induced by thermal neutron interactions with water (H(n,g)—2.23 MeV) or steel components of reactor internals (56Fe(n,g)—7.63 and 7.64 MeV; 58Ni(n,g)—8.534 MeV and 8.998 MeV, 53Cr(n,g)—8.884 MeV, 54Fe(n,g) 9.298 MeV. The fraction of high energy gamma in the spectrum is lower than the energies in interval <2 MeV (36%) and the mean energy of the gamma spectrum in concrete shielding was measured to be 1.93 MeV. Thus, the irradiation in the 60Co irradiator can be considered as the valid approach for concrete degradation studies, because it is hard to obtain sufficiently strong gamma emitters with higher energy lines corresponding to the gamma spectrum of the nuclear reactor.
Data obtained from nondestructive measurements of concrete after irradiation did not show small change in fundamental frequency after 48 days of irradiation (see Fig. 10). After LOCA, i.e., sudden increment of temperature and steam pressure, peak frequency move nearly by 1500 Hz lower. This is caused by damaged material with lower rigidity. Less stiff material has lower resonance frequencies
where ρ is the specific gravity of concrete, kg m−3; k is the coefficient of cross section; here, k = 1.187; L is the sample length, m; f is the torsion fundamental frequency determined here using resonance method impact-echo, s−1.
Decrement of dynamic shear modulus by 2 and by 42% was caused by gamma + drying effect (of radiolysis of water) and by steep increase of temperature and pressure in concrete samples, respectively.
Analyses using microscope techniques proved our hypotheses that no alternation of phase and chemical composition occurs during irradiation to such doses, as reached in the experiment. However, as shown in Fig. 11, visible widening of the microcracks present in the microstructure of concrete occurs, or moreover there is rise of new ones, especially in the interfacial zone between the cement paste and aggregate. The cracking might be caused by the effect of heating due to gamma irradiation or by hydrogen release as the results of water radiolysis .
Data obtained from destructive tests are summarized in Tables 6 and 7. More pronounced changes were recorded in the case of flexural strength (Table 6). After the irradiation experiment, the flexural strength decreased to 73 ± 9% of the initial strength and after LOCA test even to 37 ± 9%. In the means of compressive strength, only minor changes were recorded (Table 7). The residual value of compressive strength after irradiation was 94 ± 7%. The change after LOCA test was due to the temperature and pressure shock more significant—the decrease to 82 ± 2% of the reference values was recorded.
The strongest effect of the gamma irradiation seems to be
The decrement of the tensile strength by 30% (Table 1) due to the microcracks.
The values of E-moduli changed from −1 to −5% after the midterm irradiation (0.8 to 1.2 × 106 Gy) and stayed nearly the same after another 48 days in gamma cell (1.6 to 1.8 × 106 Gy).
The further effect of LOCA test can be summarized as:
The damage was caused predominantly by the temperature and pressure shock.
Decrease of tensile strength was very significant—up to 37 ± 9% of the initial values, while the drop of compressive strength was lesser—to 82 ± 2% (Table 7).
In between the sensitivity of two above-mentioned destructive methods, there is nondestructive testing method impact-echo, which gives 58 ± 2% (Table 4).
Methodologies and capabilities for complex studies of gamma irradiated concrete were introduced in this paper. Basically, when the proper neutron source will be applied, impact of the neutron dose could be studied as well.
The shape of the measured photon spectrum in lower energies fairly corresponds to calculations, but substantially larger discrepancies are indicated in higher energies. Photons in spectrum in concrete are generated predominantly by nuclei interactions with neutrons. From this point of view, fast neutron flux spectrum in concrete does not appear to be too badly described by calculation, the main issue is in neutron slowing down process in concrete, where the discrepancies reach up to 40 and 20% for thermal and epithermal neutrons in different layers of concrete. And the proper description of thermal neutrons is the key to the proper description of gamma spectrum, because photons in concrete are induced mainly by (n,g) reactions.
From the measured and calculated photon spectrum shape, it can be assumed that the 60Co lines are reasonable for concrete radiation aging simulations, because the amount of gamma particles decreases with increasing energy and the share of photons below 1 MeV forms a majority in the spectrum. Further, it was shown that the region inside the concrete in position of ionizing chamber contains approximately 25 times more photons than neutrons; thus, the study of photon interactions with concrete is important.
Concrete appears to be sensitive to irradiation by 60Co photons, which was used as a representative of photon spectrum in the biological shielding of the VVER-1000 reactor. The decrease of the tensile strength or shear modulus and appearance of microcracks have been observed after irradiation. Simulation of LOCA accident led to worsening of the postirradiation characteristics. Destructive and nondestructive methods indicated substantial changes of all investigated mechanical properties. Changes in shear modulus occurred after LOCA conditions rather than after pure gamma irradiation.
For the proper estimation of doses from gamma radiation in case of the energy producing reactors, the correct description of photon spectrum is needed. Because photons behind the reactor vessel and in biological shielding are originating mainly through (n,g) reactions, proper distribution of fast and thermalized neutrons has to be determined. It was shown that the calculations with current data libraries do not describe the photon flux correctly in the whole spectrum and therefore the dose estimate will be only rough.
The experiments and calculations were realized under contract No. Rea/11/34/996-10/INR between Research Centre Řež ltd. and National Research Centre “Kurchatov Institute” (Moscow, Russia).
The research has been financially supported by Ministry of Education, Youth and Sport Czech Republic (CZ27013111) — project LQ1603 Research for SUSEN. This work has been realized within the SUSEN Project (established in the framework of the European Regional Development Fund (ERDF) in project CZ.1.05/2.1.00/03.0108 and of the European Strategy Forum on Research Infrastructures (ESFRI) in the project CZ.02.1.01/0.0/0.0/15_008/0000293. Concrete samples were fabricated within the project VI20192022154, Biological Shielding Concrete Testing after Reactor Core's Neutrons Irradiation “with financial support from the state budget through the Ministry of the Interior of the Czech Republic (CZ00007064) in Program Security research of the Czech Republic 2015-2020.
- b =
side length in the cross section of the concrete sample, m
- E =
energy, kg m2 s−2
- f =
resonance frequency, 1/s
- F =
force, kg m/s2
- G =
shear modulus, kg/(m s2)
- h =
side length in the cross section of the concrete sample, m
- k =
coefficient of cross section
- L =
sample length, m
- σ =
strength, kg/(m s2)
- ρ =
specific gravity of concrete, kg/m3
Subscripts or Superscripts
- B28, B40 =
types of concrete samples
- CEM II 32.5 R =
type of the rapid hardening cement, strength class 32.5
- CV =
- CVR =
Research Centre Rez
- ČSN EN =
The European norm (standard) adopted by Czech Office for Standards, Metrology and Testing
- ENDF/B-VII.0 =
Evaluated Nuclear Data File, version B-7.0
- G/N =
- IRPhEP =
International Reactor Physics Experiment Evaluation Project
- LOCA =
loss of coolant accident
- LR-0 =
unique type of the light-water reactor based on fuel in triangular lattice with zero power
- MCNP6 =
Monte Carlo N-particle transport code version 6
- NDT =
- NPP =
nuclear power plant
- Pt-7 =
- Pt-8 =
- RPV =
reactor pressure vessel
- RR =
- S.D. =
- SEM =
scanning electron microscopy
- USA =
United States of America
- VVER =
water–water energetic reactor, Russian version of the pressurized water reactor