bluuebunny/crossref_metadata_2025_split huggingface dataset:
Geological Criteria for Site Selection of an LILW Radioactive Waste Repository in the Philippines
Authors: Aurelio Mario , Taguibao Kristine Joy , Vargas Edmundo , Palattao Maria Visitacion , Reyes Rolando , Nohay Carl , Luna Roy Anthony , Singayan Alfonso 
Published: February 2014

In the selection of sites for disposal facilities involving low- and intermediate-level radioactive waste (LILW), International Atomic Energy Agency (IAEA) recommendations require that “the region in which the site is located shall be such that significant tectonic and surface processes are not expected to occur with an intensity that would compromise the required isolation capability of the repository”. Evaluating the appropriateness of a site therefore requires a deep understanding of the geological and tectonic setting of the area. The Philippines sits in a tectonically active region frequented by earthquakes and volcanic activity. Its highly variable morphology coupled with its location along the typhoon corridor in the west Pacific region subjects the country to surface processes often manifested in the form of landslides.
The Philippine LILW near surface repository project site is located on the north eastern sector of the Island of Luzon in northern Philippines. This island is surrounded by active subduction trenches; to the east by the East Luzon Trough and to the west by the Manila Trench. The island is also traversed by several branches of the Philippine Fault System. The Philippine LILW repository project is located more than 100 km away from any of these major active fault systems. In the near field, the project site is located less than 10 km from a minor fault (Dummon River Fault) and more than 40 km away from a volcanic edifice (Mt. Caguas).
This paper presents an analysis of the potential hazards that these active tectonic features may pose to the project site. The assessment of such geologic hazards is imperative in the characterization of the site and a crucial input in the design and safety assessment of the repository.
Waste Handling and Storage in the Decontamination Pilot Projects of JAEA for Environments of Fukushima
Authors: Nakayama S. , Kawase K. , Iijima K. , Kato M. 
Published: February 2014

After the Fukushima Daiichi nuclear accident, Japan Atomic Energy Agency (JAEA) was chosen by the national government to conduct decontamination pilot projects at selected sites in Fukushima prefecture. Despite tight boundary conditions in terms of timescale and resources, the projects served their primary purpose to develop a knowledge base to support more effective planning and implementation of stepwise regional remediation of the evacuated zone. A range of established, modified and newly developed techniques were tested under realistic field conditions and their performance characteristics were determined. The results of the project can be summarized in terms of site characterization, cleanup and waste management.
A range of options were investigated to reduce the volumes of waste produced and to ensure that decontamination water could be cleaned to the extent that it could be discharged to normal drainage. Resultant solid wastes were packaged in standard flexible containers, labelled and stored at the remediation site (temporary storage until central interim storage becomes available). The designs of such temporary storage facilities were tailored to available sites, but all designs included measures to ensure mechanical stability (e.g., filling void spaces between containers with sand, graded cover with soil) and prevent releases to groundwater (impermeable base and cap, gravity flow drainage including radiation monitors and catch tanks). Storage site monitoring was also needed to check that storage structures would not be perturbed by external events that could include typhoons, heavy snowfalls, freeze/thaw cycles and earthquakes.
The Conceptual Solutions Concerning Decomissioning and Dismantling of Russian Civil Nuclear Powered Ships
Authors: Kulikov Konstantin N. , Nizamutdinov Rinat A. , Abramov Andrey N. 
Published: February 2014

From 1959 up to 1991 nine civil nuclear powered ships were built in Russia: eight ice-breakers and one lash lighter carrier (cargo ship). At the present time three of them were taking out of service: ice-breaker “Lenin” is decommissioned as a museum and is set for storage in the port of Murmansk, nuclear ice-breakers “Arktika” and “Sibir” are berthing. The ice-breakers carrying radwastes appear to be a possible source of radiation contamination of Murmansk region and Kola Bay because the ship long-term storage afloat has the negative effect on hull’s structures. As the result of this under the auspices of the Federal Targeted Program “Nuclear and Radiation Safety of Russia for 2008 and the period until 2015” the conception and projects of decommissioning of nuclear-powered ships are developed by the State corporation Rosatom with the involvement of companies of United Shipbuilding Corporation. In developing the principal provisions of conception of decommissioning and dismantling of ice-breakers the technical and economic assessment of dismantling options in shiprepairing enterprises of North-West of Russia was performed.
The paper contains description of options, research procedure, analysis of options of decommissioning and dismantling of nuclear ice-breakers, taking into account the principle of optimization of potential radioactive effect to personnel, human population and environment. The report’s conclusions contain the recommendations for selection of option for development of nuclear ice-breaker decommissioning and dismantling projects.
Preliminary Waste Management Plan of European Spallation Source
Authors: Ene Daniela 
Published: February 2014

The European Spallation Source (ESS) is the European common effort in designing and building a next generation large-scale user facility for studies of the structure and dynamics of materials. The proposed schematic layout of the ESS facility is based on a linear driver (linac) directing the proton beam (5 MW of 2.5 GeV) of 2.8 ms long pulses with a 20 Hz on a tungsten target where neutrons are produced via spallation reactions. Further the neutrons will be moderated to thermal and subthermal energies in a couple of moderators placed around the target. The moderators feed 22 beamlines guiding the neutrons to the scattering instruments, mainly for neutron scattering research, as has been previously mentioned.
The objective of this work is to develop a waste management plan for ESS facility. In this respect two important aspects are analyzed. First the present status of the problem is outlined as follow. Estimate types and quantities of waste that the ESS project will generate at different stages: commission, operation, decommissioning were derived using: i) precise Monte Carlo calculations ii) scaling the activity from the operation experience of the existing spallation source installations for waste such it is difficult to predict level of activation or for components of the facility in stage of the pre-conceptual model.
Associated waste treatment/conditioning options and final disposal route were further analyzed in order to define the waste type and packet descriptions in agreement with Swedish regulations and policy. It was found that the compilation of completely new waste type descriptions for qualification of the ESS waste for disposal will be necessary. Particular attention was devoted to “problematic waste” as Beryllium reflector, C-14 from graphite used as core zone of the beam-dump and collimators or waste arising from the purification systems of both Helium and water cooling circuits. Management of waste on ESS site: collection/segregation systems, characterization system, storage options, is also described.
In the second step, the acquired information is used for planning and implementing actions involving all participants (ESS, treatment facility operator, disposal operator, regulatory body and other authorized authorities).
Ultratrace Determination of Strontium-90 in Environmental Soil Samples From Qatar by Collision/Reaction Cell-Inductively Coupled Plasma Mass Spectrometry (CRC-ICP-MS/MS)
Authors: Al-Meer S. H. , Amr M. A. , Helal A. I. , Al-Kinani A. T. 
Published: February 2014

Because of the very low level of 90Sr in the environmental soil samples and its determination by beta counting may take several weeks, we developed a procedure for ultratrace determination of 90Sr using collision reaction cell-inductively coupled plasma tandem mass spectrometry (CRC-ICP-MS/MS, Agilent 8800).
Soil samples were dried at 105 °C and then heated in a furnace at 550 °C to remove any organics present. 500 g of each soil samples were aliquoted into 2000 ml glass beakers. Each Soils samples were soaked in 2 ppm Sr solution carrier to allow determination of chemical yield. The solid to liquid ratio was 1:1. Finally the soil samples were dried at 105 °C. Five hundred milliliters concentrated nitric acid and 250 ml hydrochloric acid volumes were added on 500 g soil samples. The samples were digested on hot plate at 80 °C to prevent spraying with continuous manual mixing. The leachate solution was separated. The solids were rinsed with 500 ml deionized water, warmed on a hot plate and the leachate plus previous leachate were filtered and the total volume was reduced to 500 ml by evaporation. Final leachate volume was transferred to a centrifuge tubes. The centrifuge tubes were centrifuged at 3,500 rpm for 10 min. The leachate was transferred to a 1 L beaker and heated on a hot plate to evaporate the leachate to dryness. The reside was re-dissolved in 100 ml of 2% HNO3 and reduced by evaporation to 10 mL. The solution was measured directly by CRC-ICP-MS/MS by setting the first quadruple analyzer to m/z 90 and introducing oxygen gas into the reaction cell for elimination isobar interference from zirconium-90. The method was validated by measurements of standard reference materials and applied on environmental soil samples. The overall time requirement for the measurement of strontium-90 by CRC-ICP-MS/MS is 2 days, significantly shorter than any radioanalytical protocol currently available.
Treatment of Irradiated Graphite From French Bugey Reactor
Authors: Stevens Howard , Laurent Gerard 
Published: February 2014

In 2008, following the general French plan for nuclear waste management, Électricité de France attempted to find for irradiated graphite an alternative solution to direct storage at the low-activity long-life storage center in France managed by the national agency for wastes (ANDRA). EDF management requested that its engineering arm, EDF CIDEN, study the graphite treatment alternatives to direct storage.
In mid-2008, this study revealed the potential advantage for EDF to use a steam reforming process known as THermal Organic Reduction, “THOR” (owned by Studsvik, Inc., USA), to treat or destroy the graphite matrix and limit the quantity of secondary waste to be stored. In late 2009, EDF began a test program with Studsvik to determine if the THOR steam reforming process could be used to destroy the graphite. The program also sought to determine if the graphite could be treated to release the bulk of activity while minimizing the gasification of the bulk mass of the graphite.
In October 2009, tests with non-irradiated graphite were completed and demonstrated destruction of a graphite matrix by the THOR process at satisfactory rates. After gasifying the graphite, focus shifted to the effect of roasting graphite at high temperatures in inert gases with low concentrations of oxidizing gases to preferentially remove volatile radionuclides while minimizing the graphite mass loss to 5%. A radioactive graphite sleeve was imported from France to the US for these tests. Completed in April 2010, ‘Phase I’ of testing showed that the process removed >99% of H-3 and 46% of C-14 with <6% mass loss. Completed in September 2011, ‘Phase II’ testing achieved increased removals as high as 80% C-14. During Phase II, it was also discovered that roasting in a reducing atmosphere helped to limit the oxidation of the graphite.
Future work seeks to explore the effects of reducing gases to limit the bulk oxidation of graphite. If the graphite could be decontaminated of long-lived radionuclides up to 95% for C-14 while minimizing mass loss to <5%, this would minimize the volume of any secondary waste streams and potentially lower the waste class of the larger bulk of graphite. Alternatively, if up to 95% decontamination of C-14 is achieved, the graphite may be completely gasified which could result in lower disposal volumes.
The Long-Term Safety and Performance Analyses of the Surface Disposal Facility for the Belgian Category A Waste at Dessel
Authors: Cool Wim , Vermariën Elise , Wacquier William , Perko Janez 
Published: February 2014

ONDRAF/NIRAS, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials, and its partners have developed long-term safety and performance analyses in the framework of the license application for a surface disposal facility for low level radioactive waste (category A waste) at Dessel, Belgium.
This paper focusses on the methodology of the safety assessments and on key results from the application of this methodology.
An overview is given (1) of the performance analyses for the containment safety function of the disposal system and (2) of the radiological impact analyses confirming that radiological impacts are below applicable reference values and constraints and leading to radiological criteria for the waste and the facility. In this discussion, multiple indicators for performance and safety are used to illustrate the multi-faceted nature of long-term performance and safety of the surface disposal.
This contributes to the multiple lines of reasoning for confidence building that a positive decision to proceed to the next stage of construction is justified.
Know-How and Experience of the Decommissioning of the Eurochemic Reprocessing Plant
Authors: Notelteirs W. , Ooms B. , Huys T. 
Published: February 2014

Belgoprocess started the industrial decommissioning of the main process building of the former Eurochemic reprocessing plant in 1987. In 2008 the eastern part was torn down and in 2010 the central part of the building was removed. The demolition of the third and last part is planned in 2014. At that point the area will be restored to brown field conditions.
During the 25 years of decommissioning the Eurochemic installations a lot of effort has been invested in improving the efficiency. The Belgian decommissioning policy is based on maximizing effort for decontamination with the goal of unconditional release and minimization of nuclear waste. As a result a number of installations have been implemented that reduce the processing costs, increase the output of decontamination and increase the speed of release measurements.
• The abrasive blasting installation is used to remove a thin layer from metal components and concrete objects.
• The concrete crushing and sampling installation is used as the final measurement before unconditional release of concrete debris.
• The concrete spec installation is used to separate the contaminated part from the non contaminated part of concrete debris.
As the end of the decommissioning project is approaching the time is right to look back and evaluate the course of the project. Looking back what were the key decisions during the project, what were important milestones and what could or should have been done differently knowing what we know now.
Characterisation and Evaluation of Wastes for Treatment in the Batch Pyrolysis Plant in Studsvik, Sweden
Authors: Lindberg Maria , Österberg Carl , Vernersson Thomas 
Published: February 2014

The new batch pyrolysis plant in Studsvik is built primarily for treatment of uranium containing dry active waste, “DAW”. Several other waste types have been identified that are considered or assumed suitable for treatment in the pyrolysis plant because of the possibility to carefully control the atmosphere and temperature of the thermal treatment. These waste types must be characterised and an evaluation must be made with a BAT (Best Available Technology) perspective. Studsvik have performed or plan to perform lab scale pyrolysis tests on a number of different waste types. These include
- Pyrophoric materials (uranium shavings)
- Uranium chemicals that must be oxidised prior to being deposited in repository
- Sludges and oil soaks (this category includes NORM-materials)
- Ion exchange resins (both “free” and solidified/stabilised)
- Bitumen solidified waste
Methodology and assessment criteria for various waste types, together with results obtained for the lab scale tests that have been performed, are described.
Remediation of Uranium In-Situ Leaching Area at Stráž pod Ralskem, Czech Republic
Authors: Vokál Vojtěch , Mužák Jiří , Ekert Vladimír 
Published: February 2014

A large-scale development in exploration and production of uranium ores in the Czech Republic was done in the 2nd half of the 20th century. Many uranium deposits were discovered in the territory of the Czech Republic. One of the most considerable deposits in the Czech Republic is the site Hamr na Jezere – Straz pod Ralskem where both mining methods — the underground mining and the acidic in-situ leaching — were used. The extensive production of uranium led to widespread environmental impacts and contamination of ground waters. Over the period of “chemical” leaching of uranium (ca. 32 years), a total of more than 4 million tons of sulphuric acid and other chemicals have been injected into the ground. Most of the products (approx. 99.5 %) of the acids reactions with the rocks are located in the Cenomanian aquifer. The contamination of Cenomanian aquifer covers the area larger then 27 km2. The influenced volume of groundwater is more than 380 million m3. The total amount of dissolved SO42− is about 3.6 million tons. After 1990 a large-scale environmental program was established and the Czech government decided to liquidate the ISL Mine and start the remediation in 1996.
The remediation consists of contaminated groundwater pumping, removing of the contaminants and discharging or reinjection of treated water. Nowadays four main remedial technological installations with sufficient capacity for reaching of the target values of remedial parameters in 2037 are used — the “Station for Acid Solutions Liquidation No. One”, the “Mother liquor reprocessing” station, the “Neutralization and Decontamination Station NDS 6” and the “Neutralization and Decontamination Station NDS 10”. It is expected that the amount of withdrawn contaminants will vary from 80 000 to 120 000 tons per year. Total costs of all remediation activities are expected to be in excess of 2 billion EUR.
Application of PHADEC Method for the Decontamination of Radioactive Steam Piping Components
Authors: Lo Frano R. , Pilo F. , Aquaro D. 
Published: February 2014

The dismantling of nuclear plants is a complex activity that originates often a large quantity of radioactive contaminated residue.
In this paper the attention was focused on the PHADEC (PHosphoric Acid DEContamination) plant adopted for the clearance of Caorso NPP (in Italy) metallic systems and components contaminated by Co60 (produced by the neutron capture in the iron materials), like the main steam lines, moisture separator of the turbine buildings, etc.
The PHADEC plant consists in a chemical off line treatment: the crud, deposited along the steam piping during life plant as an example, is removed by means of acid attacks in ponds coupled to a high pressure water washing.
Due to the fact that the removed contaminated layers, essentially, iron oxides of various chemical composition, depend on components geometry, type of contamination and time of treatment in the PHADEC plant, it becomes of meaningful importance to suggest a procedure capable to improve the control of the PHADEC process parameters.
This study aimed thus at the prediction and optimization of the mentioned treatment time in order to improve the efficiency of the plant itself and to achieve, in turn, the minimization of produced wastes. To the purpose an experimental campaign was carried out by analysing several samples, i.e. taken along the main steam piping line.
Smear tests as well as metallographic analyses were carried out in order to determine respectively the radioactivity distribution and the crud composition on the inner surface of the components. Moreover the radioactivity in the crud thickness was measured. These values allowed finally to correlate the residence time in the acid attack ponds to the level of the achieved decontamination.
Destructive Testings: Dry Drilling Operations With TruPro System to Collect Samples in a Powder Form, From Two Hulls Containing Immobilized Wastes in a Hydraulic Binder
Authors: Pombet Denis , Desnoyers Yvon , Charters Grant , Aggarwal Sue 
Published: February 2014

The TruPro® process enables to collect a significant number of samples to characterize radiological materials. This innovative and alternative technique is experimented for the ANDRA quality-control inspection of cemented packages. It proves to be quicker and more prolific than the current methodology.
Using classical statistics and geostatistics approaches, the physical and radiological characteristics of two hulls containing immobilized wastes (sludges or concentrates) in a hydraulic binder are assessed in this paper. The waste homogeneity is also evaluated in comparison to ANDRA criterion. Sensibility to sample size (support effect), presence of extreme values, acceptable deviation rate and minimum number of data are discussed.
The final objectives are to check the homogeneity of the two characterized radwaste packages and also to validate and reinforce this alternative characterization methodology.