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

Abstract

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.     

Transport of radioactive sources in Africa: the Nigerian experience

Abstract

In his plenary presentation at PATRAM 2010, Professor Shamsideen Elegba of the Nigerian Nuclear Regulatory Authority, Abuja, Nigeria, reviewed the regulatory framework for transport of radioactive materials in Africa.     

Grain growth of zircon pigment in tile glaze

Abstract

Glazes for ceramic materials are often opacified by a zircon pigment. The increase in optical density is related to the number and size of pigment particles, which are well dispersed in the glass. Mechanical dispersion by grinding of a fine zircon powder in a raw glaze mixture results in an unoptimised pigment dispersion in the melted glassy phase at high temperature. A significant reduction in the number of particles and a simultaneous large increase in their size is observed. In this case the optical glaze density does not attain the expected value and a very large quantity of zircon must be added. In the present work it is shown that a large part of the zircon reacts with the glass phase during firing. Simultaneously, the remaining zircon grains undergo a large increase in grain size by Ostwald ripening through the liquid phase. This process is characterised by a low activation energy (79 ± 1 kJ mol^-1) and a short diffusion path for Zr through the liquid phase between neighbouring grains. The path length is shorter than the size of the larger grains. In the usual firing range, both populations of smaller and larger grains vary in size, the smaller grains favouring growth of the larger grains.     

Fireside issues in advanced power generation systems

Abstract

The traditional trend towards the development and use of power plants with ever increasing efficiencies is now being coupled to the use of a wider range of fuels and technologies designed to minimise CO₂ emissions. Alternative solid fuels such as biomass and waste products, which can be classified as CO₂ neutral, are being used alone or cofired with fossil fuels. The cofiring of biomass and coal is currently the most efficient and effective method for using biomass to generate power. CO₂ capture technologies include systems for either precombustion or postcombustion CO₂ removal. Gasification of fuels (using either oxygen or steam as the oxidant) produces a gas that can be conditioned to enable precombustion CO₂ removal. Post-combustion CO₂ capture can be carried out using either solid or aqueous sorbent processes. Oxy firing of fuels is a technology that would enable more efficient post-combustion CO₂ capture. The various combinations of new fuels, novel technologies and higher temperature component operating conditions are producing challenging operating environments for components. Deposition, erosion and corrosion issues for hot gas path components in these advanced power generating systems, which are potentially life limiting, are reviewed. Reduction in heat transfer owing to high rates of deposition can significantly reduce heat transfer and increase the need for component cleaning. Depending on the system, component parts can include various heat exchangers, gas cleaning systems and gas turbines.     

Considerations of Irradiation-Induced Transient Creep in Fuel Particle Modeling*

An analytical assessment is made of the potential effects of irradiation-induced transient creep on the behavior of the TRISO-coated fuel particles of the New Production Modular High Temperature Gas-Cooled Reactor (NP-MHTGR). An analytical solution is presented for the three-layer particles, which includes transient creep in addition to steady-state creep behavior. The solution allows for evaluating the effects that transient creep has on individual particle stresses and for determining failure probabilities for particle batches using the Monte Carlo approach. Because experimental data needed to determine parameters for a transient component in a creep model for the pyrocarbons is not available, a range of possible parameter values were considered in the assessments. It was shown that transient creep measurably affects particle stresses early in the irradiation life of the particle. At that time, the hoop stress in the primary load bearing layer of the particle is in compression and the article is not vulnerable to pressure vessel failure. Later in irradiation, the effects of transient creep were typically shown to be less significant. Thus, transient creep had less than an order of magnitude effect on batch failure probabilities for prototypical NP-MHTGR fuel particles and was much less significant than steady-state creep. Whether the presence of transient creep increased or decreased the particle failure probability was dependent on the specific values used for the transient creep material properties.     

Spent fuel transport using cask loaded outside pool

Abstract

During the last year, Sogin (the Italian company in charge for decommissioning of Italian nuclear power plants) had to implement an accelerated decommissioning plan of a EUREX spent fuel pool due to finding a water leakage into the environment from the pool. EUREX is no longer operating a pilot reprocessing plant, which some years ago became the responsibility of Sogin. There were 52 spent fuel assemblies from the Trino Vercellese PWR nuclear power plant, 48 irradiated pins from a Garigliano BWR fuel assembly, and 10 plates from an irradiated MTR fuel assembly stored in the EUREX pool, so the first step of the accelerated decommissioning plan consisted in the evacuation of this spent fuel. Considering the necessity to start the evacuation as soon as possible, Sogin decided to use an already existing cask (AGN-1) used in the past for the transport of Trino and Garigliano fuel assemblies. This cask was requalified in order to obtain a transport licence for the fuel assemblies stored in the EUREX pool according to ADR 2005 regulation. The transport license for the AGN-1 cask loaded with EUREX fuel assemblies was released by APAT (the Italian Safety Authority) in the spring of 2007. Owing to the limited capacity of the EUREX pool crane (27 t for nuclear loads) and limited dimensions of pool operational area, it was not possible to transfer the AGN-1 cask (50 t) into the pool for fuel assemblies charging. The solution implemented to overcome this problem was the loading of the cask outside the pool. A special shielding shuttle was developed and used to allow safe spent fuel transfer between the pool and the cask. This procedure avoided also the problem of excessive contamination of cask surfaces that could have occurred due to very high level of contamination of EUREX pool water if the cask had been immersed in the pool. Additional shielding devices were developed and used to reduce dose rate during cask loading operations. Although the evacuation of spent fuel assemblies from the EUREX pool was a very challenging activity due to the short time available, unfavourable space conditions inside the pool building and handling tool limitations; all loading and transport operations were performed successfully and without particular problems. Ten transports were carried out to evacuate all of the spent fuel stored in the EUREX pool. Spent fuel was transferred to the Avogadro Deposit pool. The first loading sequence started on 2 May 2007 and the first transport was performed on 6 May 2007. The tenth and last transport was performed on 21 July 2007. A dose less than 50 μSv (neutron + gamma) was measured for the most exposed operator during a complete cask loading sequence.     

Mechanical characteristics of fuel rod claddings in transport conditions

Abstract

A synthesis on the mechanical characteristics of unirradiated and irradiated fuel rod claddings was performed by the French Institut de Radioprotection et de Sûreté Nucléaire (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 M5^TM. 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 content and creep deformation rate. The following phenomena which could influence the mechanical behaviour of the cladding were more particularly studied: oxidation which could become very important during transport in case of cladding temperatures of ～500°C; creep for which only a few data ～500°C are available and which depends in particular on the internal pressure of the rods, the cladding oxidation and the presence of the hydrides; and recrystallisation of Zry-4 at ～500°C, which could have consequences on the mechanical properties of the cladding after cooling during the storage. For other topics of interest for the study of the mechanical behaviour of the cladding, such as the fracture toughness for example, it was identified that the data available is scarce.     

Looking to the future – a UK industry perspective

Abstract

The theme of the 10th PATRAM Conference has been ‘looking to the future’. This contribution aims to give a UK perspective on future issues and challenges. In doing so, I will give a short summary of UK transport experience before going on to discuss the future and in particular the challenges and opportunities facing the UK radioactive transport industry.     

BWR spent fuel transport and storage with TN9/4 and TN24BH casks

Abstract

In 2001 the Swiss nuclear utilities started to store spent fuel in dry metallic dual purpose casks at ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd, as the owner of the Mühleberg nuclear power plant, is involved in this process and has selected to store the spent fuel in a new high capacity dual purpose cask, the TN24BH. For the transport Cogema Logistics has developed a new medium size cask, the TN9/4, to replace the NTL9 cask, which has performed numerous shipments of BWR spent fuel in past decades. Licensed by the IAEA 1996, the TN9/4 is a 40 t transport cask, for seven BWR high burnup spent fuel assemblies. The spent fuel assemblies can be transferred to the ZWILAG hot cell in the TN24BH cask. These casks were first used in 2003. Ten TN9/4 shipments were made, and one TN24BH was loaded. After a brief presentation of the operational aspects, the paper will focus on the TN24BH high capacity dual purpose cask and the TN9/4 transport cask and describe in detail their characteristics and possibilities.     

Buckling of fuel rods under inertia loading

Abstract

The buckling analysis of fuel rods during an end drop impact of a spent fuel transportation cask has traditionally been performed to demonstrate the structural integrity of the fuel rod cladding or the integrity of the fuel geometry in criticality evaluations for a cask drop event. The actual calculation of the fuel rod buckling load, however, has been the subject of some controversy, with estimates of the critical buckling load differing by as much as a factor of 5. Typically, in the buckling analysis of a fuel rod, assumptions are made regarding the percentage of fuel mass that is bonded to or that participates with the cladding during the buckling process, with estimates ranging from 0 to 100%. The greater the percentage of fuel mass that is assumed to be bonded to the cladding, the higher the inertia loads on the cladding, and, therefore, the lower the 'g' value at which buckling occurs. However, these solutions do not consider displacement compatibility between the fuel and the cladding during the buckling process. By invoking displacement compatibility between the fuel column and the cladding column, this paper presents an exact solution for the buckling of fuel rods under inertia loading. The results show that the critical inertia load magnitude for the buckling of a fuel rod depends on the weight of the cladding and the total weight of the fuel, regardless of the percentage of fuel mass that is assumed to be attached to or participate with the cladding in the buckling process. Therefore, 100% of the fuel always participates in the buckling of a fuel rod under inertia loading.