Assessing the Radiological Risks Associated with High Natural Radioactivity of Microgranitic Rocks: A Case Study in a Northeastern Desert of Egypt

This study aimed to evaluate the radiological hazards of uranium (238U), thorium (232Th), and potassium (40K) in microgranitic rocks from the southeastern part of Wadi Baroud, a northeastern desert of Egypt. The activity concentrations of the measured radionuclides were determined by using a gamma-ray spectrometer (NaI-Tl-activated detector). The mean (238U), (232Th), and (40K) concentrations in the studied rocks were found to be 3680.3, 3635.2, and 822.76 Bq/kg, respectively. The contents in these rocks were elevated, reaching up to 6.3 wt%. This indicated the alkaline nature of these rocks. The high ratios of Th/U in the mineralized rocks could be related to late magmatic mineralization, suggesting the ascent of late magmatic fluids through weak planes such as faults and the contact of these rocks with older granites. The present data were higher than those of the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) guideline limits. All the radiological hazard results indicated high human health risks. This confirmed that this area is not radiologically safe, and care must be taken when working in this area. This study showed that the area under investigation had high U content suitable for uranium extraction that could be used in the nuclear fuel cycle.

A new tool for the simulation of different nuclear fleets at equilibrium

Scenario simulations are the main tool for studying the impact of a nuclear reactor fleet on the related fuel cycle facilities. This equilibrium preliminary study aims to present the functionalities of a new tool and to show the wide variety of reactors/cycles/strategies that can be studied in steady state conditions and validated with more details thanks to dynamic code. Different types of scenario simulation tools have been developed at CEA over the years, this study focuses on dynamic and equilibrium codes. Dynamic fuel cycle simulation code models the ingoing and outgoing material flow in all the facilities of a nuclear reactor fleet and their evolutions through the different nuclear processes over a given period of time. Equilibrium fuel cycle simulation code models advanced nuclear fuel cycles in equilibrium conditions, i.e. in conditions which stabilize selected nuclear inventories such as spent nuclear fuel constituents, plutonium or some minor actinides. The principle of this work is to analyze different nuclear reactors (PWR, AMR) and several fuel types (UOX, MOX, ERU, MIX) to simulate advanced nuclear fleet with partial and fully plutonium and uranium multi-recycling strategies at equilibrium. At this first stage, selected results are compared with COSI6 simulations in order to evaluate the precision of this new tool, showing a significant general agreement.

Structural dynamics of Schottky and Frenkel defects in ThO2: a density-functional theory study

Thorium dioxide (ThO2) is a promising alternative to mixed-oxide nuclear fuels due to its longer fuel cycle and resistance to proliferation. Understanding the thermal properties, in particular the thermal conductivity,...

A scenario study on the transition to a closed nuclear fuel cycle using the nuclear energy system modelling application package (NESAPP)

The paper presents the results of a case study on evaluating performance and sustainability metrics for Russian nuclear energy deployment scenarios with thermal and sodium-cooled fast reactors in a closed nuclear fuel cycle. Ten possible scenarios are considered which differ in the shares of thermal and sodium-cooled fast reactors, including options involving the use of mixed uranium-plutonium oxide fuel in thermal reactors. The evolution of the following performance and sustainability metrics is estimated for the period from 2020 to 2100 based on the considered assumptions: annual and cumulative uranium consumption, needs for uranium enrichment capacities, fuel fabrication and reprocessing capacities, spent fuel stocks, radioactive wastes, amounts of plutonium in the nuclear fuel cycle, amounts of accumulated depleted uranium, and the levelised electricity generation cost. The results show that the sustainability of the Russian nuclear energy system can be significantly enhanced through the intensive deployment of sodium-cooled fast reactors and the transition to a closed nuclear fuel cycle. The authors have highlighted some issues for further considerations, which will lead to more rigorous conclusions regarding the preferred options for the development of the national nuclear energy system.

System levelized fuel cost of electricity generation in a two-component nuclear energy system with a closed uranium-plutonium NFC

The authors propose an approach to the calculation of the levelized unit fuel cost (LUFC) of electricity generation for a fast reactor in a two-component nuclear energy system (NES) with regard for plutonium production. The approach is based on taking into account the additional economic effect, which can be achieved through the sale at the market price of the natural uranium released due to the substitution of thermal reactors by fast reactors with MOX fuel based on the plutonium bred in a fast reactor. This requires considering simultaneously the reactor parts of the fuel cycle for fast and thermal reactors. Relationships have been obtained which connect the key neutronic and fuel characteristics with the NPP and fuel cycle economic performance. The described methodology was used for the computational study of the LUFC for a fast sodium-cooled reactor. Calculations have shown that, in the considered case, taking into account the plutonium production leads to the LUFC reduction by nearly half and, therefore, to a major decrease in the total unit cost of electricity generation (levelized cost of electricity or LCOE).

Global outlook on large-scale nuclear power development strategies

As of today, nuclear power together with hydropower provides three-quarters of global low-carbon electricity generation. Over the past 60 years since the time of its inception, the use of nuclear power has reduced CO2 emissions by over 60 gigatonnes. There is no doubt that nuclear power can play a major, and maybe even a decisive role in decarbonizing the electricity sector, as it is evident from the current energy mix of some European countries, especially France, and major economic powers like the Unites States, Russia and South Korea. It is also evident that in most advanced economies nuclear power has entered a phase of gradual decline with little new investment coming into new projects, regardless of the world’s desperate need for more low-carbon electricity. Although existing reactor and their corresponding fuel cycle technologies have enabled the global nuclear power fleet to reach ~ 400 GWe of net installed capacity, there is growing concern that the scale of NPP shutdowns expected in Europe and North America could offset new capacity additions in Asian markets. Theoretically, renewable energy could fill the void left by reactors taken offline but there is strong evidence that the potential of wind and solar for global decarbonization is limited by material, land and economic constraints. Large-scale renewable systems would also require massive energy storage capacity that would hamper economic sustainability of the energy supply for developing countries. Taking into account the potential benefits of developing nuclear power, some countries are determined to expand its share in their energy mix through technological innovation and application of new strategies, directed at improving or completely resolving current issues related to economics, environmental concerns or non-proliferation of nuclear weapons. There are many states in the world today pursuing some sort of nuclear power development. A limited number of countries envision expanding or transforming their nuclear energy system using truly game-changing strategies based on innovative reactor, fuel cycle and waste management technologies. The focus of this paper is to give an overview of the approaches to large-scale nuclear power development being applied today in Russia, China, USA and India.

Optimization of the loading pattern of the PWR core using genetic algorithms and multi-purpose fitness function

The study demonstrates an application of genetic algorithms (GAs) in the optimization of the first core loading pattern. The Massachusetts Institute of Technology (MIT) BEAVRS pressurized water reactor (PWR) model was applied with PARCS nodal-diffusion core simulator coupled with GA numerical tool to perform pattern selection. In principle, GAs have been successfully used in many nuclear engineering problems such as core geometry optimization and fuel configuration. In many cases, however, these analyses focused on optimizing only a single parameter, such as the effective neutron multiplication factor (k eff), and often limited to the simplified core model. On the contrary, the GAs developed in this work are equipped with multiple-purpose fitness function (FF) and allow the optimization of more than one parameter at the same time, and these were applied to a realistic full-core problem. The main parameters of interest in this study were the total power peaking factor (PPF) and the length of the fuel cycle. The basic purpose of this study was to improve the economics by finding longer fuel cycle with more uniform power/flux distribution. Proper FFs were developed, tested, and implemented and their results were compared with the reference BEAVRS first fuel cycle. In the two analysed test scenarios, it was possible to extend the first fuel cycle while maintaining lower or similar PPF, in comparison with the BEAVRS core, but for the price of increased initial reactivity.

Perspective Compounds for Immobilization of Spent Electrolyte from Pyrochemical Processing of Spent Nuclear Fuel

Immobilization of spent electrolyte–radioactive waste (RW) generated during the pyrochemical processing of mixed nitride uranium–plutonium spent nuclear fuel is an acute task for further development of the closed nuclear fuel cycle with fast neutron reactors. The electrolyte is a mixture of chloride salts that cannot be immobilized directly in conventional cement or glass matrix. In this work, a low-temperature magnesium potassium phosphate (MPP) matrix and two types of high-temperature matrices (sodium aluminoironphosphate (NAFP) glass and ceramics based on bentonite clay) were synthesized. Two systems (Li0.4K0.28La0.08Cs0.016Sr0.016Ba0.016Cl and Li0.56K0.40Cs0.02Sr0.02Cl) were used as spent electrolyte imitators. The phase composition and structure of obtained materials were studied by XRD and SEM-EDS methods. The differential leaching rate of Cs from MPP compound and ceramic based on bentonite clay was about 10−5 g/(cm2·day), and the rate of Na from NAFP glass was about 10−6 g/(cm2·day). The rate of 239Pu from MPP compound (leaching at 25 °C) and NAFP glass (leaching at 90 °C) was about 10−6 and 10−7 g/(cm2·day), respectively. All the synthesized materials demonstrated high hydrolytic, mechanical compression strength (40–50 MPa) even after thermal (up to 450 °C) and irradiation (up to 109 Gy) tests. The characteristics of the studied matrices correspond to the current requirements to immobilized high-level RW, that allow us to suggest these materials for industrial processing of the spent electrolyte.