Accident Conditions Of Transport Research Articles

Investigations of aged metal seals for transport package safety assessment

Abstract. Acceptable limits for activity release from transport casks for high-level radioactive material specified in the IAEA regulations must be kept by the integrity of cask body and the cask sealing system. BAM – the German competent authority for mechanical, thermal and containment assessment of packages liable for approval – verifies the activity release compliance with the regulatory limits. One of the fundamental aspects in assessment is the specification of conservative package design leakage rates. To ensure the required package tightness for both storage and transport of the cask before and after storage, usually metal seals of the HELICOFLEX® type are used. Due to long-term use, the seal behaviour is influenced by temperature and time. The mechanical and thermal loadings associated with the routine, normal and accident conditions of transport specified in the regulations can have a significant effect on the leak tightness of the sealing system. Whereas the safety for application of new, non-aged HELICOFLEX® seals is verified sufficiently, there are still technical data gaps concerning the efficiency of aged HELICOFLEX® seals. BAM performed experiments to learn more about the sealing efficiency of aged HELICOFLEX® seals with aluminium and silver outer jackets. The seals were compressed in test flanges, and for artificial ageing the complete flange systems were stored in an oven for several months at a high temperature. During the compression and decompression tests after the ageing, load-deformation characteristics of the seals and leakage rates were measured. With these tests a load situation was simulated, which can occur in the regulatory drop test of the cask: under high-impact loads the bolted lid can lift a little for a short moment, allowing a little movement of the seal so that the contact area can change before compressing again. The poster presentation will show details about test conditions and first results.

Sensitivity of SNF transport cask response to uncertainty in properties of wood inside the impact limiter under drop accident conditions

It is essential to ensure the safety of spent nuclear fuel (SNF) transport cask in drop situation that is included in transport accident scenarios. The safety of the drop situation is affected by the impact absorption performance of impact limiters. Therefore, when designing an impact limiter, the uncertainty in the material properties that affect the impact absorption performance must be considered. In this study, the material properties of the wood inside the impact limiter were selected as the variables for a parametric study. The sensitivity analysis of the drop response of the SNF transport cask with impact limiter was performed. The minimum wood strength required to prevent a direct collision between the cask and floor was derived from the analysis results. In addition, the plastic strain response was analyzed and strain-based evaluation was performed. Based on this result, the critical values of wood properties that change the impact dynamic characteristics were investigated. Finally, the optimal material properties of wood were obtained to secure the structural safety of the SNF transport cask. The results of this study can contribute to the development of SNF transport cask, thereby ensuring safety in transport accident conditions.

Numerical Simulation of Cobalt Absorber Rod Transportation Flask Under Impact Loading and Design of Impact Limiter Cage

A Lead shielded Cobalt absorber rod transportation flask has been designed for transportation of Cobalt absorber rod through public domain as Type B(U) package. As per the IAEA/AERB regulations for safe transport of radioactive materials, the flask should demonstrate compliance to the performance standards for normal and accident conditions of transport. In view of this, the ability of the flask to withstand 9m drop on an unyielding target in most damaging orientation is verified by finite element analysis and design improvement is carried out in the flask. The explicit finite element (FE) code PAM-CRASH is used for the numerical simulation. It is found that the stresses and deformations in flask components do not meet structural integrity criteria. Therefore, an impact limiter cage is conceptualized and series of design modifications in the cage are carried out through numerical simulations to meet the structural integrity requirement under different drop orientations. The cage serves the purpose of an effective impact limiter and meets the structural integrity criteria by limiting overall damage to the flask. The concepts of cage and drop test simulation results are presented in this paper along with comparison of simulation results of flask with cage and without cage.

Proposal to simplify LSA-III material requirements of IAEA Transport Regulations

AbstractThe LSA-II and LSA-III material requirements were introduced into the International Atomic Energy Agency (IAEA) Transport Regulations in the early 70s and revised in the 80s of last century. Meanwhile much progress has been achieved internationally regarding knowledge of material characteristics of LSA-II and LSA-III and their release behaviour under accident conditions of transport. In particular, in Germany, comprehensive experimental and theoretical research work was performed in recent years to investigate the release behaviour of various LSA-II and LSA-III materials under different mechanical impact conditions and to use the results to assess the potential radiation exposure resulting from such materials under severe handling and transport accident conditions. Based on this research work, a review of LSA-II and LSA-III requirements was performed, which was mainly focusing on the need and the justification of the LSA-III leaching test, which have often be questioned in the past. For packages w...

Thermal analysis on NAC-STC spent fuel transport cask under different transport conditions

Transport casks used for conveying spent nuclear fuel are inseparably related to the safety of the whole reprocessing system for spent nuclear fuel. Thus they must be designed according to rigorous safety standards including thermal analysis. In this paper, for NAC-STC cask, a finite element model is established based on some proper simplifications on configurations and the heat transfer mechanisms. Considering the complex components and gaps, the equivalence method is presented to define their material properties. Then an equivalent convection coefficient is introduced to define boundary conditions. Finally, the temperature field is captured and analyzed under both normal and accident transport conditions by using ANSYS software. The validity of numerical calculation is given by comparing its results with theoretical calculation. Obtaining the integral distribution laws of temperature and peak temperature values of all vital components, the security of the cask can be evaluated and verified.

Spent fuel transportation risk assessment: overview

The US Nuclear Regulatory Commission (NRC) is responsible for issuing regulations for the packaging of spent fuel (and other large quantities of radioactive material) for transport that provide for public health and safety during transport [Title 10 of the Code of Federal Regulations (10 CFR) Part 71, ‘Packaging and transportation of radioactive waste’, dated 26 January 2004]. In September 1977, the NRC published NUREG 0170, ‘Final environmental statement on the transportation of radioactive material by air and other modes’, which assessed the adequacy of those regulations to provide safety assurance. In that assessment, the measure of safety was the risk of radiation doses to the public under routine and accident transport conditions, and the risk was found to be acceptable. Since that time, there have been two affirmations of this conclusion for spent nuclear fuel (SNF) transportation, each using improved tools and information that supported the earlier studies. This report presents the results of a fou...

Operational Vibration Analysis of Railway Transport Complex

Packages for transporting radioactive materials have to meet requirement of normal and accident transport conditions. On the basis of methodology of experimental verification packages for transport of radioactive materials elaborated by authors were realized extensive tests of containers. In the paper is described modal and vibration analysis of transport complex during its transportation.

Mechanical assessment within type B packages approval: application of static and dynamic calculation approaches

This paper demonstrates exemplarily how numerical and experimental approaches can be combined reasonably in mechanical assessment of package integrity according to the IAEA regulations. The paper also concentrates on the question about how static mechanical approaches can be applied, and what their problems are in relation to dynamic calculation approaches. Under defined impact tests, which represent accident transport conditions, the package has to withstand impact loading, e.g. resulting from a 9 m free drop onto an unyielding target in sequence with a 1 m puncture drop test. Owing to the local character of the interaction between the puncture bar and the cask body, it is possible to develop a dynamic numerical model for the 1 m puncture drop which allows an appropriate simulation of the interaction area. Results from existing experimental drop tests with prototype or small scale cask models can be used for verification and validation of applied analysis codes and models. The link between analysis and e...

Drop test program with half scale model CASTOR® HAW/TB2

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...

Historical background: early deliberations on and assessments of need for dynamic crush test

AbstractBeginning in the late 1970s, discussions were fostered by the International Atomic Energy Agency (IAEA) on the need for additional tests for some type B packages. Consideration at the international level of these early deliberations and tests ultimately led to the inclusion in the IAEA Regulations for the Safe Transport of Radioactive Material of the third mechanical (drop) test for demonstrating the ability of the package design to withstand accident conditions of transport, commonly known as the ‘dynamic crush test’. This test included the requirement that the package be positioned so as to sustain maximum damage. Recently discussions have been occurring as to what constitutes positioning on an unyielding target, where considerations are being put forward for clarifying this phrasing and possibly changing the test requirement. Some of these proposed changes could make the test more demanding than originally envisioned. This paper, developed in support of a panel discussion at PATRAM 2010, provid...

Mechanical safety analysis for high burn-up spent fuel assemblies under accident transport conditions

AbstractTransport packages for spent fuel have to meet the requirements concerning containment, shielding and criticality as specified in the International Atomic Energy Agency regulations for different transport conditions. Physical state of spent fuel and fuel rod cladding as well as geometric configuration of fuel assemblies are, among others, important inputs for the evaluation of correspondent package capabilities under these conditions. The kind, accuracy and completeness of such information depend upon purpose of the specific problem. In this paper, the mechanical behaviour of spent fuel assemblies under accident conditions of transport will be analysed with regard to assumptions to be used in the criticality safety analysis. In particular the potential rearrangement of the fissile content within the package cavity, including the amount of the fuel released from broken rods has to be properly considered in these assumptions. In view of the complexity of interactions between the fuel rods of each fu...

Thermal analysis of a spent fuel cask in different transport conditions

Like all industries, the generation of electricity from Nuclear Power Plant produces wastes to be managed. Spent fuel element casks used for transport of nuclear materials must be designed according to rigorous acceptance criteria and standards requirements, e.g. International Atomic Energy Agency ones, in order to provide protection to people and environment against radiation exposure. The aim of this work was the evaluation of the integrity of spent fuel cask under both normal and accident transport conditions, such as impact (9 m drop impact event onto a flat, essentially unyielding, horizontal surface, in the most damaging orientation) and rigorous fire events (full exposure to an engulfing fire for 30 min (fire test) or to an environment at 800 °C temperature for a numerical simulation or for a furnace test). Using the finite element code ANSYS both steady-state and transient thermal analyses were carried out to determine the maximum fuel temperature and the temperatures behaviour into the cask, considering all the heat transfer modes between the cask and the external environment as well as inside the cask itself. Moreover, both wet and dry fuel storage inside the cavity of the body were analyzed. The obtained results, used for the new licensing approval by the Italian competent Authority of the cask for PWR spent fuel transport, are discussed.

Modelling of release of particulate material from transport containers

The Nuclear Decommissioning Authority (NDA) is developing a family of Standard Waste Transport Containers (SWTCs) for the transport of unshielded intermediate level radioactive waste packages. The SWTCs are shielded transport containers designed to carry different types of waste packages. The combination of the SWTC and the waste package is required to meet the regulatory requirements for Type B packages. One such requirement relates to the containment of the radioactive contents, with the IAEA Transport Regulations specifying release limits for normal and accident conditions of transport. In the impact tests representing accident conditions of transport, the waste package will experience significant damage and radioactive material will be released into the SWTC cavity. It is therefore necessary to determine how much of this material will be released from the cavity to the external environment past the SWTC seals. Typical assessments use the approach of assuming that the material will be evenly distributed within the cavity volume and then determining the rate at which gas will be released from the cavity, with the volume of radioactive material released with the gas based on the concentration of the material within the cavity gas. This is a pessimistic approach as various deposition processes would reduce the concentration of gas-borne particulate material and hence reduce their release rate from the SWTC. This paper assesses these physical processes that control the release rate and develops a conservative methodology for calculating the particulate releases from the SWTC lid and valve seals under normal and accident conditions of transport, in particular: a) the flows within the SWTC cavity, especially those near the cavity walls; b) the aerodynamic forces necessary to detach small particles from the cavity surface and suspend them into the cavity volume; c) the adhesive forces holding contaminant particles on the surface of a waste package; d) the breakup of waste material upon impact that will determines the volume fraction and size distribution of fine particulate released into the cavity. Three mechanisms are specifically modelled, namely Brownian agglomeration, Brownian diffusion and gravitational settling, since they are the dominant processes that lead to deposition within the cavity and the easiest to calculate with much less uncertainty than the other deposition processes. Calculations of releases under normal conditions of transport concentrate on estimating the detachment of any waste package surface contamination by inertial and aerodynamic forces and show that very little of any contamination removed from the waste package surface would be released from the SWTC. Under accident conditions of transport, results are presented for the fraction released from the SWTC to the environment as a function of the volume fraction of the waste package contents released as fine particulate matter into the SWTC cavity. These show that for typical release fractions of 10 -6 to 10 -8 for the release of radioactive material from waste packages into the SWTC cavity, the release fraction of the waste package inventory from the

Behaviour of metallic seals in CASTOR casks under normal and accident conditions of transport: qualification requirements

AbstractThis paper presents several BAM requirements for a sealing qualification process for metallic seals in transport casks. On qualification tests performed by GNS, advice will be given for the implementation of necessary investigations. This paper shows the results of special research by GNS which were used as proof concerning tightness behaviour of seals with aluminium and silver outer jackets under the influence of mechanical loading, particles and corrosion.

Multiple water barriers: alternative to assessment of fuel assemblies during accident conditions of transport

Packages for the transport of radioactive material have to comply with national and/or international regulations. These regulations are widely based on the requirements set forth by the International Atomic Energy Agency (IAEA) in the 'Regulations for the safe transport of radioactive material'. In this framework, packages to transport fuel assemblies (including spent fuel assemblies) have to meet the requirements for packages containing fissile material. In accident conditions of transport, the applicant for the package design approval has to show that the package remains subcritical taking due account of the status of the contents in these conditions. In most cases, considering water ingress in the package, it is not possible to assume that the fissile material included in the fuel assemblies is dispersed in the package with the most severe conceivable distribution regarding criticality. In order to alleviate this difficulty, during the last years, we have provided a significant better knowledge of the conditions of the fuel assemblies to be transported. This was part of the Fuel Integrity Project, whose progress was regularly reported during PATRAM 2001 and PATRAM 2004 Symposia. However, for packages which encounter a large g-load during accident conditions of transport and/or which contain spent fuel assemblies with very high burn-up, it can be difficult to demonstrate that the fuel assemblies are not significantly damaged. Then, to make the criticality assessment considering water inleakage into the flask and a large release of fissile material within its cavity will not allow meeting the subcriticality criteria. For that reason, for our package designs, which use a gas and not water as an internal coolant and which fall into that category, the author has decided to take credit of the possibilities provided by the subparagraph 677 (b) of the Regulations. This paragraph allows not taking into account water in the package, provided that the package exhibits 'multiple high standard water barriers'. The paper describes the author's experience with the implementation of this paragraph. Two different cases are considered: either a double vessel, or a double lid. It will be explained when each of these solutions is implemented, and give examples of package designs with such features, as well as the approvals which were granted for these designs in various countries.

Mechanical characteristics of fuel rod claddings in transport conditions

AbstractA synthesis on the mechanical characteristics of unirradiated and irradiated fuel rod claddings was performed by the French Institut de Radioprotection et de Surete Nucleaire (IRSN) in order to have reference data for the assessment of the safety demonstrations in normal and accident conditions of transport required by the procedure of package licensing. Indeed, the transport conditions correspond to a range of cladding temperatures (200–550°C) which is only partly covered by the data acquired within the framework of the safety demonstration relative to the reactor normal operating conditions, especially beyond 400°C. This work concerned Zircaloy-4 cladding material (Zry-4) and M5TM. Data about mechanical properties (elastic and ductile properties, creep behaviour), oxidation (in reactor and under air during transport), hydrides and fracture toughness have been collected and synthesised. The laws presented in the document can be used to obtain orders of magnitude of oxide layer thickness, hydrogen...

New metallic gaskets qualification: resistance to accident conditions of transport in TN international casks

AbstractFor dual purpose 'transport/storage' casks (like the TN24 cask family or casks for vitrified wastes), the confinement by way of metallic gaskets has long been the chosen solution. Their metallic nature leads to a good behaviour in the long term and high temperature conditions but they still have to satisfy the requirements of transport regulations, which are extremely demanding. In order to secure the supply of such strategic parts of a cask and, above all, to keep developing efficient sealing solutions, TN International has found and qualified new metallic gaskets. For this qualification, a series of tests complying with the accident conditions of transport have been designed and performed on these new metallic gaskets. Indeed, the gaskets have to withstand three major types of 'accident' conditions:(i) parallel sliding of the lid and the body, resulting in a lateral displacement of the contact surfaces(ii) separation of the lid and the body, resulting in a decompression of the gasket(iii) 30 min...

Evaluation and verification of a structural safety analysis for a type B transport package for a radioactive waste drum

We conducted and verified the structural safety analyses for a type B transport package for a radioactive waste drum. The type B package is used to transport one high-level radioactive waste drum from a waste facility to a temporary site in a nuclear power plant. It is important that the structural integrity of a type B transport package should be maintained under normal transport conditions such as the free drop test, the stacking test and the penetration test, and accident conditions of transport such as the 9 m height free drop test and the puncture test. Based on the ASME Boiler and Pressure Vessel Code, Section III, Division 3 we evaluated the structural safety integrity of the type B package for a radioactive waste drum using finite element analysis. For the stacking test condition the maximum stress is a very low value when compared with the stress limits. The impact of the penetration bar on the overpack does not affect the containment part. For the free drop test condition and the puncture test condition the local stresses appear at the impact part and the containment part connected with the inner structure. The maximum stresses for all of the conditions are lower than the stress limits, so the structural integrity of the type B package for a radioactive waste drum was maintained. Also, the structural safety analyses were verified by using the data acquired from the 9 m drop and the puncture tests of the real model. The analytic accelerations of the 9 m drop test are similar to the test results. For the puncture test condition the analytic result is conservative when compared to the test. The analysis for the structural safety of the type B package reveals conservative results and it is proved that its structural integrity is maintained under normal and accident conditions of transport.

Measurement techniques and preliminary results of drop tests with full scale spent fuel transport and storage casks

The experimental testing and quantitative stress analysis of new cask designs becomes more and more important with increasing size of casks and their impact limiters, in order to assess the safety margins as well as for calculation validation reasons. Full scale drop testing of spent fuel transport and storage casks is performed by the Federal Institute for Materials Research and Testing within the type B package approval procedure and for safety demonstration reasons respectively. In drop tests for demonstrating ability of a package to withstand normal and accident conditions of transport, the instrumentation of a specimen is an important tool to evaluate its mechanical behaviour during impact. Generally, the instrumentation incorporates the measurement of strains and accelerations at the package. Test results as deceleration–time and strain–time functions constitute a main basis for the validation of assumptions in the safety analysis, for the evaluation of calculations based on finite element methods and extrapolation of scale model testing on full sized package within approval design tests. Besides, all of the above mentioned instrumentation, the flatness of the sealing area and the circularity of the cask lid flange are determined by using three-dimensional coordinate measurements to evaluate the mechanical impact behaviour of the cask flange region after the drop test sequence. Indispensable are measurements concerning leak tightness of the closure lid system. Additionally, specific photogrammetric measurements are used to characterise the conditions of the closure lid system after drop test in more detail. The preparation and interpretation of the results is in progress and indicates possible correlations between the leak tightness rate and measures of the relative movement of the lid.

Safety of ocean shipments of plutonium between Europe and Japan

Sandia National Laboratories (SNL) has conducted an extensive study of emergency response planning applicable to sea transport of plutonium for the Japan Nuclear Cycle Development Institute (JNC). This work covered four separate areas to better define the accident environment for long range sea transport of nuclear materials. A probabilistic safety analysis evaluated technical issues for the transport of plutonium between Europe and Japan. An engine room fire aboard a purpose built ship was used to analyse the vulnerability of plutonium packaging designed to International Atomic Energy Agency (IAEA) standards. A comprehensive corrosion study estimated the time required for sea water to breach a containment boundary in submerged generic plutonium packaging. A survey of worldwide commercial recovery capabilities provided a compilation of information on the capabilities of salvaging high value cargo from sunken ships. This paper addresses salvage modes from harbour depths to the deepest ocean trenches. Previous studies (J. L. Sprung et al., SNL reports SAND98-1171/1 and SAND98-1171/2, May 1998) included a probabilistic risk assessment of the overall safety, source term evaluations and finite element structural dynamics calculations to determine the effects of ship to ship collisions on nuclear material containers and the effects of ship fires on transport packaging as determined by actual fire experiments conducted on board a test ship. The previous studies, together with this work, form a comprehensive technical basis that encompasses the overall safety of sea transport of plutonium between Europe and Japan. Based on these technical analyses, transport of nuclear materials by sea in Type B packaging, approved in accordance with US Nuclear Regulatory Commission (NRC) and IAEA regulations, and carried in purpose built ships with adequate surveillance, has a very high degree of safety for the failure modes studied. Non-purpose built ships do not have the redundancy in safety features provided by the newer purpose built ships. However, SNL studies on non-purpose built ships have shown accident environments to be within NRC and IAEA regulatory assumptions for Type B packaging. These studies were carried out for both structural ship to ship collisions and engine room fires by analysis (for the collisions) and direct experimentation and analysis (for the fires). Thus, land transport mode regulations are applicable for sea transport accident conditions.