Valence and Local Environment of Molybdenum in Aluminophosphate Glasses for Immobilization of High Level Waste from Uranium-Graphite Reactor Spent Nuclear Fuel Reprocessing

2015 ◽  
Vol 1744 ◽  
pp. 73-78 ◽  
Author(s):  
Sergey V. Stefanovsky ◽  
Andrey A Shiryaev ◽  
Michael B. Remizov ◽  
Elena A. Belanova ◽  
Pavel A. Kozlov ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Local Environment ◽  
High Level Waste ◽  
Vitreous Phase ◽  
Xafs Spectroscopy ◽  
Graphite Reactor ◽  
State And Local ◽  
High Level ◽  
A Minor

ABSTRACTTwo Mo-bearing glasses considered as candidate forms for high level waste (HLW) a uranium-graphite reactor spent nuclear fuel (SNF) reprocessing were characterized. Incorporation of Mo in sodium aluminophosphate (SAP) glass increases its tendency to devitrification with segregation of orthophosphate phases. Valence state and local environment of Mo in the materials containing ∼2 wt.% MoO3 were determined by X-ray absorption fine structure (XAFS) spectroscopy. In the quenched samples composed of major vitreous and minor AlPO4 nearly all Mo is located in the vitreous phase as [Mo6+О6] units whereas in the annealed samples Mo is partitioned among vitreous and one or two orthophosphate crystalline phases in favor of the vitreous phase. Mo predominantly exists in a hexavalent state in distorted octahedral environment. Four oxygen ions are positioned at a distance of ∼1.71-1.73 Å and two - at a distance of 2.02-2.04 Å. Minor Mo(V) is also present as indicated by a response in EPR spectra with g ≈ 1.911-1.915.

Radionuclide Inventory of Vitrified Waste After Spent Nuclear Fuel Reprocessing at La Hague

2013 ◽  
Author(s):  
A. Meleshyn ◽  
U. Noseck
Keyword(s):  
Nuclear Fuel ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Auxiliary Information ◽  
Opalinus Clay ◽  
High Level Waste ◽  
Stable Elements ◽  
High Level ◽  
Model Approach

The primary aim of the present work was to determine the inventories of the radionuclides and stable elements in vitrified high-level waste produced at La Hague and delivered to Germany, which are of importance for long-term safety assessment of final repositories for radioactive wastes. For a subset of these radionuclides and stable elements, the inventories were determined — either by direct measurements or by involving established correlations — and reported by AREVA. This allowed verification of the validity of application of a model approach utilizing the data of burnup and activation calculations and auxiliary information on the reprocessing and vitrification process operated at La Hague. Having proved that such a model approach can be applied for prediction of inventories of actinides, fission and activation products in vitrified waste, the present work estimated the minimum, average and maximum inventories of the radionuclides, which are of importance for long-term safety assessment of final repositories for radioactive waste but were not reported by AREVA for delivered CSD-V canisters. The average and maximum inventories in individual CSD-V canisters predicted in the present approach were compared to the inventories predicted by Nagra for canisters with vitrified waste delivered from La Hague to Switzerland [1]. This comparison revealed a number of differences between these inventories despite the fact that the canisters delivered to Switzerland were produced in essentially the same way and from the common reprocessing waste stock as CSD-V canisters delivered to Germany. Therefore, a further work is required in order to identify the reason for the discrepancy in the present estimation versus the Nagra estimation [1]. Such a work should also address the recommendation by the international peer review of the Safety Report of the Project Opalinus Clay to obtain estimates of the inventories of long-lived mobile radionuclides (such as 14C, 36Cl, 79Se, and 129I), which contribute most to the dose estimates in the radiological safety assessments, if possible, in agreement with other countries with similar waste streams in order for a coordinated set of data to be generated [2]. Since vitrified waste from reprocessing of spent nuclear fuel at La Hague was delivered to several countries — Belgium, France, Germany, Japan, Netherlands, and Switzerland — an international effort can be recommended.

Shared, Regional Repositories: Developing a Practical Implementation Strategy

2009 ◽  
Author(s):  
Ewoud Verhoef ◽  
Charles McCombie ◽  
Neil Chapman
Keyword(s):  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Basic Concept ◽  
Implementation Strategy ◽  
Spent Nuclear Fuel ◽  
Action Plan ◽  
High Level Waste ◽  
Practical Implementation ◽  
Strategic Action ◽  
High Level

The basic concept within both EC funded SAPIERR I and SAPIERR II projects (FP6) is that of one or more geological repositories developed in collaboration by two or more European countries to accept spent nuclear fuel, vitrified high-level waste and other long-lived radioactive waste from those partner countries. The SAPIERR II project (Strategic Action Plan for Implementation of Regional European Repositories) examines in detail issues that directly influence the practicability and acceptability of such facilities. This paper describes the work in the SAPIERR II project (2006–2008) on the development of a possible practical implementation strategy for shared, regional repositories in Europe and lays out the first steps in implementing that strategy.

Thermal Analysis for Geologic Codisposal of SNF Packages

2002 ◽  
Author(s):  
Si Y. Lee
Keyword(s):  
Thermal Analysis ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Peripheral Region ◽  
Waste Form ◽  
Temperature History ◽  
High Level Waste ◽  
Geologic Repository ◽  
Geologic Disposal ◽  
High Level

The engineering viability of disposal of aluminum-clad, aluminum-based spent nuclear fuel (Al-SNF) in a geologic repository requires a thermal analysis to provide the temperature history of the waste form. Calculated temperatures are used to demonstrate compliance with criteria for waste acceptance into the geologic disposal system and as input to assess the chemical and physical behavior of the waste form within the Waste Package (WP). The leading codisposal WP design proposes that a central DOE Al-SNF canister be surrounded by five Defense Waste Process Facility (DWPF) glass log canisters, that is, High-level Waste Glass Logs (HWGL’s), and placed into a WP in a geologic disposal system. A DOE SNF canister having about 0.4318m diameter is placed along the central horizontal axis of the WP. The five HWGL’s will be located around the peripheral region of the DOE SNF canister within the cylindrical WP container. The codisposal WP will be laid down horizontally in a drift repository. In this situation, two waste form options for Al-SNF disposition are considered using the codisposal WP design configurations. They are the direct Al-SNF form and the melt-dilute ingot. In the present work, the reference geologic and design conditions are assumed for the analysis even though the detailed package design is continuously evolved. This paper primarily dealt with the thermal performance internal to the codisposal WP for the qualification study of the WP containing Al-SNF. Thermal analysis methodology and decay heat source terms have been developed to calculate peak temperatures and temperature profiles of Al-SNF package in the DOE spent nuclear fuel canister within the geologic codisposal WP.

Deep Geological Disposal of Spent Nuclear Fuel and High-Level Waste

2015 ◽  
pp. 367-399
Author(s):  
Želimir Veinović ◽  
Biljana Kovačević Zelić ◽  
Dubravko Domitrović
Keyword(s):  
Nuclear Fuel ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
High Level Waste ◽  
Modern World ◽  
Geological Disposal ◽  
Long Term Storage ◽  
Deep Geological Disposal ◽  
Safety Cases ◽  
High Level

Management of Spent Nuclear Fuel (SF) and High-Level Waste (HLW) is one of the most important and challenging problems of the modern world. Otherwise a clean, cheap, constant, and secure way to produce electricity, nuclear power plants create large amounts of highly hazardous waste. Repositories—deep Geological Disposal Facilities (GDF)—for these types of waste must prevent radionuclides from reaching the biosphere, for up to 1,000,000 years, migrating from a deep (more than 300m), stable geological environment. At present, there are no operating GDFs for SF and/or HLW, mostly due to the difficult and complex task of preparing safety cases and licensing. The purpose of this chapter is to validate the generic R&D activities in this area and present alternative concepts of Radioactive Waste (RW) management: retrievability, reversibility, regional GDFs, long-term storage, and deep borehole disposal, demonstrating the main engineering tasks in solving the problem of RW management and disposal.

An Approach to Model Abstraction of Stress Corrosion Cracking Damage in Management of Spent Nuclear Fuel and High-Level Waste

2013 ◽  
Author(s):  
Tae M. Ahn
Keyword(s):  
Crack Propagation ◽  
Stress Corrosion ◽  
Nuclear Fuel ◽  
Stress Corrosion Cracking ◽  
Spent Nuclear Fuel ◽  
Corrosion Cracking ◽  
Propagation Model ◽  
High Level Waste ◽  
Multiple Cracks ◽  
High Level

This paper presents an approach to assess stress corrosion cracking (SCC) damage of a canister for use in confinement management (extended dry storage or geological disposal) of radionuclides from spent nuclear fuel and high-level (radioactive) waste. Localized corrosion, mainly in pitting form and fabrication flaws, were analyzed as a possible precursor to SCC using field/laboratory data. This paper assesses single crack propagation over long time periods and estimates the potential maximum opening area resulting from multiple cracks. This crack propagation model was developed by the Sandia National Laboratories (SNL) for disposal under seismic conditions, and it appears to be conservative with respect to radionuclide releases through the opening area. The SNL model could be applied to the weld and various metals for both management applications. The conservative SNL approach could be used to estimate consequences of radionuclides dispersals, if a canister failed as the confinement barrier.

Integrated Model of Korean Spent Fuel and High Level Waste Disposal Options

2009 ◽  
Author(s):  
Yongsoo Hwang ◽  
Ian Miller
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Near Field ◽  
Integrated Model ◽  
High Level Waste ◽  
Spent Fuel ◽  
Parameter Uncertainties ◽  
Design Concepts ◽  
High Level

This paper describes an integrated model developed by the Korean Atomic Energy Research Institute (KAERI) to simulate options for disposal of spent nuclear fuel (SNF) and reprocessing products in South Korea. A companion paper (Hwang and Miller, 2009) describes a systems-level model of Korean options for spent nuclear fuel (SNF) management in the 21’st century. The model addresses alternative design concepts for disposal of SNF of different types (CANDU, PWR), high level waste, and fission products arising from a variety of alternative fuel cycle back ends. It uses the GoldSim software to simulate the engineered system, near-field and far-field geosphere, and biosphere, resulting in long-term dose predictions for a variety of receptor groups. The model’s results allow direct comparison of alternative repository design concepts, and identification of key parameter uncertainties and contributors to receptor doses.

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