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- Metal seal (9)
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- Metal seals (6)
- Leak tightness (5)
- Long-term behavior (5)
- Safety assessment (5)
- Aging mechanisms (3)
- Radioactive material (3)
- Spent fuel (3)
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Organisationseinheit der BAM
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.
In Germany, the concept of dry interim storage in dual purpose metal casks before disposal is being pursued for spent nuclear fuel (SF) and high active waste (HAW) management. However, since there is no repository available today, the initially planned and established dry interim storage license duration of up to 40 years will be too short and its extension will become necessary. For such a storage license extension it is required to assess the long-term performance of SF and all safety related storage system components in order to confirm the viability of extended storage.
The main safety relevant components are the thick-walled dual purpose metal casks. These casks consist of a monolithic cask body with integrated neutron shielding components (polymers, e.g. polyethylene) and a monitored double lid barrier system with metal and elastomeric seals. The metal seals of this bolted closure system guarantee the required leak-tightness whereas the elastomeric seals allow for leakage rate measurement of the metal seals.
This paper presents an update on running long-term tests on metal seals at different temperatures under static conditions over longer periods of time. In addition, first results of our approach to understand the aging behavior of different elastomeric seals and neutron radiation shielding material polyethylene are discussed.