High Level Waste

High-level waste (HLW) is radioactive waste containing appreciable quantities of radioactive products (long-lived alpha emitters and beta-gamma emitters), that is highly radioactive and generates a significant amount of heat. This is essentially spent nuclear fuel (SF) from nuclear power plants and vitrified waste produced in the reprocessing of small quantities of SF.

Special waste (SW) is long-lived waste of a significant activity level, the temporary and definitive management of which will be similar to high-level waste.

Characteristics of spent fuel (SF)

Nuclear fuel comprises a set of cylindrical ceramic pellets of uranium oxide, U-238, with a degree of U-235 enrichment (less than 5%), positioned within tubes made from a zirconium-rich alloy known as Zircaloy, and assembled within a structure comprising the fuel element.

Once the fuel is loaded into the nuclear reactor, it undergoes certain reactions comprising neutron capture and nuclear fission in part of the uranium and other radionuclides generated, giving rise to fission products, activation products and the generation of plutonium and minority actinides, presenting a composition containing practically all elements of the periodic table. These reactions give off a great amount of heat.

The quantities and characteristics of the different components of the SF depend on its initial U-235 enrichment and the degree of burn-up of the fuel and how the reactor was operated. The SF composition varies over time because of nuclear reactions, disintegration processes and other nuclear reactions which take place inside, and also depends on the time that has passed since it was unloaded from the reactor (cooling time).

Infographic of a fuel element

Infographic of a fuel element


Nuclear Fuel Cycle




The nuclear fuel cycle covers every stage from the extraction of the uranium ore, its concentration, enrichment and production of fuel elements for use at the nuclear power plant where the reactor causes the fission reaction (first part of the cycle), to the management of the radioactive waste generated (second part of the cycle).

Explanation of the image:
Diagram explaining the open and closed spent fuel cycles; in the open cycle: the spent fuel is considered high-level waste and is managed as such, being stored at specific temporary facilities. The definitive disposal of spent fuel would be in a deep geological repository, constituting its final management as waste.
In the closed cycle, the spent fuel is partially reused by means of reprocessing and recycling. This comprises the recovery of those components of the used fuel that have energy potential, essentially the uranium and plutonium, to be reused in a reactor. This fuel, comprising plutonium and uranium oxides, is known as MOX.
The other components of the spent fuel (fission and activation products and other actinides, structural materials) are considered waste, and are conditioned and transported to a storage facility. As in the open cycle, their final management requires them to be housed in a deep geological repository.
Diagram explaining the open and closed spent fuel cycles


Once it has been extracted from the nuclear reactor, the spent fuel element must be stored underwater at all times, to be cooled in the pools at the nuclear power plant. The choice of water as host medium is based on its high heat transfer coefficient, its good shielding properties, transparency and manageability. Following several years of cooling, the following options are available:

Open cycle:

The SF is temporarily stored in the pools at the nuclear power plants, or in other (individual or centralised) temporary storage systems, awaiting final management in a deep geological repository (DGR).

Closed cycle:

The SF is reprocessed, to recover the uranium and plutonium it still contains and manufacture new fuel. The resulting radioactive waste is vitrified and managed as HLW in temporary (individual or centralised) storage facilities awaiting final management in a deep geological repository.



Spent fuel management

Spain has opted for the open cycle, in other words it does not reprocess the SF. Only the SF from the Vandellós I nuclear power plant, in its entirety, and small amounts from the José Cabrera and Santa María de Garoña plants, were reprocessed in the past. The vitrified waste resulting from the reprocessing of spent fuel from Vandellós I is stored in France, and must return to Spain.

The strategic approach addressed in the sixth Radioactive Waste Management Plan in Spain for the temporary management of SF, HLW and SW focuses on the commissioning of a Centralised Temporary Storage (CTS) facility, while for final management, a DGR is expected to be available by the 2070s.



Temporary storage systems



 
Nuclear power plant´s fuel pool

Storage underwater in nuclear power plant pools

Once the SF is unloaded from the nuclear reactor, it must be stored underwater at all times in pools at the nuclear power plant, in order to cool. The choice of water as host medium is based on its high heat transfer coefficient, its good shielding properties, transparency and manageability.

Individualised Temporary Storage (ITS) facility

Dry storage in casks at the ITS facility

At some power plants the capacity of the pools is reaching its limit, or the need to evacuate fuel from the pools to begin decommissioning has been raised. In this case, the fuel is placed in casks which are stored at an appropriate facility at the power plant site. This is known as an Individualised Temporary Storage (ITS) facility.

The temporary storage casks may be of different types, such as dual-purpose metal casks (storage and transportation) or welded metal canisters stored in concrete-metal modules that can be transported in a metal cask.



Individualised temporary storage (ITS) facility at the different nuclear power plants (NPP)


Trillo ITS facility

The Trillo nuclear power plant has since 2002 had a cask storage facility in order to temporarily hold the site's SF. This is a hall with concrete walls and roof capable of holding up to 80 dual-purpose casks (storage and transport).

Trillo´s nuclear power plant ITS facility
José Cabrera ITS facility

Facility located on the site of the power plant and designed for the dry storage of all SF unloaded from the reactor.

It comprises a reinforced concrete slab to support the storage modules, surrounded by outer radiation protection fencing and inner security fencing to delimit the storage area.

This ITS facility holds 12 casks loaded with SF and a further 4 which house the more active metal parts obtained during the internal reactor component segmentation dismantling.

José Cabrera´s nuclear power plant ITS facility
Ascó ITS facility

Facility located within the power plant site and designed for the dry storage of SF unloaded from the two reactors.

It comprises two reinforced concrete slabs with capacity for 16 storage modules per slab. It is surrounded by an outer radiation protection fencing system and an inner physical security fence to delimit the storage area.

Ascó´s nuclear power plant ITS facility
Almaraz ITS facility

This storage facility employs similar technology to the above, based on a concrete slab with dual-purpose metal casks on the surface.
It began operation in 2018 when the first ENUN-32P cask was loaded.

Almaraz´s nuclear power plant ITS facility
Santa María de Garoña ITS facility

A storage facility with similar technology to the above, based on a concrete slab with dual-purpose metal casks, which will have overpack shielding. The storage facility has been built with physical capacity for 32 ENUN-52B casks (2 slabs with 16 positions each).

In 2020, the regulatory body (the CSN) was also presented with a revised design of the ITS facility to accommodate all the fuel housed in the power plant pool.

Santa Marís de Garoña´s nuclear power plant ITS facility
Cofrentes ITS facility

A storage facility with similar technology to the above, based on a concrete slab with dual-purpose metal casks, and capacity for 24 HI-STAR 150 cask storage positions.

This facility has been built and is scheduled to begin operation in 2021.

Cofrente´s nuclear power plant ITS facility


The CTS facility as a strategic goal


The Centralised Temporary Storage (CTS) facility addressed in the sixth Radioactive Waste Management Plan is designed for temporary dry storage at one single facility housing both spent fuel (SF) produced by Spain's stock of nuclear plants, and also high-level and special waste (HLW and SW).



The CTS facility can:

  • Manage both spent fuel and high-level waste/special waste under optimal conditions at one single location.
  • Manage the vitrified waste stored in France.
  • Release all nuclear power plant sites once their decommissioning is complete.
  • Decouple temporary and definitive storage.
Meanwhile, it requires fewer resources to maintain the same standards of safety and security as distributed management at individual facilities, resulting in greater resource usage efficiency.



The CTS facility is designed for an operational life of around 60 years, following which all the waste would be transferred to the DGR and the facility would then be decommissioned.


The decision to store SF and HLW at one single location in Spain is based on a December 2004 Decision of the Industry Committee in the Lower House of Parliament, calling on the Government, in partnership with Enresa, to develop the criteria to implement a temporary storage facility for SF and HLW in Spain.

The current 6th GRWP and the draft 7th GRWP now being processed maintain the strategy of unified temporary management of SF, HLW and SW at one single facility, allowing for the operational start-up of a CTS.

In order to conduct the facility site selection process, Royal Decree 775/2006 of 23 June 2006 was approved, setting up an Inter-ministerial Committee to establish the criteria that the CTS facility site must fulfil, giving rise to a selection procedure resulting in approval of the designation of Villar de Cañas in Cuenca as the site for the facility and associated technological centre, by means of the Resolution of the Council of Ministers of 30 December 2011.

On the basis of the above, in January 2014, Enresa presented the then Ministry of Industry, Energy and Tourism with an application for prior authorisation and construction authorisation for this facility. The CSN issued its favourable report with regard to the prior authorisation request on 27 July 2015.

However, before either of these two authorisations had been granted, in July 2018 the State Secretariat for Energy requested that the CSN and the State Secretariat for the Environment suspend both the issuance of the required report on the construction authorisation, and the handling of the environmental assessment procedure, respectively, until the 7th GRWP was approved.



Definitive storage system

In accordance with Council Directive 2011/70/Euratom of 19 July 2011, establishing a Community framework for the responsible and safe management of spent fuel and radioactive waste, the broadly accepted concept at present is that a deep geological repository (DGR) represents the safest and most sustainable option as the endpoint of the management of SF and HLW.

The aim pursued by the storage of HLW in deep geological formations is to prevent the radioactive substances they contain from reaching the human environment in concentrations that could harm the natural world, and hence human health.

In order to achieve this, the waste must be isolated for lengthy time periods, allowing the level of activity of the various radioactive elements it contains to decay to sufficiently low values so as not to alter the natural background radiation, and not to increase normal doses for human beings.

The DGR is based on the so-called multi-barrier principle, which involves placing a series of artificial and natural barriers between the waste and the biosphere. Safety does not rely on one single barrier, but on the combined action of different barriers with different functions. The aim is that any deficiencies that could arise in the performance of one barrier over time would not compromise the overall safety of the system.

These barriers act in two different ways:

  • First, they serve to contain the radioactive materials
  • Furthermore, they delay and dilute potential releases into the biosphere comprising the set of ecosystems which will suffer the potential impact from the disposal facility (soil, water, living beings, etc.).

There are two types of barrier or component to this concept: artificial and natural.


Explanatory text AGP
xxxx
Combination of artificial and natural barriers on a deep geological repository

Artificial barriers

Artificial barriers or engineered barriers, are designed, built and installed in accordance with the design of the disposal facility, the specific function or functions assigned to them, and the conditions imposed in the short and long term by the other artificial and natural barriers within the system.

The components of the artificial or engineered barriers are (see figure):

  • The chemical form of the waste itself.
  • The metal storage canisters.
  • The filling and sealing materials.

Natural barriers

Natural barriers are not specified or built by human hand, but must be characterised and selected in accordance with certain criteria, or functional requirements, that would allow the artificial barriers and the whole system to function properly.

The components of the natural barriers are:

  • The geosphere. Geological formations housing the repository, and the waters and gases they contain.
  • The biosphere. Set of ecosystems (soil, water, living beings, etc.) that would suffer the impact of the repository.

A geological barrier should be understood as the geological formation where the disposal facility is located, which essentially comprises a solid part, made up of rocks and minerals, and a fluid part made up of water and gases.

The natural barrier is responsible for the long-term safety of the system, delaying the emergence of radionuclides into the human environment, and controlling their dispersal and dilution. The artificial barriers play a decisive role in short-term safety, given their capacity for containment and delay.

Since 1985, Enresa has worked on the DGR option in four basic directions:


The Site Search Plan (SSP) allowed sufficient information to be gathered to ascertain that the subsoil of Spain contains plentiful granite and clay formations capable of housing a disposal facility.


Generation of conceptual designs of a disposal facility for each of the stated lithological contexts, aiming for the greatest number of points in common among all of them.


Development of Safety Assessment exercises for the conceptual designs (granite and clay), incorporating the knowledge built up during work and projects under the successive R&D Plans undertaken, highlighting that DGRs are capable of complying with the safety and quality criteria applicable to this type of facility.


Development of successive R&D plans which have gradually evolved in line with SF and HLW management plans, serving to acquire technical know-how and participate in national/international research projects at underground laboratories abroad.



Considerable research efforts have also been dedicated to the different versions of separation and transmutation technologies with the aim of reducing the level of activity and duration of the elements and facilitating their DGR storage, although the scale of such programmes necessarily requires participation at an international level.

As a result of the studies performed between 1986 and 1996, conducting an analysis of suitable geological formations to house the DGR site, an Inventory of Suitable Formations was drawn up.

Two generic designs are available, along with the safety assessment associated with both, tailored to a granite and a clay-type host medium. These advances will constitute a sound basis for the launch of the forthcoming stages for the selection of the site and the implementation of the DGR.

The reference concept assumes final storage of the SF and other HLW in carbon steel capsules surrounded by an appropriate sealant material. These capsules will be located in horizontal tunnels positioned at sufficient depth, which will vary according to whether the formations are clay- or granite-based.

The reactivation of the DGR programme in Spain, together with its regulation and a specific implementation plan, were a recommendation of the combined IRRS/ARTEMIS mission to evaluate the Spanish regulatory framework regarding nuclear and radiological safety (IRRS), and the radioactive waste management system (ARTEMIS), so as to comply with:

  • Directive 2009/71/EURATOM establishing a Community framework for the nuclear safety of nuclear installations, and
  • Directive 2011/70/Euratom of 19 July 2011 establishing a Community framework for the responsible and safe management of spent fuel and radioactive waste.

The draft 7th GRWP, currently being processed, considers DGR disposal to be the preferred and basic option for the management of SF and HLW, and in terms of economic and planning calculations this would enter operation from 2073 onwards, following a prior period of temporary storage of SF.



Transport of high-level waste (HLW)



There is at present no transport of high-level waste and spent nuclear fuel in Spain, as such waste remains in the pools or individualised interim storage facilities at the power plants themselves.

Transport container for high-level waste and spent nuclear fuel
Preparatory transport work

Given the need for the Spanish Centralised Temporary Storage facility, a Transport Plan has been drawn up to allow high-level waste and spent nuclear fuel to be transferred from the nuclear power plants to the new site. This activity is subject to the same standards and criteria as employed in the transportation of low- and intermediate-level radioactive waste.


For the transportation of high-level waste and spent nuclear fuel, the containers used will be designed to withstand so-called "accident conditions" in accordance with the ADR (Agreement concerning the International Carriage of Dangerous Goods by Road).ADR (Agreement concerning the International Carriage of Dangerous Goods by Road).

More than 30 million kilometres of transportation of such materials have been conducted worldwide, without any radiation incident having occurred..



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