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.     

Regulations on Transport of Radioactive Material in Spain

Abstract

There are several national and international regulations that have to be considered when a shipment of radioactive material is going to take place in Spain. Some of them are specific for the transport of dangerous goods and others are general and affect all activities involving radioactive material, as for instance regulations on radiological protection or liability insurance. All these regulations are described and how those on transport follow the IAEA recommendations. Furthermore, other points considered fundamental are dealt with: competent authorities and their responsibilities, the control of package designs and shipments and training of personnel.     

Modelling of release of particulate material from transport containers

Abstract

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 SWTC of typically 10^-9 to 10^-10. Hence, the effective decontamination factor provided by the SWTC is 10² to 10³. Whilst this analysis has been carried out specifically for the SWTC carrying waste packages, it is applicable to other arrangements and its use would reduce the high degree of pessimism used in typical containment assessments, whilst still giving conservative results.     

Chapter 3: Type and Quantity of Radioactive Waste Destined for Disposal

Abstract

According to the current planning status, radioactive waste with negligible heat generation is destined for storage in the Konrad waste repository in accordance with the Provisional Waste Acceptance Requirements⁽⁴⁾. Waste of this kind occurs in the nuclear fuel cycle, research, medicine and technology. In its original state, the primary wastes of this type have various forms, such as: liquids, concentrates, sludges; ion exchange resins; compressible and/or combustible solids; incompressible solids, e.g. structural material components; filters, filter candles; other types of waste. The radioactive waste is appropriately conditioned and packaged before transport to the waste disposal site. To meet the basic requirements specified in the Waste Acceptance Requirements, primary waste must be solidified, for which the most common solidifying agents are cement and concrete although bitumen is also used. Waste products in decomposing, fermenting or liquid form or which contain a significant fraction in such states are not accepted for disposal.     

The Transport System Approval Concept

Abstract

The needs for, and merits of, a new concept for the safety assessment and approval of shipments of radioactive materials is introduced and discussed. The purpose of the new concept is to enable and encourage integration of analysis and review of transport safety with similar safety analysis and review of the handling operations involving the radioactive material at the despatching and receiving ends of a shipment. Safety contributing elements or functions of the means of transport (the Transport System) can thus readily be taken into account in the assessment. The objective is to avoid constraints — experienced or potential — introduced by the package functional provisions contained in the transport regulations, whilst maintaining safety during transport, as well as during facility handling operations, at least at the level currently established.     

Transport of Radioactive Waste in Poland

Abstract

General information is given about the regulations and limits concerning radioactive waste in Poland. Radwaste—being in 95% low level—comes at present from one research reactor and over 2000 smaller producers; there is no high level waste. The responsibility for collecting, handling and disposing of all radwaste is delegated to one organisation partially supported by the state. The frequency of transport to the Central Repository is about 50 times a year. The total volume of conditioned radwaste is about 200 m³.     

Regulations for the Transport of Radioactive Materials in Poland

Abstract

The paper gives general information on the subject of transport of radioactive material. The responsibilities and tasks of the competent authorities for transport of radioactive materials in Poland, the Ministry of Transportation and Maritime Economy, and the President of National Atomatic Agency, are specified. Regulations applied for different modes of transport and other regulatory requirements related to transport of radioactive materials are described.     

Testing of the Sealing Arrangements for the Nirex Re-Usable Shielded Transport Containers

Abstract

As part of its responsibility for the development of a deep repository for intermediate and low level radioactive waste, UK Nirex Ltd is developing a range of Type B re-usable shielded transport containers (RSTCs). A testing programme has been carried out on two alternative concepts for the RSTC sealing arrangements over the temperature range ?40^°C to 200^°C. For each sealing system, a test rig was developed to measure the performance under simulated normal and accident conditions of transport. The elastomer O-rings used for some of the tests had been irradiated to the maximum dose they might receive in normal transport. The performance of both sealing systems was good and it is concluded that either concept would meet the specified leakage criteria over the full temperature range under both normal and accident conditions of transport. However, further testing is required to confirm the performance of Concept N under accident conditions.     

A Screening Model for Calculating the Risks of Transporting High Level Radioactive Waste

Abstract

In transporting high level radioactive waste (including spent nuclear fuel), shippers (and sometimes carriers) need to evaluate the risks of potential radiation exposure to the public and transport workers. A simple model is presented that can be applied to nuclear waste transport risk assessments. The model considers radiation risks arising from incident free exposure, accidental release-caused exposure to on-link population, off-link population, crew, transport workers, etc. Important parameters and factors that affect the radiation dose level are grouped using the physics of the different exposure phenomena. The total radiation risk (in person-rems) is given by a linear combination of the groups of these factors, each representing a different type of exposure. The radiation exposure risk assessment is reduced to the evaluation of a single linear algebraic equation containing five distinct terms and each term containing the groups of parameters and a constant coefficient. The estimation of the values of the constant coefficients was accomplished by selecting a sample of 65 origin-destination (0-D) pairs and simulating the shipment of high level nuclear waste or spent nuclear fuel between each O-D pair, and evaluating the radiation dose risks to each group of population. RADTRAN 4 was used in the detailed assessments. The coefficient values were tested for statistical robustness using a sampling hypothesis and t-statistics. These values are presented. The simplified model presented here represents a viable and economical option as a radiological risk assessment tool, to be used in mode or route options screening.     

Analysis of the Regulations for the Transport of Radioactive Materials in the FRG and Application to Uranium Ore Concentrate and UF6

Abstract

A survey of the regulations ruling the transport of radioactive materials in Germany is given. Two distinct areas of legislation exist: (1) the nuclear laws which in respect to transport deal with approvals, liability, physical protection, import/export and which specify the competent authority; and (2) the Dangerous Goods legislation which regulates the transport of all dangerous goods including Class 7, Radioactive Materials. Mode-specific ordinances deal mainly with packaging requirements, labelling/marking and other controls for shipment. They are practically identical with the relevant IAEA recommendations. The paper also describes how these German regulations are applied to uranium in the form of ore concentrates of UP₆.     

Radioactive Transportation,Clarity and Consensus: IBC Conference, 5–6 October 2006, London,UK

Abstract

It is believed that there will be shipments of radioactive material and waste, as the nuclear industry moves into maturity, which may not be well accommodated within the current transport regulations. Although these shipments could be made under Special Arrangement approvals, the regulatory system would be better served if more formal requirements and criteria were included in the regulations and the shipments were considered in accordance with the regulations. A Special Arrangement approval is now defined as authorising transport of a shipment which does not satisfy all the applicable requirements of the regulations. This paper proposes the Transport System approach to regulating these types of shipments, where operational restrictions or other packaging provisions could compensate for the absence or inadequacy of packaging or other associated requirements. These shipments would require Competent Authority approval, and acceptance criteria would be included in terms of limits on probability, consequences and risk. The process would be limited to those types of shipments where the package system does not work well. The advantages of including Transport System approval within the regulations include reduction in the time required to obtain an approval, greater efficiency of decontamination and decommissioning operations, and assurance of an equivalent level of safety.     

The International Legal Position on Transboundary Shipments of Radioactive Waste

Abstract

The recent decision not to grant planning permission for construction of a Rock Characterisation Facility near Sellafield has reopened the question of long-term radioactive waste disposal policy in the UK. One possible solution would be the construction and operation of a small number of international radioactive waste disposal facilities, taking waste from several countries. Such an approach would allow pooling of international expertise; would allow the choice of excellent sites from geological and demographical standpoints; and may be economically attractive depending on economies of scale. However, the approach would also increase the amount of waste transport, and may reduce the pressure on producers to reduce the volumes of waste arising. This paper traces the development of international legal attitudes to transboundary transport of radioactive and other hazardous waste. It concludes that as international law now stands it would be very difficult to establish a network of international waste disposal facilities, and therefore strategies which are developed will be nationally based.     

Marine Transport of Irradiated Nuclear Fuel,Plutonium and High Level Wastes

Abstract

On 4 November 1993 the Assembly of the International Maritime Organization (IMO) adopted the Code for The Safe Carriage of Irradiated Nuclear Fuel, Plutonium, and High-level Radioactive Wastes in Flasks on board Ships. The Code was recommended for adoption by a Joint IAEA/IMO/UNEP Working Group and sets standards for the survival capability of ships carrying those materials. It aims to complement the packaging requirements and the shipment procedures imposed by the IAEA's Regulations for the Safe Transport of Radioactive Material and IMO's International Maritime Dangerous Goods Code (IMDG-Code). The Joint Working Group also considered a number of issues related to accidents at sea, accident statistics, risk studies and emergency response. The Group concluded that all the available information demonstrates very low levels of radiological risk and environmental consequences from the marine transport of radioactive material. In this paper the details of the Code and the highlights of the deliberations regarding marine transport of radioactive material are discussed.     

Probabilistic Safety Assessment for the Transport of Radioactive Waste to a UK Repository at Sellafield

Abstract

A study has been undertaken to provide a detailed understanding of the radiological and non-radiological risks associated with the transpott of radioactive waste from the sites at which waste is produced in the UK to a proposed deep repository at Sellafield, and to ensure that these risks meet the design targets specified by Nirex. The routine transport collective dose to members of the public was assessed to be 0.2 man.Sv per year, which is only about 0.004% of the natural background dose. Accident frequencies were calculated using event tree methodology. The radiological consequences of accidents were assessed using the probablistic computer code CONDOR. The risk expectation value was calculated to be 1.5 × 10^?5 ? 8.6 × 10^?6 latent cancer fatalities per year (depending on the transport mode scenario). These values are significantly lower than the corresponding prediciions for non-radiological accident fatality rates, 0.05 ? 0.035 fatalities per year. The radiological accident risk for the most exposed individual member of the public was assessed to be 5 × 10^?11 ? 1.7 × 10^?11 per year, very much less than the Nirex target of 5 × 10^?7 per year. Plots of societal risk were shown to lie in the region of ‘negligible risk’, as defined by the UK Health and Safety Commission for non-radioactive dangerous goods transport.     

Challenges in the renewal of the Swedish sea transport system

Abstract

Since 1985, SKB has successfully operated a sea transport system for transport of spent nuclear fuel and radioactive waste to the intermediate storage facility, Clab and the final repository, SFR, in Sweden. The main components in the system are the ship M/S Sigyn, transport casks for spent fuel and core components, IP2 containers and terminal vehicles.     

Chapter 1: Principles and Objectives

Abstract

Like all activities in our civilised world the transport of radioactive waste on public routes or routes to which the public has access (e.g. rail/road) entails some elements of risk — nuclear and nonnuclear — for man and his environment. The risks involved in waste transport comprise the radiation exposure of the population and transport personnel from normal, accident-free transport, as well as from possible transport or handling accidents with the potential of causing radiation exposure of people and contamination in the surrounding area.     

Development of Sealed Type B Transport Containers for the Transport of Intermediate Level Radioactive Waste

Abstract

UK Nirex Ltd have developed three possible concepts of sealed transport containers for the safe transport of immobilised intermediate level radioactive waste. Computer based finite element impact and thermal analysis has been carried out on each concept and compliance with both the IAEA regulatory requirements and specified Nirex design aims has been demonstrated. One single concept will be selected at a later date following further development and confirmatory testing to provide a fleet of transport containers.     

Verification of activity release compliance with regulatory limits within spent fuel transport cask assessment

Abstract

Admissible limits for activity release from type B(U) packages for spent fuel transport specified in the International Atomic Energy Agency regulations (10^?6 A₂ h^?1 for normal conditions of transport and A₂ per week for accidental conditions of transport) have to be kept by an appropriate function of the cask body and its sealing system. Direct measurements of activity release from the transport casks are not feasible. Therefore, the most common method for the specification of leak tightness is to relate the admissible limits of activity release to equivalent standardised leakage rates. Applicable procedure and calculation methods are summarised in the International Standard ISO 12807 and the US standard ANSI N14·5. BAM as the German competent authority for mechanical, thermal and containment assessment of packages liable for approval verifies the activity release compliance with the regulatory limits. Two fundamental aspects in the assessment are the specification of conservative design leakage rates for normal and accidental conditions of transport and the determination of release fractions of radioactive gases, volatiles and particles from spent fuel rods. Design leakage rates identify the efficiency limits of the sealing system under normal and accidental transport conditions and are deduced from tests with real casks, cask models or components. The releasable radioactive content is primarily determined by the fraction of rods developing cladding breaches and the release fractions of radionuclides due to cladding breaches. The influence of higher burn-ups on the failure probability of the rods and on the release fractions are important questions. This paper gives an overview about methodology of activity release calculation and correlated boundary conditions for assessment.