Journal articles: 'Nuclear reactors Shielding (Radiation)' – Grafiati (2024)

Abstract:

In nuclear reactors, inelastic neutron scattering is a significant energy-loss mechanism which has deep impacts on designments of nuclear reactor and radiation shielding. Iron is an important material in reactor. However, for the existing nuclear data for iron, there exists an obvious divergence for the inelastic scattering cross sections and the related gamma production sections. Therefore the precise measurements are urgently needed for satisfying the demanding to design new nuclear reactors (fast reactors), Accelerator Driven Subcritical System (ADS), and other nuclear apparatus. In this paper, we report a new system with an array of HPGe detectors, electronics and acquisition system. Experiments had been carried out on three neutron facilities.

Abstract:

Shielding of a reactor is required for protection of people and environment during normal operation and accidental situations. In the present paper we investigated the shielding parameters viz. mass attenuation coefficients, linear attenuation coefficients, tenth-value layer, effective atomic numbers, kerma relative to air and exposure buildup factors for gamma-ray for ordinary, heavy, and super heavy concretes. Macroscopic effective removal cross-sections for fast neutron had also been calculated. Ordinary concrete is economically suitable for mixture high energy gamma-ray and neutron as it has large weight fraction of low-Z as compared with super heavy concretes to slow down the neutron. Super heavy concretes are superior shielding for both reactor operation and accident situations. The study is useful for optimizing for shielding design and radiation protection in the reactors.

Abstract:

Abstract The most important material for shielding is concrete in nuclear facilities which performance can be improved by addition some Nanoparticles (NP) at the various concentrations. Nanoparticles, which have a distinctive potential for bio-radiation and shielding of nuclear reactors, are used in many areas due to their special characteristics, which lead to an improvement in the mechanical properties and the pore structure of the concrete shield. The aim of this research was to initiate a novel coefficient (n), experiment to theory density ratio for integrated NP at different nanoparticle concentrations (xnano), established upon purely mathematical viewpoints and some appropriate physical objectives.

Abstract:

Concrete is a relatively cheap material and easy to be cast into variously shaped structures. Its good shielding properties against neutrons and gamma-rays, due to its intrinsic water content and relatively high-density, respectively, make it the most widely used material for radiation shielding also. Concrete is so chosen as biological barrier in nuclear reactors and other nuclear facilities where neutron sources are hosted. Theoretical formulas are available in nuclear engineering manuals for the optimum thickness of shielding for radioprotection purposes; however they are restricted to one-dimensional problems; besides the basic empirical constants do not consider radiation damage effects, while its long-term performance is crucial for the safe operation of such facilities. To understand the behaviour of concrete properties, it is necessary to examine concrete strength and stiffness, water behavior, volume change of cement paste, and aggregate under irradiated conditions. Radiation damage process is not well understood yet and there is not a unified approach to the practical and predictive assessment of irradiated concrete, which combines both physics and structural mechanics issues. This paper provides a collection of the most distinguished contributions on this topic in the past 50 years. At present a remarkable renewed interest in the subject is shown.

Abstract:

Abstract Recently, technologists try to develop novel gamma radiation shielding materials instead of traditional materials such as lead and concrete with improved performance in gamma radiation shielding in medical applications and nuclear reactors. For this purpose, alloys such as stainless steel (SS) and carbon steel (CS) attracted much attention, these days. Preliminary results on such alloys have shown better attenuation of γ rays as compared to traditional shielding materials. This work aimed to conduct research on different alloy samples to evaluate their radiation attenuation efficiency and their suitability for radiation shielding when utilized in nuclear facilities. The mass attenuation coefficients for eight alloy samples were measured at different photon energies ranging from 80.997 to 1332.501 keV using transmission geometry. From the mass attenuation coefficients, different photon attenuation parameters such as half value layer, mean free path, effective atomic number, and radiation protection efficiency were evaluated. In addition, the equivalent atomic number and the exposure buildup factor were calculated using G-P fitting method for photon energy ranging from 0.015 MeV to 15 MeV at different penetration depth. The results showed that the Zeff values remain almost constant for all samples except W72/Cu28 in which the Zeff for this sample tends to decrease with the energy. The lowest value of half value layer is found for the alloy sample Ta97.5/W2.5 and the highest value is found for the alloy sample In50/Sn50. The Ta97.5/W2.5, Ta90/W10, Ta95/W5 samples demonstrated good radiation attenuation properties.

Abstract:

The nuclear reactor emits various kinds of nuclear radiations during operation. However, the main contributions to the radiation field in the reactor are given by neutrons and gamma rays. These radiation components are the principal concern of reactor shielding. In our study the neutron and gamma radiation dose rates at different depths in concrete bio-shield of the PWR were calculated based on spherical model for M-C simulation with using MCNP4C2. The simulation results were compared with the results obtained from calculations based on S$_{8}$P$_{3}$ spherical approximation with using the ANISN code.

Abstract:

The optimization of radiation protection is an important task in both the design and operation of a nuclear power plant. Although this topic has been considerably investigated for pressurized water reactors, there are very few public reports on it for pebble-bed reactors. This paper proposes a routine that jointly optimizes the system design and radiation protection of High Temperature Reactor-Pebble-Bed Module (HTR-PM) towards the As Low As Reasonably Achievable (ALARA) principle. A systematic framework is also established for the optimization of radiation protection for pebble-bed reactors. Typical calculations for the radiation protection of radioactivity-related systems are presented to quantitatively evaluate the efficiency of the optimization routine, which achieve 23.3%~90.6% reduction of either dose rate or shielding or both of them. The annual collective doses of different systems are reduced through iterative optimization of the dose rates, designs, maintenance procedures, and work durations and compared against the previous estimates. The comparison demonstrates that the annual collective dose of HTR-PM is reduced from 0.490 man-Sv/a before optimization to 0.445 man-Sv/a after optimization, which complies with the requirements of the Chinese regulatory guide and proves the effectiveness of the proposed routine and framework.

Abstract:

Renewable source of energy is the best form of consumable energy that can be harnessed from the earth’s available resources in a sustainable manner. Nuclear energy obtained from nuclear fission reactions from the nuclear reactors is an example of such kind of energy, but deeply requires caution in the handling of operations. Leakage of fast moving high energy neutrons creates harm to both the environment and mankind and hence should be shielded from all available means. One such effort to shield this high energy radiation is by using boron carbide infused cement mortar that can be used for plastering and other similar applications in the construction of nuclear reactors. The strength and performance based characteristics of such type of mortar are studied and improved by adding super-plasticiser and pozzolanic materials like microsilica and metakaolin. Also durability studies like alkalinity, water absorption etc are done to analyse the lifetime of the design mix. EDS and SEM analysis were also performed extensively to study the microstructure of the casted specimen and to analyse the elemental composition cum distribution (by EDS mapping) for the calculation of the neutron attenuation of the specimen mix. Based on these an optimum combination is arrived at for practical applications that has a desirable strength, performance, durability and neutron shielding property.

Abstract:

The paper presents a new composite material for radiation shielding properties. This material is based on super-heavy concrete reinforced with basalt fiber, which could be used in biological protection systems for neutron radiation sources. The simulation of the neutron transport in the presented material was performed using the Monte Carlo Serpent code. Two types of heavy concretes were considered in the present paper: 1) with ordinary rubble coarse aggregate and 2) with barite coarse aggregate. For each type of concrete, the basalt fiber with dosage from 1 kg/m3 to 50 kg/m3 was added. The current transmission rates were obtained as a result of the neutron-physical modelling for neutron transport from source to detector through the proposed concrete samples with different thicknesses. The obtained modelling results were analyzed from the viewpoint of effectiveness of the radiation shielding properties. Also, elastic and capture microscopic cross-sections were considered for some isotopes and as a result, some aspects of the radiation shielding properties were clarified. The concrete with ordinary rubble coarse aggregate has better radiation shielding properties in case of low concrete thicknesses due to better neutron scattering on light nucleuses. In contrast to this, the concrete with barite coarse aggregate has better radiation shielding properties in case of high concrete thicknesses due to better neutron absorption. It is shown that the addition of basalt fiber to concrete not only improves its mechanical properties and reduces the number and size of microcracks, but also increases the ability to protect against neutron flux. The proposed composite material could be recommended for use with the following neutron sources: (D, T) neutron generators, plasma focus devices, fusion reactors and fast reactors. This research was carried out with the financial support of the IAEA, within the terms and conditions of the Research Contract 20638 in the framework of the Coordinated Research Project (CRP) “Accelerator Driven Systems (ADS) Applications and use of Low-Enriched Uranium in ADS (T33002)” within the project ‘The Two-Zone Subcritical Systems with Fast and Thermal Neutron Spectra for Transmutation of Minor Actinides and Long-Lived Fission Products’.

Abstract:

The radiation environment inside and near spacecraft consists of various components of primary radiation in space and secondary radiation produced by the interaction of the primary radiation with the walls and equipment of the spacecraft. Radiation fields inside astronauts are different from those outside them, because of the body's self-shielding as well as the nuclear fragmentation reactions occurring in the human body. Several computer codes have been developed to simulate the physical processes of the coupled transport of protons, high-charge and high-energy nuclei, and the secondary radiation produced in atomic and nuclear collision processes in matter. These computer codes have been used in various space radiation protection applications: shielding design for spacecraft and planetary habitats, simulation of instrument and detector responses, analysis of absorbed doses and quality factors in organs and tissues, and study of biological effects. This paper focuses on the methods and computer codes used for radiation transport calculations on cosmic radiation, and their application to the analysis of radiation fields inside spacecraft, evaluation of organ doses in the human body, and calculation of dose conversion coefficients using the reference phantoms defined in ICRP Publication 110.

Abstract:

High Density concrete is commonly used for radiation shielding of nuclear-reactors and other structures like counter weights, coating of off-shore pipelines. High Density concrete or is designed by using heavy weight aggregates such as hematite, magnetite, barite etc. The material is called hematite is used in this special concrete. The mix used is for OPC 43grade of cement. High density concrete can be designed in same way as normal weight concretes, but the additional self-weight should be taken into account.

Abstract:

Concrete is a very important composite for making different building structures to absorb different levels of radiation. Nuclear power plants, nuclear research reactors, particle accelerators and linear accelerator in medical institution use concrete in building construction. Nanoparticles or nanocrystals have different properties than their bulk counterparts. The gamma radiation absorption characteristics and beta back scattering by nanoparticles is also different than their counterparts. In this paper, carbon nanoparticles are mixed in the concrete composite during its preparation. The concrete composite with carbon nanoparticles as admixture were analyzed to provide radiation protection. The gamma radiation absorption characteristics and beta back scattering in ordinary and carbon nanoparticles contained concretes have been studied by GM counter. The results show that using carbon nanoparticles as an admixture in the concrete is one of the solutions for gamma ray shielding and beta back scattering. Therefore, it is good to use carbon nanoparticles as admixtures in concrete composites for beta and gamma radiation scattering and absorption respectively.

Abstract:

Introduction. Dry shielding is a cylindrical structure made of serpentinite concrete in a metal casing with an inner diameter of 5.6 m, an outer diameter of 6.7 m, and a height of 5.3 m, which surrounds the VVER reactor vessel in the vicinity of the core. The purpose of serpentinite concrete, containing an increased amount of chemically bound water, is to soften the spectrum of the neutron flux outside the reactor, increasing the fraction of thermal neutrons in the spectrum, which is necessary for the operation of ionization chambers (IR) of the reactor control and protection system. Dry shielding also performs the functions of radiation and thermal protection, reducing the flux of radiation on ordinary concrete of biological protection. Before the installation of the dry shielding in the reactor shaft, heat treatment (drying) of concrete is carried out at temperatures up to 250 °С to remove unbound water in order to avoid radiolysis. Quality control of concreting and then heat treatment is carried out using a radioisotope device — a neutron moisture meter. These works are very lengthy and costly. Materials and methods. The design of the dry shielding casing was considered in order to perform additional perforation in order to avoid the formation of air pockets during concreting. The possibility of using modern plasticizing additives was considered in order to minimize the consumption of mixing water and, as a result, free water in the body of serpentinite concrete. Results. The possibility of exclusion of the stages of quality control of concreting and heat treatment in their traditional form is shown. Additional perforation of the metal casing, its internal diaphragms in problem areas, the use of a mixture of 20 cm slump or more allows you to completely eliminate the formation of internal voids. According to preliminary estimates, given the intensity of radiation in the NW for a modern reactor with a capacity of 1200 MW, the intensity of the release of hydrogen outside the shell due to radiolysis does not pose any danger. The concentration of hydrogen in the air surrounding the dry shielding is many orders less of magnitude than the dangerous 4 %. Conclusions. The cost of work on the construction of the SZ power unit of a nuclear power plant with a capacity of 1000–1200 MW can be reduced by 70–100 million rubles, the duration of work by 5 months.

Abstract:

Today, mass nuclear weapons and reactor plants are becoming more prominent. However, current methods of radiation shielding are not viable due to heavy cost and ineffective means of weakening photon momentum. Therefore, it becomes necessary to design structures resistant to the behavior of radiation from exposing to human life. Specifically, 280 computational experiments were conducted in a SPENVIS environment utilizing Multi-Layered Shielding Simulation (MULASSIS) and Geant4 Radiation Analysis for Space (GRAS) on multiple shielding models. These thirteen models tested against nuclear, artificially-generated incident particles under single and multi-ray analyses with four angular photon distributions in comparison to SHEILDOSE, a current standard for cosmic radiation shielding developed by the European Space Agency. These designs using stainless steel, lead, slightly-radioactive bismuth, and lithium-hydride prevented over 99% of particle detection compared to SHIELDOSE, which conversely increased the neutron-energy dose by over 700%, and insufficiently reduced high-energy gamma ray penetration. Per kilogram, my model is 144 times cheaper and only a small fraction of the thickness of either SHIELDOSE or metal foams. Thus, the potential of enhanced nuclear plants, further space exploration, and an overall safer approach to utilizing or preventing exposure to atomic particles such as with multi-disaster protection buildings can become more readily available, thus saving millions of lives that are in impending danger.

Abstract:

Gamma rays are an inextricable part of a nuclear reactor’s radiation field, and as such require characterization for dose rate estimations required for the radiation protection of personnel, material choices, and the design of nuclear facilities. Most commonplace radiation transport codes used for shielding calculations only included the prompt neutron induced component of the emitted gamma rays. The relative amount of gamma rays that are emitted from delayed processes – the delayed gamma fraction – amount to a significant contribution, e.g. in a typical zero power reactor at steady state is estimated to be roughly a third. Accurate predictions of gamma fields thus require an estimation of the delayed content in order to meaningfully contribute. As a consequence, recent code developments also include delayed gamma sources and require validation data. The CROCUS zero power research reactor at EPFL is part of the NEA IRPhE and has therefore been characterized for benchmark quality experiments. In order to provide the means for delayed gamma validation, a dedicated experimental campaign was conducted in the CROCUS reactor using its newly developed gamma detection capabilities based on scintillators. In this paper we present the experimental determination of the delayed gamma fraction in CROCUS using in-core neutron and gamma detectors in a benchmark reactor configuration. A consistent and flexibly applicable methodology on how to estimate the delayed gamma fraction in zero power reactors has hitherto not existed – we herein present a general experimental setup and analysis technique that can be applied to other facilities. We found that the build-up time of relevant short lived delayed gamma emitters is likely attributed to the activation of the aluminium cladding of the fuel. Using a CeBr3 scintillator in the control rod position of the CROCUS core, we determined a delayed gamma fraction of (30.6±0.6)%.

Abstract:

Photon induced reactions are called as photonuclear reactions and used in many research fields of nuclear science and nuclear physics. The photonuclear data are used in many nuclear applications such as radiation shielding and protection, radiation transport analyses, reactor core design, activation analysis and nuclear waste transmutation. In the past, many studies had been devoted to extract photonuclear data covering the isotopic chart. However, there is still lack of existing data. In the present study, we have performed photonuclear reactions on erbium (Er) target by using clinical electron linear accelerators (cLINAC). By using measured residual activity of photonuclear reaction products of Er nuclei, we have determined the half-life of [Formula: see text]Er nucleus and transition energies of [Formula: see text]Ho nucleus. Also, new measurements on gamma-ray energies of the products have been determined accurately. Furthermore, this study shows that repurposed cLINAC with limited budget can contribute to the global nuclear science knowledge.

Abstract:

ARES is a multidimensional parallel discrete ordinates particle transport code with arbitrary order anisotropic scattering. It can be applied to a wide variety of radiation shielding calculations and reactor physics analysis. To validate the applicability of the code to accelerator shielding problems, ARES is adopted to simulate a series of accelerator shielding experiments for 43 MeV proton-7Li quasi-monoenergetic neutrons, which is performed at Takasaki Ion Accelerator for Advanced Radiation Application. These experiments on iron and concrete were analyzed using the ARES code with FENDL/MG-3.0 multigroup libraries and compared to direct measurements from the BC501A detector. The simulations show good agreement with the experimental data. The ratios of calculated values to experimental data for integrated neutron flux at peak and continuum energy regions are within 64% and 25% discrepancy for the concrete and iron experiments, respectively. The results demonstrate the accuracy and efficiency of ARES code for accelerator shielding calculation.

Abstract:

Shielding analysis and design are important tools for ensuring that humans and the environment are protected from the detrimental effects of high levels of radiation. The fundamental principles and methodologies for shielding analysis and design, especially for reactor applications, have been developed and refined since the 1940s and the beginning of nuclear power research programs in Canada and internationally. Other applications are gaining importance due to both increased need and technological advances. In this work, a high-level survey of emerging areas in shielding research and development is provided. Areas of topical interest include remote reactor monitoring, source reconstruction and inverse shielding methods, waste management and decommissioning applications, accelerator, cyclotron, and other advanced medical shielding applications, space exploration, and new materials development. Each of these areas of interest is evaluated based on current capacity of the research community. They are also evaluated in terms of the benefits for the scientific community and industry arising from performing research including development of new technologies and techniques.

Abstract:

The spherical fuel elements are adopted in the high-temperature gas-cooled reactor pebble-module (HTR-PM). The fuel elements will be discharged continuously from the reactor core and transported into the fuel transport pipelines during the reactor operation, leading to spatially varying dose outside the pipeline. In this case, the dose evaluation faces two major challenges, including dynamic source terms and pipelines with varying lengths and shapes. This study tries to handle these challenges for HTR-PM through comprehensive calculations using the QAD-CGA program and to design the corresponding shielding of the pipeline. During the calculation, it is assumed that a spherical fuel element stays in different positions of the pipelines in turn, and the corresponding dose contributions were calculated. By integrating the dose contributions at different positions, the dose at the points of interest can be obtained. The total dose is further determined according to the assumed fuel elements transport speed of 5 m/s and total 6000 fuel elements transportation per day. Two types of fuel transport pipelines and two source terms were considered, i.e., the spent fuel element transport pipelines with corresponding spent fuel source term and the different burn-up fuel element transport pipelines with the average burn-up fuel source term. Doses at different points of interest were calculated with no shielding scenario and with lead shielding of different thicknesses scenario. To evaluate the shielding effect, the dose limit of the orange radiation zone of HTR-PM and the radiation damage thresholds from NCRP report No.51 were both adopted. The calculated results show that, for pipelines that transport the spent fuel, a 4 cm lead shielding will be enough. And for pipelines that transport fuel elements with different burn-up, a 5 cm lead shielding will be added. The method and results can provide valuable reference for other work of HTR-PM.

Abstract:

The National Nuclear Energy Agency (BATAN) Yogyakarta uses two kinds of paraffin for shielding radiation of Kartini reactor. For developing BNCT research, the radiation attenuation capability of paraffin has been analyzed to find out the coefficient attenuation, density, and composition of both kinds of paraffin. The components of the paraffin were analyzed using Fourier transform infrared (FTIR) spectroscopy, scanning electron microscopy (SEM), and energy dispersive X-ray (EDX) spectroscopy characterization. Paraffin P1 has a density of 0.689 gr/mL and paraffin P2 is 0.578 gr/mL. Paraffin samples P1 and P2 were the sample content of functional group CH, CH2, and OH when analyzed by FTIR. Paraffin P2 had an additional content namely CO. The concentration of carbon (C) and oxide (O) of paraffin P2 was much greater than that of paraffin P1. Hydrogen (H) in the paraffin has the function of moderating neutrons, but hydrogen content in both kinds of paraffin could not be detected by EDX. The acquired neutron coefficient attenuation of paraffin P2 was 0.0382 cm-1 and the gamma coefficient attenuation was 0.0535 cm-1.

Abstract:

ABSTRAK DESAIN KONSEPTUAL PERISAI RADIASI REAKTOR RRI-50. Salah satu parameter yang harus dipenuhi dalam mendesain reaktor nuklir adalah desain perisai radiasi yang dapat menjamin keamanan dan keselamatan radiasi bagi pekerja dan masyarakat sekitar. Pada penelitian ini dilakukan desain perisai radiasi RRI-50 dengan elemen bakar berjenis U9Mo-Al berkerapatan tinggi dengan tipe pelat sebanyak 21 buah dan berdimensi seperti elemen bakar RSG-GAS tapi panjang aktifnya 70 cm. Konfigurasi teras terdiri dari 16 elemen bakar dan 4 elemen kendali serta 5 posisi iradiasi sehingga membentuk matriks 5 x 5. Tujuan dari penelitian ini adalah mendesain perisai radiasi dan menentukan distribusi laju dosis di daerah kerja dan di lingkungan reaktor RRI-50. Tahapan awal penelitian adalah perhitungan kuat sumber dan inventori bahan radioaktif teras reaktor dengan mensimulasikan operasi 50 MW selama 20 hari tiap siklus menggunakan program ORGEN2.1. Berdasarkan kuat sumber tersebut dan model yang dibuat menggunakan program VisEd, maka dilakukan analisis penentuan parameter perisai radiasi secara iteratif menggunakan program MCNPX. Pada tahap akhir, dilakukan analisis distribusi laju dosis di seluruh ruang di dalam dan di luar gedung reaktor juga menggunakan program MCNPX. Hasil yang diperoleh menunjukkan bahwa ketinggian permukaan air adalah 1000 cm dan kombinasi 90 cm beton berat dan 60 cm beton biasa dapat digunakan sebagai perisai biologi. Desain perisai tersebut mereduksi laju dosis menjadi 0,05 µSv/jam di Balai Operasi sem*ntara di Balai Eksperimen dan di luar gedung reaktor menjadi 4,2 µSv/jam dan 0,03 µSv/jam pada saat reaktor beroperasi. Hasil penelitian juga menunjukkan pemasagan perisai tambahan setebal 280 cm berjarak 300 cm di depan tabung berkas neutron radial yang terbuka mereduksi laju dosis gamma dan neutron menjadi 3,3 µSv/jam dan 3,1x10-11 µSv/jam. Hasil-hasil penelitian ini menunjukkan bahwa desain perisai radiasi yang dibuat membuat reaktor RRI-50 menjadi aman dari bahaya radiasi bagi pekerja dan masyarakat sekitarnya. Kata kunci : Perisai radiasi, laju dosis, keselamatan radiasi, RRI-50. ABSTRACT RADIATION SHIELDING CONSEPTUAL DESIGN OF RRI-50 REACTOR. One of the parameters that must be met in the design of nuclear reactors is radiation shielding design to ensure the security and safety of workers and the surrounding community. This study has been conducted to design radiation shielding of RRI-50 with high density U9Mo-Al fuel elements that consist of 21 pieces of plate type fuel elements with dimension as same as RSG-GAS fuel elements but the active length is 70 cm. Core configurations consist of 16 fuel elements and 4 control elements and 5 irradiation positions to form a matrix of 5 x 5. The objective of this research is to design radiation shielding and determine the distribution of dose rates in the working area and the environment of RRI-50 reactor. The early stages of this research is to calculate source strength and inventory of radioactive materials within the reactor core with one operation pattern cycle of 50 MW for 20 days using ORGEN2.1 program. Based on core source strength and models that are created using the VisEd software, the analysis parameter of the shielding was determined iteratively using MCNPX program. In the final stage, an analysis of the dose rate distributions in the whole space inside and outside the reactor building was conducted also using MCNPX program. The results show that the height of the water surface is 1000 cm and the combination of heavy concrete thickness of 90 cm and ordinary concrete thickness of 60 cm can be used as an biological shield. This design can reduce the dose rate to 0.05 µSv/h in the Operations Room while in the Experiments Room and outside the reactor building to 4.2 µSv/h and 0.03 µSv/h during reactor operation. The results also suggest that the installation of additional radiation shield of 280 cm thickness within 300 cm in front of the open radial neutron beam tube can reduce gamma and neutron dose rate to 3.3 µSv/h and 3,1x10-11 µSv/h. The results of this study indicate that the radiation shield design is made to make reactor RRI-50 to be safe from radiation hazards to workers and surrounding communities. Keywords : Radiation shielding, dose rates, radiation safety, RRI-50.

Abstract:

The design of concrete for radiation shielding structures is principally based on the selection of materials of adequate elemental composition and mix proportioning to achieve the long-term durability in nuclear environment. Concrete elements may become radioactive through exposure to neutron radiation from the nuclear reactor. A selection of constituent materials of greatly reduced content of long-lived residual radioisotopes would reduce the volume of low-level waste during plant decommissioning. The objective of this investigation is an assessment of trace elements with a large activation cross section in concrete constituents and simultaneous evaluation of susceptibility of concrete to detrimental alkali-silica reaction. Two isotopes 60Co and 152Eu were chosen as the dominant long-lived residual radioisotopes and evaluated using neutron activation analysis. The influence of selected mineral aggregates on the expansion due to alkali-silica reaction was tested. The content of 60Co and 152Eu activated by neutron radiation in fine and coarse aggregates, as well as in four types of Portland cement, is presented and discussed in respect to the chemical composition and rock origin. Conflicting results were obtained for quartzite coarse aggregate and siliceous river sand that, despite a low content, 60Co and 152Eu exhibited a high susceptibility to alkali-silica reaction in Portland cement concrete. The obtained results facilitate a multicriteria selection of constituents for radiation-shielding concrete.

Abstract:

Existing situation in nuclear industry is characterized with simultaneous development of the following two processes: design and construction of new generation of nuclear installations and decommissioning of installations of older generations. Significant amounts of radioactive wastes generated during the decommissioning phase are determined both for the first and the second types of installations by the induced activity of neutron irradiated structural and shielding materials. Concentration of the so-called radioactivity-hazardous nuclides in primary building and construction materials is the most important characteristics determining the resulting levels of induced activity. Values of these concentrations for the same type of material extracted from different geological deposits may differ by one or two orders of magnitude. Information about concentrations of radiation-hazardous elements in radiation shielding materials is fragmented and, as a rule, unsuitable for practical application. The purpose of the present study was to substantiate the necessity of compiling and recording the data on the concentrations of radioactivity-hazardous nuclides for building and structural materials for nuclear installations during the phases of design, operation and decommissioning. Three types of shielding concrete compositions were selected for the investigation. Concentrations of radioactivity-hazardous nuclides were mainly obtained by neutron activation technique. Neutron transport calculations were performed in one-dimensional cylindrical geometry at the core mid-plane according to usual core-vessel-shielding model of modern VVER reactor unit including 2-m thick concrete shield. Both transport and activation calculations were undertaken using modules of SCALE system. The obtained results allow estimating neutron-induced activation levels in the material as the function of irradiation time, amounts and categories of radioactive waste and their evolution during the decay time from 1 to 100 years. It was established that neutron-induced activity of shielding concrete strongly depends on the actual concentrations of radioactivity-hazardous nuclides in the concrete including ‘trace’ concentrations (other factors being the same). It was also shown that failure to take such concentrations into account may lead to the underestimation of neutron-induced activation levels and amounts of radioactive wastes and their category. The obtained results confirmed the necessity of compiling and maintaining data records on the concentrations of radioactivity-hazardous nuclides for materials used in structural and shielding materials of nuclear installations. Proposals were formulated on the potential accumulation of information, composition and formatting of descriptors of chemical composition of shielding and structural materials of nuclear installations.

Abstract:

Concrete has been used as a shield against high-energy photons (gamma) and neutrons since the beginning of use of nuclear reaction in energy, medicine and research. State of knowledge in shielding concrete technology is that while in case of protection against gamma radiation an increase in density caused by change of aggregate type for heavy-weight one is usually an efficient solution, the protection against neutrons is more complex. It is due to the differences in interactions of neutrons with the matter, depending on their kinetic energy and cross-sections for different reactions of the component atoms of the cement paste and the aggregate. The paper presents the results of the project NGS-Concrete - New-Generation Shielding Concrete. The aim is to design the composition of concrete against ionizing radiation, achieved by the use of experiment based on multi-criteria optimization of materials supported by the Monte Carlo simulations. Better concrete is the one that absorbs more thermal neutrons and slows down more fast neutrons at the same time. In the paper both results of Monte Carlo simulations and experimental studies on ordinary and heavyweight concrete containing epoxy polymer additive are presented. Close values of thermal neutron attenuation coefficients proved good accordance between simulation and experiment. The final conclusion is that epoxy resin is an efficient additive for neutron shielding concretes improving its ability to protect mainly against low energy neutrons. In experimental measurement there has not been observed an improvement of fast neutron attenuation due to increase of hydrogen atom content introduced with epoxy.

Abstract:

Long-term exposure of concrete to nuclear reactor environments may enhance the ageing phenomena. An investigation concerning a possible deleterious alkali-silica reaction (ASR) in concrete containing high-density aggregates is presented in this paper. The scope of this investigation was limited to heavy aggregates that could be used for the construction of the first Polish nuclear power plant (NPP). Five different high-density aggregates were selected and tested: three barites, magnetite, and hematite. Mineralogical analysis was conducted using thin section microscopic observation in transmitted light. The accelerated mortar beam test and the long-time concrete prism test were applied to estimate the susceptibility of heavy aggregates to ASR. X-ray diffraction (XRD) and scanning electron microscopy (SEM) analyses were conducted on aggregates and mortars. The quartz size in aggregate grains was evaluated using image analysis. Application of the accelerated mortar beam method confirmed the observations of thin sections and XRD analysis of high-density aggregates. The microcrystalline quartz in hematite aggregate and cristobalite in one of barite aggregate triggered an ASR. The composition of ASR gel was confirmed by microscopic analysis. The long-term concrete test permitted the selection of innocuous high-density aggregates from among the other aggregates available, which showed practically no reactivity.

Abstract:

Gamma rays in nuclear reactors, arising either from fission or decay processes, significantly contribute to the heating and dose of the reactor components. Zero power research reactors offer the possibility to measure gamma rays in a purely neutronic environment, allowing for validation experiments of computed spectra, dose estimates, reactor noise and prompt to delayed gamma ratios. This data then contributes to models, code validation and photo atomic nuclear data evaluation. In order to contribute to aforementioned experimental data, gamma detection capabilities are being added to the CROCUS reactor facility. The CROCUS reactor is a two-zone, uranium-fueled light water moderated facility operated by the Laboratory for Reactor Physics and Systems Behaviour (LRS) at the Swiss Federal Institute of Technology Lausanne (EPFL). With a maximum power of 100W, it is a zero power reactor used for teaching and research, most recently for intrinsic and induced neutron noise studies. For future gamma detection applications in the CROCUS reactor, an array of four detectors - two large 5”x10” Bismuth Germanate (BGO) and two smaller Cerium Bromide (CeBr3) scintillators - was acquired. The BGO detectors are to be arbitrarily positioned in the core reflector and out of the vessel for measurements at arbitrary distances. The CeBr3 detectors on the other hand are small enough to be set in the guide tubes of the control rods for in-core measurements. We present a study of the neutron and gamma flux in the core and reflector using the MCNP 6.2 and Serpent 2 Monte Carlo codes for coupled neutron and photon transport criticality calculations. More specifically, we investigate and compare predicted spectra as well as reactivity worth of different envisioned experimental setups. We further predict pulse height spectra as well as doses to the crystals with and without cadmium shielding to estimate allowable reactor powers with respect to detector radiation hardness. The results serve as basis for calibration and aid in the design and regulatory approval of the experiments.

Abstract:

Seeking optimal material distribution in a nuclear system to maximize a response function of interest has been a subject of considerable interest in nuclear engineering. Examples are the optimal fuel distribution in a nuclear reactor core to achieve uniform burnup using minimum critical mass and the use of composite materials with an optimal mix of constituent elements in detection systems and radiation shielding. For such studies, variational methods have been found to be useful but, they have been used for standalone analyses often restricted to idealized models, while more elaborate design studies have required computationally expensive Monte Carlo simulations ill-suited to iterative schemes for optimization. Such an inherent disadvantage of Monte Carlo methods changed with the development of perturbation algorithms but, their efficiency is still dependent on the reference configuration for which a hit-and-trial approach is often used. In the first illustrative example, this paper explores the computational speedup for a bare cylindrical reactor core, achievable by using a variational result to enhance the computational efficiency of Monte Carlo design optimization simulation. In the second example, the effect of non-uniform material density in a fixed-source problem, applicable to optimal moderator and radiation shielding, is presented. While applications of this work are numerous, the objective of this paper is to present preliminary variational results as inputs to elaborate stochastic optimization by Monte Carlo simulation for large and realistic systems.

Abstract:

We developed a new version of the X-PMSP program for estimating mass stopping powers in charged particle radiotherapy, shielding of nuclear reactors and particle accelerators, and medical and energetic radioisotope production. Accordingly, we calculated the mass stopping powers of important medical, industrial, and chemical compounds and mixtures defined by International Commission on Radiation Units and Measurements using the X-PMSP program, to contribute to the existing literature and compared our results with those from the National Institute of Standards and Technology database. The application of this program, for particles in the energy range 10-50 MeV, will be especially important in the production of medical radioisotopes, as it reduces the maximum error rate to <5 % for proton, and to ~1.6 % for alpha particles. Furthermore, the maximum error rates in charged particle radiotherapy, at the energy range 1-250 MeV, are 8.2 % for protons, and 3.0 % for alpha particles.

Abstract:

Mathematical simulation of fast neutron reactors requires high-precision calculations of protection problems based on unstructured meshes. The paper considers and analyzes a parallel version of the ODETTA code (Belousov et al. 2019) with the use of the MPI (Message Passing Interface) library technology (Knyazeva et al. 2006). The code is designed for numerical simulation of neutronic processes in shielding compositions of fast neutron lead cooled reactor plants in normal operating modes, and can be used to calculate the radiation conditions of using structural components and equipment of nuclear power facilities which are assumed to be the sources of and/or exposed to ionizing radiation during their safety justification. The operation of the generated code is compared against the previous version. The MPI-based development of the ODETTA code’s algorithmic part is described. Peculiarities and specific features of the code parallelization are presented, the code modification is given, and respective algorithms are considered. The structure of the ODETTA code based on the MPI is described in brief. The results of using the ODETTA code’s serial and parallel versions in OS Linux (Kostromin 2012) for NRNU MEPhI’s HPC cluster are provided (Savchenko et al. 2020). A comparative analysis is presented for two code implementation options in terms of speed and accuracy of results when using two different clusters and different numbers of nodes for these. Peculiarities of cluster-based calculations are noted.

Abstract:

Sealing of nuclear fuel material inside the fuel element clad must create a leak-tight "safety barrier" to prevent the release of radioactive products to environment. High quality welds are mandatory to withstand harsh conditions (radiation, pressure, temperature, corrosion) making possible the safe operation of nuclear reactors. The joint design, material selection and welding technique must be combined by smartly balancing possible technological options to yield the best attainable quality for the intended purpose; these choices are discussed in the paper. For thin-walled clad to end-plug welding, heat flow pattern as determined by joint design and fitting accuracy proved to be crucial for the fusion boundary shape and moreover for the success rate in automate welding. Consequently, Finite Element Analysis of the transient thermal field during welding was performed, in order to determine the best compromise with reasonable machining precision for parts. The main features of the developed thermal model and some results illustrating its good predictions vs. actual welds are also presented. Helium-shielded pulsed welding was initially preferred to minimize HAZ, distortion and porosity but unfortunately important cast-to-cast variation in penetration was observed with Inconel-600 plugs, due to Marangoni effect. Extensive work was done to overcome this, mainly through variation of pulsing and of the shielding gas; depth-to-width ratio can be noticeably improved with no material addition. Out of welding classic cladding materials, studies were initiated at INR on joining oxide-dispersion strengthened (ODS) alloys and specialty austenitic formulations (e.g. 15/15Ti) since they are candidate materials of great interest for the next generation of nuclear reactors.

Abstract:

This contribution is aimed at designing the optimal thickness of lead-iron double-layer container to store a radioactive waste releasing the photon energy at 1.3325 MeV and initial radiation intensity at 100 mSv/hr using the optimization design by MATLAB software. This design consisted of three parts of calculations to achieve 1000 times the radiation attenuation of container. The first was the logarithmic interpolation for the mass attenuation coefficient. The second was the bilogarithmic interpolation for the exposure buildup factor. The third was the contour-plotting analytical technique for the optimal thickness of radiation container. The values of mass attenuation coefficient and exposure buildup factor were exactly validated as compared with the standard reference database. Furthermore, we have found that the optimal thickness was 3.2 cm for lead (1st layer) and 17.0 cm for iron (2nd layer). Container weight was 994.30 kg, whilst container cost was 167.30 USD. The benefit of our design can quickly and precisely apply for the radiation safety assessment of the occupational radiation workers who always work in the nuclear reactor area.

Abstract:

The radiation parameters such as radionuclide content and activities, fluxes and energy spectrum of gamma and neutrons of spent nuclear fuel are essential when planning further spent fuel management options – interim wet or dry storage or disposal into a geological repository. Radiation parameters determine the design of a storage or disposal facility, what materials, structures and thicknesses of structures should be used to provide adequate biological shielding. Experimental measurements of spent fuel radiation parameters are rather complicated and expensive, therefore numerical methods are widely used. Various computer codes (APOLLO, BOXER, CASMO, FISPACT, ORIGEN-S, WIMS, etc.) have been developed to simulate the irradiation processes of nuclear fuel and to obtain resulting radiation parameters. Irrespective of the used computer code, the input data firstly must be entered into that code. When simulating nuclear fuel irradiation and burn-up in the reactor core, the geometrical parameters of the fuel assembly, materials’ data (chemical compositions, densities), the operating parameters of the reactor (power, operation time, coolant parameters, etc.) shall be entered into the program as initial data. Fairly often approximations of the input data are performed, for example, fuel rods in a fuel assembly are hom*ogenized and geometrically described as a solid cylinder, the reactor operation time is assumed as continuous and at constant power. The particularity of the input data and accepted assumptions depend on what initial information is available and on the capabilities of the computer code. The modelled spent fuel radiation parameters depend not only on the input data and assumptions, but also on the cross-section databases that are used in computer codes. Computer codes TRITON, ORIGEN-S and FISPACT have been used to model the concentration of actinides and fission products in the spent fuel from the RBMK-1000 reactor. The obtained results are compared and possible reasons for the differences in the modelling results are discussed.

Abstract:

Bushehr Nuclear power plant (BNPP) is currently under construction. The VVER-1000 reactor will be loaded with 126 tons of about 4% enriched fuel having 3-years life cycle. The spent fuel (SF) will be transferred into the spent fuel pool (SPF), where it stays for 8 years before being transferred to Russia. The SPF plays a crucial role during 8 years when the SP resides in there. This paper investigates the shielding of this structure as it is designed to shield the SF radiation. In this study, the SF isotope inventory, for different cycles and with different burnups, was calculated using WIMS/4D transport code. Using MCNP4C nuclear code, the intensity of γ rays was obtained in different layers of SFP shields. These layers include the water above fuel assemblies (FA) in pool, concrete wall of the pool and water laid above transferring fuels. Results show that γ rays leakage from the shield in the mentioned layers are in agreement with the plant's PSAR data. Finally we analyzed an accident were the water height above the FA in the pool drops to 47 cm. In this case it was observed that exposure dose above pool, 10 and 30 days from the accident, are still high and in the levels of 1000 and 758 R/hr.

Abstract:

Decommissioning of the Ignalina Nuclear Power Plant involves multiple problems. One of them is personnel radiation safety during the performance of dismantling activities. In this paper, modeling results of radiation doses during the dismantling of the pressurized tank from the emergency core cooling system (ECCS PT) of RBMK-1500 reactor are presented. The radiological surveys indicate that the inner surface of the ECCS PT is contaminated with radioactive products of corrosion and sediments due to the radioactive water. The effective doses to the workers have been modeled for different strategies of ECCS PT dismantling. In order to select the optimal personnel radiation safety, the modeling has been performed by the means of computer code “VISIPLAN 3D ALARA Planning tool” developed by SCK CEN (Belgium). The impacts of dismantling tools, shielding types, and extract ventilation flow rate on effective doses during the dismantling of ECCS PT have been analyzed. The total effective personnel doses have been obtained by summarizing the effective personnel doses from various sources of exposure, that is, direct radiation from radioactive equipment, internal radiation due to inhalation of radioactive aerosols, and direct radiation from radioactive aerosols arising during hot cutting in premises. The uncertainty of the collective doses is also presented in this paper.

Abstract:

ARES is a multidimensional parallel discrete ordinates particle transport code with arbitrary order anisotropic scattering. It can be applied to a wide variety of radiation shielding calculations and reactor physics analysis. ARES uses state-of-the-art solution methods to obtain accurate solutions to the linear Boltzmann transport equation. A multigroup discretization is applied in energy. The code allows multiple spatial discretization schemes and solution methodologies. ARES currently provides diamond difference with or without linear-zero flux fixup, theta weighted, directional theta weighted, exponential directional weighted, and linear discontinuous finite element spatial differencing schemes. Discrete ordinates differencing in angle and spherical harmonics expansion of the scattering source are adopted. First collision source method is used to eliminate or mitigate the ray effects. Traditional source iteration and Krylov iterative method preconditioned with diffusion synthetic acceleration are applied to solve the linear system of equations. ARES uses the Koch-Baker-Alcouffe parallel sweep algorithm to obtain high parallel efficiency. Verification and validation for the ARES transport code system have been done by lots of benchmarks. In this paper, ARES solutions to the HBR-2 benchmark and C5G7 benchmarks are in excellent agreement with published results. Numerical results are presented which demonstrate the accuracy and efficiency of these methods.

Abstract:

TRIPOLI-4® is a Monte-Carlo particle-transport code developed at CEA-Saclay (France) that is employed in the domains of nuclear-reactor physics, criticality-safety, shielding/radiation protection and nuclear instrumentation. The goal of this paper is to report on current developments, validation and verification made in TRIPOLI-4 in the electron/positron/photon sector. The new capabilities and improvements concern refinements to the electron transport algorithm, the introduction of a charge-deposition score, the new thick-target bremsstrahlung option, the upgrade of the bremsstrahlung model and the improvement of electron angular straggling at low energy. The importance of each of the developments above is illustrated by comparisons with calculations performed with other codes and with experimental data.

Abstract:

Abstract The paper describes a method of pulse height spectrum measurement in a wide energy range. The LVR‑15 research reactor building was chosen to demonstrate this method. Pulse height spectra were measured on the third floor of the reactor building. Two types of scintillation detectors, NaI (Tl) and a plastic scintillator, were used. The detectors were placed for about 25 m from the reactor core, thus, separated from the primary circuit water in the reactor pool, biological shielding, building wall and other constructional materials. Spectra were measured in a wide energy range from 30 keV to 1000 MeV, in which signals were recorded from natural and man-made radionuclides, prompt gamma radiation and cosmic radiation. Experimental data were collected both while the reactor was in operation and while it was out of operation. This study confirms that differences in these spectra can be detected remotely over relatively large distances from the reactor core by adequately simple detection means.