bluuebunny/crossref_metadata_2025 huggingface dataset:
Radionuclide Source Term for the ASSE Salt Mine: Geochemical Assessment for the Use of Magnesium(II) Based Backfill Material
Authors: Metz Volker , Schu¨ßler Wolfram , Vejmelka Peter , Lu¨tzenkirchen Johannes , Kienzler Bernhard
Published: March 2009

Low- and intermediate-level radioactive waste was emplaced in the Asse salt mine. Application of backfill materials to provide mechanical stability, plugging of openings and chemical buffering is discussed for closure of the mine. If no remedial action is carried out in the emplacement rooms, degradation of the organic matter cannot be excluded which may result in acidification of the geochemical environment or actinides may be complexed by HCO3− / CO32−. Selection of backfill materials is based on geochemical modeling taking into account the corrosion of the cemented LLW / ILW and degradation of organic waste components. In the present study the evolving geochemical milieu and respective solubilities of Am, Np, Pu, U, Tc, Sr, Cs and I were modeled. The solubilities and the radionuclide inventory define the geochemical source term for each emplacement room. Laboratory experiments were undertaken to verify the modeling predictions. The modeling leads to the conclusion that Portland cement, a Mg(OH)2-based material and crushed salt should be used in different combinations as backfill materials. According to the modeling predictions, the Mg(OH)2-based backfill material provides a long-term stable geochemical environment and scavenges CO2 by formation of (Ca,Mg)-carbonates. The source term calculations were carried out for the presence of the backfill material in the emplacement rooms. The inventory of Pu and U in the waste is sufficient to allow these actinides to reach solubility limits. Maximum concentrations of the other radionuclides are limited by their inventories, which are too low to reach solubility limits. Regarding corrosion of cemented waste forms and the backfill material in MgCl2 rich brines, experimental data and calculated results agree well. Experimental results demonstrate the ability of the Mg(OH)2-based material to buffer the pH and to limit the HCO3− / CO32− concentration in the brines. Based on both the modeling and the experimental results, it is highly recommended to apply the Mg(OH)2-based backfill material in the Asse salt mine.
An Integrated Safety Assessment System for Near-Surface Disposal of Low- and Intermediate-Level Radioactive Waste in Korea
Authors: Park JooWan , Kim Chang-Lak , Park Jin Beak , Lee Eun Yong , Lee Youn Myoung , Kang Chul Hyung , Zhou Wei , Kozak Matthew W.
Published: March 2009

An integrated safety assessment system to be used for evaluation of near-surface disposal concept has been developed within the framework of safety assessment methodology taken for low- and intermediate-level radioactive waste disposal in Korea. It is to provide an evaluation of the safety of the disposal system in a clear, comprehensive and well-documented manner, and to integrate the results into a defensible package showing reasonable assurance of compliance with regulatory requirements for licensing application. This system is made up of two key components, a system-level safety assessment code and an input database/quality assurance module for safety assessment, which both are interfaced with each other.
Site Specific Sorption Data for the Asse Salt Mine
Authors: Lu¨tzenkirchen Johannes , Vejmelka Peter , Metz Volker , Kienzler Bernhard
Published: March 2009

The focus of this contribution is to report sorption data for Asse salt mine. The purpose of the experimental program was to obtain data based on site and scenario specific considerations for use in terms of Kd-values for the emplacement rooms in performance assessment calculations. A range of materials, solutions and radionuclides was studied, including the variation of solid to liquid ratio and the effect of complexing agents with EDTA chosen as the representative. Observed sorption kinetics is compared to observations in model systems, where a fast and a subsequent slow sorption step occur. Calculations on a model system are performed assuming a similar sampling procedure as in the experimental program. It is concluded that the observed kinetics can at least partially be explained by the experimental procedure. Problems in obtaining and applying the site specific sorption coefficients in performance assessment are addressed. Special focus is in this respect on the need to develop (i) more mechanistic understanding of sorption processes in complex systems (ii) more complete sorption data sets and (iii) pragmatic sorption models capable of handling high salt concentrations.
Development of Induction Melting System With Active Insulator for Radioactive Solid Waste
Authors: Matsumoto Takeshi , Kawaguchi Ichiro , Chishiro Ryo , Nishimura Motohiko , Kanai Kazuki , Yamazaki Seiichiro , Yokosawa Minoru , Mihara Shigeru
Published: March 2009

The melting treatment is suitable for reducing the volume of the wastes because of the high volume reduction ratio (the volume reduction ratio is the initial volume to the volume after treatment). We have developed a new high-frequency induction melting system for the low-level radioactive miscellaneous solid wastes. The non-conductive ceramic canister and a heat loss compensator (Active insulator) were used in this new system. It is difficult to melt a large amount of the non-metallic materials with the canister. We solved this problem by using the active insulator, which was made of the conductive material. Melting performance confirmation tests were performed in the medium-scale melting system. Based on the result of the medium-scale melting test and multi-dimensional thermal-hydraulic analysis, the full-scale melting system was designed and constructed. We performed the melting tests using the full-scale melting system. the volume ratio of the non-metallic wastes at the re-solidification was more than 70%. Behavior of nuclides was also investigated with non-radioactive Co and Cs tracers. The residual ratio of Co and Cs were 97%, 58%, respectively.
Post-Leaching Studies of Defect Rodlets
Authors: Wegen D. H. , Papaioannou D. , Glatz J.-P. , Bottomly P. D. W. , Amme M. , van Winckel S.
Published: March 2009

UO2 and MOX fuel rodlets are examined by optical microscopy after 5 years leaching in water to correlate the irradiation history of the fuels and of the fuel behaviour and with the release data for fission products and actinides. Metallographic observation revealed a strong intergranular attack of the MOX fuel in contrast to UO2. The latter effect could be responsible for the particularly high releases from the MOX fuel (upto 12% of the Cs inventory), compared to the UO2 fuel (only 0.6% of the Cs inventory). The large inventory of volatile fission products at grain boundaries can be explained by the high linear power rating (350 W/cm) with extremely high central temperatures in the MOX fuel during the irradiation.
Status of the NRC Decommissioning Program
Authors: Orlando Dominick , Craig Claudia M.
Published: March 2009

On July 21, 1997, the U.S. Nuclear Regulatory Commission (NRC) published the final rule on Radiological Criteria for License Termination (the License Termination Rule or LTR) as Subpart E to 10 CFR Part 20. NRC regulations require that materials licensees submit Decommissioning Plans to support the decommissioning of its facility if it is required by license condition, or if the procedures and activities necessary to carry out the decommissioning have not been approved by NRC and these procedures could increase the potential health and safety impacts to the workers or the public. NRC regulations also require that reactor licensees submit Post-shutdown Decommissioning Activities Reports and License Termination Plans to support the decommissioning of nuclear power facilities. This paper provides an update on the status of the NRC’s decommissioning program. It discusses the staff’s current efforts to streamline the decommissioning process, current issues being faced in the decommissioning program, such as partial site release and restricted release of sites, as well as the status of the decommissioning of complex sites and those listed in the Site Decommissioning Management Plan. The paper discusses the status of permanently shut-down commercial power reactors and the transfer of complex decommissioning sites and sites listed on the SDMP to Agreement States. Finally the paper provides an update of the status of various tools and guidance the NRC is developing to assist licensees during decommissioning, including an effort to consolidate and risk-inform decommissioning guidance.
Continuous on Site Monitoring of VOCs in Water Sources
Authors: Linenberg Amos
Published: March 2009

Public concern over the state of the environment is at an all-time high and rightfully so! Industry practice, recent government regulations and lax enforcement have allowed frightening practices to continue for too long. Industries must, by law, monitor the level of toxins they discharge into the environment. Collecting samples and sending them to an off-site laboratory for analysis is the normal practice to comply with present regulations. This protocol is not only a time-consuming and costly exercise, but does not provide continuous information for alerting the public and the authorities of a potential disaster. A water treatment plant is obligated to test water for volatile organic compounds (VOCs) at frequencies that vary from a few times per week to once every three months. Authorities may test finished water as infrequently as once per year. This means that drinking water supplied or waste water discharged, between discrete analyses, is of unknown purity. Since September 11, 2001, an additional dimension, “water security”, has been added to the need for instantaneous analysis. Protection and preservation of water sources such as reservoirs, lakes and rivers from intentional and unintentional contamination, have become an issue, which involves homeland security. Here again, obtaining a fast and accurate response at all times is extremely important. Sentex Systems, Inc., which has specialized for several years in on site VOCs analysis, has developed a system by which online continuous analysis of VOCs in water is available. This system, which is based on the principle of in-situ purge and trap Gas Chromatography, will detect and analyze VOCs on site for most industrial and environmental applications, without the need for sample preparation, such as filtration. The system can continuously monitor process streams so that at any given time plant management knows what the VOC content is. The system, called the SituProbe, is already being used successfully in various industrial plants.
Design, Loading, Transport and Storage Experience of CASTOR® Casks for Vitrified High Level Waste
Authors: Voßnacke Andre´ , Graf Wilhelm , Hu¨ggenberg Roland , Gisbertz Astrid
Published: March 2009

The revised German Atomic Act together with the Agreement between the German Government and the German Utilities of June 11, 2001 form new boundary conditions that considerably influence spent fuel strategies by stipulation of lifetime limitations to nuclear power plants and termination of reprocessing. The contractually agreed return of reprocessing residues comprises some 156 casks containing vitrified highly active waste, the so-called HAW or glass canisters, coming form irradiated nuclear fuel assemblies to be shipped from COGEMA, France and BNFL, UK to Germany presumably until 2011. Several hundred casks with compacted residues and other waste will follow. The transports are scheduled presumably beyond 2020. The central interim storage facilities in Ahaus and Gorleben, formerly intended to accumulate up to 8,000 t of heavy metal (HM) of spent fuel from German nuclear power plants, offer sufficient capacity to receive the totality of residues to be returned from reprocessing abroad. GNB has developed, tested, licensed, fabricated, loaded, transported and stored a large number of casks for spent fuel and is one of the world leaders for delivering spent fuel and high level waste casks. Long-term intermediate storage of spent fuel is carried out under dry conditions using these casks that are licensed for transport as well as for storage. Standardized high performance casks such as the types CASTOR® HAW 20/28 CG, CASTOR® V/19 and CASTOR® V/52 meet the needs of most nuclear power plants in Germany. Up to now GNS has co-ordinated the loading and transport of 27 casks loaded with 28 canisters each from COGEMA back to Germany for storage in Gorleben for up to 40 years. In all but one case the cask type CASTOR® HAW 20/28 CG has been used.
L/ILW Waste Characterisation by the ENEA Multi-Technique Gamma System SRWGA
Authors: Troiani Francesco , Cherubini Nadia , Dodaro Alessandro , Frazzoli Franco Vittorio , Remetti Romolo
Published: March 2009

The SEA Radioactive Waste Gamma Analyser (SRWGA), is the gamma assay system of the ENEA Laboratory for L/ILW Waste Characterisation, which started operating in 1996 as a simple gamma scanner and has been under a continuous improvement process to became a multi-techniques system. The SRWGA is designed for the assay of radioactive waste drums containing gamma emitting nuclides. The system operates with an XtRa (extended range) Ge coaxial detector. This is liquid nitrogen colled, and shielded by means of a lead cylinder with two collimation windows; one mixed transmission source is provided. The system allows the application of four different measuring techniques, each one with its peculiar field of application, depending on waste characteristics or measuring time: Open Geometry, Segmented Gamma Scanning (with multi-energy transmission correction), Angular Scanning and, recently, Low Resolution Emission and Transmission Tomography, as the SRWGA is now endowed with new mechanical motions systems for tomographic capabilities. Tomographic reconstructions are obtained by means of a backprojection filtered by convolution methods (for transmission tomography) and Best Likelihood Maximisation (for emission tomography). The information obtained with transmission and emission tomography allows the localisation of matrix dishomogeneities and hot spots, carrying out a strong reduction of total activity uncertainties. This work presents the experimental results obtained using certified γ sources located in known matrices.
Decommissioning Strategy for Chernobyl NPP
Authors: Baybuzenko T. Yu , Krasnjansky G. G. , Kuchinsky V. K. , Litvinsky L. L. , Lobach Yu. N. , Purtov O. A. , Rudko V. M. , Saliy V. P. , Seyda V. A. , Filatov V. I. , Shenderovich V. Ya. , Scherbin V. N. , Yakovlev A. A.
Published: March 2009

At present, Chernobyl NPP is under decommissioning after the final shutdown of Unit 3 at the end of 2000. The decommissioning activity is being carried out on the basis of the “Decommissioning Conception for ChNPP” approved in 1992. In accordance with the acting legislation, this concept should be reconsidered in 2002 at the latest. Presently, a new version of the concept is being developed. This document is based on the decommissioning strategy, which foresees the long-term safe storage of reactors’ cores (up to 100 years) and coolant circuit facilities (up to 50 years) within the existing building constructions together with dismantling of auxiliary equipment. Key decisions are the following: 1) the final goal of decommissioning is the condition described as a “brown field site”; 2) decommissioning involves measures for decontamination and dismantling of the contaminated structures; 3) structures having contamination levels at or below free release are considered as “conditionally clear” and for them the decommissioning goals have been achieved; 4) it is anticipated that the dismantling of building constructions and refinement of site will be considered in the framework of Chernobyl Exclusion Zone rehabilitation. This paper describes the decommissioning process for materials and reactor facilities and presents an overview of the decommissioning program activities.
Immobilisation of Contaminated DEHPA Waste in Portland Cement
Authors: Blagojevic Ned , Vance Lou , Aldridge Laurie , Malik Syed A.
Published: March 2009

The immobilisation of organic liquids in cement products can often be difficult when attempts are made to achieve high waste loading. In this work, diethylhexyl phosphoric acid (DEHPA) contaminated with minor amounts of U (1400 ppm), Th (100 ppm) and rare earth elements (17,900 ppm) arising from solvent extraction technology for rare earth extraction from monazite shows promise of immobilisation in ordinary Portland cement (OPC). Waste loadings of up to 50% (v/v) have been achieved at the laboratory scale. The product was allowed to set overnight and had reasonable resistance to leaching after exposure to deionised water (DIW) at 25°C. Centimetre-sized samples released <0.1% of the rare earth and U inventories after exposure to 100 ml of DIW for 7 days. Releases of Ca, Al and Si were comparable with those from DEHPA free OPC. Samples were examined by SEM to determine elemental distribution and assess the porosity. Compressive strengths and detailed leaching behaviour of sample bodies over the temperature range between 25 °C and 50 °C will be presented. Preliminary attempts with geopolymeric materials were less successful than those using cement. The relative merits of immobilisation in cement compared with other possible means of dealing with the contaminated DEHPA are discussed.
Characterization of NORM Contaminated Sites at the Syrian Oilfield: Depth Profiles and Leaching Processes
Authors: Al-Masri M. S. , Aba A. , Al-Hamwi A. , Mukallati H.
Published: March 2009

Production water containing naturally occurring radioactive materials (NORM) has been collected in unlined artificial lagoons for evaporation in some Syrian oilfields. These lagoons have become highly contaminated with NORM and the situation has urged the operating oil companies in Syria to initiate a remediation program in cooperation with the Atomic Energy Commission of Syria. Part of this national remediation program is to characterize the contaminated soil as a preliminary step for disposal of this waste. Depth profiles of radioactivity have been established and found to be variable from one field to another. Factors that influence this distribution have been evaluated and are presented. Laboratory leaching experiments were performed using six 60-cm cores collected from highly contaminated areas in the oil fields. Results show that 226Ra is transferred to deep layers via erosion caused by disposal of production water and some heavy rain water that occurred in the past. This erosion process is mainly affected by the mineralogical compositions of the contaminated soil and the particle size distribution. Gypsum present in the soil has increased transfer of 226Ra from surface layers to deeper layers; water has caused some sink holes (caves) in those soils containing high amount of gypsum. In addition, 226Ra was also determined in different particle size soil samples before and after leaching experiments where small soil particles sizes were found to be moved downwards by water. Radium was only more concentrated in smaller particle sizes than larger ones in those samples containing low concentration of gypsum. In addition, halite content in the upper soil layers has increased the radium specific activity after leaching since it dissolved in water and moved to deeper layers.
Fleetwide Low Level Radwaste Strategy
Authors: Pal Arun C.
Published: March 2009

This project was aimed at mapping a path forward for improved fleet-wide Low Level Radwaste (LLRW) handling. Thus it entailed both solid and liquid wastes and all aspects, i.e., generation, processing and storage or disposal of LLRW. The conclusions are fairly straight forward: minimization of volume generated, maximization of volume reduction of the already generated waste, disposal as soon as possible, fleet-wide standardization of processes and procedures and consolidation of contracts for the economies of scale.
The Role of the Waste Isolation Pilot Plant in the Cleanup of the U.S. Nuclear Weapons Complex
Authors: Smith Lynne K. , Bisesi Mary L.
Published: March 2009

As a result of nuclear weapons production, the United States of America produced significant quantities of transuranic waste, which consists of clothing, tools, rags, residues, debris and other items contaminated with small amounts of radioactive man-made elements — mostly plutonium — with an atomic number greater than that of uranium. Transuranic waste began accumulating in the 1940s and continued through the Cold War era. Today, most transuranic waste is stored at weapons production sites across the United States. In 1957, the National Academy of Sciences concluded that the most promising disposal option for radioactive wastes was disposal in deep geologic repositories situated in the salt formations. After nearly a decade of study, the United States Department of Energy decided in January 1981 to proceed with construction of the Waste Isolation Pilot Plant (WIPP) at a site 41.6 km (26 miles) southest of Carlsbad, New Mexico. After years of study, construction, and permitting, the WIPP facility became operational in early 1999. As the United States continues to clean up and close its former nuclear weapon facilities, the operation of WIPP will continue into the next several decades. This paper will provide on overview of the history, regulatory, and public process to permit a radioactive repository for disposal of transuranic wastes and the process to ensure its long-term operation in a safe and environmentally compliant manner.
Clearance Levels in Germany: How Do We Know They Are Right?
Authors: Thierfeldt Stefan
Published: March 2009

Clearance levels (CL) in Germany have been implemented in the Radiation Protection Ordinance (RPO / Strahlenschutzverordnung StrlSchV) of July 2001 which transforms the EURATOM Basic Safety Standards (BSS) into national legislation. Separate clearance options exist e.g. for metals, building rubble, liquids, buildings, wastes and sites. Meanwhile, the CL have found extensive application at a number of operating nuclear installations in Germany, in particular at those under decommissioning or those which will enter the decommissioning phase soon. The CL are based on extensive studies and have been approved by the competent bodies. Yet there has been and is an ongoing discussion in the interested public whether the basis for the CL is correct. In order to establish a scientific basis for this discussion, a Working Group on Clearance within the German Commission on Radiological Protection (SSK) has been established and studies have been launched by the German Federal Ministry for the Environment, Nature Conservation and Nuclear Safety. Key issues comprise inter alia: • comparison of German and international CL (scope, values, method of their derivation, etc.); • review and update of German CL for clearance for disposal as a consequence of changes in the (conventional) waste law; • leaching behaviour of relevant radionuclides from cleared building rubble, especially for the long-term behaviour of radionuclides relevant in contaminated concrete of NPPs; • the possibility of multiple exposure for a single individual from various clearance options; • evaluation of the collective dose which may be caused by clearance in Germany and comparison with the reference value of 1 manSv/a.
Integrated Approach to Remediatiion of Multiple Uranium Mill Tailing Sites for the US DOE in the Western United States
Authors: Van Dyke Bill , Dabrowski Tom
Published: March 2009

This paper provides a case history of a highly successful approach that was developed and implemented for the U.S. Department of Energy (DOE) for the cleanup and remediation of a large and diverse population of uranium mill tailings sites located in the Western United States. The paper addresses the key management challenges and lessons learned from the largest DOE Environmental Management Clean-up Project (in terms of number of individual clean-up sites) undertaken in the United States. From 1986 to 1996, the Department of Energy’s Grand Junction Projects Office (GJPO) completed approximately 4600 individual remedial action site cleanup projects for large- and small-scale properties, and sites contaminated with residual hazardous and radioactive materials from former uranium mining and milling activities. These projects, with a total value of $597 million, involved site characterization, remedial design, waste removal, cleanup verification, transportation, and disposal of nearly 2.7 million cubic yards of low-level and mixed low-level waste. The project scope included remedial action at 4,200 sites in Grand Junction, Colorado, and Edgemont, South Dakota; 412 sites in Monticello, Utah; and, 44 sites in Denver, Colorado. The projects ranged in size and complexity from the multi-year Monticello Millsite Remedial Action Project, which involved investigations, characterization, remedial design, and remedial action at this uranium millsite along with design of a 2.5 million cubic yard disposal cell, to the remediation and reconstruction of thousands of smaller commercial and residential properties throughout the Southwestern United States. Because these projects involved remedial action at a variety of commercial facilities, businesses, churches, schools and personal residences, and the transportation of the waste through towns and communities, an extensive public involvement program was the cornerstone of an effort to promote stakeholder understanding and acceptance. The Project established a DOE model for rapid, economical, and effective remedial action. During the ten years of the contract, the management operations contractor (Duratek) met all project milestones on schedule and under budget, with no cost growth from the original scope. By streamlining remediation schedules and techniques, ensuring effective stakeholder communications, and transferring lessons learned from one project to the next, the contractor achieved maximum efficiency and the lowest remediation costs of any similar DOE environmental programs at the time.
The Australian National Radioactive Waste Repository: Environmental Impact Statement and Radiological Risk Assessment
Authors: Smith-Briggs Jane , Wells Dave , Green Tommy , Baker Andy , Kelly Martin , Cummings Richard
Published: March 2009

The Environmental Impact Statement (EIS) for the proposed Australian National Repository for low and short-lived intermediate level radioactive waste was submitted to Environment Australia for approval in the summer of 2002 and has subsequently undergone a consultancy phase with comments sought from all relevant stakeholders. The consultancy period is now closed and responses to the comments have been prepared. This paper describes some of the issues relevant to determining the radiological risk associated with the repository to meet the requirements of the EIS. These include a brief description of the three proposed sites, a description of the proposed trench design, an analysis of the radioactive waste inventory, the proposed approach to developing waste acceptance criteria (WAC) and the approach taken to determine radiological risks during the post-institutional control phase. The three potential sites for the repository are located near the Australian Department of Defence site at Woomera, South Australia. One site is inside the Defense site and two are located nearby, but outside of the site perimeter. All have very similar, but not identical, topographical, geological and hydrogeological characteristics. A very simple trench design has been proposed 15 m deep and with 5 m of cover. One possible variant may be the construction of deeper borehole type vaults to dispose of the more active radioactive sources. A breakdown of the current and predicted future inventory will be presented. The current wastes are dominated in terms of volume by some contaminated soils, resulting from experiments to extract U and Th, and by the operational wastes from the HIFAR research reactor at ANSTO. A significant proportion of the radionuclide inventory is associated with small volumes of sources held by industry, medical, research and defence organisations. The proposed WAC will be described. These are based on the current Australian guidelines and best international practice. The preliminary radiological risk assessment considered the post-institutional control phase in detail with some 12 scenarios being assessed. These include the impact of potential climate change in the region. The results from the risk assessment will be presented and discussed. The assessment work is continuing and will support the license application for construction and operation of the site. Please note that this is not the final assessment for the licence application.
Application of the Dry Polyfunctional Additives at Radioactive Waste Cementation
Authors: Varlakov Andrey P. , Efimov Konstantin M. , Tchernonojkine Valeri N. , Barinov Aleksandr S. , Gorbunova Olga A.
Published: March 2009

One of the known methods directed to improving of the technological cementation process, the increasing quality of a cement compound and degree of radioactive waste incorporating into a final product, is use of the various additives to a cement compound. At present there are technological processes where one or two additives in a dry loose or liquid condition in quantity of 1–10% are used. The application of these additives is directed, as a rule, to improving of one or two properties of a cement compound. The magnification of a quantity of the additives and use of them in a different aggregation state is connected with rise in the cost of the technological process. At Institute of Ecology and Technology Problems and Moscow SIA “Radon” the polyfunctional modifying additives representing dry mixtures of original macro- and microadditives to cement have been developed. The polyfunctional additive is introduced by traditional, reliable and inexpensive equipment directly into liquid radioactive waste and intermixed together with the rest of cement. The quantity of additive varies from 5 up to 20% of cement weight. The additives considerably improve all regulated properties of a cement compound (compressive strength, radionuclides leaching, frost resistance, biological resistance, etc.) and allow modifying the required properties (penetrating ability, viability, disintegration, terms of setting, viscosity etc.). Such additives are used both at cementation of solid radioactive waste and cementation of liquid radioactive waste having a complicated chemical composition, for example, containing simultaneously boric acid, sulphates and great quantity of organic compounds. It is important, that the components of the additives did not change the action in a mixture with other ones. In the report the compositions of the polyfunctional additives developed for various waste and technological processes, their properties and results of practical application are represented.
TN™ 81: A Challenging Design
Authors: Chanzy Yves , Otton Camille
Published: March 2009

With increasing burn-up, reprocessing of spent nuclear fuel yields higher quantities of radionuclides, with a powerful source term and high heat output. Improvement in the vitrification process and environmentally sound thinking have been drivers to reduce the number of transports of vitrified residues (High Level Waste) to interim storage facilities in their owner’s country: this results in higher concentrations of nuclides in the stable glass matrix. The challenge was to create, with almost the same allowable mass and dimensions, a transport/storage casks able to transport glass canisters with this new specifications. Improving the environmental performance of the glass canisters would be of no avail without the corresponding means of transport and storage. This is why COGEMA LOGISTICS introduced the TN™ 81 concept; a dual purpose cask able to handle the most demanding canisters from reprocessing: 56 kW instead of 41 kW, and to shield with efficiency greater gamma and neutron sources, which regulations have been made more stringent regarding the neutron quality factor. The paper will comment the choices made, the drop test campaigns run specifically, and report on the loading of the first TN 81 for KernKraftwerk Go¨sgen (KKG), Switzerland.
Characterization and Decontamination of Contaminated Structures Using a Profiling Technology
Authors: Aggarwal S. , Charters G. , Thacker D.
Published: March 2009

Certain radioisotopes (tritium, radium, cobalt, plutonium, and cesium) can penetrate porous concrete and contaminate the concrete well below the easily measured surface. Certain radioisotopes can penetrate concrete and contaminate the concrete well below the surface. The challenge is to determine the extent of the contamination problem and the magnitude of the problem in a real-time. Currently, concrete core bores are shipped to certified laboratories where the concrete residue is run through a battery of tests to determine the contaminants. The existing core boring operation volatilizes some of the contaminants (like tritium) and oftentimes cross-contaminates the area around the core bore site. The volatilization of the contaminants can lead to airborne problems in the immediate vicinity of the core bore. Cross-contamination can increase the contamination area and thereby increase the amount of waste generated. The goal is to avoid those field activities that could cause this type of release. The concrete profiling technology, TRUPROSM in conjunction with portable radiometric instrumentation produces a profile of radiological or chemical contamination through the material being studied. The data quality, quantity, and representativeness may be used to produce an activity profile from the hot spot surface into the material being sampled. This activity profile may then be expanded to ultimately characterize the facility and expedite waste segregation and facility closure at a reduced cost and risk. Performing a volumetric concrete or metal characterization safer and faster (without lab intervention) is the objective of this characterization technology. This way of determining contamination can save considerable time and money.
Using the HERCULES System to Segregate Room Trash From the LANL Plutonium Facility and CMR Building
Authors: Gruetzmacher Kathleen M. , Bustos Roland M. , Ramsey Susan S. , Myers Steven C. , Gallegos Lucas E.
Published: March 2009

The Los Alamos National Laboratory (LANL) Plutonium Facility (PF-4) and the Chemistry and Metallurgy Research (CMR) building are radiological facilities operated by the Nuclear Materials Technology (NMT) Division. Combined, these facilities generate approximately 262 m3 of low-level waste (LLW) per year of routine room trash from laboratories in radiological controlled areas (RCAs). It has been estimated that 50 to 90% of this low-density room trash is free of radioactive contamination and eligible for inclusion in LANL’s Green is Clean (GIC) program. GIC waste includes all non-regulated waste from RCAs that has been actively segregated as nonradioactive through the use of the waste generator’s acceptable knowledge (AK). The High Efficiency Radiation Counter for Ultimate Low Emission Sensitivity (HERCULES) system is one of several nondestructive assay (NDA) measurement systems used in the GIC program to verify the generator’s AK. This highly sensitive system is optimized to detect very small quantities of common LANL radionuclides, especially isotopes of plutonium, americium, and uranium. A pilot project was conducted in collaboration with NMT Division waste management personnel to determine which RCAs generate the greatest volumes of potential low-density GIC wastes. The HERCULES system was used to screen a portion of the low-density laboratory room trash that NMT shipped to LANL’s low-level waste disposal facility in a six-month period from October, 2002–March, 2003. This consisted of 1350 plastic lined cardboard boxes (0.305m × 0.305m × 0.610m) containing between 2.0–13.5 kg each of room trash. Results of the HERCULES screening indicate that with no active segregation attempts by the generator, 33% of the low-level waste boxes are free of radioactive contamination. An additional 42% of the boxes exhibited very low total activities and could probably become legitimate GIC wastes with active segregation employed during the waste generation process. Thus, the expansion of the GIC program to include low-density laboratory room trash could significantly reduce the total volume of LLW generated by NMT Division.
Being Ahead of the Game: Public Involvement and Community Relations Before and During Environmental Projects
Authors: Roos Emmy , Stahl Lisa P. M.
Published: March 2009

Public involvement or community relations activities have become essential to environmental remediation projects. The key to success for these programs is to identify community and stakeholder concerns and needs early on and address them through an effective outreach program that can result in a win-win outcome for those involved. A three-phase community outreach approach is used to identify and develop proactive community outreach strategies and programs. In the first phase, a community assessment is performed to obtain the input needed to develop an effective community outreach plan. The second phase consists of providing project information and building community involvement at the beginning of site remediation. The third phase consists of continuing and dynamic two-way communication activities during site remediation, based on knowledge gathered and rapport built during Phases 1 and 2. This paper presents this three-phase approach and discusses in more detail how the information obtained from the community assessment can be transformed into a successful community involvement strategy. This paper illustrates this approach with a hypothetical military base example, based on actual projects. The Shaw Group Inc. subsidiary, Shaw Environmental & Infrastructure, Inc. (Shaw), develops a base-wide community involvement plan and begins its implementation in support of remediation activities. The community involvement plan is typically a multi-faceted approach designed to identify the most effective methods to share information between project staff and base stakeholders and to foster community involvement in a remediation project. Not only is public involvement in remediation programs mandated by federal laws, but regulators, elected officials, and military entities have learned through past experiences that, in the long run, it is better on all fronts to inform and include community stakeholders early in and throughout the remediation process. Early information and involvement educates communities about environmental impacts and provides them with opportunities to have input to remediation activities regarding land that may be turned over to the community either in the short- or long-term future.
Characteristics on High Temperature Particulate Clean Up in a Pilot Scale Vitrification Plant
Authors: Park S. C. , Hwang T. W. , Ha J. H. , Kim H. S.
Published: March 2009

Characteristics of high temperature filter system were evaluated and the operation parameters were identified through a series of tests by the ceramic candle filter elements. At the first time of the tests, the trend of the pressure drop across the filter was increased sharply. But due to the growth of dust cake on the surface of filters it became stabilized to a certain level soon. The pressure drop was increased by the face velocity strongly. The permeability of the filter element decreased rapidly in the initial time but because the deposition and it also became stable as time goes on. Dust collection efficiency of the filter was higher than the designed value of 99.9%. Dust cleaning of the filter surface was effective under the back flushing pressure range of 1 to 5 bar. Finally, at the long-term vitrification test, no specific defect or failure were occurred to the ceramic candle filter. Based on the test results, the performance and durability of the ceramic candle filter was proved as sound.
Decommissioning of a Shielded αβγ Pie Facility at Harwell
Authors: Chambers T. N.
Published: March 2009

Building 393.6 at Harwell was constructed and commissioned in the mid 1950’s as a remote handling facility used for the post irradiation examination (PIE) of a variety of radioactive materials. The facility contained a number of lead shielded cells, fume cupboards, gloveboxes and a block of three concrete shielded cells. A particular feature of the facility was a suite of ten lead shielded cells arranged as a cell line with interconnecting shielded tunnel sections, used for PIE of irradiated fuels. Operation of the plant continued until the mid-1990s supporting a variety of commercial programmes for the nuclear industry, including the assessment of fuel performance and metallurgy of a variety of cladding materials. Decommissioning of the facility commenced in 1995 with the objective of removing all radiological and toxic materials, dismantling of containments and removal of the building fabric, to enable release of the land for alternative uses. The key challenges encountered in decommissioning were: • B393.6 was the first Category 1 shielded facility to be decommissioned at the Harwell Site; • The hands on decommissioning team was divested from UKAEA to a contracting organisation part way through the decommissioning; • Development of remote decontamination techniques to ensure operator doses were ALARP; • Remote decontamination of heavily contaminated shielded enclosures to permit disposal of waste to Drigg as Low Level Waste; • Reducing the volume of Intermediate Level Waste into either Low Level Waste or free release material; • Decontamination of large volumes of lead to enable disposal of approximately 95% of lead arisings as free release. The paper will discuss the technical and managerial challenges encountered, and provide details of the solutions adopted in successfully reaching the radiological endpoint.
Two-Phase Flow Patterns During Microchannel Vaporization of CO2 at Near-Critical Pressures
Authors: Pettersen Jostein
Published: March 2009

Carbon dioxide (CO2 / R-744) is receiving renewed interest as a refrigerant, in many cases based on systems with microchannel heat exchangers that have high pressure capability, efficient heat transfer, and compact design. A good understanding of two-phase flow of evaporating CO2 in microchannels is needed to analyze and predict heat transfer. A special test rig was built in order to observe two-phase flow patterns, using a horizontal quartz glass tube with ID 0.98 mm, externally coated by a transparent resistive film. Heat flux was obtained by applying DC power to the film, and flow patterns were recorded at 4000 or 8000 frames per second by a digital video camera. Flow patterns were recorded for temperatures 20°C and 0°C, and for mass flux ranging from 100 to 580 kgm−2s−1. The observations showed a dominance of intermittent (slug) flow at low x, and wavy annular flow with entrainment of droplets at higher x. At high mass flux, the annular/entrained flow pattern could be described as dispersed. The aggravated dryout problem reported from heat transfer experiments at high mass flux could be explained by increased entrainment. Stratified flow was not observed in the tests with heat load. Bubble formation and growth could be observed in the liquid film, and the presence of bubbles gave differences in flow pattern compared to adiabatic flow. The flow pattern observations did not fit generalized maps or transition lines showed in the literature.
Frictional Characteristics of Microchannel Gas Flow
Authors: Bari Abdullahel , Koo Jae-Mo , Jiang Linan , Paidipati Jay , Goodson Kenneth E.
Published: March 2009

The improved rates of heat transfer in microchannel gas flows are promising for the design and development of microfluidic systems. This research focuses on the flow characteristics of air in rectangular micro/minichannels at moderate velocities (∼100 m/sec). The 50.8 mm long channels vary from approximately 266 μm to 1090 μm in hydraulic diameter, and the aspect ratio ranges from 0.1 to 0.75. The value of Re ranged from 250 to 4300, with the intention of studying the transition to turbulence. The friction factor is found to be higher than predicted values for Re < 1400 and lower when Re > 1400 suggesting earlier transition to turbulence.
Channel-to-Channel Pressure Differences in Serpentine Minichannel Flow Systems
Authors: Oosthuizen P. H. , Austin M.
Published: March 2009

One of the inter-connected factors that can lead to failures in the flow plates of PEM fuel cells is the pressure differences that exist between adjacent flow channels. These pressure differences lead to stresses in the channel supports, i.e., the ribs, which can be important in the presence of stresses arising due to other factors such as temperature gradients in the flow plates. In order to investigate the magnitudes of the pressure differences across the supports and the places where the maximum pressure differences occur, the flow and pressure variations in various forms of serpentine channels, these channels having a rectangular cross-sectional shape, have been numerically calculated. The presence of the diffusion layer has been ignored and the flow has been calculated using a commercial finite-element software package using the governing equations written in dimensionless form. Solutions have been obtained for various values of the Reynolds number for each of the flow geometries considered for two channel height-to width ratios (one and three). Except for the flow in the vicinity of the bends in channels, the pressure has been found, as is to be expected, not vary significantly across the channel cross-section. The difference between the center point dimensionless pressure in a given channel with those at the same longitudinal position in the adjacent channels has been determined. The dependence of the highest dimensionless pressure difference between channel on the input parameters has been examined.
Development of Interferometric Temperature Measurement Procedures for Microfluid Flow
Authors: Newport D. , Garvey J. , Whelan M. , Dalton T. , Egan V.
Published: March 2009

In order to understand heat transfer processes at the microscale, detailed temperature measurements are required. This paper begins with a review of the current state-of-the art in fluid temperature measurement at the microscale. At present, fluid temperature profiles are not measured, with verification of predicted heat transfer performance being based on global measurements. The paper describes a potential full-field technique based on micro-interferometry. The accuracy of extracting temperature data from small phase difference intensity maps is discussed, with particular reference to the high levels of signal to noise as would be found in a micro-scale flow. Benchmark optical experiments quantifying the effect of noise on phase evaluation are described and the paper concludes with an outline of the achievable resolution for a given channel length and fluid.
Creating a Successful Corporate Maintenance Council
Authors: Peterson S. Bradley
Published: March 2009

Most large, multi-plant companies have launched corporate “Maintenance Councils” in the past decade. While some have achieved a significant success, most are still floundering for direction and concrete results. What are the differences that make some successful and some status quo?