Design and licensing of UKAEA IP 2m box

Abstract

Croft Associates Ltd (Croft) was contracted by the United Kingdom Atomic Energy Authority (UKAEA), Winfrith in 2006 to design, manufacture, test and license an intermediate level waste industrial package transport packaging (UKAEA 2 m Box Design no. 3954) for decommissioning waste from the Dragon and steam generating heavy water reactors at Winfrith. The package had to meet the requirements of the existing UK. Nuclear Decommissioning Authority (NDA) (Nirex) 2 m Box specification. The UKAEA 2 m Box is intended for on-site storage at Winfrith and possibly another site for 50 years and subsequent transport to the NDA repository without further processing. The design of the box closure includes a leak testable sealing system which is capable of being replaced at a later date (e.g. before shipment), by the removal of the lid and replacement of seals remote from the box location. Initially the specification for the UKAEA 2 m Box called for rating for gross weight of 40 t, but this was raised to 50 t as it was realised that the efficient use of the capacity of the box would result in a 50 t gross weight. Two prototype 2 m Boxes were produced; one being provided with 100 mm of concrete shielding, and the other being provided without any shielding. The shielded box was filled with simulated waste and a concrete grout cap fitted and the resulting 50 t package was dropped onto the massive target at Winfrith from a height of 300 mm (this being the regulatory test height). The test showed that the box performed as expected; meeting all requirements. In particular: there was no loss of shielding and no effect on the containment system that would reduce its effectiveness.     

Improved design methodology for an existing automated transportation system with automated guided vehicles in a seaport container terminal

In this paper, we propose an improved design methodology to meet the changing demands of an existing automated container transportation system in which automated guided vehicles (AGVs) are used. This system is called an AGV transportation system. To achieve an improved design, it is essential to detect and correct any occurring bottlenecks. For this purpose, we exhaustively enumerate design proposals by constructing a logic tree. As a case study to verify the proposed methodology, we apply the methodology to an existing AGV transportation system. From the enumerated design proposals, we suggest design policies by considering the actual constraints of the transportation system. Finally, we redesign the transportation system as rapidly as possible. On the other hand, we keep system balancing into account; then, we derive a suitable demand and input number of AGVs under given specifications for a transportation system.     

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.     

Catalytic mitigation of hydrogen risk during wet transportation of radioactive materials

Abstract

Major challenges in the area of wet transportation of radioactive materials are reliability and safety of transportation casks. In most cases, the bottom part of the cask is filled with water whereas a gaseous mixture is contained in the upper part. In such a configuration, water radiolysis leads to the formation of hydrogen and oxygen, which continuously enriches the gaseous mixture. Among the functions to be satisfied, wet transportation systems shall thus allow the control of the hydrogen content below its flammability limit. This is currently achieved by limiting the transportation duration so as to reopen the cask before the critical hydrogen concentration is reached. Development of new technologies that would mitigate the hydrogen risk is all the more motivated because it would allow an extension of the transportation duration. AREVA-TN International and the Institut de recherches sur la catalyse et l'environnement de Lyon have developed a catalytic system which aims at buffering the hydrogen concentration far below the flammability limit. The principle of this catalyst is to recombine the hydrogen with the oxygen formed by water radiolysis. The present paper gives an overview of the development of this catalytic recombining system. It describes the laboratory qualification tests undertaken for the evaluation of the recombining efficiency. Particular attention is placed on the recombining efficiency after immersion of the catalyst in borated water, which would occur in a nuclear reactor pool during loading of used fuel. Laboratory investigations, carried out in an autoclave simulating a transportation cask, showed that, after immersion in borated water, the catalytic system allows the recombination of 3% hydrogen in less than 24 h at temperatures as low as 35°C.     

The French Nuclear Safety Authority's Experience with Radioactive Transport Inspection

Abstract

About 300,000 radioactive material packages are transported annually in France. Most consist of radioisotopes for medical, pharmaceutical or industrial use, but the nuclear industry deals with the transport of fuel cycle materials (uranium, fuel assemblies, etc.) andwaste from power plants, reprocessing plants and research centres. France is also a transit country for shipments such as spent fuel packages from Switzerland or Germany, which are bound for Sellafield in the United Kingdom. The French nuclear safety authority(DGSNR, Directorate General for Nuclear Safety and Radioprotection) has since 1997 been responsible for the safety of radioactive material transport. This paper presents DGNSR's experience with transport inspection: a feedback of key points based on 300 inspections achieved during the past 5 years is given.     

Ten years of Romanian experience on risk assessment owing to radioactive materials transport in Romania

Abstract

The transport of radioactive materials is a very important problem considering the potential risks and radiological consequences in carrying out the present activity. Based on the International Atomic Energy Agency (IAEA)'s Safety Standard TS-R-1 (1996 edition, as amended 2003), Romanian National Nuclear Regulatory Body – Romanian National Commission for Nuclear Activities Control (CNCAN) was adopted and implemented by act no. 374/October 2001, the safety regulations for the transport of radioactive materials in Romania under the title 'Fundamental regulations for a safe transport of radioactive materials, in Romania'. The present paper will present the main sources of radioactive materials in Romania, their transport routes with a particular interest paid to the radioactive wastes. Hypothetical scenarios for specific problems related to the identification and evaluation of the risks and potential radiological consequences associated with the transport of radioactive materials in Romania, for all these situations: routine transport (incident free) and possible accidents.     

Mixed-Load Transportation Scheduling of Multiple Agents in a Warehouse Environment

In a floor warehouse environment, where many agents transport numbers of products iteratively, efficient product transportation is required for high productivity. In general, the following facts are recognized. (i) Three kinds of jobs exist: picking, replenishment and arrangement. (ii) Mixed-load transportation can decrease the completion time of vehicles. In many former methods, however, the above facts are not considered. We propose a design methodology using a dispatching rule and a local search method based on a simulated annealing algorithm. The rule is applied to obtain the initial solution and the local search method is adopted to obtain a semi-optimal solution in consideration of mixed-load transportation. To evaluate the applicability of the proposed algorithm, we have implemented the algorithm in several product environments. The simulation results verify the effectiveness of the proposed method.     

Transportation safety risk for source recovery versus consequence of leaving radioactive sources in place and vulnerable due to limited transport options or denial of shipment

Abstract

Potential risks associated with transportation safety of recovered radioactive sources in normal commerce are rhetorically compared to the latent risk of not recovering disused radioactive sources due to limited transport options or outright denial of shipment. It is essential, during each phase of the recovery process, to ensure secure, timely, cost effective and reliable means to return vulnerable radioactive sources to safe and protected locations by land, sea and/or air transport. In some cases, only limited transport options exist or denials of shipment may occur that impede the recovery process. Risks associated with normal transportation of recovered sources are considered less significant than the risks related to leaving disused radioactive sources at their original location.