Surveillance of Quality Assurance in the Manufacture of Transport and Storage Casks for Spent Fuel and High-level Waste

2005 ◽  
Vol 16 (1) ◽  
pp. 43-48
Author(s):  
M. Baden ◽  
T. Nitz
Keyword(s):  
Quality Assurance ◽  
High Level Waste ◽  
Spent Fuel ◽  
High Level ◽  
And Storage

Canister Sealing for High Level Waste Encapsulation

2001 ◽  
Author(s):  
Richard E. Andrews
Keyword(s):  
Electron Beam Welding ◽  
Deep Level ◽  
High Level Waste ◽  
Spent Fuel ◽  
Reduced Pressure ◽  
Friction Stir ◽  
The Uk ◽  
High Level ◽  
Welding Processes ◽  
And Storage

Abstract Sweden has chosen to manage spent fuel rods by direct encapsulation and storage in a deep level repository. Two welding processes are being investigated for the sealing of copper vessels that form the outer barrier of the disposal canisters. TWI Ltd in the UK has developed Reduced Pressure Electron Beam Welding and Friction Stir Welding for 50mm thick copper. This paper describes some of the investigations and compares the techniques. Over the past 3 years a full-size canister welding machine has been designed and built. Specialised tools have been developed for the welding of thick sections in copper with very encouraging results.

Developing a Standard Code for Spent Fuel and High Level Waste Containment Baskets

2008 ◽  
Author(s):  
Donald Wayne Lewis
Keyword(s):  
Radioactive Material ◽  
High Level Waste ◽  
Plant Design ◽  
Support Structures ◽  
Spent Fuel ◽  
Waste Containment ◽  
Internal Support ◽  
Division 1 ◽  
High Level ◽  
And Storage

ASME Section III, Division 3, “Containments for Transportation and Storage of Spent Nuclear Fuel and High Level Radioactive Material and Waste” currently addresses the design of transportation and storage containment shells but it has yet to address the containment internal support structure that holds the spent fuel or high level waste in place. However, the code for internal support structures, hereafter referred to by its common name “basket”, has been under development by ASME for the past 2 years. Development of the new code, to be known as Subsection WD, “Internal Support Structures” was deemed necessary because current containment system basket construction is a piecemeal approach using ASME Section III, Division 1, Subsection NF, “Supports” and/or ASME Section III, Division 1, Subsection NG, “Core Support Structures” or some other engineering method. Approvals for the various combinations are granted from the regulatory authority. The piecemeal approach tries to capture the critical elements important for a containment basket. However, Subsections NF and NG are based on nuclear power plant design which has different design goals than for a spent fuel or high level waste containment. The issuance of Subsection WD will ensure standardization of future containment baskets, assist the regulatory agency in the review and approval of the baskets, and ensure that the essential criteria in the basket related to spent fuel and high level waste storage transportation and disposal is adequately addressed. The purpose of the basket is primarily to ensure that the radioactive components in the containment are supported in a way as not to create a criticality event. Current acceptance is typically based on a no yield design that the containment manufactures all say is too conservative and based on unreasonable criteria. What should the basket design be based on, how should Subsection WD address them, etc.? The purpose of this paper is to inform interested parties of the progress that has been made in development of Subsection WD, what construction provisions it will initially include and what is planned for it, and when is it scheduled to be issued.

TRU Recycling Options for Environmentally Friendly and Proliferation-Resistant Nuclear Fuel Cycle of FBR

2010 ◽  
Author(s):  
Sidik Permana ◽  
Mitsutoshi Suzuki
Keyword(s):  
High Performance ◽  
Fuel Cycle ◽  
Operation Time ◽  
Nuclear Nonproliferation ◽  
High Level Waste ◽  
Minor Actinide ◽  
Spent Fuel ◽  
Closed Fuel Cycle ◽  
The Core ◽  
High Level

The embodied challenges for introducing closed fuel cycle are utilizing advanced fuel reprocessing and fabrication facilities as well as nuclear nonproliferation aspect. Optimization target of advanced reactor design should be maintained properly to obtain high performance of safety, fuel breeding and reducing some long-lived and high level radioactivity of spent fuel by closed fuel cycle options. In this paper, the contribution of loading trans-uranium to the core performance, fuel production, and reduction of minor actinide in high level waste (HLW) have been investigated during reactor operation of large fast breeder reactor (FBR). Excess reactivity can be reduced by loading some minor actinide in the core which affect to the increase of fuel breeding capability, however, some small reduction values of breeding capability are obtained when minor actinides are loaded in the blanket regions. As a total composition, MA compositions are reduced by increasing operation time. Relatively smaller reduction value was obtained at end of operation by blanket regions (9%) than core regions (15%). In addition, adopting closed cycle of MA obtains better intrinsic aspect of nuclear nonproliferation based on the increase of even mass plutonium in the isotopic plutonium composition.

scholarly journals Spent Fuel as a Waste Form - Data Needs to Allow Long Term Performance Assessment Under Repository Disposal Conditions.

1986 ◽  
Vol 84 ◽  
Author(s):  
V. M. Oversby
Keyword(s):  
Performance Assessment ◽  
Waste Form ◽  
High Level Waste ◽  
Detailed Knowledge ◽  
Spent Fuel ◽  
Expected Performance ◽  
High Level ◽  
Term Performance ◽  
Physical And Chemical ◽  
Waste Repositories

AbstractPerformance assessment calculations are required for high level waste repositories for a period of 10,000 years under NRC and EPA regulations. In addition, the Siting Guidelines (IOCFR960) require a comparison of sites following site characterization and prior to final site selection to be made over a 100,000 year period. In order to perform the required calculations, a detailed knowledge of the physical and chemical processes that affect waste form performance will be needed for each site. While bounding calculations might be sufficient to show compliance with the requirements of IOCFR60 and 40CFRI91, the site comparison for 100,000 years will need to be based on expected performance under site specific conditions. The only case where detailed knowledge of waste form characteristics in the repository would not be needed would be where radionuclide travel times to the accessible environment can be shown to exceed 100,000 years. This paper will review the factors that affect the release of radionuclides from spemt fuel under repository conditions, summarize our present state of knowledge, and suggest areas where more work is needed in order to support the performance assessment calculations.

Assessment of Redox Conditions in the Near Field of Nuclear Waste Repositories: Application to the Swiss high-level and intermediate level waste disposal concept

2003 ◽  
Vol 807 ◽  
Author(s):  
Paul Wersin ◽  
Lawrence H. Johnson ◽  
Bernhard Schwyn
Keyword(s):  
Solid Phase ◽  
Near Field ◽  
Intermediate Level ◽  
Redox Conditions ◽  
High Level Waste ◽  
Redox Potentials ◽  
Spent Fuel ◽  
Reducing Conditions ◽  
High Level ◽  
Waste Repositories

ABSTRACTRedox conditions were assessed for a spent fuel and high-level waste (SF/HLW) and an intermediate-level waste (ILW) repository. For both cases our analysis indicates permanently reducing conditions after a relatively short oxic period. The canister-bentonite near field in the HLW case displays a high redox buffering capacity because of expected high activity of dissolved and surface-bound Fe(II). This is contrary to the cementitious near field in the ILW case where concentrations of dissolved reduced species are low and redox reactions occur primarily via solid phase transformation processes.For the bentonite-canister near field, redox potentials of about -100 to -300 mV (SHE) are estimated, which is supported by recent kinetic data on U, Tc and Se interaction with reduced iron systems. For the cementitious near field, redox potentials of about -200 to -800 mV are estimated, which reflects the large uncertainties related to this alkaline environment.

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.

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