Concrete Biological Shield Research Articles

Measurement of gamma field inside the biological concrete shielding of VVER-1000 Mock-Up at the LR-0 reactor

The long-term operation of existing nuclear power reactors is a crucial concern due to the complexities and expenses associated with replacing key components, such as the reactor pressure vessel and reactor internals. Gamma radiation, a byproduct of nuclear reactions and radioactive decay, significantly influences the lifetime of these components. This radiation is responsible for various degradation pathways leading to void swelling in steel reactor components and cracking or other radiation damage in concrete structures.A study conducted at a full-scale mock-up of the VVER-1000 reactor at the LR-0 zero-power reactor employed HPGe and stilbene measurements to analyze gamma spectra behind the reactor pressure vessel and within concrete biological shielding. While simulations behind the reactor pressure vessel aligned with measurements, notably, amarked overestimation of stilbene spectrum calculations occurred deep in concrete, suggesting potential inaccuracies in radiation predictions for power plant structures.

The influence of tritium behaviour on spent fuel pool concrete

This paper investigates the effects caused by Tritium released from spent nuclear fuel (SNF) stored into a spent nuclear pool made of (reinforced) concrete. The release of tritium from fuel elements to coolant during normal LWRs operation is of 1 % while during 6–20 years of wet storage is assessed to 13–45 %. Permanent Tritium concentration in coolant is assessed to n·109 Bq·dm−3.The water contamination first and the filtration of the concrete pool walls then pose important safety problems to face from both radiological and structural point of view: during Tritium filtration, radiation affects significantly the structural and mechanical properties of concrete due to the breaking of atomic bonds.Under the effect of radiation concrete microstructure modifies; collision determines atom dislocation and so a lattice defect. Ionized rays may cause the decay of free or bonded water leading to the formation of H2 and O2 and produce explosive mixture.The damage from “soft” β-decay of Tritium inside the concrete could be harder due to short track of the particle and the resulting “volume” effect. Neither this effect nor to what extent Tritium movement in/through concrete material would take place over long-term are yet completely studied.Due to Helium (He) formation, when Tritium decays, the restructuring of water molecules with generation of reactive free radicals occurs. We have assessed these effects: the irradiation and accompanying volume expansion, formation of free radicals and corresponded chemical corrosion are not considerably influencing on the structure and sustainability of concrete biological shield. The most significant effect is the radiation breaking of van der Waals bonds that within 10 years can lead to the destruction of about 2 % of concrete.Concrete shield does not provide a complete barrier against Tritium release into the environment. The main threat of Tritium diffusion to coolant in SNF storage pools is caused by filtration of tritiated water through the concrete walls that may provide significant health hazards for employees of the nuclear industry. Results show that β-decay of tritiated water adsorbed in 1 m3 of concrete may destroy about of 0.05 mol (11.4 g) of cement stone within 10 years, which is less than 0.001%. Moreover, with 3. With reference to the 5th power unit of the Zaporizhzhia NPP, the volume of Tritiated water released from the pool is 0.67 m3 annually. Resultsshow that the tritium waste management concept needs to be revised.

Mesoscale 3D model of irradiated concrete

The concrete biological shield in light-water reactors is exposed to neutron and gamma irradiation, which deteriorates the concrete’s mechanical properties in the long term. To assess the irradiation-induced damage, predictive mechanical models are developed and used in parallel with the characterization of irradiated concrete samples. Realistic 3D simulation domains can drastically improve a model’s prediction. In this work, we utilized x-ray computed tomography (XCT) data of a concrete specimen to reconstruct its 3D microstructure. The XCT data shows low contrast between the concrete’s aggregates and cement paste, resulting in poor image segmentation when using traditional unsupervised techniques. To address this issue, we developed and trained a 2.5D U-Net model on only 24 pre-labeled XCT layers to segment 651 layers of the XCT data. The overall F1-score of the model is approximately 96%. Then, we created a 3D finite element (FE) mesh based on the stack of segmented images. The FE model contains radiation-induced expansion, damage, and creep. The constitutive equations are adapted to each phase (aggregates and cement paste). We simulated the effects of neutron irradiation in the concrete specimen as well as the specimen’s mechanical response to uniaxial compression. Finally, model validation was performed using experimental data on similar concrete specimens in the literature.

Calculation of radiation fields in the VVER-1000 concrete biological shield using Monte Carlo code Serpent

To calculate radiation fields in the concrete biological shield (CBS) of the VVER-1000 reactor in this work, we have developed and applied the Monte Carlo code Serpent simulation framework based on the variance reduction technique. We have quantified the radial, axial, and azimuthal variation of neutron and gamma-ray fluxes and the absorbed dose rate in the CBS. Using the calculation results, we estimate maximum neutron fluence and maximum absorbed dose in the VVER-1000 CBS over the period of 60 and 80 years of the reactor operation and localize the domains of highest radiation exposure. The obtained results are in good agreement with the available data on the VVER-1000 and other pressurized water reactors. We show that the fluence of neutrons with energy above 0.1 MeV decreases by half at a depth of 4 cm of concrete, and the gamma-ray absorbed dose decreases by 40 % at a depth of 13.5 cm. The outcomes from this research will help to assess the effects of prolonged irradiation of the VVER-1000 CBS, which is required for reliable risk assessment for extended operation of nuclear power plants.

Dynamic responses of radiation-induced heavyweight concrete subjected to biaxial compression

As a good shielding material, heavyweight concrete is largely utilized for building the Concrete Biological Shielding (CBS) wall of the commercial nuclear power plant, and is often subjected to the biaxial loadings with a strain rate possibly varying from 10−5s−1 to 10−2s−1. This paper assesses the combined effects of the lateral stress and the strain rate on the compressive strength of radiation-induced heavyweight concrete on the example of serpentine concrete. For this goal, a 3-dimensional rectangular numerical concrete specimen containing special aggregates and ITZs was firstly established. Exposed to neutrons of different fluences, the initial damages caused by the fast neutron were assessed by a thermal-mechanical approach. Then, a series of static and dynamic biaxial compressive tests on the rectangular specimens having initial damages of different degrees were carried out by finite element analysis via ABAQUS. The failure modes, the crack patterns, and the ultimate compressive strengths of the pre-irradiated serpentine concrete were obtained numerically. Finally, based on the numerical results and the classical "Kupfer-Gerstle" ("K-G") criterion, this paper derives the biaxial strength criterion of the radiation-induced heavyweight concrete undergoing both lateral stress and strain rate. Through the formula, it was found that the fast neutron fluence of high level reduces significantly the biaxial compressive strength of the heavyweight concrete. Besides, if the heavyweight concrete undergoes additionally lateral stress and the strain rate, the biaxial compressive strength could be enhanced. The enhancement effect was maximized at the lateral stress ratio of 0.5 and the strain rate of 10−2s−1. The current numerical approach was developed with a potential application to the lifetime extension of the heavyweight concrete infrastructures used in commercial nuclear reactors.

A structural model of the long-term degradation of the concrete biological shield

The concrete biological shield (CBS) of light water reactors is exposed to high neutron radiation dose in the long term, which may lead to the degradation of the concrete’s mechanical properties. Given the important shielding role of the CBS, it is necessary to investigate the irradiation effects at the structural scale and provide estimates of the damage extent from the wall’s inner surface to study potential license renewals. For this purpose, we developed a mechanical model accounting for radiation-induced expansion, creep, and damage in concrete using the Grizzly finite element code, informed by ex-core neutron flux calculations using the VERA tool. The model was applied to a 3D CBS structure represented by the CBS wall, a steel liner, reinforcement bars, and a concrete base mat and evaluated damage at 40, 60, and 80 years of operation. The VERA model predicted a maximum fluence of approximately 2×1019 n cm−2 at 80 years of operation. The results showed that damage is highest at the inner surface of the CBS wall and gradually decreases with depth. It extends beyond the rebar after 60 years and reaches a depth of approximately 12 cm at 80 years.

Ultra-high gamma irradiation of calcium silicate hydrates: Impact on mechanical properties, nanostructure, and atomic environments

The concrete biological shield in a nuclear power plant receives ~100–200 MGy gamma dosage during an 80-year design life. However, precise changes in the mechanical properties and atomic environments of C-S-H at ultrahigh irradiation dosages have not been systematically documented. Here, we report that irradiation decreases C-S-H basal spacing (~ 0.6 ± 0.1 Å for 189 MGy) and increases its Young's modulus, which is attributed to the lower basal spacing as the nano porosity potentially increased and microporosity remained unchanged. Irradiation also decreased the molecular water content and increased hydroxyl groups in C-S-H, showing that interlayer water removal reduces the basal spacing. Finally, 1H and 29Si NMR results indicate some disorder in the local proton CaO-H species and slight depolymerization of the silicate structure. Together, these results indicate that the C-S-H gel stiffens upon ultrahigh gamma irradiation dosage, a finding which concerns long-term nuclear power plants operations worldwide.

Validation and Optimization of Activity Estimates of the FiR 1 TRIGA Research Reactor Biological Shield Concrete

FiR 1 is a TRIGA Mark II-type research reactor in Finland. It was in operation between 1962 to 2015 and will be dismantled in 2022 to 2023. Preliminary calculations of the activities in the reactor main structures were performed in an earlier stage of the decommissioning project. Samples of the activated parts of the reactor biological shield concrete were drilled in December 2018 to validate these estimates. This paper describes the calculations and gamma activity measurements performed for the activated concrete samples to determine the boundary between radioactive parts and concrete that can plausibly be free-released from regulatory control. The activities have been estimated with a two-step calculation process using the MCNP and ORIGEN-S calculation codes and measurements using an ISOCS gamma spectrometer with a high-purity germanium detector.

Modeling long-term distribution of fast and thermal neutron fluence in degraded concrete biological shielding walls

Since neutron radiation can deteriorate mechanical properties of the concrete materials, it’s critical to obtain accurate neutron radiation levels in concrete structures during their service live. A two-group neutron diffusion model for fast and thermal neutron transport is developed in this study which considers the kinetic energy of the neutron more realistically. Based on the cross-property correlation theory, the variation of transport properties of concrete can be estimated using the degradation of mechanical properties of concrete, then, the thermal and fast neutron radiation fields with two representative energy levels as well as the thermal field in a concrete bioshield can be calculated. Coupled radio-thermo analysis of a section of simplified concrete bioshield was performed to 80 years as a case study, and the results were discussed and compared with the results from the one-speed model. The degradation of concrete induced by neutron radiation and elevated temperature can lead to considerable increases of thermal and fast neutron flux/fluence in the concrete. The profiles of thermal neutron flux/fluence are very different from the profiles of fast neutron flux/fluence. A proper treatment of the neutron energy dependence is needed to predict long-term neutron radiation levels in concrete biological shields accurately.

Assessment of a polymeric composite as a radiation attenuator and a restoration mortar for cracking in biological shields

This work is dedicated to figuring out robust epoxy/magnetite/boron carbide (EP/Mag/B4C) composite for radiation attenuation at multiple applications related to nuclear installations, as well as restoration mortar for cracking developed in concrete biological shields. The mechanical properties (flexural, compressive, and impact strengths) and the physical properties (water absorption, porosity, and dry bulk density), each, have been performed to label the composite integrity for practical application. In practice, attenuation properties have been performed by using a collimated beam emitted from spontaneous fission 252Cf (100 μg) neutron source and neutron gamma spectrometer with stilbene scintillator. The pulse shape discrimination technique which would come of the zero cross over method was used to measure the fast neutron and gamma-ray spectra. Thermal neutron fluxes have been measured by using the thermal neutron detection system and the BF-3 detector. The attenuation parameters: precisely, macroscopic effective removal cross-sections ΣR (cm-1), macroscopic cross-sections Σ (cm-1), and total attenuation coefficients μ (cm-1) of fast and thermal neutrons and total gamma-rays respectively were evaluated using the attenuation relations. Also, the MCNP5 code and MERCSF-N program have been used to compute the parameters theoretically. When applicable, measured and calculated results were compared, and it tells us a comprehensive agreement.

Application of Nonlinear Elastic Wave Spectroscopy in the Field of Nuclear Energy

Abstract In the field of nuclear energy, two particular applications are considered—the measurement of biological shielding concrete and measurement of high-temperature high-pressure piping. For the measurement of biological shielding concrete, a special manipulator was developed, which is applied through the ionization channels of VVER-440 power plants. Biological shielding concrete is covered on all sides with an 11 mm thick steel liner preventing its assessment using any conventional test method. For the measurement of high-temperature high-pressure piping, acoustic emission sensors installed in the LEMOP system (NPP Temelin) are used for both excitation and data acquisition. In operation, this piping is covered with glass wool insulation, but installed sensors on the waveguides make it possible to assess pipelines without removing the insulation. To demonstrate the material degradation, a set of concrete test specimens and a set of steel test specimens were made. Nonlinear elastic wave spectroscopy is a test method based on measuring the elastic–plastic response of the material whose plastic component is caused by microcracks. Part of this response may be caused by inhomogeneity of the material itself or the nonlinearity of the test system used, which is to be minimized in the measurement. Due to degradation, the number of microcracks increases, increasing the total nonlinear response of the material. The aim of this approach is to monitor the degradation of concrete and metallic materials based on the response of the test system to ultrasonic excitation.

Testing Oxygen Nuclear Data Libraries in Research Center Rez

Abstract Oxygen belongs to the group of the most important isotopes in the nuclear data field. The aim of this paper is validate various oxygen nuclear data libraries in different scenarios with high content of oxygen. For this purpose, fast neutron spectra were measured by a stilbene scintillation detector in the region of 1–10 MeV in the three model cases involving 252Cf neutron source and light water reactor. The cases include measurements of leakage spectra using 252Cf neutron source placed in the centers of the light water and heavy water spheres of 0.30 m diameter. Following measurements were carried out inside the concrete biological shielding of the VVER-1000 mock-up simulator in the LR-0 reactor and in the dry channel located in the center of the special core placed in the LR-0 reactor. In the case of the special core, symmetric active core consisted of six standard fuel assemblies which surround the experimental dry module, where the fast neutron spectrum was measured. The measured neutron spectra were compared with MCNP6 transport code calculations in ENDF/B-VII.1, ENDF/B-VIII.0, JENDL-4.0, and IRSN 16O nuclear data evaluations. Experimental results for all cases follow similar trend. All considered libraries underestimate experimental measurement in the region of 3–4 MeV in all cases.

Numerical Analysis of a Concrete Biological Shielding Wall under Neutron Irradiation by 3D RBSM

Numerical Analysis of a Concrete Biological Shielding Wall under Neutron Irradiation by 3D RBSM

Development of the Pushing Type Cutting Device to Dismantle Concrete Structure for Decommissioning of Nuclear Power Plant

Pulling-type cutting devices, which use a diamond wire saw, have been used generally for cutting concrete structures. In this study, a pushing-type cutting device with a collection cover was developed by overcoming the disadvantages of pulling-type devices. In this device, dry or liquid methods can be selected to cool frictional heat. Operation and leakage tests of the dust generated during the dismantling of a concrete structure were carried out, confirming the suitable operation of the fabricated cutting device; the leakage rate was approximately 1.7%. For a conservative evaluation, the internal dose of workers was estimated in dismantling the core center part of biological shield concrete with a specific activity of 99.5 Bq·g-1. The committed effective dose per worker was 0.25 mSv. The developed cutting device contributed to reducing radioactive concrete waste and minimizing worker exposure due to its easy installation. Therefore, it can be utilized as a cutting apparatus for dismantling not only reinforced concrete structures but also radioactive biological shield concrete in nuclear power plant decommissioning efforts.

Numerical analysis of VVER-440/213 concrete biological shield under normal operation

The numerical approach for the coupled damage-creep modeling of concrete biological shield, which combines the current and the past knowledge regarding the effects of irradiation and temperature on concrete with the real measured and calculated neutron fluence and temperature distribution for VVER-440/213 reactors, is described in detail in this study. The proposed approach takes into account the real structural geometry as well as the real neutron fluence and temperature distribution and the latest knowledge about the effect of irradiation and temperature on concrete strength and stiffness. The radiation induced volumetric expansion and the thermal expansion of concrete are modeled. According to the results of the numerical simulation, the analyzed structure reaches critical damage within the time interval from 10.00 to 35.25 years of normal operation. The damage of the concrete biological shield of the VVER reactor will not affect the load-bearing function of the containment building, since the biological shield is self-bearing. The shielding properties of the biological shield may be reduced due to the appearance of the radial cracks, however, the concrete wall, which is situated right behind the biological shield, will ensure the necessary shielding. Therefore, the concrete biological shield of the VVER reactors can be considered as sacrificial structure and can be damaged without significant consequences. However, this study implies the importance of capability to predict the behavior of those PWR reactor biological shields which serve both the load-bearing and shielding purposes.

Structural assessment of radiation damage in light water power reactor concrete biological shield walls

An important topic for long-term operation of nuclear plants is aging of plant concrete structures. The containment building, biological shielding, and support concrete are examples of concrete structures that are of primary importance in the operation of a nuclear plant. These and other safety-related structures must be capable of maintaining structural capability for the operating life of the plant. Demonstration of the satisfactory condition of the concrete structures is required for safe operation, particularly when plant operation beyond 60 years is considered.The concrete biological shield (CBS) wall is particularly important because of its shielding and structural functions and that it is irradiated due to its proximity to the reactor pressure vessel. Demonstrating that the concrete remains structurally able to perform its intended function is vital in effective aging management.Neutron irradiation above certain thresholds can cause loss of tensile and compressive strength of the concrete and volumetric expansion. If the irradiation level exceeds the threshold, both loss of strength and volumetric expansion should be evaluated. Methods to reliably evaluate the effects of the loss of strength and volumetric expansion are required.

Long-term neutron radiation levels in distressed concrete biological shielding walls.

Neutron radiation can deteriorate mechanical properties of the concrete materials, and thus it is questionable that neutron transport properties of concrete can remain unchanged during the life span of biological shielding walls. one-speed neutron diffusion equation and heat conduction equation were used as governing equations for prediction of neutron radiation and thermal field in concrete, respectively. The potential variations of transport properties due to neutron radiation and elevated temperature were estimated. A simplified example of a typical concrete biological shielding wall was analyzed up to 80 years, and the results were discussed. The radiation damage and radiation heating lead to minor changes of the temperature profile in the concrete. However, neutron radiation and elevated temperature can result in considerable increases of neutron flux and fluence in the concrete. The damage of concrete induced by neutron radiation and elevated temperature can considerably accelerate the penetration of neutron radiation into the concrete. This work is the first attempt to deal with the degradation of neutron and heat transport properties of concrete and its effect on neutron fluence distribution in concrete, and provides a possible way to determine the long-term neutron and thermal fields in concrete biological shielding walls.

Facile emulsion mediated synthesis of phase-pure diopside nanoparticles

Diopside is a common natural pyroxene that is rarely found in a pure state, since magnesium is often partially substituted by iron, and other elements (sodium and aluminum) are often present. This pyroxene, along with feldspars and olivines, is common in concrete. As the prospective license renewal of light water reactors to 80 years of operation has raised concerns on the effects of radiation in the concrete biological shield surrounding the reactors, mineral nanoparticles can be valuable to perform amorphization studies to inform predictive models of mechanical properties of irradiated concrete. The synthesis of diopside nanoparticles was achieved in this study using a reverse-micelle sol-gel method employing TEOS, calcium chloride and Mg(MeO)2 in a methanol/toluene solution. Tert-butylamine and water were used as hydrolysis agents, and dodecylamine as a surfactant. The resulting amorphous precursor was centrifuged to remove organics and fired at 800 °C. Additional reaction with hydrogen peroxide was used to remove amine remnants. TEM and SEM examinations revealed a product comprised of 50–100 nm diameter nanoparticles. XRD indicated phase pure diopside and BET indicated a surface area of 63.5 m2/g before peroxide treatment, which at a bulk density of 3.4 g/cm3 is equivalent to particles with diameter of 28 nm.

Loss of Core Cooling Test With One Cooling Line Inactive in Vessel Cooling System of High-Temperature Engineering Test Reactor

In the high-temperature engineering test reactor (HTTR), the vessel cooling system (VCS) which is arranged around the reactor pressure vessel (RPV) removes residual heat and decay heat from the reactor core when the forced core cooling is lost. The test of loss of forced cooling (LOFC) when one of two cooling lines in VCS lost its cooling function was carried out to simulate the partial loss of cooling function from the surface of RPV using the HTTR at the reactor thermal power of 9 MW, under the condition that the reactor power control system and the reactor inlet coolant temperature control system were isolated, and three helium gas circulators (HGCs) in the primary cooling system (PCS) were stopped. The test results showed that the reactor power immediately decreased to almost zero, which is caused by negative feedback effect of reactivity, and became stable as soon as HGCs were stopped. On the other hand, the temperature changes of permanent reflector block, RPV, and the biological shielding concrete were quite slow during the test. The temperature decrease of RPV was several degrees during the test. The numerical result showed a good agreement with the test result of temperature rise of biological shielding concrete around 1 °C by the numerical method that uses a calibrated thermal resistance by using the measured temperatures of RPV and the air outside of biological shielding concrete. The temperature increase of water cooling tube panel of VCS was calculated to be about 15 °C which is sufficiently small in the view point of property protection. It was confirmed that the sufficient cooling capacity of VCS can be maintained even in case that one of two water cooling lines of VCS loses its function.

Characterization of Radiation Fields for Assessing Concrete Degradation in Biological Shields of NPPs

Life extensions of nuclear power plants (NPPs) to 60 years of operation and the possibility of subsequent license renewal to 80 years have renewed interest in long-term material degradation in NPPs. Large irreplaceable sections of most nuclear generating stations are constructed from concrete, including safety-related structures such as biological shields and containment buildings; therefore, concrete degradation is being considered with particular focus on radiation-induced effects. Based on the projected neutron fluence values (E > 0.1 MeV) in the concrete biological shields of the US pressurized water reactor fleet and the currently available data on radiation effects on concrete, some decrease in mechanical properties of concrete cannot be ruled out during extended operation beyond 60 years. An expansion of the irradiated concrete database is desirable to ensure reliable risk assessment for extended operation of nuclear power plants.