bluuebunny/crossref_metadata_2025 huggingface dataset:
Stepwise Site Selection in Switzerland: Sectoral Plan Status and Outlook
Authors: Ernst Thomas , Fritschi Markus , Vomvoris Stratis
Published: April 2011

The stepwise site selection process in Switzerland is governed by the Sectoral Plan. It is divided in three stages, narrowing down the number of potential siting regions (Stage 1) to at least two sites for each of the two geological repositories, the low and intermediate-level waste site (LLW) and the high-level waste site (HLW), in Stage 2, leading to the final selection of a site for each repository (Stage 3) for the application of a general license. In October 2008, Nagra proposed a total of 6 siting regions for the LLW repository and three for the HLW repository; the latter would also be suitable for shared use with the surface facilities and part of the access tunnels for the two types of repositories. The review of Nagra’s proposals by the safety authority (ENSI — the Swiss Federal Nuclear Safety Inspectorate) and its supporting commission was completed in February 2010. ENSI found Nagra’s analysis justified comprehensive and transparent and approved the six proposed regions. A similar overall conclusion was reached by KNS, the Swiss Commission for Nuclear Safety. The decision-making process continues in 2010 with an evaluation of the review and recommendations of ENSI by the various agencies at the governmental, cantonal and local level, an open public consultation and finally, a resolution of the comments received by the Swiss Federal Office of Energy (SFOE). Stage 1 will be concluded in 2011 with the decision by the Swiss Federal Government. Stage 2 foresees a provisional safety analysis for each potential site as an additional criterion to be used in the narrowing down to two sites for each type of waste repository. The guidelines for this have been recently published by ENSI. The main findings of the review by the authorities, the next steps and preparatory activities for the initiation of Stage 2, as well as how the criteria and guidelines specified by ENSI will be applied by Nagra in order to meet the requirements for a successful completion of Stage 2 are described below.
THOR® Steam Reforming Technology for the Treatment of Ion Exchange Resins and More Complex Wastes Such as Fuel Reprocessing Wastes
Authors: Mason J. Brad , Myers Corey A.
Published: April 2011

The THOR® fluidized bed steam reforming process has been successfully operated for more than 10 years in the United States for the treatment of low- and intermediate-level radioactive wastes generated by commercial nuclear power plants. The principle waste stream that has been treated is ion exchange resins (IER) and Dry Active Waste (DAW), but various liquids, sludges, and solid organic wastes have also been treated. The principle advantages of the THOR® process include: (a) high volume reduction on the order of 5:1 to 10:1 for IER and up to 50:1 for high plastic content DAW streams depending on the waste type and waste characteristics, (b) environmentally compliant off-gas emissions, (c) reliable conversion of wastes into mineralized products that are durable and leach-resistant, and (d) no liquid effluents for treatment of most radioactive wastes. Over the past ten years, the THOR® process has been adapted for the treatment of more complex wastes including historic defense wastes, reprocessing wastes, and other wastes associated with the fuel cycle. As part of the U.S. Department of Energy (DOE) environmental remediation activities, the THOR® dual bed steam reforming process has successfully processed: (a) Idaho National Laboratory (INL) Sodium-Bearing Waste (SBW), (b) Savannah River Tank 48 High Level Waste (HLW), (c) Hanford Low Activity Waste (LAW), and (d) Hanford Waste Treatment Plant Secondary Waste (WTP-SW) liquid slurry simulants. The THOR® process has been shown in pilot plant operations to successfully process various simulated liquid, radioactive, nitrate-containing wastes into environmentally safe, leach-resistant, solid mineralized products. These mineralized products incorporate normally soluble ions (e.g. - Na, K, Cs, Tc), sulfates, chloride salts, and fluoride salts into an alkali alumino-silicate mineral matrix that inhibits the leaching of those ions into the environment. The solid mineralized products produced by the THOR® process exhibit durability and leach resistance characteristics superior to borosilicate waste glasses. As a result of this work, a full-scale THOR® process facility is currently under construction at the DOE’s Idaho site for the treatment of SBW and a full-scale facility is in the final design stage for the DOE’s Savannah River Site for the treatment of Tank 48 high level waste. Recent work has focused on the development of new monolithic waste formulations, the extension of the THOR® process to new waste streams, and the development of modular THOR® processes for niche waste treatment applications. This paper will provide an overview of current THOR® projects and summarize the processes and outcomes of the regulatory and safety reviews that have been necessary for the THOR® process to gain acceptance in the USA.
Feasibility Study on the Nuclide Analysis of the Radwaste Drum Using the Spectrum to Dose Conversion Factor
Authors: Ji Young-Yong , Hong Dae-Seok , Kim Tae-Kuk , Ryu Woo-Seog
Published: April 2011

Estimating the radioisotope inventory of a drum based on the measured dose rate information, which is called as the dose to curie (DTC) conversion [1–3], has been known that there could be extremely high uncertainty associated with establishing the radioactivity of gamma emitters in a drum. However, the DTC method is still an effective assay method to calculate the radioisotope inventory because of their simple and easy procedures to be applied. To make the DTC method practical, numerous assumptions have to be made and limitations placed on its use. These assumptions and limitations were related to the dose rate measurement and the relative abundance of gamma emitters in a drum. These two variables were generally obtained from the different detection mechanism. Unfortunately, that expanded the limitation of using the DTC method. The spectrum to dose (STD) conversion factor [4,5] that was calculated from the measure pulse height spectrum of the gamma ray detector could be made obtaining two variables from the drum to be assayed at once. This method could be made estimating the radioisotope inventory of a drum more practical.
Long Term Storage of Nuclear Spent Fuel as Key Role of Japan’s Nuclear Fuel Cycle Until 2100: Cost and Benefit
Authors: Katsuta Tadahiro
Published: April 2011

Political and technical advantages to introduce spent nuclear fuel interim storage into Japan’s nuclear fuel cycle are examined. Once Rokkasho reprocessing plant starts operation, 80,000 tHM of spent Low Enriched Uranium (LEU) fuel must be stored in an Away From Reactor (AFR) interim storage site until 2100. If a succeeding reprocessing plant starts operating, the spent LEU will reach its peak of 30,000 tHM before 2050, and then will decrease until the end of the second reprocessing plant operation. Throughput of the second reprocessing plant is assumed as twice of that of Rokassho reprocessing plant, indeed 1,600tHM/year. On the other hand, tripled number of final disposal sites for High Level Nuclear Waste (HLW) will be necessary with this condition. Besides, large amount of plutonium surplus will occur, even if First Breeder Reactors (FBR)s consume the plutonium. At maximum, plutonium surplus will reach almost 500 tons. These results indicate that current nuclear policy does not solve the spent fuel problems but rather complicates them. Thus, reprocessing policy could put off the problems in spent fuel interim storage capacity and other issues could appear such as difficulties in large amount of HLW final disposal management or separated plutonium management. If there is no reprocessing or MOX use, the amount of spent fuel will reach over 115,000 tones at the year of 2100. However, the spent fuel management could be simplified and also the cost and the security would be improved by using an interim storage primarily.
The Radioactivity of 3H in Metals by a High Temperature Furnace and a Liquid Scintillation Counter
Authors: Kim Hee Reyoung , Choi Geun Sik , Park Sang Yun , Lee Chang Woo , Han Moon Hee
Published: April 2011

The radioactivity of 3H of the metal samples from the nuclear sites was analyzed by using a commercialized high temperature furnace and a Liquid Scintillation Counter (LSC). The 3H activity of the sample was measured according to the duration of the high temperature combustion and the oxidation temperature. Basically, the recovery from the furnace was 90% for 3H and the LSC had a quenching efficiency of approximately 30 %. HNO3 was used as a trapping solution for 3H and the solution was cocktailed with a scintillator. The activity extracted from the sample was increased till the combustion time elapsed 60 minutes and the increasing rate was reduced continuously thereafter at 600 °C whereas 80% of radioactivity was extracted during the first 15 minutes at 900 °C. Also, the pretreatment for the metal sample, which included a high temperature combustion and trapping, had the time required of at least four hours at 900 °C. Finally, it was suggested that this high temperature combustion method could be applied to analyze the activity of the radioactive metal waste from the nuclear power plants.
A Milestone in Vitrification: The Replacement of a “Hot Metallic Crucible” With a “Cold Crucible Melter” in a Hot Cell at the La Hague Plant, France
Authors: Robert S. , Gassot Guilbert F. , Carpentier B. , Naline S. , Gouyaud F. , Girold C.
Published: April 2011

Vitrification of high-level waste is the internationally recognized way to optimize the conditioned high level waste to be disposed of. AREVA’s successful operation of AVM at the Marcoule plant, R7 and T7 at the La Hague Plants demonstrates the capabilities and experience of the Group to deploy innovative high level waste (HLW) processing technologies in industrial facilities, in partnership with R&D (CEA) and engineering (SGN)., CEA and AREVA engineering team have continuously improved the hot metallic crucible melter vitrification technology through operational feedback as well as ongoing research and development. They have led the way in the development and implementation of Cold Crucible Induction Melter technology (CCIM). The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass are molten by direct high frequency induction. This technology can handle with highly corrosive solutions and high operating temperature, which offers, among others advantages: • A great flexibility in matrix compositions, • The increase of the industrial glass production capacity, • Increase of melting temperature, • Allow higher wastes loading, • Extend to a wide and varied waste type. In order to take advantage of CCIM, a new project, called “Vitrification 2010”, was launched in 2005 with the objective to put in commercial operation a CCIM in one of the existing active R7 vitrification line in 2010. The main challenge for the project team is to replace the hot metallic crucible melter with the CCIM: • In an existing hot cell which is in operation for 20 years with high level waste solutions, • With respect to the current hot cell’s layout (calciner, through-wall for utility for example...). The paper will illustrate why this event is a major milestone in vitrification and how the “Vitrification 2010” project succeeds in record time and without any impact on the La Hague Plant production.
Dismantling Method of Fuel Cycle Facilities Obtained by Dismantling of the JRTF
Authors: Kanayama Fumihiko
Published: April 2011

The Japan Atomic Energy Research Institute Reprocessing Test Facility (JRTF) was the first reprocessing facility which was constructed by applying only Japanese technology to establish basic technology on wet reprocessing. JRTF had been operated since 1968 to 1969 using spent fuels (uranium metal/aluminum clad, about 600kg as uranium metal and 600MWD/T) from the Japan Research Reactor No.3 (JRR-3). Reprocessing testings on PUREX process were implemented at 3 runs, so that, 200g of plutonium dioxide were extracted. After JRTF was shut down at 1970, it had been used for research and development of reprocessing since 1971. The more mature research and development of nuclear are, the more opportunity of dismantling of old nuclear facilities would be. Japan Atomic Energy Agency (JAEA) has an experience of full scale of dismantling through decommissioning of Japan Power Demonstration Reactor (JPDR)1). On the other hand, we didn’t have that of fuel cycle facility. Moreover, it is considered that dismantling methods of nuclear reactor and fuel cycle facility are different for following reason, components contaminated TRU nuclide including Pu, and components installed inside narrow cells. Dismantling methods are important factor to decide manpower and time for dismantling. So, it is indispensable to optimize dismantling method in order to minimize manpower and time for dismantling. Considering the background mentioned above, the decommissioning project of JRTF was started in 1990. The decommissioning project of JRTF is carried out phase by phase. Phase 1; Investigation for dismantling of the JRTF2)3)4). Phase 2; R&D of decommissioning technologies for dismantling of the JRTF5)6)7)8). Phase 3; Actual dismantling of the JRTF9)10). There were several components used for reprocessing and a system for liquid radwaste storage, and those were installed inside of each of several thick concrete cells. The inner surfaces of each cell were contaminated by TRU nuclides including Pu. In phase 3, components used in reprocessing and a system for liquid radwaste storage were dismantled. Moreover, opening was made in concrete walls (including ceiling) for this work. Effective practices for dismantling fuel cycle facilities were obtained through these works. On this report, effective dismantle methods obtained by actual dismantling activities in JRTF are introduced.
Verification of Source Term Analysis System for Decommissioning Wastes From a CANDU Reactor
Authors: Cho Dong-Keun , Sun GwangMin , Choi JongWon , Hwang Donghyeun , Kim Hak-Soo , Hwang Tae-Won
Published: April 2011

There are now twenty commercial nuclear power reactors operating as of May 2010 in South Korea. As nuclear capacity becomes higher and installations age, the Korean government and industry have launched R&D to estimate appropriate decommissioning costs of power reactors. In this paper, MCNP/ORIGEN2 code system which is being developed as a source term evaluation tool was verified by comparing the estimated nuclide inventory from MCNP/ORIGEN2 simulation with the measured nuclide inventory from chemical assay in an irradiated pressure tube discharged from Wolsong Unit 1 in 1994. Equilibrium core model of Wolsoung unit 1 was used as a neutron source to activate in-core and ex-core structural components. As a result, the estimated values from the analysis system agreed with measured data within 20% difference. Therefore, it can be concluded that MCNP/ORIGEN system could be a reliable tool to estimate source terms of decommissioning wastes from CANDU reactor, although this system assumes constant flux irradiation and snapshot equilibrium core model as a reference core.
Integrated Model for the Near Field of a Repository in Granite Host-Rock: Probabilistic Approach
Authors: Duro Lara , Valls Alba , Riba Olga , Bruno Jordi , Marti´nez-Esparza Aurora
Published: April 2011

The application of probabilistic approaches to the performance of underground repositories for long-lived radioactive waste has received special attention in the last years. Numerous exercises have been developed in order to elicit the Probability Distribution Functions (PDFs) of the several parameters needed for these developments. Several integrated models allow the implementation of PDFs in the long-term simulations needed for Performance Assessment. In this work we present how the deterministic compartmental model for a repository of high level nuclear waste (HLNW) located in a crystalline host-rock has been modified to include PDFs for some of the parameters. The implementation of probabilistic approaches gives also information on the most influencing parameter on the migration of the different radionuclides from a deep repository concept.
Exposure Dose Evaluation of Worker at Radioactive Waste Incineration Facility on KAERI
Authors: Park Sang Kyu , Jeon Jong Seon , Kim Youn Hwa , Lee Jae Min , Lee Gi Won
Published: April 2011

An incineration treatment of inflammable radioactive wastes leads to have a reduction effect of disposal cost and also to contribute an enhancement of safety at a disposal site by taking the advantage of stabilization of the wastes which is accomplished by converting organic materials into inorganic materials. As it was required for an incineration technology, KAERI (Korea Atomic Energy Research Institute) has developed a pilot incineration process and then constructed a demonstration incineration facility having based on the operating experiences of the pilot process. In this study, worker exposure doses were evaluated to confirm safety of workers before the demonstration incineration facility will commence a commercial.
Scenario Development for Safety Assessment of Waste Repository for Feasibility Study on Transmutation of Spent Nuclear Fuel Into LILW Using PEACER
Authors: Kim Sung-yeop , Lee Kun Jai
Published: April 2011

PEACER (Proliferation-resistant, Environmental-friendly, Accident-tolerant, Continuable-energy and Economical Reactor) is a conceptual liquid metal fast reactor using Pb-Bi as a coolant and feasibility study on transmutation of spent nuclear fuel into LILW (Low and Intermediate Level Waste) using PEACER is in progress. Safety assessment of repository is essential for this feasibility study with assumption that we dispose the wastes from PWRs and PEACERs with established decontamination factors. Scenario development is one of important step for carrying out reliable and comprehensive safety assessment. This study adopted scenario development methodology from H12 report (JNC, 2000) and classified assessment scenarios into base scenario, perturbation scenarios and isolation failure scenarios. Scenarios are established by classifying, screening out and selecting FEPs with concepts and conditions of disposal for feasibility study.
RFID Technology for Environmental Remediation and Radioactive Waste Management
Authors: Tsai Hanchung , Liu Yung Y. , Shuler James
Published: April 2011

An advanced Radio Frequency Identification (RFID) system capable of tracking and monitoring a wide range of materials and components—from fissionable stocks to radioactive wastes—has been developed. The system offers a number of advantages, including enhanced safety, security and safeguards, reduced personnel exposure to radiation, and improved inventory control and cost-effectiveness. Using sensors, RFID tags can monitor the state of health of the tracked items and trigger alarms instantly when the normal ranges are violated. Nonvolatile memories in the tags can store sensor data, event records, as well as a contents manifest. Gamma irradiation tests showed that the tag components possess significant radiation resistance. Long-life batteries and smart management circuitries permit the tags to operate for up to 10 years without battery replacement. The tags have a near universal form factor, i.e., they can fit different package types. The read range is up to >100 m with no line-of-sight required. With careful implementation, even a large-size processing or storage facility with a complex configuration can be monitored with a handful of readers in a network. In transportation, by incorporating Global Positioning System (GPS), satellite/cellular communication technology, and secure Internet, situation awareness is assured continuously. The RFID system, when integrated with Geographic Information System (GIS) technology, can promptly provide content- and event-specific information to first responders and emergency management teams in case of incidents. In stand-alone applications, the monitoring and tracking data are contained within the local computer. With a secure Internet, information can be shared within the complex or even globally in real time. As with the deployment of any new technology, overcoming the cultural resistance is part of the developmental process. With a strong institutional support and multiple successful live demonstrations, the cultural resistance has been mostly overcome. As a result, implementation of the RFID technology is taking place at several of U.S. Department of Energy sites and laboratories for processing, storage, and transportation applications.
A Comparative Study of Stakeholder Participation in the Cleanup of Radioactive Wastes in the US, Japan and UK
Authors: Lawless W. F. , Akiyoshi Mito , Whitton John , Angjellari-Dajci Fjorentina , Poppeliers Christian
Published: April 2011

We review case studies of stakeholder participation in the environmental cleanup of radioactive wastes in the United States, Japan and United Kingdom (e.g., [21,26,27,66,78]). Citizen participation programs in these three countries are at different stages: mature in the US, starting in Japan, and becoming operational in the UK. The US issue at the US Department of Energy’s (DOE) Savannah River Site (SRS) in South Carolina (SC) had been focused on citizens encouraging Federal (DOE; US Environmental Protection Agency, or EPA; and the US Nuclear Regulatory Commission, or NRC) and State (SC’s Department of Health and Environmental Compliance, or DHEC) agencies to pursue “Plug-in-RODs” at SRS to simplify the regulations to accelerate closing seepage basins at SRS. In Japan, the Reprocessing of spent fuel and deep geological disposal of vitrified high-level waste have been among Japan’s priorities. A reprocessing plant in Rokkasho, Aomori Prefecture is expected to commence operations in October 2010. The search of a site for a deep geological disposal facility has been ongoing since 2002. But the direct engagement of stakeholders has not occurred in Japan. Indirectly, stakeholders attempt to exert influence on decision-making with social movements, local elections, and litigation. In the UK, the issue is gaining effective citizen participation with the UK’s Nuclear Decommissioning Authority (NDA). We hope that the case studies from these countries may improve citizen participation.
An Analyse of Components and Impact Factors Related to Spent Fuel Transportation Plans
Authors: Shin SangWon , Lee JaMin , Lee SuHong , Baeg ChangYeal , Yoon JeongHyoun
Published: April 2011

The simulation tool of spent fuel transportation plan is being developed to support the development of core technology for transporting and storing spent fuel developed by the Korea Radioactive Waste Management Corporation (KRMC). So we extracted the major activities involved in spent fuel transport — such as setting-up a transport volume, infrastructure, processes and systems. And then we analyzed the components and properties of each activity. The impact factors related to spent fuel transport were determined to be maximum loading capacity of the INF ship and interim storage facility location, and it was around these that the preliminary scenario was generated. This will be used as a conceptual data model for the development of a simulation tool for spent fuel transportation plans.
Application of Lifecycle Management to Design of the UK Geological Disposal Facility
Authors: Rendell Philip G. P. , O’Grady Henry J. P. , Currie Malcolm F.
Published: April 2011

The Radioactive Waste Management Directorate (RWMD) of the United Kingdom’s (UK) Nuclear Decommissioning Authority (NDA) has been given the responsibility for delivery of a Geological Disposal Facility (GDF) for the UK’s higher activity wastes in accordance with government policy. As part of this process, the RWMD has developed a project lifecycle, which addresses the overall lifecycle of the GDF in terms of five phases, from Preparatory Studies through to Operation and finally Closure, and is developing a staged approach to engineering design. The Engineering Design Process is broken down into seven stages, encompassing option development, requirements definition and preliminary and detailed design through to “design development during closure”. Each stage finishes with a formally defined milestone (a “gate”) comprising a technical review and a specific set of engineering deliverables. This paper describes the background to the UK GDF development programme, the organisational issues associated with the RWMD’s evolving role, the relationship between the top-level UK Government’s Managing Radioactive Waste Safely programme [1] and the RWMD engineering lifecycle, the formal reviews, the milestones and the overall contribution this makes to RWMD organisational development and UK regulatory approval. It also describes some of the lessons learnt.
Tokai-1 Decommissioning Project: The First Challenge in Japan
Authors: Yoshino Keizaburou
Published: April 2011

Tokai-1 is the oldest and historical commercial Magnox reactor in Japan, which had started commercial operation in 1966. The unit had helped introduction and establishment of the construction and operation technologies regarding nuclear power plant at early stage in Japan by its construction and operating experiences. However, The Japan Atomic Power Company (JAPC), the operator and owner of Tokai-1, decided to cease its operation permanently because of a fulfillment of its mission and economical reason. The unit was finally shut down in March 1998 after about 32 year operation. It took about three years for removal of all spent fuels from the site, and then decommissioning started in 2001. JAPC, always on the forefront of the nation’s nuclear power generation, is now grappling Japan’s first decommissioning of a commercial nuclear power plant, striving to establish effective, advanced decommissioning. The decommissioning for Tokai-1 was scheduled as 20 years project. At the beginning, the reactor was started to be in a static condition for ten years (“safe storage period”). While the reactor had been safely stored, the phased decommissioning works started from non-radioactive or low radioactive equipments toward high radioactive equipments. First five years of the project, JAPC concentrated to drain and clean spent fuel cartridge cooling pond and to remove conventional equipments such as turbine, feed water pump and fuel charge machine as planed and budgeted. From 2006, the project came into new phase. JAPC has been trying to remove four Steam Raising Units (SRUs). The SRUs are huge component (750ton, φ 6.3m, H24.7m) of the Gas Cooling Reactor (GCR) and inside of the SRUs are radioactively contaminated. Major concerns are workers safety and minimizing contamination areas during SRU removal. Therefore, JAPC is developing and introducing Jack-down method and remote control multi-functional dismantling system. This method is to cut and remove the SRUs in turn from the bottom to top remotely while lifting the SRU by a large jack system. The system enables cutting and holding not only the SRU body but also internals. This technology and experiences would be useful for the reactor removal in the near future.
Portable Non-Destructive Assay Methods for Screening and Segregation of Radioactive Waste
Authors: Simpson Alan , Jones Stephanie , Clapham Martin , Lucero Randy
Published: April 2011

Significant cost-savings and operational efficiency may be realised by performing rapid non-destructive classification of radioactive waste at or near its point of retrieval or generation. There is often a need to quickly categorize and segregate bulk containers (drums, crates etc.) into waste streams defined at various boundary levels (based on its radioactive hazard) in order to meet disposal regulations and consignor waste acceptance criteria. Recent improvements in gamma spectroscopy technologies have provided the capability to perform rapid in-situ analysis using portable and hand-held devices such as battery-operated medium and high resolution detectors including lanthanum halide and high purity germanium (HPGe). Instruments and technologies that were previously the domain of complex lab systems are now widely available as touch-screen “off-the-shelf” units. Despite such advances, the task of waste stream screening and segregation remains a complex exercise requiring a detailed understanding of programmatic requirements and, in particular, the capability to ensure data quality when operating in the field. This is particularly so when surveying historical waste drums and crates containing heterogeneous debris of unknown composition. The most widely used portable assay method is based upon far-field High Resolution Gamma Spectroscopy (HRGS) assay using HPGe detectors together with a well engineered deployment cart (such as the PSC TechniCART™ technology). Hand-held Sodium Iodide (NaI) detectors are often also deployed and may also be used to supplement the HPGe measurements in locating hot spots. Portable neutron slab monitors may also be utilised in cases where gamma measurements alone are not suitable. Several case histories are discussed at various sites where this equipment has been used for in-situ characterization of debris waste, sludge, soil, high activity waste, depleted and enriched uranium, heat source and weapons grade plutonium, fission products, activation products, americium, curium and other more exotic nuclides. The process of acquiring and analyzing data together with integration of historical knowledge to resolve and delineate waste streams (for example between low-level waste and transuranic waste) is described.
Phase Behavior and Reverse Micelle Formation in Supercritical CO2 With DTAB and F-Pentanol for Decontamination of Radioactive Wastes
Authors: Kurahashi Kensuke , Tomioka Osamu , Meguro Yoshihiro
Published: April 2011

To develop a metal separation method using supercritical CO2 (scCO2) solvent for the decontamination process of radioactive wastes, the reverse micelle formation in scCO2 was investigated. Dodecyltrimethylammonium bromide (DTAB) as a surfactant to form the reverse micelles and 2,2,3,3,4,4,5,5-octafluoro-1-pentanol (F-pentanol) as a modifier to increase the solubility of DTAB into scCO2 were used. The reverse micelles could be formed by using 0.02 mol/dm3 DTAB and 0.45 mol/dm3 F-pentanol. A water concentration dissolved in scCO2 was increased with an increase of pressure, and 0.42 mol/dm3 water, which was 3 times larger than that in the neat CO2, could be dissolved in scCO2 at 38 MPa. Moreover, 0.1 mol/dm3 HNO3 could dissolve at the same pressure as water. On the other hand, it was found that the solubility of water at outside of reverse micelles increased with F-pentanol. The ratio of water and F-pentanol affected the phase behavior of water in scCO2.
Removal of Fission Products in the Spent Electrolyte Using Iron Phosphate Glass as a Sorbent
Authors: Amamoto Ippei , Mitamura Naoki , Tsuzuki Tatsuya , Takasaki Yasushi , Shibayama Atsushi , Yano Tetsuji , Nakada Masami , Okamoto Yoshihiro
Published: April 2011

This study is carried out to make the pyroprocessing hold a competitive advantage from the viewpoint of environmental load reduction and economical improvement. As one of the measures to reduce the volume of the high-level radioactive waste (HLW), the phosphate conversion method is applied for removal of fission products (FP) from the melt, referring to the spent electrolyte in this paper. Among the removing target chlorides in the spent electrolyte i.e., alkali metals, alkaline earth metals and rare earth elements, only the rare earth elements and lithium form the precipitates as insoluble phosphates by reaction with Li3PO4. The sand filtration method was applied to separate FP precipitates from the spent electrolyte. The iron phosphate glass (IPG) powder, which is a compatible material for the immobilization of FP, was used as a filter medium. After filtration experiment, it was proven that insoluble FP could almost be completely removed from the spent electrolyte. Subsequently, we attempted to separate the dissolved FP from the spent electrolyte. The IPG was being used once again but this time as a sorbent instead. This is possible because the IPG has some unique characteristics, e.g., changing the valence of iron, which is one of its network modifiers due to its manufacturing temperature. Therefore, it would be likely to sorb some FP when the chemical condition of IPG is unstable. We produced three kinds of IPG under different manufacturing temperature and confirmed that those glasses could sorb FP as anticipated. According to the experimental result, its sorption efficiency of metal cations was attained at around 20–40%.
Considerations for Grout Formulations for Facility Closures Using In Situ Strategies
Authors: Serrato Michael J. , Langton Christine A. , Gladden John B. , Long John T. , Blankenship John K. , Szilagyi Andrew P. , Hannah George R. , Stubblefield Rita B.
Published: April 2011

The U.S. Department of Energy (DOE) is conducting in situ closures (entombment) at a large number of facilities throughout the complex. Among the largest closure actions currently underway are the closures of the P and R Reactors at the Savannah River Site (SRS), near Aiken, South Carolina. In these facilities, subgrade open spaces are being stabilized with grout; this ensures the long term structural integrity of the facilities and permanently immobilizes and isolates residual contamination. The large size and structural complexity of these facilities present a wide variety of challenges for the identification and selection of appropriate fill materials. Considerations for grout formulations must account for flowability, long term stability, set times, heat generation and interactions with materials within the structure. The large size and configuration of the facility necessitates that grout must be pumped from the exterior to the spaces to be filled, which requires that the material must retain a high degree of flowability to move through piping without clogging while achieving the required leveling properties at the pour site. Set times and curing properties must be controlled to meet operations schedules, while not generating sufficient heat to compromise the properties of the fill material. The properties of residual materials can result in additional requirements for grout formulations. If significant quantities of aluminum are present in the facility, common formulations of highly alkaline grouts may not be appropriate because of the potential for hydrogen generation with the resultant risks. SRS is developing specialized inorganic grout formulations that are designed to address this issue. One circum-neutral chemical grout formulation identified for initial consideration did not possess the proper chemical characteristics, having exceptionally short set times and high heat of hydration. Research efforts are directed toward developing grout formulations that can meet operational requirements for chemical compatibility, extended set times and reduced heat generation.
Safety Assessment of Disposal Container for Higher Activity Low Level Waste
Authors: Nakagami Motonori , Komatsuki Seiji , Fujisawa Kyosuke , Nishio Takashi , Quercetti Thomas , Musolff Andre´ , Mu¨ller Karsten
Published: April 2011

As one of the studies on “yoyushindo disposal” whose concept is similar to an intermediate disposal, the development of a disposal container has been conducted by the Federation of Electric Power Companies of Japan. To assess a drop event of a waste package in which stored the radioactive wastes from nuclear power plants, the toughness of the disposal container was evaluated by drop tests using three specimens which have actual dimensions, drop analysis, fracture mechanics assessment and macroscopic tests. The three specimens for drop tests were manufactured in consideration of the design specifications and the manufacture operations in nuclear power plants. The lid plates of the specimens were welded to the body plates without pre- and post-weld heat treatment by using a remote automated welding machine. The drop tests showed that no penetration cracks or splash of its content occurred in the disposal container under conservative conditions such as the maximum weight and height in the handling. Drop analysis and the fracture mechanics assessment indicate that the strain induced by the drop impact did not exceed the fracture strain and an unstable fracture did not occur. And macroscopic tests showed that penetration cracks did not occur at 8m drop events. These tests and evaluations confirmed that the disposal container had sufficient toughness.
Evaluation of the Activated Radioactivity of Turbine Equipments in BWR
Authors: Watanabe Masato , Nakagami Motonori
Published: April 2011

The activated radioactivity of turbine equipments irradiated by neutron originating from 17N in the main stream is evaluated for an introduction of clearance system to boiling-water reactor (BWR) plant. The 17N, main neutron source is generated by 17O(n, p)17N reaction in the core region. The evaluation results clarified that the activated radioactivity of the turbine equipment is extremely small comparing to the clearance level. The feature of the evaluation is as follows. (1) Actual radioactive concentration of the 17N in the main steam in Hamaoka nuclear power station unit 5 (Hamaoka-5) which is an advanced boiling-water reactor (ABWR) was measured with solid-state track detector (SSTD). The 17N concentration is used for the neutron transport calculation as initial neutron sources. (2) The turbine equipments were modeled as two-dimensional geometry for DORT code. (3) Activation cross-sections for major nuclides subject to the clearance evaluation were based on JENDL3.3 on 175 energy group structure (VITAMIN-J). (4) Minor nuclides subject to the clearance evaluation were calculated with ORIGEN-S code.
Cement Based Solidification / Stabilization of Industrial Contaminated Soil Using Various Cement Additives
Authors: Voglar Grega E. , Lestan Domen
Published: April 2011

In a laboratory study, 15% (w/w) of ordinary portland cement (OPC), black portland cement (BPC) and puculanic cement (PC) combined with various cement additives were used for solidification/stabilization (S/S) of Cd, Pb, Zn, Cu, Ni and As contaminated soils from the former industrial site. Soil formed solid monoliths with all cements and cement additives. S/S effectiveness was assessed by measuring the mechanical strength of the monoliths, concentrations of metals in deionised water and TCLP (Toxicity Characteristic Leaching Procedure) soil extracts, and mass transfer of metals. Concentrations of Cd, Pb, Zn and Ni in water extracts from S/S soils generally decreased, while concentrations of As and Cu increased. Concentrations of Cd, Pb, Zn, Cu and Ni in the TCLP extracts from S/S mixes were lower than from original soil, while the extractability of As from S/S mixes increased. Overall, the concentration of metals in TCLP solution, obtained after extraction of the S/S mixes, was below the regulatory limits. S/S greatly reduced the mass transfer of Cd (up to 21.5-times), Pb (up to 114-times) and Zn (up to 999-times). Mass transfer of Ni was generally also reduced, while that of Cu and As increased in some S/S mixes. In general, monoliths made with OPC present higher cumulative mass of PTMs leached than those with BPC and PC.
Latex Particles Functionalized With Transition Metals Ferrocyanides for Cesium Uptake and Decontamination of Solid Bulk Materials
Authors: Avramenko Valentin , Bratskaya Svetlana , Zheleznov Veniamin , Sheveleva Irina , Marinin Dmitry , Sergienko Valentin
Published: April 2011

Decontamination of spent ion-exchange resins, corrosion-unstable metal structures, soil, ground, and construction materials contaminated by fission, corrosion and transuranic radionuclides remains one of the most urgent and complicated ecological problems. Among the existing methods having different efficiencies in regard to such materials decontamination, application of selective sorbents put into a humid medium to be decontaminated (ground, bulk materials) appears to be rather extensive. However, the efficiency of such an approach is significantly limited by difficulties concerned with uniform sorbent distribution in porous media and completeness of spent sorbents removal for final disposal. In this paper we suggest a principally new approach to preparation of colloid-stable selective sorbents for cesium uptake using immobilization of transition metals (cobalt, nickel, and copper) ferrocyanides in nanosized carboxylic latex emulsions. The effects of ferrocyanide composition, pH, and media salinity on the sorption properties of the colloid-stable sorbents toward cesium ions were studied in solutions containing up to 200 g/1 sodium nitrate or potassium chloride. The sorption capacities of the colloid sorbents based on mixed potassium/transition metals ferrocyanides were in the range 1.45–1.86 mol Cs/mol ferrocyanide with the highest value found for the copper ferrocyanide. It was shown that the obtained colloid-stable sorbents were capable to penetrate through bulk materials without filtration that makes them applicable for decontamination of solids, e.g. soils, zeolites, spent ion-exchange resins contaminated with cesium radionuclides. After decontamination of liquid or solid radioactive wastes the colloid-stable sorbents can be easily separated from solutions by precipitation with cationic flocculants providing localization of radionuclides in a small volume of the precipitates formed. Besides, functionalized latex particles can be used for preparation of carbon fiber/ferrocyanide composite materials for cesium uptake using electrodeposition method. Application of the carbon fibers as an inert support for ferrocyanides, in general, significantly improves the sorption kinetics, but washing out of ferrocyanide fines from the fiber surface limits the potential of such materials. When ferrocyanides are deposited in a form of nanocrystals stabilized by latexes which undergo electropolymerization on the fiber surface, the thin polymeric film formed substantially improves the stability of the composite and prevents loss of ferrocyanide during sorbent application. The effect of electrodeposition conditions on composite morphology, ferrocyanide loading and cesium distribution coefficient in media with different salinity has been discussed.
Repository Engineering and Demonstration: Special Challenges for TRU
Authors: McKinley Ian G. , Takase Hiroyasu
Published: April 2011

The diverse range of long-lived radioactive wastes without significant heat output specified for deep geological disposal (here termed TRU) pose challenges that are potentially more serious than those from vitrified high-level waste and spent fuel. Despite this, the latter tend to be the focus of R&D in national programmes. Such challenges are particularly severe for the case for countries that are not considering evaporite host rocks or have a volunteering approach to siting and for those with inventories of TRU resulting from reprocessing of spent fuel. While there is little doubt that safe disposal of TRU is feasible, it is tricky to develop a convincing safety case for a site during early stages of characterisation as, compared to HLW/SF, less credit can be taken for robust, long-term performance of current designs of the engineered barrier systems. In order to improve this situation and increase flexibility with respect to host rock properties, two different options are available — improving the conditioning of particular waste streams or improving the overall repository safety concept. Although the former has been a focus for work in some countries (particularly Japan), much less effort has been invested in the latter and hence this will be illustrated by some examples. These options are compared in terms of their pros and cons with respect to practicality of implementation, environmental impact and cost. Additionally, the ease with which the resulting safety case can be supported by demonstrations of key arguments will be discussed, which may indicate the likely degree of acceptance by stakeholders.
Neural Network Based Predictive Emission Monitoring Module for a GE LM2500 Gas Turbine
Authors: Botros K. K. , Cheung M.
Published: April 2011

A Predictive Emission Monitoring (PEM) model has been developed for a non-DLE GE LM2500 gas turbine used on a natural gas compressor station on the TransCanada Pipeline System in Alberta. The PEM model is based on an optimized Neural Network (NN) architecture which takes four fundamental engine parameters as input variables. The model predicts NOx emission in ppmv-dry-O2 corrected and in kg/hr as NO2. The NN was trained using Continuous Emission Monitoring (CEM) measurements comprising two sets of actual emission data collected over two different dates in 2009, when the ambient ambient temperatures were vastly different (∼1° C and 24 °C), respectively. These training data were supplemented by other emission data generated by GE ‘Cycle-Deck’ tool to generate emission data at different ambient temperatures ranging from −30 to +30 °C. The outcome is a total of 1872 emission data of engine emissions at different operating conditions covering the range of the engine operating parameters (402 data points from CEM and 1470 data points from GE Cycle-Deck). The PEM model comprises a simple single hidden layer perceptron type NN with only two neurons in it. The performance of the NN-based model showed a correlation coefficient greater than 0.99, and error standard deviation of 4.5 ppmv of NOx and 1.4 kg/hr as NO2. Uncertainty analysis was conducted to assess the effects of uncertainties in the engine parameters on the NOx predictions by PEM. It was shown that for uncertainty in the ambient temperature of ±1 °C, the uncertainty in the NOx prediction is ± 0.9 to ±3.5%. Uncertainties of the order of ±1% in the other three input parameters results in uncertainties in NOx predictions by ±2.5 to ±6%. Finally, the PEM model was implemented in the station CEHM (Compressor Equipment Health Monitoring) system and NOx prediction were reported online on a minutely basis. These data are presented here over the first three months since implementation.
The Application of Quantitative X-Ray Diffraction (Rietveld Refinement) in Characterizing the Microstructure and Precipitates in Microalloyed Steels
Authors: Wiskel J. B. , Li X. , Lu J. , Ivey D. G. , Henein H.
Published: April 2011

Quantitative X-ray diffraction (the Rietveld Method) was used to quantify both the crystallite size (diffracting length) and precipitate size in several grades of microalloyed steels. The crystallite size calculated from diffraction profiles obtained from the rolling faces of X80 and X100 was found to decrease with decreasing coiling temperature. Quantitative X-ray diffraction analysis was also conducted on precipitate residues extracted (by matrix dissolution) from a Grade 100 microalloyed steel. A mean precipitate size of 4.8 nm was calculated for the fine (Nb, Mo, Ti, V) carbides observed in this steel. This value compares favourably with precipitate size measured using TEM and SANS analysis.