Failure model and Monte Carlo simulations for titanium (grade-7) drip shields under Yucca Mountain repository conditions

A failure model was developed for the titanium alloy drip shield proposed for the Yucca Mountain nuclear waste repository. The degradation modes considered are hydrogen-induced cracking (HIC) and general aqueous corrosion, processes which are inextricably linked. Failure by HIC is controlled by the environment, the corrosion rate, the material properties, and the hydrogen absorption efficiency which is assumed to decrease parabolically with time. This model includes both oxygen and water reduction coupled to corrosion, and allows for the release of the absorbed hydrogen as the alloy containing the hydrogen converts to oxide. Monte Carlo simulations were employed to predict drip shield lifetimes, and to investigate the effects of the hydrogen absorption efficiency, the critical HIC concentration, the corrosion rate, and the fraction of corrosion supported by water reduction, on the susceptibility of the material to HIC.     

Second generation waste package design and storage concept for the Yucca Mountain Repository

The reference waste package design and operating mode to be used in the Yucca Mountain Repository is reviewed. An alternate (second generation) operating concept and waste package design is proposed to reduce the risk of localized corrosion of waste packages and to reduce repository costs. The second generation waste package design and storage concept is proposed for implementation after the initial licensing and operation of the reference repository design. Implementation of the second generation concept at Yucca Mountain would follow regulatory processes analogous to those used successfully to extend the design life and uprate the power of commercial light water nuclear reactors in the United States. The second generation concept utilizes the benefits of hot dry storage to minimize the potential for localized corrosion of the waste package by liquid electrolytes. The second generation concept permits major reductions in repository costs by increasing the number of fuel assemblies stored in each waste package, by eliminating the need for titanium drip shields and by fabricating the outer container from corrosion resistant low alloy carbon steel.     

Copper canister lifetime limited by a sulfide intrusion in a deep geologic repository

Korea has continuously implemented an ambitious nuclear energy deployment program since 1978. Korea currently operates 20 units, 16 PWRs and four CANDUs and constructs four and reviews license application of two more units. Also, Korea plans to build two more units by 2016. In addition, according to the new “Green Growth Plan while reducing the emission of carbon dioxide” Korea will introduce 10–12 units by 2030. This will inevitably result in more burdens on the safe management of spent nuclear fuels. Korea Atomic Energy Research Institute has developed a final disposal concept for Spent Nuclear Fuel (SNF) named KRS. KRS proposes to emplace SNF in a deep geologic formation such as a crystalline rock. Two key engineered barriers are applied to retard the potential release of a radionuclide from an embedded SNF; a waste container and an engineered barrier. Such an engineered barrier is composed of domestic calcium bentonite and the waste container is composed of an outer copper layer and an inner steel layer. The outer layer, a copper layer is dedicated to protect a waste container against corrosion. The main corrosion mechanism to corrode a copper waste container is a pitting whose speed of corrosion is 5–25 times higher than that of a uniform corrosion. In this paper, a special mass transfer resistance model is developed to predict the migration of sulfide from a fracture to a waste container surface via a bentonite layer. Based on it the lifetime of a copper canister layer limited by a pitting corrosion is estimated. Results show that under normal conditions, a copper layer can sustain its integrity for up to more than millions of years.     

Corrosion issues in nuclear waste disposal

This paper summarized some corrosion issues specific to nuclear waste disposal and illustrates them by the French geological clay concept for the reliable prediction of container degradation rate and engineering barrier integrity over extended periods, up to several thousands years. Among the items, the following are included:

• The importance of the underground repository conditions.

• The necessity of developing comprehensive semi-empirical models and also predictive models that must be based on the mechanisms of corrosion phenomena.

• The use of archaeological artefacts to demonstrate the feasibility of long term storage and to provide a database for testing and validating the models.

Investigations concerning the semihydrostatic emplacement of waste packages in vertical boreholes

The semihydrostatic emplacement of waste packages in unlined vertical boreholes in salt domes involves the uncontrolled deposition of the packages in the borehole and their complete enclosure in crushed salt. In order to calculate the pressure distribution in the backfilled borehole section, the borehole filling composed of packages and crushed salt is treated by approximation as a homogeneous medium in the theoretical model. The eigenvalues of the pressure tensor depend on the radial distance r from the borehole centre and on the depth z in the borehole and are thus directionally dependent. Emplacement-typical crushed-salt parameters such as lateral pressure coefficient and friction coefficients were experimentally determined. The calculations show that, according to the semihydrostatic model, the maximum pressure in the borehole does not exceed 107 kPa for a 4001 waste package, that this maximum pressure is already reached at a depth of 9.26 m on the borehole wall and that there is an optimum borehole radius for which stressing of the emplaced packages is minimized.     

An experimental procedure for measuring accelerations and strains from a tie down system of a heavy nuclear transport package during a rail journey

Abstract

The transportation of nuclear waste and new nuclear fuel is an important aspect in sustaining the generation of electricity by nuclear power. The design of packages that satisfy regulatory requirements for normal operating and accident conditions is a complex engineering challenge. The ancillary equipment used to constrain the packages to their conveyance, a tie down system, is part of a multicomponent system used to transport packages. Traditionally, the individual components of the transport system have been designed in isolation. This approach does not account for the interaction between components of the system such as the conveyance, tie down system and package. The current design process for tie down systems is well established but, due to its heuristic development, suffers from uncertainties over which loading conditions should be applied. This paper presents a method for collecting measured acceleration and strain data that can be used to derive customised load cases for the design of tie down systems during rail transportation. The data was collected from a tie down system that restrained an empty TN81 package, weighing 99·7 tonnes during a routine rail journey from Barrow-in-Furness to Sellafield. Furthermore, the data can be used to validate modern computer models, allowing for the development of the previously described holistic approach to tie down system design. The results are unique because an ensemble of acceleration and strain time histories from a transport system laden with a nuclear package is unprecedented. A visual examination indicates that the loading a tie down system incurs during a rail journey consists of low magnitude accelerations. The measurement points also show that the general trend of acceleration levels is highest nearest the track and is attenuated by the package. The implications for the design of tie down systems are that two potential failure modes, fatigue and static strength, have been identified. The data provides scope for customising accurate static strength and fatigue calculations using modern computational techniques. This allows for the safety margins inherent in new designs to be determined and optimised design solutions made possible.

INS makes no representations or warranties or any kind concerning this article, express or implied, statutory or otherwise, including without limitation, warranties of accuracy or the absence of errors.     

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.     

Waste Package and Shipment Aspects in Relation to Transport and Disposal Requirements

Abstract

In the management of radioactive waste, different processes have to be considered such as conditioning, interim storage and final disposal together with transport as the linking process. Attention should be paid to all the relevant steps within these processes, in particular to derive appropriate waste package requirements for a safe waste management system as well as to obtain a consistent regulatory framework. Radioactive waste arising from research and development centres, nuclear power plant operation, decommissioning, the nuclear fuel cycle industry, and applications of radioisotopes in medicine, industry and research, has finally to be shipped to a final disposal site. Therefore waste packages are subject to both the regulatory requirements of transport and the requirements of disposal. Resulting consequences for waste package limitations will be discussed, in particular for low and intermediate level waste taking into account LSA/SCO regulations for transport and waste acceptance criteria for disposal in Germany. Some aspects of different package concepts, like the use of non-reusable or reusable packages, will be considered as well as the application of LSAISCO regulations and further development of LSA/SCO criteria.     

Simulation of particle impact on protective coating of high-level waste storage packages

Integrity and survivability of high-level waste packages are critical for their storage and during their transport. Multi-layer, multi-component coatings composed of TiCN/ZrO₂–TiO₂–Al₂O₃/MoS₂ on the outer shield material can provide engineered barriers resistant to corrosion; radiation, diffusion, and thermal cycling effect that are also wear tolerant and mechanically robust. While waste packages are designed to survive some structural damage, potential coatings applied to future packages may be affected by the development of micro-cracks. In such a case neutrons and gamma rays might interact with the external coatings. In this research, particle impact with multi-layered, multi-component coatings is studied to assess the damage expected in the coatings if micro cracking would happen and heavy particles (neutrons) leak into the coatings. As a first step to investigate this scenario, the open source code SRIM has been used to perform the study using protons as a simulation of the heavy particle interaction. The simulation provides a tool to determine the optimal coating thickness to be manufactured in order to limit the coating surface damage to within minimum values.     

Revised guidelines for the procedure for approval of package designs in Germany

Abstract

The IAEA Regulations for the Safe Transport of Radioactive Material TS-R-1 are applied in Germany through the implementation of the Dangerous Goods Transport Regulations for Class 7 of the International Modal Organisations (ADR, RID, IMDG-Code, ICAO-TI). Based on this the procedures for the approval of package designs used in Germany are in compliance with the provisions of TS-R-1. BfS is the competent authority for the approval of Type B(U), Type B(M) and Type C packages and all packages containing fissile material, and BAM is the competent authority for approval of H(U)/H(M) packages for UF6, special form and low-dispersible radioactive material. The basis for the procedure for approval of package design in Germany are the R 003 guidelines, first issued by the Ministry of Transport, Building and Housing (BMVBW) in 1991. These guidelines have been reviewed and revised to reflect the latest developments in the regulations as well as in regulatory practice. In particular they have been extended to the procedures for approval of Type C packages and packages subject to transitional arrangements, special form and low-dispersible radioactive material, and provide more detailed information to the applicant about the requested documentation. This paper gives an overview of the main parts and provisions of the revised R 003 guidelines issued in December 2004 including scope, responsibilities, application, documentation, evaluation and certification for the various approval procedures.     

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.     

Review of waste glass corrosion and associated radionuclide release as a part of safety assesment of entire disposal system

Current knowledge on high-level nuclear waste glass corrosion is summarized, and remaining problems are discussed for meaningful predictions of the glass corrosion and associated radionuclide release as a part of safety assessment of entire disposal system. In recent years, much progress has been made in understanding the mechanism of waste glass corrosion in aqueous environments. Glass corrosion models based on the mechanism have been developed for predicting the long-term glass performance, and they are incorporated as part of radionuclide source term in safety assessments of the disposal system. However, these results have not yet allowed meaningful predictions for the long-term release of individual radionuclides from the glass in repository environments, because mechanism of the long-term glass corrosion has not been fully understood and solubilities of actinoids and fission products under disposal conditions are rather uncertain. In addition, the most serious problem is that the effects of various reactions and interactions occurring in the engineered barrier system, such as corrosion of overpack, alteration of backfill and chemical interactions of the released glass constituents with them have not been fully coupled with the glass performance. These reactions may be dominant processes controlling the glass corrosion and associated radionuclide release for the long-term. For the meaningful predictions, we must evaluate the waste glass performance in combination with the effects of various reactions and interactions occurring in the engineered barrier system on the basis of fully understanding of the chemical and geochemical mechanisms.     

碳钢容器腐蚀模式

碳钢是用于低中放废液贮存或低中放固体废物处置包装容器的一种常用材料。容器的完整性是阻滞废液向外泄漏或者被处置废物中放射性核素向外释放的一个重要因素。包装容器损坏的主要机制是腐蚀。在废液储存或废物处置条件下，所关心的是容器的点蚀和均匀腐蚀。本文介绍了有关这方面的情况和一些点蚀模式和均匀腐蚀模式。     

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.     

The Swiss Concept—Container Design and the Transport of L/ILW to the Planned Final Repository

Abstract

The design of the Swiss final repository for short lived L/ILW is based on a Nagra container and package concept. The package handling operations have been restricted to a minimum through the design of special handling tools. e.g. a gripper for 9 drums. The routine transport weight by rail is 56 t, and for non-routine transport 80 t (maximum). The transport of drums and reprocessing waste will be in re-usable steel containers and that of decommissioning waste in dual purpose transport and disposal containers. Most of the containers have standardised dimensions and corner fittings which are based on the ISO dimensions. The modes of transport for the containers and packages within the repository include overhead cranes, an air cushion platform for precise manoeuvering in limited spaces and internal rail transport. The handling and transport will mostly be remotely controlled and monitored by video cameras from the control room. Hence, the exposure times of the operating personnel in the radiation environment is minimised.     

Demonstration of structural performance of IP-2 packages by advanced analytical simulation and full-scale drop test

Two new types of IP-2 (Industrial Package Type 2) to transport low and intermediate level radioactive waste (LILW) steel drums from nuclear power plants to a disposal facility have been developed in accordance with the IAEA and Korean regulations for radioactive materials. According to the regulations, both packages must preserve their structural performance after they are subjected to 0.9 m free drop tests, which are prescribed as normal conditions.In this study, an advanced analytical simulation and an evaluation process using the finite element (FE) method have been developed for the design assessment of the newly developed IP-2s. Then, analytical simulations for the various drop orientations were performed to evaluate the structural performance of the packages and demonstrate their compliance with the regulatory requirements. Also, full-scale drop tests were carried out to verify the numerical tools and modeling methodology used in the analyses and to confirm the performance of the IP-2s. In addition, parametric studies are carried out to investigate the sensitivity of the analytical variables, such as the material model and modeling methodology.In addition, this paper intends to provide basic guidance on the analytical simulation and evaluation process specifically for Korean types of transport packages, because numerous transport packages must now be developed for the various kinds of LILW that have accumulated in temporary storage facilities in Korea.     

The Swedish nuclear waste program and the long-term corrosion behaviour of copper

The principal strategy for high-level radioactive waste disposal in Sweden is to enclose the spent fuel in tightly sealed copper canisters that are embedded in bentonite clay about 500 m down in the Swedish bedrock. Besides rock movements, the biggest threat to the canister in the repository is corrosion. ‘Nature’ has proven that copper can last many million of years under proper conditions, bentonite clay has existed for many million years, and the Fennoscandia bedrock shield is stable. The groundwater may not stay the very same over very long periods considering glaciations, but this will not have dramatic consequences for the canister performance. While nature has shown the way, research refines and verifies. The most important task from a corrosion perspective is to ascertain a proper near-field environment. The background and status of the Swedish nuclear waste program are presented together with information about the long-term corrosion behaviour of copper with focus on the oxic period.     

Feasibility Study on Transportation Techniques Employing Air-Bearing System for Pre-Assembled Waste Package at Drift Tunnel of HLW Repository

Full-scale tests were performed to evaluate the technical feasibility of a transport system with air-bearings at the underground HLW disposal tunnel for pre-assembled heavy disposal packages, which consist of a waste package and buffer material. Transport conditions in the disposal tunnel, such as roughness and unevenness of the curved surface, make it difficult to achieve smooth movement using the commercial airbearing transport system. In order to evaluate the applicability of the air-bearing transport system to such conditions, tests using a full-scale test device (modified package) and simulated tunnel surface were conducted. Based on the tests, the applicability of this transport system to a disposal tunnel was confirmed.     

EDF Waste Packages Transport and Compliance with Regulations: Prediction of Retention of Radionuclides by Cementitious Materials

Abstract

Different concrete waste packages have been designed by Electricite de France (EDF) for the long-term storage of radioactive Low Level Waste (LLW). Their main function is to confine radionuclides from the biosphere for three hundred years in a near-surface disposal. According to the transport regulation, a Type B package is needed for some waste like water filters. The water filters from EDF nuclear power plants are encapsulated in mortar and placed in a concrete container. Transport regulations for these containers have required the development of a methodology for safety assessment. The reference scenario of container degradation during transport considers a 9 m drop and a 800°C fire for 30 min. First, the different chemical and physical processes involved in the containment of radionuclides are analysed. In particular, the radionuclide transport mechanisms in cement-based materials have been reviewed. Secondly, the effects of a container drop on the mortar and concrete retention are discussed. Thirdly, in order to prove compliance with the regulations, a simplified model is proposed to predict the radionuclides release with time. It is concluded that cement-based materials offer high performance as a mechanical and chemical barrier to radionuclide releases for Type B packages.