Fractional Scaling Analysis for IRIS pressurizer reduced scale experiments

About twenty organizations joined in a consortium led by Westinghouse to develop an integral, modular and medium size pressurized water reactor (PWR), known as international reactor innovative and secure (IRIS), which is characterized by having most of its components inside the pressure vessel, eliminating or minimizing the probability of severe accidents.     

Determination of a test section parameters for IRIS nuclear reactor pressurizer

An integral, modular and medium size nuclear reactor, known as IRIS, is being developed by Westinghouse and by research centers. IRIS is characterized by having most of its components inside the pressure vessel, eliminating the probability of accidents. Due to its integral configuration, there is no spray system for boron homogenization, which may cause power transients. Thus, boron mixing must be investigated. The aim of this paper is to establish the conditions under which a test section has to be built for boron dispersion analysis inside IRIS reactor pressurizer. Through Fractional Scaling Analysis, which is a new methodology of similarity, the main parameters for a test section are obtained. By combining Fractional Scaling Analysis with local scaling for the densimetric Froude number and a previously established volumetric scale factor, the values of recirculation orifices, inlet water temperature, time scale factor and recirculation flow for the test section (model) are determined so that boron distribution is well represented in IRIS reactor pressurizer (prototype). Analytical solutions were used to validate the adopted methodology and when the results simulated in the model are compared to those that characterize the prototype, the agreement for both systems is absolute. The thermal power also influences boron distribution inside the test section. This power is determined by condensation laws in the vapor region and by suitable correlations for free convection. The fractions for rising inlet recirculation water enthalpy and vapor formation are also considered.     

The artificial neural network used in the study of sensitivities in the IRIS reactor pressurizer

The technique of sensibility analysis studies the behavior of the ratio between the variation of output results and the variation of input parameters in general. This study performed in the reactor pressurizer, which is a component responsible for controlling of the pressure inside the vessel, has the fundamental importance in designing the security of any concept of an advanced reactor. In fact, for its feature of passive action of the pressurizer (there is no spray), this analysis becomes a necessary step for safety and performance of the plant. The direct method through code MODPRESS, which represents the pressurizer model of the International Reactor Innovative and Secure (IRIS), has required a huge computational effort. To solve this problem, artificial neural networks (ANNs), beyond faster, has been used to replace the MODPRESS in this article. The ANNs do not require linear behavior of the system and can use both, simulated or experimental data for their training and learning. In order this, we adopted a classical non-supervised training ANN for mapping and forecasting of the pressurized using initially simulated data. In next future, we will incorporate the experimental data from the operation of the CRCN-NE reduced-scale test facility mapping. Moreover, based on the results obtained in this study, one can conclude that the artificial neural networks are presented as an alternative to MODPRESS code, and artificial neural networks are actually a great tool to calculate the sensitivity coefficient.     

INPRO economic assessment of the IRIS nuclear reactor for deployment in Brazil

This paper presents the results of the economic assessment of the International Reactor Innovative and Secure (IRIS) for deployment in Brazil using the assessment methodology developed under the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), co-ordinated by the International Atomic Energy Agency (IAEA). INPRO initiated in 2001 and has the main objective of helping to ensure that nuclear energy will be available to contribute in a sustainable manner to the energy needs of the 21st century. Among its missions is the development of a methodology to assess innovative nuclear energy systems (INSs) on a global, regional and national basis. In 2005, Brazil submitted a proposal for the assessment of two small-size reactors as components of an INS, completed with a conventional open nuclear fuel cycle based on enriched uranium. One of the reactors assessed was IRIS, a small-size, modular, integral-type PWR reactor. IRIS was evaluated with regard to the areas of reactor safety and economics only. This paper outlines the rationale for the study and summarizes the results of the economic assessment. The study concluded that the reference design of IRIS complies with most of INPRO economics criteria and has potential to comply with the remaining ones.     

Generalized linear model for identifying the kinetic parameters of a nuclear reactor

Methods of real-time identification of the kinetic parameters of a reactor on the basis of an analysis of the linear couplings in transient neutron processes are examined. The algorithms described can be used in various combinations during reactor operation to identify typical transient processed according to their traces.     

New concept for the vessel of a nuclear power reactor with supercritical coolant parameters

A multiwall design, akin to a Russian nested doll, of a nuclear reactor vessel is described. This design is intended for reactors with supercritical coolant parameters. The interwall gaps of a multiwall vessel are hydraulically connected with the reactor coolant via a separative device. A system of pressure regulators distributes the coolant pressure successively over the gaps in a manner so that in the operating regime a wall would be under the optimal nominal stress. To eliminate any danger of the vessel collapsing in the event that the counter pressure system malfunctions, ribbed rings with axial channels are tightly installed in the gaps between the walls. This makes it possible to freely fill the gaps in the vessel with a working body, for example, water. In the event that the counterpressure system fails the multiwall vessel is automatically converted into a multilayer wall, which eliminates the possibility of the vessel being destroyed. To increase the reliability of such a vessel, one or two additional walls, which can carry a load, for example, at the end of operation when the inner walls, which are subjected to the highest neutron irradiation, partially lose their strength, are installed in the vessel. Translated from Atomnaya énergiya, Vol. 105, No. 6, pp. 322–325, December, 2008.     

Structural homogenized analysis for a nuclear reactor core

A nuclear reactor core is composed of a great number of tubular beams with periodic structure, which are immersed in an acoustic fluid. In the present paper, a 3-D homogenization model is developed to predict its overall dynamic behavior. An approximate solution to the local problem is given. The application to an 1-D example shows that approximate expressions of the natural frequency, the added fluid mass and the equivalent sound speed can be used in engineering estimation.     

Kinetic parameters of a low enriched uranium fuelled material test research reactor at end-of-life

The kinetic parameters at end-of-life of a material test reactor fuelled with low enriched uranium fuel were calculated. The reactor used for the study was the IAEA’s 10 MW benchmark reactor. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time and effective delayed neutron fraction. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that in comparison with the beginning-of-life values, at end-of-life, the neutron flux increased throughout the core, the prompt neutron generation time increased by 3.68% while the effective delayed neutron fraction decreased by 0.35%.     

Computational analysis of neutronic parameters for TRIGA Mark-II research reactor using evaluated nuclear data libraries

The aim of this study is to analyze the neutronic parameters of TRIGA Mark-II research reactor using the chain of NJOY-WIMS-CITATION computer codes based on evaluated nuclear data libraries CENDL-2.2 and JEFF-3.1.1. The nuclear data processing code NJOY99.0 has been employed to generate the 69 group WIMS library for the isotopes of TRIGA core. The cell code WIMSD-5B was used to generate the cross sections in CITATION format and then 3-dimensional diffusion code CITTATION was used to calculate the neutronic parameters of the TRIGA Mark-II research reactor. All the analyses were performed using the 7-group macroscopic cross section library. The CITATION test-runs using different cross section sets based on different models applied in WIMS calculations have shown a strong influence of those models on the final integral parameters. Some of the cells were specially treated with PRIZE options available in WIMSD-5B to take into account the fine structure of the flux gradient in the fuel-reflector interface region. It was observed that two basic parameters, the effective multiplication factor, k_eff and the thermal neutron flux, were in good agreement among the calculated results with each other as well as the measured values. The maximum power densities at the hot spot were 1.0446E02 W/cc and 1.0426E02 W/cc for the libraries CENDL-2.2 and JEFF-3.1.1 respectively. The calculated total peaking factors 5.793 and 5.745 were compared to the original SAR value of 5.6325 as well as MCNP result. Consequently, this analysis will be helpful to enhance the neutronic calculations and also be used for the further thermal–hydraulics study of the TRIGA core.     

Choice of the capacity of a fast reactor for nuclear power

Analytical assessments, associated with the choice of the unit capacity of a serially built fast reactor under conditions of the future advancement of nuclear power, are presented. It is shown that considering the limited resources of natural uranium, the development of a reliable raw materials base must be based on the development of fast reactors with expanded breeding of fuel and fuel cycle closure. Since fast reactors, together with energy production, are also producers of new fuel, their parameters must be optimized taking account of this factor on the basis of systems analysis. Calculations show that the optimal capacity for fast reactors is in the 1 GW range. __________ Translated from Atomnaya énergiya, Vol. 103, No. 2, pp. 83–88, August, 2007.     

Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl_x-Al) containing 4.40 gU/cm³ of an MTR was replaced with silicide (U₃Si-Al and U₃Si₂-Al) and oxide (U₃O₈-Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U₃Si₂-Al followed by 0.03% for U₃Si-Al, and 0.01% for U₃O₈-Al fuel. The U₃O₈-Al fueled reactor gave the maximum ρ_excess at BOL which was 21.67% more than the original fuel followed by U₃Si-Al which was 2.55% more, while that of U₃Si₂-Al was 2.50% more than the original UAl_x-Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U₃O₈-Al followed by U₃Si-Al and then U₃Si₂-Al fuel.     

A scheme for the evaluation of dominant time-eigenvalues of a nuclear reactor

This paper presents a scheme to obtain the fundamental and few dominant solutions of the prompt time eigenvalue problem (referred to as α-eigenvalue problem) for a nuclear reactor using multi-group neutron diffusion theory. The scheme is based on the use of an algorithm called Orthomin(1). This algorithm was originally proposed by Suetomi and Sekimoto [Suetomi, E., Sekimoto, H., 1991. Conjugate gradient like methods and their application to eigenvalue problems for neutron diffusion equations. Ann. Nucl. Energy 18 (4), 205–227] to obtain the fundamental K-eigenvalue (K-effective) of nuclear reactors. Recently, it has been shown that the algorithm can be used to obtain the further dominant K-modes also. Since α-eigenvalue problem is usually more difficult to solve than the K-eigenvalue problem, an attempt has been made here to use Orthomin(1) for its solution. Numerical results are given for realistic 3-D test case.     

Design of an intelligent fuzzy logic controller for a nuclear research reactor

The main objective of this paper is to design an intelligent controller system based on the concepts of fuzzy logic. This latter will be used to control the power of a nuclear reactor. The principle of this controller is based on rules established from experiments used with a classical controller and from the knowledge and the expertise of the operators of the reactor. This intelligent controller could be used in parallel with the actual system, which is semiautomatic, as a decision aided system to assist the operators in the control room.