Testing and Numerical Simulation of Elastomers: From Specimen Tests to Simulation of Seal Behavior Under Assembly Conditions

2017 ◽  
Author(s):  
Mike Weber ◽  
Anja Kömmling ◽  
Matthias Jaunich ◽  
Dietmar Wolff ◽  
Uwe Zencker ◽  
...  
Keyword(s):  
Finite Element ◽  
Spent Nuclear Fuel ◽  
Circular Cross Section ◽  
Computer Code ◽  
Intermediate Level ◽  
Sensitivity Analyses ◽  
Spent Fuel ◽  
Test Configuration ◽  
Waste Container

Extended periods of interim storage are more relevant in Germany due to delays in the siting procedure to establish a deep geological repository for spent nuclear fuel, high level radioactive waste and in low/intermediate level waste container storage designated for the Konrad repository. BAM is involved in national cask licensing procedures and responsible for the evaluation of cask-related long-term safety issues. The long-term performance of elastomer seals for lid systems of transport and storage casks, whether used as auxiliary seals in spent fuel casks or as primary seals for low and intermediate level waste packages, is an important issue in this context. The polymeric structure of these seals causes a complex mechanical behavior with time-dependent sealing force reduction. The results of a comprehensive purpose-designed test program consisting of basic compression and tension tests as well as relaxation tests on unaged specimens of representative types of elastomers (fluorocarbon rubber (FKM) and ethylene propylene diene rubber (EPDM)) at different temperatures and strain rates are presented. They were used to identify the constitutive behavior and to obtain parameters for finite element material models provided by the computer code ABAQUS®. After estimating the influence of parameters such as Poisson’s ratio and friction coefficient by sensitivity analyses, the chosen values for the finite element simulation were validated by comparison with specimen test results. Based on this preliminary work the simulation of a specific laboratory test configuration containing a typical elastomer seal with circular cross section is presented. The chosen finite element material model and the related parameters had to show that they are able to represent not only the specimen behavior under predominantly uniaxial load but also the more complex stress states in real components. Deviations between the measured and calculated results are pointed out and discussed. The results from this work will be utilized in future simulations of aged elastomer behavior.

Testing and Numerical Simulation of Elastomeric Seals Under Consideration of Time Dependent Effects

2016 ◽  
Author(s):  
Mike Weber ◽  
Anja Kömmling ◽  
Matthias Jaunich ◽  
Dietmar Wolff ◽  
Uwe Zencker ◽  
...  
Keyword(s):  
Finite Element ◽  
Spent Nuclear Fuel ◽  
Viscoelastic Behavior ◽  
Element Model ◽  
Intermediate Level ◽  
High Level Waste ◽  
Spent Fuel ◽  
Safety Issues ◽  
Elastomeric Seals

Due to delays in the siting procedure to establish a deep geological repository for spent nuclear fuel and high level waste and in construction of the already licensed Konrad repository for low and intermediate level waste, extended periods of interim storage will become more relevant in Germany. BAM is involved in most of the cask licensing procedures and is responsible for the evaluation of cask-related long-term safety issues. Elastomeric seals are widely used as barrier seals for containers for low and intermediate level radioactive waste. In addition they are also used as auxiliary seals in spent fuel storage and transportation casks (dual purpose casks (DPC)). To address the complex requirements resulting from the described applications, BAM has initiated several test programs for investigating the behavior of elastomeric seals. These include experiments concerning the hyperelastic and viscoelastic behavior at different temperatures and strain rates, the low temperature performance down to −40°C, the influence of gamma irradiation and the aging behavior. The first part of the paper gives an overview of these tests, their relevant results and their possible impact on BAM’s work as a consultant in the framework of approval and licensing procedures. The second part presents an approach of the development of a finite element model using the finite element code ABAQUS®. The long-term goal is to simulate the complex elastomeric behavior in a complete lid closure system under specific operation and accident conditions.

Na/Ca selectivity coefficients of montmorillonite in perchlorate solution at different temperatures

2012 ◽  
Vol 1475 ◽  
Author(s):  
Aku Itälä ◽  
Arto Muurinen
Keyword(s):  
Cation Exchange ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Selectivity Coefficients ◽  
Crystalline Bedrock ◽  
Bentonite Buffer ◽  
Different Temperatures ◽  
C 50 ◽  
Exchange Selectivity

ABSTRACTThe Finnish spent nuclear fuel disposal is based on the Swedish KBS-3 concept in crystalline bedrock. The concept aims at long-term isolation and containment of spent fuel in copper canisters surrounded by bentonite buffer which mostly consists of montmorillonite. For the long-term modelling of the chemical processes in the buffer, the cation-exchange selectivity coefficients have to be known at different temperatures. In this work, the cation-exchange selectivity coefficients and cation-exchange isotherms were determined in batch experiments for montmorillonite at three different temperatures (25 °C, 50 °C and 75 °C). Five different ratios of NaClO4/Ca(ClO4)2 were used in the experimental solutions. After equilibration the solution and montmorillonite were separated and the solution analysed to get the desired exchange parameters. The experiments were modelled with a computational model capable of taking into account the physicochemical processes that take place in the experiment.

Stress Concentration Factor Calculation for the Hooks in Lifting Appliance of Spent Fuel Well Cover

2014 ◽  
Author(s):  
Xiang Liu ◽  
Yue Li ◽  
Jinhua Wang ◽  
Bin Wu
Keyword(s):  
Finite Element Method ◽  
Finite Element ◽  
Stress Concentration ◽  
Stress Concentration Factor ◽  
Concentration Factor ◽  
Spent Nuclear Fuel ◽  
Curved Beam ◽  
Spent Fuel ◽  
The Finite Element Method ◽  
High Temperature Reactor

The spent nuclear fuel of HTR-PM (High Temperature Reactor–Pebblebed Modules) will be dry stored in wells. In the mouth of each well, there is a cover weighing 11 tons. A lifting appliance with three hooks is used to open and close the covers. The hooks are L-shaped with fillet at the inside corner. The stress concentration at the corner has a significant impact on the strength and fatigue life of hooks. For optimizing the structure of the hook, the stress concentration factor related to the radius of fillet is calculated by both theoretical and numerical methods. The theoretical calculation is based on the Saint-Venant’s Principle and the analytical solution of a curved beam. The result is consistent with the numerical calculation performed by the finite element method.

Behavior of UO2 in the RBMK-1000 Spent Fuel under Oxidizing Conditions

2004 ◽  
Vol 824 ◽  
Author(s):  
A. B. Kolyadin ◽  
V. Ya. Mishin ◽  
K. Ya. Mishin ◽  
A. S. Aloy ◽  
T. I. Koltsova
Keyword(s):  
Thermal Oxidation ◽  
Spent Nuclear Fuel ◽  
Oxidation Process ◽  
Spent Fuel ◽  
X Ray Diffraction ◽  
X Ray ◽  
Dry Storage ◽  
Different Temperatures ◽  
Xrd Techniques

AbstractThe oxidation of UO2–type spent nuclear fuel (SNF) in gaseousmedia was studied at different temperatures and oxygen contents using gravimetric and powder X-ray diffraction (XRD) techniques. The aim of the study was to determine the mechanism(s) of thermal-oxidation alteration of SNF during long-term dry storage. The samples used in the experiments were chips of RBMK-1000 fuel rods.Oxidation of UO2with a mean burn-up of 10.7 and 19.73 MW d/kg in humid air was observed at a temperature as low as 150°C. At 200°C nearly all of the UO2was transformed into U3O8 between 3500-4000 hours. In a humid nitrogen environment containing of 0.05-1.3 vol. % oxygen at 300°C, the UO2 completely transformed to U3O8 between 2500-3000 hours. Oxidation of UO2in samples with small amounts of jacket damage (e.g., <0.04 MM2)ll progresses more slowly and after â3000 hours the oxygen-to-uranium ratio was 2.56.Stabilization of the oxidation process was not observed in the fuel samples upto an O/U ratio of 2.4, which may be attributed to the smallburn-up of the fuel under investigation.

scholarly journals Numerical Analyses of Locomotive Impacts on a Spent Fuel Truck Cask and Trailer

2005 ◽  
Author(s):  
Doug Ammerman ◽  
Dave Stevens ◽  
Matt Barsotti
Keyword(s):  
Finite Element ◽  
High Speed ◽  
Applied Research ◽  
Spent Nuclear Fuel ◽  
Numerical Study ◽  
Numerical Approach ◽  
Finite Element Models ◽  
Spent Fuel ◽  
Physical Test ◽  
Grade Crossing

During the transportation of spent nuclear fuel by truck, the possibility exists that a train could run into the spent fuel cask at a grade crossing. Sandia National Laboratories has conducted a numerical study to assess the possibility of cask breach or material release in the event of a high-speed, broadside locomotive collision. A numerical approach has the advantage over conducting a physical test as was done in the 1970s [1] in that varying parameters can be examined. For example, one of the criticisms of the 1970s test was the height of the cask. In the test, the centerline of the cask was above the main frame-rails of the locomotive. In this study the position of the cask with respect to the locomotive was varied. The response of the cask and trailer in different collision scenarios was modeled numerically with LS-DYNA [2]. The simulations were performed as a collaborative endeavor between Sandia National Laboratories (SNL), Applied Research Associates, Inc. (ARA) and Foster-Miller, Inc (FMI). ARA developed the GA-4 Spent Fuel Cask and Cask Transporter models described in this report. These models were then combined with two existing FMI heavy freight locomotive finite element models to create the overall simulation scenarios. The modeling effort, results, and conclusions are presented in this paper.

Sorption of Actinides in Well-Defined Oxidation States on Geologic Media

1981 ◽  
Vol 11 ◽  
Author(s):  
B. Allard ◽  
U. Olofsson ◽  
B. Torstenfelt ◽  
H. Kipatsi ◽  
K. Andersson
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Oxidation States ◽  
Spent Fuel ◽  
Long Term Storage ◽  
Term Storage ◽  
Geologic Media

The long-lived actinides and their daughter products largely dominate the biological hazards from spent nuclear fuel already from some 300 years after the discharge from the reactor and onwards . Therefore it is essential to make reliable assessments of the geochemistry of these elements in any concept for long-term storage of spent fuel or reprocessing waste, etc.

Radionuclides Release Model for Performance Assessment Studies of Spent Nuclear Fuel in Geological Disposal

2004 ◽  
Vol 824 ◽  
Author(s):  
Christophe Poinssot ◽  
Patrick Lovera ◽  
Cécile Ferry
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Hydrogen Effect ◽  
Spent Fuel ◽  
Geological Disposal ◽  
The Matrix ◽  
Simple Kinetic Model ◽  
Scientific Results ◽  
Release Model

AbstractIn the framework of the research conducted on the long term evolution of spent nuclear fuel in geological disposal conditions, a source term model has been developed to evaluate the instantaneous release of RN (Instant Release Fraction IRF) and the delayed release of the RN which are embedded within the matrix. This model takes into account all the scientific results currently available in the literature except the hydrogen effect. IRF was assessed by considering the evolution with time of the RN inventories located within the fuel microstructure to which no confinement properties can be allocated on the long term (rim, gap, grain boundaries). It allows to propose some reference bounding values for the IRF as a function of time of canister breaching and burnup. The matrix radiolytic dissolution was modeled by a simple kinetic model neglecting the radiolytic species recombination and the influence of aqueous ligands and radiolytic oxidants were supposed to completely react with the fuel surface. Spent fuel performance was therefore demonstrated to deeply depend on the reactive surface area.

scholarly journals Recycling versus Long-Term Storage of Nuclear Fuel: Economic Factors

2013 ◽  
Vol 2013 ◽  
pp. 1-7 ◽  
Author(s):  
B. Yolanda Moratilla Soria ◽  
Maria Uris Mas ◽  
Mathilde Estadieu ◽  
Ainhoa Villar Lejarreta ◽  
David Echevarria-López
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Economic Factors ◽  
International Studies ◽  
Closed Cycle ◽  
Spent Fuel ◽  
Long Term Storage ◽  
Open Cycle ◽  
Term Storage

The objective of the present study is to compare the associated costs of long-term storage of spent nuclear fuel—open cycle strategy—with the associated cost of reprocessing and recycling strategy of spent fuel—closed cycle strategy—based on the current international studies. The analysis presents cost trends for both strategies. Also, to point out the fact that the total cost of spent nuclear fuel management (open cycle) is impossible to establish at present, while the related costs of the closed cycle are stable and known, averting uncertainties.

The effects of alpha-radiolysis on UO2 dissolution determined from electrochemical experiments with 238Pu-doped UO2

2002 ◽  
Vol 90 (9-11) ◽  
Author(s):  
S. Stroes-Gascoyne ◽  
F. King ◽  
J. S. Betteridge ◽  
F. Garisto
Keyword(s):  
Dose Rate ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Dissolution Rates ◽  
Term Stability ◽  
Long Term Stability ◽  
Alpha Radiolysis

SummaryThe long-term stability of spent nuclear fuel under deep geologic repository conditions will be determined mostly by the influence of α-radiolysis, since the dose-rate for α-radiolysis will exceed that for γ/β-radiolysis beyond a fuel age of ∼100 years and will persist for more than 10000 years. Dissolution rates derived from studies with currently available spent fuel include radiolysis effects from γ/β- as well as α-radiolysis. The use of external α-sources and chemically added H

scholarly journals Analysis of Some Computer Codes Used for Evaluation of Spent Nuclear Fuel Radiation Parameters

2019 ◽  
pp. 44-50
Author(s):  
A. Smaizys ◽  
E. Narkunas ◽  
V. Rudychev ◽  
Y. Rudychev
Keyword(s):  
Fuel Assembly ◽  
Nuclear Fuel ◽  
Input Data ◽  
Spent Nuclear Fuel ◽  
Operation Time ◽  
Computer Code ◽  
Chemical Compositions ◽  
Spent Fuel ◽  
Management Options ◽  
Computer Codes

The radiation parameters such as radionuclide content and activities, fluxes and energy spectrum of gamma and neutrons of spent nuclear fuel are essential when planning further spent fuel management options – interim wet or dry storage or disposal into a geological repository. Radiation parameters determine the design of a storage or disposal facility, what materials, structures and thicknesses of structures should be used to provide adequate biological shielding. Experimental measurements of spent fuel radiation parameters are rather complicated and expensive, therefore numerical methods are widely used. Various computer codes (APOLLO, BOXER, CASMO, FISPACT, ORIGEN-S, WIMS, etc.) have been developed to simulate the irradiation processes of nuclear fuel and to obtain resulting radiation parameters. Irrespective of the used computer code, the input data firstly must be entered into that code. When simulating nuclear fuel irradiation and burn-up in the reactor core, the geometrical parameters of the fuel assembly, materials’ data (chemical compositions, densities), the operating parameters of the reactor (power, operation time, coolant parameters, etc.) shall be entered into the program as initial data. Fairly often approximations of the input data are performed, for example, fuel rods in a fuel assembly are homogenized and geometrically described as a solid cylinder, the reactor operation time is assumed as continuous and at constant power. The particularity of the input data and accepted assumptions depend on what initial information is available and on the capabilities of the computer code. The modelled spent fuel radiation parameters depend not only on the input data and assumptions, but also on the cross-section databases that are used in computer codes. Computer codes TRITON, ORIGEN-S and FISPACT have been used to model the concentration of actinides and fission products in the spent fuel from the RBMK-1000 reactor. The obtained results are compared and possible reasons for the differences in the modelling results are discussed.

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