Filtern
Erscheinungsjahr
- 2018 (9) (entfernen)
Dokumenttyp
- Beitrag zu einem Tagungsband (9) (entfernen)
Sprache
- Englisch (9)
Schlagworte
- Radioactive material (2)
- Wood (2)
- Assessment (1)
- Containment (1)
- Drop testing (1)
- Dual purpose cask (1)
- FEM (1)
- Fire (1)
- Fire test (1)
- IAEA (1)
Organisationseinheit der BAM
Accident safe packages for the transport of spent nuclear fuel and high-level waste shall fulfil international IAEA safety requirements. Compliance is shown by consecutive mechanical and thermal testing. Additional numerical analysis are usually part of the safety evaluation. For damage protection some package designs are equipped with wood filled impact limiters encapsulated by steel sheets. The safety of these packages is established in compliance with IAEA regulations. Cumulative mechanical and fire tests are conducted to achieve safety standards and to prevent loss of containment. Mechanical reliability is proven by drop tests. Drop testing might cause significant damage of the impact limiter steel sheets and might enable sufficient oxygen supply to the impact limiter during the fire test to ignite the wood filling. The boundary conditions of the fire test are precisely described in the IAEA regulatory. During the test the impact limiter will be subjected to a 30 minute enduring fire phase. Subsequent to the fire phase any burning of the specimen has to extinguish naturally and no artificial cooling is allowed. At BAM a large-scale fire test with a real size impact limiter and a wood volume of about 3m3 was conducted to investigate the burning behaviour of wood filled impact limiters in steel sheet encapsulation. The impact limiter was equipped with extensive temperature monitoring equipment. Until today burning of such impact limiters is not sufficiently considered in transport package design and more investigation is necessary to explore the consequences of the impacting fire. The objective of the large scale test was to find out whether a self-sustaining smouldering or even a flaming fire inside the impact limiter was initiated and what impact on the cask is resulting. The amount of energy, transferred from the impact limiter into the cask is of particular importance for the safety of heavy weight packages. With the intention of heat flux quantification a new approach was made and a test bench was designed.
Numerical approach to determine the correct puncture bar length for the IAEA puncture bar drop test
(2018)
Among other mechanical tests the 1 meter drop onto a steel puncture bar shall be considered for accident safe packages for the transport of radioactive material. According with the IAEA regulations “the bar shall be of solid mild steel of circular section, 15.0 ± 0.5 cm in diameter and 20 cm long, unless a longer bar would cause greater damage…”. The most damaging puncture bar length can be estimated by iterative processes in numerical simulations. On the one hand, a sufficient puncture bar length has to guarantee that shock absorbers or other attachments do not prevent or reduce the local load application to the package, on the other hand, a longer and thus less stiff bar causes a smaller maximum contact force. The contrary influence of increasing puncture bar length and increasing effective drop height shall be taken into account if a shock absorber is directly placed in the target area. The paper presents a numerical approach to identify the bar length that causes maximum damage to the package. Using the example of two typical package masses the sensitivity of contact forces and puncture bar deformations to the initial length are calculated and assessed with regard to the international IAEA package safety requirements.
The containment systems of transport casks for spent fuel and high radioactive waste usually include bolted lids with metallic or elastomeric seals.
The mechanical and thermal loadings associated with the routine, normal and accident conditions of transport can have a significant effect on the leak tightness of such containment system.
Scaled cask models are often used for providing the required mechanical and thermal tests series. Leak tests have been conducted on those models.
It is also common practice to use scaled component tests to investigate the influence of deformations or displacements of the lids and the seals on the standard leakage rate as well as to study the temperature and time depending alteration of the seals.
In this paper questions of the transferability of scaled test results to the full size design of the containment system will be discussed.
Packages for the transport of radioactive material are often equipped with impact limiters consisting of wood. Mostly this wood is encapsulated by steel sheets. The impact limiters are needed to ensure that the transport casks meet the IAEA safety requirements. According to the IAEA safety requirements a package has to withstand consecutively severe mechanical tests followed by a thermal test. The mechanical tests have to produce maximum damage concerning the thermal test. Following this, the impact limiters may have serious pre-damage when the thermal tests begins. The IAEA safety requirements state that during and following the fire test, the specimen shall not be artificially cooled and any combustion of materials of the package shall be permitted to proceed naturally.
Small scale fire tests with wood filled metal drums by BAM and works of the French Institute for Radiological Protection and Nuclear Safety (IRSN) showed that pre-damaged steel encapsulated wooden structures could start smoldering initiated by the thermal test. These processes supply additional energy to the cask which should be considered within the safety assessment of the package.
As not much is known about smoldering processes in encapsulated wooden structures with a reduced oxygen supply the need for a test was identified. To investigate the influence of a smoldering impact limiter concerning the amount of energy supplied to the cask in dependence of the time BAM conducted a large scale impact limiter thermal test. For that, a pre-damaged impact limiter with a diameter of 2,3 m was mounted on a water tank simulating a cask. A complex system of a regulated pump, a heater, a cooler, a slide valve, a flow meter and numerous thermocouples were installed and connected to a control unit to ensure all needed operating conditions.
After a pre-heating compared to typical SNF decay-heat, the 30 min lasting fire phase of the thermal test was started. After that, the expected and initiated smoldering began. The results of the large scale test are presented in this poster. Systematic small scale tests will follow to identify the influence of different parameters, e.g. moisture content and scale effects. The tests took place at BAM Test Site for Technical Safety (TTS) with its various possibilities for mechanical and thermal tests. The results of these tests will have direct influence in the safety assessment of transport cask for the transport of radioactive material
Specific attention should be paid on safety demonstrations transmitted by applicants in the case of approval request for the package designs containing enriched UF6. Concerning the shipment of enriched UF6, the package designs consist in general of a filled 30B cylinder surrounded by an overpack.
The description of the content, considering the UF6 origin, i.e. natural or reprocessed, shall be clearly justified especially when the UF6 isotopic composition exceeds the limits specified in ASTM standards.
Concerning the containment of the UF6, the applicant shall demonstrate in all conditions of Transport the leak-tightness of the valve and plug of the cylinders filled with enriched UF6. In this regard, when justifications are based on numerical calculations, the absence of contact between These components of the cylinder and the internal surfaces of the overpack after the regulatory drop tests shall be shown. In particular, absence of contact between the valve and any other component of the packaging shall be confirmed to respect the current IAEA regulations [3]. If complementary calculations show a contact between the plug and the internal surfaces of the overpack, additional tests are required to confirm that the strength resulting from this contact will not affect the plug leak-tightness. It can be noticed that the future revision of the IAEA regulations will include additional provision in case of contact of the plug with any other component of the packaging. In addition, the applicant shall demonstrate that the melting temperature of the valve, including the tinned joint, will not be exceeded during the regulatory fire test.
Furthermore, the representativeness of the ballast used to simulate the behaviour of the UF6 loaded within the cylinder shall be justified if drop tests are performed.
Finally, specific provisions relative to the use of plugs and the maintenance of cylinders should be included in the safety analysis reports.
Recent developments in standards and IAEA guidance material for package load attachment points
(2018)
For transport package design and operation according to IAEA regulations, the package shall be securely stowed and its retention system shall be capable to withstand load conditions of routine transport. The supporting IAEA Advisory Material SSG-26 provides information how to do that.
Up to now package designers in different countries use other load factors for the design of attachment points than those specified in the IAEA guidance material. In particular the acceleration values vary between different countries and lead to difficulties during the validation of foreign approval certificates. Therefore the IAEA started a discussion process to review the existing guidance text. An international working group was constituted in 2013. Representatives came from different stakeholders, e.g. transport operators, competent authorities and modal organizations. The discussions concluded especially on the transport conditions which has to be considered for stowage design, including on the one hand the relevance of the load factors used for strength and fatigue analysis and on the other hand the criteria which have to be considered for the attachment points.
Another standard with relevance to the load attachment of packages is ISO 10276. This standard deals with trunnion design, manufacturing and operational aspects. The regular standard revision phase started in 2017. An expert group discussed new state-of-the-art technology and proposed revised text for the ISO standard for international discussion.
The paper describes relevant tie-down aspects, gives background argumentation, and tries to support harmonized application of the revised IAEA guidance material and future ISO 10276 standard.
Assessment experience on packages loaded with damaged spent nuclear fuel for transport after storage
(2018)
In 2017 the first German package approval certificate was issued for a dual purpose cask (DPC) design with encapsulated damaged spent nuclear fuel. At the Bundesanstalt für Materialforschung und -prüfung (BAM) a comprehensive assessment procedure was carried out with respect to the mechanical and thermal design, the containment design and quality assurance for manufacturing and operation. Main objective of this procedure was to verify the Package Design Safety Report (PDSR) fulfils the requirements according to the IAEA regulations SSR-6.
Until now only standard spent nuclear fuel assemblies were designated for interim storage and transports. Due to nuclear phase out in Germany all other kinds of SNF in particular damaged fuel has to be packed. Therefore specific requirements have to be considered in accordance with international experiences written in IAEA technical reports. In Germany damaged spent nuclear fuel (DSNF) needs a tight encapsulation with special encapsulations and clearly defined properties.
Due to the limited amount of DSNF these encapsulations are designed for storage and transport in existing packages. From the assessment experience it has been seen, corresponding PDSR need an extensively expansion to cover the design of these encapsulations and their influences on the package. Then such well-defined encapsulations can be handled like standard fuel assemblies. The main difference to standard package components is, encapsulations with permanent closure achieve their specified condition not after manufacturing but only during operation after loading and closing.
Thus specific handling instruction and test procedures are necessary especially for welding, where BAM is able to survey the quality of this first part of operation.
Typical transport packages used in Germany are equipped with encapsulated wooden impact limiting devices. We would like to present the current status regarding the development of a Finite Element (FE) material model for the crush of wood for the FE-code LS-DYNA. The crush of is a phenomenon governed by macroscopic fracture. Here, we would like to reproduce fracture and failure mechanisms over the continuous volume. In a first step we altered an existing LS-DYNA material model for foams, which considers an ellipse shaped yield surface. For the use for longitudinal compression of wood, we modified the existing material model to consider the deviatoric strain for the evolution of the yield surface as well. This is in accordance with the results of crush tests with spruce wood specimens, where the crushing deformation was rather deviatoric for uniaxial stress states and rather volumetric for multiaxial stress states We rate the basic idea of this approach to be reasonable, though other problems exist regarding the shape of the yield surface and the assumption of isotropic material properties. Therefore we developed a new transversal isotropic material model with two main directions, which considers different yield curves according to the multiaxiality of the stress state via a multi-surface yield criterion and a non-associated flow rule. The results show the ability to reproduce the basic strength characteristics of spruce wood. Nevertheless, problems with regularization etc. show that additional investigations are necessary.
Assessment of quality management for transport packages not requiring authority design approval
(2018)
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.