Corrosion of Carbon Steel Containers During Open Site Tests of Vitrified Radioactive Waste

1997 ◽  
Vol 506 ◽  
Author(s):  
A.S. Barinov ◽  
M.I. Ojovan ◽  
N.V. Ojovan
Keyword(s):  
Carbon Steel ◽  
Radioactive Waste ◽  
Nuclear Power ◽  
Atmospheric Corrosion ◽  
Power Plants ◽  
Matrix Material ◽  
Waste Glass ◽  
Testing Site ◽  
Intermediate Level Radioactive Waste

ABSTRACTIn the middle of 80-s the Moscow SIA “Radon” initiates a program on long term tests of vitrified intermediate level radioactive waste from nuclear power plants with different types'of reactors and with different composition of waste. Glass blocks were prepared and disposed for tests in experimental shallow ground repository and on open testing area. Borosilicate glass as matrix material is investigated in order to obtain data on real behaviour of glass [1]. The waste glass was poured in carbon steel rectangular containers therefore containers also were subjected to aqueous and atmospheric corrosion. Exposure conditions at testing site are varying depending on season. They can be described only as average parameters with some uncertainties.

Heating Impact on Corrosion Mechanism of Carbon Steel Surrounded by Bentonite

2020 ◽  
Author(s):  
Masao Uyama ◽  
Hiroyuki Saito ◽  
Tomoya Iioka
Keyword(s):  
Carbon Steel ◽  
Radioactive Waste ◽  
Frequency Characteristic ◽  
Power Plants ◽  
Electric Heater ◽  
Dry Density ◽  
Corrosion Mechanism ◽  
Low Carbon ◽  
Corrosion Condition

Abstract The long-term safety of geological disposal of radioactive waste is studied through several simulations. Before underground disposal, radioactive waste is stored for 30 to 50 years at facilities near nuclear power plants to cool it down to around 100 degrees. It is then placed in steel canisters surrounded by artificial materials such as bentonite and concrete. To determine the long-term safety and stability of this disposal method, we’ve studied the corrosion rate of the steel canisters under different conditions using electro-chemical impedance spectroscopy (EIS). This paper describes the corrosion of the carbon steel and elucidates the corroded condition using EIS measurement. EIS was adopted to estimate the corrosion condition from the impedance frequency characteristic. In our experiment, samples of bentonite and carbon steel were compacted to Kunigel V1 of 1.37Mg/m3 dry density with several different water contents, and SM400 as a low carbon steel inside compacted bentonite. An electric heater was set inside the steel canister to maintain the temperature at 100 degrees Celsius. This model was made to a scale of around 1/120 as a current concept of a vertical disposal plan and reproduced the enclosed situation after underground emplacement of the radioactive waste. During heating, we conducted EIS measurements and set this data result as an equivalent circuit. We noted some different trends of impedance frequency characteristic depending on the bentonite’s water content and the heating time. From this result, we estimated the corrosion condition to analyze the corrosion products.

Commercialization Project of Ulchin Vitrification

2010 ◽  
Author(s):  
Hyun-Jun Jo ◽  
Cheon-Woo Kim ◽  
Tae-Won Hwang
Keyword(s):  
Radioactive Waste ◽  
Nuclear Power ◽  
Power Plants ◽  
Cold Test ◽  
Intermediate Level ◽  
Performance Tests ◽  
Cold Crucible ◽  
Commercial Operation ◽  
Intermediate Level Radioactive Waste ◽  
Hot Test

The Ulchin Vitrification Facility (UVF), to be used for the vitirification of low- and intermediate-level radioactive waste (LILW) generated by nuclear power plants (NPPs), is the world’s first commercial facility using Cold Crucible Induction Melter (CCIM) technology. The construction of the facility was begun in 2005 and was completed in 2007. From December 2007 to September 2009, all key performance tests, such as the system functional test, the cold test, the hot test, and the real waste test, were successfully carried out. The UVF commenced commercial operation in October 2009 for the vitrification of radioactive waste.

Optimization of Concrete Container Composition in Radioactive Waste Management

1985 ◽  
Vol 50 ◽  
Author(s):  
I. B. Plecas ◽  
Li. L. Mihajlovic ◽  
A. M. Kostadinovic
Keyword(s):  
Mechanical Properties ◽  
Radioactive Waste ◽  
Waste Management ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Leakage Rate ◽  
Intermediate Level ◽  
Radioactive Waste Management ◽  
Intermediate Level Radioactive Waste

AbstractIn this paper an optimization of concrete container composition, used for storing low and intermediate level radioactive waste from nuclear power plants in Yugoslavia, is presented.Mechanical properties 37−52 MPa, permeability 1.07. 10−13 - 1.50. 10−11cm2 and leakage rate 3.66. 10−6 - 1.77. 10−4 cm/d for concrete made of commercial materials, were tested.

Long-term safety of the extended SFR - Methodology and conclusions from the SR-PSU project

MRS Advances ◽  
2016 ◽  
Vol 1 (61) ◽  
pp. 4075-4080
Author(s):  
Fredrik Vahlund
Keyword(s):  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Surface System ◽  
Radiological Safety ◽  
The Impact

ABSTRACTSince 1988 the Swedish Nuclear Fuel and Waste Management Co. operates a repository for low- and intermediate-level short-lived radioactive waste, SFR, in Forsmark, Sweden. Due to decommissioning of the nuclear power plants additional storage capacity is needed. In December 2014, an application to extend the repository was therefore submitted. One key component of this application was an assessment of post-closure safety of the extended SFR. For this safety assessment, a methodology based on that developed by SKB for the spent nuclear fuel repository was used and the impact of the degradation of repository components, the evolution of the surface system and changes of future climate on the radiological safety of the repository was assessed over a period of 100,000 years. The central conclusion of the SR-PSU safety assessment is that the extended SFR repository meets requirements on protection of human health and of the environment that have been established by the Swedish radiation safety authority for the final disposal of radioactive waste. Furthermore, the design of the repository was shown suitable for the waste selected and the applied methodology suitable for the safety assessment.

scholarly journals Simulation of MASPn Experiments in MISTRA Test Facility with COCOSYS Code

2008 ◽  
Vol 2008 ◽  
pp. 1-7 ◽  
Author(s):  
Mantas Povilaitis ◽  
Egidijus Urbonavičius
Keyword(s):  
Boundary Conditions ◽  
Nuclear Power ◽  
Power Plants ◽  
Volume Fraction ◽  
Characteristic Length ◽  
Initial Conditions ◽  
Test Facility ◽  
Work Package ◽  
Acceptable Agreement

An issue of the stratified atmospheres in the containments of nuclear power plants is still unresolved; different experiments are performed in the test facilities like TOSQAN and MISTRA. MASPn experiments belong to the spray benchmark, initiated in the containment atmosphere mixing work package of the SARNET network. The benchmark consisted of MASP0, MASP1 and MASP2 experiments. Only the measured depressurisation rates during MASPn were available for the comparison with calculations. When the analysis was performed, the boundary conditions were not clearly defined therefore most of the attention was concentrated on MASP0 simulation in order to develop the nodalisation scheme and define the initial and boundary conditions. After achieving acceptable agreement with measured depressurisation rate, simulations of MASP1 and MASP2 experiments were performed to check the influence of sprays. The paper presents developed nodalisation scheme of MISTRA for the COCOSYS code and the results of analyses. In the performed analyses, several parameters were considered: initial conditions, loss coefficient of the junctions, initial gradients of temperature and steam volume fraction, and characteristic length of structures. Parametric analysis shows that in the simulation the heat losses through the external walls behind the lower condenser installed in the MISTRA facility determine the long-term depressurisation rate.

scholarly journals Stochastic Optimization for Long Term Capital Structures, Systems, and Components Refurbishment and Replacement

2020 ◽  
Author(s):  
Congjian Wang ◽  
Diego Mandelli ◽  
Shawn St Germain ◽  
Curtis Smith ◽  
David Morton ◽  
...  
Keyword(s):  
Stochastic Optimization ◽  
Nuclear Power ◽  
Power Plants ◽  
Capital Investments ◽  
Priority List ◽  
Multi Stage ◽  
Plant Operations ◽  
Economic Constraints ◽  
Returns On Investment

Abstract As commercial nuclear power plants (NPPs) pursue extended plant operations in the form of Second License Renewals (SLRs), opportunities exist for these plants to provide capital investments to ensure long-term, safe, and economic performance. Several utilities have already announced their intention to pursue extended operations for one or more of their NPPs via SLR2. The goal of this research is to develop a risk-informed approach to evaluate and prioritize plant capital investments made in preparation for, and during the period of, extended plant operations to support decisions in NPP operations. In order to prioritize project selection via a risk-informed approach we developed a single decision-making tool that integrates safety/reliability, cost, and stochastic optimization models to provide users with data analysis capabilities to more cost effectively manage plant assets. Both stochastic analysis methods — such as Monte Carlo-based sampling strategies — and multi-stage stochastic optimization strategies are employed to provide priority lists to decision-makers in support of risk-informed decisions. We applied the proposed method to a trial application of projected replacement/refurbishment expenditures for plant capital assets (i.e., structures, systems, and components [SSCs]). The objective is to optimize the SSC replacement/refurbishment schedule in terms of economic constraints, data uncertainties, and SSC reliability data, as well to generate a priority list for maximizing returns on investment.

Load Factor for Evaluating Non-Planar Flaws in Carbon Steel Piping Using Limit Load Methodology

2016 ◽  
Author(s):  
Phuong H. Hoang
Keyword(s):  
Carbon Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Pipe Wall ◽  
Local Strain ◽  
Limit Load ◽  
Evaluation Methodology ◽  
Load Factor ◽  
Wall Thinning ◽  
Steel Piping

Non-planar flaw such as local wall thinning flaw is a major piping degradation in nuclear power plants. Hundreds of piping components are inspected and evaluated for pipe wall loss due to flow accelerated corrosion and microbiological corrosion during a typical scheduled refueling outage. The evaluation is typically based on the original code rules for design and construction, and so often that uniformly thin pipe cross section is conservatively assumed. Code Case N-597-2 of ASME B&PV, Section XI Code provides a simplified methodology for local pipe wall thinning evaluation to meet the construction Code requirements for pressure and moment loading. However, it is desirable to develop a methodology for evaluating non-planar flaws that consistent with the Section XI flaw evaluation methodology for operating plants. From the results of recent studies and experimental data, it is reasonable to suggest that the Section XI, Appendix C net section collapse load approach can be used for non-planar flaws in carbon steel piping with an appropriate load multiplier factor. Local strain at non-planar flaws in carbon steel piping may reach a strain instability prior to net section collapse. As load increase, necking starting at onset strain instability leads to crack initiation, coalescence and fracture. Thus, by limiting local strain to material onset strain instability, a load multiplier factor can be developed for evaluating non-planar flaws in carbon steel piping using limit load methodology. In this paper, onset strain instability, which is material strain at the ultimate stress from available tensile test data, is correlated with the material minimum specified elongation for developing a load factor of non-planar flaws in various carbon steel piping subjected to multiaxial loading.

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