Ductile Cast Iron for Transportation Cask Bodies

2003 ◽  
Author(s):  
E. P. Warnke ◽  
W. Steinwarz ◽  
W. Sowa ◽  
R. Hu¨ggenberg
Keyword(s):  
Cast Iron ◽  
Radioactive Waste ◽  
Ductile Cast Iron ◽  
Nuclear Radiation ◽  
Spent Fuel ◽  
Accident Conditions ◽  
Fuel Elements ◽  
And Storage

Casks for transportation and storage are necessary for the handling of spent fuel elements and other radioactive waste. These casks must provide sufficient shielding against nuclear radiation, safe confinement even under severe hypothetical mechanical and thermal accident conditions and assurance of the subcriticality of the content under these conditions. Since more than 20 years ductile cast iron is established as a well qualified and proven material for the fabrication of the cask body of the CASTOR and MOSAIK type casks.

Simulation and Analysis: The Dose Distribution of KBS-3 Spent Nuclear Fuel Canister by MCNP

2010 ◽  
Author(s):  
Bo Yang ◽  
He-xi Wu ◽  
Yi-bao Liu
Keyword(s):  
Cast Iron ◽  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Nuclear Power ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Ductile Cast Iron ◽  
Nuclear Industry ◽  
Spent Fuel ◽  
High Level

With the sustained and rapid development of the nuclear power plants, the spent fuel which is produced by the nuclear power plants will be rapidly rising. Spent fuel is High-level radioactive waste and should be disposed safely, which is important for the environment of land, public safety and health of the nuclear industry, the major issues of sustainable development and it is also necessary part for the nuclear industry activities. It is important to study and resolve the high-level radioactive waste repository problem. Spent nuclear fuel is an important component in the radioactive waste, The KBS-3 canister for geological disposal of spent nuclear fuel in Sweden consists of a ductile cast iron insert and a copper shielding. The ductile cast iron insert provides the mechanical strength whereas the copper protects the canister from corrosion. The canister inserts material were referred to as I24, I25 and I26, Spent nuclear fuel make the repository in high radiant intensity. The radiation analysis of canister insert is important in canister transport, the dose analysis of repository and groundwater radiolysis. Groundwater radiolysis, which produces oxidants (H2O2 and O2), will break the deep repository for spent nuclear fuel. The dose distribution of canister surface with different kinds of canister inserts (I24, I25 and I26) is calculated by MCNP (Ref. 1). Analysing the calculation results, we offer a reference for selecting canister inserts material.

Application of Leakage Rates Measured on Scaled Cask or Component Models to the Package Containment Safety Assessment

2018 ◽  
Author(s):  
Annette Rolle ◽  
Viktor Ballheimer ◽  
Tino Neumeyer ◽  
Frank Wille
Keyword(s):  
Test Results ◽  
Spent Fuel ◽  
Thermal Tests ◽  
Containment System ◽  
Leak Tightness ◽  
Accident Conditions ◽  
High Level ◽  
Elastomeric Seals ◽  
Component Models ◽  
And Storage

The containment systems of transport and storage casks for spent fuel and high level radioactive waste usually include bolted lids with metallic or elastomeric seals. The mechanical and thermal loadings associated with the routine, normal and accident conditions of transport can have a significant effect on the leak tightness of such containment system. Scaled cask models are often used for providing the required mechanical and thermal tests series. Leak tests have been conducted on those models. It is also common practice to use scaled component tests to investigate the influence of deformations or displacements of the lids and the seals on the standard leakage rate as well as to study the temperature and time depending alteration of the seals. In this paper questions of the transferability of scaled test results to the full size design of the containment system will be discussed.

Assessment of Ductile Cast Iron Fracture Mechanics Analysis Within Licensing of German Transport Packages

2012 ◽  
Author(s):  
Steffen Komann ◽  
Yusuf Kiyak ◽  
Frank Wille ◽  
Uwe Zerbst ◽  
Mike Weber ◽  
...  
Keyword(s):  
Cast Iron ◽  
Fracture Mechanics ◽  
Mechanical Load ◽  
Complex Structure ◽  
Ultrasonic Inspection ◽  
Ductile Cast Iron ◽  
Radioactive Material ◽  
Detailed Model ◽  
Mechanics Analysis ◽  
Accident Conditions

In recent years BAM was involved in several licensing procedures of new package designs for the transport of radioactive material, where the cask body was made of Ductile Cast Iron (DCI). According to IAEA regulations type B(U) packages must withstand the defined accident conditions of transport. For the cask material DCI, it is necessary to determine the brittle fracture behaviour. Due to the complex structure of the cask body and the dynamic loading a fracture mechanical assessment in an analytical way is not always possible. Numerical calculations are necessary to determine the fracture mechanical load in the component. At the first step a numerical analysis has to be done to identify the loading state at the cask body. Secondly an analysis of a detail of the cask body is made considering the boundary conditions of the global model. An artificial flaw is considered in this detailed model to calculate the fracture mechanical loading state. The size of the artificial flaw is characterized by the ultrasonic inspection used for the quality assurance of the package. The applicant developed additional analysis tools for calculation of stress intensity factor and/or J-Integral. The paper describes the authority assessment approach for the DCI fracture mechanics analysis.

Recent Development of Code Case on Use of Ductile Cast Iron for Transport and Storage Cask for Spent Nuclear Fuel

2008 ◽  
Author(s):  
Taku Arai ◽  
Toshiari Saegusa ◽  
Roland Hueggenberg
Keyword(s):  
Fracture Toughness ◽  
Cast Iron ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fatigue Curve ◽  
External Pressure ◽  
Ductile Cast Iron ◽  
Fracture Toughness Test ◽  
Technical Basis ◽  
And Storage

Code Case N-670 “Use of Ductile Cast Iron Conforming to ASTM A874/A 874M-98 or JIS G5504-1992 for Transport Containments, Section III, Division 3” which permits use of ductile cast iron for transport containments of spent nuclear fuel was revised to the Code Case N-670-1, “Use of Ductile Cast Iron Conforming to ASTM A874/A 874M-98 or JIS G5504-2005 for Transport and Storage Containments, Section III, Division 3”. Items revised were as follows: (a) Scope was expanded to use for transport and storage, and changed to conform year edition of JIS G5504, (b) The elongation requirement was deleted form the code case to reflect the change of year edition of JIS G5504, (c) Temperature condition of −40 °C was clearly provided for fracture toughness test, (d) Design fatigue curve was re-established, (e) External pressure chart was re-established. Technical basis of the revised code case are described in this paper.

The Issue of Underground Depositing of High Radioactive Waste

2016 ◽  
Vol 722 ◽  
pp. 59-65
Author(s):  
Markéta Kočová ◽  
Zdeňka Říhová ◽  
Jan Zatloukal
Keyword(s):  
Radioactive Waste ◽  
Nuclear Power ◽  
Power Plants ◽  
High Level Waste ◽  
Spent Fuel ◽  
High Level Radioactive Waste ◽  
Medium Level ◽  
Long Time ◽  
High Level ◽  
Fuel Elements

Nowadays manipulation and depositing of high-level radioactive waste has become the most important issue, which needs to be solved. High-level radioactive waste consists mainly of spent fuel elements from nuclear power plants, which cannot be deposited for long time in surface repositories in the same way as it is possible in case of low and medium level radioactive waste. The most effective and safe solution in longer time horizon seems to be deep geological repository of high level waste. In this process of deposition, large amount of specific conditions needs to be taken into account while designing the whole underground complex, because the materials and structures must fulfil all necessary requirements. Then adequate safety will be ensured.

Numerical Analysis of the Residual Heat Removal for the Spent Fuel Canister of HTR-PM in Fuel Loading Process

2014 ◽  
Author(s):  
Jinhua Wang ◽  
Bing Wang ◽  
Bin Wu
Keyword(s):  
High Temperature ◽  
Natural Ventilation ◽  
Maximum Temperature ◽  
Spent Fuel ◽  
Fuel Loading ◽  
Pebble Bed ◽  
Loading Process ◽  
Accident Conditions ◽  
Fuel Elements ◽  
Residual Heat

High Temperature Gas Cooled Reactor (HTGR) has inherent safety, and has been selected as one of the candidates for the Gen-IV nuclear energy system. In china, the project of the High Temperature Reactor Pebble bed Module (HTR-PM) is in design and construction process. Spherical fuel elements are chosen for the HTR-PM and the spent fuel elements will be stored in canister. The spent fuel canister will be delivered to wells for storage when fully loaded. The canister is covered by a steel cask for radiation shielding, and the cask is covered by a boron polyethylene sleeve to absorb neutrons from decay in fuel loading process. Normally, the residual heat is discharged by forced ventilation in fuel loading process. An auxiliary fan is set on top of the cask considering the possible mechanical failure for the operating fan. When losing normal power supply, the emergency power will be provided to the fans by the two line diesel generators respectively. In extreme conditions of mechanical failure for both fans, the residual heat could be discharged by natural ventilation. The temperature profiles of the different structures were studied in this paper with CFD method for both normal and accident conditions. The calculation results showed that, the maximum temperature of all of the structures are lower than the safety temperature limits in either normal or accident conditions; the temperature decreases rapidly with radial distance in the canister, and the maximum temperature is located at the center of the fuel pebble bed. So it is feasible to remove the residual heat of the spent fuel by natural ventilation in accident condition, and in the natural ventilation condition, the maximum temperature of the spent fuel, the canister shell, the shielding cask, and the boron polyethylene sleeve are lower than their safety temperature limits.

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