bluuebunny/crossref_metadata_2025 huggingface dataset:
Optimizing Liquid Waste Treatment Processing in PWRS: Focus on Modeling of the Variation of Ion-Exchange Resins Selectivity Coefficients
Authors: Gressier Fre´de´ric , van der Lee Jan , Schneider He´le`ne , Bachet Martin , Catalette Hubert
Published: June 2009

A bibliographic survey has highlighted the essential role of selectivity on resin efficiency, especially the variation of selectivity coefficients in function of the resin saturation state and the operating conditions. This phenomenon has been experimentally confirmed but is not yet implemented into an ion-exchange model specific for resins. This paper reviews the state of the art in predicting sorption capacity of ion-exchange resins. Different models accounting for ions activities inside the resin phase are available. Moreover, a comparison between the values found in the literature and our results has been done. The results of sorption experiments of cobalt chloride on a strong cationic gel type resin used in French PWRs are presented. The graph describing the variation of selectivity coefficient with respect to cobalt equivalent fraction is drawn. The parameters determined by the analysis of this graph are injected in a new physico-chemical law. Implementation of this model in the chemical speciation simulation code CHESS enables to study the overall effect of this approach for the sorption in a batch.
Application of Clearance Automatic Laser Inspection System to Clearance Measurement of Concrete Waste
Authors: Sasaki Michiya , Ogino Haruyuki , Hattori Takatoshi
Published: June 2009

Recently, the Clearance Automatic Laser Inspection System (CLALIS) has been developed for the clearance measurement of metal scraps. It utilizes three-dimensional (3D) laser scanning, γ-ray measurement and Monte Carlo calculation, and has outstanding detection ability. For the clearance measurement of concrete segments, the effect of background (BG) gamma rays from natural radionuclides in the measurement target, such as K-40 and the radioactive decay products of Th-232 and U-238, should be compensated for to ensure adequate waste management. Since NE102A plastic scintillation detectors are used for γ-ray measurement in CLALIS, it is impossible to distinguish between count rates of natural radionuclides and contaminants on the basis of gamma-ray energy information. To apply CLALIS to the clearance measurement of concrete segments, the original activity evaluation method was improved by adding a new compensation procedure. In this procedure, BG count rate due to natural radionuclides is estimated by a Monte Carlo calculation with preanalyzed data of a representative sample of the measurement target. The activity concentration of natural radionuclides in concrete differs markedly depending on the production location of its components, such as cement and aggregates. In this study, using six mock concrete waste samples, which were composed of cement and fine aggregate from various production locations, the accuracy of BG compensation was experimentally estimated. In addition, the accuracy of calibration for concrete waste was also estimated using a number of mock concrete segments of small and large triangular prisms. By considering the uncertainties of BG compensation and calibration, the detection limit of CLALIS for concrete waste was estimated. As a result, it was revealed that CLALIS could be applied to the clearance measurement of concrete segments when the mass of the measurement target is greater than approximately 1.1 kg and the key radionuclide is Co-60.
Decommissioning of the BR3 Reactor: Status and Perspectives
Authors: Noynaert L. , Verstraeten I.
Published: June 2009

The BR3 plant at Mol in Belgium built at the end of the fifties was the first PWR plant built outside the USA. The reactor had a small net power output (10 MWe) but comprised all the loops and features of a commercial PWR plant. The BR3 plant was operated with the main objective of testing advanced PWR fuels under irradiation conditions similar to those encountered in large commercial PWR plants. The reactor was started in 1962 and shut down in 1987 after 25 years of continuous operation. Since 1989, SCK•CEN is decommissioning the BR3 PWR research reactor. The dismantling of the metallic components including reactor pressure vessel and internals is completed and extensively reported in the literature. The dismantling of auxiliary components and the decontamination of parts of the infrastructure are now going on. The decommissioning progress is continuously monitored and costs and strategy are regularly reassessed. The first part of the paper describes the main results and lessons learned from the reassessment exercises performed in 1994, 1999, 2004 and 2007. Impacts of changes in legal framework on the decommissioning costs will be addressed. These changes concern e.g. licensing aspects, clearance levels, waste management… The middle part of the paper discusses the management of activated and/or contaminated concrete. The costing exercise performed in 1995 highlighted that the management of activated and contaminated concrete is the second main cost item after the dismantling of the reactor pressure vessel and internals. Different possible solutions were studied. These are evacuation as radioactive waste with or without supercompaction, recycling this “radioactive” grout or concrete for conditioning of radioactive waste e.g. conditioning of metallic waste. The paper will give the results of the cost-benefit analysis made to select the solution retained. The last part of the paper will discuss the end goal of the decommissioning of the BR3. In the final decommissioning plan approved by ONDRAF/NIRAS, it was mentioned that the final goal of the BR3 decommissioning will be the “green field” unless opportunities for reuse of the BR3 site will occur during decommissioning. A strategy of partial reuse of the BR3 facility is proposed and being discussed with the main stakeholders. The paper will give the present state of the discussion.
Advances in the Application of In Situ Electrical Resistance Heating
Authors: Beyke Gregory , Smith Gregory J.
Published: June 2009

Electrical Resistance Heating (ERH) is an aggressive in situ thermal remediation technology that was developed by the U.S. Department of Energy from the original oil production technology to enhance vapor extraction remediation technologies in low permeability soils. Soil and groundwater are heated by the passage of electrical current through saturated and unsaturated soil between electrodes, not by the electrodes themselves. It is the resistance to the flow of electrical current that results in increased subsurface temperatures, and this is typically applied to the boiling point of water. It is estimated that more than 75 ERH applications have been performed. Capacity to perform these projects has increased over the years, and as many as 15 to 20 of these applications now being performed at any given time, mainly in North America, with some European applications. While the main focus has been to vaporize volatile organic compounds, as one would expect other semi-volatile and non-volatile organic compounds have also been encountered, resulting in observations of chemical and physical reactions that have not been normally incorporated into environmental restoration projects. One such reaction is hydrolysis, which is slow under normal groundwater temperatures, becomes very rapid under temperatures that can easily be achieved using ERH. As a result, these chemical and physical reactions are increasing the applicability of ERH in environmental restoration projects, treating a wider variety of compounds and utilizing biotic and abiotic mechanisms to reduce energy costs. For the treatment of oil and coal tar residues from manufactured gas plants, a process TRS has called steam bubble floatation is used to physically remove the coal and oil tar from the soils for collection using conventional multi-phase collection methods. Heat-enhanced hydrolysis has been used to remediate dichloromethane from soils and groundwater at a site in Illinois, while heat-enhanced biotic and abiotic dehalogenation has been observed at the vast majority of the sites where ERH has been applied. With disposal options becoming more limited around the world, alternate in situ treatment methods for soil and groundwater restoration are becoming more important. Over the 10 years of commercialization of the ERH technology, soil and groundwater remediation mechanisms and processes that were not envisioned by the technology’s developers expand the range of chemicals that have successfully been treated. This paper will discuss these processes and how these processes have been used to effect remediation of soil and groundwater where ERH has been employed.
Experimental Investigation on the Volume Reduction of Irradiated Graphite Arising From the Decommissioning of KRR-2
Authors: Lee D. G. , Cho Y. J. , Yang H. C. , Lee K. W. , Jung C. H.
Published: June 2009

Graphite has been used as a moderator and reflector of neutrons in more than 100 nuclear power plants as well as many experimental reactors and plutonium production reactors in various countries. Most of the older graphite moderated reactors are already shut down and are awaiting decommissioning planning and preparation. The graphite waste has different characteristics than other decommissioning waste due to its physical and chemical properties and also because of the presence of tritium and carbon-14. Therefore radioactive graphite dismantling, handling, conditioning and disposal are a common part of the decommissioning activities. A volume reduction of the waste is needed to reduce disposal cost of radioactive waste. However the existing processing technologies are based mostly on the isolation of radioactive graphite from the environment, they are not able to provide for a significant volume reduction. For this reason, the high-temperature thermal treatment process such as an incineration or a pyrolysis is considered as promising technologies, since it provides a substantial volume reduction. Currently, the fluidized bed incineration is considered as efficient technology for the treatment of radioactive graphite waste. In this study, the fluidized bed incineration condition and the radioisotopes behavior were experimentally investigated by using irradiated graphite waste which has arisen from the decommissioning of Korean Research Reactor 2 (KRR-2).
Development of Improved Methodology for the Comparative Assessment of Potential Repository Concepts and Locations
Authors: Tsuchi Hiroyuki , Koike Akihisa , Sato Shoko , Kawamura Hideki
Published: June 2009

NUMO has adopted a volunteering approach to siting a geological repository for high-level radioactive waste (HLW). It is important for this process that the pros and cons of volunteers can be assessed from literature data in a clear and transparent manner, prior to the very careful selection of those sites that will be carried forward for more detailed characterisation. For this purpose, multi-attribute analysis (MAA) methodology has been developed that allows the technical assessment of criteria to be represented as scoring models. The trickier job of weighting different criteria involves expert opinion, which can be solicited by different methods. In particular, weighting of top-level attributes involves balancing a range of technical and socio-economic issues, which can be examined by considering the viewpoint of different stakeholders. The applicability of the MAA tool and its sensitivity to stakeholder viewpoints have been examined by simple case studies.
Reduction of Radioactive Secondary Waste With Steam Reforming in Treatment of Waste TBP/Dodecane
Authors: Sone Tomoyuki , Sasaki Toshiki , Yamaguchi Hiromi
Published: June 2009

Waste tributyl phosphate (TBP) and normal dodecane generated from R&D activities on recycle of nuclear fuel has been stored in Japan Atomic Energy Agency (JAEA). If it is incinerated, a large quantity of contaminated phosphorous compounds will be generated as radioactive secondary wastes. The objective of this study is to reduce the generation of the radioactive secondary wastes by the treatment of the waste TBP/dodecane using steam reforming system. We constructed the demonstration scale steam reforming system which consists of a gasification chamber for vaporization of wastes, a metal mesh filter for removal of radioactive nuclides from gasified wastes, a combustion chamber, and scrubbers for removal of phosphorous oxides. We conducted process demonstration tests using waste TBP/dodecane with 0.07 g/L of uranium. We studied the temperature dependence of the gasification ratio of inorganic phosphorus compounds formed by pyrolysis of TBP in the gasification chamber and removal of uranium by the filter. As the results, more than 90% of phosphorus compounds were gasified from the gasification chamber at temperature of 600°C or more, and the uranium concentration in the waste water generated from the off-gas treatment system is under the detection limits. The waste water containing the separated phosphorus compounds can be discharged into the river or the sea as the liquid wastes in which uranium concentration is under the regulatory level. These results show the steam reforming system is effective in the reduction of radioactive secondary waste in the treatment of TBP/dodecane.
The Radioactivity Analysis of 14C in the Graphite Samples From the Dismantled KRR-1&2 Sites by a High Temperature Furnace and a LSC
Authors: Kim Hee Reyoung , Lee Wanno , Chung Kun Ho , Kang Mun Ja , Lee Dong Gyu , Choi Geun Sik , Lee Chang Woo
Published: June 2009

The radioactivity of 14C of the graphite samples from the dismantled Korea Research Reactor 1&2 (the KRR-1&2) site was analyzed and proposed to be disposed of as a low level radioactive waste rather than self-disposed of. The graphite wastes, with a weight of seven tons, have been generated during the dismantling of a research reactor with a capacity of one MW from 1995 to 2006. The graphite was used as a moderator for the research reactor and so has been radioactivated by thermal neutron. It was thought that the graphite wastes mainly included a radioisotope of stable carbon, 14C, a pure beta emitter with a half life of 5,730 years and with a maximum decay energy of 156 keV. Therefore, it has been requested to see whether the dismantled graphite radioactive wastes including 14C can be self-disposed of or not. In the present study, the radioactivity of 14C in the graphite sample used in the research reactor was analyzed by using a commercialized high temperature furnace and a Liquid Scintillation Counter (LSC). The combustion temperature of the furnace was five hundred degrees centigrade and especially the temperature in the catalyst region was eight hundred degrees centigrade. The recovery from the furnace was 95% for 14C and the LSC had a quenching efficiency of approximately 66%. Carbosorb was used as a trapping solution for 14C. The radioactivity of 14C was measured by a LSC through the procedure of a pre-treatment such as the combustion of a sample in the temperature range of 500–800 degrees centigrade by a high temperature furnace, trapping of 14C into carbosorb and cocktailing it with a scintillator. The radioactivity was analyzed to have a concentration with a value of much more than a domestic legal limit for a self-disposal. And an individual effective dose rate estimation was also carried out. Finally, it is suggested that the graphite wastes from the dismantled research reactor should be disposed of at a low level radioactive waste disposal site and monitored.
Radiolytic Decomposition of Organic C-14 Released From TRU Waste
Authors: Kani Yuko , Noshita Kenji , Kawasaki Toru , Nishimura Tsutomu , Sakuragi Tomofumi , Asano Hidekazu
Published: June 2009

It has been found that metallic TRU waste releases considerable portions of C-14 in the form of organic molecules such as lower molecular weight organic acids, alcohols and aldehydes. Due to the low sorption ability of organic C-14, it is important to clarify the long-term behavior of organic forms under waste disposal conditions. From investigations on radiolytic decomposition of organic carbon molecules into inorganic carbonic acid, it is expected that radiation from TRU waste will decompose organic C-14 into inorganic carbonic acid that has higher adsorption ability into the engineering barriers. Hence we have studied the decomposition behavior of organic C-14 by gamma irradiation experiments under simulated disposal conditions. The results showed that organic C-14 reacted with OH radicals formed by radiolysis of water, to produce inorganic carbonic acid. We introduced the concept of “decomposition efficiency” which expresses the percentage of OH radicals consumed for the decomposition reaction of organic molecules in order to analyze the experimental results. We estimated the effect of radiolytic decomposition on the concentration of organic C-14 in the simulated conditions of the TRU disposal system using the decomposition efficiency, and found that the concentration of organic C-14 in the waste package will be lowered when the decomposition of organic C-14 by radiolysis was taken into account, in comparison with the concentration of organic C-14 without radiolysis. Our prediction suggested that some amount of organic C-14 can be expected to be transformed into the inorganic form in the waste package in an actual system.
Estimating Fate and Transport of Multiple Contaminants in the Vadose Zone Using a Multi-Layered Soil Column and Three-Phase Equilibrium Partitioning Model
Authors: Rucker Gregory G.
Published: June 2009

Soils at waste sites must be evaluated for the potential of residual soil contamination to leach and migrate to the groundwater beneath the disposal area. If migration to the aquifer occurs, contaminants can travel vast distances and pollute drinking water wells, thus exposing human receptors to harmful levels of toxins and carcinogens. To prevent groundwater contamination, a contaminant fate and transport analysis is necessary to assess the migration potential of residual soil contaminants. This type of migration analysis is usually performed using a vadose zone model to account for complex geotechnical and chemical variables including: decay processes, infiltration rate, soil properties, vadose zone thickness, and chemical behavior. The distinct advantage of using a complex model is that less restrictive, but still protective, soil threshold levels may be determined avoiding the unnecessary and costly remediation of marginally contaminated soils. However, the disadvantage of such modeling is the additional cost for data collection and labor required to apply these models. In order to allay these higher costs and to achieve a less restrictive but still protective clean-up level, a multiple contaminant and multi layered soil column equilibrium partitioning model was developed which is faster, simpler and less expensive to use.
Segmentation and Removal of the Carolinas-Virginia Tube Reactor (CVTR) Moderator Tank
Authors: Anderson Michael G.
Published: June 2009

Special tooling has been deployed to segment the Moderator Tank (MT) at the Carolinas-Virginia Tube Reactor (CVTR) Parr site near Jenkinsville, South Carolina. The MT or reactor vessel, the most activated component remaining on site which included over 1,000 Ci of activation products, has been segmented into sections to fit within three hardware liners and three custom boxes. This work has been completed in approximately 12 months from tool conception to final packaging with no spread of contamination, no generation of secondary wastes and minimizing personnel radiological exposure. With contact dose readings in excess of 90 R/hr, segmentation of the MT had to be performed remotely and with the assurance that the spread of contamination to otherwise clean areas of the reactor building did not occur. Additionally, since the MT was entombed within a bioshield not capable of containing water, cutting had to be performed dry without benefit of shielding typically provided by the water of a spent fuel pool. In addition, the component removal scope included the removal, packaging and disposal of other activated components including thermal shields and the steel liner from the internal face of the bioshield. Concept engineering began in January 2006. Tools were tested and delivered in May 2006. Segmentation was completed in December 2006, followed by the removal of the thermal shields and bioshield liner. The component removal work was completed without the spread of contamination, no generation of secondary waste and an exposure total of 17 person rem.
Management of Liabilities at SCK•CEN: Lessons Learned
Authors: Noynaert L. , Cornelissen R.
Published: June 2009

SCK•CEN launched its technical liabilities and waste management program in 1989. This program refers 3 research reactors, 1 zero power reactor and nuclear laboratories buildings. The main decommissioning activities at SCK•CEN are focused on the BR3 reactor, but decommissioning activities are also carried out in other SCK•CEN facilities. These activities mainly concern old equipments and experiments which have to be decommissioned to make room for new R&D projects. In the past, 4 laboratory buildings of SCK•CEN were fully cleaned before they were transferred for unrestricted reuse to a non nuclear institute. The management of spent fuel and nuclear material is also part of this program. It mainly concerns the back end of BR2 HEU spent fuel, the BR3 LEU and MOX spent fuel. The Technical Liabilities and Waste Management Program are continuously monitored. Technical Liabilities costs were estimated in 1989, 1995, 2000 and 2005. This regular reassessment of the liability costs allows identifying the key issues to be tackled for the sound management of the liabilities. The key issues are listed hereafter by order of decreasing importance: • the differences between the assumption made by the State to secure the Technical Liabilities Fund and the observed economical conditions; • the drastic increase of the waste tariff; • the decrease of the clearance levels; • the development of the legislation regarding decommissioning and waste production activities.
Experiences in the Field of Radioactive Materials Seizures in the Czech Republic
Authors: Svoboda Karel , Podlaha Josef , Sˇi´r David , Mudra Josef
Published: June 2009

In recent years, the amount of radioactive materials seizures (captured radioactive materials) has been rising. It was above all due to newly installed detection facilities that were able to check metallic scrap during its collection in scrap yards or on the entrance to iron-mills, checking municipal waste upon entrance to municipal disposal sites, even incineration plants, or through checking vehicles going through the borders of the Czech Republic. Most cases bore a relationship to secondary raw materials or they were connected to the application of machines and installations made from contaminated metallic materials. However, in accordance to our experience, the number of cases of seizures of materials and devices containing radioactive sources used in the public domain was lower, but not negligible, in the municipal storage yards or incineration plants. Atomic Act No. 18/1997 Coll. will apply to everybody who provides activities leading to exposure, mandatory assurance as high radiation safety as risk of the endangering of life, personal health and environment is as low as reasonably achievable in according to social and economic aspects. Hence, attention on the examination of all cases of the radioactive material seizure based on detection facilities alarm or reasonably grounds suspicion arising from the other information is important. Therefore, a service carried out by group of workers who ensure assessment of captured radioactive materials and eventual retrieval of radioactive sources from the municipal waste has come into existence in the Nuclear Research Institute Rez plc. This service has covered also transport, storage, processing and disposal of found radioactive sources. This service has arisen especially for municipal disposal sites, but later on even other companies took advantage of this service like incineration plants, the State Office for Nuclear Safety, etc. Our experience in the field of ensuring assessment of captured radioactive materials and eventual retrieval of radioactive sources will be presented in the paper.
The Community Environmental Monitoring Program: A Model for Stakeholder Involvement in Environmental Monitoring
Authors: Hartwell William T. , Shafer David S.
Published: June 2009

Since 1981, the Community Environmental Monitoring Program (CEMP) has involved stakeholders directly in its daily operation and data collection, as well as in dissemination of information on radiological surveillance in communities surrounding the Nevada Test Site (NTS), the primary location where the United States (US) conducted nuclear testing until 1992. The CEMP is funded by the US Department of Energy’s National Nuclear Security Administration, and is administered by the Desert Research Institute (DRI) of the Nevada System of Higher Education. The CEMP provides training workshops for stakeholders involved in the program, and educational outreach to address public concerns about health risk and environmental impacts from past and ongoing NTS activities. The network includes 29 monitoring stations located across an approximately 160,000 km2 area of Nevada, Utah and California in the southwestern US. The principal radiological instruments are pressurized ion chambers for measuring gamma radiation, and particulate air samplers, primarily for alpha/beta detection. Stations also employ a full suite of meteorological instruments, allowing for improved interpretation of the effects of meteorological events on background radiation levels. Station sensors are wired to state-of-the-art dataloggers that are capable of several weeks of on-site data storage, and that work in tandem with a communications system that integrates DSL and wireless internet, land line and cellular phone, and satellite technologies for data transfer. Data are managed through a platform maintained by the Western Regional Climate Center (WRCC) that DRI operates for the U.S. National Oceanic and Atmospheric Administration. The WRCC platform allows for near real-time upload and display of current monitoring information in tabular and graphical formats on a public web site. Archival data for each station are also available on-line, providing the ability to perform trending analyses or calculate site-specific exposure rates. This configuration also allows for remote programming and troubleshooting of sensors. Involvement of stakeholders in the monitoring process provides a number of benefits, including increased public confidence in monitoring results, as well as decreasing costs by more than 50 percent from when the program was managed entirely by U.S. federal employees. Additionally, the CEMP provides an ideal platform for testing new environmental sensors.
Fire as a Long-Term Stewardship Issue for Soils Contaminated With Radionuclides in the Western U.S.
Authors: Shafer David S. , DuBois David , Etyemezian Vic , Kavouras Ilias , Miller Julianne J. , Nikolich George , Stone Mark
Published: June 2009

On both U.S. Department of Energy (DOE) and U.S. Department of Defense sites in the southwestern United States (U.S.), significant areas of surface soils are contaminated with radionuclides from atmospheric nuclear testing, and with depleted uranium, primarily from military training. At DOE sites in Nevada, the proposed regulatory closure strategy for most sites is to leave contaminants in place with administrative controls and periodic monitoring. Closure-in-place is considered an acceptable strategy because the contaminated sites exist on access-restricted facilities, decreasing the potential risk to public receptor, the high cost and feasibility of excavating contaminated soils over large areas, and the environmental impacts of excavating desert soils that recover very slowly from disturbance. The largest of the contaminated sites on the Tonopah Test Range in Nevada covers over 1,200 hectares. However, a factor that has not been fully investigated in the long-term stewardship of these sites is the potential effects of fires. Because of the long half-lives of some of the contaminants (e.g., 24,100 years for 239Pu) and changes in land-cover and climatic factors that are increasing the frequency of fires throughout the western U.S., it should be assumed that all of these sites will eventually burn, possibly multiple times, during the timeframe when they still pose a risk. Two primary factors are contributing to increased fire frequency. The first is the spread of invasive grasses, particularly cheatgrass (Bromus tectorum and Bromus rubens), which have out-competed native annuals and invaded interspaces between shrubs, allowing fires to burn easier. The second is a sharp increase in fire frequency and size throughout the western U.S. beginning in the mid-1980s. This second factor appears to correlate with an increase in average spring and summer temperatures, which may be contributing to earlier loss of soil moisture and longer periods of dry plant biomass (particularly from annual plants). The potential risk to site workers from convective heat dispersion of radionuclide contaminants is an immediate concern during a fire. Long-term, post-fire concerns include potential changes in windblown suspension properties of contaminated soil particles after fires because of loss of vegetation cover and changes in soil properties, and soil erosion from surface water runoff and fluvial processes.
Safety Analysis of Irradiated Nuclear Fuel Transportation Container
Authors: Uspuras E. , Rimkevicius S.
Published: June 2009

Ignalina NPP comprises two Units with RBMK-1500 reactors. After the Unit 1 of the Ignalina Nuclear Power Plant was shut down in 2004, approximately 1000 fuel assemblies from Unit were available for further reuse in Unit 2. The fuel-transportation container, vehicle, protection shaft and other necessary equipment were designed in order to implement the process for on-site transportation of Unit 1 fuel for reuse in the Unit 2. The Safety Analysis Report (SAR) was developed to demonstrate that the proposed set of equipment performs all functions and assures the required level of safety for both normal operation and accident conditions. The purpose of this paper is to introduce the content and main results of SAR, focusing attention on the container used to transport spent fuel assemblies from Unit 1 on Unit 2. In the SAR, the structural integrity, thermal, radiological and nuclear safety calculations are performed to assess the acceptance of the proposed set of equipment. The safety analysis demonstrated that the proposed nuclear fuel transportation container and other equipment are in compliance with functional, design and regulatory requirements and assure the required safety level.
Decommissioning Strategy Options for MR Reactor at the Kurchatov Institute, Moscow
Authors: Cross M. T. , Harman N. F. , Charles D. , Harper A. , Bylkin B. K. , Gorlinksy Yu. E. , Kolyadin V. I. , Kutkov V. A. , Pavlenko V. I. , Sivintsev Yu. V. , Lobach Yu. N.
Published: June 2009

The principal decommissioning goal for the nuclear installations of the Kurchatov Institute is the removal of spent fuel, reactor facilities and radioactive waste from the Institute’s site. As the result of decommissioning, the buildings, constructions and areas should be cleaned to residual contamination levels acceptable to the stakeholders. These levels are determined in view of possible options for the rehabilitation of the Institute’s areas under conditions of compliance with acting legislation for safety provisions for staff, population noting the proximity of the site to the local urban environment. Research reactor MR was commissioned in 1963 for reactor materials testing and finally shutdown in 1993. The reactor power with the experimental loops was 50 MWt. Several features were identified for the development of a decommissioning strategy for this reactor, namely: – the strategy should consider many factors in a broad approach with international, inter-industry and long-term perspectives; – the current situation for decommissioning is uncertain and must account for the views of a variety of stakeholders on possible final conditions and further use of the site and the route to achieve these; and – a lack of sufficiency in the national legislation base for execution of the work and the possible options for its completion. On the basis of worldwide experience, the strategy for decommissioning of reactor MR was determined as follows: – determination of the options for the final rehabilitation of the Institute’s areas; – determination of the stakeholders and their priority concerns; – determination of the strategy options for achievement of the final status; – determination of the main factors influencing the selection of the decommissioning strategy; – selection of the most acceptable strategies on the basis of a multi-attribute analysis; – determination of the main stages and principles of implementation of the selected strategy; and – development of the decommissioning activities considering the work that will be required. As the result of the multi-attribute analysis, the following conclusion has been made: – the preferred final status of the facility is for a nuclear re-use application; – the preferred decommissioning option is immediate dismantling.
The Strategy on Rehabilitation of the Former Uranium Facilities at the “Pridneprovsky Chemical Plant” in Ukraine
Authors: Voitsekhovich O. , Lavrora T. , Skalskiy A. S. , Ryazantsev V. F.
Published: June 2009

This paper describes current status of the former Uranium Facilities at the Pridneprovsky Chemical Plant in Ukraine, which are currently under development of action plan for its territory rehabilitation. The monitoring data carried out during recent several years show its impact to the Environment and gives a basis for justification of the number of measures aiming to reduce radiological and ecological risks of the Uranium tailings situated at the territory of PChP. The monitoring data and strategy for its remediation are considered in the presentation. Uranium mining has been intensively conducted in Ukraine since the end of the 40-s. Most of the uranium deposits have been explored in the Dnieper river basin, shown in Figure 1, while some smaller deposits can be found within the basins of the Southern Bug and Severskiy Donets rivers. There also several large Uranium Milling facilities were in operation since the end of the 40-s till 1991, when due to disintegration of the former Soviet Union system the own uranium production has been significantly declined. The Milling Plant and Uranium extraction Facilities in ZhevtiVody is still in operation with UkrAtomprom Industrial Consortium. Therefore rehabilitation programme for all Uranium facilities in this site are in duty of the East Mining Combine and the Consortium. The most difficult case is to provide rehabilitation Action Plan for Uranium tailings and number of other facilities situated in Dnieprodzerzhinsk town and which were in operation by the former State Industrial Enterprise Pridneprovskiy Chemical Plant (PChP).
Innovative Integration of Decommissioning and Deactivation Program With Soil-Groundwater Clean Up Program Has Positive Results on Budget and Schedule: A Case Study
Authors: Schappell Bruce G. , Rucker Gregory G.
Published: June 2009

An innovative approach to integrate the activities of a decommissioning and deactivation program (D&D) with a soil-groundwater clean up program has had significant positive results saving both money and time at the Department of Energy’s Savannah River Site. The accomplishments that have been achieved by the combining the two programs have been remarkable including significant cost savings, economies of scale for sampling and document generation, and alignment of common objectives. Because of the coordination of both activities area-wide “end states” can be formulated and be consistent with the customers’ cleanup goals and federal regulations. This coordinates and aligns both the environmental clean up and D&D objectives because each must be addressed simultaneously and comprehensively. In this respect, resources from both organizations can be pooled to take advantage of the strengths of each. The new approach allows more efficient use of lean financial resources and optimizes workforce activities to attain the common objectives while being more cost effective, more protective of the environment, and optimizing the use existing resources.
Phased Implementation of In Situ Chemical Oxidation for a Large TCE DNAPL Source Area at the Portsmouth Gaseous Diffusion Plant, USA
Authors: Cross Paul E. , Baird Del
Published: June 2009

This paper describes the In Situ Chemical Oxidation (ISCO) remediation being implemented for the X-701B groundwater plume at the Department of Energy (DOE) Portsmouth Gaseous Diffusion Plant (PORTS). Modified Fenton’s reagent is the principal oxidant for the remedy, and Direct Push Technology (DPT) is being used for delivery of the oxidant. Trichloroethene (TCE) is the primary contaminant of concern and is present within the unit as a dense non-aqueous phase liquid (DNAPL). A phased approach is being implemented to optimize the type, location, and mass of the oxidant injections. During Phase I, a unique near-real time monitoring approach was utilized to observe the transient effects of the oxidant injections on the formation. As a result of the positive results from Phase I, Ohio EPA has approved the final work plan for the remedy, and the approach is now being applied to the source area of the plume. The results from Phase I and the layout for the first series of Phase II injections are presented in this paper. Previous testing at the site has shown that the shallow, water-bearing formation is primarily composed of silty gravel and clay, and is both heterogeneous and anisotropic. These factors have significantly compromised earlier attempts to remediate the unit. A patented ISCO process from In-Situ Oxidative Technologies, Inc. (ISOTEC) was selected for the remediation of the plume. Phase I results indicate that oxidant delivery via DPT is feasible for the unit. Contaminant reduction to date has been minimal due to the small quantity of oxidant injected during Phase I. Contaminant rebound in the aqueous phase remains a concern and will be monitored closely during the remedy.
Removal of Mercury From Aqueous Solutions by ETS-4 Microporous Titanosilicate: Effect of Contact Time, Titanosilicate Mass and Initial Metal Concentration
Authors: Lopes C. B. , Otero M. , Lin Z. , Pereira E. , Silva C. M. , Rocha J. , Duarte A. C.
Published: June 2009

Mercury is one of the most toxic heavy metals present in the environment and therefore is extremely important develop new, simple and reliable techniques for its removal from aqueous solutions. A recent line of research within this context is the application of microporous materials. The use of these materials for removing heavy metals from solutions may become a potential clean-up technology in the field of wastewater treatment. In this work it is reported the application of microporous titanosilicate ETS-4 as ion exchanger to remove Hg2+ from aqueous solution. Under batch conditions, we studied the effect of contact time, titanosilicate mass and initial Hg2+ concentration. Only 5 mg of ETS-4 are required to purify 2 litres of water with 50 μg L−1 of metal. Under the experimental conditions, the initial Hg2+ concentration and ETS-4 mass have strong influence on the sorption process, and it is proved that 24 h are almost always sufficient to attain ion exchange equilibrium. Langmuir and Freundlich isotherms were used to fit equilibrium experimental results. The kinetics of mercury removal was reliably described by a pseudo second-order model. On the whole, ETS-4 shows considerable potential to remove Hg2+ from wastewaters.
Modelling Multi-Phase Flow Phenomena in Concrete Barriers Used for Geological Disposal of Radioactive Waste
Authors: Mallants Dirk , Jacques Diederik , Perko Janez
Published: June 2009

Gas generation and gas transport phenomena occur in geological repositories of radioactive waste. This has been extensively studied over the past ten years, usually within the framework of international projects (MEGAS, PROGRESS, etc.). These studies indicate that the production of hydrogen by anaerobic corrosion of metals is the most important source for gas generation. Laboratory and in situ experiments carried out at SCK•CEN indicate that, in the presence of Boom Clay (the reference geologic formation for deep disposal studies in Belgium), carbon steel suffers generalised corrosion estimated conservatively at 1 μm y−1. Simulations with the finite difference multi-phase flow code TOUGH2 were carried out in an attempt to quantify the effects of hydrogen gas generation on desaturation of initially saturated concrete components of the disposal gallery and the concomitant expulsion of cementitious pore-water into the surrounding host formation. Several simulation cases were considered and addressed differences in initial water saturation degree of concrete, hydrogen gas generation rate, and material porosity. Several conceptual models have been developed to better understand the phenomena at work in the transport of gas in the cementitious engineered barriers and Boom Clay. Multi-phase flow modelling was found to be helpful to get insight into the phenomenology of coupled water-gas flow in the cementitious engineered barriers. However, modeling the discontinuous variation in the conductivity of the clay relative to the gas (creation of preferential pathways) requires incorporation of geomechanical processes in conventional models based on the laws of two-phase flow.
Development of a Processing and Treatment Solution for a Thoria Waste Stream
Authors: Anderson Andy , Jenkins Jon , Mitchell Charles , Simmons Richard
Published: June 2009

Waste Management Technology Ltd (WMT) has developed the optimal process for immobilizing a solid waste contaminated with thorium dioxide (thoria). The physical and chemical characteristics of the waste present challenges to producing a wasteform acceptable for disposal. Also, high-energy radiation from thorium’s decay progeny requires a treatment plant with shielding and remote handling facilities. Key points of the paper are as follows. 1. Treatment options were investigated and the best practicable means identified as intimate mixing of the waste with cementitious grout. 2. Samples were analysed for particle size and organic contamination. 3. Small-scale active mixes resulted in a single treatment formulation for all the waste. Leach tests confirmed that the organic material is adequately retained within the immobilised waste provided activated carbon is included in the formulation. 4. Active mixes at the two litre scale confirmed that the formulation is mixable and the product acceptable and consistent with expectations from the earlier work. 5. WMT is constructing a treatment plant at its Winfrith site, based on remote grouting in a 200 litre drum with a sacrificial mixer. Inactive full-scale trials with such 200 litre drums were carried out after selection of simulants with the appropriate physical properties.
Release and Disposal of Materials During Decommissioning of Siemens MOX Fuel Fabrication Plant at Hanau, Germany
Authors: Koenig Werner , Baumann Roland
Published: June 2009

In September 2006, decommissioning and dismantling of the Siemens MOX Fuel Fabrication Plant in Hanau were completed. The process equipment and the fabrication buildings were completely decommissioned and dismantled. The other buildings were emptied in whole or in part, although they were not demolished. Overall, the decommissioning process produced approximately 8500 Mg of radioactive waste (including inactive matrix material); clearance measurements were also performed for approximately 5400 Mg of material covering a wide range of types. All the equipment in which nuclear fuels had been handled was disposed of as radioactive waste. The radioactive waste was conditioned on the basis of the requirements specified for the projected German final disposal site “Schachtanlage Konrad”. During the preconditioning, familiar processes such as incineration, compacting and melting were used. It has been shown that on account of consistently applied activity containment (barrier concept) during operation and dismantling, there has been no significant unexpected contamination of the plant. Therefore almost all the materials that were not a priori destined for radioactive waste were released without restriction on the basis of the applicable legal regulations (§ 29 of the Radiation Protection Ordinance), along with the buildings and the plant site.
The Crucial Importance of the Back-End in Multinational Initiatives to Enhance Fuel Cycle Security
Authors: McCombie Charles , Chapman Neil , Isaacs Thomas H.
Published: June 2009

There have been repeated proposals for establishing multinational cooperation approaches that could reduce the security concerns of spreading nuclear technologies. Most recently, there have been initiatives by both Russia (GNPI) and the USA (GNEP) – each aimed at promoting nuclear power whilst limiting security concerns. In practice, both initiatives place emphasis on the supply of reactors and enriched fuel but neither has made clear and specific proposals about the back-end part of the arrangement. The primary incentive offered to the user countries is “security of supply” of the front end services. However, there is no current shortage of supply of front end services, so that the incentives are not large. A much greater incentive could be the provision of a spent fuel or waste disposal service. The fuel supplied to Tier 2 countries could be shipped back (with no return of wastes) to the supplier or else to an accepted third party country that is trusted to operate safe and secure disposal facilities. If a comprehensive service that obviates the need for a national deep repository is offered to small countries then there will be a really strong incentive for them to sign up to GNEP or GNPI type deals.