The Zwilag Plasma Facility: Five Years of Successful Operation

2010 ◽  
Author(s):  
Walter Heep
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Pure Metal ◽  
Radioactive Wastes ◽  
Operating Conditions ◽  
Plasma Synthesis ◽  
Plasma Technology ◽  
Process Technology ◽  
Successful Operation ◽  
Low Level

This paper is about a treatment facility for low level radioactive wastes that operates with plasma technology. The first processing of low-level radioactive wastes from Swiss nuclear power plants marked the successful completion of the commissioning of this facility in March 2004. This process technology is derived from metallurgy. Its theoretical principles are based on plasma technology, which has been applied for quite some time outside the field of nuclear technology for the production of highly pure metal alloys and for the plasma synthesis of acetylene. The commercial operation of the plasma plant owned by Zwilag Zwischenlager Wu¨renlingen AG (Zwilag) has also enabled this technology to be used successfully for the first time in the nuclear field in a way that addresses the issue of radiation protection. In addition to a brief presentation of the technology used in the plant, the paper explains in detail how the process works under operating conditions. The plasma facility has been operating now for five years and was granted an unrestricted operating license in September 2009.

The ZWILAG Interim Storage Plasma Plant: Technology to Handle Operational Waste From Nuclear Plants

2007 ◽  
Author(s):  
Walter Heep
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Melting Process ◽  
Pure Metal ◽  
Radioactive Wastes ◽  
Operating Conditions ◽  
Plasma Synthesis ◽  
Plasma Technology ◽  
Process Technology ◽  
Plasma Plant

The first processing of low level radioactive wastes from Swiss nuclear power plants marks the successful completion of commissioning in March 2004 of a treatment facility for low and intermediate level radioactive wastes, which is operated with the help of plasma technology. The theoretical principles of this metallurgy-derived process technology are based on plasma technology, which has already been used for a considerable period outside of nuclear technology for the production of highly pure metal alloys and for the plasma synthesis of acetylene. The commercial operation of the Plasma Plant owned by Zwischenlager Wuerenlingen AG (ZWILAG) has also enabled this technology to be used successfully for the first time in the nuclear field, especially in compliance with radiation protection aspects. In addition to a brief presentation of the technology used in the plant, the melting process under operating conditions will be explained in more detail. The separation factors attained and volume reductions achieved open interesting perspectives for the further optimisation of the entire process in the future.

scholarly journals Radiochemical Methodologies Applied to Analytical Characterization of Low and Intermediate Level Wastes from Nuclear Power Plants

2019 ◽  
Vol 7 (2A) ◽  
Author(s):  
Roberto Pellacani Monteiro ◽  
Aluísio Souza Reis Junior ◽  
Geraldo Frederico Kastner ◽  
Eliane Silvia Codo Temba ◽  
Thiago César De Oliveira ◽  
...  
Keyword(s):  
Radiochemical Separation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Radioactive Wastes ◽  
Intermediate Level ◽  
Analytical Characterization ◽  
Low Level ◽  
Nuclear Techniques

The aim of this work is to present radiochemical methodologies developed at CDTN/CNEN in order to answer a program for isotopic inventory of radioactive wastes from Brazilian Nuclear Power Plants.  In this program  some radionuclides, 3H, 14C, 55Fe, 59Ni, 63Ni, 90Sr, 93Zr, 94Nb, 99Tc, 129I, 235U, 238U, 238Pu, 239+240Pu, 241Pu, 242Pu, 241Am, 242Cm e 243+244Cm, were determined  in Low Level Wastes (LLW) and Intermediate Level Wastes (ILW) and a protocol of analytical methodologies based on radiochemical separation steps and spectrometric and nuclear techniques was stablished.

A Study on the Vitrification Tests of the Low Level Radioactive Wastes Using Plasma Arc Melter System

2004 ◽  
Author(s):  
Boo Ho Yoon ◽  
Jae Hak Cho ◽  
Sang Chul Lee ◽  
Dong Woo Kang ◽  
Yong Joon Choi ◽  
...  
Keyword(s):  
Compressive Strength ◽  
Nuclear Power ◽  
Power Plants ◽  
Radioactive Wastes ◽  
Plasma Arc ◽  
Micro Hardness ◽  
Radioactive Materials ◽  
Low Level ◽  
Glass Density ◽  
Glass Formulation

For the research on the vitrification of the low-level radioactive wastes (LLRW) produced in nuclear power plants, one pilot plant with plasma arc melter system was built and several tests were done on it. Some surrogate wastes, which were spiked with several materials and were made very similar to the real LLRW, were used for these tests. For the vitrification of the surrogate wastes, the surrogate wastes were classified into the combustible, the non-combustible and the resin. Then each waste was spiked with special materials and was melted in separate. Off-gases produced in each test were picked up and analyzed. Real radioactive materials cesium (Cs) and cobalt (Co) were spiked in each wastes. Data gained from the final glass formulation were follows. Glass density is 2.42 ∼ 2.95(g/cm3), the compressive strength is 30 ∼ 175 Mpa, micro hardness is 5.5 ∼ 5.8 Gpa. The leaching ratio for Co is 1.27×10−4 ∼ 1.08×10−3 (10mL/g) and that for Cs is 2.46×10−3 ∼ 3.23×10−2 (10mL/g). The leaching speed for Co is 4.14×10−7 ∼ 5.53×10−6 (g/m2) and that for Cs is 4.58×10−5 ∼ 3.87×10−4 (g/m2). In off-gas, dioxin & furan is 0.016 mano gram on the average, CO is about 20 ppm, NO2 is about 15 ppm and SO2 is about 15 ppm.

scholarly journals Application of a Specifi c Indicator for the Estimation of Radioactive Wastes Generation Volumes During Normal Operation of Nuclear Power Plants in Russia

2020 ◽  
Vol 11 (2) ◽  
pp. 66-74
Author(s):  
A. A. Ekidin ◽  
◽  
K. L. Antonov ◽  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Radiation Safety ◽  
Radioactive Wastes ◽  
Intermediate Level ◽  
Normal Operation ◽  
Low Level ◽  
Retrospective Assessment ◽  
High Level

Generation of radioactive wastes (RW) is viewed a most urgent problem of radiation safety under normal operation of nuclear power plants (NPP). The paper demonstrates the application of a specifi c indicator (rate) of RW generation per unit of generated power (m3/GW·h) for a retrospective assessment and forecasting of RW generation volumes at Russian NPPs. Mean and median values of annual specifi c RW generation rates were calculated for each NPP based on published environmental reports of JSC Rosenergoatom Concern for the period of 2008—2018. Advantage of applying median values in retrospective and forecast assessments was shown. Medians for solid very low-level, low-level, intermediate-level and high-level radioactive waste amounted to 1.5·10−2 m3/GW·h, 3.3·10−2 m3/GW·h, 3.3·10−3 m3/GW·h and 2.8·10−4 m3/GW·h, respectively; for liquid low-level and intermediate-level waste these values accounted for 1.4·10−3 m3/GW·h, 2.5·10−3 m3/GW·h, respectively. NPPs with RBMK reactor units are characterized by the highest mean and median values of specifi c RW generation rates for all RW categories. Given various types of reactor facilities and their characteristic specifi c rates, retrospective estimates for the total volume of liquid RW was increased by 8 % and for solid RW — by 12 %. The forecast estimates based on specifi c rate medians, as well as on increased power generation planned for Russian NPPs indicates probable increase in RW generation volumes by 0.8—7.1 % (depending on waste category) from 2020 to 2027.

ISO Standardization of the Scaling Factor Method for Low- and Intermediate Level Radioactive Wastes Generated at Nuclear Power Plants

2007 ◽  
Author(s):  
Makoto Kashiwagi ◽  
Hideki Masui ◽  
Yasutaka Denda ◽  
David James ◽  
Bertrand Lante`s ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Scaling Factor ◽  
Project Team ◽  
Radioactive Wastes ◽  
Intermediate Level ◽  
International Standard ◽  
Factor Method ◽  
International Standardization

Low- and intermediate-level radioactive wastes (L-ILW) generated at nuclear power plants are disposed of in various countries. In the disposal of such wastes, it is required that the radioactivity concentrations of waste packages should be declared with respect to difficult-to-measure nuclides (DTM nuclides), such as C-14, Ni-63 and α-emitting nuclides, which are often limited to maximum values in disposal licenses, safety cases and/or regulations for maximum radioactive concentrations. To fulfill this requirement, the Scaling Factor method (SF method) has been applied in various countries as a principal method for determining the concentrations of DTM nuclides. In the SF method, the concentrations of DTM nuclides are determined by multiplying the concentrations of certain key nuclides by SF values (the determined ratios of radioactive concentration between DTM nuclides and those key nuclides). The SF values used as conversion factors are determined from the correlation between DTM nuclides and key nuclides such as Co-60. The concentrations of key nuclides are determined by γ ray measurements which can be made comparatively easily from outside the waste package. The SF values are calculated based on the data obtained from the radiochemical analysis of waste samples. The use of SFs, which are empirically based on analytical data, has become established as a widely recognized “de facto standard”. A number of countries have independently collected nuclide data by analysis over many years and each has developed its own SF method, but all the SF methods that have been adopted are similar. The project team for standardization had been organized for establishing this SF method as a “de jure standard” in the international standardization system of the International Organization for Standardization (ISO). The project team for standardization has advanced the standardization through technical studies, based upon each country’s study results and analysis data. The conclusions reached by the project team was published as ISO International Standard 21238:2007 “The Scaling Factor method to determine the radioactivity of low- and intermediate-level radioactive waste packages generated at nuclear power plants” [1]. This paper gives an introduction to the international standardization process for the SF method and the contents of the recently published International Standard.

Design, Analysis and System-Level Modelling of a Single Axis Capacitive Accelerometer

2016 ◽  
Author(s):  
Zakriya Mohammed ◽  
Owais Talaat Waheed ◽  
Ibrahim (Abe) M. Elfadel ◽  
Aveek Chatterjee ◽  
Mahmoud Rasras
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Noise Analysis ◽  
Linear Displacement ◽  
Operating Conditions ◽  
System Level ◽  
Complete Analysis ◽  
Capacitive Accelerometer ◽  
Static Capacitance ◽  
Cross Axis

The paper demonstrates the design and complete analysis of 1-axis MEMS capacitive accelerometer. The design is optimized for high linearity, high sensitivity, and low cross-axis sensitivity. The noise analysis is done to assure satisfactory performance under operating conditions. This includes the mechanical noise of accelerometer, noise due to interface electronics and noise caused by radiation. The latter noise will arise when such accelerometer is deployed in radioactive (e.g., nuclear power plants) or space environments. The static capacitance is calculated to be 4.58 pF/side. A linear displacement sensitivity of 0.012μm/g (g = 9.8m/s2) is observed in the range of ±15g. The differential capacitive sensitivity of the device is 90fF/g. Furthermore, a low cross-axis sensitivity of 0.024fF/g is computed. The effect of radiation is mathematically modelled and possibility of using these devices in radioactive environment is explored. The simulated noise floor of the device with electronic circuit is 0.165mg/Hz1/2.

Utilization of Radioactive Waste for Solidifying Material to Fill Waste Forms

2001 ◽  
Author(s):  
Takeshi Ishikura ◽  
Daiichiro Oguri
Keyword(s):  
Radioactive Waste ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Filling Ratio ◽  
Improved Method ◽  
Low Level ◽  
Waste Forms ◽  
Disposal Facility ◽  
Low Level Radioactive Waste

Abstract Minimizing the volume of radioactive waste generated during dismantling of nuclear power plants is a matter of great importance. In Japan waste forms buried in shallow burial disposal facility as low level radioactive waste (LLW) must be solidified by cement with adequate strength and must extend no harmful openings. The authors have developed an improved method to minimize radioactive waste volume by utilizing radioactive concrete and metal for mortar to fill openings in waste forms. Performance of a method to pre-place large sized metal or concrete waste and to fill mortar using small sized metal or concrete was tested. It was seen that the improved method substantially increases the filling ratio, thereby decreasing the numbers of waste containers.

Thermal Design Options Using Uranium Carbide and Uranium Dicarbide in SCWR Uniformly-Heated Fuel Channel

2009 ◽  
Author(s):  
Leyland J. Allison ◽  
Lisa Grande ◽  
Sally Mikhael ◽  
Adrianexy Rodriguez Prado ◽  
Bryan Villamere ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Operating Conditions ◽  
Thermal Design ◽  
Nuclear Fuels ◽  
Viable Option ◽  
Bulk Fluid ◽  
Fuel Channel ◽  
Uranium Carbide

SuperCritical Water-cooled nuclear Reactor (SCWR) options are one of the six reactor options identified in Generation IV International Forum (GIF). In these reactors the light-water coolant is pressurized to supercritical pressures (up to approximately 25 MPa). This allows the coolant to remain as a single-phase fluid even under supercritical temperatures (up to approximately 625°C). SCW Nuclear Power Plants (NPPs) are of such great interest, because their operating conditions allow for a significant increase in thermal efficiency when compared to that of modern conventional water-cooled NPPs. Direct-cycle SCW NPPs do not require the use of steam generators, steam dryers, etc. allowing for a simplified NPP design. This paper shows that new nuclear fuels such as Uranium Carbide (UC) and Uranium Dicarbide (UC2) are viable option for the SCWRs. It is believed they have great potential due to their higher thermal conductivity and corresponding to that lower fuel centerline temperature compared to those of conventional nuclear fuels such as uranium dioxide, thoria and MOX. Two conditions that must be met are: 1) keep the fuel centreline temperature below 1850°C (industry accepted limit), and 2) keep the sheath temperature below 850°C (design limit). These conditions ensure that SCWRs will operate efficiently and safely. It has been determined that Inconel-600 is a viable option for a sheath material. A generic SCWR fuel channel was considered with a 43-element bundle. Therefore, bulk-fluid, sheath and fuel centreline and HTC profiles were calculated along the heated length of a fuel channel.

Environmental Fatigue Test of CF8M With a Small Autoclave Simulating PWR Conditions

2008 ◽  
Author(s):  
Il-Seok Jeong ◽  
Gag-Hyeon Ha ◽  
Tae-Ryoung Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Low Cycle Fatigue ◽  
Test Procedure ◽  
Cyclic Strain ◽  
Fatigue Tests ◽  
Operating Conditions ◽  
Measuring Instruments ◽  
Pressurized Water Reactor ◽  
Pressurized Water

To develop a fatigue design curve of cast stainless steel CF8M used in primary piping material of nuclear power plants, low-cycle fatigue tests have been conducted by Korea Electric Power Research Institute (KEPRI). A small autoclave simulated the environment of a pressurized water reactor (PWR), 15 MPa and 315 °C. Fatigue life was measured in terms of the number of cycles with the variation of strain amplitudes at 0.04%/s strain rate. A small autoclave of 1 liter and cylindrical solid fatigue specimens were used for the strain-controlled low cycle environmental fatigue tests to make the experiments convenient. However, it was difficult to install displacement measuring instruments at the target length of the specimens inside the autoclave. To mitigate the difficulty displacement data measured at the shoulders of the specimen were calibrated based on the data relation of the target and shoulder length of the specimen during hot air test conditions. KEPRI developed a test procedure to perform low cycle environmental fatigue tests in the small autoclave. The procedure corrects the cyclic strain hardening effect by performing additional tests in high temperature air condition. KEPRI verified that the corrected test result agreed well with that of finite element method analysis. The process of correcting environmental fatigue data would be useful for producing reliable fatigue curves using a small autoclave simulating the operating conditions of a PWR.

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