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RELAP5 is a system program for analysis of transients during normal operation and emergency processes in thermohydraulic systems using fluid. The fluid may be a mixture of steam, water, non-condensable substance (gas) and non-volatile solute (boric acid). Fluid and energy flows are approximated in one-dimensional flow-pipe and conductive models. The program contains models of equipment components specific to pressurized water reactors (PWR). The program contains an ejector type component and can be used to model systems with boiling reactors (BWR).
In the case of pressurized water reactors for which the program is intended, heat loss accidents (LOCA), operational events and transient emergency processes such as:

  • transient emergency modes without reactor silencing;
  • loss of primary circuit or GHG nutrition;
  • loss of external power supply;
  • reactor supercooling.

The thermohydraulic behavior of the reactor plant and the behavior of the unit systems is usually tested until a maximum fuel element shell temperature ≥ 1200 ° C is reached. The deformation of the shell of the heat dissipating elements is not modeled. A model for estimating the oxidation of fuel element shells is included.


TRACE is the latest in a series of advanced thermohydraulic codes developed by U.S. Nuclear Regulatory Commission (NRC) for the analysis of stationary neutron and thermohydraulic processes, as well as for the analysis of transients in light water reactors. This computer program is the product of a long-term effort to combine the capabilities of the four basic system codes of NRC (TRAC-P, TRAC-B, RELAP5 and RAMONA) in one modernized computing tool.

TRACE is designed to perform best-rated analyzes such as coolant losses (LOCA), operational transients and other emergency scenarios in pressurized water reactors (PWR and WWER) and boiling reactors (BWR). The code can also be used to analyze phenomena in experimental installations that are designed to simulate transients in reactor systems. The models used include the consideration of two-phase flows in three-dimensional coordinate systems, nonequilibrium thermodynamics, various cases of heat exchange, re-flooding, level monitoring, neutron reactor kinetics and others.


MELCOR is a fully integrated, engineering-based computer code that allows modeling the development of severe accidents at NPPs with water reactors. MELCOR was developed by the US NRC (United States Nuclear Regulatory Commission) as a second generation risk assessment tool and as a follower of the computer code Source Term. With the help of the MELCOR program code, a wide range of phenomena accompanying the severe accident is analyzed, both in boiling and in pressurized water reactors. This includes thermohydraulic analysis (response) of the unit, reactor shaft, hermetic structure, heating, degradation and relocation of the core, interaction of the melt with the concrete, generation, transport and combustion of hydrogen, disposal and transport of radioactive products. The current version of MELCOR includes sensitivity and uncertainty in the assessment of radioactive discharges for a number of applications.


The PARCS package (Purdue Advanced Reactor Core Simulator) provides a non-stationary three-dimensional simulation of the core of a coolant reactor. PARCS calculates the equation of neutron transfer in the diffusion approximation, in a small number of energy groups, in order to track the change in the state of the reactor installation in the event of external disturbances such as movement OR or change in thermohydraulic parameters in the reactor core. PARCS is applicable to both pressurized water reactors (PWR, WWER) and fluidized bed reactors (BWR) and can be used in conjunction with system thermohydraulic codes such as TRAC / RELAP (TRACE, formerly known as TRAC-M) or RELAP5. The thermohydraulic code provides information on the thermohydraulic parameters (coolant and fuel temperatures, physical state and coolant composition, flow rate, etc.) required for PARCS during the simulated transient. PARCS uses the thermohydraulic parameters to update the neutron interaction cross sections and, in turn, provides the thermohydraulic code with the spatial distribution of energy release in the reactor core.

HRA Calculator

The EPRI HRA Calculator is a software tool designed to facilitate a standardized approach to human reliability analysis (HRA). Wide varieties of methodologies are used for HRA in probabilistic risk assessment (PRA). The results from these differing methodologies can vary considerably when comparing results between similar plants, or even when comparing the actions within the same plant that are evaluated by different analysts.

The EPRI HRA Calculator includes modules for preinitiator, post‐initiator and dependency analysis. The pre‐initiator module can encapsulate the procedure review and historical events review through which preinitiator human failure events (HFEs) are identified. For the quantification of pre‐initiator HFEs, the user can select ASEP or THERP.

The post‐initiator module functions on the basis that each post‐initiator HFE comprises a cognitive error and an execution error. The cognitive error can be analyzed using the EPRI CBDTM and HCR/ORE, THERP or SPAR‐H methods. The execution error is analyzed using THERP. The user can analyze an HFE using any or all of the applicable methods.

The dependency analysis module identifies combinations of HFEs in cutsets from WinNUPRA or CAFTA, and Riskman sequence data that can be imported directly. The user can specify to analyze combinations of preinitiators, post‐initiators or both pre and post‐initiators. The HFE combinations and associated cutsets can be sorted in order of importance to focus the analysis and optimize resources. Basic dependency rules are applied to the HFE combinations to calculate conditional and joint human error probabilities (HEPs).


RiskSpectrum PSA delivers blue-chip fault tree and linked event tree modelling and analysis capabilities. It’s considered the nuclear industry standard in many countries, licensed for use at 60% of commercial nuclear power plant operators World-Wide.

Via its intuitive user interface, we can model everything from the basic fault tree with and and or gates, right through to advanced fault tree and event tree integration of sequences in linked event trees with boundary conditions and CCF events. It is designed with special features to cover internal, area (fire and flooding) and external events like seismic shocks.

The integrated RSAT analysis tool is specially designed for solving large PSA models and offers MCS (Minimal Cut Set), BDD, sensitivity, importance and time-dependent analysis.


Fire Dynamics Simulator (FDS) is a computational fluid dynamics (CFD) model of fire-driven fluid flow. The software solves numerically a form of the Navier-Stokes equations appropriate for low-speed, thermally-driven flow, with an emphasis on smoke and heat transport from fires.

Smokeview (SMV) is a visualization program that is used to display the output of FDS and CFAST simulations


ACS SASSI is a highly specialized soil-structure interaction (SSI) finite element analysis software. The software is a finite element computer code on the MS Windows PC platforms for performing efficiently linear or nonlinear 3D dynamic soil-structure interaction (SSI) analyses for complex geometry foundations subjected to spatially varying incoherent motions or multiple support seismic excitations.

PSCAD v 2.6 and Design Expert v 3.1

are used by the construction and design team of the project to develop strength calculations and sizing of building structures;

Autodesk Nastran

used to generate a computational model for pipelines, equipment, structures and perform the necessary static, modal, dynamic and temperature analyzes;

Autodesk Inventor Professional

used as a pre- and post-processor of Autodesk Nastran 2019, as well as for three-dimensional (3D) design, automatic three-dimensional pipeline generation, spatial (3D) layout and design of heating equipment.