Support to plant operation and safety (ANAV)

For the last 15 years GET has been working in the field of full power plant models along with Asociación Nuclear Ascó–Vandellòs (ANAV), which is a utility that presently runs three operating PWRs. In this framework, GET has developed and maintained plant models for the three power plants.


Extensive technical literature exists aimed at establishing the requirements needed to qualify a Nuclear Power Plant model. Most of this literature is focused on qualifying a model for licensing uses. Less documentation is available nowadays on the requirements needed when an Integral Plant Model is used for supporting plant operation and control of an actual commercial facility, while fulfilling its goals of safety and competitiveness. GET has developed an advanced qualification process (AQP) of plant models for operation support, introduced the concept of plant configuration.

SCDAP Development Traninig Program (SDTP)

GET participates to the international SCDAP Development Traninig Program (SDTP) which consists of more than 90 organizations in 30 countries supporting the development of technology, software, and training materials for the nuclear industry. The program members and licensed software users include universities, research organizations, regulatory organizations, vendors, and utilities located in Europe, Asia, Latin America, and the United States. The software supported through this program includes RELAP/SCDAPSIM code versions, a best estimate thermal hydraulic code designed to predict the behaviour of reactor systems during normal and accident conditions, including severe accident stages.

Innovative Systems Software (ISS) administers the program and is responsible for the configuration control and distribution of the RELAP/SCDAPSIM code.

Coupled calculations (TH-Neutron kinetics)

The group has developed the competences to perform calculations of thermal hydraulic codes coupled with 3D neutron kinetics. In particular calculations have been performed with TRACE/PARCS and RELAP/PARCS.


In addition, BEPU calculations have been carried out with a TRACS/PARCS model of a full NPP including uncertainties of both thermal-hydraulics parameters and neutron kinetics parameters.

Integral Test Matrix and participation to international experimental programs

GET participates actively to several international projects launched by the OECD/NEA. The participation of GET is funded through the "Consejo de Seguridad Nuclear". The main activities of GET in these projects are the pre and post-test calculations of the experiments performed at Integral Test Facilities like LSTF and PKL. The simulation of these experiments allow both the assessment of the codes used for future application to commercial power plant safety analyses and the training and learning in the use of the codes for complex systems. GET has developed a database in a form of a wikipedia of available experiments. The experiments are categorized in scenarios and phenomena. In this way, the available information can be then used in the following areas:

  • Improvement and validation of NPP models through our own scaling methodology
  • selection of uncertainty parameters for BEPU analyses

Validation of NPP models through Scaling of experiments

System codes along with necessary nodalizations are valuable tools for thermal hydraulic safety analysis. Qualifying both codes and nodalizations is an essential step prior to their use in any significant study involving code calculations. Since most existing experimental data come from tests performed on the small scale, any qualification process must therefore address scale considerations.

GET has developed a Scaling-up Methodology to allow the qualification of the NPP models in certain scenarios. An important development has been performed in order to identify and justify discrepancies that appear between counterpart simulations at different scales and designs. The methodology is a systematic procedure for qualifying NPP nodalizations taking advantage of the experience acquired through the post-test analysis of ITF experiments. It is devoted to the modeling qualification, which implies that the methodology can only be applied to those phenomena that have been well reproduced in ITF post-test analyses, and that scaling analyses are only performed through code simulations (and do not involve experimental data)

RELAP/SCDAPSIM/MOD4.0 modification for transient accident scenario of Test Blanket Modules in ITER

GET colaborates with The Institute for Plasma Research (IPR), India, for safety analysis of the Indian Test Blanket Module (TBM) for ITER (International Thermo nuclear Experimental Reactor). The Indian TBM concept is a Lead-Lithium cooled Ceramic Breeder (LLCB), which utilizes lead-lithium eutectic alloy (LLE) as tritium breeder, neutron multiplier and coolant. The first wall facing the plasma is cooled by helium gas. Thermal hydraulic safety analyses of the TBM system will be carried out with the system code RELAP/SCDAPSIM/MOD4.0 which was initially designed to predict the behavior of light water reactor systems. In order to analyze some of the postulated off-normal events, there is the need to simulate the mixing of Helium and Lead-Lithium fluids. The Technical University of Catalonia is cooperating with IPR to implement the necessary changes in the code to allow for the mixing of helium and liquid metal. A new flow regime map for LLE and helium flows is being developed on the basis of numerical simulations with the OpenFOAM CFD toolkit.

BEPU Analyses (Best Estimate Plus Uncertainties)

The use of best estimate codes in combination with evaluation of uncertainties, the so-called BEPU methodologies, is an accepted procedure by the regulatory authorities to carry out deterministic safety analyses.


GET has participated and lead phases of the BEMUSE program, an international project on BEPU methodologies. Through this participation, GET has developed its own BEPU methodology in cooperation with the Spanish Nuclear Regulatory body(CSN). The methodology is based on the German methodology developed at GRS, which uses a name-order statistics for the Wilks formula (Wilks, 1941). Based on experimental data and expert information, a list of uncertain parameters is defined. Each uncertain parameter is considered as an input value with an associated probability density function (pdf). Once the parameters have been specified, a certain number (N) of sets of parameter's samples are randomly generated according to the pdf of each single parameter. These sets of parameter samples are used to define N code runs of the same transient. The Wilks formula gives the number N of runs to be performed in order to obtain an estimation of a given simulation output within a-percentile with a 95% confidence level.

Quantification of input parameters uncertainty through SET experiments (PREMIUM)

PREMIUM (Post BEMUSE Reflood Models Input Uncertainty Methods) is an activity launched with the aim to push forward the methods of quantification of physical models uncertainties in thermal-hydraulic codes. It is endorsed by OECD/NEA/CSNI/WGAMA.


PREMIUM deals with uncertainty evaluation methods based on input uncertainties quantification and propagation. The benchmark is based on a selected case of uncertainty analysis application to the simulation of quench front propagation in an experimental test facility. The Application to an experiment enables evaluation and confirmation of the quantified probability distribution functions on the basis of experimental data. The scope of the benchmark comprises a review of the existing methods, selection of potentially important uncertain input parameters, preliminary quantification of the ranges and distributions of the identified parameters, evaluation of the probability density function using experimental results of tests performed on FEBA test facility and confirmation/validation of the performed quantification on the basis of blind calculation of Reflood 2-D PERICLES experiment.