Biological Shield Research Articles

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.

Radiological characterisation in view of nuclear reactor decommissioning: On-site benchmarking exercise of a biological shield

Nearly all decommissioning and dismantling (D&D) projects are steered by the characterisation of the plant being dismantled. This radiological characterisation is a complex process that is updated and modified during the course of the D&D. One of the tools for carrying out this characterisation is the performance of in-situ measurements.There is a wide variety of equipment and methodologies used to carry out on-site measurements, depending on the environment in which they are to be carried out and also on the specific objectives of the measurements and the financial and personnel resources available. The extent to which measurements carried out with different types of equipment or methodologies providing comparable results can be crucial in view of the D&D strategy development and the decision-making process.This paper concerns an on-site benchmarking exercise carried out at the activated biological shield of Belgian Reactor 3 (BR3). This activity allows comparison and validation of characterisation methodologies and different equipment used as well as future interpretation of final results in terms of uncertainties and sensitivities. This paper describes the measurements and results from the analysis of this exercise. Other aspects of this exercise will be reported in separate papers. This paper provides an overview of the on-site benchmarking exercise, outlines the participating organisations and the measurement equipment used for total gamma, dose rate and gamma spectrometry measurements and finally, results obtained and their interpretations are discussed for each type of measurement as a function of detector type.Regarding the dose measurements, results obtained by using a large variety of equipment are very consistent. In view of mapping the inner surface of the biological shield the most appropriate equipment tested might be the organic scintillator, the BGO or even the ionisation chamber. In addition, for mapping this surface, the most appropriate total gamma equipment tested might be the LaBr3(Ce), the thick organic scintillator or the BGO. These measurements can only be used as a secondary parameter in a relative way. Results for the gamma spectrometry are very consistent for all the equipment used and the main parameters to be determined.

Neutron beam preparation for soil moisture measurements: Nedis-PHITS and artificial neural networks study

The angular distribution of thermal neutrons scattered from the surface of a soil sample was used to determine soil moisture content. All simulations were performed assuming optimal dimensions of the designed collimator and AmO2-Be neutron source. Moreover, an assembly consisting of five Amersham X.l4 type capsular neutron sources was studied to obtain faster, more accurate data. The results showed promising agreement with previously published data and provided insights into potential applications for the construction of biological shielding of nuclear reactors.

Research on biological sealing for openings in nuclear power plant

Civil opening sealing is an important work during the construction stage of nuclear power projects, which directly affects the follow-up nuclear cleaning and system transferred to commissioning. The sealing of biological shielding openings mainly exists in the walls and floors across the boundary of the nuclear island buildings, which is with great significance to ensure the integrity of the boundary function of the radiation zone. The configurations for openings in concrete shield walls shall provide adequate attenuation to reduce the exposure of the workers. Through the research of the biological sealing for civil opening and the investigation of good practice for foreign nuclear facilities, the purpose of this paper is to optimize the opening sealing scheme based on construction and operation experience, and give the guidance for the construction application and save the cost for the nuclear project.

Radiation dose distribution of liquid fueled thorium molten salt reactor

A liquid fueled thorium molten salt reactor (TMSR-LF), one of the Generation IV reactors, was designed by the Shanghai Institute of Applied Physics, Chinese Academy of Sciences. This study uses the`rt code to calculate the neutron and gamma dose rate distributions around the reactor. Multiple types of tallies and variance reduction techniques were employed to reduce calculation time and obtain convergent calculation results. Based on the calculation and analysis results, the TMSR-LF1 radiation shield with a 60-cm serpentine concrete layer and a 120-cm ordinary concrete layer is able to meet radiation requirements. The gamma dose rate outside the reactor biological shield was 16.1 mSv h−1; this is higher than the neutron dose rate of 3.71 × 10–2 mSv h−1. The maximum thermal neutron flux density outside the reactor biological shield was 1.89 × 103 cm−2 s−1, which was below the 1 × 105 cm−2 s−1 limit.

VVER-1000 BENCHMARK INTERPRETATION WITH MONTE CARLO CODE MCS

An interpretation of the NEA-1517/82 benchmark from the SINBAD shielding database has been conducted with the MCS Monte Carlo code developed at the Ulsan National Institute of Science and Technology (UNIST) and the ENDF/B-VII.1 nuclear data library. The NEA-1517/82 benchmark corresponds to experiments on a VVER-1000 critical mock-up (thermal reactor with hexagonal fuel lattice) inside the LR-0 research reactor operated by the Nuclear Research Institute (NRI) in the Czech Republic. A new 3D model of the VVER-1000 mock-up core is developed for MCS based on the SINBAD documentation. The model includes the top and bottom parts of fuel pins, the spacer grids and core components: baffle, barrel, downcomer, tank, reactor pressure vessel (RPV) and concrete block used as biological shielding. The quality of the model is verified first by code/code comparison of MCS against MCNP6 for criticality and power distributions (pin-by-pin and axial power). The validation of MCS results is then performed against six critical cases, 260 measured pin powers and benchmark calculations of the axial power profile. Finally, a comparison of calculated and measured neutron spectra inside the mock-up core is presented as a preliminary study for upcoming works on the deep-penetration shielding capability of MCS.

Probing neutron-hidden neutron transitions with the MURMUR experiment

MURMUR is a new passing-through-walls neutron experiment designed to constrain neutron-hidden neutron transitions allowed in the context of braneworld scenarios or mirror matter models. A nuclear reactor can act as a source of hidden neutrons, such that neutrons travel through a hidden world or sector. Hidden neutrons can propagate out of the nuclear core and far beyond the biological shielding. However, hidden neutrons can weakly interact with usual matter, making possible for their detection in the context of low-noise measurements. In the present work, the novelty rests on a better background discrimination and the use of a mass of a material – here lead – able to enhance regeneration of hidden neutrons into visible ones to improve detection. The input of this new setup is studied using both modelizations and experiments, thanks to tests currently performed with the experiment at the BR2 research nuclear reactor (SCKcdot CEN, Mol, Belgium). A new limit on the neutron swapping probability p has been derived thanks to the measurements taken during the BR2 Cycle 02/2019A: p<4.0times 10^{-10} ; text {at 95}%text { CL}. This constraint is better than the bound from the previous passing-through-wall neutron experiment made at ILL in 2015, despite BR2 is less efficient to generate hidden neutrons by a factor of 7.4, thus raising the interest of such experiment using regenerating materials.

IMPROVEMENTS TO THE MODELING OF THE TREAT REACTOR AND EXPERIMENTS

This paper summarizes the latest improvements and lessons learned from the modeling and simulation of the transient test reactor at Idaho National Laboratory using the MAMMOTH reactor physics application. MAMMOTH is a MOOSE-based, Finite Element Method application that specializes in the analysis of the spatial dynamics behavior of nuclear reactors. Since early 2018 several transient tests have been conducted at TREAT, thus providing the opportunity to apply and benchmark modern modeling and simulation tools. MAMMOTH was used to provide predictions of the power coupling factor between the core and the experiment for various experiments. Even though the power coupling factor predictions agree very well with the experimental data, within the bounds of the experimental uncertainty, one shortcoming was the underprediction of the total energy deposited in the core and experiment. Determination of the sources for this discrepancy is ongoing, but several key problems have been identified and resolved, thus providing valuable insights for future research. This paper discusses several of these lessons learned. First, the heat capacity data for the TREAT fuel has some significant problems due to limitations of the measurement techniques used circa 1960s. The sensitivity of the peak power and the total energy deposition to various representations of the heat capacity is approximately 5%. Second, the effects of the biological shield and thermal column on the modeling of the core are non-negligible, since they affect the mean generation time and the effective reflection of neutrons back into the core, which is suspected to be important during the core heat up. Matching the reactor period resolves the fact that the reduced spatial domain used in the MAMMOTH model underpredicts the mean generation time. The neutron reflection from these regions is marginally improved with the use of an albedo boundary condition. Third, modeling of the control rod movement with a multi-scheme method is introduced and its current limitations are exposed. Fourth, we explore the effects of using a homogenized model with Superhomogenization equivalence and how that differs from fully heterogeneous simulations. Finally, the energy condensation effects for this graphite core are significant. Solutions with 10 and 26 energy groups show the benefits of using a finer coarse group structure.

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.

Providing rationale for the possibility of decommissioning Bilibino nuclear cogeneration plant based on the onsite disposal option

The problem of the NPP decommissioning after the end of the specified or extended life has reached the practical solution stage for countries possessing a nuclear power industry. The major decommissioning options, both in Russia and abroad, include immediate dismantling and deferred dismantling. At the same time, there are NPP units for which, for a number of reasons, none of the two options are acceptable in terms of ensuring the safety of the personnel, the public and the environment. Disposal, the third and a more rare option, shall be used for decommissioning in this case. The purpose of the work is to provide rationale for the possibility of decommissioning Bilibino Nuclear Cogeneration Plant based on the Onsite Disposal option by covering the main building with an inert material with the formation of a mound. The option has been selected considering the results of an integrated analysis taking into account the geographical, operational, radiological, and socioeconomic factors, as well as based on a limited experience of decommissioning commercial uranium-graphite reactors both within and outside Russia. In accordance with Russian law, the decommissioning stage will start after spent nuclear fuel is withdrawn from the unit and removed. Emphasis is placed on the proposed option preparation and implementation issues. Dates and sequences for the performance of operations to dismantle the components and civil works of buildings and structures, as well as the onsite protective mound formation structure and composition are discussed. The geometrical dimensions, as well as the quantities and types of the mound-forming materials have been estimated. The key mound-forming materials will be fragments of the components, the biological shielding, and the civil works, as well as local materials.

Study of a Gamma Radiation Impact on Concrete Properties Under Severe Accident Conditions

Abstract 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.

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

Dynamic crack propagation in the turtle carapace

The turtle carapace is a biological composite shield consisting of a hard boney layer covered with a soft keratin-collagen bi-layer. The unique microstructure enables the turtle carapace to protect the underlying soft tissues from predator assaults. Recent experiments have shown that when the turtle carapace is subjected to dynamic loading, brittle fracture of the boney layer occurs, which is accompanied by delamination at the keratin-collagen and collagen-bone interfaces. To unveil the role of interfacial delamination, we propose a model of fracture in the multilayered microstructure of turtle carapace, in which brittle cracking in the boney layer, delamination of the soft bi-layer and plastic deformation in the bi-layer are taken into account. Calculations are carried out for crack propagation in the turtle carapace under dynamic loading; it is found that low strength of the keratin-collagen interface can activate large plastic deformation in the collagen layer and promote spreading of damage region in the keratin-collagen interface, which delays fracture of the boney layer. High toughness of the keratin-collagen interface increases plastic deformation in the soft bi-layer and suppresses crack propagation in the boney layer. We further reveal the role of mechanical properties of the collagen-bone interface. It is identified that high strength of the collagen-bone interface enables enhanced plastic dissipation in the soft bi-layer, retarding crack growth in the boney layer. The collagen-bone interface with high toughness plays a role in enhancing plastic deformation in the soft bi-layer, which potentially increases the resistance to crack growth in the boney layer. In addition, we explore the effect of strain rate sensitivity of the soft bi-layer. We show that decreasing the degree of strain rate sensitivity of the keratin layer can mitigate fracture of the boney layer. The findings of this study shed new light on the protection mechanisms of turtle carapace.

Positron production using a 9 MeV electron linac for the GBAR experiment

For the GBAR (Gravitational Behaviour of Antihydrogen at Rest) experiment at CERN’s Antiproton Decelerator (AD) facility we have constructed a source of slow positrons, which uses a low-energy electron linear accelerator (linac). The driver linac produces electrons of 9 MeV kinetic energy that create positrons from bremsstrahlung-induced pair production. Staying below 10 MeV ensures no persistent radioactive activation in the target zone and that the radiation level outside the biological shield is safe for public access. An annealed tungsten-mesh assembly placed directly behind the target acts as a positron moderator. The system produces 5×107 slow positrons per second, a performance demonstrating that a low-energy electron linac is a superior choice over positron-emitting radioactive sources for high positron flux.

Will 37Ar emissions from light water power reactors become an obstacle to its use for nuclear explosion monitoring?

37Ar is a promising candidate for complementing radioxenon isotopes as indicators of underground nuclear explosions. This study evaluates its potential anthropogenic background caused by emissions from commercial pressurized water reactors. Various 37Ar production pathways, which result from activation of 36Ar and of 40Ca, respectively, are identified and their emissions quantified. In-core processes include (1) the restart of operation and degassing of the primary cooling water after maintenance and refueling shutdown, (2) the replacement of primary coolant water for limiting its tritium concentrations, and (3) the leakage of 37Ar produced from calcium impurities in UO2 after fuel rod failures. Activation of air and of calcium in concrete within the biological shield are major out-of-core production pathways. Whereas emissions from in-core processes are transient, a rather constant 37Ar source term results from its out-of-core production. Generic atmospheric dispersion simulations indicate that already at moderate distances from the emitter, concentrations of 37Ar caused by routine reactor operations are far below its cosmogenic background in air. The only exception results from an inadvertent reactor re-start without operation of the primary cooling water degassing system for prolonged time. Such an event also causes high emissions of 41Ar which can be used for discriminating its 37Ar signal from an underground nuclear explosion.

Influence of Slag Cement on the Permeability of Concrete for Biological Shielding Structures

Durability of concrete designed for radiation shielding structures is an important issue in nuclear power plant safety. An investigation of the permeability of concrete containing heavyweight aggregates and water-bearing aggregates was performed with respect to gaseous and liquid media. Mix design was developed using Portland and slag cement, crushed magnetite and serpentine aggregate. The use of slag cement in concrete containing magnetite and serpentine aggregates resulted in the substantial improvement of the compressive strength in comparison with Portland cement concrete. The application of slag cement was found to reduce the chloride ingress, regardless of the special aggregate use. The coefficient of chloride migration was within the range 5 ÷ 8 × 10−12 m2/s and 17 ÷ 25 × 10−12 m2/s for slag cement concrete and Portland cement concrete, respectively. At the same time, the carbonation depth was increased twice for slag cement concrete in comparison to Portland cement concrete. However, the maximum carbonation depth after one year of exposure to 1% CO2 was only 14 mm for slag cement concrete, and 7 mm for reference concrete. The total pore volume evaluated using mercury intrusion porosimetry was influenced by the type of special aggregate used. It was shown that concrete with various contents of magnetite aggregate and slag cement achieved the smallest total pore volume. While serpentine coarse aggregate caused an increase in total pore volume in comparison to concrete with magnetite aggregate.

Biofilm-encapsulated nano drug delivery system for the treatment of colon cancer

Aim In this study, 5-fluorouracil (5-FU) is delivered to target colon without the interference of mononuclear phagocyte system (MPS). Methods Outer membrane vesicles (OMVs) were used as the biological shield to disguise mesoporous silica (MSN) and 5-FU. OMVs-MSN-5-FU were prepared by high pressure co-extrusion, and characterised on the basis of size, drug loading, transmission electron microscope, infra-red spectroscopy, differential scanning calorimetry, thermal gravity analysis, % in vitro release, MTT assay, cell uptake and in vivo imaging. Results OMVs-MSN-5-FU with −18.22 ± 0.17 mV zeta potential and 90.4 ± 9.1 nm size were used for oral treatment of colon cancer. Drug loading of the drug was 50.22%±0.17 (w/w). The cumulative release of OMVs-MSN-5-FU reached 75.07%±0.94 in tumour microenvironment. The percentage of cell viability of OMVs-MSN-5-FU was 33.75%±2.73. In vivo experiments results confirmed that OMVs-MSN-5-FU could be taken up by colon cancer cells. Conclusions The study provided a promising nano platform for the targeting treatment of colon cancer.

Simulations of scattered neutrons for the time-of-flight enhanced diagnostics (TOFED) neutron spectrometer on EAST

The time-of-flight enhanced diagnostics (TOFED) neutron spectrometer has been installed in TOFED house, which is a newly-built neutron diagnostic lab outside the EAST experimental hall. The line of sight (LOS) of TOFED is defined by the collimator going through the 150 cm biological shielding wall, which can reduce background neutrons very effectively for the neutron spectral measurements. The Monte Carlo code MCNP5 is used in the simulations to characterize collimation effects and the radiation safety of the new house. The collimator makes neutron flux almost fully perpendicular to S1 detectors. In addition, the proportion of scattered neutrons in the LOS of the TOFED is obtained for the measured spectral data interpretation. The simulation results show that the TOFED house satisfies the requirement for the TOFED operations at EAST and gives a safe environment for personnel access during the EAST operation.