C. A. Silva

Safety Studies and General Simulations of Research Reactors Using Nuclear Codes

Antonella L. Costa1, Patricia A. L. Reis1, Clarysson A. M. Silva1, Claubia Pereira1, Maria Auxiliadora F. Veloso1, Bruno T. Guerra1, Humberto V. Soares1 and Amir Z. Mesquita2 1Departamento de Engenharia Nuclear – Escola de Engenharia Universidade Federal de Minas Gerais Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq 2Centro de Desenvolvimento da Tecnologia Nuclear/Comissao Nacional de Energia Nuclear – CDTN/CNEN Brasil

Neutronic Evaluation of a MHR System to Transmutation of Minor Actinides

The goal is to simulate the modular helium reactor (MHR) core to analyze the neutronic parameters behavior due the insertion of Pu isotopes and minor actinides (MAs) using shuffling scheme without compromising the safety parameters. Initially the core is filled with driver fuel (DF). After the burn-up, these fuels are then reprocessed and used to produce the transmutation fuel (TF). Some cycles after, the core is filled with DF and TF fuels. DF fuel is composed of Pu and Np while TF fuel is a mixture of Pu and MAs. The shuffling scheme was evaluated after each cycle. It was verified that neutronic parameters and isotopic composition reach equilibrium when this scheme is used. The WIMS code was used to perform the simulations and the following neutronic parameters were evaluated: infinite multiplication factor, spectrum hardening, and reactivity temperature coefficients.

A neutronic evaluation of VHTR and LS-VHTR

VHTR (Very High Temperature Reactor) and LS-VHTR (Liquid Salted - Very High Temperature Reactor) reactors were analyzed and compared from the viewpoint of the neutronic parameters. The multiplication coefficient (kinf), the fuel and the moderator temperature coefficients (αTF and αTM, respectively) and the hardening spectrum (øF/øT, fast flux/total flux) parameters were analyzed focusing to understanding better the VHTR and LS-VHTR neutronic behaviors. The experience obtained performing this work will be used in the future to study and to simulate the possibility of such systems for fuel reprocessing and recycling always analyzing how the process can affect the hydrogen production efficiency. The preliminary analyses have demonstrated that the VHTR reactor has a safer behavior in comparison with to the LS-VHTR one considering the evolution of the safety parameters along the burnup using WIMSD simulations.

A Neutronic Evaluation of the HTR-10 Using Scale, MCNPX and MCNP5 Nuclear Codes

The HTR-10 (High Temperature Gas-cooled Test Reactor) is a 10 MW modular pebble bed type reactor, built by the Institute of Nuclear Energy Technology (INET), Tsinghua University, China. As an advanced reactor, it has good passive safety characteristics: capacity of retaining all fission products inside the coated particles (up to 1,600° C), passive decay heat removal, large heat capacity of the core to mitigate temperature transition, large fuel temperature margin and negative temperature reactivity coefficient sufficient to accommodate reactivity insertion and small amount of excess reactivity in the core. This reactor, which core is filled with 27,000 spherical fuel elements, e.g. TRISO coated particles, is used to test and develop fuel, verify PBR safety features, demonstrate combined electricity production and cogeneration of heat, and provide experience in PBR design, operation and construction. Using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation), the MCNPX 2.6.0 (Monte Carlo N-Particle eXtended) and the MCNP 5 (Monte Carlo N-Particle) nuclear codes, the HTR-10 first critical core described in the Evaluation of The Initial Critical Configuration of The HTR-10 Pebble-Bed Reactor was modeled and analyzed. A three-dimension model was simulated and the keff was obtained and compared with the reference. The result presents good agreement with experimental value. The goal is to validate the DEN/UFMG model to be applied in transmutation studies changing the fuel.Copyright

Sensitivity Analysis of the TRIGA IPR-R1 Reactor Models Using the MCNP Code

In the process of verification and validation of code modelling, the sensitivity analysis including systematic variations in code input variables must be used to help identifying the relevant parameters necessary for a determined type of analysis. The aim of this work is to identify how much the code results are affected by two different types of the TRIGA IPR-R1 reactor modelling processes performed using the MCNP (Monte Carlo N-Particle Transport) code. The sensitivity analyses included small differences of the core and the rods dimensions and different levels of model detailing. Four models were simulated and neutronic parameters such as effective multiplication factor (), reactivity (), and thermal and total neutron flux in central thimble in some different conditions of the reactor operation were analysed. The simulated models presented good agreement between them, as well as in comparison with available experimental data. In this way, the sensitivity analyses demonstrated that simulations of the TRIGA IPR-R1 reactor can be performed using any one of the four investigated MCNP models to obtain the referenced neutronic parameters.

Research Reactor Analysis Using Thermal Hydraulic and Neutron Kinetic Coupling

Simulations of complex scenarios in nuclear power plants have been improved by the utilization of coupled thermal hydraulic (TH) and neutron kinetics (NK) system codes with the development of computer technology and new calculation methodology which made it possible to perform transport calculation schemes with accurate solutions. This paper presents a model for the IPR-R1 TRIGA research reactor using the RELAP5-3D 3.0.0 code. By using this code, a multi-dimensional neutron kinetics model based on the NESTLE code can be implemented also. In this way, during a 3D TH/NK coupled simulation, RELAP5-3D calls the appropriate NESTLE subroutines to perform the calculations. The development and the assessment of the thermal hydraulic RELAP5 code model for the IPR-R1 TRIGA have been validated for steady state and transient situations and the results were published in preceding works. The model has been adapted to RELAP5-3D code and was verified to point kinetic calculations. After this, adequate cross sections to the NK code were supplied using the WIMSD5 code. The results of steady state and transient calculations using the 3D neutron modeling to the IPR-R1 are being presented in this paper.Copyright

Thermal Modeling of the HTR-10 Using the RELAP5-3D Code

Several efforts have been considered in the development of the modular High Temperature Gas cooled Reactor (HTGR) planned to be a safe and efficient nuclear energy source for the production of electricity and industrial applications. In this work, the RELAP5-3D thermal hydraulic code was used to simulate the steady state behavior of the 10 MW pebble bed high temperature gas cooled reactor (HTR-10), designed, constructed and operated by the Institute of Nuclear and New Energy Technology (INET), in China. The reactor core is cooled by helium gas. In the simulation, results of temperature distribution within the pebble bed, inlet and outlet coolant temperatures, coolant mass flow, and others parameters have been compared with the data available in a benchmark document published by the International Atomic Energy Agency (IAEA) in 2013. This initial study demonstrates that the RELAP5-3D model is capable to reproduce the thermal behavior of the HTR-10.Copyright