Use In Light Water Reactors Research Articles

Cross-jurisdictional analysis and forecasting of North American nuclear fuel inventory using a standardized unit

This study fills the noted gap in comparative analyses of spent nuclear fuel (SNF) by assessing inventories from two key nuclear power regions, the USA and Canada, using a comprehensive analytical framework and standardized data from 2009 to 2021. In the USA, SNF inventory increased by 14.7 % in fuel assembly weight and 47 % in residual heavy metal content compared to Canada, in line with their use of light water reactors. Canada's SNF production is directly correlated to its nuclear power output, influenced by the lower burnup of natural uranium fuel used in CANDU reactors (R2 = 0.57; p-value < 0.05) while the USA shows insignificant correlation, likely due to a variety of reactor types and higher burnup rates (R2 = 0.008; p-value > 0.05). Further, the study identifies a strong negative correlation between uranium mine production and SNF inventory in the USA, indicating a reliance on imports amidst negligible domestic mining. In contrast, Canada also exhibits moderate negative dependency due to its position as a major uranium exporting jurisdiction. The obtained negative correlations with coal rents in both countries indicate a shift towards more nuclear energy use, impacting economic growth and energy patterns. The developed predictive models indicate a higher future SNF increase in Canada than in the USA. These findings are essential for planning the transition from temporary to permanent SNF disposal, ensuring safe long term management of radioactive waste.

Performance evaluation of a currently in-use dry storage cask design for spent accident tolerant fuel loading case under normal operation condition

Abstract The aim of this study is to assess the performance of an existing dry storage cask design for accident tolerant fuel loading case and to examine the compliance with the safety limits applied currently for dry storage. In the study, for a dry storage cask design currently in use, criticality calculations, dose rate evaluation and thermal analyses are performed in case of loading with the accident tolerant fuel discharged from a PWR. Firstly, for an accident tolerant fuel selected among the concepts proposed for use in light water reactors, burnup analyses are performed by utilizing the Serpent continuous energy code and spent fuel characteristics are determined. Then, criticality analyzes are carried out by using the Serpent Monte Carlo code for the case of loading the accident tolerant fuel into the selected dry storage cask design. Gamma and neutron dose rates at the outer surface and close distances of the storage cask are determined with the Serpent code. To evaluate the thermal performance of the storage cask, thermal analyzes are performed by using the ANSYS Fluent computational fluid dynamics code. The analysis results are compared with the nuclear safety criteria applied to dry storage casks. Results of the analysis show that the dry storage cask design currently in-use does not exceed the criticality, dose rate and maximum surface temperature limits when loaded with spent accident tolerant fuel.

Impact of MOX fuel use in light-water reactors: long-term radiological consequences of disposal of high-level waste in a geological repository

ABSTRACT As a series of studies to evaluate impact of mixed-oxide (MOX) fuel in light-water reactors (LWRs), post-closure long-term safety for various vitrified high-level radioactive waste (HLW) arising from the different fuel cycle intends to recycle Pu are examined. In this study, four fuel cycle scenarios with different ratio of spent MOX generated and two reprocessing options for each fuel cycle scenario are considered. One reprocessing option considers disposal of vitrified HLW generated separately from the reprocessing of spent UO2 fuel and MOX fuel (separated HLW), and the other is blended vitrified UO2-MOX HLW (blended HLW) generated during reprocessing whereby MOX spent fuel is diluted by UO2 spent fuel. First, the radionuclide inventories of those vitrified HLWs are discussed. Next, radionuclide migration analyses for geological disposal of those vitrified HLWs are evaluated. It has revealed that the disposal of blended HLW will not have an adverse effect on the long-term radiological impact compared to separated HLW. Results of this study can be used as a basis for considering the blending option as a viable alternative approach in the future for managing MOX fuel used in light-water reactors.

Oxidation of UN/U2N3-UO2 composites: an evaluation of UO2 as an oxidation barrier for the nitride phases

Composite fuels such as UN-UO2 are being considered to address the lower oxidation resistance of the UN fuel from a safety perspective for use in light water reactors, whilst improving the in-reactor behaviour of the more ubiquitous UO2 fuel. An innovative UN-UO2 accident tolerant fuel has recently been fabricated and studied: UN microspheres embedded in UO2 matrix. In the present study, detailed oxidative thermogravimetric investigations (TGA/DSC) of high-density UN/U2N3-UO2 composite fuels (91-97 %TD), as well as post oxidised microstructures obtained by SEM, are reported and analysed. Triplicate TGA measurements of each specimen were carried out at 5 K/min up to 973 K in a synthetic air atmosphere to assess their oxidation kinetics. The mass variation due to the oxidation reactions (%), the oxidation onset temperatures (OOTs), and the maximum reaction temperatures (MRTs) are also presented and discussed. The results show that all composites have similar post oxidised microstructures with mostly intergranular cracking and spalling. The oxidation resistance of the pellet with initially 10 wt% of UN microspheres is surprisingly better than the UO2 reference. Moreover, there is no significant difference in the OOT (~557 K) and MRT (~615 K) when 30 wt% or 50 wt% of embedded UN microspheres are used. Therefore, the findings in this article demonstrate that the UO2 matrix acts as a barrier to improve the oxidation resistance of the nitride phases at the beginning of life conditions.

Hydrothermal corrosion of 2nd generation FeCrAl alloys for accident tolerant fuel cladding

As part of an effort to develop advanced fuel cladding for use in light water reactors, several FeCrAl alloys have been developed at Oak Ridge National Laboratory. A second generation of these alloys bearing additional alloying elements over the previously tested model alloys were tested in boiling water reactor (BWR) conditions to determine their resistance to hydrothermal corrosion. Coupons with alloy compositions Fe–10Cr–6Al–2Mo, Fe–13Cr–5Al–2Mo, Fe–13Cr–6Al–2Mo, Fe–13Cr–7Al–2Mo, and Fe–13Cr–5Al–2Mo–1Nb were tested in normal water chemistry (NWC) and hydrogen water chemistry (HWC) for 9 months in continuously refreshing autoclaves. Commercial FeCrAl alloy APMT and Zircaloy-2 were also tested for comparison. Among samples exposed to HWC, Fe–13Cr–7Al–2Mo performed worst with an average mass loss of 1.3 mg/cm2 over 9 months. This mass loss represents an estimated thickness loss of approximately 60 μm over 6 years. Samples exposed to NWC had very small mass losses of less than 0.15 mg/cm2 or mass gains up to 0.05 mg/cm2. Based on the results of this testing, the 2nd generation FeCrAl alloys tested exhibit low wall thickness loss and are suitable in terms of corrosion resistance for use as LWR cladding in BWR-HWC and BWR-NWC under normal operating conditions.

Thermal–mechanical performance modeling of thorium–plutonium oxide fuel and comparison with on-line irradiation data

Thorium-plutonium Mixed OXide (Th-MOX) fuel is considered for use in light water reactors fuel due to some inherent benefits over conventional fuel types in terms of neutronic properties. The good material properties of ThO2 also suggest benefits in terms of thermal–mechanical fuel performance, but the use of Th-MOX fuel for commercial power production demands that its thermal–mechanical behavior can be accurately predicted using a well validated fuel performance code. Given the scant operational experience with Th-MOX fuel, no such code is available today.This article describes the first phase of the development of such a code, based on the well-established code FRAPCON 3.4, and in particular the correlations reviewed and chosen for the fuel material properties. The results of fuel temperature calculations with the code in its current state of development are shown and compared with data from a Th-MOX test irradiation campaign which is underway in the Halden research reactor. The results are good for fresh fuel, whereas experimental complications make it difficult to judge the adequacy of the code for simulations of irradiated fuel.

High temperature oxidation of molybdenum in water vapor environments

Molybdenum has recently gained attention as a candidate cladding material for use in light water reactors. Its excellent high temperature mechanical properties and stability under irradiation suggest that it could offer benefits to performance under a wide range of reactor conditions, but little is known about its oxidation behavior in water vapor containing atmospheres. The current study was undertaken to elucidate the oxidation behavior of molybdenum in water vapor environments to 1200°C in order to provide an initial assessment of its feasibility as a light water reactor cladding. Initial observations indicate that at temperatures below 1000°C, the kinetics of mass loss in water vapor would not be detrimental to cladding integrity during an off-normal event. Above 1000°C, degradation is more rapid but remains slower than observed for optimized zirconium cladding alloys. The effect of hydrogen–water vapor and oxygen–water vapor mixtures on material loss was also explored at elevated temperatures. Parts-per-million levels of either hydrogen or oxygen will minimally impact performance, but hydrogen contents in excess of 1000ppm were observed to limit volatilization at 1000°C.

Innovation in Nuclear Power

Innovation in nuclear energy centres principally on four areas: Isotopic separation to enrich uranium for use in light-water reactors, reactor designs of diverse kinds, robust fuel designs, and the recycling of most of what is in used fuel. Allied to these are metallurgical and design improvements to basic concepts that have not changed fundamentally in decades. An analogy here is the contrast between the car of the 1960s and the car of today – the same idea, but very different in every detailed respect. In uranium enrichment we have come from very energy-intensive gaseous diffusion technology to evolving centrifuges and now seem to be on the threshold of commercial laser isotope separation. Reactor designs include those for fast neutron operation, high-temperature helium-cooled systems and others with the fuel carried in molten fluoride salt. There is more. Fuel design has progressively improved and provides for more efficient reactor operation with longer periods between refueling. Beyond this, electrometallurgical processes for recycling most of what is in used fuel promise significant advantages over the 1940s processes that remain predominant.

Thermal and mechanical properties of (U,Er)O 2

Erbium is considered as a slow burnable poison suitable for use in light water reactors (LWRs). Addition of a small amount of Er 2O 3 to all UO 2 pellets will make it possible to develop super high burnup fuels in Japanese nuclear facilities which are now under the restriction of the upper limit of 235U enrichment. When utilizing the (U,Er)O 2 fuels, it is very important to understand the thermal and mechanical properties. Here we show the characterization results of (U 1− x Er x )O 2 (0 ⩽ x ⩽ 0.1). We measured their thermal and mechanical properties and investigated the effect of Er addition on these properties of (U,Er)O 2. All Er completely dissolved in UO 2, and the lattice parameter decreased linearly with the Er content. Both the thermal conductivity and Young’s modulus of (U,Er)O 2 decreased with the Er content. These results would be useful for us in evaluating the performance of the (U,Er)O 2 fuels in LWRs.

Neutronic and Logistic Proposal for Transmutation of Plutonium from Spent Nuclear Fuel as Mixed-Oxide Fuel in Existing Light Water Reactors

The use of light water reactors (LWRs) for the destruction of plutonium and other actinides [especially those in spent nuclear fuel (SNF)] is being examined worldwide. One possibility for transmutation of this material is the use of mixed-oxide (MOX) fuel, which is a combination of uranium and plutonium oxides. MOX fuel is used in nuclear reactors worldwide, so a large experience base for its use already exists. However, to limit implementation of SNF transmutation to only a fraction of the LWRs in the United States with a reasonable number of license extensions, full cores of MOX fuel probably are required. This paper addresses the logistics associated with using LWRs for this mission and the design issues required for full cores of MOX fuel. Given limited design modifications, this paper shows that neutronic safety conditions can be met for full cores of MOX fuel with up to 8.3 wt% of plutonium.

Modeling of Irradiation Embrittlement of Reactor Pressure Vessel Steels

A model to describe the change in the inelastic and fracture properties of reactor pressure vessel steels due to neutron irradiation in the ductile region (i.e., irradiation embrittlement) is developed. First, constitutive equations for unirradiated elastic-viscoplastic-damaged materials are developed within the framework of the irreversible thermodynamics theory. To take into account the effect of hydrostatic pressure on the nucleation and growth of microvoids, properly defined dissipation potential is used. Then, the effect of irradiation on the material behavior is incorporated into the proposed model as a function of neutron fluence Φ by taking into account the interaction between irradiation-induced defects and movable dislocations. As regards the damage strain threshold pD, the mechanism of void nucleation due to pile-up of dislocations at the inclusions in the material is proposed first under unirradiated-condition, and then the effect of irradiation on the mechanism is formulated. In order to demonstrate the validity of this model, it is applied to the case of uniaxial tensile loading of a low alloy steel A533B cl. 1 for the pressure vessel use of light-water reactors at 260°C. The resulting model can describe the increase in yield stress and ultimate tensile strength, the decrease in total elongation and strain hardening, and the strain rate dependence of yield stress due to neutron irradiation. [S0094-4289(00)00901-4]

原子炉圧力容器鋼の照射ぜい化機構とモデル化

A model to describe the change in the properties of the inelastic deformation and the fracture of reactor pressure vessel steels due to neutron irradiation, i.e., irradiation embrittlement, in the ductile region is developed. First, constitutive equations for unirradiated elastic-viscoplastic-damaged materials are developed within the irreversible thermodynamics theory. To take into account the effect of hydrostatic pressure on the growth of microvoids, suitable dissipation potential is used. Then, the effect of irradiation on the material behavior is incorporated into the proposed model as function of neutron fluence Φ taking into account the interactin of irradiation induced defects and movable dislocations. Especially for the effect on damage strain threshold pD, the mechanism of void nucleation due to stress concentration responsible for pile-up of dislocations at the inclusions in the material is proposed under unirradiated condition, then the effect of irradiation on that mechanism is considered. To demonstrate the validity of this model, it is applied to the case of uniaxial tensile loading of a low alloy steel A533B cl. 1 for the pressure vessel use of light-water reactors at 260°C. The resulting model can describe the increase of yield stress, ultimate tensile strength and the decrease of total elongation, strain hardening and strain rate dependence of yield stress due to neutron irradiation.

Physics Methods for Calculating Light Water Reactor Increased Performances

The intensive use of light water reactors (LWRs) has induced modification of their characteristics and performances in order to improve fissile material utilization and to increase their availability and flexibility under operation. From the conceptual point of view, adequate methods must be used to calculate core characteristics, taking into account present design requirements, e.g., use of burnable poison, plutonium recycling, etc. From the operational point of view, nuclear plants that have been producing a large percentage of electricity in some countries must adapt their planning to the need of the electrical network and operate on a load-follow basis. Consequently, plant behavior must be predicted and accurately followed in order to improve the plant's capability within safety limits. The Belgonucleaire code system has been developed and extensively validated. It is an accurate, flexible, easily usable, fast-running tool for solving the problems related to LWR technology development. The methods and validation of the two computer codes LWR-WIMS and MICROLUX, which are the main components of the physics calculation system, are explained.

Relative Worth of233U and235U in Light Water Reactor Fuel Cycles

AbstractThe equivalence between 235U and 233U for use in light water reactors has been determined by comparing the performance of the two isotopes over the complete length of the fuel cycle. The method developed in this work has been applied to a uranium cycle as well as to a denatured thorium cycle. A basic price based on the initial substitution method has also been estimated for 233U.