I-129 DETERMINATION IN EVAPORATOR CONCENTRATE USING OXIDATIVE EXTRACTION AND CHROMATOGRAPHIC RESIN

2021 ◽  
Vol 9 (1A) ◽  
Author(s):  
Thiago César De Oliveira ◽  
Roberto Pellacani Guedes Monteiro ◽  
Geraldo Frederico Kastner ◽  
Arno Heeren De Oliveira
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Selective Extraction ◽  
Thermal Neutrons ◽  
X Ray ◽  
Tracer Solution ◽  
Spectrometry Technique ◽  
Chromatographic Resin ◽  
Set Up

Determination of long-lived radionuclides is very important for study of the radioactive waste final deposition. In this work will be studied 129I radionuclide which present 1.6 x 107 years half-life, β-particle-emitting (Emax = 194 keV) and X-ray-emitting (E = 29.78 keV). It’s produced primarily from fission of 235U and 239Pu and for fission induced by thermal neutrons. For this reason, radiochemical procedures for 129I determination in evaporator concentrate wastes from nuclear power plants were carried out. The first procedure was based on oxidative extraction and alkaline absorption and the second one, was based on selective extraction using a chromatographic resin in order to separate iodine from its interferents. After the separation steps, the iodine activity was measured by ultra low gamma spectrometry technique. To set up the yield recovery for 129I, a tracer solution of 129I was used in order to follow the behavior of iodine during the separation steps. The yield recovery for iodine was around 75-80% for the first procedure and 80-85% for the second. The two procedures used mutually, ensure a greater efficiency in the separation of iodine from their respective interferents.

scholarly journals Ion chromatographic analysis of aqueous iodine species by conductivity and radioactivity measurements

2005 ◽  
Vol 93 (9-10) ◽  
Author(s):  
Dorothea Schumann ◽  
R. Grasser ◽  
R. Dressler ◽  
H. Bruchertseifer
Keyword(s):  
Chromatographic Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Radioactive Iodine ◽  
Cooling Water ◽  
Molecular Iodine ◽  
New Device ◽  
Iodine Species ◽  
Radioactivity Measurements

SummaryA new device was developed for the identification of several iodine species in aqueous solution using ion chromatography. Iodide, iodate and molecular iodine can be determined. (The equipment allows both conductivity and radioactivity detections.) The method is applicable for the determination of radioactive iodine contaminations in the cooling water of nuclear power plants.

scholarly journals Structural Modeling and Dynamic Analysis of a Nuclear Reactor Building

2020 ◽  
Author(s):  
Evrim Oyguc ◽  
Abdul Hayır ◽  
Resat Oyguc
Keyword(s):  
Nonlinear Dynamic ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Energy Sources ◽  
Finite Element Program ◽  
Design Spectrum ◽  
Reactor Building ◽  
Set Up

Increasing energy demand urge the developing countries to consider different types of energy sources. Owing the fact that the energy production capacity of renewable energy sources is lower than a nuclear power plant, developed countries like US, France, Japan, Russia and China lead to construct nuclear power plants. These countries compensate 80% of their energy need from nuclear power plants. Further, they periodically conduct tests in order to assess the safety of the existing nuclear power plants by applying impact type loads to the structures. In this study, a sample third-generation nuclear reactor building has been selected to assess its seismic behavior and to observe the crack propagations of the prestressed outer containment. First, a 3D model has been set up using ABAQUS finite element program. Afterwards, modal analysis is conducted to determine the mode shapes. Nonlinear dynamic time history analyses are then followed using an artificial strong ground motion which is compatible with the mean design spectrum of the previously selected ground motions that are scaled to Eurocode 8 Soil type B design spectrum. Results of the conducted nonlinear dynamic analyses are considered in terms of stress distributions and crack propagations.

scholarly journals Control Experiences Regarding Clearable Materials from Nuclear Power Plants and Nuclear Installations: Scaling Factors Determination and Measurements’ Acceptance Criteria Definition

Environments ◽  
2019 ◽  
Vol 6 (11) ◽  
pp. 120
Author(s):  
Luca Albertone ◽  
Massimo Altavilla ◽  
Manuela Marga ◽  
Laura Porzio ◽  
Giuseppe Tozzi ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fission Products ◽  
Scaling Factors ◽  
Activation Products ◽  
Long Time ◽  
Nuclear Installation ◽  
Enriched Uranium

Arpa Piemonte has been carrying out, for a long time, controls on clearable materials from nuclear power plants to verify compliance with clearance levels set by ISIN (Ispettorato Nazionale per la Sicurezza Nucleare e la Radioprotezione - National Inspectorate for Nuclear Safety and Radiation Protection) in the technical prescriptions attached to the Ministerial Decree decommissioning authorization or into category A source authorization (higher level of associated risk, according to the categorization defined in the Italian Legislative Decree No. 230/95). After the experience undertaken at the “FN” (Fabbricazioni Nucleari) Bosco Marengo nuclear installation, some controls have been conducted at the Trino nuclear power plant “E. Fermi,” “LivaNova” nuclear installation based in Saluggia, and “EUREX” (Enriched Uranium Extraction) nuclear installation, also based in Saluggia, according to modalities that envisage, as a final control, the determination of γ-emitting radionuclides through in situ gamma spectrometry measurements. Clearance levels’ compliance verification should be performed for all radionuclides potentially present, including those that are not easily measurable (DTM, Difficult To Measure). It is therefore necessary to carry out upstream, based on a representative number of samples, those radionuclides’ determination in order to estimate scaling factors (SF), defined through the logarithmic average of the ratios between the i-th DTM radionuclide concentration and the related key nuclide. Specific radiochemistry is used for defining DTMs’ concentrations, such as Fe-55, Ni-59, Ni-63, Sr-90, Pu-238, and Pu-239/Pu-240. As a key nuclide, Co-60 was chosen for the activation products (Fe-55, Ni-59, Ni-63) and Cs-137 for fission products (Sr-90) and plutonium (Pu- 238, Pu-239/Pu-240, and Pu-241). The presence of very low radioactivity concentrations, often below the detection limits, can make it difficult to determine the related scaling factors. In this work, the results obtained and measurements’ acceptability criteria are presented, defined with ISIN, that can be used for confirming or excluding a radionuclide presence in the process of verifying clearance levels’ compliance. They are also exposed to evaluations regarding samples’ representativeness chosen for scaling factors’ assessment.

Thermoelastoplastic Stress Analysis of a Thick-Walled Tube of Nonlinear Strain Hardening

1996 ◽  
Vol 118 (3) ◽  
pp. 340-346 ◽  
Author(s):  
S. Jahanian
Keyword(s):  
Strain Hardening ◽  
Nuclear Power ◽  
Power Plants ◽  
Rapid Heating ◽  
Numerical Procedure ◽  
Temperature Dependent ◽  
Nonlinear Strain ◽  
Elastic Approximation

In pressure vessel technology or nuclear power plants, some of the mechanical components are often subjected to rapid heating. If the temperature gradient during such process is high enough, thermoelastoplastic stresses may be developed in the components. These plastic deformations are permanent and may result in the incremental deformation of the structure in the long term. Accordingly, determination of thermoelastoplastic stresses during this process is an important factor in design. In this paper, a thick-walled cylinder of nonlinear strain hardening is considered for the thermoelastoplastic analysis. The properties of the material are assumed to be temperature dependent. The cylinder is subject to rapid heating of the inside surface while the outside surface is kept at the room temperature. A quasi-static and uncoupled thermoelastoplastic analysis based on incremental theory of plasticity is developed and a numerical procedure for successive elastic approximation is presented. The thermoelastoplastic stresses developed during this process are also presented. The effect of strain hardening and temperature dependency of material on the results are investigated.

Effect of Opening on the Response of Containment Model to Pressure Loading

2006 ◽  
Author(s):  
M. K. Agrawal ◽  
A. Ravi Kiran ◽  
A. K. Ghosh ◽  
H. S. Kushwaha
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Dead Weight ◽  
Modeling Methodology ◽  
Containment Model ◽  
Scale Ratio ◽  
Heavy Water Reactor ◽  
Set Up ◽  
Accident Conditions ◽  
Design Pressure

The Containment Studies Facility (CSF) is being set up in BARC for studying various containment related thermal hydraulic and other phenomena which occur during simulated accident conditions in Nuclear power Plants. The facility consists of a concrete containment model having a volumetric scale ratio of 200:1 with respect to the actual containment of Indian Pressurized Heavy Water Reactor. The structure is designed for pressure of 1.73 Kg/cm2 for specified leak tightness. Adequacy to withstand design pressure is checked by test as well as numerical analysis before commissioning of the facility. Accordingly Containment building model has been analyzed by finite element method for internal design pressure and dead weight. Analysis has been carried out for the structure with and without the opening in the containment. Effect of opening on the response of containment has been studied. The paper includes the modeling methodology, maximum deflection and stress amplification around the opening for various models.

Proposed Improvements of the ASME B&PV Code and Future Directions of CIWG

2013 ◽  
Author(s):  
Jinquan Yan ◽  
Yinbiao He ◽  
Gang Li ◽  
Hao Yu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Third Generation ◽  
Future Directions ◽  
The Past ◽  
Development Direction ◽  
Asme Code ◽  
Set Up ◽  
Near Future ◽  
First Time

The ASME B&PV Code, Section III, is being used as the design acceptance criteria in the construction of China’s third generation AP1000 nuclear power plants. This is the first time that the ASME Code was fully accepted in Chinese nuclear power industry. In the past 6 years, a few improvements of the Code were found to be necessary to satisfy the various requirements originated from these new power plant (NPP) constructions. These improvements are originated from a) the stress-strain curves needed in elastic-plastic analysis, b) the environmental fatigue issue, c) the perplexity generated from the examination requirements after hydrostatic test and d) the safe end welding problems. In this paper, the necessities of these proposed improvements on the ASME B&PV code are further explained and discussed case by case. Hopefully, through these efforts, the near future development direction and assignment of the ASME B&PV-III China International Working Group can be set up.

Experimental Study on the Interaction of Molten Sn With Water

2018 ◽  
Author(s):  
Longkun He ◽  
Pengfei Liu ◽  
Xisi Zhang ◽  
Wenjun Hu ◽  
Bo Kuang ◽  
...  
Keyword(s):  
Water Temperature ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Dynamic Pressure ◽  
Ambient Pressure ◽  
Influence Factors ◽  
Video Camera ◽  
Metallic Oxide ◽  
Set Up

In nuclear power plants, fuel-coolant interaction (FCI) often accompanied with core melt accidents, which may escalate to steam explosion destroying the integrity of structural components and even the containment under certain conditions. In the present study, a new facility for intermediate-scaled experiments named ‘Test for Interaction of MELt with Coolant’ (TIMELCO) has been set up to study FCI phenomena and thermal-hydraulic influence factors in metal or metallic oxide/water mixtures with melt at maximum 2750°C. The first series of tests was performed using 3kg of Sn which was heated to 800°Cand jetted into a column of 1m water depth (300mm in diameter) under 0.1MPa ambient pressure. The main changing parameter was water temperature, at 60 °C and 72 °C respectively. From the high-speed video camera, violent explosion phenomenon occurred at water temperature of 60°C, while no evident explosion observed at 72°C. The size of melt debris at 60°C is smaller than this at 72°C.On the contrary, the dynamic pressure at 60°C is larger. The results indicate that water temperature has an important effect on FCI and decreasing the temperature of the coolant is advantageous to the explosion.

Analysis of Non-Uniformly Degraded Pipe Sections

2003 ◽  
Author(s):  
Amy J. Smith ◽  
Keshab K. Dwivedy
Keyword(s):  
Wall Thickness ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Operating Cycle ◽  
Moment Capacity ◽  
Asme Code ◽  
Uniform Cross Section ◽  
Uniform Wall ◽  
Set Up

The management of flow assisted corrosion (FAC) has been a part of the maintenance of piping in nuclear power plants for more than 15 years. Programs have been set up to identify vulnerable locations, perform inspections, characterize the degraded configurations, and evaluate the structural integrity of the degraded sections. The section of the pipe is repaired or replaced if the structural integrity cannot be established for the projected degraded section at the next outage. During the past 15 years, significant improvements have been made to every aspect of the program including structural integrity evaluation. Simplified methods and rules are established in ASME Section XI code and in several code cases for verifying structural integrity. The evaluation of structural integrity is performed during the plant outage prior to a decision for repair or replacement. Any improvement in structural integrity evaluation to extend the life of a component by one additional operating cycle can help in performance of repair/replacement of component in a planned manner. Simplified methods and rules provided in the code can be easily used for analysis of pipe sections with degraded area with uniform wall thickness and for non-uniformly degraded sections, provided the degraded portions are modeled with uniform wall thickness equal to the lowest thickness of the section. The representation of a non-uniformly degraded section in this manner is necessarily conservative. The purpose of this paper is to develop methodology to analyze the non-uniformly degraded sections subjected to pressure and moment loading by modeling it in a manner that accounts for the non-uniform cross-section. The formulation developed here is more realistic than the code methodology and is still conservative. The results are presented in form of charts comparing the limit moment capacity of the degraded sections calculated by the formulation in this paper with that using ASME code formulation. The paper concludes that the proposed formulation can be used to supplement the ASME Code method to extend the remaining life of FAC degraded components.

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