The Cumulative Beta Decay and the Effect of Electron Radiation on a Simulated HLW Sintered Glass

1997 ◽  
Vol 506 ◽  
Author(s):  
A.M. Bevilacqua ◽  
N.B. Messi de Bernasconi ◽  
D.O. Russo ◽  
S. Prastalo ◽  
M. Sanfilippo ◽  
...  
Keyword(s):  
Heavy Metal ◽  
Beta Decay ◽  
Internal Conversion ◽  
Fission Products ◽  
Electron Radiation ◽  
Spent Fuel ◽  
Sintered Glass ◽  
Computer Codes ◽  
The Mean ◽  
High Level

The first aim of the work was to calculate, by means of the computer codes ORIGEN2. 1, ITS 3.0 and TRIM95, the beta/gamma cumulative dose and displacements produced by the fission products hold in the high level wastes (HLW) from reprocessed CANDU spent fuel, up to one million years, with a burnup of 7,000 MWd/tU and a cooling time of 20 years. Based on that results were calculated and graphically presented the mean electron energy (β−, internal conversion and Auger) and the cumulative beta events and ionization dose due to electrons, per metric ton of heavy metal and per cubic centimeter of the sintered HLW glass form, as a function of the time up to one million year.

Characteristics and quantities of radioactive wastes

1986 ◽  
Vol 319 (1545) ◽  
pp. 5-16 ◽  
Keyword(s):  
Nuclear Power ◽  
Small Volume ◽  
Fuel Cycle ◽  
Residual Activity ◽  
Fission Products ◽  
High Level Waste ◽  
Spent Fuel ◽  
Fuel Cladding ◽  
Waste Streams ◽  
High Level

Estimates are given of the total quantities of radioactivity, and of the contribution from different isotopes to this total, arising in the wastes from civil nuclear power generation; the figures are normalized to 1 GW (e) y of power production. The intensity of the heat and y-radiation emitted by the spent fuel has been calculated, and their decrease as the radioactivity decays. Reprocessing the spent fuel results in 95% or more of the fission products and higher actinides being concentrated in a small volume of high-level, heat-emitting waste. The total decay curve of unreprocessed spent fuel or of the separated high-level waste is dominated by the decay of some fission products for a few hundred years and then by the decay of some actinide isotopes for some tens of thousands of years. The residual activity is compared with that of the original uranium ore. Some of the long-lived activity will appear in other waste streams, particularly on the fuel cladding, and the volumes and activities of these wastes arising in this country are recorded. The long-lived activity arising from reactor decommissioning will be small compared with the annual arisings from the fuel cycle.

The Application Research of Concentration & Denitration Technology in Advanced Reprocessing Process

2012 ◽  
Vol 560-561 ◽  
pp. 637-643
Author(s):  
Yong Li ◽  
Xue Gang Liu ◽  
Jin Chen
Keyword(s):  
Nitric Acid ◽  
Nuclear Power ◽  
Liquid Waste ◽  
Fission Products ◽  
Extraction Process ◽  
Purex Process ◽  
Current Status ◽  
Spent Fuel ◽  
High Level Liquid Waste ◽  
High Level

The proper management of spent fuel arising from nuclear power production is a key issue for the sustainable development of nuclear energy. While conventional reprocessing process, PUREX process, was successful to recover uranium and plutonium, in recent years some countries have turned to focus on advanced reprocessing process, which features of partitioning of minor actinides (MA) and long-lived fission products(LLFP). Most advanced reprocessing processes under development involve new extractants and additional extraction cycles. In China, TRPO extraction process has been developed to partition MA/LLFP from high-level liquid waste(HLLW) since early 1980’s. In parallel to R&D work on separation technologies, studies on concentration & denitration process have been evolved to prepare feed solutions to suit qualifications of extraction. Industrially, concentration & denitration is the internationally recognized standard to treat HLLW released from PUREX before vitrification. It enables to minimize the volume of interim storage, to restrain the corrosion of storage tank, to recover nitric acid in HLLW and to reduce the required evaporation duty of the vitrification process. Generally, the constitution of concentrated HLLW has little impact on the following vitrification process. But when concentration & denitration acts as pretreatment process of partitioning, the composition of actinides, fission products, and nitric acid in concentrated HLLW solution plays significant role in extraction process. A series of technical issues relevant to the connection between concentration ﹠denitration and extractions should be solved. This paper describes current status of concentration & denitration technology utilized in industry and under reprocessing plants. The specific separation requirements in advanced reprocessing process and challenges to apply concentration & denitration process are addressed. Besides, concentration & denitration process was tested in laboratory to adjust feed solutions for TRPO and Cyanex301 partitioning. Results demonstrate its promising prospect in advanced reprocessing process.

Hot Cell Preparation of Tesiing Materials

1984 ◽  
Vol 44 ◽  
Author(s):  
R. E. Thornhill ◽  
C. A. Knox
Keyword(s):  
Fission Products ◽  
Materials Characterization ◽  
High Level Waste ◽  
Nuclear Waste Repository ◽  
Spent Fuel ◽  
Cell Preparation ◽  
High Radiation ◽  
High Level ◽  
Commercial Reactor ◽  
Simulated High Level Waste

AbstractIt is important in nuclear waste repository development that testing be done with materials containing a radionuclide spectrum representative of actual wastes. To meet the need for such materials, the Materials Characterization Center (MCC) has prepared simulated high level waste (HLW) glasses with radionuclides representative of about 10-, 300-, and 1000-year-old waste. A quantity of well characterized spent fuel also has been acquired for the same purpose. Glasses containing 10- and 300-year-old wastes, and the spent fuel specimens, must be fabricated in a hot cell. Hot cell conditions (high radiation field, remote operation, and difficulty of repairs) require that procedures and equipment normally used in materials preparation out-of-cell be modified for hot cell applications.This paper discusses the fabrication of two glasses, and the preparation of test specimens of these glasses and spent fuel. One of the glasses is a 76–68 composition, which is fully loaded with actual commercial reactor fission product waste. The other glass contains simulated Barnwell Nuclear Fuel Plant waste, doped with different combinations of fission products and actinides. The spent fuel is a 10-year-old PWR material. Special techniques have been used to achieve high quality, well characterized testing materials, including specimens in the form of segments, wafers, cylinders, and powders of these materials.

German Nuclear High-Level Waste Program - Key Research Areas

1985 ◽  
Vol 50 ◽  
Author(s):  
Helmut Geipel
Keyword(s):  
Mine Safety ◽  
High Level Waste ◽  
Salt Mine ◽  
Federal Republic Of Germany ◽  
Spent Fuel ◽  
Research Areas ◽  
The Status ◽  
Safety Studies ◽  
Computer Codes ◽  
High Level

AbstractBased on more than 10 years of research and development, vitrification and high level waste disposal have reached the status of demonstration projects in the Federal Republic of Germany: hot operation of the PAMELA vitrification plant is scheduled for october 1985, and a disposal test with 30 canisters of high active glass is being prepared in the ASSE salt mine. Safety studies for a model repository led to a good understanding of the relevant phenomena; they will continue using sitespezific parameters. Modelling and computer codes will be further developed in international cooperation. In addition to reprocessing, the technology for direct disposal of spent fuel will be developed and demonstrated in the next years.

scholarly journals Transmutations of Long-Lived and Medium-Lived Fission Products Extracted from CANDU and PWR Spent Fuels in an Accelerator-Driven System

2019 ◽  
Vol 2019 ◽  
pp. 1-13
Author(s):  
Alper Buğra Arslan ◽  
İlayda Yilmaz ◽  
Gizem Bakir ◽  
Hüseyin Yapici
Keyword(s):  
Fission Products ◽  
Time Dependent ◽  
Thermal Reactor ◽  
Beam Power ◽  
Spent Fuel ◽  
Driven System ◽  
Helium Gas ◽  
Computer Codes ◽  
Accelerator Driven System ◽  
Additional Fuel

This study presents the time-dependent analyses of transmutations of long-lived fission products (LLFPs) and medium-lived fission products (MLFPs) occurring in thermal reactors in a conceptual helium gas-cooled accelerator-driven system (ADS). In accordance with this purpose, the CANDU-37 and PWR 15 × 15 spent fuels are separately considered. The ADS consists of LBE-spallation neutron target, subcritical fuel zone, and graphite reflector zone. While the considered ADS is fueled with the spent nuclear fuels extracted from each thermal reactor without the use of additional fuel, fission products extracted from same thermal reactor are also placed into transmutation zone in graphite reflector zone. The LLFP transmutation performance of the modified ADS is analyzed by considering three different spent fuels extracted from the thermal reactors. Spent fuels are extracted from CANDU-37 in case A, from PWR-15 × 15 in case B, and from CANDU-37 fueled with mixture of PWR 15 × 15 spent fuel and 46% ThO2 in case C. The LBE target is bombard with protons of 1000 MeV. The proton beam power is assumed as 20 MW, which corresponds to 1.24828·1017 protons per second. MCNPX 2.7 and CINDER 90 computer codes are used for the time-dependent burn calculations. The ADS is operated under subcritical mode until the value of keff increases to 0.984, and the maximum operation times are obtained as 3400, 3270, and 5040 days according to the spent fuel cases of A, B, and C, respectively. The calculations bring out that in the modified ADS, LLFPs and MLFPs, which are extracted from thermal reactors, can be transformed to stable isotopes in significant amounts along with energy production.

Advanced ORIENT Cycle: Progress on Fission Product Separation and Utilization

2010 ◽  
Author(s):  
Isao Yamagishi ◽  
Masaki Ozawa ◽  
Hitoshi Mimura ◽  
Shohei Kanamura ◽  
Koji Mizuguchi
Keyword(s):  
Fuel Cycle ◽  
Liquid Waste ◽  
Fission Products ◽  
Spent Fuel ◽  
High Level Liquid Waste ◽  
Pt Electrode ◽  
Acid Media ◽  
Electrolytic Hydrogen ◽  
Product Separation ◽  
High Level

Fission reaction of U-235 and/or plutonium generates more than 40 elements and 400 nuclides in the spent fuel. Among them, 31 elements are categorized as rare metals. In a conventional fuel cycle U and Pu are reused but others are vitrified for disposal. Adv.-ORIENT (Advanced Optimization by Recycling Instructive Elements) Cycle strategy was drawn up for the minimization of radio-toxicity and volume of radioactive waste as well as the utilization of valuable elements/nuclides in the waste. The present paper describes the progress on Fission Products (FP) separation in this Cycle. Highly functional inorganic adsorbent (AMP-SG, silica gel loaded with ammonium molybdophosphate) and organic microcapsule (CE-ALG, alginate gel polymer enclosed with crown ether D18C6) were developed for separation of heat-generating Cs and Sr nuclides, respectively. The AMP-SG adsorbed more than 99% of Cs selectively from a simulated High-level Liquid Waste (HLLW). The ALG microcapsule adsorbed 0.0249 mmol/g of Sr and exhibited the order of its selectivity; Ba > Sr > Pd >> Ru > Rb > Ag. The electrodeposition is advantageous for both recovery and utilization of PGMs (Ru, Rh, Pd) and Tc because PGMs are recovered as metal on Pt electrode. Among PGMs, Pd was easily deposited on the Pt electrode. In the presence of Pd or Rh the reduction of Ru and Tc was accelerated more in hydrochloric acid media than in nitric acid. In the simulated HLLW, the redox reaction of Fe(III)/Fe(II) disturbed deposition of elements except for Pd. The deposits on Pt electrode showed higher catalytic reactivity on electrolytic hydrogen production than the original Pt electrode. The reactivity of deposits prepared from the simulated HLLW was higher than that from solution containing only PGM.

scholarly journals German Spent Nuclear Fuel Legacy: Characteristics and High-Level Waste Management Issues

2013 ◽  
Vol 2013 ◽  
pp. 1-11 ◽  
Author(s):  
A. Schwenk-Ferrero
Keyword(s):  
Nuclear Fuel ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Fission Products ◽  
High Level Waste ◽  
Minor Actinides ◽  
Spent Fuel ◽  
Cycle Simulation ◽  
High Level

Germany is phasing-out the utilization of nuclear energy until 2022. Currently, nine light water reactors of originally nineteen are still connected to the grid. All power plants generate high-level nuclear waste like spent uranium or mixed uranium-plutonium dioxide fuel which has to be properly managed. Moreover, vitrified high-level waste containing minor actinides, fission products, and traces of plutonium reprocessing loses produced by reprocessing facilities has to be disposed of. In the paper, the assessments of German spent fuel legacy (heavy metal content) and the nuclide composition of this inventory have been done. The methodology used applies advanced nuclear fuel cycle simulation techniques in order to reproduce the operation of the German nuclear power plants from 1969 till 2022. NFCSim code developed by LANL was adopted for this purpose. It was estimated that ~10,300 tonnes of unreprocessed nuclear spent fuel will be generated until the shut-down of the ultimate German reactor. This inventory will contain ~131 tonnes of plutonium, ~21 tonnes of minor actinides, and 440 tonnes of fission products. Apart from this, ca.215 tonnes of vitrified HLW will be present. As fission products and transuranium elements remain radioactive from 104to 106years, the characteristics of spent fuel legacy over this period are estimated, and their impacts on decay storage and final repository are discussed.

Salt waste treatment from a LiCl-KCl based pyrochemical spent fuel treatment process

2008 ◽  
Vol 96 (4-5) ◽  
Author(s):  
Mike T. Harrison ◽  
Howard E. Simms ◽  
Angela Jackson ◽  
Robert G. Lewin
Keyword(s):  
Alkali Metals ◽  
Treatment Process ◽  
Spent Nuclear Fuel ◽  
Waste Treatment ◽  
Electrochemical Techniques ◽  
Fission Products ◽  
High Level Waste ◽  
Spent Fuel ◽  
High Level ◽  
Salt Waste

Spent nuclear fuel may be treated using molten salt electrochemical techniques to separate fission products and actinide metals. Salt waste arising from the electrorefining process contains alkali metals, alkaline-earth and rare earth fission products, along with residual actinides. The removal of fission product elements has been investigated using zeolite ion exchange and phosphate precipitation, which allow the salt electrolyte to be recycled back into the main electrorefining vessel. Recycling the salt minimizes the volume of high level waste (HLW) generated and yields the fission products in a form more amenable to immobilization in a final disposal matrix. Several sets of experiments have been completed, all of which have significant implications for the use of these techniques on an industrial scale, as well as their ability to clean up the salt, and potentially produce robust and durable waste forms.

Dissolution Kinetics of UO2. I. Flow-Through tests on UO2.00 Pellets and Polycrystalline Schoepite Samples in Oxygenated, Carbonate/Bicarbonate Buffer Solutions at 25°C

1991 ◽  
Vol 257 ◽  
Author(s):  
Son N. Nguyen ◽  
Homer C. Weed ◽  
Herman R. Leider ◽  
Ray B. Stout
Keyword(s):  
Fission Products ◽  
Dissolution Kinetics ◽  
Spent Fuel ◽  
Bicarbonate Buffer ◽  
Buffer Solutions ◽  
Flow Through ◽  
High Level ◽  
Radionuclide Release ◽  
Kinetics Of

ABSTRACTThe modelling of radionuclide release from waste forms is an important part of the performance assessment of a potential, high-level radioactive waste repository. Since spent fuel consists of UO2 containing actinide elements and other fission products, it is necessary to determine the principal parameters affecting UO2 dissolution and quantify their effects on the dissolution rate before any prediction of long term release rates of radionuclides from the spent fuel can be made.

DFT Benchmark Study of the O--O Bond Dissociation Energy in Peroxides Validated with High-Level Ab-Initio Calculations

2019 ◽  
Author(s):  
Danilo Carmona ◽  
Pablo Jaque ◽  
Esteban Vöhringer-Martinez
Keyword(s):  
Reference Value ◽  
Basis Set ◽  
Basis Sets ◽  
Mean Deviation ◽  
Bond Dissociation ◽  
Dissociation Energies ◽  
The Mean ◽  
Experimental Values ◽  
High Level ◽  
Almost All

<div><div><div><p>Peroxides play a central role in many chemical and biological pro- cesses such as the Fenton reaction. The relevance of these compounds lies in the low stability of the O–O bond which upon dissociation results in radical species able to initiate various chemical or biological processes. In this work, a set of 64 DFT functional-basis set combinations has been validated in terms of their capability to describe bond dissociation energies (BDE) for the O–O bond in a database of 14 ROOH peroxides for which experimental values ofBDE are available. Moreover, the electronic contributions to the BDE were obtained for four of the peroxides and the anion H2O2− at the CBS limit at CCSD(T) level with Dunning’s basis sets up to triple–ζ quality provid- ing a reference value for the hydrogen peroxide anion as a model. Almost all the functionals considered here yielded mean absolute deviations around 5.0 kcal mol−1. The smallest values were observed for the ωB97 family and the Minnesota M11 functional with a marked basis set dependence. Despite the mean deviation, order relations among BDE experimental values of peroxides were also considered. The ωB97 family was able to reproduce the relations correctly whereas other functionals presented a marked dependence on the chemical nature of the R group. Interestingly, M11 functional did not show a very good agreement with the established order despite its good performance in the mean error. The obtained results support the use of similar validation strategies for proper prediction of BDE or other molecular properties by DF Tmethods in subsequent related studies.</p></div></div></div>

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