Summary of Results From the Series 2 and Series 3 NNWSI Bare Fuel Dissolution Tests

1987 ◽  
Vol 112 ◽  
Author(s):  
C. N. Wilson
Keyword(s):  
Spent Nuclear Fuel ◽  
Fine Particles ◽  
Fission Products ◽  
Test Site ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Nuclear Waste Storage ◽  
Waste Storage ◽  
Rate Of Increase ◽  
In Series

AbstractThe Nevada Nuclear Waste Storage Investigations (NNWSI) Project is studying dissolution and radionuclide release behavior of spent nuclear fuel in Nevada Test Site groundwater. Specimens prepared from pressurized water reactor (PWR) fuel rod segments were tested for multiple cycles in J-13 well water. The Series 2 tests were run in unsealed silica vessels under ambient hot cell air (25°C) for five cycles for a total of 34 months. The Series 3 tests were run in sealed stainless steel vessels at 25°C and 85°C for three cycles for a total of 15 months. Selected summary results from Series 2 and Series 3 tests with bare fuel specimens are reported.Actinide concentrations tended to saturate and then often decreased during test cycles. Uranium concentrations in later test cycles ranged from 1 to 2 μg/ml in the Series 2 Tests versus about 0.1 to 0.4 μg/ml in Series 3 with the lowest concentrations occurring in the 85°C tests. Formation of a calciumuranium-silicate phase identified as uranophane in the 85°C Series 3 Tests is thought to have limited uranium concentration in these tests. Americium-241, Pu-239 and Pu-240 activities measured in filtered solution decreased to less than 1 pCi/ml in the 85°C tests. Preferential release of fission products Cs, I, Sr and Tc, and activation product C-14, was indicated relative to the actinides. Tc-99 and Cs-137 activities measured in solution after Cycle 1 increased linearly with time, with the rate of increase greater at 85°C than at 25°C. Continuous preferential release of soluble fission products is thought to result primarily from the dissolution of fine particles of fission product phases concentrated on grain boundaries.

Study of the release of the fission gases (Xe and Kr) and the fission products (Cs and I) under anoxic conditions in bicarbonate water

2015 ◽  
Vol 1744 ◽  
pp. 35-41 ◽  
Author(s):  
Ernesto González-Robles ◽  
Markus Fuß ◽  
Elke Bohnert ◽  
Nikolaus Müller ◽  
Michel Herm ◽  
...  
Keyword(s):  
Gas Phase ◽  
Nuclear Fuel ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Collaborative Project ◽  
Fuel Rod ◽  
Release Ratio

ABSTRACTFor safety assessment analyses of the disposal of spent nuclear fuel (SNF) in deep geological repositories it is indispensable to evaluate the contribution of fission products to the instant release fraction (IRF). During the last three years the EURATOM FP7 Collaborative Project, “Fast / Instant Release of Safety Relevant Radionuclides from Spent Nuclear Fuel (CP FIRST-Nuclides)” was carried out to get a better understanding of the IRF.Within CP FIRST-Nuclides, a leaching experiment with a cladded SNF pellet was performed in bicarbonate water (19 mM NaCl + 1 mM NaHCO3) under Ar /H2 atmosphere over 333 days. The cladded SNF pellet was obtained from a fuel rod segment which was irradiated in the Gösgen pressurized water reactor; the average burn-up of the segment was 50.4 MWd/kgUO2. In the multi-sampling experiment, gaseous and liquid samples were taken periodically. The moles of the fission gases Kr and Xe released in the gas phase and those of 129I and 137Cs released in solution were measured. Cumulative release fractions of (1.6 ± 0.2)·10-1 fission gases, (1.6 ± 0.1)·10-1129I and (3.9 ± 0.2)·10-2 137Cs, respectively, were achieved after 333 days of leaching. Accordingly the release ratio of fission gases to 129I was 1:1 and the release ratio of fission gases to 137Cs was 4:1, respectively.

scholarly journals Management of the Energy and Economic Potential of Nuclear Waste Use

Energies ◽  
2021 ◽  
Vol 14 (12) ◽  
pp. 3709
Author(s):  
Bader Alshuraiaan ◽  
Sergey Pushkin ◽  
Anastasia Kurilova ◽  
Magdalena Mazur
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Waste ◽  
Spent Nuclear Fuel ◽  
Integrated Approach ◽  
Economic Potential ◽  
Nuclear Waste Storage ◽  
Scientific Discussion ◽  
Management Algorithm ◽  
Waste Storage ◽  
The Eu

Recently, issues related to the effects (benefit or harm) of processing nuclear waste and its further use as fuel have been increasingly often raised in the scientific discussion. In this regard, the research aims to investigate issues related to the assessment of the economic potential of nuclear waste use, as well as the cooperation between states in the context of the reduction of risks associated with nuclear waste storage and processing. The research methodology is based on an integrated approach, including statistical, factor analysis, and the proposed system of performance indicators for managing spent nuclear fuel use. The research was carried out on the basis of materials from Russia and the EU countries. In the course of the study, a model of cooperation between states has been developed (based on the example of technologies and methods of processing nuclear waste used in the EU and Russia) according to the nuclear waste (spent nuclear fuel) management algorithm. The model considers the risks and threats associated with ecology and safety. The developments and other results described in the study should be used in further research devoted to the use of nuclear waste as heat-producing elements.

scholarly journals Muon radiography to visualise individual fuel rods in sealed casks

2021 ◽  
Vol 7 ◽  
pp. 12
Author(s):  
Thomas Braunroth ◽  
Nadine Berner ◽  
Florian Rowold ◽  
Marc Péridis ◽  
Maik Stuke
Keyword(s):  
Spent Nuclear Fuel ◽  
Cosmic Ray ◽  
Scattering Data ◽  
Pressurized Water Reactor ◽  
Test Statistic ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Acceptance Criterion ◽  
Muon Radiography ◽  
Kolmogorov Smirnov

Cosmic-ray muons can be used for the non-destructive imaging of spent nuclear fuel in sealed dry storage casks. The scattering data of the muons after traversing provides information on the thereby penetrated materials. Based on these properties, we investigate and discuss the theoretical feasibility of detecting single missing fuel rods in a sealed cask for the first time. We perform simulations of a vertically standing generic cask model loaded with fuel assemblies from a pressurized water reactor and muon detectors placed above and below the cask. By analysing the scattering angles and applying a significance ratio based on the Kolmogorov-Smirnov test statistic we conclude that missing rods can be reliably identified in a reasonable measuring time period depending on their position in the assembly and cask, and on the angular acceptance criterion of the primary, incoming muons.

scholarly journals Contribution to the study of fission products release from nuclear fuels in severe accident conditions: effect of the pO2 on Cs, Mo and Ba speciation

2020 ◽  
Vol 6 ◽  
pp. 2 ◽  
Author(s):  
Claire Le Gall ◽  
Fabienne Audubert ◽  
Jacques Léchelle ◽  
Yves Pontillon ◽  
Jean-Louis Hazemann
Keyword(s):  
Oxygen Potential ◽  
Fission Products ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Mox Fuel ◽  
Chemical Behaviour ◽  
Accident Conditions ◽  
The Impact

The objective of this work is to experimentally investigate the effect of the oxygen potential on the fuel and FP chemical behaviour in conditions representative of a severe accident. More specifically, the speciation of Cs, Mo and Ba is investigated. These three highly reactive FP are among the most abundant elements produced through 235U and 239Pu thermal fission and may have a significant impact on human health and environmental contamination in case of a light water reactor severe accident. This work has set out to contribute to the following three fields: providing experimental data on Pressurized Water Reactor (PWR) MOX fuel behaviour submitted to severe accident conditions and related FP speciation; going further in the understanding of FP speciation mechanisms at different stages of a severe accident; developing a method to study volatile FP behaviour, involving the investigation of SIMFuel samples manufactured at low temperature through SPS. In this paper, a focus is made on the impact of the oxygen potential towards the interaction between irradiated MOX fuels and the cladding, the interaction between Mo and Ba under oxidizing conditions and the assessment of the oxygen potential during sintering.

scholarly journals Pressurized water reactor spent nuclear fuel data library produced with the Serpent2 code

Data in Brief ◽  
2020 ◽  
Vol 33 ◽  
pp. 106429
Author(s):  
Zsolt Elter ◽  
Li Pöder Balkeståhl ◽  
Erik Branger ◽  
Sophie Grape
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Data Library

Physico-chemical characterization of a spent UO2 fuel with respect to its stability under final disposal conditions

2014 ◽  
Vol 1665 ◽  
pp. 283-289 ◽  
Author(s):  
Ernesto González-Robles ◽  
Detlef H. Wegen ◽  
Elke Bohnert ◽  
Dimitrios Papaioannou ◽  
Nikolaus Müller ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Chemical Characterization ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Final Disposal ◽  
Fuel Rod ◽  
Physico Chemical ◽  
The Stability ◽  
Fission Gas Release

ABSTRACTTwo adjacent fuel rod segments were irradiated in a pressurized water reactor achieving an average burn-up of 50.4 GWd/tHM. A physico-chemical characterisation of the high burn-up fuel rod segments was performed, to determine properties relevant to the stability of the spent nuclear fuel under final disposal conditions. No damage of the cladding was observed by means of visual examination and γ-scanning. The maximal oxide layer thickness was 45 µm. The relative fission gas release was determined to be (8.35 ± 0.66) %. Finally, a rim thickness of 83.7 µm and a rim porosity of about 20% were derived from characterisation of the cladded pellets.

scholarly journals Analysis of Release Model Effect in the Transport of Fission Products Simulating the FPT3 Test Using MELCOR 2.1 and MELCOR 2.2

2021 ◽  
Vol 13 (14) ◽  
pp. 7964
Author(s):  
Alain Flores y Flores ◽  
Danilo Ferretto ◽  
Tereza Marková ◽  
Guido Mazzini
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Nuclear Industry ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Accident Simulation ◽  
Experimental Facilities ◽  
The Difference

The severe accident integral codes such as Methods for Estimation of Leakages and Consequences of Releases (MELCOR) are complex tools used to simulate and analyse the progression of a severe accident from the onset of the accident up to the release from the containment. For this reason, these tools are developed in order to simulate different phenomena coupling models which can simulate simultaneously the ThermoHydraulic (TH), the physics and the chemistry. In order to evaluate the performance in the prediction of those complicated phenomena, several experimental facilities were built in Europe and all around the world. One of these facilities is the PHEBUS built by Institut de Radioprotection et de Sûrete Nucléaire (IRSN) in Cadarache. The facility reproduces the severe accident phenomena for a pressurized water reactor (PWR) on a volumetric scale of 1:5000. This paper aims to continue the assessment of the MELCOR code from version 2.1 up to version 2.2 underlying the difference in the fission product transport. The assessment of severe accident is an important step to the sustainability of the nuclear energy production in this period where the old nuclear power plants are more than the new reactors. The analyses presented in this paper focuses on models assessment with attention on the influence of B4C oxidation on the release and transport of fission products. Such phenomenon is a concern point in the nuclear industry, as was highlighted during the Fukushima Daiichi accident. Simulation of the source term is a key point to evaluate the severe accident hazard along with other safety aspects.

Composition Changes and Future Challenges for the Sellafield Waste Vitrification Plant

2009 ◽  
Vol 1193 ◽  
Author(s):  
A. Riley ◽  
S. Walker ◽  
Nick R. Gribble
Keyword(s):  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Feed Composition ◽  
Power Stations ◽  
Highly Active ◽  
Long Term Storage ◽  
Waste Storage ◽  
Waste Vitrification ◽  
Future Challenges ◽  
Future Work

AbstractThe Sellafield Waste Vitrification Plant (WVP) immobilises highly active liquors produced during reprocessing of spent nuclear fuel by bonding the fission products as metal oxides into a borosilicate glass matrix. This provides a stable and durable waste form suitable for safe long term storage and ultimate disposal.WVP was commissioned with feed from reprocessing of Magnox reactor fuel. This material is relatively low in fission product content per tonne of fuel, but contains significant Al and Mg from fuel cladding. WVP also routinely treats a blended feed made from a mixture of Magnox and Oxide reprocessing products. The Oxide fuel from Light Water Reactor (LWR) and Advanced Gas Cooled (AGR) power stations is of higher burnup and contains more fission products per tonne of fuel, also Gd and other process additives. Blending allows 25% incorporation of waste oxides by weight in glass to be achieved routinely.Recent programmes of development work in WVP have been aimed at increasing incorporation rates for these feeds, to reduce the number of waste containers produced for disposal. Work has also focussed on increasing the throughput of WVP, to more rapidly treat current stocks of liquid reprocessing waste, both by increasing the feed rate and by improving the lifetime of key components to improve plant availability.Future challenges for WVP include flowsheet changes to treat historic stocks of reprocessing wastes containing high U, Fe and Cr. Washout of solids from the base of waste storage tanks in preparation for decommissioning is also likely to give high Mo feeds. Development of flowsheet and glass formulation to accept these changes in feed composition will be a key objective of future work.

Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel—I: Methodology Overview

1999 ◽  
Vol 125 (3) ◽  
pp. 255-270 ◽  
Author(s):  
Dale B. Lancaster ◽  
Emilio Fuentes ◽  
Chi H. Kang ◽  
Meraj Rahimi
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor
Sign in / Sign up

Export Citation Format

Share Document