General description

COSAC is used to study the flow and the storage of nuclear materials in a nuclear fleet composed of the following installations:

• mines
• fresh fuel manufacturing plants.
• reactors
• cooling pools
• spent fuel reprocessing facilities
• interim and ultimate storage facilities

COSAC enables the modelling of each of these installations and their connections between each other. Modelling the nuclear installations consists of taking account of the parameters that have an incidence on the evolution of the amount and the isotopia of nuclear materials.

For instance, for a cooling pool, the duration of the stay of the nuclear material in the pool will have an incidence on the isotopia; for a fuel manufacturing plant, U235-enrichment of the UOX fuel as well as enrichment of the tails will have an incidence on the amount of natural uranium required and on the amount of depleted uranium produced.

The user can integrate as many as wanted installations to build a scenario and introduce specific parameters in each of these installations to model the impact of the scenario on the nuclear materials.

No predefined scenario is integrated in the code.

COSAC can compute several physics outputs: mass and isotopia of the nuclear materials, decay power and radiotoxicity. The physics outputs are time-dependent and can be edited for a peculiar installation, a group of installations or for the whole scenario. No economics computations are done in COSAC.

The programing language is C++ and a typical run lasts a few minutes.

Main features

The development of COSAC was based on some basic following principles:

• the ability to model almost¹ any kind of scenarios
• the ability to add any isotope to the computation (either fission product or heavy nuclide)
• a simple Graphical User Interface to use the code: the GUI allows the user to enter input data to define a scenario and to edit the results in the form of tables or graphics
• a short run time so that a scenario (even complex) can be computed in a few minutes

The code handles the data that have been entered by the user. It does not contain physics models. All the physics is done outside the code and then introduced in the code by the means of matrices.

For instance, radioactive decay is modelled in COSAC by applying a decay matrix to the mass vector. In the same way, fuel depletion under irradiation is modelled by applying an irradiation matrix to the mass vector.

The decay matrices and irradiation matrices are computed outside the code and introduced in the code by the user as any other input data.

Decay matrices can be obtained by an external code using the Bateman equations. Irradiation matrices can be obtained by using an external depletion code such as APOLLO or a similar code.

Modelling the fuel depletion under irradiation by an irradiation matrix is a simplified way to take account of the incidence of the fresh fuel composition (especially, Pu isotopia when fresh fuel is a MOX fuel) on the spent fuel composition. Even if simplified, this approach is accurate enough to run calculations with acceptable precision as far as the discharge burnup of the spent fuel doesn’t vary and the composition of the fresh fuel varies inside a given range of variation. This range of variation is to be defined at the same time as the irradiation matrix is defined.

The Pu content of a MOX fresh fuel can also be adjusted according to the Pu isotopia: the user can indeed introduce an equivalency formula in COSAC allowing the code to adjust the Pu content to the fissile quality of the plutonium and to the performances the MOX fuel management is expected to reach.

Two ways of managing the spent fuel reprocessing are available in COSAC: the user can either define an annual capacity of reprocessing (limited capacity) or let the reactor demand define itself the right capacity of reprocessing to feed all the reactors at each time (unlimited capacity). In the first case, the not-used Pu after reprocessing is stored in an interim storage, waiting to be used later by the reactors.

The minimum time step in COSAC is fixed to one month. The transportation of the nuclear materials from an installation to another is not modelled in COSAC so that all the transports are supposed to be done “instantaneously”.

The renewal of the reactor fleet after a given lifetime can be modelled or not in COSAC. This choice is decided by the user when defining the scenario. More generally speaking, most of the input parameters of a scenario can be easily changed by the user, so that the tune-up of a scenario can be greatly simplified by a short run time after each change.

¹ The scenarios involving continuous reloading reactors, such as bullet-High Temperature Reactors or Molten Salt Reactors, have not been tested in COSAC. Modelling such scenarios could require some new developments in COSAC.