Biological Shield Research Articles

CHARACTERISTICS OF NEUTRON FIELDS AROUND BIOLOGICAL SHIELDING OF RESEARCH NUCLEAR REACTORS IN RUSSIA

The paper presents the results of a study of the neutron field characteristics surrounding biological shielding of Russian research nuclear reactors to establish their expected behavior and enhance personnel neutron monitoring. The study covered five nuclear research reactors: IVV-2M, IRT-T, IRT-MEPhI, RBT-6, and SM-3. All reactors are pool-typed; some (RBT-6 and SM-3) have pressurized water, while others (IVV-2M, IRT-T, and IRT-MEPhI) have water under normal pressure. The neutron fields analyzed are located at the tank cover of reactors, in front of the reactor core covered with biological shielding, and in front of the horizontal experimental channel. The spectrum average neutron energy ranges from 0.01 MeV (RBT-6) to 1.0 MeV (SM-3). The fluence-to-ambient dose conversion coefficient varies from 15 pSv cm2 (RBT-6) to 260 pSv cm2 (SM-3). The spectra at the other measurement points exhibit the energy distribution characteristics of the neutron radiation flux density behind the biological shielding.

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.

Mass Optimization of a Multilayered Shield for Transportable Microreactors

The ability to easily transport microreactors is a major selling point for deploying microreactors to remote areas. However, this creates a unique shielding challenge, especially when the microreactor is being shipped after irradiation. A traditional reactor configuration utilizes a separate biological shield and pressure vessel to meet radiological shielding and pressure needs. The limited space available for transportable microreactors for both shielding and pressure vessels requires a revised assessment of separating out the biological shield and pressure vessel. To address these concerns, we examine a nuclear-grade sandwich composite (NGSC) that combines the reactor pressure vessel and biological shielding functions into a single component. Through a series of optimization problems for both transportation and operational use cases, the NGSC is able to minimize dose, minimize the vessel cost, and ensure that weight requirements are met for transportation. Initial results show that using a tungsten-tetraboride cermet in the first two layers of a six-layer NGSC provides adequate shielding for both use cases. These results show promise that an NGSC has enough overlap between operational and transportation cases to help reduce the design space for future analysis and assessment.

Measurements of the Spectral and Dose Characteristics of the Neutron Field Behind the Biological Shielding of the IREN Source for 60-MeV Electrons

Measurements of the Spectral and Dose Characteristics of the Neutron Field Behind the Biological Shielding of the IREN Source for 60-MeV Electrons

Methodology for shielding design and evaluation for container scanners.

There has been an increase in the use of high energy photon beam for container scanners in many countries for multi purposes such as detecting high atomic number materials which might be nuclear materials, drugs, high explosive materials and other contrabands etc. High energy photon beams generally 6 and 9 MV can be used for scanning such materials. However, it is important to ensure that radiation level beyond the container scanner installation is within the permissible dose limit specified by the national competent authority for the protection of public and radiation workers. In this paper, challenges in the biological shielding during the installation of high energy X-ray system for scanning vehicles containing suspected materials are discussed. The purpose of the present study is to develop a methodology for shielding design and evaluation for container scanner installations. The basic concept pertaining to shielding evaluation of radiotherapy installations provided in National Council on Radiation Protection and Measurements (NCRP)/International Atomic Energy Agency (IAEA) reports are referred, and appropriately used to calculate optimized shielding thicknesses requirements for container scanner installation. Workload is estimated based on number of containers scanned, machine ON time and dose rate at 1m. The shielding evaluation includes use of beam stopper in the primary beam, scattering by heterogeneous metallic scrap materials or any other suspected materials contained in the vehicle and their impact on the thickness of shielding walls. A model lay out plan to be used for installation of container scanner is developed. A methodology for shielding evaluation for various protective walls and ceiling of this model is also discussed. The study provides basic requirement for designing a structural room for installing 9MV container scanner from radiological safety view point.

Determination of the Activity of Long-Lived $${}^{{41}}$$Ca in Reactor Biological Shielding Materials by Photoactivation Method

Determination of the Activity of Long-Lived $${}^{{41}}$$Ca in Reactor Biological Shielding Materials by Photoactivation Method

Characterisation of the neutron field for streaming analyses in TT operations at JET

Assessing radiation fields in the biological shield penetrations of fusion reactors is a challenging task. At the Joint European Torus (JET) the neutron field at larger distances from the torus has been calculated and measured. JET operated in 2021–22 with a tritium-tritium plasma and neutronics experiments were performed for validating in a real fusion environment the neutronics codes and nuclear data applied in ITER nuclear analyses. In particular, the fluence of neutrons passing through the penetrations of the JET vacuum vessel and the torus hall was measured and compared with calculations in order to assess the capability of state-of-the-art numerical tools to correctly predict the radiation streaming in large and complex geometries. The neutron fluence was monitored at several locations inside the torus hall at larger distances from the tokamak with activation foils and thermo-luminescent detectors. The calculations have been performed in a two-step process using the deterministic code ADVANTG to determine the variance reduction parameters and with MCNP for subsequent calculation of the neutron field with the Monte Carlo method. The paper presents results of calculations and the first comparison to experimentally obtained values.

Updates of the removable biological shielding blocks inside of the test cell, part of the test systems of IFMIF-DONES

The International Fusion Materials Irradiation Facility - DEMO Oriented Neutron Source (IFMIF-DONES) is a facility which is designed under the framework of the EU fusion roadmap. It is going to be an essential irradiation facility for testing and qualifying candidate materials under severe irradiation conditions of a neutron field having an neutron irradiation effect on materials like the one expected in a commercial fusion power reactor. The material specimens are irradiated in a containment structure named Test Cell (TC), which is part of the Test Systems (TS).The first protecting „barrier” against irradiation affecting the surrounding of the TC, are the Removable Biological Shielding Blocks (RBSBs). These ∼9 m height, ∼80 tons elements, which are formed by stainless steel liners filled with heavy concrete, need to be remotely handled as after the first experiments they will get dose rates above the hands-on limit. Irradiation will also result in a large nuclear heat power deposited in the shielding blocks, therefore needs to be actively water cooled by a system of embedded pipes to control the temperature.In this paper the updated design of the RBSBs is described, including the latest achievements and proposals for feasible manufacturing, lifting and positioning possibilities of the blocks inside of the TC respecting the given tolerances.

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.

Improving Neutron Shielding Capacities of Galena and Barite Hybrid Fiber Heavyweight Aggregate Concrete for Nuclear Reactor

Biological shielding of nuclear reactors is an important interest and diminishing the intricacy and cost of these installations is an important interest. In this paper, we used galena minerals and barite for the production of a hybrid fiber heavyweight aggregate concrete. Galena minerals that exist in many parts of Iran were used in the concrete mix design. Barite is an important chemical element for neutron absorption processes. The cross-section in matter and neutron capture is a utilizable cause to explain the neutron shielding characteristics of samples. Neutron cross-section measurements of samples were done by using a source of 4.5 MeV neutrons. Cross-section and neutron capture of each sample were calculated by using the Geant 4 Monte Carlo code. As a result, the use of appropriate galena concentration and barite and the use of monofilament polypropylene fiber with steel fiber can improve cross cross-section value of hybrid fiber heavyweight aggregate concrete and enhance the properties of neutron shielding.

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.

Pump availability prediction using response surface method in nuclear plant

The safety-related raw water system's strong operational condition supports the radiation defense and biological shield of nuclear plant containment structures. Gaps and failures in maintaining proper working condition of main equipment like pump were among the most common causes of unavailability of safety related raw water systems. We integrated the advanced data analytics tools to evaluate the maintenance records of water systems and gave special consideration to deficiencies related to pump. We utilized maintenance data over a three-and-a-half-year period to produce metrics like MTBF, MTTF, MTTR, and failure rate. The visual analytic platform using tableau identified the efficacy of maintenance & deficiency in the safety raw water systems. When the number of water quality violation was compared to the other O&M deficiencies, it was discovered that water quality violations account for roughly 15% of the system's deficiencies. The pumps were substantial contributors to the deficit. Pump availability was predicted and optimized with real time data using response surface method. The prediction model was significant with r-squared value of 0.98. This prediction model can be used to predict forth coming pump failures in nuclear plant.

Reliability and availability simulation for vessel cooling system of the high temperature engineering test reactor

The article deals with the problem of risk analysis for new and unique nuclear facilities where safety-related systems are continuously operated during normal exploitation. A non-standard approach to system modeling was applied which goes beyond traditional method suitable for on-demand safety systems with a fixed mission time. As an example, the Vessel Cooling System (VCS) of the Japanese High Temperature Engineering Test Reactor (HTTR) has been analyzed in terms of life-cycle reliability and availability. Normal operation and emergency conditions have been considered, differing in the requirements for the system performance. In the former, the VCS reliability for cooling the biological shield around the Reactor Pressure Vessel (RPV) has been examined. In the latter, the VCS availability to remove the residual heat from the reactor core and RPV has been simulated. The results are of high importance for the safety and profitability of the HTTR-based electricity-hydrogen cogeneration plant (HTTR-GT/H2). The high reliability of the VCS under emergency conditions has been confirmed, thus contributing to the probabilistic safety assessment (PSA). Moreover, the fraction of the Forced Outage Rate (FOR) caused by VCS failures has been determined, thus contributing to the plant profitability studies.

Remote infrared view of JET divertor compatible with D-T operations

A remote viewing system located outside the biological shield wall to image the divertor region in JET is presented here. This paper focuses on the optical and mechanical designs of the installation of a line of sight which is capable of imaging the divertor over an optical distance of approximately 31 m. In addition, it includes the experience obtained during operations and it demonstrates its successful performance. This line of sight was the second viewing system used during the D-T campaign and it follows from the paper dedicated to the remote wide angle view line of sight which is also compatible with D-T operations and includes common elements to the construction of the two lines of sight. The main novel feature is the design in house of two glass reinforced plastic kinematic mounts standing vertically to each accommodate a mirror of 500 mm in diameter.

Aspergillus niger as a cell factory for the production of pyomelanin, a molecule with UV-C radiation shielding activity.

Melanins are complex pigments with various biological functions and potential applications in space exploration and biomedicine due to their radioprotective properties. Aspergillus niger, a fungus known for its high radiation resistance, is widely used in biotechnology and a candidate for melanin production. In this study, we investigated the production of fungal pyomelanin (PyoFun) in A. niger by inducing overproduction of the pigment using L-tyrosine in a recombinant ΔhmgA mutant strain (OS4.3). The PyoFun pigment was characterized using three spectroscopic methods, and its antioxidant properties were assessed using a DPPH-assay. Additionally, we evaluated the protective effect of PyoFun against non-ionizing radiation (monochromatic UV-C) and compared its efficacy to a synthetically produced control pyomelanin (PyoSyn). The results confirmed successful production of PyoFun in A. niger through inducible overproduction. Characterization using spectroscopic methods confirmed the presence of PyoFun, and the DPPH-assay demonstrated its strong antioxidant properties. Moreover, PyoFun exhibited a highly protective effect against radiation-induced stress, surpassing the protection provided by PyoSyn. The findings of this study suggest that PyoFun has significant potential as a biological shield against harmful radiation. Notably, PyoFun is synthesized extracellularly, differing it from other fungal melanins (such as L-DOPA- or DHN-melanin) that require cell lysis for pigment purification. This characteristic makes PyoFun a valuable resource for biotechnology, biomedicine, and the space industry. However, further research is needed to evaluate its protective effect in a dried form and against ionizing radiation.

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.

Reproduction of shielding concrete activation measurements by simulations.

With the increasing number of decommissioning projects, radioactive waste estimation of biological shielding concretes is becoming more and more important. Simulation tools supporting this activity, like MCNP and Cinder are already available; however, publicly available neutron spectra in shielding concretes are limited. The aim of the study was to present and evaluate possible model arrangements for accurate neutron transport to deeper points of shielding concretes from the reactor pressure vessel. Representation of reality, neutron behavior and activity generation of seven long-lived radioisotopes (54Mn, 60Co, 65Zn, 133Ba, 134Cs, 152Eu and 154Eu) were assessed in each arrangement. After evaluating several model geometries, a conical shape neutron reflecting surface was shown to be the most suitable one to reproduce neutron fields in deeper layers of shielding concretes from a monodirectional initial neutron source.

Remote wide angle view broad wavelength viewing system compatible with D-T operations in JET

Imaging diagnostics in Joint European Torus have become essential in the study of plasma wall interactions and the protection of the plasma facing components. During deuterium operations, the location of these diagnostics was at close proximity to the vessel. However, for the 50–50 deuterium-tritium operation, the neutron yield impact onto the electronics of these imaging diagnostics would have caused them to fail completely at these locations. Shielding these systems from neutrons is achieved by relocating a selected number of these imaging systems to outside the Torus Hall (i.e. to the other side of the biological shield wall). The relocation encompasses the construction of two new lines of sight crossing through the biological shield wall and each of them using an optical relay which extends to separate ports in the machine. This article starts with a brief description of two new viewing systems. However, it focuses on the optical modelling and engineering design of one of them which is a multiple camera system producing a wide angle view (WAV) of the inside of the vessel. This WAV system has been used successfully during the recent D-T campaign. It operates in a broadband wavelength range extending from the visible to the middle infrared. It also incorporates a novel in-house horizontal kinematic mirror mount made from glass reinforced plastic as part of the optical relay. The operational stability of the WAV system has also been assessed and some post design improvements implemented.

Investigation of the effect of cement type on nuclear shield performance of heavy concrete

This article studies the radiation shielding performance of three different concrete-added samples. First, the mass attenuation coefficient values of three distinct concrete materials were computed using the MATLAB code at energies ranging from 0.015 to 15 MeV, and the findings obtained were verified by theoretical WinXcom results. Half-Value Layer (HVL), Tenth-Value Layer (TVL), Mean Free Path (MFP), Effective Atomic Number (Zeff), Exposure Build-Up Factor (EBF), Energy Absorption Build-Up Factor (EABF) are the variables computed to assess available concrete samples in terms of radiation attenuation characteristics. The results demonstrated that the concrete sample with CAC possesses better gamma-radiation shielding characteristics. Concrete samples’ removal cross-sections (∑R) were obtained to analyze their ability to stop fast neutrons. Iron, aluminum, and lead-added concretes are better in terms of their density than the other types of concretes. Given the relationship between the density of the material and neutron shielding propensities, iron, aluminum, and lead-added materials are expected to better serve for shielding purposes. SRIM Monte Carlo code, as it is standard, was utilized to compute the mass stopping power (MSP) and projection range (PR) values. The concrete sample with CAC has shown better shielding properties compared to the others in the sample of the concretes having the lowest HVL, TVL, and MFP and highest ∑R. For proton and alpha particles, the concrete sample with CAC sample shows the best radiation shielding and protection properties. The concretes selected for the study are useful for shielding purposes compared to many other materials used in industry. In addition, radiation shielding concrete is cheaper than other shielding materials. It is easier to mold into different geometric shapes and is suitable for shielding from protons or alpha particles. Among the different concretes investigated in this study, it can be used for structures such as radiation oncology and nuclear medicine clinics in hospitals. Additionally, it can act as a biological shield for nuclear power plants, particle accelerators, and radionuclide laboratories, as well as other sources of radiation.