TY - CONF A1 - Rolle, Annette A1 - Buhlemann, Lothar T1 - Quality assurance in production and use of special form radioactive material - focal points in BAM approvals N2 - BAM, as the competent authority for approval of special form radioactive material, attaches great importance to a detailed audit of the required quality assurance programmes for design, manufacture, testing, documentation, use, maintenance and inspection. Applicants have to submit, together with their application documentation, information on general arrangements for quality assurance, as well as on quality assurance in production and in operation. Fields where BAM has often found deficiencies are leak test methods, weld seam quality and the safety level after use. T2 - 14th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Berlin, Germany DA - 2004-09-20 KW - Zulassung KW - Special form KW - Qualitätssicherung PY - 2004 IS - 235 SP - 1(?) EP - 5(?) PB - BAM CY - Berlin AN - OPUS4-11190 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Rolle, Annette A1 - Buhlemann, Lothar T1 - Quality assurance in production and use of special form radioactive material - focal points in BAM approvals N2 - BAM, as the competent authority for approval of special form radioactive material, attaches great importance to a detailed audit of the required quality assurance programmes for design, manufacture, testing, documentation, use, maintenance and inspection. Applicants have to submit, together with their application documentation, information on general arrangements for quality assurance, as well as on quality assurance in production and in operation. Fields where BAM has often found deficiencies are leak test methods, weld seam quality and the safety level after use. PY - 2004 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 15 IS - 2 SP - 101 EP - 103 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-7278 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Rolle, Annette A1 - Droste, Bernhard T1 - Mechanical behaviour of high burn-up SNF under normal and accident transport conditions - present approaches and perspectives N2 - 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. T2 - ASME 2012 Pressure vessels & piping division conference CY - Toronto, Ontario, Canada DA - 15.07.2012 KW - Fuel rods KW - High burn-up KW - Mechanical behaviour KW - Transport conditions PY - 2012 SP - PVP2012-78302, 1-7 PB - American Society of Mechanical Engineers CY - New York AN - OPUS4-26320 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Rolle, Annette A1 - Droste, Bernhard A1 - Dombrowski, H. T1 - Type testing of devices with inserted radioactive sources T2 - 2nd European IRPA Congress on Radiation Protection CY - Paris, France DA - 2006-05-15 KW - Radiation protection KW - Type approval PY - 2006 SP - 1(?) EP - 6(?) PB - IRPA CY - Paris AN - OPUS4-12815 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Rolle, Annette A1 - Droste, Bernhard A1 - Schubert, Sven A1 - Probst, Ulrich A1 - Wille, Frank T1 - Verification of activity release compliance with regulatory limits within spent fuel transport cask assessment N2 - 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. KW - Containment compliance KW - Seals KW - Regulations KW - Activity release PY - 2012 U6 - https://doi.org/10.1179/1746510913Y.0000000012 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 23 IS - 3-4 SP - 149 EP - 152 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-29244 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Rolle, Annette A1 - Droste, Bernhard T1 - Safety during whole life time: important aspect in safety assessment of sealed radioactive sources N2 - Many sealed sources with long halflife isotopes commonly used in industry or medicine have a long working life, up to several decades. Source integrity must be guaranteed in transport and use at any time. On the one hand, safety during the working life has to be ensured by the source design. Its strain has to be tested. On the other hand, source durability depends on the specific operating conditions. BAM as the competent authority in Germany has to assess the suitability of a source design for safe transport and use also for a longer service life for: (a) sources approved as special form radioactive material according to the regulations for the safe transport of radioactive material, (b) sources in approved devices for licence free use according to the Radiation Protection Ordinance, Para 25, and (c) sources with an extended leak test period according to Radiation Protection Ordinance, Para 66. In all these domains BAM has to assess if design and additional arrangements are qualified and guaranteed to prevent a release of radioactive content under the mechanical, chemical and physical operating conditions of the specified working life of a sealed radioactive source. As a result, limits for the duration of validity of the special form status of a source or a type approval of a device are specified and, in many cases, special additional responsibilities for users, such as periodical control and test measurements, have to be specified in approval certificates as binding conditions to satisfy the required safety standards in regulations. This paper presents BAM's experiences and shows which aspects should be considered in assessment of a lifetime limit of sealed sources. KW - Radioactive KW - Sealed sources KW - Aging PY - 2008 U6 - https://doi.org/10.1179/174651008X344458 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 19 IS - 3 SP - 151 EP - 154 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-18741 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Rolle, Annette A1 - Droste, Bernhard T1 - Safety during whole life time - An important aspect in safety assessment of sealed radioactive sources T2 - PATRAM 2007, 15th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Miami, Florida, USA DA - 2007-10-21 PY - 2007 IS - (Abstract # 53) SP - 1 EP - 7 PB - Institute of Nuclear Materials Management AN - OPUS4-18742 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Rolle, Annette A1 - Probst, Ulrich A1 - Ballheimer, Viktor A1 - Müller, Karsten A1 - Winkler, H. P. T1 - Transport condition effects on the sealing system and activity release within spent fuel transport casks T2 - RAMTRANSPORT 2012 - 9th International Conference on the Radioactive Materials Transport and Storage CY - London, UK DA - 2012-05-22 KW - Radioactive material KW - Safety assessment KW - Spent fuel KW - Leak tightness PY - 2012 SP - 1 EP - 9(?) AN - OPUS4-26498 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Rolle, Annette A1 - Neumeyer, Tino A1 - Wille, Frank T1 - Safety assessment of aged metal seals N2 - With our experiments we want to simulate a load situation which can result from a cask vertical drop test. The lid can lift a little for a very short moment and that can perhaps result in a little movement of the seal, a little rotation or a little lateral movement, so that the contact section can change. The aim of the project is to learn more about the sealing efficiency after the seal is compressed again. What we know from former component tests with non-aged seals is, that the leakage rate measured after one or more decompression /compression cycles with seal movement becomes significant higher. And this result is considered for specification of design leakage rates for release calculation in safety cases. In the recent years we all have learned more about the significant influence of temperature and time on seal behavior and so the plan was to do similar experiments with aged seals, just to get an idea about seal behavior and achievable leakage rates. What we did in detail is: We compressed Helicoflex-seals of both design types, with Aluminum and Silver outer jackets, in testflanges and aged them in an oven at a temperature of 125°C for 3 months to produce seal properties comparable with properties after several years use at more typical operating temperatures between 90 and 100°C. After this ageing procedure we opened the flanges completely, moved the seal a little to vary the contact area and compressed the flanges again. What we measured during all compression and decompression cycles was the load, the deformation and -as long as possible- the leakage rates. T2 - Workshop IRSN/BAM CY - Cadarache, France DA - 12.10.2022 KW - Transport KW - Radioactive KW - Seal behavior KW - Ageing PY - 2022 AN - OPUS4-56141 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Linnemann, Konrad A1 - Ballheimer, Viktor A1 - Sterthaus, Jens A1 - Rolle, Annette A1 - Wille, Frank A1 - Vlassopoulos, Efstathios A1 - Papaioannou, Dimitrios T1 - Numerical Simulation of Spent Fuel Segments under Transport Loads N2 - 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. T2 - 17th International High-Level Radioactive Waste Management Conference (IHLRWM 2019) CY - Knoxville, Tennessee, USA DA - 14.04.2019 KW - Spent Nuclear Fuel KW - Finite Element Simulation KW - Transport packages PY - 2019 SN - 978-1-51088-669-8 SP - 1 EP - 7 AN - OPUS4-52046 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -