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].

Design, calibration, and application of an airborne gamma spectrometer system in Switzerland

Geophysics ◽  
1997 ◽  
Vol 62 (5) ◽  
pp. 1369-1378 ◽  
Author(s):  
Georg F. Schwarz ◽  
Ladislaus Rybach ◽  
Emile E. Klingelé
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Gamma Ray ◽  
Point Sources ◽  
Swiss Alps ◽  
Artificial Radioactivity ◽  
Radiation Level ◽  
Satellite Navigation System ◽  
Emergency Situations

Airborne radiometric surveys are finding increasingly wider applications in environmental mapping and monitoring. They are the most efficient tool to delimit surface contamination and to locate lost radioactive sources. To secure radiometric capability in survey and emergency situations, a new sensitive airborne system has been built that includes an airborne spectrometer with 256 channels and a sodium iodide detector with a total volume of 16.8 liters. A rack mounted PC with memory cards is used for data acquisition, with a GPS satellite navigation system for positioning. The system was calibrated with point sources using a mathematical correction to take into account the effects of gamma‐ray scattering in the ground and in the atmosphere. The calibration was complemented by high precision ground gamma spectrometry and laboratory measurements on rock samples. In Switzerland, two major research programs make use of the capabilities of airborne radiometric measurements. The first one concerns nuclear power plant monitoring. The five Swiss nuclear installations (four power plants and one research facility) and the surrounding regions of each site are surveyed annually. The project goal is to monitor the dose‐rate distribution and to provide a documented baseline database. The measurements show that all sites (with the exception of the Gösgen power plant) can be identified clearly on the maps. No artificial radioactivity that could not be explained by the Chernobyl release or earlier nuclear weapons tests was detected outside of the fenced sites of the nuclear installations. The second program aims at a better evaluation of the natural radiation level in Switzerland. The survey focused on the crystalline rocks of the Central Massifs of the Swiss Alps because of their relatively high natural radioactivity and lithological variability.

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.

scholarly journals Estimation of Tritium Production in VVER-440 Reactor Core During Normal Operation

2019 ◽  
Vol 9 (2) ◽  
pp. 10-16
Author(s):  
Štefan Čerba ◽  
Jakub Lüley ◽  
Branislav Vrban ◽  
Filip Osuský ◽  
Vladimír Nečas
Keyword(s):  
Thermal Neutron ◽  
Cross Section ◽  
Nuclear Power ◽  
Power Plants ◽  
Normal Operation ◽  
Boron Steel ◽  
Detailed Calculation ◽  
Primary Coolant ◽  
Tritium Production ◽  
Operation Conditions

Slovakia as one of the world leading countries in the share of nuclear power in electricityproduction and currently operates 2 nuclear power plants, each with 2 VVER-440 units. In addition to these reactors there are 2 VVER-440 units under construction and 2 units in decommissioning. The VVER-440 technology features thermal neutron spectrum, low enriched uranium dioxide fuel and light-water coolant, diluted boric acid and 37 emergency reactivity control assemblies with boron steel absorber. Due to the presence of 10B in the coolant/moderator which has high thermal neutron capture cross-section, the absorption of neutron on these atoms may lead to tritium production. Tritiumstrongly contributes to the level of radioactivity of the primary coolant, therefore the NPP staff must have appropriate knowledge of its production during operation. This paper focuses on the estimation of the tritium production for a specific scenario of the operation of the 3rd unit of Mochovce NPP. For simulations the SCALE6 system is used with the detailed calculation model developed at the B&J NUCLEAR ltd. company. The calculations presented in the paper are performed using self-shielded multi-group cross-section libraries, taking into account the operation conditions of Mochovce unit 3 NPP in the first fuel campaign.

scholarly journals The Characterization of Silicone-Tungsten-Based Composites as Flexible Gamma-Ray Shields

Materials ◽  
2021 ◽  
Vol 14 (20) ◽  
pp. 5970
Author(s):  
Jie Wang ◽  
Haoyu Zhou ◽  
Yong Gao ◽  
Yupeng Xie ◽  
Jing Zhang ◽  
...  
Keyword(s):  
High Temperature ◽  
Tensile Property ◽  
Nuclear Power ◽  
Power Plants ◽  
Degrees Of Freedom ◽  
Gamma Ray ◽  
Tensile Elongation ◽  
Nano Tio2 ◽  
Attenuation Coefficients ◽  
Structural Investigations

Robots are very essential for modern nuclear power plants to monitor equipment conditions and eliminate accidents, allowing one to reduce the radiations on personnel. As a novel robot, a soft robot with the advantages of more degrees of freedom and abilities of continuously bending and twisting has been proposed and developed for applications in nuclear power industry. Considering the radiation and high-temperature environment, the overall performance improvement of the flexible materials used in the soft nuclear robot, such as the tensile property and gamma-ray shielding property, is an important issue, which should be paid attention. Here, a flexible gamma-ray shielding material silicone-W-based composites were initially doped with nano titanium oxide and prepared, with the composition of 20 silicone-(80-x) W-(x) TiO2, where x varied from 0.1 to 2.0 wt.%. Structural investigations on SEM and EDS were performed to confirm the structure of the prepared composites and prove that all the chemicals were included in the compositions. Moreover, the tensile property of the composites at 25, 100, and 150 °C were investigated to study the effect of working temperature on the flexibility of the compositions. The attenuation characteristics including the linear attenuation coefficients and mass attenuation coefficients of the prepared silicone-W or silicone-W-TiO2-based composites with respect to gamma ray were investigated. The stability of the silicone–tungsten-TiO2-based composite at high temperature was studied for the first time. In addition, the influence of nano TiO2 additive on the property’s variation of silicone-W-based composites was initially studied. The comparison of the properties such as the tensile elongation, thermal stability, and gamma-ray shielding of the synthesized silicone-W and silicone-W-TiO2 composites showed that the addition of nano TiO2 powders could be useful to develop novel gamma-ray-shielding materials for radiation protection of soft robots or other applications for which soft gamma-ray-shielding materials are needed.

Development of Approaches to Simulate Fuel Rod Destruction With Different Fuel Type

2020 ◽  
Author(s):  
Eduard Usov ◽  
Nikolay Pribaturin ◽  
Vladimir Chukhno ◽  
Ilya Klimonov ◽  
Anton Butov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Liquid Metals ◽  
Numerical Algorithms ◽  
Severe Accident ◽  
Theoretical Research ◽  
Detailed Calculation ◽  
Fuel Rods ◽  
Fuel Rod ◽  
The Core

Abstract Due to the revival of interest to the development of fast reactors cooled by liquid metals, the problem of carrying out theoretical research in support of their safety is actual. A detailed calculation of all stages of the accident from the beginning to the end requires knowledge of the laws for modeling physical processes occurring in the reactor in an emergency. The most serious are accidents with the destruction of the core. Simulation of severe accident in nuclear reactor is the key element in safety analysis of nuclear power plants. Destruction of fuel rods is one of the most important processes that should be calculated during core degradation. For different type of fuels the mechanism of the degradation are different too. For example, oxide and metallic fuels usually melt congruently at high temperature, but nitride fuel dissociates. The main objective of the proposed research is developing of models and numerical algorithms for calculation fuel rods destruction with oxide, metallic and nitride fuels. The models of the destruction processes and some calculation results are presented in the paper. The processes are investigated for the first phase of severe accidents covering the period from the onset of fuel-rod melting to the melt escape from the core center.

Characterization Of A Lanthanum Bromide Detector For Eye Lens Dosimetry At The Candu Nuclear Power Plants Based On Direct Measurements Of The Gamma-Ray Spectra

2020 ◽  
Author(s):  
A S Laranjeiro ◽  
F Bohra ◽  
S H Byun ◽  
J Atanackovic ◽  
A R Hanu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Pulse Shaping ◽  
Nuclear Power Plants ◽  
Gamma Ray ◽  
Absorbed Dose ◽  
Eye Lens ◽  
High Dose ◽  
Candu Reactors ◽  
Radiation Fields

Abstract Gamma-ray spectra were measured using a LaBr$_{3}$(Ce) spectrometer during the outage periods, aiming at quantifying the gamma source term of radiation workers’ exposure, at the CANDU nuclear power reactors, for the purposes of eye lens dosimetry. The spectra were measured inside the boiler rooms, of the Bruce Power and Ontario Power Generation (OPG) CANDU nuclear power plants, where workers are exposed to relatively high dose rates radiation fields during the maintenance work. Prior to measurements at the CANDU reactors, the pulse shaping parameters of the gamma spectrometer were optimised for high rates gamma fields, up to an input rates of 120 kcps, in order to accomplish a high output rate with a reasonable energy resolution. In parallel, the response of the LaBr$_{3}$(Ce) detector was characterized by experiments and Monte Carlo simulations. The gamma spectra measured at the CANDU reactors were reported in terms of the gamma-ray fluence rate spectrum. In all measured data, $^{60}$Co and $^{95}$Nb were main contributors of the gamma fields. The measured spectra have been used to calculate the dosimetric quantities of interest: personal dose equivalents H$_{p}$(10) and H$_{p}$(0.07) and eye lens absorbed dose.

scholarly journals Hydrazine Radiolysis by Gamma-Ray in the N2H4–Cu+–HNO3 System

2021 ◽  
Vol 22 (14) ◽  
pp. 7376
Author(s):  
Naon Chang ◽  
Huijun Won ◽  
Sangyoon Park ◽  
Heechul Eun ◽  
Seonbyeong Kim ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gamma Ray ◽  
Uv Spectroscopy ◽  
Radiolysis Product ◽  
Decomposition Mechanism ◽  
Irradiation Condition ◽  
High Dose ◽  
Hno3 Solution ◽  
Γ Ray

Radiolysis of chemical agents occurs during the decontamination of nuclear power plants. The γ-ray irradiation tests of the N2H4–Cu+–HNO3 solution, a decontamination agent, were performed to investigate the effect of Cu+ ion and HNO3 on N2H4 decomposition using a Co-60 high-dose irradiator. After the irradiation, the residues of N2H4 decomposition were analyzed by Ultraviolet-visible (UV) spectroscopy. NH4+ ions generated from N2H4 radiolysis were analyzed by ion chromatography. Based on the results, the decomposition mechanism of N2H4 in the N2H4–Cu+–HNO3 solution under γ-ray irradiation condition was derived. Cu+ ions form Cu+N2H4 complexes with N2H4, and then N2H4 is decomposed into intermediates. H+ ions and H● radicals generated from the reaction between H+ ion and eaq− increased the N2H4 decomposition reaction. NO3− ions promoted the N2H4 decomposition by providing additional reaction paths: (1) the reaction between NO3− ions and N2H4●+, and (2) the reaction between NO● radical, which is the radiolysis product of NO3− ion, and N2H5+. Finally, the radiolytic decomposition mechanism of N2H4 obtained in the N2H4–Cu+–HNO3 was schematically suggested.

scholarly journals A Study on Steel-Concrete-Steel Wall to Resist Perforation from Rigid Projectile Impact

2021 ◽  
Vol 2021 ◽  
pp. 1-12
Author(s):  
Pengfei Han ◽  
Jingbo Liu ◽  
Bigang Fei ◽  
Fei Wang
Keyword(s):  
Yield Strength ◽  
Nuclear Power ◽  
Power Plants ◽  
Steel Plate ◽  
Steel Plates ◽  
Dissipated Energy ◽  
Calculation Formula ◽  
Rigid Projectile ◽  
Calculation Results ◽  
Energy Dissipated

A calculation method of SCS wall which is used in the third generation of nuclear power plants to resist perforation from rigid projectile based on energy method is proposed in this paper. The energy is divided into four parts including the energy dissipated by front steel plate, concrete, back steel plate, and tie bars. The method accounts for the perforation of the concrete and steel plates separately and accounts for the interaction between them, and a practical antiperforation calculation formula of SCS wall with tie bars is given. The most formular results are close to the test results and the FEM results with a deviation less than 10%, which shows that the calculation formula given in this paper is reasonable and credible to effectively evaluate the perforation failure of the SCS wall and carry out a relevant design. The energy dissipated by the steel plate is much larger than that of the tie bars through a comparative analysis of dissipated energy. The effects of various factors on perforation velocity are analyzed according to finite element calculation results, which can be roughly divided into three categories: the influence of the thickness of steel plate and distance of tie bar is the largest effect, followed by that of yield strength of steel plate, yield strength of tie bar and diameter of tie bar, and that of compressive strength of concrete is the smallest effect.

Industrial Application of Research Results of Packings on Safety Relevant Valves in Nuclear Power Plants

2006 ◽  
Author(s):  
Reinhard Koring
Keyword(s):  
Industrial Application ◽  
Nuclear Power ◽  
Power Plants ◽  
Design Calculation ◽  
Detailed Calculation ◽  
Research Results ◽  
Packing Rings ◽  
Leak Tightness ◽  
Specific Material ◽  
Packing Assembly

In a NPP approximately 380 safety relevant valves are installed. The main requirements are the safety and operational availability as well as a adequate leak tightness of the sealings and stem packings. Based upon the research results of the MPA-Stuttgart, Germany on packing features specific procedures for handling, assembly and maintenance of packings and analytical inputs for the design calculation have been developed. Depending on the valve size, especially on the stem diameter, the packing influences the necessary stem force for the closing and opening function due to the amount of friction force. The complying equations of the valve calculation guideline were derived from the research results and rely on covering material characteristics of the approved packing rings such as friction coefficient and the vertical/horizontal force transfer factor of the stressing force. The industrial application often requires a more detailed calculation and handling of the actually installed packing configurations especially if existing valves are considered for redesign measures and recalculation. Therefore additionally mock-up tests of existing packing configurations have been performed in cooperation with the packing manufacturer in order to get very specific material coefficients as input data for the calculations. This paper presents the application of research results to design calculation of safety relevant valves as well as the development of procedures for the packing assembly and maintenance.

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