scholarly journals Monte Carlo Computational Software and Methods in Radiation Dosimetry

2021 ◽  
Vol 5 (3) ◽  
pp. 36-51
Author(s):  
Nikolaos Chatzisavvas ◽  
Georgios Priniotakis ◽  
Michael Papoutsidakis ◽  
Dimitrios Nikolopoulos ◽  
Ioannis Valais ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Nuclear Power ◽  
Radiation Dosimetry ◽  
Power Plants ◽  
Radiation Treatment ◽  
Space Station ◽  
Radiation Shielding ◽  
X Ray ◽  
Nuclear Medical Imaging ◽  
Treatment Planning Systems

The fast developments and ongoing demands in radiation dosimetry have piqued the attention of many software developers and physicists to create powerful tools to make their experiments more exact, less expensive, more focused, and with a wider range of possibilities. Many software toolkits, packages, and programs have been produced in recent years, with the majority of them available as open source, open access, or closed source. This study is mostly focused to present what are the Monte Carlo software developed over the years, their implementation in radiation treatment, radiation dosimetry, nuclear detector design for diagnostic imaging, radiation shielding design and radiation protection. Ten software toolkits are introduced, a table with main characteristics and information is presented in order to make someone entering the field of computational Physics with Monte Carlo, make a decision of which software to use for their experimental needs. The possibilities that this software can provide us with allow us to design anything from an X-Ray Tube to whole LINAC costly systems with readily changeable features. From basic x-ray and pair detectors to whole PET, SPECT, CT systems which can be evaluated, validated and configured in order to test new ideas. Calculating doses in patients allows us to quickly acquire, from dosimetry estimates with various sources and isotopes, in various materials, to actual radiation therapies such as Brachytherapy and Proton therapy. We can also manage and simulate Treatment Planning Systems with a variety of characteristics and develop a highly exact approach that actual patients will find useful and enlightening. Shielding is an important feature not only to protect people from radiation in places like nuclear power plants, nuclear medical imaging, and CT and X-Ray examination rooms, but also to prepare and safeguard humanity for interstellar travel and space station missions. This research looks at the computational software that has been available in many applications up to now, with an emphasis on Radiation Dosimetry and its relevance in today's environment.

scholarly journals Mathematical Formulation and Analytic Solutions for Uncertainty Analysis in Probabilistic Safety Assessment of Nuclear Power Plants

Energies ◽  
2021 ◽  
Vol 14 (4) ◽  
pp. 929
Author(s):  
Gyun Seob Song ◽  
Man Cheol Kim
Keyword(s):  
Monte Carlo ◽  
Uncertainty Analysis ◽  
Probability Density ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Mathematical Formulation ◽  
Analytic Solutions ◽  
Probabilistic Safety Assessment ◽  
Probabilistic Safety

Monte Carlo simulations are widely used for uncertainty analysis in the probabilistic safety assessment of nuclear power plants. Despite many advantages, such as its general applicability, a Monte Carlo simulation has inherent limitations as a simulation-based approach. This study provides a mathematical formulation and analytic solutions for the uncertainty analysis in a probabilistic safety assessment (PSA). Starting from the definitions of variables, mathematical equations are derived for synthesizing probability density functions for logical AND, logical OR, and logical OR with rare event approximation of two independent events. The equations can be applied consecutively when there exist more than two events. For fail-to-run failures, the probability density function for the unavailability has the same probability distribution as the probability density function (PDF) for the failure rate under specified conditions. The effectiveness of the analytic solutions is demonstrated by applying them to an example system. The resultant probability density functions are in good agreement with the Monte Carlo simulation results, which are in fact approximations for those from the analytic solutions, with errors less than 12.6%. Important theoretical aspects are examined with the analytic solutions such as the validity of the use of a right-unbounded distribution to describe the uncertainty in the unavailability/probability. The analytic solutions for uncertainty analysis can serve as a basis for all other methods, providing deeper insights into uncertainty analyses in probabilistic safety assessment.

Research Progress on Radiation Shielding Concrete

2012 ◽  
Vol 253-255 ◽  
pp. 303-307 ◽  
Author(s):  
Jing Yang ◽  
Zhen Fu Chen ◽  
Yuan Chu Gan ◽  
Qiu Wang Tao
Keyword(s):  
Nuclear Waste ◽  
Material Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Radiation Shielding ◽  
Nuclear Industry ◽  
Research Progress ◽  
Concrete Durability ◽  
Technology Research ◽  
Shielding Material

Radiation shielding concrete is widely used in nuclear power plants, accelerators, hospitals, etc. With the development of nuclear industry technology, research on radiation shielding material properties is of great importance. Research on properties of radiation shielding concrete with different aggregates or admixtures and the effect of high temperature on the performance of shielding concrete are introduced. Along with the nuclear waste increase, shielding concrete durability and nuclear waste disposal are getting paramount.

scholarly journals Estimation of the X/γ- ray shielding properties of different glass system using Monte Carlo Simulation

2021 ◽  
Vol 12 (2) ◽  
pp. 1117-1122
Author(s):  
Ayano Shanko, MD, Et. al.
Keyword(s):  
Monte Carlo Simulation ◽  
Monte Carlo ◽  
Attenuation Coefficient ◽  
Mass Attenuation Coefficient ◽  
Radiation Shielding ◽  
Glass System ◽  
Effective Electron ◽  
X Ray ◽  
Shielding Properties ◽  
Mass Attenuation

The aim of the research is to estimate the X-ray shielding properties of different glass systems using Monte Carlo Simulation. X-ray glass is also known as radiation shielding glass. Glass provides protection against the absorption of energy radiation. The shielding layer is formed by a high concentration of lead and barium. The mass attenuation coefficient, the effective atomic number and the effective electron density are used to determine the position of gamma-ray photons in matter. Shield characterization in terms of mass attenuation coefficient (μm), transmission fraction (T), effective atomic numbers (Zeff), half-value layer (HVL) and exposure build-up. factor (EBF) of a glass system is estimated by the Monte Carlo Simulation. The random sampling and statistical analysis are computed using the monte carlo simulation. Various external factors are considered as the input parameters. The different composition of the glass will be examined using the Monte Carlo simulation and the shielding capability would be determined for the various samples.

scholarly journals Radiation shielding structures : Concepts, behaviour and the role of the heavy weight concrete as a shielding material - Rewiev

2020 ◽  
Vol 21 ◽  
pp. 24-30
Author(s):  
Suha Ismail Ahmed Ali ◽  
Éva Lublóy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Building Materials ◽  
Physical And Mechanical Properties ◽  
Cost Effective ◽  
Radiation Shielding ◽  
Radiation Emission ◽  
Neutron Shielding ◽  
Heavy Weight ◽  
Industrial Buildings

The construction of radiation shielding buildings still developed. Application of ionizing radiations became necessary for different reasons, like electricity generation, industry, medical (therapy treatment), agriculture, and scientific research. Different countries all over the world moving toward energy saving, besides growing the demand for using radiation in several aspects. Nuclear power plants, healthcare buildings, industrial buildings, and aerospace are the main neutrons and gamma shielding buildings. Special design and building materials are required to enhance safety and reduce the risk of radiation emission. Radiation shielding, strength, fire resistance, and durability are the most important properties, cost-effective and environmentally friendly are coming next. Heavy-weight concrete (HWC) is used widely in neutron shielding materials due to its cost-effectiveness and worthy physical and mechanical properties. This paper aims to give an overview of nuclear buildings, their application, and behaviour under different radiations. Also to review the heavy-weight concrete and heavy aggregate and their important role in developing the neutrons shielding materials. Conclusions showed there are still some gaps in improving the heavy-weight concrete (HWC) properties.

Operational Experience of TN 24 Dual Purpose Spent Fuel Storage Casks at the DOEL Nuclear Power Plants in Belgium

2001 ◽  
Author(s):  
C. Baroux ◽  
M. Detrilleaux ◽  
G. Demazy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
High Capacity ◽  
Radiation Shielding ◽  
Ring Closure ◽  
Operational Experience ◽  
Spent Fuel ◽  
Dual Purpose ◽  
New Members

Abstract Spent nuclear fuel has been stored at the DOEL power station in Belgium in dual-purpose metal casks since 1995. The casks were procured from TRANSNUCLEAIRE by SYNATOM to meet the operational demands for on-site dry storage solutions for fuel arising from the four PWR reactors at DOEL. The TN 24 type of cask was chosen and a range of different cask types were developed. The initial requirement was for dual purpose cask to contain fuel from the DOEL units 3 and 4, these having similar fuel types but different lengths, and thus two new members of the TN 24 family were developed; the TN 24 D and TN 24 XL with capacities of 28 and 24 SFA’s. These casks were licensed as B(U) fissile packagings with approval certificates granted by the French and validated by the Belgium competent authorities for the transport configurations. Both cask designs were also analyzed by TRANSNUCLEAIRE in their storage configurations to ensure that the criteria for safe interim storage could be met. Since 1995, a total of 18 TN 24 D and TN 24 XL casks have been loaded with spent fuel assemblies with an average burn-up of 40,000 MWd/tU. SYNATOM subsequently decided to purchase further casks for DOEL 3 and 4 fuels with higher enrichments, higher burn-ups and shorter cooling times. TRANSNUCLEAIRE developed the TN 24 DH and TN 24 XLH casks within the similar envelope size and weight limits. The increase in performance was achieved by an in-depth optimization of each design in terms of radiation shielding, heat transfer and criticality safety. This paper shows how this optimization process was undertaken for the TN 24 DH and TN 24 XLH casks, 16 of which have been ordered by SYNATOM. DOEL 1 and 2 units use much shorter PWR fuel and it was decided to ship the fuel to unit 3 with an internal transfer cask because the handling limitations in the DOEL 1 and 2 pool prohibited the loading of a high capacity dual purpose transport/storage cask. The TN 24 SH cask was subsequently designed for DOEL 1 and 2 PWR fuel with a capacity of 37 assemblies and nine of there casks have been ordered by SYNATOM. The casks are fitted with monitoring devices to detect any change in the performance of the double metal O ring closure system and none of the casks has shown any deterioration in leaktightness. This paper examines the operation experience of loading and storing more than 30 TN 24 dual purpose casks and compares the performance with design expectations.

scholarly journals On Capabilities and Limitations of Current Fast Neutron Flux Monitoring Instrumentation for the Demo LFR ALFRED

2016 ◽  
Vol 2 (4) ◽  
Author(s):  
Luigi Lepore ◽  
Romolo Remetti ◽  
Mauro Cappelli
Keyword(s):  
Monte Carlo ◽  
Neutron Flux ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Power Plants ◽  
Fuel Cycle ◽  
Signal To Noise Ratio ◽  
Reactor Power ◽  
Fast Reactors ◽  
Flux Monitoring

Among GEN IV projects for future nuclear power plants, lead-cooled fast reactors (LFRs) seem to be a very interesting solution due to their benefits in terms of fuel cycle, coolant safety, and waste management. The novelty of this matter causes some open issues about coolant chemical aspects, structural aspects, monitoring instrumentation, etc. Particularly, hard neutron flux spectra would make traditional neutron instrumentation unfit to all reactor conditions, i.e., source, intermediate, and power range. Identification of new models of nuclear instrumentation specialized for LFR neutron flux monitoring asks for an accurate evaluation of the environment the sensor will work in. In this study, thermal hydraulics and chemical conditions for the LFR core environment will be assumed, as the neutron flux will be studied extensively by the Monte Carlo transport code MCNPX (Monte Carlo N-Particles X-version). The core coolant’s high temperature drastically reduces the candidate instrumentation because only some kinds of fission chambers and self-powered neutron detectors can be operated in such an environment. This work aims at evaluating the capabilities of the available instrumentation (usually designed and tailored for sodium-cooled fast reactors) when exposed to the neutron spectrum derived from the Advanced Lead Fast Reactor European Demonstrator, a pool-type LFR project to demonstrate the feasibility of this technology into the European framework. This paper shows that such a class of instrumentation does follow the power evolution, but is not completely suitable to detect the whole range of reactor power, due to excessive burnup, damages, or gamma interferences. Some improvements are possible to increase the signal-to-noise ratio by optimizing each instrument in the range of reactor power, so to get the best solution. The design of some new detectors is proposed here together with a possible approach for prototyping and testing them by a fast reactor.

Optimization of the Radiation Shielding Program QAD

2016 ◽  
Author(s):  
Wenyi Wang ◽  
Liguo Zhang ◽  
Jianzhu Cao ◽  
Feng Xie
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Discrete Source ◽  
Radiation Shielding ◽  
Application Range ◽  
Calculation Results ◽  
Γ Rays ◽  
Kernel Integration ◽  
Shielding Design ◽  
Original Edition

The QAD program, based on the point kernel integration method, is widely used in the radiation shielding design of nuclear power plants and related fields. However, QAD-CGA, as the latest version of QAD program, still has some problems, which may affect calculation results and limit the application range. In this paper, the features, principles, and algorithms of QAD-CGA program will be described and several optimization will be introduced. The quantity of γ rays considered in each calculation has been expanded, which can supply more accurate results than those from the original program. Furthermore, the number of dose receivers has been increased, which can provide detailed distribution of the dose field. In addition, a method has been put forward to realize the discretization of source intensity automatically which can simplify the input of the program. Meanwhile, the compartmentalization of the discrete source in the program has been improved. If the size of the discrete source can be minimized small enough to be served as an ideological core, the accuracy of calculations of QAD-CGA program would be guaranteed. However, with the increase of the radius of a sphere or cylinder, the volume of the discrete source will be enlarged and the precondition “small enough” will be lost gradually which can result in the increase of the inaccuracy of calculations. A superior algorithm to solve the coordinate distribution of point kernel which is nonuniform has been proposed. It can reduce the inaccuracy from the discretization of the source intensity in spherical and cylindrical geometry effectively. The optimization of QAD-CGA program has been implemented, analyzed and compared to the original edition with a numerical example.

Application of the Monte Carlo Method for Capacitive Pressure Transmitters Surveillance in Nuclear Power Plants

2012 ◽  
Vol 134 (3) ◽  
Author(s):  
C. Montalvo ◽  
A. García-Berrocal ◽  
J. Blázquez ◽  
M. Balbás
Keyword(s):  
Monte Carlo ◽  
Time Delay ◽  
Monte Carlo Method ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Noise Analysis ◽  
Ar Model ◽  
Akaike Criterion ◽  
New Criterion

In nuclear power plants (NPPs), according to current regulations, the response time of capacitive pressure transmitters is used as an index for surveillance. Such measurement can be carried out in situ applying the noise analysis techniques to the sensor output signal. The method is well established, and it is based on the autoregressive (AR) fitting optimized by the Akaike criterion (AIC). The sensor response is influenced by the sensing line, and its length is different in each plant. Recent empirical research has proved that the sensor inner structure can be modeled with a two real poles transfer function. In the present work, it has been proved that the noise analysis applied to the simulated response of a transmitter, modeled with two poles coupled with a sensing line, gives erroneous values for the ramp time delay when the sensing line is long. Specifically, the order of the AR model supplied by the Akaike criterion is not appropriate. Therefore, a Monte Carlo method is proposed to be applied in order to establish a new criterion, based on the statistical analysis of the repeatability of the ramp time delay obtained with the AR model.

scholarly journals Review on the Change Law of the Properties and Performance of Polymer-Matrix Nuclear Radiation Shielding Materials Under the Coupling of Nuclear Radiation and Thermal Effects

2021 ◽  
Vol 9 ◽  
Author(s):  
Guang Hu ◽  
Weiqiang Sun ◽  
Yihong Yan ◽  
Rongjun Wu ◽  
Hu Xu
Keyword(s):  
Mechanical Properties ◽  
Polymer Matrix ◽  
Thermal Effects ◽  
Nuclear Power ◽  
Power Plants ◽  
Radiation Shielding ◽  
Nuclear Radiation ◽  
And Performance ◽  
Shielding Material

The polymer-matrix nuclear radiation shielding material is an important component of nuclear power plants. However, its mechanical properties and shielding performance gradually deteriorate due to the long-term synergy of nuclear radiation and thermal effects, which brings hidden dangers to the safe operation of the device. Based on this problem, this article makes a comprehensive review. First, the degradation of mechanical properties and shielding performance of polymer-matrix nuclear radiation materials in service is briefly described. Then, the research methods adopted by scholars to study the change law of properties and performance are introduced, and the main existing difficulties encountered by the study are summarized. Finally, the physical mechanism of the change of material properties is explained in detail, and a reference approach to solving the problem is proposed.

Application of Monte Carlo Methods in Reactor Protection System Reliability Research

2018 ◽  
Author(s):  
Duo Li ◽  
Zhaojun Hao ◽  
Shuqiao Zhou ◽  
Chao Guo
Keyword(s):  
Monte Carlo ◽  
Reliability Analysis ◽  
Monte Carlo Methods ◽  
Nuclear Power ◽  
Power Plants ◽  
Fault Tree Analysis ◽  
Engineering Application ◽  
Protection System ◽  
System States ◽  
Reactor Protection System

Digital Reactor Protection System (RPS) is one of the most important systems in instrumentation and control systems of Nuclear Power Plants (NPP). The reliability analysis of RPS plays an important role both in theory and engineering application. Traditional reliability methods, such as fault tree analysis and Markov chain theory, have many limitations in the research of RPS reliability, since the number of system states increases exponentially with the growth of system complexity. Aiming at the reliability analysis of complex system like RPS, the Monte Carlo method simulates the system behaviors and obtains the reliability calculations through a large number of simulations. This paper takes a preliminary research of RPS reliability based on Monte Carlo Methods, including static reliability analysis based on Monte Carlo simulation of the behavior of every equipment in the RPS, and dynamic characters of the RPS based on the simulation of RPS period tests.

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