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6 PRE-DISPOSAL MANAGEMENT OF RADIOACTIVE WASTE

This Section focuses on the pre-disposal management of all classes of radioactive waste (see Section 3, “The Classification of Radioactive Waste”). The major topics discussed are waste minimization (subsection 6.1), waste processing (subsection 6.2) and waste storage (subsection 6.3). Partitioning and transmutation are discussed as a topical issue (subsection 6.4).

6.1 Waste Minimization The Agency defines waste minimization as [5.1]:

The process of reducing the amount and activity of radioactive waste to a level as low as reasonably achievable, at all stages from the design of a facility or activity to decommissioning, by reducing waste generation and by means such as recycling and reuse, and treatment, with due consideration for secondary as well as primary waste.

The previous issue of this Status and Trends report provided an overview of major aspects of waste minimization and it highlighted considerations that could be taken into account in decision making for minimization. Of note, the previous issue indicated that waste minimization encompasses organizational, technological, and economical aspects. The following components are usually considered in the planning and implementation of complex, significant waste minimization projects:

• waste minimization strategy, • minimization of radioactive waste arisings (waste avoidance), and • minimization of the volume of waste that has been generated and requires

disposal.

Until recently, waste managers’ efforts were concentrated mostly on the minimization of nuclear facilities’ operational liquid waste generation and the reduction of the volumes of conditioned waste for disposal. Deep evaporation of liquid waste with subsequent calcination of the evaporate and packaging of the calcinate in high integrity containers has been routinely used at some Nuclear Power Plants (NPP) in Europe.

A new approach to liquid waste minimization has brought the wide utilization of various selective sorbents and special filtration procedures for the separation of radioactive components from liquid waste. The methods include re-utilization of the purified water and the advanced processing of sorbents and filtration materials. In some NPP in the USA, the liquid waste to be conditioned has been practically eliminated and the trend to this approach is increasing. Instead of large volumes of conditioned low activity liquid waste, substantially smaller volumes of higher activity conditioned liquid waste, sorbents and filtration materials remain for disposal.

Most of the volume of solid, operational waste from NPP is very low level (VLLW, see subsection 7.4.1) and LILW (see Table 3-1). Typically, most solid waste consists of protective clothing, mops, wipes and other incinerable materials. In spite of its high volume reduction factor, incineration as a treatment process has not become very popular because of problems with ash conditioning and secondary waste generation mainly from complicated

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exhaust air filtration systems. In addition, incinerator ash is often classified as toxic and/or hazardous material [6.1] and, therefore, may be subject to non-nuclear regulation.

An option for minimizing solid waste arisings is through material substitution. One example is the use of dissolvable protective clothing and wipe materials, such as those based on polyvinyl alcohols (PVA). The Electric Power Research Institute (EPRI) launched a special project to evaluate benefits and constraints of replacing classical cotton materials by PVA based materials [6.2]. At a recent EPRI conference (EPRI International Low Level Waste Conference and Exhibit Show, June 19-21 2002, Colorado Springs), one session was dedicated to a presentation of project results as well as presentations of PVA protective clothing manufacturer and potential users. Sufficient contamination protection, high worker comfort and lower heat stress, high performance and simple waste processing are the main features of the new material. In addition, clear economical benefit of cotton replacement by PVA was demonstrated. The PVA material is launderable and can be used several times. However when protective clothing is classified as waste, it is dissolved in simple autoclaves using the “MicrobasixTM” processing system (see Figure 6-1), based on aqueous media. Solid particles are processed with other solid waste, organic components in the liquid phase are destroyed to CO2 and the resulting aqueous solution is processed by standard liquid waste processing technologies. Total volume reduction factors of up to 10 000 can be achieved.

Figure 6-1: Before and After Photos Showing Volume Reduction of PVA Clothing

The EPRI study stated that “If a decision is made to replace existing clothing inventories with PVA alternatives, it is recommended that a phased approach be given primary consideration.”

6.2 Waste Processing Processing involves the treatment and/or conditioning of waste, where treatment and conditioning are defined as follows [5.1].

treatment: operations intended to benefit safety and/or economy by changing the characteristics of the waste. Three basic treatment objectives are (a) volume reduction, (b) removal of radionuclides from the waste, and (c) change of composition of the waste.

conditioning: operations that produce a waste package suitable for handling, transport, storage and/or disposal. Conditioning may include the conversion of the waste to a solid waste form, enclosure of the waste in containers and, if necessary, providing an overpack.

The previous issues of this Status and Trends report gave an overview of radioactive waste processing and highlighted topics such as the minimization of waste arising from nuclear facility decommissioning and innovative approaches to waste processing. In the current issue, an example of a waste processing facility is presented. For upcoming issues, additional

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examples will be solicited for presentation. The following subsection describes the collection and dissemination of waste processing information at the international level.

6.2.1 Collection and Dissemination of Radwaste Processing Information at the International Level

The Agency’s Net Enabled Waste Management Database (NEWMDB, see subsection 11.1) is used to collect and disseminate information on national radioactive waste management programmes, plans and activities, relevant laws and regulations, policies and radioactive waste inventories. Specifically, the NEWMDB is used to collect information about waste processing facilities and information about waste treatment and/or conditioning methods. See Figure 6-2 and Figure 6-3.

Figure 6-2: Example of Processing Facility Information in the NEWMDB

At time of writing, participation by Agency Member States in submitting information to the NEWMDB was low (see subsection 2.3). However, participation is expected to increase during future data collection cycles. The intent is for the NEWMDB to be the most comprehensive source of information about waste processing facilities and methods. For up-to-date information about waste processing in Agency Member States, please refer to NEWMDB reports, which are publicly accessible and cost free (http://www-newmdb.iaea.org/reports.asp).

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Figure 6-3: Example of Processing Methods Information in the NEWMDB

6.2.2 Example(s) of Radioactive Waste Processing

6.2.2.1 Overview of Belgoprocess

based on input provided by B. Thieren, Public Relations Officer, Belgoprocess, Belgium

Belgoprocess is major provider of radioactive waste management and nuclear facility decommissioning services and expertise in Belgium. Belgoprocess was set up in 1984, in Dessel, in the Antwerp region. It is a subsidiary of NIRAS/ONDRAF, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials. NIRAS/ONDRAF is responsible for outlining the Belgian policy on radioactive waste. Belgoprocess is currently processing all categories of radioactive waste resulting from both the nuclear fuel cycle activities and from the production and uses of isotopes in medicine, agriculture and industry. It has a workforce of some 240 people.

Part of the company’s knowledge and experience is drawn from work carried out by the former Eurochemic reprocessing company, the activities and personnel of which were taken over in 1985. Part of it is drawn from work carried out by the former waste treatment department of the Belgian Nuclear Research Centre SCK•CEN, the activities and personnel of which were taken over in 1989.

The industrial experience of Belgoprocess has grown on top of its know-how in the development of waste treatment processes, the operation and maintenance of nuclear fuel cycle facilities and practical experience in the dismantling of alpha contaminated facilities.

Treatment and conditioning of radioactive waste in Belgium

Radioactive waste arises from the nuclear fuel cycle, industrial and medical applications of radioisotopes, and research and development. In order to choose the appropriate treatment, the

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waste is classified into several categories, all related to a specific treatment and conditioning process.

The main criteria are:

• the level of radioactivity (low, medium or high);

• the state of the waste (solid, liquid or gaseous);

• some physical properties (burnable, compressible or otherwise); and

• the type of radionuclides present (α or β-γ nuclides, short or long lived).

Compliance with the specifications

Belgoprocess treats radioactive waste in a way that complies with the licenses for storage and processing and with specifications stipulated by NIRAS/ONDRAF. With its extensive storage and processing infrastructure, Belgoprocess can receive most types of radioactive waste.

Several techniques are used to optimize the overall concept of waste treatment, among them incineration, precompaction, supercompaction (see Figure 6-4), mechanical and thermal size reduction techniques for solid wastes and flocculation and evaporation for liquid wastes.

Figure 6-4: Result of Supercompaction of 220-l drums in the CILVA Installation

Residues from the treatment of the various radioactive waste categories are encapsulated in one of three possible matrices: cement, bitumen or glass. According to the physical state, the chemical composition and the radiological content of the waste residue, the appropriate matrix is chosen and the waste is encapsulated.

Since the late 1960’s, an extensive infrastructure for the processing of low, intermediate and high level (classifications used in Belgium, see subsection 3.2) solid and liquid waste has been built up. By constructing additional plants for special batches of waste, the capabilities of Belgoprocess will be strengthened. In this way, plants are brought into operation for the processing of low, intermediate and high level solid and liquid waste, α-bearing organic waste

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and α-bearing solid waste. This versatility makes Belgoprocess a potential partner for the treatment of foreign waste.

Governments and the nuclear sector increasingly realize the importance of waste characterization in the waste management cycle. Its role is clear when one considers all aspects: legal requirements of the entrusted agency (NIRAS/ONDRAF in Belgium), safety permits of treatment installations and storage/disposal buildings, economical optimization of waste streams, quality control, and information for future generations, especially in regards to long term storage and disposal.

With its NDA-IPAN/GEA system, Belgoprocess has an important tool for the radiological characterization of (possible) alpha contaminated waste. The system was specially designed for Belgoprocess and is continuously upgraded (see Figure 6-5).

Figure 6-5: Non-destructive Analysis of Conditioned Waste

While qualification/validation clearly defines the operational range, the ISO 9001 certificate proves the QA/QC applied in routine operations.

6.3 Waste Storage The Agency defines storage as [5.1]:

the holding of spent fuel or of radioactive waste in a facility that provides for its containment, with the intention of retrieval Storage is by definition an interim measure, and the term interim storage would therefore be appropriate only to refer to temporary, short-term storage when contrasting with the longer-term fate of waste. Storage as defined above should not be described as interim storage”.

The purpose of this subsection is to discuss issues related to radioactive waste storage, not interim storage. It is realized, however, that not all countries follow the Agency’s recently

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stated definition in a consistent manner. This subsection contains information about radioactive waste storage in some Member States that use the term interim storage. Within this subsection, any and all references to interim storage are to be taken to mean storage, per the Agency definition.

6.3.1 Examples of Radioactive Waste Storage

6.3.1.1 Radioactive Waste Storage In The Netherlands

based on input provided by H. Codée, COVRA, The Netherlands

All radioactive waste produced in The Netherlands is managed by COVRA, the Central Organization for Radioactive Waste. COVRA operates a facility at the industrial area Vlissingen-Oost in the southwest of the country.

Waste conditioning and long term storage (at least 100 years) takes place on a 25 hectare site. Facilities for low and medium level waste (classifications used in the Netherlands, see subsection 3.2) were erected between 1990 and 1992. In 2000, a building for storing VLLW (see subsection 7.4.1) from ore processing industries was commissioned for technologically enhanced, naturally occurring radioactive material (TE-NORM) waste. The construction of a naturally cooled storage facility for HLW started in 1999 and will be commissioned in 2003. Construction of a storage facility for depleted uranium will start in 2003 and will be commissioned in 2004.

A layout of the COVRA facilities is given in Figure 6-6. All storage facilities are modular buildings. The storage building for low and medium level waste is H-shaped (number 3 in Figure 6-6) and it consists of a central reception bay surrounded by four storage modules. Each storage module has a storage capacity for ten years of waste production at the present rate. In Figure 6-6, the situation is shown for a fully developed site, which is expected after 2100. As can be seen from this figure, in total 16 storage modules for low and medium level waste can be constructed, which represents some 160 years of waste production.

Only one third of the storage capacity for TE-NORM waste (number 5 in Figure 6-6) is currently in operation. Three more buildings of approximately the same size can be constructed in the future as can be seen from the situation after 2100 in Figure 6-6. One or possibly two of these buildings will be used for the storage of depleted uranium. It is expected that the potential storage capacity here will be sufficient for fifty to one hundred years.

The storage building for high level waste (number 4 in Figure 6-6) can be doubled in capacity. The present capacity is sufficient for the existing nuclear programme.

Since all wastes will be stored for a period of at least 100 years, this has to be taken into account in the design of the storage.

Low and medium level waste

All waste is conditioned in cement in relatively small units. Cement is a very stable product and creates an alkaline environment for the waste materials. This will prevent or slow down the degradation of the waste materials. Producing relatively small units of 200 or 1 000 litres facilitiates repairs if needed.

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In the storage building, blocks of waste packages are placed in rows, which leaves open corridors for inspection. Lower dose rate packages are stored along the outer walls of the modules, and on the top layers in order to provide additional shielding for higher dose rate packages at the interior. See Figure 6-7.

The storage building is a simple concrete building; there is no mechanical ventilation. Using mobile equipment, humidity in the building is kept at a low level to prevent condensation of air moisture on packages. The storage area is classified as a contamination free area.

Figure 6-6: Layout of the COVRA Facilities in 2003 and the Situation after 2100

1 – office building and exhibition centre; 2 – building for the treatment of low and medium level waste; 3 – storage building for conditioned low and medium level waste; 4 – storage building for high level waste; 5 - storage building for low level waste from the ore processing industry; 6 - storage building for depleted uranium.

Figure 6-7: Storage of Low and Medium Level Waste at COVRA

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TE-NORM and depleted Uranium

The TE-NORM is a calcinate with only Po-, Bi- and Pb-210. It is a stable product that does not need to be conditioned to assure safe storage. Any additional conditioning would enlarge the volume and would add to the costs. The calcinate is collected in a specially designed ~6 m (20 foot) container. The container can be filled with up to 30 tonnes of material. The containers are stacked four high in the container storage building (Figure 6-8).

The container storage building is a steel construction frame with steel insulation panels. High quality criteria were set for the construction and for the type of materials used in order to meet a 150 year lifetime with practical maintenance. The building is modular to allow expansion. Each module has an overhead crane. Technical provisions inside the building are minimal. With mobile equipment, the air humidity in the storage building is kept below 60%. All containers must be free of outside contamination.

In 2003, the construction of a storage building for depleted U3O8 will start. This will be a concrete building again with minimal fixed installations or equipment, comparable to the store for low and medium level waste.

The depleted U3O8 is a stable product to store. Because of its potential future use, the material will not be conditioned in a fixed matrix. When judged necessary in the future, for instance when the material will be brought into a geologic disposal facility, then this can be done according to applicable standards at that time. Money for this treatment and for the final disposal will be set aside in a capital growth fund in the same way as is done for all other waste stored at COVRA.

Figure 6-8: The Storage of Radioactive Calcinate from Phosphor Production

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High level waste

Because of the long term storage requirement, a system was chosen that is as passive as possible and where precautions are taken to prevent degradation of the waste packages. The heat generating waste is stored in an inert noble gas atmosphere and cooled by natural convection. In the design of the storage vault, all accidents with a frequency of occurrence larger than once per million years were taken into account. The design requirement was that these accidents must not cause radiological damage to the environment.

The non-heat generating waste is remotely controlled and stacked in well-shielded storage areas. The heat generating waste, such as the vitrified residues, will be put into vertical storage wells cooled by natural ventilation. This method is a proven technology in the storage facilities of BNFL at Sellafield and Cogéma at La Hague.

The spent fuel elements of research reactors are delivered to COVRA in a cask containing a basket with about 30 elements. The basket, with elements, is removed from the cask and placed in a steel canister, which is filled with an inert gas and welded tight. These sealed canisters are placed in wells, in the same way as the vitrified residues. The wells will be filled with an inert gas to prevent corrosion of canisters with spent fuel elements or vitrified waste (Figure 6-9).

Construction of the storage vault started in 1999 and it will be commissioned in 2003.

Figure 6-9: Emplacement of the Wells during Construction

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Policy for long term storage followed by disposal (also see subsection 6.3.3, “Topical Issue: Long Term Storage”)

The policy in the Netherlands is that all hazardous waste must be isolated, controlled and monitored. In principle, this can be done by storage in buildings and control by society. This can also be done by shallow land burial and control of society, or by deep geologic disposal and control of society. For the options mentioned, the degree of societal control is the highest for storage in buildings and the lowest for deep disposal. When containment is required over periods of time longer than the existence of the current society, doubt may arise on the capability of future societies to fulfil the control requirement.

The Netherlands has a very high ground water table and under those circumstances shallow land burial is not acceptable for low and medium level waste. As a consequence, for all waste categories deep geologic disposal will be required as the ultimate solution Fortunately, deep lying, large salt formations with good potential as disposal sites, as well as deep clay formations, are available. Unfortunately, however, public acceptance for deep disposal is lacking. Some comfort can be given when disposal is retrievable and this is a requirement for The Netherlands.

The volume of waste accumulated to date is only a few thousand m3 and it is not economically feasible to construct a deep geologic disposal facility for such a small volume. The waste volume expected over 100 years should be large enough to make a disposal facility viable. Therefore, a period of at least 100 years of storage in buildings will be required. This creates at least six positive effects:

• Public acceptance is currently quite high for long term storage. The general public has more confidence in physical control by today’s society than in long term risk calculations for repositories even when the outcome of the latter is a negligible risk,

• There is a period of 100 years available to allow the money in the capital growth fund to grow to the desired level. This brings the financial burden for today’s waste to an acceptable level,

• During the next 100 years, an international or regional solution may become available. For many countries the total volume of radioactive waste is small. Co-operation creates financial benefits that can result in a higher safety standard and more reliable control,

• Over 100 years, the heat generating waste will cool down to a condition where cooling is no longer required. In addition, a substantial volume of the waste will decay to a non-radioactive level over 100 years, and

• A little more than 100 years ago, societies were not even aware of the existence of radioactivity. In 100 years from now, new techniques or management options can become available.

The solution for The Netherlands is, therefore, to store the waste in buildings for a period of at least 100 years and to prepare financially, technically and socially for deep disposal during this period in such a way that it can be implemented after the storage period. Of course, at that time society has the freedom of choice between a continuation of the storage for another 100 years or implementation of disposal.

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6.3.1.2 Interim Safe Storage (ISS) in the UK

based on a presentation by D. Bonser (BNFL, UK) at the International Conference on Issues and Trends in Radioactive Waste Management December 2002 (see subsection 12.1)

Historic Intermediate Level Waste (ILW, classification used in the UK, see subsection 3.2) resulted from reprocessing and other treatment processes that were carried out prior to the introduction of modern waste encapsulation facilities. Much of this waste is poorly documented, difficult to characterize and in facilities that were designed with little attention paid to emptying and decommissioning. These wastes are a result of strategic decisions taken over the past fifty years. Against this background, the concept of Interim Safe Storage (ISS) was introduced as an approach to the safe, timely, and effective retrieval and treatment of historic ILW. See Page 54 regarding “interim storage”.

The ILW disposal criteria established by Nirex [6.3] require that waste to be encapsulated is well characterized and, in many cases, that mixed wastes are segregated. However, given the current status of historic ILW, the ISS concept sets retrieval and improved storage as the near term priority while continuing the search for new techniques that would allow the safe retrieval, characterization, segregation and packaging that would fully meet the Nirex disposal criteria, or the disposal criteria that might be ultimately established.

The ISS concept does not preclude or trivialize the Nirex disposal criteria as a desirable objective. However, what it leads to is a set of principles that should be observed by those responsible for managing the historic ILW liability. These principles are aimed at ensuring the overall risk and hazard is reduced in a timely manner, see Figure 6-10, and that appropriate regulatory requirements are in place.

ISS Principles

In order to manage and, over the earliest possible timescale, reduce the risk from historic ILW, the following principles should be considered:

• Deploy simple, proven technologies for the retrieval and treatment of historic wastes,

• Recognize that some of the historic wastes as stored at Sellafield cannot, in general, be representatively sampled or comprehensively characterized,

• Seek waste treatment processes that can accept the inevitable uncertainties in waste characteristics or mix and that require, at worst, only minimal waste segregation,

• Maintain that it is inappropriate to re-store historic wastes, untreated, in new stores,

• Maintain that it is impractical to develop new or untested technologies for waste characterization, segregation or treatment in order to satisfy the requirements of Nirex disposal criteria,

• Establish a radioactive inventory of waste packages before storage but not to segregate wastes before treatment and packaging.

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• Develop waste packages that are demonstrably safe for extended storage (of the order of 100 years) but that can be inspected and re-packaged should their long term integrity prove poorer than predicted,

• Limit the fissile content of waste packages to meet the requirements of safe storage and not the post closure repository model,

• Ensure that, if necessary, design for safety in storage takes primacy over requirements to assure repository post-closure performance (as currently modelled),

• Treat the Nirex disposal criteria as a sound assessment process to be used to provide guidance on the quality of waste package design,

• Minimize the integrated risk during ISS by seeking to enhance the passivity of waste packages, and

• Develop an overall model of the magnitude and timing of the secondary waste streams (principally air borne and liquid discharges) that will result from retrieval and treatment operations.

These principles allow the work to be discharged in a variety of ways. British Nuclear Fuels is developing a variety of approaches to address historic ILW in line with these principles.

Figure 6-10: Overall Risk and Hazard Reduction with Time

One approach under consideration is that historic ILW is retrieved and processed as necessary in order to package in approved Nirex containers. Every effort will be made to conform to Nirex ILW disposal criteria and produce drummed, cemented waste packages that are sanctioned by Letters of Comfort (LoC). However, for waste streams where it becomes evident that Nirex ILW disposal criteria cannot be fully satisfied, work would be carried out to understand the potential consequences of the non-conformance.

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Those non-conformances that could potentially breach safety criteria will be of most concern, and strategies would be required to mitigate such safety concerns. Where non-conformances raised repository operability concerns then the priority would be to retrieve and store the waste. Non-conformances of any of the Nirex criteria would have to be assessed and the likely impact understood. Nothing will be done that is not acceptably safe and that does not meet regulatory requirements.

A second approach considers the risk of rework to drummed, cemented waste. The approach, known as design for storage, centres on the safe, timely, retrieval of ILW wastes. Processing of these wastes focuses on stabilizing and rendering the wastes passive. Once processed, wastes would be transferred to suitable containers to allow safe mechanical handling and storage for a period of the order of 100 years. Design for storage would require specific purpose built storage facilities. These would include modern security, environmental monitoring, and mitigation systems (including nuclear ventilation and fire systems).

Design for storage will almost inevitably require waste containers to be reworked in order to meet disposal requirements. Storage should be designed to allow relatively straightforward re-work. This is in contrast to cemented, drummed waste that may prove extremely difficult, costly, and slow to re-work.

Timescales of the order of a hundred years are needed for the ISS concept. These timescales may be reduced in the event of significant progress being made on the provision of a final disposal route. Similarly, the timescales may be extended. However, ISS is not envisaged as an indefinite solution. The protracted storage above ground of ILW would demand continuous attention by subsequent generations.

By following these principles, BNFL seeks to manage the retrieval, treatment and storage of historic wastes in ways that will produce the greatest time integrated reduction in risk while avoiding grossly disproportionate expenditure. In doing so, it recognizes that in deciding to adopt these principles, there will be difficulties in ensuring that the waste packages produced fully satisfy the Nirex assessment criteria.

This will create regulatory difficulties, as there is a requirement for all waste packages to have an LoC before approval is given for construction, commissioning or operation.

6.3.1.3 Storage of Spent Nuclear Fuel in Germany

based on a paper by B.R. Thomauske (BfS, Germany) [6.4]

The German Government has decided to phase out nuclear power by limiting the lifetime of nuclear power plants (NPP) to 32 years from their commissioning. In accordance with this policy, a new waste management concept was developed. The goal is to establish a geological repository for radioactive waste starting operation by 2030. Until then, spent fuel (SF) has to be stored. The Federal Government and the utilities have agreed that the utilities will set up storage facilities on their sites or near their NPP.

Spent fuel elements are initially stored under water in cooling basins inside NPP as well as in external central or decentralized storage facilities. There are currently two central and two decentralized storage facilities in operation. The central storage facilities Gorleben and Ahaus are able to take about 8 000 Mg heavy metal (Gorleben: 3 800 Mg heavy metal, Ahaus: 3 960 Mg heavy metal). The storage facility at Gorleben is the only one in the Federal Republic of

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Germany that has a license for the storage of vitrified high level radioactive waste from reprocessing of German fuel rods abroad.

At two sites, decentralized storage facilities for spent fuel elements from NPP have been in operation since 1999; one wet storage facility near Obrigheim as well as the “Interim Storage Facility North” (ZLN, 585 Mg heavy metal) at Rubenow/Greifswald for spent fuel originating from the decommissioned nuclear power plants at Greifswald and Rheinsberg. Another decentralized storage facility for spent fuel elements from the decommissioned AVR research reactor is located at Jülich. See Page 54 regarding “interim storage”.

From the 19 German NPP, approximately 500 Mg of spent fuel elements are discharged annually. In addition to spent fuel sent for reprocessing, approximately 12 000 Mg of heavy metal will be produced during the phase of nuclear energy use, which is clearly limited now. A capacity of 7 760 Mg of heavy metal currently exists at the Ahaus and Gorleben central storage facilities. Therefore, additional storage capacity for fuel elements is needed. The question is, therefore, not if but only how would this be accomplished, what technical design(s) are needed and where further storage capacity should be created.

To minimize the future transport of fuel, it would make sense to establish new storage capacity at the site of a future repository. However, currently no site has been identified.

The Gorleben salt dome has been explored since 1979. The Federal Government has doubts about Gorleben’s suitability as a fuel repository. Until planning and safety-related issues have been clarified, its exploration was suspended in October 2000 for at least 3 years to at most 10 years. This time will also be used to determine the criteria and a selection procedure for a repository site. The Federal Ministry for the Environment, Nature Conservation and Reactor Safety has appointed a group of scientists called Arbeitskreis Auswahlverfahren Endlagerstandorte (AkEnd) to develop scientifically founded criteria as well as a selection procedure for appropriate siting of a repository for radioactive waste. AkEnd reports are available on the Internet [6.5].

Concept of Decentralized Storage

Fuel storage at or near its point of generation minimizes transportation. When a repository is available, the fuel can be transported directly to the repository, thereby avoiding the intermediate step of shipping the fuel to a centralized storage facility.

From the end of 1998 to the end of 2000, new applications were filed for on-site storage facilities (12 storage buildings, one storage tunnel and 5 storage areas) at 13 NPP sites in five Federal States (see Figure 6-11). The operational life of the SF storage facilities is limited to 40 years beginning with the emplacement of the first spent fuel elements. On-site storage facilities are limited to the storage of spent fuel elements from the neighbouring NPP. It is not planned to store spent fuel elements from other plants or high level radioactive waste from reprocessing. The application for the Biblis fuel storage facility, where fuel elements from Biblis as well as fuel elements from the decommissioned Mülheim-Kärlich power plant are to be stored, is an exception.

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storage area applied NPP storage area approved

storage facility applied application withdrawn

Figure 6-11: On-Site Dry Storage in Germany

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Technical Concepts For On-Site Storage Facilities

The applications for licenses represent three basic alternatives – buildings, tunnels and areas:

Storage building

The storage building has two concepts: the STEAG concept and the WTI concept:

• STEAG concept (a storage hall designed by the company STEAG encotec GmbH) - characteristics: thick concrete structures, wall thickness approximately 1.2 m, roof thickness approximately 1.3 m, one-nave building, and

• WTI concept (a storage hall designed by the company of consulting engineers Wissenschaftlich-Technische Ingenieurberatung GmbH), hall similar to the storage facilities at Gorleben, Ahaus and Lubmin/Greifswald - characteristics: wall thickness approximately 0.7 m or approximately 0.85 m, respectively, roof thickness approximately 55 cm, two-nave building, consisting of two halls separated by a wall.

The STEAG concept has been applied for the 6 North German sites of Brokdorf, Krümmel, Brunsbüttel, Grohnde, Lingen and Unterweser. The WTI concept has been applied for the 5 sites of Biblis, Philippsburg, Grafenrheinfeld, Isar and Gundremmingen located in the southern part of Germany.

Supporters of the WTI concept argue that the concept has proven to be worthwhile for approximately 15 years. Operational experience has been gained for many years on the basis of which technical and economic optimizations could be made. The STEAG concept was developed with regard to the use of future and maybe cheaper generations of casks. While current transport and storage casks already guarantee safety in the case of an airplane crash, the utilities plan to take credit of the thick concrete structure of the STEAG concept when using future cask designs. The strengthened building structure should serve as an additional barrier corresponding to the multiple barrier principle in nuclear engineering.

As in the existing central storage facilities, the control of pressure and, thus the tightness of the storage casks, is guaranteed in on-site storage facilities and storage areas as well.

For licensing and safety, there is no difference in the requirements for decentralized and centralized storage facilities. In the former case, possible interactions with the neighbouring NPP must be examined and verified.

Storage Tunnel

In Neckarwestheim, on-site storage in two tunnel tubes lined with concrete is planned. This special solution below ground was developed due to site-specific conditions. It has the advantage that no increase of radiation exposure will occur in the environment, in spite of the increase of activity at the site. This solution, nevertheless, does not, at all, mean final disposal – as is feared in the region. At this site neither the rock formation is suitable nor is there a depth of several hundred meters in the underground that would be necessary for a repository.

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“Interim Storage Area” (See Page 54 regarding “interim storage”)

In “interim storage areas”, storage casks (type CASTOR or type TN) are stored on a defined area on the NPP site. In contrast to storage in upright positions in storage buildings, horizontal storage of the casks on concrete slabs has been applied for “interim storage areas”. To shield gamma and neutron radiation and as protection against the weather, each cask is covered by prefabricated concrete elements (Figure 6-12).

Since the only purpose of an “interim storage area” is to bridge the period until a 40-year, on-site storage facility is ready for operation, a low number of casks, 12 to 28, is typical. Due to the flat storage, this concept requires considerable space and is, therefore, only suitable for a low number of casks because of the limited space available at NPP sites. It has the advantage of a short construction period of about 1 to 2 months.

Figure 6-12: “Interim Storage Area” Neckarwestheim

6.3.2 Collection and Dissemination of Radwaste Storage Information at the International Level

The Agency’s Net Enabled Waste Management Database (NEWMDB, see subsection 11.1) is used collect and disseminate information on national radioactive waste management programmes, plans and activities, relevant laws and regulations, policies and radioactive waste inventories. Specifically, the NEWMDB is used to collect information about waste storage facilities and the inventories of wastes in these facilities. See Figure 6-13 and Figure 6-14.

At time of writing, the participation by Agency Member States in submitting information to the NEWMDB was low (see subsection 2.3). However, participation is expected to increase during future data collection cycles. The intent is for the NEWMDB to be the most comprehensive source of information about waste storage facilities and stored waste inventories. For up-to-date information about waste storage in Agency Member States, please refer to NEWMDB reports, which are publicly accessible and cost free (http://www-newmdb.iaea.org/reports.asp)

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Figure 6-13: Example of Storage Facility Information in the NEWMDB

Figure 6-14: Example of Waste Storage Inventory Data in the NEWMDB

6.3.3 Topical Issue: Long Term Storage

Long term surface storage in advance of the disposal of radioactive waste and/or spent nuclear fuel is formally becoming a reality in some countries (see subsections 6.3.1.1 and 6.3.1.2 for examples). In some countries, there is an unstated policy of “wait and see”. This is occurring for various reasons, including delays in repository programmes, lack of resources, uncertainties about whether spent fuel should be considered a waste or a resource, lack of public acceptance of disposal and lack of political will to proceed with disposal. At the December 2002 international conference on Issues and Trends in Radioactive Waste Management (see subsection 12.1), it was concluded that long term surface storage is unsustainable because of the need to maintain institutional control to guarantee the safety of the storage facility. This is an important issue on which a clear, internationally agreed policy would be valuable.

An Agency Technical Meeting was held 20 – 22 November 2002 to “Assess the Safety Implications of Extended Storage of Radioactive Waste”. Agreement was reached on the following conclusions:

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• Storage and disposal are complementary rather than competing activities, and both are needed. The timing and duration of the process of moving from storage to disposal is influenced by various factors including the types of wastes being considered and national circumstances,

• Storage is a necessary phase in safely managing most types of radioactive waste. During the storage phase, for example, the radiation levels and heat generation intensities may decrease to levels that are more manageable. Storage is also a necessary part of waste treatment and conditioning programmes. Storage has been carried out safely within the past few decades and there is a high confidence that it can be continued safely for limited periods of time,

• For the types of radioactive wastes considered, perpetual surface or near surface storage is not considered to be either feasible or acceptable because of the difficulties in assuring active control over the time periods for which these wastes remain potentially hazardous,

• Strategies for storage and disposal need careful consideration in respect of many issues. These include, transport of radioactive wastes from storage to a disposal site, security of the waste, the retrievability of the waste, availability of suitable disposal sites, confidence that adequate levels of safety can be achieved, and the availability of finance, and

• Waste acceptance criteria are needed for storage facilities. These criteria may be different from acceptance criteria for disposal facilities. For example, there may be differences in requirements for package stackability and durability, which would be related to requirements to inspect and retrieve waste from storage as opposed to enabling the possible retrieval of waste from disposal.

6.4 Topical Issue: Partitioning and Transmutation Two Agency projects are approaching partitioning and transmutation (P&T) of spent nuclear fuel from the strategic standpoints of waste management and fuel utilization, respectively. The Agency's Nuclear Fuel Cycle and Waste Technology Division is finalizing a draft technical document describing the benefits and difficulties of P&T technologies as a part of waste management strategies, while the Nuclear Power Division has a coordinated research project (CRP) focused on advanced technologies for transmutation (additional information about the CRP is given later in this subsection).

The Nuclear Energy Agency of the Organization of Economic Co-operation and Development (OECD/NEA) has recently initiated a new task on the assessment of the impact of fuel cycle options on waste management facilities. The NEA task is complementary to the Agency’s tasks. The most important part of the NEA task is to assess the potential role of P&T in selected fuel cycles. The Agency is a participating organization.

The Agency’s projects respond to concerns among countries with nuclear power programs, not only about increasing spent fuel inventories but also about the prospect that nuclear energy growth would be curbed by lack of public acceptance of current solutions to the back end of the fuel cycle. P&T is one promising answer.

The once-through fuel cycle requires disposal of spent fuel as waste in a repository. The classic closed fuel cycle, which relies on reprocessing spent fuel, requires final disposal of HLW and long lived waste. In either case, the material disposed of would remain radioactive

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for very long times. Implementation of P&T would significantly decrease the radiotoxicity of waste and it would decrease the time needed to reduce the radioactivity to harmless levels (from hundreds of thousands of years down to a few thousand).

The final draft of a technical document entitled “Implications of Partitioning and Transmutation in Radioactive Waste Management”, has been completed by an international specialist group. The document is scheduled to be published in 2003. The main aim of the document is to assess recent and future availability of technologies and to estimate the impact P&T could make on strategies, planning and the management of radioactive waste, rather than to contribute to development of systems or procedures.

According to the draft, world nuclear generating capacity of about 350 gigawatts generates some 10 500 metric tons heavy metal (MTHM) of spent fuel (SF) a year. Available reprocessing capacity covers barely a third of the SF generated. The total inventory of stored SF has increased to 130 000 MTHM, while only 70 000 MTHM has been reprocessed.

Spent fuel accumulation is expected to continue for decades, given low uranium prices, and thus increasing the amounts of storage and repository space that will be required. P&T assumes reprocessing. Partitioning is just the additional separation of specific long lived radionuclides from the mixture of fission products in high level waste (HLW), and managing them separately. The rest would decay to acceptable radiation levels in about 500 years. Transmutation involves transforming long lived isotopes into shorter lived or inactive ones.

Partitioning and transmutation are "fundamentally different" concepts, both technologically and in their roles in waste management, the draft report says. Partitioning is "a form of ‘super-reprocessing’ by which, in addition to uranium, plutonium and iodine (I-129), the minor actinides and the long lived fission products (technetium-99, zirconium-93, cesium-135, palladium-107 and selenium-79) would be extracted from the liquid high level waste" from reprocessing.

Transmutation, on the other hand, requires completely new fuel fabrication plants and irradiation technologies that are yet to be developed and implemented on an industrial scale. Existing nuclear power plants could, in principle, be used for transmutation, but many practical obstacles may arise, such as interference with daily operation of the plants, core safety considerations, and low transmutation or incineration yields.

The objective of P&T is to reduce the long term hazard of SF and HLW associated with the radioactive source term itself. For waste management, on the other hand, the stress is on long term radiological risk, which combines the potential hazard with the confining properties of the geologic repository strata. The measures to reduce hazard and risk are very different and fundamental: hazard has to do with actinides, particularly transuranic elements, risk with the mobility of fission products in the geosphere and possible migration to the biosphere.

The document would propose where and to what effect P&T could be applied in the "reasonable" future. While transmutation facilities do not yet exist, it is reasonable to pursue partitioning as a waste management tool. It could be industrially applicable within five to ten years. Then minor actinides and even long lived fission products can be stored in isolation or transmuted when transmutation devices become available. Repositories for remaining shorter lived radioactive waste could be then constructed in many more geological formations than are acceptable today.

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In spite of optimistic expectations related to potential applications of P&T, it should be noted that the real benefit of both processes can be achieved only in connection with extensive and long term utilization of nuclear energy. Several transmutation cycles are needed to achieve the desired results (reduction of the quantities of long lived radionuclides). Taking this into account and considering the extremely high investment costs, P&T can only be considered to be a realistic option if applied over a long time (about 100 years). Without such a long term application, P&T may only remain a scientific dream.

Meanwhile, the CRP on “Studies of Advanced Power Reactor Technology Options for Effective Incineration of Radioactive Waste” was kicked off in November 2002 with the first Research Co-ordination Meeting hosted by Forschungszentrum Karlsruhe, Germany. Up to date information on the CRP, including a list of participating organizations, is available on line [6.6].

The teams of the Commissariat à l’Énergie Atomique (CEA), France, reported significant progress made in their laboratories to develop chemical methods to separate minor actinides from lanthanides and other fission products in the waste stream of PUREX reprocessing. On the one hand, using very selective extractant molecules, one can separate americium, curium and cesium. On the other hand, with some modifications to the industrial process implemented in La Hague, one can separate neptunium, iodine and technetium. The scientific feasibility of these methods is now established. Very good partitioning factors have been obtained. Figure 6-15 describes the CEA "reference scheme" for advanced partitioning. There is still a long way to go to turn such a scheme into viable industrial processes.

Some advanced research is also carried out internationally on non-aqueous separation processes based on high temperature pyroprocessing technologies [6.7]. Further development and implementation of pyroprocessing is an unavoidable condition for really efficient, safe and proliferation resistant fuel cycles. For overviews of P&T, see references [6.8] and [6.9].

(http://www.cea.fr/gb/publications/Clefs46/pdfg/05-radionuclide.pdf)

Figure 6-15: CEA Reference Scheme for Complete Partitioning

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References for Section 6

6.1 K.L. Gering, “Problematic Incinerator Ash: A Case Study in Finding a Successful Treatment Approach”, Waste Management 1999 Symposium, Tucson, Arizona, USA, 28 February - 4 March 1999.

6.2 Electric Power Research Institute, “Dissolvable Clothing: Cost-Effective, Emerging Technology for Low Level Waste Management, EPRI Online Journal article. http://www.epri.com/journal/details.asp?id=471&doctype=field

6.3 United Kingdom Nirex Limited, “Generic Phased Disposal System Documentation”, Harwell, United Kingdom, March 2001.

6.4 Thomauske, B.R., “Interim Storage of Spent Fuel in Germany – Situation, State of Licensing Procedures, Prospects -”, Waste Management 2002 Symposium, Tucson, Arizona, USA, 24 - 28 February 2002.

6.5 Home page (English) for AkEnd (Arbeitskreis Auswahlverfahren Endlagerstandorte) http://www.akend.de/englisch/home/index_1024.htm

6.6 Web site “Fast Reactor & Accelerator Driven Systems Technology Development” http://www.iaea.org/inis/aws/fnss/

6.7 World Nuclear Association, “Processing of Nuclear Wastes”, December 2001. http://www.world-nuclear.org/info/inf69.htm

6.8 Commissariat à l’Énergie Atomique, “Long-lived Radionuclides Partitioned at Will”, Clefs-CEA, No. 46, Spring 2002. http://www.cea.fr/gb/publications/Clefs46/pdfg/05-radionuclide.pdf

6.9 Commission pour l'Analyse des Modes de Production de l'Electricité et le Redéploiement des Energies (AMPERE Commission), Electricité Nucléaire Rapport Principal, Appendix A, “Partitioning and Transmutation as a Waste Management Option in a Future Nuclear Era”. http://mineco.fgov.be/energy/ampere_commission/Append.pdf

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