Filtern
Erscheinungsjahr
Dokumenttyp
- Beitrag zu einem Tagungsband (17)
- Zeitschriftenartikel (3)
- Posterpräsentation (2)
- Vortrag (1)
Schlagworte
- Transport packages (7)
- Leakage rate (4)
- Radioactive material (4)
- Safety assessment (4)
- Spent fuel assessment (4)
- Ageing (3)
- Seals (3)
- Spent fuel (3)
- Aging mechanisms (2)
- Beförderung (2)
Organisationseinheit der BAM
Eingeladener Vortrag
- nein (1)
Transport packages for spent fuel have to meet the International Atomic Energy Agency requirements for different transport conditions. Physical state of spent fuel and fuel rod cladding as well as geometric configuration of fuel assemblies are important inputs for the evaluation of package capabilities under these conditions. In this paper, the mechanical behaviour of high burn-up spent fuel assemblies under transport conditions is analysed 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 (fine cracks or complete breakage), which can cause release of gas, volatiles, fuel particles or fragments have to be properly considered in these assumptions. In view of the complexity of interactions between the fuel rods as well as between fuel assemblies, basket, and cask containment, the exact mechanical analysis of such phenomena is nearly impossible. 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. In this context some practical approaches based on experiences of BAM Federal Institute for Material Research and Testing within safety assessment of packages for transport of spent fuel are discussed.
Admissible limits for activity release from type B(U) packages for spent fuel transport specified in the International Atomic Energy Agency regulations (10-6 A2 h-1 for normal conditions of transport and A2 per week for accidental conditions of transport) have to be kept by an appropriate function of the cask body and its sealing system. Direct measurements of activity release from the transport casks are not feasible. Therefore, the most common method for the specification of leak tightness is to relate the admissible limits of activity release to equivalent standardised leakage rates. Applicable procedure and calculation methods are summarised in the International Standard ISO 12807 and the US standard ANSI N14·5. BAM as the German competent authority for mechanical, thermal and containment assessment of packages liable for approval verifies the activity release compliance with the regulatory limits. Two fundamental aspects in the assessment are the specification of conservative design leakage rates for normal and accidental conditions of transport and the determination of release fractions of radioactive gases, volatiles and particles from spent fuel rods. Design leakage rates identify the efficiency limits of the sealing system under normal and accidental transport conditions and are deduced from tests with real casks, cask models or components. The releasable radioactive content is primarily determined by the fraction of rods developing cladding breaches and the release fractions of radionuclides due to cladding breaches. The influence of higher burn-ups on the failure probability of the rods and on the release fractions are important questions. This paper gives an overview about methodology of activity release calculation and correlated boundary conditions for assessment.
Für die Beförderung von umschlossenen radioaktiven Stoffen (Strahler) können Transporterleichterungen gelten, wenn das Strahlerdesign nachweislich unfallsicher ausgelegt ist und eine Zulassung als radioaktiver Stoff in besonderer Form vorliegt. Die in Deutschland zuständige Behörde für die Prüfung und Zulassung radioaktiver Stoffe in besonderer Form ist die Bundesanstalt für Materialforschung und –prüfung (BAM).
Eine oft langfristige Nutzung der Strahler kann eine alterungsbedingte Schwächung der Umschließung des radioaktiven Stoffes zur Folge haben. Jeder Strahler einer zugelassenen Bauart muss jedoch im Beförderungsfall zu jeder Zeit, auch nach längerer Nutzung, den vorgeschriebenen Prüfanforderungen genügen. Die Alterungsbewertung auf der Grundlage einer vom Antragsteller zu spezifizierenden Nutzungsdauer des radioaktiven Stoffes in besonderer Form ist seit langer Zeit Bestandteil des Zulassungsverfahrens in Deutschland. Ein von Deutschland eingebrachter Vorschlag im Rahmen der anstehenden Revision der IAEA Empfehlungen für die sichere Beförderung von radioaktiven Stoffen (SSR-6) soll die Lesbarkeit des Regelwerkes bezüglich der Alterungsbewertung verbessern, Klarheit über die damit verbundenen Anforderungen erzeugen und damit zu einer Harmonisierung der Zulassungsverfahren auf internationaler Ebene beitragen. In diesem Beitrag sollen wichtige Einflussfaktoren auf das Alterungsverhalten von radioaktiven Stoffen in besonderer Form aufgezeigt und die Notwendigkeit für die Spezifikation einer Nutzungsdauer als Grundlage für die Alterungsbewertung begründet werden. Der deutsche Vorschlag für die Revision des IAEA Regelwerks zum sicheren Transport radioaktiver Stoffe (SSR-6) wird vorgestellt und erläutert.
Packages for the transport of spent nuclear fuel shall meet the International Atomic Energy Agency regulations to ensure safety under different transport conditions. The physical state of spent fuel and the fuel rod cladding as well as the geometric configuration of fuel assemblies are important inputs for the evaluation of package capabilities under these conditions. Generally, the mechanical behavior of high burn-up spent fuel assemblies under transport conditions shall be analyzed with regard to the assumptions which are used in the containment and criticality safety analysis. Considering the complexity of the interactions between the fuel rods as well as between the fuel assemblies, basket, and cask containment, the exact mechanical analysis of such phenomena is nearly impossible. The gaps in Information concerning the material properties of cladding and pellet behavior, especially for the high burn-up fuel, make the analysis more complicated additionally. As a result, enveloping analytical approaches are usually used by BAM within the safety assessment of packages approved for transport of spent nuclear fuel.
To justify the safety margins of such approaches additional analyses are necessary. In this paper, numerical simulations of a spent fuel assembly Segment are presented. The segment modeled represents the part of a generalized BWR fuel assembly between two spacers.
Dynamic and quasi-static finite element calculations are performed to simulate the spent fuel behavior under regulatory defined accident conditions of transport. Beam elements are used for the modeling of the fuel rods representing the compound consisting of claddings and fuel pellets. The dynamic load applied is gathered from an experimental drop test with a spent fuel cask performed at BAM. A hot cell bending test performed at JRC Karlsruhe is the basis for obtaining the material behavior of the fuel rods. The material properties are determined by simulating the test setup of JRC and optimizing the results to fit the experimental load deflection curve. The simulations of the fuel assembly segment are used to get a better understanding about the loads on fuel rods under accident conditions of transport.
Transport packages for spent nuclear fuel have to be 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 BAM for the authority assessment of spent fuel behavior within the package design approval procedure of German package designs. In particular, cracks or failures in the fuel rod cladding can occur under regulatory transport conditions. These defects can cause the release of gas, volatiles, fuel particles or fragments into the package cavity and have to be considered properly in the safety analysis. Another issue is the transport of defective fuel rods. One concept is to use special canisters which can be handled like fuel assemblies. This concept requires additional assessment concerning drying, sealing and the mechanical and thermal design of such canisters. 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 simplified approaches to consider conservatively spent fuel behavior currently accepted by BAM are presented here.
Verification of activity release compliance with regulatory limits within spent fuel transport casks
(2010)
When storage of spent nuclear fuel or high level waste is carried out 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 (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. Consideration of these issues in the DPC transport safety case will be addressed. Special attention is given to all cask components that cannot be directly inspected or changed without opening the cask cavity, like 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 the 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 license holders and in appropriate design approval update processes. The paper addresses issues that are subject of an actual International Atomic Energy Agency 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.
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'.