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"Spent nuclear fuels"
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A new approach to the nuclear fuel cycle : best practices for security, nonproliferation, and sustainable nuclear energy
\"In the past decade, a resurgence of enthusiasm for nuclear power has rekindled interest in efforts to manage the fuel cycle. The 2011 accident at the Fukushima Daiichi nuclear power plants in Japan and current proliferation crises in North Korea and Iran raise this question: Is the current approach on the fuel cycle -- leaving uranium enrichment and spent fuel reprocessing capabilities in the hands of national governments -- too risky on proliferation grounds? In early 2011, the Nuclear Threat Initiative and the Center for Strategic and International Studies launched the New Approaches to the Fuel Cycle (NAFC) project. This project, led by Corey Hinderstein and Sharon Squassoni, sought to build consensus on common goals,address practical challenges, and engage a spectrum of actors who influence nuclear energy policymaking. Drawing from industry, government, and NGO community expertise in the United States and abroad, the NAFC project worked to outline a vision for an integrated approach to nuclear supply and demand. The result, presented in this report, is the first comprehensive approach that contains guidelines for shaping a sustainable nuclear supply system and leverages existing trends in nuclear industry, with 'best practices' to help implement that sustainable system\"--Publisher's web site.
Book
The Advancement of Neutron-Shielding Materials for the Transportation and Storage of Spent Nuclear Fuel
In this paper, the mechanism of neutron absorption and common reinforced particles is introduced, and recent research progress on different types of neutron-shielding materials (borated stainless steels, B/Al Alloy, B4C/Al composites, polymer-based composites, and shielding concrete) for transportation and wet or dry storage of spent fuel is elaborated, and critical performance is summarized and compared. In particular, the most widely studied and used borated stainless steel and B4C/Al composite neutron-absorption materials in the field of spent fuel are discussed at length. The problems and solutions in the preparation and application of different types of neutron-shielding materials for spent fuel transportation and storage are discussed, and their research priorities and development trends are proposed.
Journal Article
Towards a more realistic MELCOR model for a dry cask for spent nuclear fuel. Part II: application
Nowadays, a great deal of attention is devoted to the development of best-estimate models able to produce more realistic outcomes. This is also the case for system codes, such as MELCOR, that are being mostly used in a conservative way especially when dealing with the licensing process. The above-mentioned need for more realistic results is at the core of this two-paper series related to the creation of a more accurate MELCOR model for the HI-STORM 100S dry cask. The findings obtained from the sensitivity studies carried out in the Part I are leveraged to set up an improved MELCOR model, the characteristics of which are consistent with the typical features of Spent Nuclear Fuel (SNF), and with geometrical and material properties of the cask itself. The addition of an axial power profile in the Fuel Assembly (FA), the better characterization of the flow losses in the air gap between internal metallic canister and external concrete-based overpack, and the choice of an appropriate value for the concrete thermal conductivity, are taken into account conjointly in this Part II. The outcomes from the improved MELCOR simulation are reported mainly in terms of the Peak Cladding Temperature (PCT), being the variable under regulatory surveillance. However, in addition to PCT, calculated temperature profiles are displayed and compared against the ones resulting from the previous model.
Journal Article
Towards a more realistic MELCOR model for a dry cask for spent nuclear fuel. Part I: sensitivity studies
The United States (US) Department Of Energy (DOE) has addressed the thermal analysis of the Spent Nuclear Fuel (SNF) stored within a dry cask system as a matter of high priority. In this regard, it is of utmost importance that simulation tools effectively reproduce the general thermal behavior of the modelled cask, including heat exchange and removal. Temperature distribution in the different components of the system is usually the focus of performed thermal analyses. In particular, attention is paid to the maximum temperature reached in the fuel cladding, namely the Peak Cladding Temperature (PCT). Within this framework, the present paper is the first of a two-paper series aimed at developing a more accurate model for the HI-STORM 100S cask. The dry cask in question is modelled and its behavior is simulated by means of the MELCOR code (version 2.2.18019). Stressing the need for a more realistic model rather than a conservative one, this paper reports the efforts undertaken to evaluate the influence of some specific modelling choices on the PCT. The study of the cask performance is therefore conducted taking into consideration three main factors: the axial power distribution in the Fuel Assembly (FA), the flow losses in the air gap between the internal canister and the external overpack, and the conductivity of the overpack concrete.
Journal Article
Extraction of uranyl from spent nuclear fuel wastewater via complexation—a local vibrational mode study
Context
The efficient extraction of uranyl from spent nuclear fuel wastewater for subsequent reprocessing and reuse is an essential effort toward minimization of long-lived radioactive waste. N-substituted amides and Schiff base ligands are propitious candidates, where extraction occurs via complexation with the uranyl moiety. In this study, we extensively probed chemical bonding in various uranyl complexes, utilizing the local vibrational modes theory alongside QTAIM and NBO analyses. We focused on (i) the assessment of the equatorial O-U and N-U bonding, including the question of chelation, and (ii) how the strength of the axial U
=
O bonds of the uranyl moiety changes upon complexation. Our results reveal that the strength of the equatorial uranium-ligand interactions correlates with their covalent character and with charge donation from O and N lone pairs into the vacant uranium orbitals. We also found an inverse relationship between the covalent character of the equatorial ligand bonds and the strength of the axial uranium-oxygen bond. In summary, our study provides valuable data for a strategic modulation of N-substituted amide and Schiff base ligands towards the maximization of uranyl extraction.
Method
Quantum chemistry calculations were performed under the PBE0 level of theory, paired with the relativistic NESCau Hamiltonian, currently implemented in Cologne2020 (interfaced with Gaussian16). Wave functions were expanded in the cc-pwCVTZ-X2C basis set for uranium and Dunning’s cc-pVTZ for the remaining atoms. For the bonding properties, we utilized the package LModeA in the local modes analyses, AIMALL in the QTAIM calculations, and NBO 7.0 for the NBO analyses.
Graphical abstract
Journal Article
Peening Techniques for Mitigating Chlorine-Induced Stress Corrosion Cracking of Dry Storage Canisters for Nuclear Applications
Fusion-welded austenitic stainless steel (ASS) was predominantly employed to manufacture dry storage canisters (DSCs) for the storage applications of spent nuclear fuel (SNF). However, the ASS weld joints are prone to chloride-induced stress corrosion cracking (CISCC), a critical safety issue in the nuclear industry. DSCs were exposed to a chloride-rich environment during storage, creating CISCC precursors. The CISCC failure leads to nuclear radiation leakage. Therefore, there is a critical need to enhance the CISCC resistance of DSC weld joints using promising repair techniques. This review article encapsulates the current state-of-the-art of peening techniques for mitigating the CISCC in DSCs. More specifically, conventional shot peening (CSP), ultrasonic impact peening (UIP), and laser shock peening (LSP) were elucidated with a focus on CISCC mitigation. The underlying mechanism of CISCC mitigation in each process was summarized. Finally, this review provides recent advances in surface modification techniques, repair techniques, and developments in welding techniques for CISCC mitigation in DSCs.
Journal Article
Behaviour of ruthenium in nitric media (HLLW) in reprocessing plants: a review and some perspectives
The reprocessing of spent nuclear fuel produces high level liquid waste (HLLW). Due to the decay heat, the concentrated nitric solutions containing fission products (FP) are stored in cooled tanks. Actually, the loss-of-cooling accident on these HLLW storage tanks could lead to releases of radioactive materials to the environment, especially ruthenium volatile species. After a general presentation on the physico-chemical behaviour of ruthenium in concentrated nitric acid, this paper successively details the different transfer modes of ruthenium reported in the literature, and the available trapping techniques. Based on this review, research perspectives are presented aiming at updating data concerning ruthenium behaviour in this context.
Journal Article
Assessment of Closing the Fuel Cycle in WWER Reactors: A Neutronics Perspective
This study evaluates the potential benefits of closing the fuel cycle of WWER reactors by reprocessing spent nuclear fuel (SNF) and utilizing MOX (Mixed Oxide) fuel for multiple irradiation cycles. The analysis quantifies reductions in natural uranium consumption and the volume of spent nuclear fuel through modeling the fraction of MOX fuel assemblies in the core and its impact on isotopic composition and neutron economy. Key equations are presented alongside results derived using computational tools.
Journal Article
Radiolysis and radiolytic by-product of N, N-diethylhydroxylamine in HNO3 at lower dose
N, N-diethylhydroxylamine (DEHA) is a novel salt-free reducing reagent used in the separation of Pu and Np from U in the treatment of used nuclear fuel. This paper reports on the radiation damage and radiolytic by-product of 0.5 mol L
−1
DEHA in 0.3 mol L
−1
~ 1.0 mol L
−1
HNO
3
at dose up to 25 kGy. Results show that the radiolysis rate of DEHA is less than 10%. The main radiolytic products are hydrogen, acetaldehyde, acetic acid and nitrous acid, which increase with the dose. The concentration of acetaldehyde and acetic acid is much higher than that of nitrous acid.
Journal Article