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.

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 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.

scholarly journals Effects of different nuclear evaluation data on the RMC keff calculation

2020 ◽  
Vol 239 ◽  
pp. 19005
Author(s):  
Zhang Wenxin ◽  
Qiang shenglong ◽  
Yin qiang ◽  
Cui Xiantao
Keyword(s):  
Cross Section ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Neutron Cross Section ◽  
Nuclear Data ◽  
Section Data ◽  
Calculation Results ◽  
The Impact ◽  
Physical Calculation

Neutron cross section data is the basis of nuclear reactor physical calculation and has a decisive influence on the accuracy of calculation results. AFA3Gassemble is widely used in nuclear power plants. CENACE is an ACE format multiple-temperature continuous energy cross section library that developed by China Nuclear Data Centre. In this paper, we calculated the AFA3G assemble by RMC.We respectively used ENDF6.8/, ENDF/7 and CENACE data for calculation. The impact of nuclear data on RMC calculation is studied by comparing the results of different nuclear data.

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 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.

Development of advanced, non-toxic, synthetic radiation shielding aggregate

2018 ◽  
Vol 106 (1) ◽  
pp. 59-68 ◽  
Author(s):  
Manish Mudgal ◽  
Ramesh Kumar Chouhan ◽  
Sarika Verma ◽  
Sudhir Sitaram Amritphale ◽  
Satyabrata Das ◽  
...  
Keyword(s):  
Red Mud ◽  
Nuclear Power ◽  
Power Plants ◽  
Xrd Analysis ◽  
Radiation Shielding ◽  
Industry Waste ◽  
Hematite Ore ◽  
First Time ◽  
Different Sources ◽  
Aggregate Impact Value

AbstractFor the first time in the world, the capability of red mud waste has been explored for the development of advanced synthetic radiation shielding aggregate. Red mud, an aluminium industry waste consists of multi component, multi elemental characteristics. In this study, red mud from two different sources have been utilized. Chemical formulation and mineralogical designing of the red mud has been done by ceramic processing using appropriate reducing agent and additives. The chemical analysis, SEM microphotographs and XRD analysis confirms the presence of multi-component, multi shielding and multi-layered phases in both the different developed advance synthetic radiation shielding aggregate. The mechanical properties, namely aggregate impact value, aggregate crushing value and aggregate abrasion value have also been evaluated and was compared with hematite ore aggregate and found to be an excellent material useful for making advanced radiation shielding concrete for the construction of nuclear power plants and other radiation installations.

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.

FCC-NG: The AREVA Safe Fresh Fuel Shipping Cask to Deliver the Future EPR™ Reactors

2010 ◽  
Author(s):  
M. Doucet ◽  
P. Faye ◽  
Ch. Faignet ◽  
R. Babut ◽  
M. Landrieu ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Nuclear Power ◽  
Power Plants ◽  
Neutron Absorber ◽  
Balsa Wood ◽  
Array Configuration ◽  
Safety Margins ◽  
Safety Case ◽  
Calculation Results ◽  
Set Up

AREVA as a worldwide PWR fuel provider has to have a fleet of fresh UO2 shipping casks being agreed within a lot of countries and to accommodate foreseen EPR™ Nuclear Power Plants fuel buildings. To reach this target the AREVA NP Fuel Sector together with TN International (a subsidiary of AREVA NC) decided to develop an up to date shipping cask gathering experience feedback of the today fleet and an improved safety allowing the design to comply with international regulations (NRC and IAEA) and other local foreign Safety Authorities requirements. Based on pre design features a safety case was set up to highlight safety margins. Criticality hypothetical accidental assumptions were defined: • Preferential flooding, • Fuel rod array expansion (so called “bird caging” effect), • Fuel sliding, • Neutron absorber penalty, • …. The French criticality code package CRISTAL is used to check several configurations reactivity and derived safety margins. The CRISTAL code package relies on a wide validation process helping defining uncertainties as required by regulations in force. The shipping cask, containing two fuel assemblies, is designed to maximize fuel isolation inside the cask and with neighboring ones even for large array configuration cases. Few and proven industrial products are used: • Stainless steel for the structural frame, • Balsa wood for impact limiters, • BORA® resin as neutron absorber. The cask is designed to handle mainly the EPR™ fuel assembly type and may be extended to other contents such as APWR fuel assembly type. After a brief presentation of the computer codes and the description of the shipping cask, the CRISTAL calculation results as well as the applied uncertainties will be discussed.

Development of an Efficient Tool for Evaluating Dose at Through-Holes

2020 ◽  
Author(s):  
Tatsuya Watanabe ◽  
Hironobu Iwanami ◽  
Tomoharu Hashimoto ◽  
Ryuichi Tayama
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gamma Ray ◽  
Integration Method ◽  
Computation Time ◽  
Simple Calculation ◽  
Detailed Calculation ◽  
Calculation Formula ◽  
Evaluation Point ◽  
Kernel Integration

Abstract In the design of nuclear power plants, it is demanded to quickly and calculate gamma ray scattering line (streaming) from the penetrating portion provided in the shielding such as electrical cables and ducts. However, when conducting gamma-ray streaming calculations from multiple penetrations, MCNP, a detailed calculation code, requires a long calculation time. This is due to the nature of MCNP, where many particles must reach the evaluation point when calculating in order for the results to be within an acceptable accuracy. To shorten the computation time, an analysis code utilizing a simple calculation method is necessary. Thus, we have developed a new method and a simple calculation tool (SVD-Dorc) for streaming computation. This method combines dose rate at an evaluation point with point kernel integration method and a simple streaming calculation formula for straight cylindrical ducts. Properties of SVD-Dorc are as follows: • Point kernel integration method • Simple streaming calculation formula for straight cylindrical ducts • Manual and automatic meshing of rectangular and cylindrical sources • Differentiation between direct line and non-direct sources • 3D drawing of input data • File output The validity of SVD-Dorc was confirmed by comparison with MCNP calculations and measured values from benchmark tests [2].

scholarly journals Method for Prototyping and Testing the Information Minimalism Concept of Nuclear Power Plants

2020 ◽  
Vol 225 ◽  
pp. 03004
Author(s):  
Gwi-sook Jang ◽  
Gee-yong Park
Keyword(s):  
Information Structure ◽  
Nuclear Power ◽  
Power Plants ◽  
Operation Mode ◽  
Careful Analysis ◽  
Test Methods ◽  
Application Range ◽  
New Concepts ◽  
Operational Test ◽  
And Control

he information structure and visualization design in nuclear power plant is based on careful analysis and understanding of the work domain of the operator. Piping and instrumentation diagram (P&ID) based layouts tend to require more display space. Overuse of mimic layouts can result in visual clutter. P&ID based layouts of controls are usually less easily operated configurations than those provided by other array conventions. Thus, a display method is emerging to minimize P&ID based display. The information minimalism concept is a monitoring and controller display method for each operation mode based on log analyses of operator actions. This method provides the monitoring information and control means necessary for the operator, according to operation mode, to perform a specific operation. This method can reduce the time spent searching for information and the transition between display pages. The aim of this paper was to verify the feasibility of implementing new concepts by establishing an information minimalism prototype. The validity of the function, performance, and operational test methods were verified by testing one such information minimalism prototype. In this paper, we describe prototyping and testing methods for the information minimalism concept for NPPs. In the future, this concept is to be added to the concept of operator support with the existing display configuration and navigation, and thereby extend its application range while actually utilizing it.

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