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Substantial Variability Exists in Utilities' Nuclear Decommissioning Funding Adequacy: Baseline Trends (1997-2001); and Scenario and Sensitivity Analyses (Year 2001)

Description: This paper explores the trends over 1997-2001 in my baseline simulation analysis of the sufficiency of electric utilities' funds to eventually decommission the nation's nuclear power plants. Further, for 2001, I describe the utilities' funding adequacy results obtained using scenario and sensitivity analyses, respectively. In this paper, I focus more on the wide variability observed in these adequacy measures among utilities than on the results for the ''average'' utility in the nuclear industry. Only individual utilities, not average utilities -- often used by the nuclear industry to represent its funding adequacy -- will decommission their nuclear plants. Industry-wide results tend to mask the varied results for individual utilities. This paper shows that over 1997-2001, the variability of my baseline decommissioning funding adequacy measures (in percentages) for both utility fund balances and current contributions has remained very large, reflected in the sizable ranges and frequency distributions of these percentages. The relevance of this variability for nuclear decommissioning funding adequacy is, of course, focused more on those utilities that show below ideal balances and contribution levels. Looking backward, 42 of 67 utility fund (available) balances, in 2001, were above (and 25 below) their ideal baseline levels; in 1997, 42 of 76 were above (and 34 below) ideal levels. Of these, many utility balances were far above, and many far below, such ideal levels. The problem of certain utilities continuing to show balances much below ideal persists even with increases in the adequacy of ''average'' utility balances.
Date: February 26, 2003
Creator: Williams, D. G.
Partner: UNT Libraries Government Documents Department

Evaluation of pipe whip impacts on neighboring piping and walls of the Ignalina nuclear power plant.

Description: Presented in this paper is the transient analysis of a Group Distribution Header (GDH) following a guillotine break at the end of the header. The GDH is the most important component of reactor safety in case of accidents. Emergency Core Cooling System (ECCS) piping is connected to the GDH piping such that, during an accident, coolant passes from the GDH into the ECCS. The GDH that is propelled into motion after a guillotine break can impact neighboring GDH pipes or the nearest wall of the compartment. Therefore, two cases are investigated: GDH impact on an adjacent GDH and its attached piping; and GDH impact on an adjacent reinforced concrete wall. A whipping RBMK-1500 GDH along with neighboring concrete walls and pipelines is modeled using finite elements. The finite element code NEPTUNE used in this study enables a dynamic pipe whip structural analysis that accommodates large displacements and nonlinear material characteristics. The results of the study indicate that a whipping GDH pipe would not significantly damage adjacent walls or piping and would not result in a propagation of pipe failures.
Date: February 26, 2002
Creator: Dundulis, G.; Kulak, R.F.; Marchertas, A.; Narvydas, E.; Petri, M.C. & Uspusas, E.
Partner: UNT Libraries Government Documents Department

ENHS : the encapsulated nuclear heat source - a nuclear energy concept for emerging worldwide energy markets.

Description: A market analysis is presented which delineates client needs and potential market size for small turnkey nuclear power plants with full fuel cycle services. The features of the Encapsulated Nuclear Heat Source (ENHS) which is targeted for this market are listed, and the status of evaluation of technological viability is summarized.
Date: February 26, 2002
Creator: Wade, D.C.; Feldman, E.; Sienicki, J.; Sofu, T.; Brown, N.W.; Hossain, Q. et al.
Partner: UNT Libraries Government Documents Department

STAR - H2 : the secure transportable autonomous reactor for hydrogen production and desalinization.

Description: The Secure Transportable Autonomous Reactor for Hydrogen production is a modular fast reactor intended for the mid 21st century energy market wherein electricity and hydrogen are employed as complementary energy carriers and nuclear energy contributes to sustainable energy supply based on full transuranic recycle in a passively safe, environmentally friendly and proliferation-resistant manner suitable for widespread worldwide deployment.
Date: February 26, 2002
Creator: Wade, D.C.; Doctor, R. & Peddicord, K.L.
Partner: UNT Libraries Government Documents Department

DEVELOPMENT OF HFE SECTIONS OF DG-1145.

Description: For the licensing of the current fleet of commercial nuclear power plants (NPPs), the Nuclear Regulatory Commission (NRC) used two key documents, NUREG-0800 and Regulatory Guide (RG) 1.70. RG 1.70 provided guidance to applicants on the contents needed in their Safety Analysis Reports (SARs) submitted as part of their application to construct or operate an NPP. NUREG-0800, the NRC Standard Review Plan (SRP), provides guidance to the NRR staff reviewers on performing their safety reviews of these applications. As part of the preparation for a new wave of improved NPP designs the NRC is in the process of updating the SRP and is also developing a new RG designated as draft RG or DG-1145, ''Combined License Applications for Nuclear Power Plants (LWR Edition).'' This will eventually become RG 1.206 and will take the place of RG 1.70. This will provide guidance for combined license (COL) applicants, as well as for other 10CFR Part 52 variations that are permitted.
Date: March 26, 2007
Creator: HIGGINS,J.C.; OHARA, J.M. & BONGARRA, J.
Partner: UNT Libraries Government Documents Department

COUNTERCURRENT FLOW LIMITATION EXPERIMENTS AND MODELING FOR IMPROVED REACTOR SAFETY

Description: This project is investigating countercurrent flow and “flooding” phenomena in light water reactor systems to improve reactor safety of current and future reactors. To better understand the occurrence of flooding in the surge line geometry of a PWR, two experimental programs were performed. In the first, a test facility with an acrylic test section provided visual data on flooding for air-water systems in large diameter tubes. This test section also allowed for development of techniques to form an annular liquid film along the inner surface of the “surge line” and other techniques which would be difficult to verify in an opaque test section. Based on experiences in the air-water testing and the improved understanding of flooding phenomena, two series of tests were conducted in a large-diameter, stainless steel test section. Air-water test results and steam-water test results were directly compared to note the effect of condensation. Results indicate that, as for smaller diameter tubes, the flooding phenomena is predominantly driven by the hydrodynamics. Tests with the test sections inclined were attempted but the annular film was easily disrupted. A theoretical model for steam venting from inclined tubes is proposed herein and validated against air-water data. Empirical correlations were proposed for air-water and steam-water data. Methods for developing analytical models of the air-water and steam-water systems are discussed, as is the applicability of the current data to the surge line conditions. This report documents the project results from July 1, 2005 through June 30, 2008.
Date: September 26, 2008
Creator: Vierow, Karen
Partner: UNT Libraries Government Documents Department

AFCI-2.0 Library of Neutron Cross Section Covariances

Description: Neutron cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The primary purpose of the library is to provide covariances for the Advanced Fuel Cycle Initiative (AFCI) data adjustment project, which is focusing on the needs of fast advanced burner reactors. The covariances refer to central values given in the 2006 release of the U.S. neutron evaluated library ENDF/B-VII. The preliminary version (AFCI-2.0beta) has been completed in October 2010 and made available to the users for comments. In the final 2.0 release, covariances for a few materials were updated, in particular new LANL evaluations for {sup 238,240}Pu and {sup 241}Am were adopted. BNL was responsible for covariances for structural materials and fission products, management of the library and coordination of the work, while LANL was in charge of covariances for light nuclei and for actinides.
Date: June 26, 2011
Creator: Herman, M.; Herman,M.; Oblozinsky,P.; Mattoon,C.; Pigni,M.; Hoblit,S. et al.
Partner: UNT Libraries Government Documents Department

Probabilistic Seismic Hazard Characterization and Design Parameters for the Sites of the Nuclear Power Plants of Ukraine

Description: The U.S. Department of Energy (US DOE), under the auspices of the International Nuclear Safety Program (INSP) is supporting in-depth safety assessments (ISA) of nuclear power plants in Eastern Europe and the former Soviet Union for the purpose of evaluating the safety and upgrades necessary to the stock of nuclear power plants in Ukraine. For this purpose the Hazards Mitigation Center at Lawrence Livermore National Laboratory (LLNL) has been asked to assess the seismic hazard and design parameters at the sites of the nuclear power plants in Ukraine. The probabilistic seismic hazard (PSH) estimates were updated using the latest available data and knowledge from LLNL, the U.S. Geological Survey, and other relevant recent studies from several consulting companies. Special attention was given to account for the local seismicity, the deep focused earthquakes of the Vrancea zone, in Romania, the region around Crimea and for the system of potentially active faults associated with the Pripyat Dniepro Donnetts rift. Aleatory (random) uncertainty was estimated from the available data and the epistemic (knowledge) uncertainty was estimated by considering the existing models in the literature and the interpretations of a small group of experts elicited during a workshop conducted in Kiev, Ukraine, on February 2-4, 1999.
Date: January 26, 2000
Creator: Savy, J.B. & Foxall, W.
Partner: UNT Libraries Government Documents Department

A Study on the Tritium Behavior in the Rice Plant after a Short-Term Exposure of HTO

Description: In many Asian countries including Korea, rice is a very important food crop. Its grain is consumed by humans and its straw is used to feed animals. In Korea, there are four CANDU type reactors that release relatively large amounts of tritium into the environment. Since 1997, KAERI (Korea Atomic Energy Research Institute) has carried out the experimental studies to obtain domestic data on various parameters concerning the direct contamination of plant. In this study, the behavior of tritium in the rice plant is predicted and compared with the measurement performed at KAERI. Using the conceptual model of the soil-plant-atmosphere tritiated water transport system which was suggested by Charles E. Murphy, tritium concentrations in the soil and in leaves to time were derived. If the effect of tritium concentration in the soil is considered, the tritium concentration in leaves is described as a double exponential model. On the other hand if the tritium concentration in the soil is disregarded, the tritium concentration in leaves is described by a single exponential term as other models (e.g. Belot's or STAR-H3 model). Also concentration of organically bound tritium in the seed is predicted and compared with measurements. The results can be used to predict the tritium concentration in the rice plant at a field around the site and the ingestion dose following the release of tritium to the environment.
Date: February 26, 2002
Creator: Yook, D-S.; Lee, K. J. & Choi, Y-H.
Partner: UNT Libraries Government Documents Department

Waste Minimization Policy at the Romanian Nuclear Power Plant

Description: The radioactive waste management system at Cernavoda Nuclear Power Plant (NPP) in Romania was designed to maintain acceptable levels of safety for workers and to protect human health and the environment from exposure to unacceptable levels of radiation. In accordance with terminology of the International Atomic Energy Agency (IAEA), this system consists of the ''pretreatment'' of solid and organic liquid radioactive waste, which may include part or all of the following activities: collection, handling, volume reduction (by an in-drum compactor, if appropriate), and storage. Gaseous and aqueous liquid wastes are managed according to the ''dilute and discharge'' strategy. Taking into account the fact that treatment/conditioning and disposal technologies are still not established, waste minimization at the source is a priority environmental management objective, while waste minimization at the disposal stage is presently just a theoretical requirement for future adopted technologies . The necessary operational and maintenance procedures are in place at Cernavoda to minimize the production and contamination of waste. Administrative and technical measures are established to minimize waste volumes. Thus, an annual environmental target of a maximum 30 m3 of radioactive waste volume arising from operation and maintenance has been established. Within the first five years of operations at Cernavoda NPP, this target has been met. The successful implementation of the waste minimization policy has been accompanied by a cost reduction while the occupational doses for plant workers have been maintained at as low as reasonably practicable levels. This paper will describe key features of the waste management system along with the actual experience that has been realized with respect to minimizing the waste volumes at the Cernavoda NPP.
Date: February 26, 2002
Creator: Andrei, V. & Daian, I.
Partner: UNT Libraries Government Documents Department

Radioactive Releases Impact from Kozloduy Nuclear Power Plant, Bulgaria into the Environment

Description: The aim of this paper is to present a general overview of the radioactive releases impact generated by Kozloduy Nuclear Power Plant (KNPP), Bulgaria to the environment and public. The liquid releases presented are known as the so called controlled water discharges, that are generated after reprocessing of the inevitable accumulated liquid radioactive waste in the plant operation process. The radionuclides containing in the liquid releases are given in the paper as a result of systematic measuring. Database for radiation doses evaluation on the public around Kozloduy NPP site is developed using IAEA LADTAP computerized program. The computer code LADTAP represents realization of a model that evaluates the public dose as a result of NPP releases under normal operation conditions. The results of this evaluation were the basic licensing document for a new liquid release limit.
Date: February 26, 2002
Creator: Genchev, G. T.; Kuleff, I.; Tanev, N. T.; Delistoyanova, E. S. & Guentchev, T.
Partner: UNT Libraries Government Documents Department

Decommissioning Project of Bohunice A1 NPP

Description: The first (pilot) nuclear power plant A1 in the Slovak Republic, situated on Jaslovske Bohunice site (60 km from Bratislava) with the capacity of 143 MWel, was commissioned in 1972 and was running with interruptions till 1977. A KS 150 reactor (HWGCR) with natural uranium as fuel, D2O as moderator and gaseous CO2 as coolant was installed in the A1 plant. Outlet steam from primary reactor coolant system with the temperature of 410 C was led to 6 modules of steam generators and from there to turbine generators. Refueling was carried out on-line at plant full power. The first serious incident associated with refueling occurred in 1976 when a locking mechanism at a fuel assembly failed. The core was not damaged during that incident and following a reconstruction of the damaged technology channel, the plant continued in operation. However, serious problems were occurring with the integrity of steam generators (CO2 gas on primary side, water and steam on secondary side) when the plant had to be shut down frequently due to failures and subsequent repairs. The second serious accident occurred in 1977 when a fuel assembly was overheated with a subsequent release of D2O into gas cooling circuit due to a human failure in the course of replacement of a fuel assembly. Subsequent rapid increase in humidity of the primary system resulted in damages of fuel elements in the core and the primary system was contaminated by fission products. In-reactor structures had been damaged, too. Activity had penetrated also into certain parts of the secondary system via leaking steam generators. Radiation situation in the course of both events on the plant site and around it had been below the level of limits specified. Based on a technical and economical justification of the demanding character of equipment repairs for the restoration ...
Date: February 26, 2002
Creator: Stubna, M.; Pekar, A.; Moravek, J. & Spirko, M.
Partner: UNT Libraries Government Documents Department

Anomalies in Proposed Regulations for the Release of Redundant Material from Nuclear and Non-nuclear Industries

Description: Now that increasing numbers of nuclear power stations are reaching the end of their commercially useful lives, the management of the large quantities of very low level radioactive material that arises during their decommissioning has become a major subject of discussion, with very significant economic implications. Much of this material can, in an environmentally advantageous manner, be recycled for reuse without radiological restrictions. Much larger quantities--2-3 orders of magnitude larger--of material, radiologically similar to the candidate material for recycling from the nuclear industry, arise in non-nuclear industries like coal, fertilizer, oil and gas, mining, etc. In such industries, naturally occurring radioactivity is artificially concentrated in products, by-products or waste to form TENORM (Technologically Enhanced Naturally Occurring Radioactive Material). It is only in the last decade that the international community has become aware of the prevalence of T ENORM, specially the activity levels and quantities arising in so many nonnuclear industries. The first reaction of international organizations seems to have been to propose ''double'' standards for the nuclear and non-nuclear industries, with very stringent release criteria for radioactive material from the regulated nuclear industry and up to a hundred times more liberal criteria for the release/exemption of TENORM from the as yet unregulated non-nuclear industries. There are, however, many significant strategic issues that need to be discussed and resolved. An interesting development, for both the nuclear and non-nuclear industries, is the increased scientific scrutiny that the populations of naturally high background dose level areas of the world are being subject to. Preliminary biological studies have indicated that the inhabitants of such areas, exposed to many times the permitted occupational doses for nuclear workers, have not shown any differences in cancer mortality, life expectancy, chromosome aberrations or immune function, in comparison with those living in normal background areas. The paper discusses these ...
Date: February 26, 2002
Creator: Menon, S.
Partner: UNT Libraries Government Documents Department

Recent Trends in the Adequacy of Nuclear Plant Decommissioning Funding

Description: Concerned about the potential cost and sufficiency of funds to decommission the nation's nuclear power plants, the Congress asked the U.S. General Accounting Office (GAO) to assess the adequacy, as of December 31, 1997, of electric utilities'; funds to eventually decommission their plants. GAO's report (GAO/RCED-99-75) on this issue addressed three alternative assumption scenarios--baseline (most likely), optimistic, and pessimistic; and was issued in May 1999. This paper updates GAO's baseline assessment of fund adequacy in 1997, and extends the analysis through 2000. In 2000, we estimate that the present value cost to decommission the nation's nuclear plants is about $35 billion; utility fund balances are about $29 billion. Both our two measures of funding adequacy for utilities are on average not only much above ideal levels, but also overall have greatly improved since 1997. However, certain utilities still show less than ideal fund balances and annual contributions. We suggest that the range of these results among the individual utilities is a more important policy measure to assess the adequacy of decommissioning funding than is the funding adequacy for the industry as a whole.
Date: February 26, 2002
Creator: Williams, D. G.
Partner: UNT Libraries Government Documents Department

Preparation for Early Termination of Ignalina NPP Operation

Description: Seimas (Parliament of Lithuania) approved updated National Energy strategy where it is indicated that first Unit will be shutdown before the year 2005 and second Unit in 2009 if funding for decommissioning is available from EU and other donors. In accordance to Ignalina NPP Unit 1 Closure Law the Government of Lithuania approved the Ignalina NPP Unit 1 Decommissioning Program until year 2005. For enforcement of this program, the plan of measures for implementation of the program was prepared and approved by the Minister of Economy. The plan consists of two parts, namely technical- environmental and social-economic. Technical-environmental measures are mostly oriented to the safe management of spent nuclear fuel and operational radioactive waste stored at the plant and preparation of licensing documents for Unit 1 decommissioning. Social-economic measures are oriented to mitigate negative social and economic impact on Lithuania, inhabitants of the region, and, particularly, o n the staff of Ignalina NPP by means of creating favorable conditions for a balanced social and economic development of the region. In this paper analysis of planned activities, licensing requirements for decommissioning, progress in preparation of the Final Decommissioning Plan is discussed.
Date: February 26, 2003
Creator: Poskas, P. & Poskas, R.
Partner: UNT Libraries Government Documents Department

Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

Description: In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on a production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration in Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste ...
Date: February 26, 2002
Creator: Kudinov, K. G.; Tretyakov, A. A.; Sorokin, Yu. P.; Bondin, V. V.; Manakova, L. F. & Jardine, L. J.
Partner: UNT Libraries Government Documents Department

Removal of Retired Alkali Metal Test Systems

Description: This paper describes the successful effort to remove alkali metals, alkali metal residues, and piping and structures from retired non-radioactive test systems on the Hanford Site. These test systems were used between 1965 and 1982 to support the Fast Flux Test Facility and the Liquid Metal Fast Breeder Reactor Program. A considerable volume of sodium and sodium-potassium alloy (NaK) was successfully recycled to the commercial sector; structural material and electrical material such as wiring was also recycled. Innovative techniques were used to safely remove NaK and its residues from a test system that could not be gravity-drained. The work was done safely, with no environmental issues or significant schedule delays.
Date: February 26, 2003
Creator: Brehm, W. F.; Church, W. R. & Biglin, J. W.
Partner: UNT Libraries Government Documents Department

Maine Yankee: Making the Transition from an Operating Plant to an Independent Spent Fuel Storage Installation (ISFSI)

Description: The purpose of this paper is to describe the challenges faced by Maine Yankee Atomic Power Company in making the transition from an operating nuclear power plant to an Independent Spent Fuel Storage Installation (ISFSI). Maine Yankee (MY) is a 900-megawatt Combustion Engineering pressurized water reactor whose architect engineer was Stone & Webster. Maine Yankee was put into commercial operation on December 28, 1972. It is located on an 820-acre site, on the shores of the Back River in Wiscasset, Maine about 40 miles northeast of Portland, Maine. During its operating life, it generated about 1.2 billion kilowatts of power, providing 25% of Maine's electric power needs and serving additional customers in New England. Maine Yankee's lifetime capacity factor was about 67% and it employed more than 450 people. The decision was made to shutdown Maine Yankee in August of 1997, based on economic reasons. Once this decision was made planning began on how to accomplish safe and cost effective decommissioning of the plant by 2004 while being responsive to the community and employees.
Date: February 26, 2002
Creator: Norton, W. & McGough, M. S.
Partner: UNT Libraries Government Documents Department

Environmental Remediation Data Management Tools

Description: Computer software tools for data management can improve site characterization, planning and execution of remediation projects. This paper discusses the use of two such products that have primarily been used within the nuclear power industry to enhance the capabilities of radiation protection department operations. Advances in digital imaging, web application development and programming technologies have made development of these tools possible. The Interactive Visual Tour System (IVTS) allows the user to easily create and maintain a comprehensive catalog containing digital pictures of the remediation site. Pictures can be cataloged in groups (termed ''tours'') that can be organized either chronologically or spatially. Spatial organization enables the user to ''walk around'' the site and view desired areas or components instantly. Each photo is linked to a map (floor plan, topographical map, elevation drawing, etc.) with graphics displaying the location on the map and any available tour/component links. Chronological organization enables the user to view the physical results of the remediation efforts over time. Local and remote management teams can view these pictures at any time and from any location. The Visual Survey Data System (VSDS) allows users to record survey and sample data directly on photos and/or maps of areas and/or components. As survey information is collected for each area, survey data trends can be reviewed for any repetitively measured location or component. All data is stored in a Quality Assurance (Q/A) records database with reference to its physical sampling point on the site as well as other information to support the final closeout report for the site. The ease of use of these web-based products has allowed nuclear power plant clients to plan outage work from their desktop and realize significant savings with respect to dose and cost. These same tools are invaluable for remediation and decommissioning planning of any scale ...
Date: February 26, 2002
Creator: Wierowski, J. V.; Henry, L. G. & Dooley, D. A.
Partner: UNT Libraries Government Documents Department

Immobilization of Fast Reactor First Cycle Raffinate

Description: This paper describes the results of work to bring forward the timing for the immobilization of first cycle raffinate from reprocessing fuel from the Dounreay Prototype Fast Reactor (PFR). First cycle raffinate is the liquor which contains > 99% of the fission products separated from spent fuel during reprocessing. Approximately 203 m3 of raffinate from the reprocessing of PFR fuel is held in four tanks at the UKAEA's site at Dounreay, Scotland. Two methods of immobilization of this high level waste (HLW) have been considered: vitrification and cementation. Vitrification is the standard industry practice for the immobilization of first cycle raffinate, and many papers have been presented on this technique elsewhere. However, cementation is potentially feasible for immobilizing first cycle raffinate because the heat output is an order of magnitude lower than typical HLW from commercial reprocessing operations such as that at the Sellafield site in Cumbria, England. In fact, it falls within the upper end of the UK definition of intermediate level waste (ILW). Although the decision on which immobilization technique will be employed has yet to be made, initial development work has been undertaken to identify a suitable cementation formulation using inactive simulant of the raffinate. An approach has been made to the waste disposal company Nirex to consider the disposability of the cemented product material. The paper concentrates on the process development work that is being undertaken on cementation to inform the decision making process for selection of the immobilization method.
Date: February 26, 2003
Creator: Langley, K. F.; Partridge, B. A. & Wise, M.
Partner: UNT Libraries Government Documents Department

Emptying of the Storage for Solid Radioactive Waste in the Greifswald Nuclear Power Plant

Description: On the Greifswald site, 8 WWER 440 reactor units are located and also several facilities to handle fuel and radwaste. After the reunification of Germany, the final decision was taken to decommission all these Russian designed reactors. Thus, EWN is faced with a major decommissioning project in the field of nuclear power stations. One of the major tasks before the dismantling of the plant is the complete disposal of the operational waste. Among other facilities, a store for solid radioactive waste is located on the site, which has been filled over 17 years of operation of units 1 to 4. The paper presents the disposal technology development and results achieved. This activity is the first project in the operational history of the Russian type serial reactor line WWER-440.
Date: February 26, 2002
Creator: Hartmann, B. & Fischer, J.
Partner: UNT Libraries Government Documents Department