The Experts below are selected from a list of 303 Experts worldwide ranked by ideXlab platform
Frank Wille - One of the best experts on this subject based on the ideXlab platform.
-
Requirements for management systems for manufacturing of transport packages: the new revision of BAM-GGR 011 guideline
2020Co-Authors: Martin Neumann, O. Schilling, Christian Kuschke, A. Darnstädt, Sven Schubert, U. Günther, Frank WilleAbstract:In accordance with IAEA SSR-6 para 306 a management system shall be established and implemented to ensure compliance with the relevant provisions of the IAEA regulations. BAM has issued an update of the guideline: the BAM-GGR 011. The new revision describes necessary quality assurance measures for Design, manufacture, testing, documentation, use, maintenance and inspection of packagings for package Designs requiring competent authority Approval for the transport of radioactive material. The measures can be categorised as system-related and Design-related. They are independently approved and monitored by the German competent authority BAM and its authorised expert (BAM/T). The qualification of the organisation applying for the Design Approval certificate is reviewed in the context of the Design Approval procedure. The quality assurance measures for manufacture consist of three main steps. Pre-assessment of manufacturing documents such as quality plans, specifications etc., Manufacturing inspections according the pre-assessed documents and inspection before commissioning including documentation review. Periodic inspections during operation as well as relevant specifications for use and maintenance ensure that the properties specified in the Approval certificate are preserved over the package life time. Special provisions for the return on experience regarding operational feedback for Design, manufacture, use, maintenance and inspection are given. Special focus shall be given here to the rearranged and meanwhile established system of manufacturing inspections. This includes more transparent roles for a) the Producers authorised inspection 11282 representative, b) the independent inspection expert (S), acting on behalf of the manufacturer with acceptance of BAM, and c) BAM or its authorised expert (BAM/T). Additional attention shall be drawn to the management of deviations during manufacturing and provisions for maintenance and periodic inspections.
-
MECHANICAL/THERMAL PACKAGE Design SAFETY ASSESSMENT AND MANUFACTURING QUALITY ASSURANCE OF SPENT FUEL TRANSPORT CASK NCS 45
2020Co-Authors: Christian Kuschke, Tino Neumeyer, Frank Wille, Bernhard DrosteAbstract:The NCS 45 package is Designed for a variety of radioactive contents ranging from irradiated PWR, BWR and FBR fuel rods, parts of fuel rods as well as non-fissile radioactive material. BAM assessed the applicants safety analysis concept and the appropriate report with respect of requirements according to transport regulations under routine, normal and accident transport conditions related to mechanical Design and thermal Design, activity release and quality assurance aspects. The assesment concept of the safety report was a combination of drop test program, analytic and numerical calculations as well as various components tests. The paper will give details of BAM activities in the official safety assessment that had been summarized in the BAM Design examination certificate which was a prerequisite for the Package Design Approval Certificate issued by the German competent authority.
-
authority experience during Design Approval procedure for packages loaded with special encapsulations for damaged spent nuclear fuel
2019Co-Authors: Konrad Linnemann, Steffen Komann, Lars Muller, Thorsten Schonfelder, Frank WilleAbstract:The first German package Design Approval certificate for a dual purpose cask intended for loading with damaged spent nuclear fuel was issued recently. BAM as part of the competent authority system in Germany carried out a comprehensive assessment procedure with respect to the mechanical and thermal Design, the release of radioactive material and the quality assurance aspects of manufacturing and operation. Packages for the transport and storage of radioactive material have been assessed by BAM for many years, thus the common assessment procedure is well-known and good practice. Up to now only SNF without defects or HLW with well-defined properties were Designated for long-term Interim storage and transports afterwards. Due to Germany’s nuclear phase out all other kinds of spent nuclear fuel in particular damaged spent nuclear fuel shall be packed as well. Damaged spent nuclear fuel needs a tight closure with Special encapsulations and clearly defined properties in Germany. In addition, these encapsulations shall be long-term durable, because they are not accessible after loading in a packaging within periodical inspections. The main difference to Standard package components is that encapsulations with a permanent closure achieve their specified conditions not after manufacturing but only during operation, after loading and closing. To ensure compliance with the specific conditions, special measures for quality assurance are necessary during operation of each encapsulation, e.g. drying and sealing, which were assessed by BAM. The present paper gives an overview of the conducted assessment from BAM and point out the findings concerning to the special closure lid of the approved encapsulation, which is screwed and welded. A wide verification concept is necessary to show the specific tightness under transport conditions. Together with quality assurance measures during first operation steps these encapsulations with damaged spent nuclear fuel can be handled like standard fuel assemblies in approved package Designs.
-
assessment of quality management for transport packages not requiring authority Design Approval
2018Co-Authors: Steffen Komann, Carsten Groke, Frank WilleAbstract:The majority of transports of radioactive materials are carried out in packages which don’t need a package Design Approval of a competent authority. Low active radioactive materials are transported in such kind of packages e.g. in the medical and pharmaceutical industry and in the nuclear industry as well. In Germany the decision to phase out nuclear energy leads to a strong demand for packages to transport low and middle active radioactive waste due to the dismantling and decommissioning of nuclear power plants. 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 of Type A. For the packages of Type IP-2, IP-3 and Type A an assessment by the German competent authority is required for the quality management for the Design, manufacture, testing, documentation, use, maintenance and inspection. In general a compliance audit of the manufacturer of the packaging is required during this assessment procedure.
-
Assessment of Ductile Cast Iron Fracture Mechanics Analysis Within Licensing of German Transport Packages
Volume 3: Design and Analysis, 2012Co-Authors: Steffen Komann, Frank Wille, Yusuf Kiyak, Uwe Zerbst, Mike Weber, Dietmar KlingbeilAbstract:In the Design Approval of transport packages for radioactive materials, the mechanical and thermal safety assessment is carried out in Germany by competent authority BAM. In recent years BAM was involved in several licensing procedures of new spent fuel and HLW package Designs, where the cask body is of Ductile Cast Iron (DCI). According to IAEA regulations package Designs have to fulfill requirements for specific conditions of transport. Type B(U) packages must withstand the defined accident conditions of transport. The temperature range from -40°C up to the operational temperature has to be considered. For the cask material DCI, it is necessary to determine safety against brittle fracture. The German guideline BAM-GGR 007 defines requirements for fracture mechanics of packagings made of DCI. Due to complex cask body structure and the dynamic loading a fracture mechanical assessment by analytical approaches is not always possible. Experience of recent Design Approval procedures show that the application of numerical calculations are applicable to determine the stresses and stress intensity factors in the cask body. At the first step a numerical analysis has to be done to identify the loading state at the whole cask body. Secondly an analysis of a detail of the cask body is made considering the displacement boundary conditions of the global model. An artificial flaw is considered in this detailed model to calculate the fracture mechanical loading state. The finite element mesh was strongly refined in the area of the flaw. The size of the artificial flaw is based on the ultrasonic inspection acceptance criteria applied for cask body manufacture.. The applicant (GNS) developed additional analysis tools for calculation of stress intensity factor and/or J-Integral. The assessment approach by BAM led to the decision to develop own tools to the possibility for independent proof of the results. The paper describes the authority assessment approach for DCI fracture mechanics analysis. The validation procedure incl. the development of own tools is explained. BAM developed a postprocessor to determine the fracture mechanical loads. A horizontal 1 m puncture bar drop test is used to give a detailed description of the assessment procedure.
Bernhard Droste - One of the best experts on this subject based on the ideXlab platform.
-
Experimental testing of impact limiters for RAM packages under drop test conditions
Packaging Transport Storage and Security of Radioactive Material, 2020Co-Authors: André Musolff, Bernhard Droste, Thomas Quercetti, Karsten Müller, Klaus-peter GründerAbstract:In context with new cask Designs and their Approval procedure, the experimental testing of impact limiters under drop test conditions becomes more and more important in order to assess the damage mechanics behaviour and safety margins for validation reasons. In recent years, various Designs of impact limiters have been tested by the Federal Institute for Materials Research and Testing within specific component testing and particularly with regard to type B package Design Approval procedures. The paper focuses on the experimental realisation of impact limiter tests and presents implemented measurement techniques to determine the amount of deformation and to explain the impact behaviour by means of photogrammetric metrology and three-dimensional fringe projection method, high speed motion analysis and adjusted deceleration measurements.
-
MECHANICAL/THERMAL PACKAGE Design SAFETY ASSESSMENT AND MANUFACTURING QUALITY ASSURANCE OF SPENT FUEL TRANSPORT CASK NCS 45
2020Co-Authors: Christian Kuschke, Tino Neumeyer, Frank Wille, Bernhard DrosteAbstract:The NCS 45 package is Designed for a variety of radioactive contents ranging from irradiated PWR, BWR and FBR fuel rods, parts of fuel rods as well as non-fissile radioactive material. BAM assessed the applicants safety analysis concept and the appropriate report with respect of requirements according to transport regulations under routine, normal and accident transport conditions related to mechanical Design and thermal Design, activity release and quality assurance aspects. The assesment concept of the safety report was a combination of drop test program, analytic and numerical calculations as well as various components tests. The paper will give details of BAM activities in the official safety assessment that had been summarized in the BAM Design examination certificate which was a prerequisite for the Package Design Approval Certificate issued by the German competent authority.
-
Drop test program with half scale model CASTOR® HAW/TB2
Packaging Transport Storage and Security of Radioactive Material, 2020Co-Authors: André Musolff, Bernhard Droste, Thomas Quercetti, Karsten Müller, Steffen KomannAbstract:AbstractFederal 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 fur 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 aft...
-
Design Approval of Special Form Radioactive Material- Important Aspects
2016Co-Authors: Annette Rolle, Tino Neumeyer, Bernhard DrosteAbstract:The Design of a special form radioactive material has to resist a severe transport accident without undue loss or dispersal of radioactive material. Safety assessment by authorities competent for Design Approval has to include besides the required test program (impact, percussion, bending and heat test) also the evaluation of the quality management system for Design, manufacture, testing, documentation, use, maintenance and inspection. These 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 without undue loss or dispersal of radioactive material at any time of its working life. All important aspects in the Design Approval procedure by BAM as the competent authority for Approvals of special form radioactive material in Germany are summarized in a guideline published in 2014. This paper will give additional explanations to some aspects in safety assessment, e.g.: the applicability of leak test methods and the need to consider ageing aspects.
-
assessment of quality assurance measures for radioactive material transport packages not requiring competent authority Design Approval 13282
2013Co-Authors: Steffen Komann, Carsten Groke, Bernhard DrosteAbstract: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, Chap. 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 themore » 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 chap. 801 requires documentary verification by the consignor concerning package compliance with the requirements. (authors)« less
Steffen Komann - One of the best experts on this subject based on the ideXlab platform.
-
Drop test program with half scale model CASTOR® HAW/TB2
Packaging Transport Storage and Security of Radioactive Material, 2020Co-Authors: André Musolff, Bernhard Droste, Thomas Quercetti, Karsten Müller, Steffen KomannAbstract:AbstractFederal 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 fur 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 aft...
-
authority experience during Design Approval procedure for packages loaded with special encapsulations for damaged spent nuclear fuel
2019Co-Authors: Konrad Linnemann, Steffen Komann, Lars Muller, Thorsten Schonfelder, Frank WilleAbstract:The first German package Design Approval certificate for a dual purpose cask intended for loading with damaged spent nuclear fuel was issued recently. BAM as part of the competent authority system in Germany carried out a comprehensive assessment procedure with respect to the mechanical and thermal Design, the release of radioactive material and the quality assurance aspects of manufacturing and operation. Packages for the transport and storage of radioactive material have been assessed by BAM for many years, thus the common assessment procedure is well-known and good practice. Up to now only SNF without defects or HLW with well-defined properties were Designated for long-term Interim storage and transports afterwards. Due to Germany’s nuclear phase out all other kinds of spent nuclear fuel in particular damaged spent nuclear fuel shall be packed as well. Damaged spent nuclear fuel needs a tight closure with Special encapsulations and clearly defined properties in Germany. In addition, these encapsulations shall be long-term durable, because they are not accessible after loading in a packaging within periodical inspections. The main difference to Standard package components is that encapsulations with a permanent closure achieve their specified conditions not after manufacturing but only during operation, after loading and closing. To ensure compliance with the specific conditions, special measures for quality assurance are necessary during operation of each encapsulation, e.g. drying and sealing, which were assessed by BAM. The present paper gives an overview of the conducted assessment from BAM and point out the findings concerning to the special closure lid of the approved encapsulation, which is screwed and welded. A wide verification concept is necessary to show the specific tightness under transport conditions. Together with quality assurance measures during first operation steps these encapsulations with damaged spent nuclear fuel can be handled like standard fuel assemblies in approved package Designs.
-
assessment of quality management for transport packages not requiring authority Design Approval
2018Co-Authors: Steffen Komann, Carsten Groke, Frank WilleAbstract:The majority of transports of radioactive materials are carried out in packages which don’t need a package Design Approval of a competent authority. Low active radioactive materials are transported in such kind of packages e.g. in the medical and pharmaceutical industry and in the nuclear industry as well. In Germany the decision to phase out nuclear energy leads to a strong demand for packages to transport low and middle active radioactive waste due to the dismantling and decommissioning of nuclear power plants. 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 of Type A. For the packages of Type IP-2, IP-3 and Type A an assessment by the German competent authority is required for the quality management for the Design, manufacture, testing, documentation, use, maintenance and inspection. In general a compliance audit of the manufacturer of the packaging is required during this assessment procedure.
-
assessment of quality assurance measures for radioactive material transport packages not requiring competent authority Design Approval 13282
2013Co-Authors: Steffen Komann, Carsten Groke, Bernhard DrosteAbstract: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, Chap. 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 themore » 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 chap. 801 requires documentary verification by the consignor concerning package compliance with the requirements. (authors)« less
-
Assessment of Ductile Cast Iron Fracture Mechanics Analysis Within Licensing of German Transport Packages
Volume 3: Design and Analysis, 2012Co-Authors: Steffen Komann, Frank Wille, Yusuf Kiyak, Uwe Zerbst, Mike Weber, Dietmar KlingbeilAbstract:In the Design Approval of transport packages for radioactive materials, the mechanical and thermal safety assessment is carried out in Germany by competent authority BAM. In recent years BAM was involved in several licensing procedures of new spent fuel and HLW package Designs, where the cask body is of Ductile Cast Iron (DCI). According to IAEA regulations package Designs have to fulfill requirements for specific conditions of transport. Type B(U) packages must withstand the defined accident conditions of transport. The temperature range from -40°C up to the operational temperature has to be considered. For the cask material DCI, it is necessary to determine safety against brittle fracture. The German guideline BAM-GGR 007 defines requirements for fracture mechanics of packagings made of DCI. Due to complex cask body structure and the dynamic loading a fracture mechanical assessment by analytical approaches is not always possible. Experience of recent Design Approval procedures show that the application of numerical calculations are applicable to determine the stresses and stress intensity factors in the cask body. At the first step a numerical analysis has to be done to identify the loading state at the whole cask body. Secondly an analysis of a detail of the cask body is made considering the displacement boundary conditions of the global model. An artificial flaw is considered in this detailed model to calculate the fracture mechanical loading state. The finite element mesh was strongly refined in the area of the flaw. The size of the artificial flaw is based on the ultrasonic inspection acceptance criteria applied for cask body manufacture.. The applicant (GNS) developed additional analysis tools for calculation of stress intensity factor and/or J-Integral. The assessment approach by BAM led to the decision to develop own tools to the possibility for independent proof of the results. The paper describes the authority assessment approach for DCI fracture mechanics analysis. The validation procedure incl. the development of own tools is explained. BAM developed a postprocessor to determine the fracture mechanical loads. A horizontal 1 m puncture bar drop test is used to give a detailed description of the assessment procedure.
Fabiano Hernandes - One of the best experts on this subject based on the ideXlab platform.
-
Widespread Fatigue Damage Evaluation for Multiple Elements Based on Probabilistic Approach
ICAF 2019 – Structural Integrity in the Age of Additive Manufacturing, 2019Co-Authors: Fabiano HernandesAbstract:In order to obtain a Type Certificate for civil transport category aircraft the Applicant must to comply with some Design rules - airworthiness requirements. Among those, the one associated with structural fatigue (§25.571), whose main objective is related to prevent a catastrophic event, from structural damage, during the operational life of the aircraft. That section requires special attention for Widespread Fatigue Damage (WFD). For that, the Design Approval holder (DAH) must establish a Limit of Validity (LOV) of the engineering data that supports the structural maintenance program. Up to the LOV, DAH must demonstrate that the aircraft will be free from WFD. A Widespread Fatigue condition can be originated from: Multiple Site Damage, Multiple Element Damage or a combination of both. The objective of this work is to propose a probabilistic approach to define maintenance actions to prevent widespread fatigue damage condition from Multiple Element Damage up to the LOV.
Sungjae Yi - One of the best experts on this subject based on the ideXlab platform.
-
contribution of thermal hydraulic validation tests to the standard Design Approval of smart
Nuclear Engineering and Technology, 2017Co-Authors: Hyunsik Park, Taesoon Kwon, Sangki Moon, Sungjae YiAbstract:Abstract Many thermal–hydraulic tests have been conducted at the Korea Atomic Energy Research Institute for verification of the SMART (System-integrated Modular Advanced ReacTor) Design, the standard Design Approval of which was issued by the Korean regulatory body. In this paper, the contributions of these tests to the standard Design Approval of SMART are discussed. First, an integral effect test facility named VISTA-ITL (Experimental Verification by Integral Simulation of Transients and Accidents-Integral Test Loop) has been utilized to assess the TASS/SMR-S (Transient and Set-point Simulation/Small and Medium) safety analysis code and confirm its conservatism, to support standard Design Approval, and to construct a database for the SMART Design optimization. In addition, many separate effect tests have been performed. The reactor internal flow test has been conducted using the SCOP (SMART COre flow distribution and Pressure drop test) facility to evaluate the reactor internal flow and pressure distributions. An ECC (Emergency Core Coolant) performance test has been carried out using the SWAT (SMART ECC Water Asymmetric Two-phase choking test) facility to evaluate the safety injection performance and to validate the thermal–hydraulic model used in the safety analysis code. The Freon CHF (Critical Heat Flux) test has been performed using the FTHEL (Freon Thermal Hydraulic Experimental Loop) facility to construct a database from the 5 × 5 rod bundle Freon CHF tests and to evaluate the DNBR (Departure from Nucleate Boiling Ratio) model in the safety analysis and core Design codes. These test results were used for standard Design Approval of SMART to verify its Design bases, Design tools, and analysis methodology.
-
Contribution of thermal–hydraulic validation tests to the standard Design Approval of SMART
Nuclear Engineering and Technology, 2017Co-Authors: Hyunsik Park, Taesoon Kwon, Sangki Moon, Sungjae YiAbstract:Abstract Many thermal–hydraulic tests have been conducted at the Korea Atomic Energy Research Institute for verification of the SMART (System-integrated Modular Advanced ReacTor) Design, the standard Design Approval of which was issued by the Korean regulatory body. In this paper, the contributions of these tests to the standard Design Approval of SMART are discussed. First, an integral effect test facility named VISTA-ITL (Experimental Verification by Integral Simulation of Transients and Accidents-Integral Test Loop) has been utilized to assess the TASS/SMR-S (Transient and Set-point Simulation/Small and Medium) safety analysis code and confirm its conservatism, to support standard Design Approval, and to construct a database for the SMART Design optimization. In addition, many separate effect tests have been performed. The reactor internal flow test has been conducted using the SCOP (SMART COre flow distribution and Pressure drop test) facility to evaluate the reactor internal flow and pressure distributions. An ECC (Emergency Core Coolant) performance test has been carried out using the SWAT (SMART ECC Water Asymmetric Two-phase choking test) facility to evaluate the safety injection performance and to validate the thermal–hydraulic model used in the safety analysis code. The Freon CHF (Critical Heat Flux) test has been performed using the FTHEL (Freon Thermal Hydraulic Experimental Loop) facility to construct a database from the 5 × 5 rod bundle Freon CHF tests and to evaluate the DNBR (Departure from Nucleate Boiling Ratio) model in the safety analysis and core Design codes. These test results were used for standard Design Approval of SMART to verify its Design bases, Design tools, and analysis methodology.