IP-2 package containing plutonium contaminated slag pots

Abstract

Croft Associates were approached by PC Richardson to carry out the licensing of an industrial package, IP-2 package containing plutonium contaminated slag pots. This project provided several unique challenges for all those involved. The slag pots belonged to Outokumpu a steel company based in Sheffield. During use, sampling indicated a high level of plutonium in the form of Pu-238 was present in a slag pot. Further sampling identified that four slag pots were contaminated with Pu-238 and required disposal. The source of the Pu-238 contamination was believed to be a single heart pacemaker present in scrap that was melted down for reuse. All four slag pots weighed between 50 and 65 t and three of the pots had large patched cracks. The size and nature of the contamination prevented the slag pots being broken up therefore they had to be packaged and transported intact. REMAC designed a steel box to package the slag pots. It was the responsibility of Croft to license this package. Several challenges were faced when licensing: first, the weight, size and contents prevented any physical drop testing of the package; second, the box was to be assembled around the slag pot, limiting the leak testing that could be carried out. All of the evaluation of the ability of the package to withstand the regulatory drop tests was therefore carried out by finite element analysis by AMEC NNC. This approach was also used to check the tie down and lifting systems. Leak testing was carried out via the soap bubble method on the assembled box before grouting. The transportation box, once assembled and licensed as an IP-2 package, was transported to Drigg from Sheffield by road.     

Development of a 500 litre drum for the encapsulation of steam-generating heavy-water reactor sludge at Winfrith

Abstract

Winfrith Technology Centre was once a leading UKAEA development site for nuclear technology. UKAEA's task now is to decommission the nuclear reactors and other facilities and restore the site for alternative use. On the site is the prototype steam-generating heavy-water reactor (SGHWR) that produced 100 MW(e) of electricity during its 22 year operational life. During this period the reactor produced large quantities of radioactive sludge and there are also the remains of ion exchange resins from various clean-up operations including the circuit decontamination campaigns at each annual shutdown. These sludges were directed to and stored in four external tanks and over the years there has been a steady build-up of sludge in these facilities, until 1990 when the reactor shut down permanently. Plans were made for the sludge in these tanks to be retrieved, encapsulated in drums and stored on site until a permanent national repository became available. Due to changes in circumstances following the relatively sudden closure of the SGHWR in 1990 and additional requirements from the operators of the Drigg low-level waste site, the original encapsulation plans of UKAEA had to be set aside. Following further considerations, RWE NUKEM, working in partnership with UKAEA, is now contracted to retrieve, condition and encapsulate the sludge in a new plant currently being constructed, and to export the drums to an existing refurbished on-site store. The Winfrith treated radwaste store (TRS) was constructed to store 500 I drums of intermediate-level waste in a matrix stacked nine high. This paper describes the drum development work undertaken prior to the introduction of RWE NUKEM and completion of the revised design of drum for use within the TRS. It also briefly describes the process for which the drum is being utilised in the newly designed sludge treatment plant. The drum design has had a number of iterations from a concept that was first drafted in 1989 to the present design that has undergone extensive finite-element modelling to support its introduction. Although the drum has a fairly standard body, the flange and lid have been modified to suit a revision to the original concept of attaching the lid by welding. This paper tracks the development of the lid bolting arrangements and the impact on the revised design of parts of the new encapsulation facility. The means of satisfying the requirements for the introduction of the new drum as an IP2 package are also described.     

Pilot study on return of reprocessed wastes from Dounreay,Scotland to Germany

Abstract

The German Federal Ministry of Education and Research (BMBF) assigned DBE Technology GmbH with a project to review the prerequisites and contractual boundary conditions for the return of cemented residues from the reprocessing plant at Dounreay to Germany. For this purpose, the bilateral contracts between the German research facilities and the operator of the reprocessing plant at Dounreay, the UK Atomic Energy Authority (UKAEA), were examined. Possible interim storage sites in Germany were sought, flasks suitable for transport and casks suitable for interim storage and final disposal were researched, and transportation options were explored. Based on the results of theses investigations, strategies for the return of the drums containing cemented residues were developed, including time and effort estimates.     

Development of packaging to transport new standardised range of disposal canisters to geological disposal facility in UK

Radioactive Waste Management Limited (RWM) of the Nuclear Decommissioning Authority (NDA) is developing concepts to demonstrate the viability of using a standardised range of disposal canister (DC) designs for geological disposal of high level waste and spent fuel in the UK. The standardised DC are designed for disposal in a geological disposal facility with integrity requirements in the range 10?000 to 100?000 years. International Nuclear Services (INS) is also a subsidiary of the NDA and working with RWM to develop a design of packaging for transporting these DC, which is called the disposal canister transport container (DCTC). Initial studies undertaken by INS focused on optimising payload and geometry for the canister designs. Subsequent studies focused on achieving criticality safety requirements for transport, which established the use of multiple water barriers, were required for higher enriched spent fuels. The results of this initial work were presented at the International Nuclear Engineering society conference at London in 2012. Subsequently, RWM commissioned INS to develop the design of DCTC to a level where it would be viable for licensing as a transport package with appropriate level of technical understanding. A specific requirement of RWM was that the loaded DCTC should be capable of transportation on an existing design of four axle rail wagon, within a gross mass of 90 t, this giving considerable logistic and overall cost benefits. Recent development work has focused on detailed impact, thermal and shielding analysis and how these influence the DCTC transport mass and the position of that mass in relation to the four axle rail wagon, both of which influence its capability for the required transport. In terms of meeting mass limits, achieving the specified radiation shielding performance (neutron and gamma) for the spent fuel was found to be a major challenge. However, of equal challenge was to accommodate the high forces generated under impact accident conditions due to the high mass ratio of contents to container. In order to mitigate these forces, the shock absorber designs needed to be carefully judged because their dimensions were restricted by the rail wagon design. This paper describes the DCTC development work, how the design challenges were addressed and the conclusions reached.     

Conceptual Design of Large Boxes for the Packaging of Intermediate and Low Level Radioactive Waste

Abstract

United Kingdom Nirex Limited (Nirex) is developing standard containers for the packaging of radioactive waste for disposal in a deep underground repository. Waste Package Specifications have been produced for each standard package to provide the essential link between waste package design and the design of the deep repository. Previous studies carried out by Nirex identified the dimensions and key features of standard boxes for decommissioning intermediate level waste and for low level waste: the 4 m ILW box, 4 m LLW box·and 2 m LLW box. Nirex has now produced conceptual designs for these boxes.     

Evaluation of safety of French type B package designs in severe accident environments other than regulatory

Abstract

Packages used to transport radioactive materials in France must be designed to meet the safety performance requirements when subject to the test conditions set forth in the International Atomic Energy Agency (IAEA) Regulations. During actual use, the packages may be subject to quite different accident conditions. The Institut de Radioprotection et de Sûreté Nucléaire (IRSN) has evaluated the behaviour of typical packages designed to transport spent fuel, high activity waste, fresh mixed oxide (MOX) fuel and plutonium oxide powder under realistic conditions of mechanical impact and fire. The studied designs remain safe after impact onto targets present in the real environment of transport. The energy absorption by the package ancillary equipment (transport frame) compensates for the kinetic energy increase by comparison to the energy expended during the regulatory tests. New software was developed to correctly simulate the thermal behaviour of the neutron shielding materials. The studied package designs exhibit large margins of safety concerning resistance to fire. The results obtained have been used to develop tools in support of the appraisal of emergency situations.     

A summary of recent package testing activities at Oak Ridge National Laboratory

Abstract

The history of testing of radioactive material packages at Oak Ridge National Laboratory (ORNL) dates back to the early 1960s, and includes the testing of hundreds of different packages of all shapes and sizes. This paper provides an overview of ORNL's new Packaging Research Facility at the National Transportation Research Center (NTRC), and describes recent package testing successes conducted at the NTRC from September 2002 to September 2003. This paper also provides an overview of the package testing capabilities available at NTRC. Between 2002 and 2003, ORNL conducted tests on the following packages: rackable can storage box (RCSB); ES-2100; DT-20; DPP-2; BRM shielded overpack; Fernald Silos IP-2 waste package; and RAJ II BWR fresh fuel package. Tests of the RCSB, a storage package for highly enriched uranium, involved two test specimens, dropped from 28 ft(8.4 m) in different orientations. The ES-2100 and DPP-2 involved four and six test units, respectively, subjected to the entire Type B normal conditions of transport and hypothetical accident conditions testing sequence, including thermal tests. A single DT-20 package was subjected to a subset of the Type B tests to confirm package performance. The BRM shielded overpack, weighing about 500 kg, was subject to the Type A package tests. Three Fernald Silos waste package test units — a large package weighing about 10,000 kg for shipping grouted waste removed from the Fernald site — were subjected to IP-2 tests. And finally, two RAJ II boiling water reactor fresh fuel test units were subjected to Type B 9 m drop and 1 m puncture tests.     

Design and Testing of the Nupak-200 Type-B(U)F Packaging

Abstract

The design and testing of the Nupak-200 Type-B(U)F packaging developed by AEA Technology with support from the United Kingdom Atomic Energy Authority (UKAEA) is described. The design concept and operational criteria are discussed, and the packaging's principal features are reviewed. The adaptability of the two-box assembly is examined, as is its application to current and future radioactive materials transport requirements. The paper highlights the test programme that was devised to simulate the impalement of composite steel and cork panels on a punch of the type used for testing to IAEA standards. This was done in order to generate adequate data about the materials' characteristics for the finite analysis programs which supported the design. The paper concludes by reviewing the drop-test. performance of the package and compares the computer analyses with the actual test results.     

Regulatory Aspects of the Transport of High Radiation Level and Alpha Waste in France

Abstract

The introduction of the 10 mSv.h^?1 at 3 m limit for LSA unshielded material makes it impossible to transport, as LSA material, the highest radiation level wastes from EdF PWR's operations. At present, the EdF's waste blocks can be transported as LSA III material by special arrangement. A new package design, equivalent to a Type B package, will be available for their transport before the end of the year 1995. It consists of a re-usable steel cylinder over-packing each block. Compliance of this package model with transport safety requirements will be demonstrated by taking into account the non-dispersability, as LSA III material, of the irradiating waste. The transport of LSA combustible material is affected, in the revised regulation, by the new limitation to 100 A2 of the total activity per conveyance. Complying with this limit would greatly reduce the quantity transported by conveyance and therefore multiply the number of journeys, which is not desirable. A two-step approach has been accepted by the French Competent Authority for the transport of these wastes: (1) A specific ISO 20 container, thermally insulated, can be used by special arrangement for the transport of LSA combustible material having a total activity per conveyance higher than 100 A2. Furthermore, additional safety measures have to be implemented for these consignments. (2) After the end of 1995, a Type B package must be used for activity contents per conveyance higher than 100 A2. A specific 20′ ISO container, complying with Type B requirements, is being developed for that purpose. The total plutonium mass transported per conveyance will be limited to 400 g for criticality and physical protection considerations. An interpretation of the general LSA requirements formulated in the current IAEA Regulations is presented with respect to the homogeneity of the radioactive material, and the definition of the unshielded material.     

New generation standard nuclear material container

Abstract

Nuclear Filter Technology (NucFil) is working with the Los Alamos National Laboratory (LANL) to design a nuclear material storage container that complies fully with the requirements of DOE M 441·1-1. LANL provided NucFil with a specification that outlines requirements to comply with the manual, as well as to satisfy specific needs of their own. NucFil has taken this specification and designed a container known as the new generation standard nuclear material container (NG SNMC). The premise of the design is a simple, robust container that is easy to use. The sealing mechanism is a single large cross-section, low durometer o-ring. The large cross-section provides a tight seal that has enough elastic rebound to compensate for any distortion of the sealing faces after a potential drop. The low durometer keeps the force required to open and close the container low. Once compressed, the seal is kept in place by a bayonet style closure that is locked in place by a positive mechanical engagement. The components of the container exposed to the load are manufactured of corrosion resistant 316L stainless steel. The container has a filter made of a heat resistant ceramic fibre to retain particles after a fire, and a water resistant membrane to keep moisture out of the container. Pewter shielding can be attached and is latched in place. These features are present in all seven sizes of the NG SNMCs, including 1, 3, 5, 8 and 12 quart and 5 and 10 gallon.     

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.     

Mechanical design assessment approaches of actual spent fuel and HLW transport package designs

Abstract

In recent years, BAM Federal Institute for Materials Research and Testing finalised the competent authority assessment of the mechanical and thermal package design in several German approval procedures of new spent fuel and high level waste package designs. The combination of computational methods and experimental investigations in conjunction with materials and cask components testing is the most common approach to mechanical safety assessment. The methodology in the field of safety analysis, including associated assessment criteria and procedures, has evolved rapidly over the last years. The design safety analysis must be based on a clear and comprehensive safety evaluation concept, including defined assessment criteria and constructional safety goals. In general, for new package designs, the implementation of experimental package drop tests in the approval process should be obligatory. Additionally, pre- and post-test calculations as well as components or material testing could be important. The extent to which drop tests are necessary depends on the individual package construction, the materials used and identified safety margins in the design.     

An overview of 20 years' Romanian experience on packaging and components testing according to national and international regulations

Abstract

More than 20 years ago, the Institute for Nuclear Research (INR) Pitesti in Romania, through its Reliability and Testing Laboratory, was licensed by the Romanian Nuclear Regulatory Body – CNCAN to carry out qualification tests for packages intended for the transport and storage of radioactive materials. The radioactive material is placed in packaging which are designed in accordance with national and the International Atomic Energy Agency's (IAEA's) Regulations for safe transport to the disposal centre. A broad range of verification and certification tests are performed at INR on radioactive material packages or component sections, such as packages used for the transport of radioactive sources to be used for industrial or medical purposes. This paper describes some of the various tests, which have been performed, and how they relate to normal conditions and minor mishaps during transport. Quality assurance and quality control measures taken in order to meet technical specification provided by design there are also presented.     

Service-oriented architecture for an overall radioactive waste package record management system

In Taiwan, there are a few radioactive waste package record management systems independently implemented by radioactive waste generators, operators of waste management facilities, transport organizations and storage facilities. To claim compliance of a radioactive waste package record meets with waste acceptance criteria for disposal, an overall radioactive waste package record management system which records and tracks all relevant information, from raw waste characteristics, through changes related to waste processing, to final checking and verification of waste package parameters should be constructed in accordance with IAEA recommendation. Service-Oriented Architecture (SOA) is widely accepted as a new paradigm for integrating heterogeneous systems in an effective way. It has formed a new trend being adopted by organizations in mitigating legacy system problems as in to maximizing interoperability, reusability and flexibility. Based on the current radioactive waste management processes, this paper proposes a three-tier SOA for the further overall radioactive waste package record management system design.     

Ten years of transport of vitrified high level waste (HLW) from Areva La Hague

Abstract

Areva has extensive experience in reprocessing irradiated fuel from both gas cooled and water cooled reactors and has legal obligations to return residues to respective foreign owners, usually in the form of vitrified waste. Such returns are subject to appropriate quality assurance controls. Appropriate transport flasks have been designed and certified. Transport operations are routinely conducted to Belgium, Germany, Japan, Switzerland and the Netherlands. The present paper reviews the experiences over a ten year period from 1995 to 2005.     

Preliminary Shielding Analysis for HCCB TBM Transport

A preliminary shielding analysis on the transport of the Chinese helium cooled ce?ramic breeder test blanket module (HCCB TBM) from France back to China after being irradiated in ITER is presented in this contribution. Emphasis was placed on irradiation safety during trans?port. The dose rate calculated by MCNP/4C for the conceptual package design satisfies the relevant dose limits from IAEA that the dose rate 3 m away from the surface of the package con?taining low specific activity III materials should be less than 10 mSv/h. The change with location and the time evolution of dose rates after shutdown have also been studied. This will be helpful for devising the detailed transport plan of HCCB TBM back to China in the near future.     

中放废物坑扩孔操作的石英砂屏蔽实践

为了回取小坑口废物坑中的废物,利用了石英砂的流动性和吸尘器的抽吸功能,用石英砂屏蔽表面辐射水平为4.28 Gy/h的坑内中放废物,将坑口辐射水平降至4.49 μGy/h 后,安全地实现了扩孔操作,实施屏蔽和拆除屏蔽过程都实现了远距离操作,极大降低了人员所受剂量,屏蔽实施与拆除操作人员的集体剂量仅为0.45人·mSv。     

IRSN's experience feedback list for assessment of transport package design safety reports

Abstract

Since 1996, the Institute for Radiation Protection and Nuclear Safety (IRSN), the technical support organisation of the French Competent Authority for the safety of transport of radioactive materials, has recorded the list of the difficulties most frequently encountered during the assessment of the safety reports of package designs. This experience feedback list takes into account the most recent evolutions of the regulations and the latest technological knowledge. For instance the safety reports should include the analysis of the most unfavourable configurations such as the 1 m free drop onto the bar when the package is in oblique position, the 9 m drop test of a package with slapdown, the thermal dissipation under a tarpaulin or canopies, the brittle fracture analysis at –40°C. IRSN's experience feedback list for transport package designs, which is published annually, is used as a guide by applicants to improve their package design safety reports and by IRSN for their assessments. Recently, it has been integrated in the French transport applicant guide and in the European technical guide for drafting the package design safety reports.     

Qualification Tests of Packages used for Transport and Storage of Low Activity Radioactive Wastes In INR Pitesti

Abstract

Radioactive wastes generated by the TRIGA INR research reactor are packaged according to the national and international rules and standards. The technology for packaging and treatment of radioactive wastes can also be used at the Nuclear Power Plant Cernavoda. The qualification tests for the package used for transport and storage of radioactive wastes (low activity, up to 6·07 GBq (0·164 Ci) per drum) are described. The package used is a drum manufactured from 1 mm thick mild steel with the dimensions: height 915 ± 10 mm; diameter 600 ± 5 mm; volume 220 litres. To achieve adequate safety in the transport of radioactive wastes strict precautions must be taken to protect transport workers, the general public and the environment by ensuring, both in normal and in accident conditions, adequate containment and shielding of the materials, according to the IAEA Regulations for the Safe Transport of Radioactive Materials. The adequacy of the package design is therefore of primary importance, the design requirements being supplemented by careful construction, quality assurance and inspection procedures. Taking into consideration the above requirements, qualification tests for the prototype package were carried out. These tests include compression, penetration, free fall, leaching, safety in use (biological protection), checking of chemical and mechanical characteristics, and the effect of the product on the environment. Performance of these tests, and the results obtained, prove that our technology for treatment and packaging of radioactive waste is in accordance with international rules.     

The top-down reflooding model in the CATHARE code

A top-down reflooding model was developed for the French best-estimate thermal hydraulic code cathare. The paper presents the current state of development of this model. Based on a literature survey and on compatibility considerations with respect to the existing cathare bottom reflooding package, a falling film top-down reflooding model was developed and implemented into cathare version 1.3U. Following a brief review of previous work, the paper describes the most important features of the model. The model was validated with the Winfrith single-tube top-down reflooding experiment and with the REWET-II simultaneous bottom and top-down reflooding experiment in a rod bundle geometry. The results demonstrate the ability of the new package to describe the falling film rewetting phenomena and the main parametric trends both in a simple analytical experimental set-up and in a much more complex rod bundle reflooding experiment.