TY - CONF A1 - Komann, Steffen A1 - Gröke, Carsten A1 - Droste, Bernhard T1 - Assessment of quality assurance measures for radioactive material transport packages not requiring competent authority design approval N2 - The majority of transports of radioactive materials are carried out in packages which don’t need a package design approval by a competent authority. Low-active radioactive materials are transported in such packages e.g. in the medical and pharmaceutical industry and in the nuclear industry as well. Decommissioning of NPP’s leads to a strong demand for packages to transport low and middle active radioactive waste. According to IAEA regulations the “non-competent authority approved package types” are the Excepted Packages and the Industrial Packages of Type IP-1, IP-2 and IP-3 and packages of Type A. For these types of packages an assessment by the competent authority is required for the quality assurance measures for the design, manufacture, testing, documentation, use, maintenance and inspection (IAEA SSR 6, § 306). In general a compliance audit of the manufacturer of the packaging is required during this assessment procedure. Their regulatory level in the IAEA regulations is not comparable with the “regulatory density” for packages requiring competent authority package design approval. Practices in different countries lead to different approaches within the assessment of the quality assurance measures in the management system as well as in the quality assurance program of a special package design. To use the package or packaging in a safe manner and in compliance with the regulations a management system for each phase of the life of the package or packaging is necessary. The relevant IAEA-SSR6 § 801 requires documentary verification by the consignor concerning package compliance with the requirements. T2 - WM2013 Conference CY - Phoenix, Arizona, USA DA - 24.02.2013 PY - 2013 SN - 978-0-9836186-2-1 SP - 1 EP - 8 CY - Phoenix, Arizona, USA AN - OPUS4-28504 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Droste, Bernhard A1 - Komann, Steffen A1 - Wille, Frank A1 - Rolle, Annette A1 - Probst, Ulrich A1 - Schubert, Sven T1 - Consideration of aging mechanism influence on transport safety of dual purpose casks for spent nuclear fuel of HLW N2 - 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'. KW - Transport KW - Storage KW - Spent fuel KW - High-level waste KW - Aging KW - Metal seals KW - Transport and storage casks KW - Spent nuclear fuel KW - Aging mechanisms KW - Corrosion KW - Safety assessment PY - 2014 U6 - https://doi.org/10.1179/1746510914Y.0000000070 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 25 IS - 3-4 SP - 105 EP - 112 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-33428 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Droste, Bernhard A1 - Komann, Steffen A1 - Wille, Frank A1 - Rolle, Annette A1 - Probst, Ulrich A1 - Schubert, Sven T1 - Considerations of aging mechanisms influence on transport safety and reliability of dual purpose casks for spent nuclear fuel or HLW N2 - 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'. T2 - PSAM 12 - Probabilistic safety assessment and management CY - Honolulu, Hawaii, USA DA - 22.06.2014 KW - Transport and storage casks for spent nuclear fuel or high level waste KW - Aging mechanisms KW - Corrosion KW - Safety assessment KW - Metal seals KW - Closure system KW - Spent fuel/high-level waste KW - Dual purpose casks KW - Metal seals reliability KW - Cesium corrosion PY - 2014 UR - http://psam12.org/proceedings/paper/paper_180_1.pdf SP - 1 EP - 10 AN - OPUS4-32518 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Droste, Bernhard A1 - Koch, Frank A1 - Komann, Steffen T1 - Design assessment of spent fuel and HLW Transport Casks T2 - 49th INMM Annual Meeting, July 13-17, 2008 CY - Nashville, USA DA - 2008-07-13 KW - Safety Analysis KW - Transport Packages KW - Drop tests KW - Finite Element Analysis KW - Thermal Design KW - Containment KW - Quality Management PY - 2008 SP - 1 EP - 8 AN - OPUS4-17780 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Wille, Frank T1 - Dicht und sicher abgeschirmt N2 - BAM-GGR Transportverpackungen für radioaktive Stoffe müssen den BAM-Gefahrgutregeln entsprechen. Ein kurzer Einblick in die Richtlinien. PY - 2014 SN - 0944-6117 VL - 3 SP - 18 EP - 20 PB - Vogel CY - München AN - OPUS4-30487 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Musolff, André A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Müller, Karsten A1 - Quercetti, Thomas T1 - Drop test program with a HLW cask model - performance, measurements and results N2 - BAM (Federal Institute for Material Research and Testing) is the competent authority for mechanical and thermal safety assessment of transport packages for spent fuel and high level waste (HLW) in Germany. In context with package design approval of the new German HLW cask CASTOR HAW28M, BAM performed several drop tests with a half-scale model of the CASTOR HAW/TB2. The test model was manufactured by GNS (Gesellschaft fur Nuklear Service mbH) and tested under accident transport conditions on the 200 tons BAM drop test facility at the BAM Test Site Technical Safety. For this comprehensive test program the test specimen CASTOR HAW/TB2 was instrumented at 21 measurement planes with altogether 23 piezo-resistive accelerometers, five temperature sensors and 131 tri-axial strain gauges in the container interior and exterior, respectively. The strains of four representative lid bolts were recorded by four uniaxial strain gauges per each bolt. Helium leakage rate measurements were performed before and after each test in the above noted testing sequence. The paper presents some experimental results of the half-scale CASTOR HAW/TB2 prototype (14,500 kg) and measurement data logging. T2 - 52nd INMM Annual meeting CY - Palm Desert, CA, USA DA - 17.07.2011 KW - Drop test results KW - Spent fuel transport cask KW - Impact limiter KW - Measurement methods KW - Drop test program PY - 2011 SP - 1 EP - 9 AN - OPUS4-24245 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Musolff, André A1 - Quercetti, Thomas A1 - Müller, Karsten A1 - Droste, Bernhard A1 - Komann, Steffen T1 - Drop test program with half scale model CASTOR HAW/TB2 N2 - Federal Institute for Materials Research and Testing (BAM) is the competent authority for mechanical and thermal safety assessment of transport packages for spent fuel and high level waste in Germany. In context with package design approval of the new German high level waste cask CASTOR® HAW28M, BAM performed several drop tests with a half scale model of the CASTOR® HAW/TB2. The cask is manufactured by Gesellschaft für Nuklear Service mbH and was tested under accident transport conditions on the 200 tons BAM drop test facility at the BAM Test Site Technical Safety. For this comprehensive test program, the test specimen CASTOR® HAW/TB2 was instrumented at 21 measurement planes with altogether 23 piezo resistive accelerometers, five temperature sensors and 131 triaxial strain gauges in the container interior and exterior respectively. The strains of four representative lid bolts were recorded by four uniaxial strain gauges per each bolt. Helium leakage rate measurements were performed before and after each test in the above noted testing sequence. The paper presents some experimental results of the half scale CASTOR® HAW/TB2 prototype (14 500 kg) and measurement data logging. It illustrates the extensive instrumentation and analyses that are used by BAM for evaluating the cask performance to the mechanical tests required by regulations. Although some of the quantitative deceleration, velocity and strain values cannot be shown because of confidentially issues, they are provided qualitatively to illustrate the types of measurements and methodologies used at BAM. KW - IAEA drop testing KW - Half scale model KW - Experimental test PY - 2011 U6 - https://doi.org/10.1179/1746510911Y.0000000013 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 22 IS - 3 SP - 154 EP - 160 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-25293 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Musolff, André A1 - Quercetti, Thomas A1 - Müller, Karsten A1 - Droste, Bernhard A1 - Komann, Steffen T1 - Drop test program with the half-scale model CASTOR HAW/TB2 T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - Spent fuel transport cask KW - Drop test KW - Impact limiter KW - Measurement procedure PY - 2010 SP - 1 EP - 8 AN - OPUS4-22771 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Wille, Frank A1 - Droste, Bernhard T1 - Gap analysis by periodical reviews of transport package design safety reports of German SNF/HLW dual purpose casks N2 - Storage of spent nuclear fuel and high-level waste in dual purpose casks (DPC) is related with the challenge of maintaining safety for transportation over several decades of storage. Beside consideration of aging mechanisms by appropriate design, material selection and operational controls to assure technical reliability by aging management measures, an essential issue is the continuous control and update of the DPC safety case. Not only the technical objects are subject of aging but also the safety demonstration basis is subject of “aging” due to possible changes of regulations, standards and scientific/technical knowledge. The basic document, defining the transport safety conditions, is the package design safety report (PDSR) for the transport version of the DPC. To ensure a safe transport in future to a destination which is not known yet (because of not yet existing repository sites) periodical reviews of the PDSR, in connection with periodic renewals of package design approval certificates, have to be carried out. The main reviewing tool is a gap analysis. A gap analysis for a PDSR is the assessment of the state of technical knowledge, standards and regulations regarding safety functions of structures, systems and components. T2 - 18th International Symposium on the Packaging and Transportation of Radioactive Materials PATRAM2016 CY - Kobe, Japan DA - 18.09.2016 KW - Radioactive material KW - Gap analysis KW - Periodical review KW - Approved packages PY - 2016 SP - Paper 5004, 1 EP - 9 AN - OPUS4-37641 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Wille, Frank A1 - Droste, Bernhard T1 - Gap analysis examples from periodical reviews of transport package design safety reports of SNF/HLW dual purpose casks N2 - Storage of spent nuclear fuel and high-level waste in dual purpose casks (DPC) is related with the challenge of maintaining safety for transportation over several decades of storage. Beside consideration of aging mechanisms by appropriate design, material selection and operational controls to assure technical reliability by aging management measures, an essential issue is the continuous control and update of the DPC safety case. Not only the technical objects are subject of aging but also the safety demonstration basis is subject of "aging" due to possible changes of regulations, standards and scientific/technical knowledge. The basic document, defining the transport safety conditions, is the package design safety report (PDSR) for the transport version of the DPC. To ensure a safe transport in future to a destination which is not known yet (because of not yet existing repository sites) periodical reviews of the PDSR, in connection with periodic renewals of package design approval certificates, have to be carried out. The main reviewing tool is a gap analysis. A gap analysis for a PDSR is the assessment of the state of technical knowledge, standards and regulations regarding safety functions of structures, systems and components. T2 - PSAM 12 - Probabilistic safety assessment and management CY - Honolulu, Hawaii, USA DA - 22.06.2014 KW - Dual Purpose Casks KW - Aging KW - Transportation KW - Periodical Review PY - 2014 UR - http://psam12.org/proceedings/paper/paper_259_1.pdf SP - Paper 259, 1 EP - 9 AN - OPUS4-32574 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -