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- Radioactive material (6)
- Transport packages (6)
- Safety assessment (5)
- Leakage rate (4)
- Spent fuel assessment (4)
- Spent fuel (3)
- Ageing (2)
- Finite element methods (2)
- Gas generation (2)
- High burn-up (2)
Organisationseinheit der BAM
Packages for intermediate level waste (ILW) often contain residual water besides the actual waste. The water either exists as obvious free water or it may be bound physically or chemically, e.g. as pore water.
Water driven gas generation could occur by vaporisation and by radiolysis.
Steam as the result of vaporisation causes an increasing pressure inside a package and can affect corrosion. Vaporisation and condensation processes itself change the thermal behaviour of the content especially during strongly unsteady thermal situations like accident fire situations.
Radiolysis changes the chemical composition of the content which could cause an unexpected interaction, e.g. hydrogen embrittlement. Besides the pressure build-up the radiolysis of water generates hydrogen and oxygen, which can be highly flammable respectively explosive.
The gas generation caused by vaporisation and radiolysis must be taken into account during the design and the safety assessment of a package. Pressure build-up, a changed thermal behaviour and content chemistry, and especially the risk of accumulation of combustible gases exceeding the limiting concentration for inflammability has to be considered in the safety assessment. Approaches to ensure the transportability of stored packages due to radiolysis will be discussed.
When storage of spent nuclear fuel (SNF) or high-level waste (HLW) is done in dual purpose casks (DPC), the effects of aging on safety relevant DPC functions and properties have to be managed in a way that a safe transport after the storage period of several decades is capable, and can be justified and certified permanently throughout that period. The effects of aging mechanisms (like e.g. radiation, different corrosion mechanisms, stress relaxation, creep, structural changes and degradation) on the transport package design safety assessment features have to be evaluated. The consideration of these issues in the DPC transport safety case will be addressed. Special attention is given to all cask components which cannot be directly inspected or changed without opening the cask cavity, what are the inner parts of the closure system and the cask internals, like baskets or spent fuel assemblies. The design criteria of that transport safety case have to consider the operational impacts during storage. Aging is not subject of technical aspects only, but also of 'intellectual' aspects, like changing standards, scientific/ technical knowledge development and personal as well as institutional alterations. Those aspects are to be considered in the management system of the license holders and in appropriate design approval update processes. The paper addresses issues which are subject of an actual IAEA TECDOC draft 'Preparation of a safety case for a dual purpose cask containing spent nuclear fuel'.
Packages for intermediate level waste (ILW) often contain residual water besides the actual waste. The water either exists as obvious free water or it may be bound physically or chemically, e.g. as pore water. A water driven gas generation could occur by vaporisation and by radiolysis. Steam as the result of vaporisation causes an increasing pressure inside a package and can affect corrosion. Vaporisation and condensation processes itself change the thermal behavior of the content especially during strongly unsteady thermal situations like accident fire situations. Radiolysis changes the chemical composition of the content which could cause an unexpected interaction, e.g. hydrogen embrittlement. Besides the pressure build-up the radiolysis of water generates hydrogen and oxygen, which can be highly flammable respectively explosive. The gas generation caused by vaporisation and radiolysis must be taken into account during the design and the safety assessment of a package. Pressure build-up, a changed thermal behavior and content chemistry, and especially the risk of accumulation of combustible gases exceeding the limiting concentration for flammability has to be considered in the safety assessment. Approaches to ensure the transportability of stored packages due to radiolysis will be discussed.
Transport packages for spent nuclear fuel have to be assessed with respect to specific transport conditions which are defined in the regulations of the International Atomic Energy Agency. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important Inputs for the evaluation of the package capabilities under These conditions. Cracks or failures in the fuel rod cladding can cause the release of gas, volatiles or fuel particles into the cavity. The amount of substances in the cavity has to be considered in the assessment of the activity release and criticality safety.
The mechanical analysis of the compound system formed by the fuel rod cladding and the spent fuel pellets is very difficult due to the limited knowledge of the material properties and the insufficient understanding of the interaction between pellets and cladding and between adjacent pellets. The variation of fuel assembly properties regarding cladding material, burn-up and the history of usage makes reliable predictions of the fuel rod behavior even harder.
For a better understanding about the behavior of spent fuel rods, JRC and BAM have started a joint research project. In this context, JRC has developed a test device which allows quasi-static 3-point-bending test on fuel rod segments in the hot cell. The loads are applied with respect to the boundary conditions of the activity release assessment. This paper deals with the numerical calculation of a single fuel rod segment under bending load. The aim is to identify the governing mechanical parameters by the variation of constitutive assumptions, contact conditions, inner constraints, etc. This knowledge helps for the interpretation of the experimental results. Furthermore, the improved understanding about the behavior of the cladding-pellets system will be beneficial for the assessment of spent fuel transport conditions.
The containment systems of transport and storage casks for spent fuel and highlevelradioactive 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.
In accordance with the IAEA transport regulations Special Form Radioactive Material (SFRM) is either an indispersible solid radioactive material or a sealed capsule containing radioactive material.
The design of special form radioactive material has to resist a severe transport accident without undue loss or dispersal of radioactive material. The safety assessment for design approval shall include besides the required test program (impact, percussion, bending and heat test) also the evaluation of the management system for design, manufacture, testing, documentation, use, maintenance and inspection. The specified quality assurance measures have to assure, that every specimen of the approved design is produced in the same verified quality and every specimen must be able to survive the severe mechanical and thermal tests at any time of its working life.
Due to the long-term use of SFRM the consideration of ageing is an important aspect in the approval procedure by BAM, the competent authority for SFRM approval in Germany. Different fields of application imply a wide range of environmental conditions, from clean room atmosphere to highly aggressive industrial conditions. So, besides of radioactive content, corrosion is a main factor for possible design degradation.
This paper will describe major influencing factors to be taken into account to assess the ageing behavior of a SFRM design and will emphasize that there is a need for a regulatory specification of a SFRM-working life as basis for the aging evaluations.
Transport packages for spent nuclear fuel have to be assessed with respect to specific transport conditions which are defined in the regulations of the International Atomic Energy Agency. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important inputs for the evaluation of the package capabilities under these conditions. Cracks or failures in the fuel rod cladding can cause the release of gas, volatiles or fuel particles into the cavity. The amount of substances in the cavity has to be considered in the assessment of the activity release and criticality safety.
The mechanical analysis of the compound system formed by the fuel rod cladding and the spent fuel pellets is very difficult due to the limited knowledge of the material properties and the insufficient understanding of the interaction between pellets and cladding and between adjacent pellets. The variation of fuel assembly properties regarding cladding material, burn-up and the history of usage makes reliable predictions of the fuel rod behavior even harder.
For a better understanding about the behavior of spent fuel rods, JRC-ITU and BAM have started a joint research project. In this context, JRC-ITU has developed a test device which allows quasi-static 3-point-bending test on fuel rod segments in the hot cell. The loads are applied with respect to the boundary conditions of the activity release assessment. This paper deals with the numerical calculation of a single fuel rod segment under bending load. The aim is to identify the governing mechanical parameters by the variation of constitutive assumptions, contact conditions, inner constraints, etc. This knowledge helps for the interpretation of the experimental results. Furthermore, the improved understanding about the behavior of the cladding-pellets system will be beneficial for the assessment of spent fuel transport conditions.
This paper presents data obtained from experiments performed using a bending test set-up developed at the Joint Research Centre (JRC) – Karlsruhe, for spent fuel segment testing. Adjustable sample holders, loading modes and other experimental conditions can be im- plemented in the experiments to study the effects of different deformation ranges up to cladding failure. The experimental set-up has been adapted to hot cell remote controlling and has a modular configuration, which allows manual and motor-driven loading option. The device has been calibrated on hydrogenated, unirradiated cladding tube segments filled with alumina pellets. The final application of present set-up is to test non-defueled spent fuel rod segments, pressurized to the original spent fuel rod pressure level. The range of applicability of this device, the scope of the experimental program and the first results from actual bending tests will be discussed.
German packages for the transport of spent nuclear fuel are assessed with respect to specific transport conditions which are defined in the safety regulations of the International Atomic Energy Agency.
In general, gastight fuel rods constitute the first barrier of the containment system. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important inputs for the evaluation of the package safety under transport conditions. The objective of this paper is to discuss the methodologies accepted by German authority BAM for the evaluation of spent fuel behavior within the package design approval procedure.
Specific test conditions will be analyzed with regard to assumptions to be used in the activity release and criticality safety analysis. In particular the different failure modes of the fuel rods, which can cause release of gas, volatiles, fuel particles or fragments, have to be properly considered in these assumptions.
The package as a mechanical system is characterized by a complex set of interactions, e.g. between the fuel rods within the assembly as well as between the fuel assemblies, the basket, and the cask containment. This complexity together with the limited knowledge about the material properties and the variation of the fuel assemblies regarding cladding material, burn-up and the operation history makes an exact mechanical analysis of the fuel rods nearly impossible.
The application of sophisticated numerical models requires extensive experimental data for model verification, which are in general not available. The gaps in information concerning the material properties of cladding and pellets, especially for the high burn-up fuel, make the analysis more complicated additionally, and require a conservative approach.
In this context some practical approaches based on experiences by BAM within safety assessment of packages for transport of spent fuel will be discussed.
Ongoing research activities to investigate SNF mechanical behavior in view of gas and fissile material release under transport loads are presented.