The 1979-1983 European Communities R&D Programme on Decommissioning of Nuclear Installations was adopted by the EC Council on the 27 March 1979 [1].

The objectives, scope and implementation of this first five-year programme are given hereafter.

Objectives of the programme

The general objective of the programme has been set out by the Council of the European Communities, when it adopted the programme in 1979, considering:

"Certain parts of nuclear power plants inevitably become radioactive during operation; it is therefore essential to find effective solutions which are capable of ensuring the safety and protection of both mankind and the environment against the potential hazards involved in the decommissioning of these plants".

The radioactivity associated with a nuclear power plant, once the plant has been finally shut down, decays according to strict laws and there is no practical way to influence the amount of radioactivity subsisting at a given time.

Consequently, protection against the potential hazards due to this radioactivity means to isolate the latter from man and his environment by appropriate barriers. Since strong radioactivity barriers are already existing in every nuclear power plant, in-situ storage of the radioactive components is a possible means to isolate the radioactivity till the initially dominating shorter-lived radio-nuclides have died away.

Eventually, however, the remaining radioactive material is to be transferred to a final radioactive waste repository, because its isolation in the very long term is best achieved there, and in order to make it possible to release the plant site for other uses. This transfer, which involves the breaking-up of the existing radioactivity barriers and the creation of new ones, is a critical phase in decommissioning.

Accordingly, the major part of the Community programme has been devoted to the operations required in this phase, i.e. dismantling and decontamination of plant components, and treatment and packaging of the arising waste materials.

The main effort in the programme has been aimed at providing efficient techniques and procedures for decommissioning, considering radiation protection, restriction of radioactive waste arisings, and cost effectiveness.

Though a number of conventional techniques appear to be applicable in decommissioning, there is a need to improve their performances and to optimize and characterize them with a view to the special requirements due to radioactivity, e.g., control of airborne contamination and remote operation.

Another aim of the programme has been to provide basic information needed in decommissioning, such as data on the distribution, the level and the radionuclide composition of the radioactivity associated with nuclear power plants finally shut down.

A third aim has been to identify and develop plant design features facilitating decommissioning.

Scope of the programme

The subjects of the programme are:

• A. Research and development projects
• Project n° 1: Long-term integrity of building and systems
• Project n° 2: Decontamination for decommissioning purposes
• Project n° 3: Dismantling techniques
• Project n° 4: Treatment of specific waste materials: steel, concrete and graphite
• Project n° 5: Large containers for radioactive waste produced in the dismantling of nuclear power-plants
• Project n° 6: Estimation of the quantities of radioactive wastes arising from the decommissioning of nuclear power-plants in the Community
• Project n° 7: Influence of nuclear power-plant design features on decommissioning
• B. Identification of guiding principles

The various R&D Projects relate to successive steps of decommissioning were defined. Concerning radioactive waste management, only aspects and procedures that are specific to the waste arising from the dismantling of nuclear power plants have been investigated under the programme on decommissioning; other procedures, in particular waste disposal methods, are being developed in the framework of the Community Programme on Radioactive Waste Management.

The scope and the particular objectives of the various projects are outlined in the following.

Implementation of the programme

The research work has been carried out by organizations and companies in the Member States under 51 research contracts, most of them cost sharing.

The Commission had a budget of 4.7 million ECU for the five-year programme. In the implementation of the programme, the Commission has been assisted by an Advisory Committee on Programme Management, which is composed of experts appointed by the Member States governments and of Commission officials.

Due to various reasons, the contractual research activities could not start before 1980. On the other side, a number of contracts have been prolonged on request of the contractors and are now running till mid 1984.

Emphasis has been laid on the exchange and dissemination of the results obtained.

The common action to collect data relevant to cost, occupational doses, working time and waste arisings is now a fully operational part of the programme.

Subjects of the programme

Section A: Research and development projects

Project/Area Nr. 1: Long-term integrity of buildings and systems

In-situ storage appears to be the preferred procedure for the commercial nuclear power plants that have already been finally shut down.

The implications of storage over extended time periods, ranging from several decades to about a hundred years, have been investigated under this programme.

Measures to put plants into conditions assuring safety essentially through passive means and with a minimum of continued surveyance and maintenance have been studied.

The operations required are not technically demanding, but they need to be carefully designed and planned in order to minimize radiation exposures and later interventions. Plant systems which are to be maintained in operation or in operable conditions (to serve in the dismantling of the plants) were identified.

The mode and pace of degradation of various materials as they exist in nuclear power plants have been studied. This involved inspections of plants, to identify critical areas, and examinations of samples. Measures to prevent or control degradation have been proposed.

These studies showed no problem which would make the feasibility of in-situ storage of finally shutdown nuclear power plants doubtful. The radiological consequences and costs of in-situ storage have been estimated. It certainly allows to reduce the occupational radiation exposure during dismantling of the plants.

To back up these studies, the experience being gained with the plants that have already been finally shut down should be continuously analysed.

It has been proposed that the dismantling of nuclear installations be delayed for periods ranging from several decades to about a hundred years.

Thereupon, the radioactivity having largely died away, dismantling would be easier and the radiation exposure of the dismantling personnel would be less.

The objective of this area is to determine the measures required for maintaining shutdown plants in a safe condition and to assess the radiological consequences of costs.

Project/Area Nr. 2: Decontamination for decommissioning purposes

Surface contamination represents only a minor amount of radioactivity compared with the internal activation by neutron exposure of the components near the reactor core, but it is spread over a wider range of components and more readily releasable.

Therefore, surface contamination is an important source of low-level radioactive waste and a concern for radiation protection in decommissioning operations.

Surface decontamination, i.e. removal of the radioactive surface layer from a plant component, facilitates the further handling and processing of the component. The removed radioactive substance should be concentrated into a small volume of secondary waste that is easy to condition for disposal.

If the decontamination is efficient enough, it makes it possible to consider the treated component as non-radioactive material and is then a way to reduce the overall radioactive waste volume and to recover valuable raw materials.

Decontamination for the purpose of unrestricted release is not yet a common procedure, and for the components considered border-line cases the available information is not sufficient to decide whether decontamination is appropriate, taking into account the attendant radiation exposure, secondary waste production and cost.

A low level and penetration depth of the contamination and a low surface to volume ratio of the component are characteristics which are favourable for decontamination. The range of components for which decontamination for unrestricted release is feasible is being enlarged by the development of more efficient decontamination procedures, which may be more aggressive than those currently used in operating plants, since weakening of the components treated can be accepted.

The following techniques have been developed for the decontamination of metals:

• techniques using chemically aggressive decontaminants in liquid and in gel-like form;
• electrochemical techniques;
• hydromechanical techniques, i.e., high-pressure water lance and erosion by cavitation.

For the decontamination of concrete, flame scarifying has been developed. With this new technique, which is based on rapid heating using an oxyacethylene torch, thin surface layers (about 1 mm in each pass) can be removed from concrete structures.

Both electropolishing and flame scarifying have been successfully tested at a significant scale in a shut-down Boiling Water Reactor plant.

Also in the framework of this Project, valuable information has been obtained on the distribution, the radionuclide composition, the chemical nature and the penetration depth of surface contamination in nuclear power plants.

The objective of this research is to develop and assess techniques for decontaminating surfaces of components and structures of nuclear installations that are past use.

The main purpose of decontamination would be reduction of the occupational radiation exposure during dismantling of the contaminated item and/or reduction of the volume of radioactive waste.

Project/Area Nr. 3: Dismantling techniques

In the decommissioning of nuclear power plants, large radioactive components and structures must be dismantled.

Important examples are:

• the highly radiating reactor internals of light-water reactors, which are made of stainless steel with thicknesses up to a hundred millimetres;
• the reactor pressure vessels of light-water reactors, which consist of stainless steel clad carbon steel, with thicknesses up to 300 mm in the wall and up to 600 mm at the vessel flange;
• biological shielding structures, consisting of reinforced concrete of several metres thickness.

The dismantling of substantially radioactive components must be carried out remotely with adequate radiation shielding for personnel protection. Where possible, segmenting may be performed under water.

There exist various conventional dismantling techniques which may be employed in decommissioning.

Generally speaking, thermal segmenting techniques tend to achieve high cutting rates but pose problems due to the production of fine particles. Mechanical techniques tend to be time-consuming and to need heavy manipulators.

Various thermal and mechanical cutting techniques have been tested on components made of stainless steel, carbon steel and concrete, using existing equipment. These tests have served to optimize process parameters, to assess and compare cutting performances and by-product characteristics, and to examine the suitability of filter systems.

Three thermal cutting techniques have been developed in particular, i.e., plasma cutting (including underwater operation), plasma-oxygen cutting and laser cutting.

For the dismantling of radioactive reinforced concrete structures, a diamond-tipped circular saw has been developed which achieves substantially higher cutting rates than equipment available before.

In a recent test of the saw, a one metre cube was cut out from a vertical reinforced concrete wall. As an alternative to sawing, explosive techniques have been investigated with a view to employ them with a high degree of control.

Systems for remotely controlled decotrmissioning operations have been reviewed in order to identify the areas needing future research.

The objective of this research is the development of the special techniques needed for dismantling the large steel components (e.g. reactor pressure vessel) and reinforced-concrete structures (e.g. reactor shielding) of redundant nuclear installations, account being taken of the particular requirements due to radioactivity.

Project/Area Nr. 4: Treatment of specific waste materials: steel, concrete and graphite

The waste from the dismantling of nuclear power plants consists mainly of steel, concrete and - for gas-cooled reactors - graphite. Methods for treating these materials with a view to disposal or, for steel, possibly reuse have been investigated.

One of these methods is melting of steel, which offers several advantages with regard to waste disposal, i.e. reduction of the bulk volume, reduction of the surface exposed to corrosion, immobilization of surface contamination by incorporation into the base metal.

On the other hand, the possible decontamination effect of melting is important with a view to the recycling of steel.

Melt tests using low-level radioactive steel from different reactor types have been carried out in order to determine the distribution of the various radionuclides on the metal product, the slag and the off-gas.

An installation for melting radioactive steel waste, which can be repeatedly mounted and operated within the controlled area of nuclear power plants finally shut down, is being designed. Experiments aimed at separating cobalt from steel, using special melting processes, failed to achieve separation factors high enough to encourage further work in tnis direction.

Coating techniques using epoxy or polyester resins to durably fix surface contamination on metal and concrete have been developed. The immobilization of concrete dust with silicate solutions has been investigated.

Various management modes have been assessed for graphite waste, considering incineration, sea disposal, shallow land burial and deep geological-disposal. This research involved in particular determination of the radionuclide inventory of Magnox Reactor and Advanced Gas-cooled Reactor graphite, conceptual studies of the incinerator flow-sheet and of waste packages, leach tests of irradiated graphite samples, and radiological assessment of the management modes considered.

The area has been strictly delimited to preclude overlapping with the Community research programme on radioactive waste management.

Project/Area Nr. 5: Large waste containers - Qualification and adaptation of remote-controlled semi-autonomous manipulator systems

Radioactive waste resulting from the dismantling of major reactor components should be transported and disposed of in larger units than those used for other types of radioactive waste, in order to reduce the required amount of segmenting and, consequently, the radiological impact and cost of decommissioning.

As a first step, a system study has been carried out, which has led to the definition of four types of waste container for further elaboration, all as large as possible within the limits set by the railway freight gauge.

Because of radiation fields, some decommissioning tasks must be performed with "remote control", in order to minimise occupational exposure. This requirement forms a major technical challenge in decommissioning.

The objective of this research is to qualify and adapt remote-controlled semi-autonomous systems for manipulation of decommissioning tools and instruments.

Project/Area Nr. 6: Estimation of the quantities of radioactive wastes arising from the decommissioning of nuclear installations in the Community

The low-level radioactive waste produced in the dismantling of nuclear installations will ultimately constitute a substantial part of the overall volume of radioactive waste generated by nuclear industry.

Estimates of the future arisings of such waste are therefore needed for the planning of waste disposal facilities. The radioactivity associated with the waste must be known, with particular reference to long-lived radionuelides, in order to classify the various types of waste with respect to the appropriate disposal modes.

This knowledge is also needed to predict the decrease of radiation levels as a function of time, which is an important factor to be considered in the timing of dismantling. This timing will in turn determine the point in time where radioactive waste from decommissioning is to be finally disposed of.

Since this Project involves the definition of reference strategies for decommissioning, it is regarded as a long-term task. The work carried out under the 1979-1983 programme has been focused on two particular aspects, i.e. long-lived activation products in concrete and steel, and basic scientific information related to the border-line between radioactive and non-radioactive material.

Radionuclide contents have been measured on activated steel and concrete samples from Boiling Water Reactor Diarits finally shut down. In particular, the thickness of the activated inner layer of biological shields has been determined by the examination of boring samples.

A number of relevant source elements for the activation of steel and concrete are trace elements whose concentrations are not defined in material specifications. Data on these concentrations, sometimes at the ppm level, are needed for the calculation of activation.

In order to explore the ranges of the relevant trace element concentrations to be expected in the materials used in nuclear power plants, non-radioactive samples from a large number of origins were analysed. These samples included reactor-graphite stainless steel and carbon steel from various manufacturers as well as concrete from nuclear power plants in various Community countries.

A methodology for evaluating the radiological consequences of the management of very low-activity steel and concrete from the dismantling of nuclear power plants has been prepared.

Such a methodology is necessary if "de minimis" radioactivity levels are to be defined. The measuring techniques required for proving that a material is non-radioactive have been reviewed.

The objective of this area is to estimate the quantities of various categories of radioactive waste that will arise from the decommissioning of nuclear installations in the Community. This involves the definition of reference strategies for decommissioning and is therefore to be regarded as a long-term task.

Project/Area Nr. 7: Plant design features facilitating decommissioning

The design of nuclear power plants has continuously evolved and many improvements made with a view to operation of the plants will also facilitate decommissioning.

An example are improvements of radioactivity barriers, such as fuel claddings and steam generators, which reduce the contamination of the plants.

On the other side, increased safety requirements tended to increase the volume of the components and structures to be dealt with in decommissioning.

The objective of this Project has been to identify and develop reasonable improvements in the design of nuclear power plants with a to decommissioning. Though requirements related to the safety and of reactor operation, as well as cost-effectiveness aspects latitude for modifying plant designs, the following, specific features appeared to merit closer investigation:

• control of the cobalt and niobium content of reactor steels;
• cobalt-free materials to substitute cobalt alloys used, e.g., in valve seatings;
• coatings, in particular removable ones, to protect concrete against contamination;
• reactor shielding design features facilitating dismantling;
• documentation system for deferred decommissioning.

Section B: Identification of guiding principles

Project/Area Nr. 8: Identification of guiding principles

The identification of guiding principles in the field of decommissioning has been considered a long-term task, since it must be based on an appropriate body of technical information. The "1979-1983 programme" comprised a number of relevant studies in this regard.

Moreover, available material in the Member States that could serve as basis for guiding principles in the field of decommissioning has been assembled. It appeared that the existing systems of authorisation and control, together with the radiological protection standards in force, already make it possible to decommission nuclear power plants on a case-by-case basis.

However, no specific technical regulations on decommissioning exist. A basic uncertainty as regards final release of sites for re-use, lies in the absence of specified criteria for distinguishing non-radioactive material from radioactive material.

References

1. Council (1979). Council Decision of 27 March 1979 adopting a research programme concerning the decommissioning of nuclear power plants, OJ N° L 83, 3.4.1979, p.19.
2. LORCHER, G. and PIEL, W. (1983). Dekontamination von Komponenten stillgelegter Kernkraftwerke für die freie Beseitigung, EUR 8704.
3. Commission (1981). The Community's research and development programme on decommissioning of nuclear power plants. First annual progress report (year 1980), EUR 7440.
4. Commission (1982). The Community's research and development programme on decommissioning of nuclear power plants. Second annual progress report (year 1981), EUR 8343.
5. Commission (1984). The Community's research and development programme on decommissioning of nuclear power plants. Third annual progress report (year 1982), EUR 8962.
6. Council (1984). Council Decision of 31 January 1984 adopting a research programme concerning the decommissioning of nuclear installations, OJ N° L 36, 8.2.1984, p. 23.

List of projects per Project/Work Area

Project/Area n° 1: Long-term Integrity of Buildings and Systems
DE-A-001-UK	United Kingdom	Degradation of building plant and materials
DE-A-002-F	France	Long-term integrity of buildings and systems
Project/Area n° 2: Decontamination for Decommissioning Purposes
DE-B-001-I	Italy	Delegation of an Expert to the USNRC in Relation to the Clean-up of the TMI-2 Plant
DE-B-002-D	Germany	Decontamination of concrete surfaces by flame-scarfing
DE-B-003-F	France	Erosion of metal surfaces by cavitation at very high velocity
DE-B-004-D	Germany	Composition of contamination layers and efficiency of decontamination
DE-B-005-I	Italy	Vigorous decontamination tests of steel samples in a special test loop (Deco loop)
DE-B-006-D	Germany	Development of economical decontamination procedures
DE-B-007-F	France	Research and development on decontamination by gel-based decontaminants
DE-B-008-F	France	Metal decontamination by chemical and electrochemical methods and by water lance
DE-B-009-D	Germany	Economic Assessment of Decontamination for Unrestricted Release
DE-B-010-F	France	Erosion of metal surfaces by cavitation at very high velocity: Testing of new cavitating elements
DE-B-011-B	Belgium	Delegation of an expert to the US-NRC in relation to the clean-up of the TMI-2 plant (US-NRC: United States Nuclear Regulatory Commission)
Project/Area n° 3: Dismantling Techniques
DE-C-001-D	Germany	Comparative Assessment of Various Thermal and Mechanical Dismantling Techniques
DE-C-002-I	Italy	Plasma techniques for cutting mineral and metal materials
DE-C-003-UK	United Kingdom	Large diamond-tipped circular saws for cutting reinforced concrete structures
DE-C-004-D	Germany	Combined plasma-oxygen torch for cutting steel pressure vessels
DE-C-005-I	Italy	Dismantling of metal components and concrete structures using laser beam
DE-C-006-UK	United Kingdom	Explosive demolition techniques for concrete structures
DE-C-007-F	France	Segmenting of steel components using intergranular fissuration
DE-C-008-D	Germany	Review of systems for remotely controlled decommissioning operations (1)
DE-C-009-B	Belgium	Review of systems for remotely controlled decommissioning operations (2)
Project/Area n° 4: Treatment of Specific Waste Materials: Steel, Concrete and Graphite
DE-D-001-UK	United Kingdom	Assessment of various management modes for radioactive graphite waste
DE-D-002-UK	United Kingdom	Treatment of contaminated steel waste by melting
DE-D-003-F	France	Immobilization of contamination on metals by coating with thermosetting resins
DE-D-004-F	France	Removal of cobalt from steel using special melting processes
DE-D-005-UK	United Kingdom	Treatment of concrete with silicate solutions to prevent dusting
DE-D-006-I	Italy	Coating of materials to prevent corrosion, fix contamination, avoid powder formation
DE-D-007-F	France	Immobilization of contamination on metals by coating with thermosetting resins
DE-D-008-D	Germany	Design of an installation for melting radioactive steel waste
Project/Area n° 5: Large Containers for Radioactive Waste Produced in the Dismantling of Nuclear Power plants - Qualification and adaptation of remote-controlled semi-autonomous manipulator systems
DE-E-001-UK	United Kingdom	System of large transport containers for waste from dismantling light water reactors and advanced gas-cooled reactors (1)
DE-E-002-B	Belgium	System of large transport containers for waste from dismantling light water reactors and advanced gas-cooled reactors (2)
Project/Area n° 6: Estimation of the Quantities of Radioactive Wastes Arising from the Decommissioning of Nuclear Installations in the Community
DE-F-001-D	Germany	Activation products in the biological shield of the Lingen reactor
DE-F-002-D	Germany	Activation products in the biological shield of the KRB-A reactor
DE-F-003-I	Italy	Activation and radiation at the Garigliano reactor pressure vessel
DE-F-004-UK	United Kingdom	Trace element content and activation of low-alloy and stainless reactor steels
DE-F-005-F	France	Determination of trace elements on concrete samples from various nuclear reactors
DE-F-006-UK	United Kingdom	Methodology for evaluating radiological consequences of the management of low-level radioactive waste from the dismantling of nuclear power plants (1)
DE-F-007-F	France	Methodology for evaluating radiological consequences of the management of low-level radioactive waste from the dismantling of nuclear power plants (2)
DE-F-008-D	Germany	Review of techniques for measuring very low-level radioactivity in relation to decommissioning (1)
DE-F-009-F	France	Review of techniques for measuring very low-level radioactivity in relation to decommissioning (2)
Project/Area n° 7: Influence of Installation Design Features on Decommissioning
DE-G-001-UK	United Kingdom	Catalogue of design features facilitating decommissioning of AGRs
DE-G-002-UK	United Kingdom	Design features of civil works of nuclear installations facilitating refurbishing or dismantling
DE-G-003-D	Germany	Erosion-corrosion testing of cobalt-free materials to substitute cobalt alloys
DE-G-004-UK	United Kingdom	Control of the cobalt content of reactor-grade steels
DE-G-005-I	Italy	Removable coatings for the protection of concrete against contamination
DE-G-006-F	France	Characterization and improvement of coatings protecting concrete against contamination
DE-G-007-I	Italy	Evaluation of the design features facilitating decommissioning of PWRs
DE-G-008-UK	United Kingdom	Concepts minimizing the activation of the biological shield
DE-G-009-D	Germany	Biological shield design with dose-reducing effect for decommissioning
DE-G-010-D	Germany	Documentation system for decommissioning of nuclear power plants

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