• Energy Research

Abstract The decommissioning of a nuclear power plant/site is a very complex process that is necessary to be performed at the end of its useful life to allow its release from regulatory control and a new use of the site by the society. Typically, a decommissioning project requires the management of several tasks and resources. In case of multiple reactors site, this management becomes harder since the interdependencies between each plant could influence the way that the tasks could be executed as well as its schedules. In literature is possible to find different kinds of tools for planning and management purposes, some of them specific for nuclear installations decommissioning projects. This is the case of tools based on the ISDC. However, these tools do not allow to consider rightly the interdependencies and previous learning. Notwithstanding, most part of existent tools/methods requires a large quantity of data/information to produce consistent results. However, this data often is unknown, requiring assumptions to justify some values, which enhances the uncertainties of the results. To overcome these difficulties, the present work will introduce a management tool and detail its mathematical model. This mathematical model aims to cost estimate the decommissioning cost for budget/bid purposes. The main differential of the model is that it requires a significantly less data compared with models available in literature. As part of the work objectives, a novel method for attribute values to the difficulty values and contingencies is presented. In addition, a validation simulation is presented as well as a sensibility analysis. The validation was performed considering a Brazilian nuclear power plant as case study, which had its decommissioning cost estimated in about US$761.7 million. The sensibility analysis indicates that is interesting to postpone the beginning of decommissioning some years after plant shutdown since it allows not only some radiological results but also a cost reduction.

Abstract This work presents the neutronics and thermal hydraulics feasibility to convert the UO 2 core of the Westinghouse AP1000 in a (U-Th)O 2 core by performing a parametric study varying the type of geometry of the pins in fuel elements, using the heterogeneous seed blanket concept and the homogeneous concept. In the parametric study, all geometry and materials for the burnable poison were kept the same as the AP 1000, and the only variable was the fuel pin material, in which we use several mass proportion of uranium and thorium but keeping the enrichment in 235 U, as LEU (20 w/o). The neutronics calculations were made by SERPENT code, and to validate the thermal limits we used a homemade code. The optimization criteria were to maximize the 233 U, and conversion factor, and minimize the plutonium production. The results obtained showed that the homogeneous concept with three different mass proportion zones, the first containing (32% UO 2 -68%ThO 2 ); the second with (24% UO 2 -76% ThO 2 ), and the third with (20% UO 2 -80% ThO 2 ), using 235 U LEU (20 w/o), and corresponding with the 3 enrichment zones of the AP 1000 (4.45 w/o; 3.40 w/o; 2.35 w/o), satisfies the optimization criteria as well as attending all thermal constrain. The concept showed advantages compared with the original UO 2 core, such a lower power density, and keeping the same 18 months of cycle a reduction of B-10 concentration at the soluble poison as well as eliminating in the integral boron poison coated (IFBA).

Abstract Most part of nuclear power sites around the world have multiple reactors, requiring different approaches to deal with plants interdependencies during its decommissioning. Usually, decommissioning strategies are proposed based on previous experiences. A strategy that would organize the decommissioning process is often proposed based on previous experience of similar projects. Different management tools for decommissioning cost estimation are found in literature but none of them focus on strategy proposal, which considers in a properly way plants interdependencies as well as mathematical models for costing the process. To overcome of these difficulties, the present work presents a method for decommissioning strategy proposal which could be used integrated with a mathematical model for cost estimation. Five strategies were obtained with the proposed method. The cost to decommissioning a case study site and its plants is also provided considering the Brazilian context.

An investigation on a flywheel is presented based on finite element modelling simulations for different geometries. The goal was to optimise the energy density (rotational energy-to-mass ratio) and, at the same time, the rotational energy of a flywheel rotor. The stress behaviour of flywheel rotors under the rotational speed at the maximum stress achievable by the flywheel was analysed. Under this condition, the energy density was obtained for the different geometries, as well as the rotational energy. The best energy density performance due to geometry was achieved with a flywheel rotor presenting a new Gaussian section, which is different from the known Laval disk shape. The best results using a single disk involved a rotational speed of nearly 279,000 rpm and a rotational energy density around 1584 kJ/kg (440 Wh/kg). These values still yielded low total energy; to increase its value, two or three rotors were added to the flywheel, which were analysed in regard to stability. In particular, the triple rotor energy density was ≈ 1550 kJ/kg (431 Wh/kg). As some instability was found in these rotors, a solution using reinforcement was developed to avoid such instabilities. The energy density of such a reinforced double rotor neared 1451 kJ/kg (403 Wh/kg), and the system achieved higher total energy. The material assumed for the devices was carbon fibre Hexcel UHM 12,000, a material kept constant throughout the simulations to allow comparison among the different geometries.

This article describes the electricity consumption in Brazilian residences between 1985 and 2013 through linear regressions. The explanatory variables considered were the number of households, effective consumption of families as a proxy for family income, and electricity tariff for households. To deal with the power generation crisis of 2001 we have introduced a dummy variable in the form of a step function. With such explanatory variables, we were able to account for the reduction of household electricity consumption caused by the policies conducted in 2001 and their permanent consequences. The regression presented coefficient of determination of 0.9892, and the several statistic tests conducted assured the existence of long-term relation between the electricity consumption in residences and the explanatory variables. The obtained elasticities for the household consumption of electricity with respect to number of residences, family income and residential tariff of electricity were 1.534±0.095, 0.189±0.049, and −0.230±0.060, respectively. These results allowed understanding the evolution over time of the household consumption of electricity in Brazil. They suggest that the electric sector in Brazil should pursue an active policy to manage demand of residential electricity using tariffs as a means to control it.

Control rod worth measurements through the inverse kinetics equation depend on accurate determination of the amplitude function from detector signals. The modal-local method introduced in a previou...

Abstract The construction time of PWRs is studied considering published data about nuclear power plants in the world. For the 268 PWRs in operation in 2010, the mode of the construction time distribution is around 5–6 years, and 80% of the plants were built in less than 120 months. To circumvent the problem of comparing plants with different size we normalized the construction time to plants with 1 GW. We restricted the analysis to 201 PWRs which suffered less from external factors that were beyond the control of the management from 1965 to 2010. The results showed that the normalized construction time did not increase over the years and nor with the plants’ gross power level. The learning rate of the industry regarding normalized construction times showed a reduction with 95% confidence level of about 0.56±0.07 months for each 10 GW of installed capacity. Over the years the normalized construction time decreased and became more predictable. The data showed that countries with more centralized regulatory, construction and operation environments were able to build PWRs in shorter times. Countries less experienced with the nuclear technology built PWRs in longer times.

This article presents the validation of the Code for Thermal-hydraulic Evaluation of Nuclear Reactors with Plate Type Fuels (COTENP), a subchannel code which performs steady-state thermal-hydraulic analysis of nuclear reactors with plate type fuel assemblies operating with the coolant at low pressure levels. The code is suitable for design analysis of research, test, and multipurpose reactors. To solve the conservation equations for mass, momentum, and energy, we adopt the subchannel and control volume methods based on fuel assembly geometric data and thermal-hydraulic conditions. We consider the chain or cascade method in two steps to facilitate the analysis of whole core. In the first step, we divide the core into channels with dimensions equivalent to that of the fuel assembly and identify the assembly with largest enthalpy rise as the hot assembly. In the second step, we divide the hot fuel assembly into subchannels with size equivalent to one actual coolant channel and similarly identify the hot subchannel. The code utilizes the homogenous equilibrium model for two-phase flow treatment and the balanced drop pressure approach for the flow rate determination. The code results include detailed information such as core pressure drop, mass flow rate distribution, coolant, cladding and centerline fuel temperatures, coolant quality, local heat flux, and results regarding onset of nucleate boiling and departure of nucleate boiling. To validate the COTENP code, we considered experimental data from the Brazilian IEA-R1 research reactor and calculated data from the Chinese CARR multipurpose reactor. The mean relative discrepancies for the coolant distribution were below 5%, for the coolant velocity were 1.5%, and for the pressure drop were below 10.7%. The latter discrepancy can be partially justified due to lack of information to adequately model the IEA-R1 experiment and CARR reactor. The results show that the COTENP code is sufficiently accurate to perform steady-state thermal-hydraulic design analyses for reactors with plate type fuel assemblies.

The nuclear industry requires specific safety qualification of electrical equipment in order to assure that they perform their function without experiencing common-cause failures. Electrical equipment used in the nuclear industry have been qualified according to IEEE 323 and IEC 60780 standards and their derivatives until 2016. Currently these two organizations consolidated their requirements and unified them in a common document, the IEC/IEEE 60780-323 standard. It defines the different methods of qualification, a qualification program plan and the documentation to be delivered demonstrating successful qualification of electrical equipment to be used in nuclear installations. Based on this standard, we present an approach focusing on the qualification by type-testing. The article details the necessary tests to be performed for this method, the equipment necessary to be used in the type-testing method and the documentation expected to be delivered by the supplier. To implement this qualification approach and obtain such capacity, it is necessary to perform large expenditures in equipment and in the improvement of the workforce. On the other hand, this capacity gives the companies knowledge, experience and opens new market opportunities in other areas, such as the naval and oil industries. With such expertise, Brazilian companies may become active suppliers in the world industry of electrical equipment.