WorldWideScience

Sample records for reactor decommissioning components

  1. Reactor decommissioning

    International Nuclear Information System (INIS)

    Lawton, H.

    1984-01-01

    A pioneering project on the decommissioning of the Windscale Advanced Gas-cooled Reactor, by the UKAEA, is described. Reactor data; policy; waste management; remote handling equipment; development; and recording and timescales, are all briefly discussed. (U.K.)

  2. Thermal aging of some decommissioned reactor components and methodology for life prediction

    International Nuclear Information System (INIS)

    Chung, H.M.

    1989-03-01

    Since a realistic aging of cast stainless steel components for end-of-life or life-extension conditions cannot be produced, it is customary to simulate the thermal aging embrittlement by accelerated aging at ∼400 degree C. In this investigation, field components obtained from decommissioned reactors have been examined after service up to 22 yr to provide a benchmark of the laboratory simulation. The primary and secondary aging processes were found to be identical to those of the laboratory-aged specimens, and the kinetic characteristics were also similar. The extent of the aging embrittlement processes and other key factors that are known to influence the embrittlement kinetics have been compared for the decommissioned reactor components and materials aged under accelerated conditions. On the basis of the study, a mechanistic understanding of the causes of the complex behavior in kinetics and activation energy of aging (i.e., the temperature dependence of aging embrittlement between the accelerated and reactor-operating conditions) is presented. A mechanistic correlation developed thereon is compared with a number of available empirical correlations to provide an insight for development of a better methodology of life prediction of the reactor components. 18 refs., 18 figs., 5 tabs

  3. Packaging, Transportation, and Disposal Logistics for Large Radioactively Contaminated Reactor Decommissioning Components

    International Nuclear Information System (INIS)

    Lewis, Mark S.

    2008-01-01

    The packaging, transportation and disposal of large, retired reactor components from operating or decommissioning nuclear plants pose unique challenges from a technical as well as regulatory compliance standpoint. In addition to the routine considerations associated with any radioactive waste disposition activity, such as characterization, ALARA, and manifesting, the technical challenges for large radioactively contaminated components, such as access, segmentation, removal, packaging, rigging, lifting, mode of transportation, conveyance compatibility, and load securing require significant planning and execution. In addition, the current regulatory framework, domestically in Titles 49 and 10 and internationally in TS-R-1, does not lend itself to the transport of these large radioactively contaminated components, such as reactor vessels, steam generators, reactor pressure vessel heads, and pressurizers, without application for a special permit or arrangement. This paper addresses the methods of overcoming the technical and regulatory challenges. The challenges and disposition decisions do differ during decommissioning versus component replacement during an outage at an operating plant. During decommissioning, there is less concern about critical path for restart and more concern about volume reduction and waste minimization. Segmentation on-site is an available option during decommissioning, since labor and equipment will be readily available and decontamination activities are routine. The reactor building removal path is also of less concern and there are more rigging/lifting options available. Radionuclide assessment is necessary for transportation and disposal characterization. Characterization will dictate the packaging methodology, transportation mode, need for intermediate processing, and the disposal location or availability. Characterization will also assist in determining if the large component can be transported in full compliance with the transportation

  4. Decommissioning a nuclear reactor

    International Nuclear Information System (INIS)

    Montoya, G.M.

    1991-01-01

    The process of decommissioning a facility such as a nuclear reactor or reprocessing plant presents many waste management options and concerns. Waste minimization is a primary consideration, along with protecting a personnel and the environment. Waste management is complicated in that both radioactive and chemical hazardous wastes must be dealt with. This paper presents the general decommissioning approach of a recent project at Los Alamos. Included are the following technical objectives: site characterization work that provided a thorough physical, chemical, and radiological assessment of the contamination at the site; demonstration of the safe and cost-effective dismantlement of a highly contaminated and activated nuclear-fuelded reactor; and techniques used in minimizing radioactive and hazardous waste. 12 figs

  5. TA-2 Water Boiler Reactor Decommissioning Project

    International Nuclear Information System (INIS)

    Durbin, M.E.; Montoya, G.M.

    1991-06-01

    This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m 3 of low-level solid radioactive waste and 35 m 3 of mixed waste. 15 refs., 25 figs., 3 tabs

  6. The potential for reducing the radiological consequences of reactor decommissioning through selection of construction materials for activated components

    International Nuclear Information System (INIS)

    Woollam, P.B.

    1984-08-01

    This report considers whether it may be possible to reduce the radiological consequences of reactor decommissioning by careful attention to the specification of the elemental concentration of materials used in the reactor's construction. In particular, consideration is given to the potential for reduction of the concentration of elements known to activate to long lived daughter isotopes. Two particular areas are addressed, both applied to Sizewell 'B' PWR. The first is the choice of raw materials for the construction of the concrete bioshield to minimise future waste arisings. The second is the specification of some trace element concentrations in the steel pressure vessel and reactor internal structures to minimise personnel exposure at decommissioning time. The report presents extensive analyses of many of the candidate raw materials for Sizewell 'B' concrete, including PFA, and derives the radiological consequences for the eventual disposal of these materials to a hypothetical municipal land fill waste site. Data are also presented on the concentrations of important elements activating to gamma emitting daughters in type 304 stainless steels, leading to an assessment of likely dose equivalent rates at decommissioning time from the pressure vessel and from the internal components. (author)

  7. The brief introduction to decommissioning of nuclear reactor projects

    International Nuclear Information System (INIS)

    Zhao Shixin

    1991-01-01

    The basic concept and procedure of the decommissioning of nuclear reactor project and the three stages of decommissioning defined by IAEA are introduced. The main work of decommissioning of nuclear reactor are as following: (1) the documentary and technological preparation; (2) the site preparation of decommissioning project; (3) the dismantling of equipment piping system and components; (4) the decontamination of the piping system before and after decomminssioning; (5) the storage and disposal of the operational and decommissioning waste

  8. The brief introduction to decommissioning of nuclear reactor projects

    Energy Technology Data Exchange (ETDEWEB)

    Shixin, Zhao [Beijing Inst. of Nuclear Engineering (China)

    1991-08-01

    The basic concept and procedure of the decommissioning of nuclear reactor project and the three stages of decommissioning defined by IAEA are introduced. The main work of decommissioning of nuclear reactor are as following: (1) the documentary and technological preparation; (2) the site preparation of decommissioning project; (3) the dismantling of equipment piping system and components; (4) the decontamination of the piping system before and after decomminssioning; (5) the storage and disposal of the operational and decommissioning waste.

  9. Decommissioning strategy for reactor AM, Russian Federation

    International Nuclear Information System (INIS)

    Suvorov, A.P.; Mukhamadeev, R.I.

    2002-01-01

    This paper presents the results of studies into the various aspects of decommissioning the oldest Russian research reactor, the AM reactor. Experimental and calculation results of a study to determine the inventory of long lived radioactive materials at the AM reactor are presented, along with a comparison to comparable data for other similar reactors. An analysis, by calculation, of the decay time needed to allow manual dismantling of the reactor vessel and stack, without remote operated equipment, defined it as 90 years. The possibility of burning most of the irradiated graphite to decrease the amount of long lived radioactive wastes was confirmed. The problems associated with the dismantling of the reactor components, contaminated with radioactive corrosion products, were analyzed. A decommissioning strategy for reactor AM was formed which is deferred dismantling, placing most of the radiological areas into long term safe enclosure. An overall decommissioning plan for reactor AM is given. (author)

  10. Decommissioning three nuclear reactors at Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Montoya, G.M.; Salazar, M.

    1992-01-01

    Three nuclear reactors, including the historic water boiler reactor, were decommissioned at Los Alamos National Laboratory (LANL). The decommissioning of the facilities involved removing the reactors and their associated components. Planning for the decommissioning operation included characterizing the facilities, estimating the costs of decommissioning operations, preparing environmental documentation, establishing systems to track costs and work progress, and preplanning to correct health and safety concerns in each facility

  11. Planning of the BN-350 reactor decommissioning

    International Nuclear Information System (INIS)

    Klepikov, A.Kh.; Tazhibayeva, I.L.; Zhantikin, T.M.; Baldov, A.N.; Nazarenko, P.I.; Koltyshev, S.M.; Wells, P.B.

    2002-01-01

    The experimental and commercial BN-350 NPP equipped with a fast neutron sodium cooled reactor is located in Kazakhstan near the Aktau city on the Caspian Sea coast. It was commissioned in 1973 and intended for weapon-grade plutonium production and as stream supply to a water desalination facility and the turbines of the Mangyshlak Atomic Energy Complex. Taking into account technical, financial and political issues, the Government of Kazakhstan enacted the Decree no. 456 'On Decommissioning of the Reactor BN-350 in the Aktau City of the Mangystau Region'. Because the decision on reactor decommissioning was adopted before the end of scheduled operation (2003), the plan to decommission the BN-350 reactor has not yet been developed. To determine the activities required for ensuring reactor safety and in preparation for decommission in the period prior, the development and ensuring approval by the Republic of Kazakhstan Government of the decommissioning plan, a 'Plan of Priority Actions for BN-350 Reactor Decommissioning' was developed and approved. Actions provided for in the plan include the following: Development of BN-350 Reactor Decommissioning Plan; Accident prevention during the period of transition; Unloading nuclear fuel from reactor and draining the coolant from the heat exchange circuits. Decommission is defined as a complex of administrative and technical actions taken to allow the removal of some or all of regulatory controls over a nuclear facility. These actions involve decontamination, dismantling and removal of radioactive materials, waste, components and structures. They are carried out to achieve a progressive and systematic reduction in radiological hazards and are undertaken on the basis of planning and assessment in order to ensure safety decommissioning operations. In accordance with the decision of Kazakhstan Government, three basic stages for BN-350 reactor decommissioning are envisaged: First stage - Placement of BN-350 into long-term storage

  12. Decommissioning of the Neuherberg Research Reactor (FRN)

    International Nuclear Information System (INIS)

    Demmeler, M.; Rau, G.; Strube, D.

    1982-01-01

    The Neuherberg Research Reactor is of type TRIGA MARK III with 1 MW steady state power and pulsable up to 2000 MW. During more than ten years of operation 12000 MWh and 6000 reactor pulses had been performed. In spite of its good technical condition and of permanent safe operation without any failures, the decommissioning of the Neuherberg research reactor was decided by the GSF board of directors to save costs for maintaining and personnel. As the mode of decommissioning the safe enclosure was chosen which means that the fuel elements will be transferred back to the USA. All other radioactive reactor components will be enclosed in the reactor block. Procedures for licensing of the decommissioning, dismantling procedures and time tables are presented

  13. Decommissioning activities for Salaspils research reactor - 59055

    International Nuclear Information System (INIS)

    Abramenkovs, A.; Malnacs, J.

    2012-01-01

    In May 1995, the Latvian government decided to shut down the Salaspils Research Reactor (SRR). The reactor is out of operation since July 1998. A conceptual study for the decommissioning of SRR has been carried out by Noell-KRC-Energie- und Umwelttechnik GmbH at 1998-1999. The Latvian government decided to start the direct dismantling to 'green field' in October 26, 1999. The upgrade of decommissioning and dismantling plan was performed in 2003-2004 years, which change the main goal of decommissioning to the 'brown field'. The paper deals with the SRR decommissioning experience during 1999-2010. The main decommissioning stages are discussed including spent fuel and radioactive wastes management. The legal aspects and procedures for decommissioning of SRR are described in the paper. It was found, that the involvement of stakeholders at the early stages significantly promotes the decommissioning of nuclear facility. Radioactive waste management's main efforts were devoted to collecting and conditioning of 'historical' radioactive wastes from different storages outside and inside of reactor hall. All radioactive materials (more than 96 tons) were conditioned in concrete containers for disposal in the radioactive wastes repository 'Radons' at Baldone site. The dismantling of contaminated and activated components of SRR systems is discussed in paper. The cementation of dismantled radioactive wastes in concrete containers is discussed. Infrastructure of SRR, including personal protective and radiation measurement equipment, for decommissioning purposes was upgraded significantly. Additional attention was devoted to the free release measurement's technique. The certified laboratory was installed for supporting of all decommissioning activities. All non-radioactive equipments and materials outside of reactor buildings were released for clearance and dismantled for reusing or conventional disposing. Weakly contaminated materials from reactor hall were collected

  14. Decommissioning of fast reactors after sodium draining

    International Nuclear Information System (INIS)

    2009-11-01

    Acknowledging the importance of passing on knowledge and experience, as well mentoring the next generation of scientists and engineers, and in response to expressed needs by Member States, the IAEA has undertaken concrete steps towards the implementation of a fast reactor data retrieval and knowledge preservation initiative. Decommissioning of fast reactors and other sodium bearing facilities is a domain in which considerable experience has been accumulated. Within the framework and drawing on the wide expertise of the Technical Working Group on Fast Reactors (TWG-FR), the IAEA has initiated activities aiming at preserving the feedback (lessons learned) from this experience and condensing those to technical recommendations on fast reactor design features that would ease their decommissioning. Following a recommendation by the TWG-FR, the IAEA had convened a topical Technical Meeting (TM) on 'Operational and Decommissioning Experience with Fast Reactors', hosted by CEA, Centre d'Etudes de Cadarache, France, from 11 to 15 March 2002 (IAEA-TECDOC- 1405). The participants in that TM exchanged detailed technical information on fast reactor operation and decommissioning experience with various sodium cooled fast reactors, and, in particular, reviewed the status of the various decommissioning programmes. The TM concluded that the decommissioning of fast reactors to reach safe enclosure presented no major difficulties, and that this had been accomplished mainly through judicious adaptation of processes and procedures implemented during the reactor operation phase, and the development of safe sodium waste treatment processes. However, the TM also concluded that, on the path to achieving total dismantling, challenges remain with regard to the decommissioning of components after sodium draining, and suggested that a follow-on TM be convened, that would provide a forum for in-depth scientific and technical exchange on this topic. This publication constitutes the Proceedings of

  15. Study on the decommissioning of research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Doo Hwan; Jun, Kwan Sik; Choi, Yoon Dong; Lee, Tae Yung; Kwon, Sang Woon; Lee, Jong Il [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-01-01

    Currently, KAERI operates TRIGA Mark-II and TRIGA Mark-III research reactors as a general purpose research and training facility. As these are, however, situated at Seoul office site of KAERI which is scheduled to be transferred to KEPCO as well as 30 MW HANARO research reactor which is expected to reach the first criticality in 1995 is under construction at head site of KAERI, decommissioning of TRIGA reactors has become an important topic. The objective of this study is to prepare and present TRIGA facility decontamination and decommissioning plan. Estimation of the radioactive inventory in TRIGA research reactor was carried out by the use of computational method. In addition, summarized in particular were the methodologies associated with decontamination, segmenting processes for activated metallic components, disposition of wastes. Particular consideration in this study was focused available technology applicable to decommissioning of TRIGA research reactor. State-of-the-art summaries of the available technology for decommissioning presented here will serve a useful document for preparations for decommissioning in the future. 6 figs, 41 tabs, 30 refs. (Author).

  16. UK reactor decommissioning strategy

    International Nuclear Information System (INIS)

    Woollam, P.B.

    2004-01-01

    With the cessation of electricity generation, nuclear power stations move into the next stage of the overall life cycle of the facility: decommissioning. Decommissioning is defined as the process whereby a nuclear facility, at the end of its economic life, is taken permanently out of service and its site made available for other purposes. This involves the implementation of a structured and safe programme for dismantling and clearing the site and making it available for alternative use in the future. In practical terms, 'decommissioning' means the systematic and progressive reduction of hazards to the point where the site could eventually be de-licensed. (author)

  17. Shippingport station decommissioning project irradiated components transfer: Topical report

    International Nuclear Information System (INIS)

    1988-01-01

    This topical report is a synopsis of the transfer of irradiated components into the Shippingport Reactor Pressure Vessel (RPV) performed at the Shippingport Station Decommissioning Project (SSDP). The information is provided as a part of the Technology Transfer Program to document the preparation activities for the decommissioning of a nuclear power reactor to be removed in one piece

  18. Decommissioning of reactor facilities (2). Required technology

    International Nuclear Information System (INIS)

    Yanagihara, Satoshi

    2014-01-01

    Decommissioning of reactor facilities was planned to perform progressive dismantling, decontamination and radioactive waste disposal with combination of required technology in a safe and economic way. This article outlined required technology for decommissioning as follows: (1) evaluation of kinds and amounts of residual radioactivity of reactor facilities with calculation and measurement, (2) decontamination technology of metal components and concrete structures so as to reduce worker's exposure and production of radioactive wastes during dismantling, (3) dismantling technology of metal components and concrete structures such as plasma arc cutting, band saw cutting and controlled demolition with mostly remote control operation, (3) radioactive waste disposal for volume reduction and reuse, and (4) project management of decommissioning for safe and rational work to secure reduction of worker's exposure and prevent the spreading of contamination. (T. Tanaka)

  19. Decommissioning of Salaspils nuclear reactor

    International Nuclear Information System (INIS)

    Abramenkovs, A.; Malnachs, J.; Popelis, A.

    2002-01-01

    In May 1995, the Latvian Government decided to shut down the Research Reactor Salaspils (SRR) and to dispense with nuclear energy in future. The reactor has been out of operation since July 1998. A conceptual study for the decommissioning of SRR has been carried out by Noell-KRC-Energie- und Umwelttechnik GmbH from 1998-1999. he Latvian Government decided on 26 October 1999 to start the direct dismantling to 'green field' in 2001. The results of decommissioning and dismantling performed in 1999-2001 are presented and discussed. The main efforts were devoted to collecting and conditioning 'historical' radioactive waste from different storages outside and inside the reactor hall. All radioactive material more than 20 tons were conditioned in concrete containers for disposal in the radioactive waste depository 'Radons' in the Baldone site. Personal protective and radiation measurement equipment was upgraded significantly. All non-radioactive equipment and material outside the reactor buildings were free-released and dismantled for reuse or conventional disposal. Weakly contaminated material from the reactor hall was collected and removed for free-release measurements. The technology of dismantling of the reactor's systems, i.e. second cooling circuit, zero power reactors and equipment, is discussed in the paper. (author)

  20. Nuclear data requirements for fission reactor decommissioning

    International Nuclear Information System (INIS)

    Kocherov, N.P.

    1993-01-01

    The meeting was attended by 13 participants from 8 Member States and 2 International Organizations who reviewed the status of the nuclear data libraries and computer codes used to calculate the radioactive inventory in the reactor unit components for the decommissioning purposes. Nuclides and nuclear reactions important for determination of the radiation fields during decommissioning and for the final disposal of radioactive waste from the decommissioned units were identified. Accuracy requirements for the relevant nuclear data were considered. The present publication contains the text of the reports by the participants and their recommendations to the Nuclear Data Section of the IAEA. A separate abstract was prepared for each of these reports. Refs, figs and tabs

  1. Decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Walton, G.R.

    1993-01-01

    The Tokamak Fusion Test Reactor (TFTR) at Princeton Plasma Physics Laboratory (PPPL) will complete its experimental lifetime with a series of deuterium-tritium pulses in 1994. As a result, the machine structures will become radioactive, and vacuum components will also be contaminated with tritium. Dose rate levels will range from less than 1 mr/h for external structures to hundreds of mr/h for the vacuum vessel. Hence, decommissioning operations will range from hands on activities to the use of remotely operated equipment. After 21 months of cool down, decontamination and decommissioning (D and D) operations will commence and continue for approximately 15 months. The primary objective is to render the test cell complex re-usable for the next machine, the Tokamak Physics Experiment (TPX). This paper presents an overview of decommissioning TFTR and discusses the D and D objectives

  2. Decommissioning of Salaspils Research Reactor

    International Nuclear Information System (INIS)

    Abramenkovs, A.; Popelis, A.; Abramenkova, G

    2008-01-01

    The Salaspils Research Reactor (SRR) is out of operation since July 1998 and the decommissioning of SRR was started in 1999 according to the decision of the Government of Latvia. The main decommissioning activities up to 2006 were connected with collecting and conditioning of historical radioactive wastes from different storages outside and inside of reactor hall. The total amount of dismantled materials was about 700 tons, more than 77 tons were conditioned in concrete containers for disposal in repository. The radioactive wastes management technology is discussed in the paper. It was found, that additional efforts must be spent for immobilization of radionuclides in cemented matrix to be comply with the wastes acceptance criteria. The investigations of mechanical stability of water-cement matrix are described and discussed in the paper

  3. Decommissioning of multiple-reactor stations: facilitation by sequential decommissioning

    International Nuclear Information System (INIS)

    Moore, E.B.; Smith, R.I.; Wittenbrock, N.G.

    1982-01-01

    Reductions in cost and radiation dose can be achieved for decommissionings at multiple reactor stations because of factors not necessarily present at a single reactor station: reactors of similar design, the opportunity for sequential decommissioning, a site dedicated to nuclear power generation, and the option of either interim or permanent low-level radioactive waste storage facilities onsite. The cost and radiation dose reductions occur because comprehensive decommissioning planning need only be done once, because the labor force is stable and need only be trained once, because there is less handling of radioactive wastes, and because central stores, equipment, and facilities may be used. The cost and radiation dose reductions are sensitive to the number and types of reactors on the site, and to the alternatives selected for decommissioning. 3 tables

  4. Decommissioning the UHTREX Reactor Facility at Los Alamos, New Mexico

    International Nuclear Information System (INIS)

    Salazar, M.; Elder, J.

    1992-08-01

    The Ultra-High Temperature Reactor Experiment (UHTREX) facility was constructed in the late 1960s to advance high-temperature and gas-cooled reactor technology. The 3-MW reactor was graphite moderated and helium cooled and used 93% enriched uranium as its fuel. The reactor was run for approximately one year and was shut down in February 1970. The decommissioning of the facility involved removing the reactor and its associated components. This document details planning for the decommissioning operations which included characterizing the facility, estimating the costs of decommissioning, preparing environmental documentation, establishing a system to track costs and work progress, and preplanning to correct health and safety concerns in the facility. Work to decommission the facility began in 1988 and was completed in September 1990 at a cost of $2.9 million. The facility was released to Department of Energy for other uses in its Los Alamos program

  5. Planning the Decommissioning of Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Podlaha, J., E-mail: pod@ujv.cz [Nuclear Research Institute Rez, 25068 Rez (Czech Republic)

    2013-08-15

    In the Czech Republic, three research nuclear reactors are in operation. According to the valid legislation, preliminary decommissioning plans have been prepared for all research reactors in the Czech Republic. The decommissioning plans shall be updated at least every 5 years. Decommissioning funds have been established and financial resources are regularly deposited. Current situation in planning of decommissioning of research reactors in the Czech Republic, especially planning of decommissioning of the LVR-15 research reactor is described in this paper. There appeared new circumstances having wide impact on the decommissioning planning of the LVR-15 research reactor: (1) Shipment of spent fuel to the Russian Federation for reprocessing and (2) preparation of processing of radioactive waste from reconstruction of the VVR-S research reactor (now LVR-15 research reactor). The experience from spent fuel shipment to the Russian Federation and from the process of radiological characterization and processing of radioactive waste from reconstruction of the VVR-S research reactor (now the LVR-15 research reactor) and the impact on the decommissioning planning is described in this paper. (author)

  6. Cost estimation for decommissioning of research reactors

    International Nuclear Information System (INIS)

    Grossi, Pablo Andrade; Tello, Cledola Cassia Oliveira de; Segabinaze, Roberto de Oliveira; Daniska, Vladimir

    2013-01-01

    In the case of research reactors, the limited data that is available tends to provide only overall decommissioning costs, without any breakdown of the main cost elements. In order to address this subject, it is important to collect and analyse all available data of decommissioning costs for the research reactors. The IAEA has started the DACCORD Project focused on data analysis and costing of research reactors decommissioning. Data collection is organized in accordance with the International Structure for Decommissioning Costing (ISDC), developed jointly by the IAEA, the OECD Nuclear Energy Agency and the European Commission. The specific aims of the project include the development of representative and comparative data and datasets for preliminary costing for decommissioning. This paper will focus on presenting a technique to consider several representative input data in accordance with the ISDC structure and using the CERREX (Cost Estimation for Research Reactors in Excel) software developed by IAEA. (author)

  7. US DOE Idaho national laboratory reactor decommissioning

    International Nuclear Information System (INIS)

    Szilagyi, Andrew

    2012-01-01

    The United States Department of Energy (DOE) primary contractor, CH2M-WG Idaho was awarded the cleanup and deactivation and decommissioning contract in May 2005 for the Idaho National Lab (INL). The scope of this work included dispositioning over 200 Facilities and 3 Reactors Complexes (Engineering Test Reactor (ETR), Materials Test Reactor (MTR) and Power Burst Facility (PBF) Reactor). Two additional reactors were added to the scope of the contract during the period of performance. The Zero Power Physics Reactor (ZPPR) disposition was added under a separate subcontractor with the INL lab contractor and the Experimental Breeder Reactor II (EBR-II) disposition was added through American Recovery and Reinvestment Act (ARRA) Funding. All of the reactors have been removed and disposed of with the exception of EBR-II which is scheduled for disposition approximately March of 2012. A brief synopsis of the 5 reactors is provided. For the purpose of this paper the ZPPR reactor due to its unique design as compared to the other four reactors, and the fact that is was relatively lightly contaminated and irradiated will not be discussed with the other four reactors. The ZPPR reactor was readily accessible and was a relatively non-complex removal as compared to the other reactors. Additionally the EBR-II reactor is currently undergoing D and D and will have limited mention in this paper. Prior to decommissioning the reactors, a risk based closure model was applied. This model exercised through the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA), Non-Time Critical Removal Action (NTCRA) Process which evaluated several options. The options included; No further action - maintain as is, long term stewardship and monitoring (mothball), entombment in place and reactor removal. Prior to commencing full scale D and D, hazardous constituents were removed including cadmium, beryllium, sodium (passivated and elemental), PCB oils and electrical components, lead

  8. Decommissioning of the CANDU-PHW reactor

    International Nuclear Information System (INIS)

    Unsworth, G.N.

    1977-04-01

    This report contains the results of a study of various aspects of decommissioning of reactors. The study places in perspective the size of the job, the hazards involved, the cost and the environmental impact. The three internationally agreed ''stages'' of decommissioning, namely, mothballing, entombment, and dismantling are defined and discussed. The single unit 600 MW(e) CANDU is chosen as the type of reactor on which the discussion is focussed but the conclusions reached will provide a basis for judgement of the costs and problems associated with decommissioning reactors of other sizes and types. (author)

  9. Waste Management During RA Reactor Decommissioning

    International Nuclear Information System (INIS)

    Markovic, M.; Avramovic, I.

    2008-01-01

    The objective of radioactive waste management during the RA reactor decommissioning is to deal with radioactive waste in a manner that protects human health and the environment now and in the future. The estimation of waste quantities to be expected during decommissioning is a very important step in the initial planning. (author)

  10. Facilitation of decommissioning light water reactors

    International Nuclear Information System (INIS)

    Moore, E.B. Jr.

    1979-12-01

    Information on design features, special equipment, and construction methods useful in the facilitation of decommissioning light water reactors is presented. A wide range of facilitation methods - from improved documentation to special decommissioning tools and techniques - is discussed. In addition, estimates of capital costs, cost savings, and radiation dose reduction associated with these facilitation methods are given

  11. Characterization and management of radioactive sodium and other reactor components as input data for the decommissioning of liquid metal-cooled fast reactors. A compilation of data produced of data produced by members of the IAEA technical working group on fast reactors (TWG-FR) at two consultancies and one technical committee meeting. Working material

    International Nuclear Information System (INIS)

    2002-01-01

    A number of liquid metal cooled fast reactors (LMFRs) are in operation and, some have already been shut down; other reactors will reach the end of their design lifetime in a few years and become candidates for decommissioning. It is unfortunate that little consideration was devoted to decommissioning of reactors at the plant design and construction stage. It is with this focus that the Technical Working Group on Fast Reactors (TWGFR) recommended that the IAEA organize the exchange of information on LMFRs decommissioning technology. It was pointed out that the decommissioning of small sodium-cooled reactors has shown that there are two basic differences between thermal and fast reactors decommissioning: on the one side, the treatment and disposal of radioactive sodium coolant, and on the other side, the management of reactor components, for which the structural materials are activated in depth by fast neutrons. To this end, a Technical Committee Meeting on Sodium Removal and Disposal from LMFRs in Normal Operation and in the framework of Decommissioning (Aix-en-Provence, France, November 1997) and two Consultancies on Decommissioning of the Kazakh BN-350 LMFR (Vienna, Austria, October 1996; Obninsk, Russian Federation, February 1998) were convened by the IAEA. These Meetings brought together a group of experts from France, Russia, Kazakhstan, the UK, and the USA to exchange information on, and to review current technical knowledge and experience in the management of radioactive coolant and reactor components following closing of LMFRs, as well as their design features and operating experience relevant for decommissioning procedures. The report provides general and detailed information on activation characteristics of the primary coolant; treatment and disposal of the spent sodium; removal of the residual sodium deposits and decontamination; the activation characteristics of the reactor components and the management of the latter. The recurring theme is finding

  12. Decommissioning and decontrolling the R1-reactor

    International Nuclear Information System (INIS)

    Bergman, C.; Holmberg, B.T.

    1985-01-01

    Sweden's first nuclear reactor - the research reactor R1 - situated in bedrock under the Royal Technical Institute of Stockholm, has in the period 1981-1983 been subject to a complete decommissioning. The National Institute for Radiation Protection has followed the work in detail, and has after the completion of the decommissioning performed measurements of radioactivity on site. The report gives an account of the work the Institute has done in preparation for- and during decommissioning and specifically report on the measurements for classification of the local as free for non-nuclear use. (aa)

  13. Regulatory aspects of nuclear reactor decommissioning

    International Nuclear Information System (INIS)

    Ross, W.M.

    1990-01-01

    The paper discusses the regulatory aspects of decommissioning commercial nuclear power stations in the UK. The way in which the relevant legislation has been used for the first time in dealing with the early stages of decommissioning commercial nuclear reactor is described. International requirements and how they infit with the UK system are also covered. The discussion focusses on the changes which have been required, under the Nuclear Site Licence, to ensure that the licensee carries out of work of reactor decommissioning in a safe and controlled manner. (Author)

  14. Status of the RA research reactor decommissioning project

    International Nuclear Information System (INIS)

    Ljubenov, V.; Nikolic, D.; Pesic, M.; Milosevic, M.; Kostic, Lj.; Steljic, M.; Sotic, O.; Antic, D. . E-mail address of corresponding author: vladan@vin.bg.ac.yu; Ljubenov, V.)

    2005-01-01

    The 6.5 MW heavy water RA research reactor at the VINCA Institute of Nuclear Sciences operated from 1959 to 1984. After 18 years of extended shutdown in 2002 it was decided that the reactor shutdown should be final. Preliminary decommissioning activities have been initiated by the end of 2002 under the Technical Co-operation Programme of the International Atomic Energy Agency. The objective of the project is to implement safe, timely and cost-effective decommissioning of the RA reactor up to unrestricted use of the site. Decommissioning project is closely related to two other projects: Safe Removal of the RA Reactor Spent Nuclear Fuel and Radioactive Waste Management in VINCA Institute. The main phases of the project include preparation of the detailed decommissioning plan, radiological characterization of the reactor site, dismantling and removal of the reactor components and structures, decontamination, final radiological site survey and the documentation of all the activities in order to obtain the approval for unrestricted use of the facility site. In this paper a review of the activities related to the preparation and realization of the RA reactor decommissioning project is given. Status of the project's organizational and technical aspects as for July 2004 are presented and plans for the forthcoming phases of the project realization are outlined. (author)

  15. Safety in decommissioning of research reactors

    International Nuclear Information System (INIS)

    1986-01-01

    This Guide covers the technical and administrative considerations relevant to the nuclear aspects of safety in the decommissioning of reactors, as they apply to the reactor and the reactor site. While the treatment, transport and disposal of radioactive wastes arising from decommissioning are important considerations, these aspects are not specifically covered in this Guide. Likewise, other possible issues in decommissioning (e.g. land use and other environmental issues, industrial safety, financial assurance) which are not directly related to radiological safety are also not considered. Generally, decommissioning will be undertaken after planned final shutdown of the reactor. In some cases a reactor may have to be decommissioned following an unplanned or unexpected event of a series or damaging nature occurring during operation. In these cases special procedures for decommissioning may need to be developed, peculiar to the particular circumstances. This Guide could be used as a basis for the development of these procedures although specific consideration of the circumstances which create the need for them is beyond its scope

  16. Planning and management for reactor decommissioning

    International Nuclear Information System (INIS)

    Miyasaka, Yasuhiko

    2001-01-01

    This report describes decommissioning strategy, planning process, regulation, management and organization, radiological characterization and safety. Planning is used to identify, define and organize the requirements for decommissioning including decommissioning options, items to be accomplished (objective, scope), to solve problems of how it is to be accomplished (methods, means and procedures), questions of who will execute it (resources, organization and responsibilities, interfacing), and time when it will be executed (schedule for meeting the objectives). A plan is highly dependent on the quality of the management team assembled to carry it out. Radiological characterization involves a survey of existing data, calculation, in situ measurements and/or sampling and analyses. Using this databases decommissioning planner may assess options, considering: decontamination processes, dismantling procedures, tools required, radiological protection of workers and public/environment, waste classification, and resulting costs. Comparison and optimization of these factors will lead to selection of a decommissioning strategy, i.e. typically, immediate or deferred dismantling. The planning and implementation of decommissioning for nuclear reactors should be referred both recent dismantling techniques and many decommissioning experiences. The technical lessons learned from many projects will help in the planning for future decommissioning projects. And systematic planning and management are essential to successful completion of a decommissioning project. (author)

  17. The decommissioning of a small nuclear reactor

    International Nuclear Information System (INIS)

    Neset, K.; Christensen, G.C.; Lundby, J.E.; Roenneberg, G.A.

    1990-02-01

    The JEEP II reactor at Kjeller, Norway has been used as a model for a study of the decommissioning of a small research reactor. A radiological survey is given and a plan for volume reducing, packaging, certifying, classifying and shipping of the radioactive waste is described. 23 refs., 4 figs

  18. Decommissioning of TRIGA Mark II type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dooseong; Jeong, Gyeonghwan; Moon, Jeikwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    The first research reactor in Korea, KRR 1, is a TRIGA Mark II type with open pool and fixed core. Its power was 100 kWth at its construction and it was upgraded to 250 kWth. Its construction was started in 1957. The first criticality was reached in 1962 and it had been operated for 36,000 hours. The second reactor, KRR 2, is a TRIGA Mark III type with open pool and movable core. These reactors were shut down in 1995, and the decision was made to decommission both reactors. The aim of the decommissioning activities is to decommission the KRR 2 reactor and decontaminate the residual building structures and site, and to release them as unrestricted areas. The KRR 1 reactor was decided to be preserve as a historical monument. A project was launched for the decommissioning of these reactors in 1997, and approved by the regulatory body in 2000. A total budget for the project was 20.0 million US dollars. It was anticipated that this project would be completed and the site turned over to KEPCO by 2010. However, it was discovered that the pool water of the KRR 1 reactor was leaked into the environment in 2009. As a result, preservation of the KRR 1 reactor as a monument had to be reviewed, and it was decided to fully decommission the KRR 1 reactor. Dismantling of the KRR 1 reactor takes place from 2011 to 2014 with a budget of 3.25 million US dollars. The scope of the work includes licensing of the decommissioning plan change, removal of pool internals including the reactor core, removal of the thermal and thermalizing columns, removal of beam port tubes and the aluminum liner in the reactor tank, removal of the radioactive concrete (the entire concrete structure will not be demolished), sorting the radioactive waste (concrete and soil) and conditioning the radioactive waste for final disposal, and final statuses of the survey and free release of the site and building, and turning over the site to KEPCO. In this paper, the current status of the TRIGA Mark-II type reactor

  19. Risk Management of Large Component in Decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    Nah, Kyung Ku; Kim, Tae Ryong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-10-15

    The need for energy, especially electric energy, has been dramatically increasing in Korea. Therefore, a rapid growth in nuclear power development has been achieved to have about 30% of electric power production. However, such a large nuclear power generation has been producing a significant amount of radioactive waste and other matters such as safety issue. In addition, owing to the severe accidents at the Fukushima in Japan, public concerns regarding NPP and radiation hazard have greatly increased. In Korea, the operation of KORI 1 has been scheduled to be faced with end of lifetime in several years and Wolsong 1 has been being under review for extending its life. This is the reason why the preparation of nuclear power plant decommissioning is significant in this time. Decommissioning is the final phase in the life-cycle of a nuclear facility and during decommissioning operation, one of the most important management in decommissioning is how to deal with the disused large component. Therefore, in this study, the risk in large component in decommissioning is to be identified and the key risk factor is to be analyzed from where can be prepared to handle decommissioning process safely and efficiently. Developing dedicated acceptance criteria for large components at disposal site was analyzed as a key factor. Acceptance criteria applied to deal with large components like what size of those should be and how to be taken care of during disposal process strongly affect other major works. For example, if the size of large component was not set up at disposal site, any dismantle work in decommissioning is not able to be conducted. Therefore, considering insufficient time left for decommissioning of some NPP, it is absolutely imperative that those criteria should be laid down.

  20. Risk Management of Large Component in Decommissioning

    International Nuclear Information System (INIS)

    Nah, Kyung Ku; Kim, Tae Ryong

    2014-01-01

    The need for energy, especially electric energy, has been dramatically increasing in Korea. Therefore, a rapid growth in nuclear power development has been achieved to have about 30% of electric power production. However, such a large nuclear power generation has been producing a significant amount of radioactive waste and other matters such as safety issue. In addition, owing to the severe accidents at the Fukushima in Japan, public concerns regarding NPP and radiation hazard have greatly increased. In Korea, the operation of KORI 1 has been scheduled to be faced with end of lifetime in several years and Wolsong 1 has been being under review for extending its life. This is the reason why the preparation of nuclear power plant decommissioning is significant in this time. Decommissioning is the final phase in the life-cycle of a nuclear facility and during decommissioning operation, one of the most important management in decommissioning is how to deal with the disused large component. Therefore, in this study, the risk in large component in decommissioning is to be identified and the key risk factor is to be analyzed from where can be prepared to handle decommissioning process safely and efficiently. Developing dedicated acceptance criteria for large components at disposal site was analyzed as a key factor. Acceptance criteria applied to deal with large components like what size of those should be and how to be taken care of during disposal process strongly affect other major works. For example, if the size of large component was not set up at disposal site, any dismantle work in decommissioning is not able to be conducted. Therefore, considering insufficient time left for decommissioning of some NPP, it is absolutely imperative that those criteria should be laid down

  1. Decommissioning of the Northrop TRIGA reactor

    International Nuclear Information System (INIS)

    Cozens, George B.; Woo, Harry; Benveniste, Jack; Candall, Walter E.; Adams-Chalmers, Jeanne

    1986-01-01

    An overview of the administrative and operational aspects of decommissioning and dismantling the Northrop Mark F TRIGA Reactor, including: planning and preparation, personnel requirements, government interfacing, costs, contractor negotiations, fuel shipments, demolition, disposal of low level waste, final survey and disposition of the concrete biological shielding. (author)

  2. Decommissioning of the Salaspils Research Reactor

    Directory of Open Access Journals (Sweden)

    Abramenkovs Andris

    2011-01-01

    Full Text Available In May 1995, the Latvian government decided to shut down the Salaspils Research Reactor and to dispense with nuclear energy in the future. The reactor has been out of operation since July 1998. A conceptual study on the decommissioning of the Salaspils Research Reactor was drawn up by Noell-KRC-Energie- und Umwelttechnik GmbH in 1998-1999. On October 26th, 1999, the Latvian government decided to start the direct dismantling to “green-field” in 2001. The upgrading of the decommissioning and dismantling plan was carried out from 2003-2004, resulting in a change of the primary goal of decommissioning. Collecting and conditioning of “historical” radioactive wastes from different storages outside and inside the reactor hall became the primary goal. All radioactive materials (more than 96 tons were conditioned for disposal in concrete containers at the radioactive wastes depository “Radons” at the Baldone site. Protective and radiation measurement equipment of the personnel was upgraded significantly. All non-radioactive equipment and materials outside the reactor buildings were released for clearance and dismantled for reuse or conventional disposal. Contaminated materials from the reactor hall were collected and removed for clearance measurements on a weekly basis.

  3. Decommissioning of a small reactor (BR3 reactor, Belgium)

    International Nuclear Information System (INIS)

    Dadoumont, J.; Massaut, V.; Klein, M.; Demeulemeester, Y.

    2002-01-01

    Since 1989, SCK-CEN has been dismantling its PWR reactor BR3 (Belgian Reactor No. 3). After gaining a great deal of experience in remote dismantling of highly radioactive components during the actual dismantling of the two sets of internals, the BR3 team completed the cutting of its reactor pressure vessel (RPV). During the feasibility phase of the RPV dismantling, a decision was made to cut it under water in the refuelling pool of the plant, after having removed it from its cavity. The RPV was cut into segments using a milling cutter and a bandsaw machine. These mechanical techniques have shown their ability for this kind of operations. Prior to the segmentation, the thermal insulation situated around the RPV was remotely removed and disposed of. The paper will describe all these operations. The BR3 decommissioning activities also include the dismantling of contaminated loops and equipment. After a careful sorting of the pieces, optimized management routes are selected in order to minimize the final amount of radioactive waste to be disposed of. Some development of different methods of decontamination were carried out: abrasive blasting (or sand blasting), chemical decontamination (Oxidizing-Reducing process using Cerium). The main goal of the decontamination program is to recycle most of the metallic materials either in the nuclear world or in the industrial world by reaching the respective recycling or clearance level. Overall the decommissioning of the BR3 reactor has shown the feasibility of performing such a project in a safe and economical way. Moreover, BR3 has developed methodologies and decontamination processes to economically reduce the amount of radwaste produced. (author)

  4. 78 FR 64028 - Decommissioning of Nuclear Power Reactors

    Science.gov (United States)

    2013-10-25

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0035] Decommissioning of Nuclear Power Reactors AGENCY... the NRC's regulations relating to the decommissioning process for nuclear power reactors. The revision... Commission (NRC) is issuing Revision 1 of regulatory guide (RG) 1.184 ``Decommissioning of Nuclear Power...

  5. Cost Estimation for Research Reactor Decommissioning

    International Nuclear Information System (INIS)

    2013-01-01

    One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world'. One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property.' The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. The purpose of this publication is to develop a costing methodology and a software tool in order to support cost estimation for research reactor decommissioning. The costing methodology is intended for the preliminary cost estimation stages for research reactor decommissioning with limited inventory data and other input data available. Existing experience in decommissioning costing is considered. As the basis for the cost calculation structure, the costing model uses the International Structure for Decommissioning Costing (ISDC) that is recommended by the IAEA, the Organisation for

  6. Reactor vessel decommissioning project. Final report

    International Nuclear Information System (INIS)

    Schoonen, D.H.

    1984-09-01

    This report describes a reactor vessel decommissioning project; it documents and explains the project objectives, scope, performance results, and sodium removal process. The project was successfully completed in FY-1983, within budget and without significant problems or adverse impact on the environment. Waste generated by the operation included the reactor vessel, drained sodium, and liquid, solid, and gaseous wastes which were significantly less than project estimates. Personnel radiation exposures were minimized, such that the project total was one-half the predicted exposure level. Except for the sodium removed, the material remaining in the reactor vessel is essentially the same as when the vessel arrived for processing

  7. Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Perry, E.; Chrzanowski, J.; Gentile, C.; Parsells, R.; Rule, K.; Strykowsky, R.; Viola, M.

    2003-01-01

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D and D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D and D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D and D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget

  8. The Preliminary Decommissioning Plan of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lam, Pham Van; Vien, Luong Ba; Vinh, Le Vinh; Nghiem, Huynh Ton; Tuan, Nguyen Minh; Phuong, Pham Hoai [Nuclear Research Institute, Da Lat (Viet Nam)

    2013-08-15

    Recently, after 25 years of operation, a preliminary decommissioning plan for the Dalat Nuclear Research Reactor (DNRR) has been produced but as yet it has not been implemented due to the continued operations of the reactor. However, from the early phases of facility design and construction and during operation, the aspects that facilitate decommissioning process have been considered. This paper outlines the DNRR general description, the organization that manages the facility, the decommissioning strategy and associated project management, and the expected decommissioning activities. The paper also considers associated cost and funding, safety and environmental issues and waste management aspects amongst other considerations associated with decommissioning a nuclear research reactor. (author)

  9. Decontamination and decommissioning techniques for research reactors

    International Nuclear Information System (INIS)

    Oh, Won Zin; Won, H. J.; Jung, C. H.; Choi, W. K.; Kim, G. N.; Lee, K. W.

    2002-05-01

    Evaluation of soil decontamination process and the liquid decontamination waste treatment technology are investigation of organic acid as a decontamination agent, investigation of the liquid waste purification process and identification of recycling the decontamination agents. Participation on IAEA CRP meeting are preparation of IAEA technical report on 'studies on decommissioning of TRIGA reactors and site restoration technologies' and exchange the research result, technology, experience and safety regulation of the research reactor D and D of USA, Great Britain, Canada, Belgium, Italy, India and so forth

  10. Decommissioning technology development for research reactors

    International Nuclear Information System (INIS)

    Lee, K. W.; Kim, S. K.; Kim, Y. K.

    2004-03-01

    Although it is expected that the decommissioning of a nuclear power plant will happen since 2020, the need of partial decommissioning and decontamination for periodic inspection and life extension has been on an increasing trend and domestic market has gradually been extended. Therefore, in this project the decommissioning DB system on the KRR-1 and 2 was developed as establishing the information classification system of the research reactor dismantling and the structural design and optimization of the decommissioning DB system. Also in order to secure the reliability and safety about the dismantling process, the main dismantling simulation technology that can verify the dismantling process before their real dismantling work was developed. And also the underwater cutting equipment was developed to remove these stainless steel parts highly activated from the RSR. First, the its key technologies were developed and then the design, making, and capability analysis were performed. Finally the actual proof was achieved for applying the dismantling site. an automatic surface contamination measuring equipment was developed in order to get the sample automatically and measure the radiation/radioactivity

  11. Initial decommissioning planning for the Budapest research reactor

    Directory of Open Access Journals (Sweden)

    Toth Gabor

    2011-01-01

    Full Text Available The Budapest Research Reactor is the first nuclear research facility in Hungary. The reactor is to remain in operation for at least another 13 years. At the same time, the development of a decommissioning plan is a mandatory requirement under national legislation. The present paper describes the current status of decommissioning planning which is aimed at a timely preparation for the forthcoming decommissioning of the reactor.

  12. Computer System Analysis for Decommissioning Management of Nuclear Reactor

    International Nuclear Information System (INIS)

    Nurokhim; Sumarbagiono

    2008-01-01

    Nuclear reactor decommissioning is a complex activity that should be planed and implemented carefully. A system based on computer need to be developed to support nuclear reactor decommissioning. Some computer systems have been studied for management of nuclear power reactor. Software system COSMARD and DEXUS that have been developed in Japan and IDMT in Italy used as models for analysis and discussion. Its can be concluded that a computer system for nuclear reactor decommissioning management is quite complex that involved some computer code for radioactive inventory database calculation, calculation module on the stages of decommissioning phase, and spatial data system development for virtual reality. (author)

  13. Research reactor back-end options - decommissioning: a necessary consideration

    International Nuclear Information System (INIS)

    England, M.R.; Parry, D.R.; Smith, C.

    1998-01-01

    Decommissioning is a challenge, which all radioactive site licensees eventually need to face and research reactors are no exception. BNFL has completed numerous major decommissioning projects at its own operational sites and has undertaken similar works at customers' sites including the decommissioning of the Universities Research Reactor (URR), Risley and the ICI TRIGA 1-Mk I Reactor at Billingham. Based on the execution of such projects BNFL has gained an understanding of the variety of customer requirements and the effectiveness of specific decommissioning techniques for research reactors. This paper addresses factors to be considered when reviewing the way forward following shut down and how these affect the final decisions for fuel management and the extent of decommissioning. Case studies are described from BNFL's recent experience decommissioning both the URR and ICI TRIGA reactors. (author)

  14. Reactor decommissioning in a deregulated market

    International Nuclear Information System (INIS)

    Beverridge, George; Cooper, T.

    2002-01-01

    Full text: Deregulation of the electricity markets in North America and Western Europe has had many profound effects on the electric utilities and the nuclear industry. Deregulation has led to cost transparency, increased competition, and a drive by the utilities to reduce costs in order to maintain market share and margins. In the context of this more competitive and dynamic market having a clear picture of decommissioning liabilities and their successful discharge has a material impact on the financial performance of a utility. This paper will summarise BNFL Environmental Services' experience with regard to its experience in both the planning and implementation phases of a reactor decommissioning project. In particular it will demonstrate how commercial projects in crucial areas of strategy development, project implementation and site restoration, can be combined with an approach that is both commercial and innovative to reduce the risks to a utility. This paper sets out to demonstrate this viewpoint. (author)

  15. Plan for Moata reactor decommissioning, ANSTO

    International Nuclear Information System (INIS)

    Kim, S.

    2003-01-01

    'Moata' is an Argonaut type 100 kW reactor that was operated by Australian Nuclear Science and Technology Organisation for 34 years from 1961 to 1995. It was initially used as a reactor-physics research tool and a training reactor but the scope of operations was extended to include activation analysis and neutron radiography from the mid 1970s. In 1995, the Moata reactor was shutdown on the grounds that its continued operation could no longer be economically justified. All the fuel (HEU) was unloaded to temporary storage and secured in 1995, followed by drainage of the demineralised water (primary coolant) from the reactor in 1996 and complete removal of electrical cables in 1998. The Reactor Control Room has been renovated into a modern laboratory. The reactor structure is still intact and kept under safe storage. Various options for decommissioning strategies have been considered and evaluated. So far, 'Immediate Dismantling' is considered to be the most desirable option, however, the timescale for actual dismantling needs to take account of the establishment of the national radioactive repository. This paper describes the dismantling options and techniques considered along with examples of other dismantling projects overseas. (author)

  16. Planning for the decommissioning of a research reactor

    International Nuclear Information System (INIS)

    Dodson, W.J.; Isakari, H.H.; Munro, J.F.; Lim, T.H.; Denton, M.M.; Vernig, P.G.

    1988-01-01

    This paper describes the steps that must be taken and the uncertainties and potential pitfalls that can be encountered in decommissioning a research reactor, whether owned by private industry, a university, or a government agency. The paper is based on the experience in preparing for decommissioning the TRIGA Mark III Berkeley Research Reactor (BRR). Six topics of interest to an owner-operator are addressed: task and schedule planning, decommissioning organization, cost estimating, health and safety considerations, waste management, and regulatory concerns

  17. Decommissioning of Swedish nuclear power reactors. Technology and costs

    International Nuclear Information System (INIS)

    1994-06-01

    The main topics discussed are planning, technology and costs of decommissioning nuclear power reactors. Oskarshamn-3 (BWR) and Ringhals-4 (PWR) have been used as reference reactors. 29 refs, figs, tabs

  18. Technology issues for decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Walton, G.R.

    1994-01-01

    The approach for decommissioning the Tokamak Fusion Test Reactor has evolved from a conservative plan based on cutting up and burying all of the systems, to one that considers the impact tritium contamination will have on waste disposal, how large size components may be used as their own shipping containers, and even the possibility of recycling the materials of components such as the toroidal field coils and the tokamak structure. In addition, the project is more carefully assessing the requirements for using remotely operated equipment. Finally, valuable cost database is being developed for future use by the fusion community

  19. Windscale advanced gas-cooled reactor (WAGR) decommissioning project overview

    International Nuclear Information System (INIS)

    Pattinson, A.

    2003-01-01

    The current BNFL reactor decommissioning projects are presented. The projects concern power reactor sites at Berkely, Trawsfynydd, Hunterstone, Bradwell, Hinkley Point; UKAEA Windscale Pile 1; Research reactors within UK Scottish Universities at East Kilbride and ICI (both complete); WAGR. The BNFL environmental role include contract management; effective dismantling strategy development; implementation and operation; sentencing, encapsulation and transportation of waste. In addition for the own sites it includes strategy development; baseline decommissioning planning; site management and regulator interface. The project objectives for the Windscale Advanced Gas-Cooled Reactor (WAGR) are 1) Safe and efficient decommissioning; 2) Building of good relationships with customer; 3) Completion of reactor decommissioning in 2005. The completed WAGR decommissioning campaigns are: Operational Waste; Hot Box; Loop Tubes; Neutron Shield; Graphite Core and Restrain System; Thermal Shield. The current campaign is Lower Structures and the remaining are: Pressure vessel and Insulation; Thermal Columns and Outer Vault Membrane. An overview of each campaign is presented

  20. Some studies related to decommissioning of nuclear reactors

    International Nuclear Information System (INIS)

    Bergman, C.; Menon, S.

    1990-02-01

    Decommissioning of large nuclear reactors has not yet taken place in the Nordic countries. Small nuclear installations, however, have been dismantled. This NKA-programme has dealt with some interesting and important factors which have to be analysed before a large scale decommissioning programme starts. Prior to decommissioning, knowledge is required regarding the nuclide inventory in various parts of the reactor. Measurements were performed in regions close to the reactor tank and the biological shield. These experimental data are used to verify theoretical calculations. All radioactive waste generated during decommissioning will have to be tansported to a repository. Studies show that in all the Nordic countries there are adequate transport systems with which decommissioning waste can be transported. Another requirement for orderly decommissioning planning is that sufficient information about the plant and its operation history must be available. It appears that if properly handled and sorted, all such information can be extracted from existing documentation. (authors)

  1. Large packages for reactor decommissioning waste

    International Nuclear Information System (INIS)

    Price, M.S.T.

    1991-01-01

    This study was carried out jointly by the Atomic Energy Establishment at Winfrith (now called the Winfrith Technology Centre), Windscale Laboratory and Ove Arup and Partners. The work involved the investigation of the design of large transport containers for intermediate level reactor decommissioning waste, ie waste which requires shielding, and is aimed at European requirements (ie for both LWR and gas cooled reactors). It proposes a design methodology for such containers covering the whole lifetime of a waste disposal package. The design methodology presented takes account of various relevant constraints. Both large self shielded and returnable shielded concepts were developed. The work was generic, rather than specific; the results obtained, and the lessons learned, remain to be applied in practice

  2. Present status of research reactor decommissioning programme in Indonesia

    International Nuclear Information System (INIS)

    Suripto, A.; Mulyanto, N.

    2002-01-01

    At present Indonesia has 3 research reactors, namely the 30 MW MTR-type multipurpose reactor at Serpong Site, two TRIGA-type research reactors, the first one being 1 MW located at Bandung Site and the second one a small reactor of 100 kW at Yogyakarta Site. The TRIGA Reactor at the Bandung Site reached its first criticality at 250 kW in 1964, and then was operated at 1000 kW since 1971. In October 2000 the reactor power was successfully upgraded to 2 MW. This reactor has already been operated for 38 years. There is not yet any decision for the decommissioning of this reactor. However it will surely be an object for the near future decommissioning programme and hence anticipation for the above situation becomes necessary. The regulation on decommissioning of research reactor is already issued by the independent regulatory body (BAPETEN) according to which the decommissioning permit has to be applied by the BATAN. For Indonesia, an early decommissioning strategy for research reactor dictates a restricted re-use of the site for other nuclear installation. This is based on high land price, limited availability of radwaste repository site, and other cost analysis. Spent graphite reflector from the Bandung TRIGA reactor is recommended for a direct disposal after conditioning, without any volume reduction treatment. Development of human resources, technological capability as well as information flow from and exchange with advanced countries are important factors for the future development of research reactor decommissioning programme in Indonesia. (author)

  3. Reactor decommissioning strategy: a new start for BNFL

    International Nuclear Information System (INIS)

    Woollam, P.; Nurden, P.

    2001-01-01

    The key points of BNFL Magnox Electric's revised waste management and reactor decommissioning strategy for the reactor sites are enlisted. Reactors will be defuelled as soon as practicable after shutdown. Predominantly Caesium contaminated plant will be dismantled when it is no longer needed. Cobalt contaminated plant such as boilers will remain in position until the reactors are dismantled, but appropriate decontamination technology will be regularly reviewed. All buildings except the reactor buildings will be dismantled as soon as practicable after they are no longer needed. Operational ILW, except some activated components, will be retrieved and packaged during the Care and Maintenance preparation period. All wastes will be stored on site, and handled in the long term in accordance with Government policy. Reactor buildings and their residual contents will be placed in a passive safe storage Care and Maintenance condition in a manner appropriate for the site. Contaminated land will be managed to maintain public safety. The reactors will be finally dismantled in a sequenced programme with a start date and duration to be decided at the appropriate time in the light of circumstances prevalent at that time. Currently, the Company is considering a sequenced programme across all sites, notionally beginning around 100 years from station shutdown, leading to a range of deferral periods. For provisioning purposes, the Company has costed a strategy involving reactor dismantling deferrals ranging from 85 to about 105 years in order to demonstrate prudent provisioning to meet its liabilities. A risk provision to reflect the potential for shorter deferral periods is included in the cost estimates. The end point for reactor decommissioning is site clearance and delicensing, based on the assumption that a reasonably practicable interpretation of the 'no danger' clause in the Nuclear Installations Act 1965 (as amended) can be developed. In line with Government policy, and taking

  4. Decontamination and decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Walton, G.R.; Perry, E.D.; Commander, J.C.; Spampinato, P.T.

    1994-01-01

    The Tokamak Fusion Test Reactor (TFTR) is scheduled to complete its end-of-life deuterium-tritium (D-T) experiments in September 1994. The D-T operation will result in the TFTR machine structure becoming activated, and plasma facing and vacuum components will be contaminated with tritium. The resulting machine activation levels after a two year cooldown period will allow hands on dismantling for external structures, but require remote dismantling for the vacuum vessel. The primary objective of the Decontamination and Decommissioning (D ampersand D) Project is to provide a facility for construction of a new Department of Energy (DOE) experimental fusion reactor by March 1998. The project schedule calls for a two year shutdown period when tritium decontamination of the vacuum vessel, neutral beam injectors and other components will occur. Shutdown will be followed by an 18 month period of D ampersand D operations. The technical objectives of the project are to: safely dismantle and remove components from the test cell complex; package disassembled components in accordance with applicable regulations; ship packages to a DOE approved disposal or material recycling site; and develop expertise using remote disassembly techniques on a large scale fusion facility. This paper discusses the D ampersand D objectives, the facility to be decommissioned, and the technical plan that will be implemented

  5. IAEA/CRP for decommissioning techniques for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Won Zin; Won, H. J.; Kim, K. N.; Lee, K. W.; Jung, C. H

    2001-03-01

    The following were studied through the project entitled 'IAEA/CRP for decommissioning techniques for research reactors 1. Decontamination technology development for TRIGA radioactive soil waste - Electrokinetic soil decontamination experimental results and its mathematical simulation 2. The 2nd IAEA/CRP for decommissioning techniques for research reactors - Meeting results and program 3. Hosting the 2001 IAEA/RCA D and D training course for research reactors and small nuclear facilities.

  6. IAEA/CRP for decommissioning techniques for research reactors

    International Nuclear Information System (INIS)

    Oh, Won Zin; Won, H. J.; Kim, K. N.; Lee, K. W.; Jung, C. H.

    2001-03-01

    The following were studied through the project entitled 'IAEA/CRP for decommissioning techniques for research reactors 1. Decontamination technology development for TRIGA radioactive soil waste - Electrokinetic soil decontamination experimental results and its mathematical simulation 2. The 2nd IAEA/CRP for decommissioning techniques for research reactors - Meeting results and program 3. Hosting the 2001 IAEA/RCA D and D training course for research reactors and small nuclear facilities

  7. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR. PART 6 - PRESENTATION OF THE DECOMMISSIONING DEVICE

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2015-05-01

    Full Text Available The objective of this paper is to present a possible solution for the designing of a device for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. The decommissioning activities are dismantling, demolition, controlled removal of equipment, components, conventional or hazardous waste (radioactive, toxic in compliance with the international basic safety standards on radiation protection. One as the most important operation in the final phase of the nuclear reactor dismantling is the decommissioning of fuel channels. For the fuel channels decommissioning should be taken into account the detailed description of the fuel channel and its components, the installation documents history, adequate radiological criteria for decommissioning guidance, safety and environmental impact assessment, including radiological and non-radiological analysis of the risks that can occur for workers, public and environment, the description of the proposed program for decommissioning the fuel channel and its components, the description of the quality assurance program and of the monitoring program, the equipments and methods used to verify the compliance with the decommissioning criteria, the planning of performing the final radiological assessment at the end of the fuel channel decommissioning. These will include also, a description of the proposed radiation protection procedures to be used during decommissioning. The dismantling of the fuel channel is performed by one device which shall provide radiation protection during the stages of decommissioning, ensuring radiation protection of the workers. The device shall be designed according to the radiation protection procedures. The decommissioning device assembly of the fuel channel components is composed of the device itself and moving platform support for coupling of the selected channel to be dismantled. The fuel channel decommissioning device is an autonomous device designed for

  8. Nuclear data in the problem of fission reactor decommissioning

    International Nuclear Information System (INIS)

    Manokhin, V.N.; Kulagin, N.T.

    1993-01-01

    This report presents a review of the works published in Russia during last several years and devoted to the problem of nuclear data and calculations of nuclear facilities activation for fission reactor decommissioning. 6 refs

  9. A Comparative Perspective on Reactor Decommissioning

    International Nuclear Information System (INIS)

    Devgun, J.S.; Zelmer, R.

    2006-01-01

    A comparative perspective on decommissioning, based on facts and figures as well as the national policies, is useful in identifying mutually beneficial 'lessons learned' from various decommissioning programs. In this paper we provide such a perspective on the US and European approaches based on a review of the programmatic experience and the decommissioning projects. The European countries selected for comparison, UK, France, and Germany, have nuclear power programs comparable in size and vintage to the US program but have distinctly different policies at the federal level. The national decommissioning scene has a lot to do with how national nuclear energy policies are shaped. Substantial experience exists in all decommissioning programs and the technology is in a mature state. Substantial cost savings can result from sharing of decommissioning information, technologies and approaches among various programs. However, the Achilles' heel for the decommissioning industry remains the lack of appropriate disposal facilities for the nuclear wastes. (authors)

  10. IPR-R1 TRIGA research reactor decommissioning plan

    International Nuclear Information System (INIS)

    Andrade Grossi, Pablo; Oliveira de Tello, Cledola Cassia; Mesquita, Amir Zacarias

    2008-01-01

    The International Atomic Energy Agency (IAEA) is concerning to establish or adopt standards of safety for the protection of health, life and property in the development and application of nuclear energy for peaceful purposes. In this way the IAEA recommends that decommissioning planning should be part of all radioactive installation licensing process. There are over 200 research reactors that have either not operated for a considerable period of time and may never return to operation or, are close to permanent shutdown. Many countries do not have a decommissioning policy, and like Brazil not all installations have their decommissioning plan as part of the licensing documentation. Brazil is signatory of Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management, but until now there is no decommissioning policy, and specifically for research reactor there is no decommissioning guidelines in the standards. The Nuclear Technology Development Centre (CDTN/CNEN) has a TRIGA Mark I Research Reactor IPR-R1 in operation for 47 years with 3.6% average fuel burn-up. The original power was 100 k W and it is being licensed for 250 k W, and it needs the decommissioning plan as part of the licensing requirements. In the paper it is presented the basis of decommissioning plan, an overview and the end state / final goal of decommissioning activities for the IPR-R1, and the Brazilian ongoing activities about this subject. (author)

  11. Contingency Cost estimation for Research reactor Decommissioning

    International Nuclear Information System (INIS)

    Jin, Hyung Gon; Hong, Yun Jeong

    2016-01-01

    There are many types of cost items in decommissioning cost estimation, however, contingencies are for unforeseen elements of cost within the defined project scope. Regulatory body wants to reasonable quantification for this issue. Many countries have adopted the breakdown of activity dependent and period-dependent costs to structure their estimates. Period-dependent costs could be broken down into defined time frames to reduce overall uncertainties. Several countries apply this notion by having different contingency factors for different phases of the project. This study is a compilation of contingency cost of research reactor and for each country. Simulation techniques using TRIM, MATLAB, and PSpice can be useful tools for designing detector channels. Thus far TRIM, MATLAB and PSpice have been used to calculate the detector current output pulse for SiC semiconductor detectors and to model the pulses that propagate through potential detector channels. This model is useful for optimizing the detector and the resolution for application to neutron monitoring in the Generation IV power reactors

  12. Contingency Cost estimation for Research reactor Decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Hyung Gon; Hong, Yun Jeong [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    There are many types of cost items in decommissioning cost estimation, however, contingencies are for unforeseen elements of cost within the defined project scope. Regulatory body wants to reasonable quantification for this issue. Many countries have adopted the breakdown of activity dependent and period-dependent costs to structure their estimates. Period-dependent costs could be broken down into defined time frames to reduce overall uncertainties. Several countries apply this notion by having different contingency factors for different phases of the project. This study is a compilation of contingency cost of research reactor and for each country. Simulation techniques using TRIM, MATLAB, and PSpice can be useful tools for designing detector channels. Thus far TRIM, MATLAB and PSpice have been used to calculate the detector current output pulse for SiC semiconductor detectors and to model the pulses that propagate through potential detector channels. This model is useful for optimizing the detector and the resolution for application to neutron monitoring in the Generation IV power reactors.

  13. Decommissioning the Los Alamos Molten Plutonium Reactor Experiment (LAMPRE I)

    International Nuclear Information System (INIS)

    Harper, J.R.; Garde, R.

    1981-11-01

    The Los Alamos Molten Plutonium Reactor Experiment (LAMPRE I) was decommissioned at the Los Alamos National Laboratory, Los Alamos, New Mexico, in 1980. The LAMPRE I was a sodium-cooled reactor built to develop plutonium fuels for fast breeder applications. It was retired in the mid-1960s. This report describes the decommissioning procedures, the health physics programs, the waste management, and the costs for the operation

  14. Communications programme for the RA nuclear reactor decommission

    International Nuclear Information System (INIS)

    Milanovic, S.; Antic, D.

    2002-01-01

    During the decommissioning of the RA research nuclear reactor at the VINCA Institute of Nuclear Sciences, an adequate number of radiation and contamination surveys should be conduced to assure radiological safety of the workers, the public and the environment. Public would like to know more about the nuclear and radiological safety. The communications programme defines the ways to informing the public, its representatives and the information media about the health and safety aspects of the activities during the RA nuclear reactor decommission. (author)

  15. Regulatory Framework for Controlling the Research Reactor Decommissioning Project

    International Nuclear Information System (INIS)

    Melani, Ai; Chang, Soon Heung

    2009-01-01

    Decommissioning is one of important stages in construction and operation of research reactors. Currently, there are three research reactors operating in Indonesia. These reactors are operated by the National Nuclear Energy Agency (BATAN). The age of the three research reactors varies from 22 to 45 years since the reactors reached their first criticality. Regulatory control of the three reactors is conducted by the Nuclear Energy Regulatory Agency (BAPETEN). Controlling the reactors is carried out based on the Act No. 10/1997 on Nuclear Energy, Government Regulations and BAPETEN Chairman Decrees concerning the nuclear safety, security and safeguards. Nevertheless, BAPETEN still lack of the regulation, especially for controlling the decommissioning project. Therefore, in the near future BAPETEN has to prepare the regulations for decommissioning, particularly to anticipate the decommissioning of the oldest research reactors, which probably will be done in the next ten years. In this papers author give a list of regulations should be prepared by BAPETEN for the decommissioning stage of research reactor in Indonesia based on the international regulatory practice

  16. Peach Bottom HTGR decommissioning and component removal

    International Nuclear Information System (INIS)

    Kohler, E.J.; Steward, K.P.; Iacono, J.V.

    1977-07-01

    The prime objective of the Peach Bottom End-of-Life Program was to validate specific HTGR design codes and predictions by comparison of actual and predicted physics, thermal, fission product, and materials behavior in Peach Bottom. Three consecutive phases of the program provide input to the HTGR design methods verifications: (1) Nondestructive fuel and circuit gamma scanning; (2) removal of steam generator and primary circuit components; and (3) Laboratory examinations of removed components. Component removal site work commenced with establishment of restricted access areas and installation of controlled atmosphere tents to retain relative humidity at <30%. A mock-up room was established to test and develop the tooling and to train operators under simulated working conditions. Primary circuit ducting samples were removed by trepanning, and steam generator access was achieved by a combination of arc gouging and grinding. Tubing samples were removed using internal cutters and external grinding. Throughout the component removal phase, strict health physics, safety, and quality assurance programs were implemented. A total of 148 samples of primary circuit ducting and steam generator tubing were removed with no significant health physics or safety incidents. Additionally, component removal served to provide access fordetermination of cesium plateout distribution by gamma scanning inside the ducts and for macroexamination of the steam generator from both the water and helium sides. Evaluations are continuing and indicate excellent performance of the steam generator and other materials, together with close correlation of observed and predicted fission product plateout distributions. It is concluded that such a program of end-of-life research, when appropriately coordinated with decommissioning activities, can significantly advance nuclear plant and fuel technology development

  17. Nuclear reactor decommissioning: an analysis of the regulatory environments

    International Nuclear Information System (INIS)

    Cantor, R.

    1984-04-01

    The purpose of this study is to highlight some of the current and likely regulations that will significantly affect the costs, technical alternatives and financing schemes for reactor decommissioning encountered by electric utilities and their customers. The paper includes a general review of the decommissioning literature, as well as information on specific regulations at the federal, state, and utility levels. Available estimated costs for the decommissioning of individual reactors are also presented. Finally, classification of the specific policies into common trends and practices among the various regulatory bodies is used to examine more general regulatory environments and their potential financial implications

  18. The decommissioning of the KEMA suspension test reactor

    International Nuclear Information System (INIS)

    Spruyt, A.; Peters, D.; Loon, W.M.G.M. van; Boekschoten, H.J.C.; Brugman, H.

    1991-01-01

    In this report the decommissioning of the KEMA Suspension Test Reactor (KSTR) is described. This reactor was a 1 MWth aqueous homo-geneous nuclear reactor in which a suspension of a mixed oxide UO 2 / ThO 2 in light water was circulated in a closed loop through a sphere-shaped core vessel. The reactor, located on KEMA premises, made 150 MW of heat during its critical periods. Dismantling of this reactor, with its many connected subsystems, meant the mastering of activated components which were also contaminated on inner surfaces caused by small fuel deposits (alpha contaminants) and fission products (beta, gamma contaminants). A description is given of the save removal of the fuel, the remote dismantling of systems and components and the disposal of steel scrap and other materials. Important features are the measures to be taken and provisions needed for safe handling, for the reduction of the radiation dose for the working team and the prevention of spreading of activity over the working area and the environment. It has been demonstrated that safe dismantling and disposal of such systems can be achieved. Experience gained at KEMA for the proper dismantling and for safety measures to be taken for workers and the environment can be made available for similar dismantling projects. A cost break-down is included in the report. (author). 22 refs.; 52 figs.; 12 tabs

  19. Decommissioning of the ICI TRIGA Mark I reactor

    International Nuclear Information System (INIS)

    Parry, D.R.; England, M.R.; Ward, A.; Green, D.

    2000-01-01

    This paper considers the fuel removal, transportation and subsequent decommissioning of the ICI TRIGA Mark I Reactor at Billingham, UK. BNFL Waste Management and Decommissioning carried out this work on behalf of ICI. The decommissioning methodology was considered in the four stages to be described, namely Preparatory Works, Reactor Defueling, Intermediate Level Waste Removal and Low Level Waste Removal. This paper describes the principal methodologies involved in the defueling of the reactor and subsequent decommissioning operations, highlighting in particular the design and safety case methodologies used in order to achieve a solution which was completed without incident or accident and resulted in a cumulative radiation dose to personnel of only 1.57 mSv. (author)

  20. Calculating the Unit Cost Factors for Decommissioning Cost Estimation of the Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Jeong, Kwan Seong; Lee, Dong Gyu; Jung, Chong Hun; Lee, Kune Woo

    2006-01-01

    The estimated decommissioning cost of nuclear research reactor is calculated by applying a unit cost factor-based engineering cost calculation method on which classification of decommissioning works fitted with the features and specifications of decommissioning objects and establishment of composition factors are based. Decommissioning cost of nuclear research reactor is composed of labor cost, equipment and materials cost. Labor cost of decommissioning costs in decommissioning works are calculated on the basis of working time consumed in decommissioning objects. In this paper, the unit cost factors and work difficulty factors which are needed to calculate the labor cost in estimating decommissioning cost of nuclear research reactor are derived and figured out.

  1. Technical survey of decommissioning of commercial power reactors

    International Nuclear Information System (INIS)

    Nakamura, Masahide

    2003-01-01

    The technical survey of decommissioning of commercial power reactors had been carried out from 1982 to 2003. The investigation items are scenarios, procedures, simplification and recycling. On the scenarios, the case studies on the decommissioning steps (1983 to 1984), evaluation of the prior conditions of case studies (1994 to 1998), evaluation of rationalization of the scenarios of decommissioning steps (1999 to 2001) and evaluation of the effects of investigation of clearance level (1999 to 2002) are described. Procedures (1985 to 1996) and simplification (1985 to 1987) of decommissioning are investigated. On the recycling, survey on recycle of waste produced by the decommissioning step (1985 to 1993) and recycle of demolition waste (1997 to 2002) are reported. Recycle of radioactive waste has to be controlled under lows. (S.Y.)

  2. Preliminary decommissioning plan of the reactor IPEN-MB01

    International Nuclear Information System (INIS)

    Vivas, Ary de Souza

    2014-01-01

    Around the world, many nuclear plants were built and need to be turned off at a certain time because they are close to their recommended time of use is approximately 50 years. So the IAEA (International Atomic Energy Agency), seeks to guide and recommend a set of guidelines for the conduct of activities of nuclear facilities, with special attention to countries that do not have a framework regulatory Legal that sustain the activities of decommissioning. Brazil, so far, does not have a specific standard to guide the steps of the guidelines regarding decommissioning research reactors. However, in March 2011 a study committee was formed with the main task facing the issues of decommissioning of nuclear installations in Brazil, culminating in Resolution 133 of November 8, 2012, a standard project that treat about the Decommissioning of nucleoelectric plants. O Instituto de Pesquisas Energeticas e Nucleares (IPEN) has two research reactors one being the reactor IPEN/MB-01. The purpose of this master dissertation is to develop a preliminary plan for decommissioning this research reactor, considering the technical documentation of the facility (RAS-Safety Analysis Report), the existing standards of CNEN (National Nuclear Energy Commission), as well as IAEA recommendations. In terms of procedures for decommissioning research reactors, this work was based on what is most modern in experiences, strategies and lessons learned performed and documented in IAEA publications covering techniques and technologies for decommissioning. Considering these technical knowledge and due to the peculiarities of the facility, was selected to immediate dismantling strategy, which corresponds to the start of decommissioning activities once the installation is switched off, dividing it into work sectors. As a resource for monitoring and project management of reactor decommissioning and maintenance of records, we developed a database using Microsoft Access 2007, which contain all the items and

  3. Technology, safety and costs of decommissioning nuclear reactors at multiple-reactor stations

    International Nuclear Information System (INIS)

    Wittenbrock, N.G.

    1982-01-01

    Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWR) and large (1155-MWe) boiling water reactors (BWR) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services

  4. Research reactor decommissioning experience - concrete removal and disposal -

    International Nuclear Information System (INIS)

    Manning, Mark R.; Gardner, Frederick W.

    1990-01-01

    Removal and disposal of neutron activated concrete from biological shields is the most significant operational task associated with research reactor decommissioning. During the period of 1985 thru 1989 Chem-Nuclear Systems, Inc. was the prime contractor for complete dismantlement and decommissioning of the Northrop TRIGA Mark F, the Virginia Tech Argonaut, and the Michigan State University TRIGA Mark I Reactor Facilities. This paper discusses operational requirements, methods employed, and results of the concrete removal, packaging, transport and disposal operations for these (3) research reactor decommissioning projects. Methods employed for each are compared. Disposal of concrete above and below regulatory release limits for unrestricted use are discussed. This study concludes that activated reactor biological shield concrete can be safely removed and buried under current regulations

  5. New safety experiments in decommissioned superheated steam reactor at Karlstein

    International Nuclear Information System (INIS)

    Koerting, K.

    1986-01-01

    This article gives a concise summary of the Status Report of the Superheated Steam Reactor Safety Program (PHDR) Project, held at KfK on Dec. 5, 1985. The results discussed dealt with fire experiments, shock tests simulating airplane crashes, temperature shocks in the reactor pressure vessel, studies of crack detection in pressure vessels and blasting experiments associated with nuclear plant decommissioning

  6. Management of the decommissioning of the Thetis reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ooms, Luc; Maris, Patrick; Noynaert, Luc [SCK-CEN, Mol (Belgium)

    2013-07-01

    The Thetis research reactor on the site of the Nuclear Sciences Institute of the Ghent University has been in operation from 1967 until December 2003. This light-water moderated graphite-reflected low-enriched uranium pool-type reactor has been used for various purposes e.g. the production of radioisotopes and activation analyses. During the first years its core power was 15 kW. In the early '70, a core enlargement allowed for operation at typically 150 kW, while the maximum was allowed to be 250 kW In September 2007, Ghent University entrusted to SCK-CEN the management of the back-end of the spent fuel and the decommissioning of the reactor. In 2010, the spent fuel was removed from the reactor and transported to Belgoprocess for cementation in 400 l drums and interim storage awaiting final disposal. This activity allows tackling the decommissioning of the reactor. The objective is to complete its decommissioning by the end of 2014. In the framework of the decommissioning of the Thetis reactor, SCK-CEN set-up the final decommissioning plan and the decommissioning licensing file. These documents include among others a radiological inventory of the reactor. The graphite moderator blocks, the control and the safety pates, the liner of the pool were modeled to assess the activation products (isotopic vector and intensity). At the end of the unloading of the reactor in 2010 a brief mapping of the equipment's and internals of the reactor pool was performed. In 2012, we realized a more detailed mapping. These results confirmed those performed earlier and allowed to confirm the assumptions made in the final decommissioning plan. We set-up the terms of reference for the first decommissioning phase of the reactor namely the dismantling of the reactor i.e. reactor pool, circuits and rabbit system, equipment's and ventilation ducts. The removal of asbestos is also included into this phase. We conducted the selection process and the awarding of this

  7. Technical meeting on decommissioning of fast reactors after sodium draining. Working material

    International Nuclear Information System (INIS)

    2005-01-01

    The objective of the technical meeting was to provide a forum for in-depth scientific and technical exchange on topics related to the decommissioning experience with fast reactors, in particular with regard to the decommissioning of components after sodium draining. Accordingly, the scope of the meeting covers the review and analyses of the experience gained from the decommissioning of both active sodium loops and sodium cooled fast reactors (e.g., KNK II, Superphenix, RAPSODIE, EBR-II, FERMI, BN-350, BR-10). It is expected that the outcome of the meeting will contribute to the Agency initiative to preserve fast reactor data and knowledge. The main focus of the technical meeting was given on the decommissioning of both active loop and reactor components (e.g., the primary vessel of a sodium-cooled reactor) that have been drained of sodium, but that still conserve some residual amounts of sodium (e.g., films covering the entire surface of the component, or particular sodium heels that cannot be drained)

  8. Decontamination and decommissioning of the JANUS reactor at the Argonne National Laboratory-East site

    International Nuclear Information System (INIS)

    Fellhauer, C.R.; Garlock, G.A.

    1997-05-01

    Argonne National Laboratory has begun the decontamination and decommissioning (D ampersand D) of the JANUS Reactor Facility. The project is managed by the Technology Development Division's D ampersand D Program personnel. D ampersand D procedures are performed by sub-contractor personnel. Specific activities involving the removal, size reduction, and packaging of radioactive components and facilities are discussed

  9. Preparations for decommissioning the TRIGA Mark III Berkeley Research Reactor

    International Nuclear Information System (INIS)

    Denton, Michael M.; Lim, Tek. H.

    1988-01-01

    On December 20, 1986 the chancellor of UC Berkeley announced his decision to decommission the 20 year old Berkeley Research Reactor citing as principal reasons a decline in use and a need to erect a new computer science building over the reactor's site. In order to meet the University's construction timetable for the new building, the reactor staff together with other units of the campus administration have initiated a program to remove the reactor structure and clear the room for unlicensed use as expediently as possible. Due to the sequence of events which must occur in a limited amount of time, the University adopted a policy to contract out as much of the work as possible, including generation of the defueling and decommissioning plans.The first physical step in the decommissioning project is the removal of the irradiated fuel. This task is largely contracted out to a commercial firm with experience in the transport of radioactive materials and reactor fuel. As suggested by the NRC, the reactor will be defueled under the current operating license. This requires that all fuel must be off-site before the DP can be approved. Therefore any delay in defueling in-turn delays the decommissioning. The NRC has given no commitment or date for completion of their review. Informal discussion with NRC project managers and the experience from other facilities indicate that the review process will take between six and nine months

  10. Decommissioning and re-utilization of the Musashi Reactor

    International Nuclear Information System (INIS)

    Tomio Tanzawa; Nobukazu Iijima; Norikazu Horiuchi; Tadashi Yoshida; Tetsuo Matsumoto; Naoto Hagura; Ryouhei Kamiya

    2008-01-01

    The Musashi Institute of Technology Research Reactor (the Musashi Reactor) is a TRIGA-? with maximum thermal power of 100 kW. The decommissioning was decided in May, 2003. The reactor facility is now under decommissioning. The phased decommissioning was selected. Phase 1 consists of permanent shutdown of the reactor and stopping the operational functions, and transportation of the spent nuclear fuels. After completion of the transportation, the reactor facility is characterized as the storage of low level radioactive materials. This is phase 2. Activities of phase 1 were completed and the facility is now under phase 2. Activities of phase 3 consist of dismantling the reactor tank and the shielding, and delivering the radioactive waste to a waste disposal facility. The phase 3 will be started on condition that the undertaking of the waste disposal for research reactors will be established. On the other hand, re-utilization of the facility has being studied, and 'realistic' reactor simulator was turned out by utilizing the reactor installations such as control rod drive and operation console. (authors)

  11. Scheme of database structure on decommissioning of the research reactor

    International Nuclear Information System (INIS)

    Park, H. S.; Park, S. K.; Kim, H. R.; Lee, D. K.; Jung, K. J.

    2001-01-01

    ISP (Information Strategy Planning), which is the first step of the whole database development, has been studied to manage effectively information and data related to the decommissioning activities of the Korea Research Reactor 1 and 2 (KRR-1 and 2). Since Korea has not acquired the technology of the decommissioning database management system, some record management system (RMS) of large nuclear facilities of national experience such as in the U.S.A, Japan, Belgium, and Russian were reviewed. In order to construct the database structure of the whole decommissioning activities such as the working information, radioactive waste treatment, and radiological surveying and analysis has been extracted from the whole dismantling process. These information and data will be used as the basic data to analyzed the matrix to find the entity relationship diagram and will contribute to the establishment of a business system design and the development of a decommissioning database system as well

  12. Remediation of Site of Decommissioning Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Danilovich, A.S.; Ivanov, O.P.; Lemus, A.V.; Pavlenko, V.I.; Potapov, V.N.; Semenov, S.G.; Shisha, A.D.; Chesnokov, A.V. [National Research Center ' Kurchatov Institute' , 123182, Moscow (Russian Federation)

    2014-07-01

    In the world the most widespread method of soil decontamination consists of removing the contaminated upper layer and sending it for long-term controlled storage. However, implementation of this soil cleanup method for remediation of large contaminated areas would involve high material and financial expenditures, because it produces large amounts of radioactive waste demanding removal to special storage sites. Contaminated soil extraction and cleanup performed right on the spot of remediation activities represents a more advanced and economically acceptable method. Radiological separation of the radioactive soil allows reducing of amount of radwaste. Studies performed during the liquidation of the Chernobyl accident consequences revealed that a considerable fraction of radioactivity is accumulated in minute soil grains. So, the separation of contaminated soil by size fractions makes it possible to extract and concentrate the major share of radioactivity in the fine fraction. Based on these researches water gravity separation technology was proposed by Bochvar Institute. The method extracts the fine fraction from contaminated soil. Studies carried out by Bochvar Institute experts showed that, together with the fine fraction (amounting to 10-20% of the initial soil), this technology can remove up to 85-90% of contaminating radionuclides. The resulting 'dirty' soil fraction could be packaged into containers and removed as radwaste, and decontaminated fractions returned back to their extraction site. Use of radiological and water gravity separations consequently increases the productivity of decontamination facility. Efficiency of this technology applied for contaminated soil cleanup was confirmed in the course of remediation of the contaminated territories near decommissioning research reactor in the Kurchatov Institute. For soil cleaning purposes, a special facility implementing the technology of water gravity separation and radiometric monitoring of soil

  13. 77 FR 8902 - Draft Regulatory Guide: Issuance, Availability Decommissioning of Nuclear Power Reactors

    Science.gov (United States)

    2012-02-15

    ... Decommissioning of Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide... draft regulatory guide (DG) DG-1271 ``Decommissioning of Nuclear Power Reactors.'' This guide describes... Regulatory Guide 1.184, ``Decommissioning of Nuclear Power Reactors,'' dated July 2000. This proposed...

  14. ADVANTAGES, DISADVANTAGES, AND LESSONS LEARNED FROM MULTI-REACTOR DECOMMISSIONING PROJECTS

    International Nuclear Information System (INIS)

    Morton, M.R.; Nielson, R.R.; Trevino, R.A.

    2003-01-01

    This paper discusses the Reactor Interim Safe Storage (ISS) Project within the decommissioning projects at the Hanford Site and reviews the lessons learned from performing four large reactor decommissioning projects sequentially. The advantages and disadvantages of this multi-reactor decommissioning project are highlighted

  15. Funding for reactor decommissioning: the NRC perspective

    International Nuclear Information System (INIS)

    Wood, R.S.

    1981-01-01

    The cost of decommissioning a nuclear power plant is discussed. Four funding approaches that have received the most attention from the NRC are: prepayment into a trust fund of estimated decommissioning funds at the start of facility operation; annual contributions into a trust fund outside the control of the utility over the estimated life of a facility; internal reserve or sinking fund amortizations over the estimated life of a facility; and insurance or other surety mechanisms used separately or in conjunction with any of the first three mechanisms

  16. Development of telerobotic systems for reactor decommissioning, (3)

    International Nuclear Information System (INIS)

    Usui, Hozumi; Fujii, Yoshio; Shinohara, Yoshikuni

    1991-01-01

    This paper describes the telerobotic system for reactor decommissioning in the scope of engineering demonstration of dismantling radioactive reactor internals of an experimental boiling water power reactor JPDR. The total system consists of a telerobotic manipulator system equipped with a multi-functional amphibious slave manipulator with a load capacity of 25 daN, a chain-driven transport system, and a computer-assisted monitoring and control system. Preceding to the application of the telerobotic system to actual dismantling operation, a mockup test was performed of dismantling the simulated reactor internals of actual-size by the method of underwater plasma arc cutting in order to study the performance of the telerobotic system in a realistic environment. The system was then successfully applied to dismantling the actual reactor internals according to the JPDR decommissioning program. (author)

  17. Socio-economic impact of nuclear reactor decommissioning at Vandellos I NPP

    International Nuclear Information System (INIS)

    Liliana Yetta Pandi

    2013-01-01

    Currently nuclear reactors in Indonesia has been outstanding for more than 30 years, the possibility of nuclear reactors will be decommissioned. Closure of the operation or decommissioning of nuclear reactors will have socio-economic impacts. The socioeconomic impacts occur to workers, local communities and wider society. In this paper we report on socio-economic impacts of nuclear reactors decommissioning and lesson learned that can be drawn from the socio-economic impacts decommissioning Vandellos I nuclear power plant in Spain. Socio-economic impact due to decommissioning of nuclear reactor occurs at installation worker, local community and wider community. (author)

  18. Decontamination experiments for stainless steel decommissioned components

    International Nuclear Information System (INIS)

    Stefanescu, D.; Radulescu, M.; Dragomir, M.; Velciu, L.; Dinu, A.

    2001-01-01

    This paper presents the factors which influence the decontamination conditions using the steps of CONAP process. This four phases process (alkaline pre-treatment , an oxidation phase with potassium permanganate in acid environment, a dissolution phase using a complexing agent, a rinsing phase) has been used for decontamination to recycle the stainless steel 304 L and 403 m. The attraction of this process results from the following reasons: - the volume of radioactive sludge is low comparatively with the original volume of the solutions; - the separation of the activity from the solution is very effective; - time of exposure is reduced; - it is not necessary to process the solution through evaporators. During decommissioning decontamination is used to reduce radiation field by removing some of the fission and activation products contained in deposits and oxide films to minimize the radiation exposure of the personnel and public. In this context, this hard decontamination yields the materials at a radioactivity level fulfilling the repository requirements. (authors)

  19. Enhanced productivity in reactor decommissioning and waste management

    International Nuclear Information System (INIS)

    Wasinger, Karl

    2014-01-01

    As for any industrial facility, the service live of nuclear power plants, fuel cycle facilities, research and test reactors ends. Decision for decommissioning such facilities may be motivated by technical, economical or political reasons or a combination of it. As of today, a considerable number of research reactors, fuel cycle facilities and power reactors have been completely decommissioned. However, the end point of such facilities' lifetime is achieved, when the facility is finally removed from regulatory control and the site becomes available for further economical utilization. This process is commonly known as decommissioning and involves detailed planning of all related activities, radiological characterization, dismantling, decontamination, clean-up of the site including treatment and packaging of radioactive and/or contaminated material not released for unrestricted recycling or industrial disposal. Decommissioning requires adequate funding and suitable measures to ensure safety while addressing stakeholders' requirements on occupational health, environment, economy, human resources management and the socioeconomic effects to the community and the region. One important aspect in successful management of decommissioning projects and dismantling operation relates to the economical impact of the endeavor, primarily depending on the selected strategy and, as from commencement of dismantling, on total duration until the end point is achieved. Experience gained by Areva in executing numerous decommissioning projects during past 2 decades shows that time injury free execution and optimum productivity turns out crucial to project cost. Areva develops and implements specific 'performance improvement plans' for each of its projects which follow the philosophy of operational excellence based on Lean Manufacturing principles. Means and methods applied in implementation of these plans and improvements achieved are described and examples are given on the way Areva

  20. Enhanced productivity in reactor decommissioning and waste management

    Energy Technology Data Exchange (ETDEWEB)

    Wasinger, Karl [Areva GmbH, Offenbach (Germany)

    2014-04-15

    As for any industrial facility, the service live of nuclear power plants, fuel cycle facilities, research and test reactors ends. Decision for decommissioning such facilities may be motivated by technical, economical or political reasons or a combination of it. As of today, a considerable number of research reactors, fuel cycle facilities and power reactors have been completely decommissioned. However, the end point of such facilities' lifetime is achieved, when the facility is finally removed from regulatory control and the site becomes available for further economical utilization. This process is commonly known as decommissioning and involves detailed planning of all related activities, radiological characterization, dismantling, decontamination, clean-up of the site including treatment and packaging of radioactive and/or contaminated material not released for unrestricted recycling or industrial disposal. Decommissioning requires adequate funding and suitable measures to ensure safety while addressing stakeholders' requirements on occupational health, environment, economy, human resources management and the socioeconomic effects to the community and the region. One important aspect in successful management of decommissioning projects and dismantling operation relates to the economical impact of the endeavor, primarily depending on the selected strategy and, as from commencement of dismantling, on total duration until the end point is achieved. Experience gained by Areva in executing numerous decommissioning projects during past 2 decades shows that time injury free execution and optimum productivity turns out crucial to project cost. Areva develops and implements specific 'performance improvement plans' for each of its projects which follow the philosophy of operational excellence based on Lean Manufacturing principles. Means and methods applied in implementation of these plans and improvements achieved are described and examples are given on

  1. Assessment methodology applicable to safe decommissioning of Romanian VVR-S research reactor

    International Nuclear Information System (INIS)

    Baniu, O.; Vladescu, G.; Vidican, D.; Penescu, M.

    2002-01-01

    The paper contains the results of research activity performed by CITON specialists regarding the assessment methodology intended to be applied to safe decommissioning of the research reactors, developed taking into account specific conditions of the Romanian VVR-S Research Reactor. The Romanian VVR-S Research Reactor is an old reactor (1957) and its Decommissioning Plan is under study. The main topics of paper are as follows: Safety approach of nuclear facilities decommissioning. Applicable safety principles; Main steps of the proposed assessment methodology; Generic content of Decommissioning Plan. Main decommissioning activities. Discussion about the proposed Decommissioning Plan for Romanian Research Reactor; Safety risks which may occur during decommissioning activities. Normal decommissioning operations. Fault conditions. Internal and external hazards; Typical development of a scenario. Features, Events and Processes List. Exposure pathways. Calculation methodology. (author)

  2. The Windscale Advanced Gas Cooled Reactor (WAGR) Decommissioning Project A Close Out Report for WAGR Decommissioning Campaigns 1 to 10 - 12474

    Energy Technology Data Exchange (ETDEWEB)

    Halliwell, Chris [Sellafield Ltd, Sellafield (United Kingdom)

    2012-07-01

    The reactor core of the Windscale Advanced Gas-Cooled Reactor (WAGR) has been dismantled as part of an ongoing decommissioning project. The WAGR operated until 1981 as a development reactor for the British Commercial Advanced Gas cooled Reactor (CAGR) power programme. Decommissioning began in 1982 with the removal of fuel from the reactor core which was completed in 1983. Subsequently, a significant amount of engineering work was carried out, including removal of equipment external to the reactor and initial manual dismantling operations at the top of the reactor, in preparation for the removal of the reactor core itself. Modification of the facility structure and construction of the waste packaging plant served to provide a waste route for the reactor components. The reactor core was dismantled on a 'top-down' basis in a series of 'campaigns' related to discrete reactor components. This report describes the facility, the modifications undertaken to facilitate its decommissioning and the strategies employed to recognise the successful decommissioning of the reactor. Early decommissioning tasks at the top of the reactor were undertaken manually but the main of the decommissioning tasks were carried remotely, with deployment systems comprising of little more than crane like devices, intelligently interfaced into the existing structure. The tooling deployed from the 3 tonne capacity (3te) hoist consisted either purely mechanical devices or those being electrically controlled from a 'push-button' panel positioned at the operator control stations, there was no degree of autonomy in the 3te hoist or any of the tools deployed from it. Whilst the ATC was able to provide some tele-robotic capabilities these were very limited and required a good degree of driver input which due to the operating philosophy at WAGR was not utilised. The WAGR box proved a successful waste package, adaptable through the use of waste box furniture specific to the

  3. Reactor component automatic grapple

    International Nuclear Information System (INIS)

    Greenaway, P.R.

    1982-01-01

    A grapple for handling nuclear reactor components in a medium such as liquid sodium which, upon proper seating and alignment of the grapple with the component as sensed by a mechanical logic integral to the grapple, automatically seizes the component. The mechanical logic system also precludes seizure in the absence of proper seating and alignment. (author)

  4. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices

    International Nuclear Information System (INIS)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data

  5. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data.

  6. Sodium removal from Hallam Reactor components

    International Nuclear Information System (INIS)

    Huntsman, L.K.; Meservey, R.H.

    1979-08-01

    This report discussed the removal of sodium from major components of the Hallam Nuclear Power Facility. This facility contained the experimental ractor used to test the feasibility of sodium coolant. The Idaho Operations Office of the Department of Energy assigned EG and G Idaho, Inc., the task of carrying out this decontamination and decommissioning program at the Idaho National Engineering Laboratory (INEL). Since their shipment to the INEL from Lincoln, Nebraska in 1968, the Hallam Reactor components had been stored in inert nitrogen to prevent the sodium in the components from reacting with moisture in the air. The procedure used to react the sodium in the components and to decontaminate them is discussed. Problems and unusual occurrences in the decontamination and decommissioning process are also reported

  7. Progress of decommissioning of Rikkyo reactor in FY2014

    International Nuclear Information System (INIS)

    Suzuki, M.; Kato, M.; Tanzawa, T.; Kawaguchi, K.; Terasawa, T.; Yamada, Shigeru; Nakai, Masaru

    2015-01-01

    Institute for Atomic Energy, Rikkyo University, applied in 2012 for changes in the decommissioning plan toward the abolition of the reactor facilities, and received approval. It promoted the decommissioning work of the research reactors in a plan for two years from 2012, conducted the removal of the structure installed in the reactor tank and storage management measures, and implemented the function stop of the disposal facility of liquid waste and the removal of part of them. These procedures achieved the safe storage condition of core internal structure / equipment with relatively high radioactivity due to neutron irradiation. In addition, the maintenance management of partial facilities and equipment that had been maintained in operational conditions had come to be unnecessary. Based on these results, the implementation plan for decommissioning scheduled for 2015-2016 was prepared. The contents of main works are as follows: (1) dismantling and removal of disposal facilities for liquid waste and storage management of subsequently generated radioactive waste in the reactor building control area, (2) storage management of radioactive solid waste of solid waste storage facilities in the reactor building control area, (3) dismantling and removal of solid waste storage facilities that become unnecessary, and (4) release of part of the controlled area associated with the above actions. (A.O.)

  8. Decommissioning plan for Tammuz-2 research reactor in Iraq

    International Nuclear Information System (INIS)

    Ahmed, A. A.; Jasim, H. I.

    2012-12-01

    For nuclear facilities, decommissioning is the final phase in the life cycle after sitting, design, construction, commissioning and operation. It is a process involving operations such as decontamination, dismantling of plant equipment of result in materials. All these activities take into account health and safety requirements for operating personnel and the general public, and any implications for the environment. (1) In several projects to decommission various type of nuclear facilities, it has been shown that technical methods and equipment are available today to dismantle safely nuclear facilities, of whatever type or size. Much experience in the use of these techniques has the decommissioning of prototype, demonstration, and small power reactors or other nuclear facilities. In Iraq these activities will be done by the cooperation with (IAEA) International Atomic Energy Agency and the other national regulatory bodies such as (IRSRA) Iraqi Radioactive Sources Regulatory Authority, and (RPC/MoEn) Radiation Protection Center/ Ministry of Environment in Iraq. (Author)

  9. Plan for decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Walton, G.R.

    1993-01-01

    The Tokamak Fusion Test Reactor (TFTR) Project is in the planning phase of developing a decommissioning project. A Preliminary Decontamination and Decommissioning (D ampersand D) Plan has been developed which provides a framework for the baseline approach, and the cost and schedule estimates. TFTR will become activated and contaminated with tritium after completion of the deuterium-tritium (D-T) experiments. Hence some of the D ampersand D operations will require remote handling. It is expected that all of the waste generated will be low level radioactive waste (LLW). The objective of the D ampersand D Project is to make TFTR Test Cell available for use by a new fusion experiment. This paper discusses the D ampersand D objectives, the facility to be decommissioned, estimates of activation, the technical (baseline) approach, and the assumptions used to develop cost and schedule estimates

  10. Nuclear power plant decommissioning

    International Nuclear Information System (INIS)

    Yaziz Yunus

    1986-01-01

    A number of issues have to be taken into account before the introduction of any nuclear power plant in any country. These issues include reactor safety (site and operational), waste disposal and, lastly, the decommissioning of the reactor inself. Because of the radioactive nature of the components, nuclear power plants require a different approach to decommission compared to other plants. Until recently, issues on reactor safety and waste disposal were the main topics discussed. As for reactor decommissioning, the debates have been academic until now. Although reactors have operated for 25 years, decommissioning of retired reactors has simply not been fully planned. But the Shippingport Atomic Power Plant in Pennysylvania, the first large scale power reactor to be retired, is now being decommissioned. The work has rekindled the debate in the light of reality. Outside the United States, decommissioning is also being confronted on a new plane. (author)

  11. Evaluation of nuclear facility decommissioning projects. Project summary report, Elk River Reactor

    International Nuclear Information System (INIS)

    Miller, R.L.; Adams, J.A.

    1982-12-01

    This report summarizes information concerning the decommissioning of the Elk River Reactor. Decommissioning data from available documents were input into a computerized data-handling system in a manner that permits specific information to be readily retrieved. The information is in a form that assists the Nuclear Regulatory Commission in its assessment of decommissioning alternatives and ALARA methods for future decommissionings projects. Samples of computer reports are included in the report. Decommissioning of other reactors, including NRC reference decommissioning studies, will be described in similar reports

  12. Waste generated by the future decommissioning of the Magurele VVR-S Research Reactor

    International Nuclear Information System (INIS)

    Dragolici, F.; Turcanu, C.N.; Dragolici, A.C.

    2001-01-01

    Nuclear Research Reactor WWR-S from the National Institute of Research and Development for Physics and Nuclear Engineering 'Horia Hulubei', Bucharest-Magurele, was commissioned in July 1957 and it was shut down in December 1997. At the moment the reactor is in conservation state. During its operation this reactor worked at an average power of 2MW, almost 3216 h/year, producing a total thermal power of 230 x 10 3 MWh. No major modifications or improvements were made during the 40 years of operation to the essential parts of the reactor, respective to the primary cooling system, reactor vessel, active core and electronic devices. So, all components of the measure, control and protection systems are old, generally at the technical level of the 1950s, therefore a reason why in December 1997 the operation was ceased. At present, the reactor can be considered, by IAEA definition in the first stage (reactor shut down, but the vital functions are maintained and monitored). The survey is related to the second stage - restrictive use of the area. To develop a real decommissioning project, it was first necessary to evaluate the volume and the characteristics of the radioactive waste which will be generated. Radioactive waste generated during the decommissioning of Magurele WR-S research reactor may be classified as: Activated wastes (internal structures, horizontal channels and thermal column, biological shield); Contaminated wastes (primary circuit non-activated components, hot cells, some technological rooms as main hall, pumps room, radioactive material transfer areas, ventilation building and stack); Possibly contaminated materials from any area of reactor building and ventilation building. After 40 years of nuclear research activities, all such areas are suspected of contamination. The volume of wastes that will result from WWR-S Research Reactor decommissioning is summarized

  13. Assuring the availability of funds for decommissioning nuclear reactors

    International Nuclear Information System (INIS)

    1990-08-01

    The general requirements for applications for license termination and decommissioning nuclear power, research, and test reactors are contained in 10 CFR Part 50, ''Domestic Licensing of Production and Utilization Facilities.'' On June 27, 1988, the Commission published amendments to 10 CFR Part 50 (53 FR 24018) concerning specific criteria for decommissioning nuclear facilities. Amended 10 CFR 50.33(k), 50.75, and 50.82(b) require operating license applicants and existing licensees to submit information on how reasonable assurance will be provided that funds are available to decommission the facility. Amended section 50.75 establishes requirements for indicating how this assurance will be provided, namely the amount of funds that must provided, including updates, and the methods to be used for assuring funds. This regulatory guide has been developed in conjunction with the rule amendments and was published for public comment in May 1989. This version incorporates, where appropriate, the public comments received. Its purpose is to provide guidance to applicants and licensees of nuclear power, research, and test reactors concerning methods acceptable to the NRC staff for complying with requirements in the amended rule regarding the amount of funds for decommissioning. It also provides guidance on the content and form of the financial assurance mechanisms indicated in the rule amendments. 9 refs

  14. Radiological protection of the staff during the decommissioning operations of the Romanian VVR-S research reactor

    International Nuclear Information System (INIS)

    Ene, D.C.

    2002-01-01

    Dose rate estimates for periods of 100 days and 6, 10, 25, 100 years after the shut down of the Romanian VVR-S reactor are presented in this paper for some foreseen decommissioning activities which include: i) cutting the water pipe in the pump room and the reactor sealing operations; ii) extracting reactor components; and iii) handling and dismantling the internal structures taken of from the reactor. For the reactor components extracted from the reactor, the considered calculation points were placed in the central plan of the items, on the surface and at distances from the surface which correspond to +0.2m, +1m, +2m, +8m, and +10m. Time dependence of the resulted dose rates are presented and discussed. Qualitative comparison with the measured values from other VVR-S reactors is done. The obtained results assist to develop working procedures that must be observed during the decommissioning activities. (author)

  15. Safe decommissioning of the Romanian VVR-S research reactor

    International Nuclear Information System (INIS)

    Garlea, C.; Garlea, I.; Kelerman, C.; Rodna, A.

    2002-01-01

    The VVR-S Romania research reactor was operated between 1957-1997, at 2 MW nominal power, for research and radioisotopical production. The detailed decommissioning plan was developed between 1995-1998, in the frame of the International Atomic Energy Agency Technical assistance project ROM/9/017. The proposed strategy agreed by the counterpart as well as international experts was stage 1. In 1997, an independent analysis performed by European Commission experts, in the frame of PHARE project PH04.1/1994 was dedicated to the 'Study of Soviet Design Research Reactors', had consolidated the development of the project emphasizing technical options of safe management for radioactive wastes and VVR-S spent fuel. The paper presents the main technical aspects as well as those of social impact, which lead to the establishment of strategy for safe management of decommissioning. Technical analysis of the VVR-S reactor and associated radwaste facilities (Radioactive Waste Treatment Plant - Magurele and National Repository Baita-Bihor) proved the possibility of the classical method utilization for dismantling of the facility and treatment-conditioning-disposal of the arrised wastes in safe conditions. The decommissioning plan at stage 2 has been developed based on radiological safety assessment, evaluation of radwaste inventory (removed as well as preserved on site), cost analysis and environmental impact. Technical data were provided by the R and D programme including neutron calculations and experiments, radiological characterizing (for facility and its influence area), seismic analysis and environmental balance during the operation and after shut down of the reactor. A special chapter is dedicated to regulatory issues concerning the development of decommissioning under nuclear safety. Based on the Fundamental Norms of Radiological Safety, the Regulatory Body defined the clearance levels and safety criteria for the process. The development of National Norms for the

  16. Decommissioning of the BR3 pressurized-water reactor

    International Nuclear Information System (INIS)

    Massaut, V.

    1996-01-01

    The dismantling and the decommissioning of nuclear installations at the end of their life-cycle is a new challenge to the nuclear industry. Different techniques and procedures for the dismantling of a nuclear power plant on an existing installation, the BR-3 pressurized-water reactor, are described. The scientific programme, objectives, achievements in this research area at the Belgian Nuclear Research Centre SCK-CEN for 1995 are summarized

  17. Integration of improved decontamination and characterization technologies in the decommissioning of the CP-5 research reactor

    International Nuclear Information System (INIS)

    Bhattacharyya, S. K.; Boing, L. E.

    2000-01-01

    The aging of research reactors worldwide has resulted in a heightened awareness in the international technical decommissioning community of the timeliness to review and address the needs of these research institutes in planning for and eventually performing the decommissioning of these facilities. By using the reactors already undergoing decommissioning as test beds for evaluating enhanced or new/innovative technologies for decommissioning, it is possible that new techniques could be made available for those future research reactor decommissioning projects. Potentially, the new technologies will result in: reduced radiation doses to the work force, larger safety margins in performing decommissioning and cost and schedule savings to the research institutes in performing the decommissioning of these facilities. Testing of these enhanced technologies for decontamination, dismantling, characterization, remote operations and worker protection are critical to furthering advancements in the technical specialty of decommissioning. Furthermore, regulatory acceptance and routine utilization for future research reactor decommissioning will be assured by testing and developing these technologies in realistically contaminated environments prior to use in the research reactors. The decommissioning of the CP-5 Research Reactor is currently in the final phase of dismantlement. In this paper the authors present results of work performed at Argonne National Laboratory (ANL) in the development, testing and deployment of innovative and/or enhanced technologies for the decommissioning of research reactors

  18. Decommissioning of the pool reactor Thetis in Ghent, Belgium

    Energy Technology Data Exchange (ETDEWEB)

    Cortenbosch, Geert; Mommaert, Chantal [Bel V, Brussels (Belgium); Tierens, Hubert; Monsieurs, Myriam; Meierlaen, Isabelle; Strijckmans, Karel [Ghent Univ. (Belgium)

    2016-11-15

    The Thetis research pool reactor (with a nominal power of 150 kW) of the Ghent University was operational from 1967 till December 2003. The first phase of the decommissioning of the reactor, the removal of the spent fuel from the site, took place in 2010. The cumulative dose received was only 404 man . μSv. During the second phase, the transition period between the removal of the spent fuel in 2010 and the start of the decommissioning phase in March 2013, 3-monthly internal inspections and inspections by Bel V, were performed. The third and final decommissioning phase started on March 18, 2013. The total dose received between March 2013 and August 2013 was 1561 man . μSv. The declassification from a Class I installation to a Class II installation was possible by the end of 2015. The activated concrete in the reactor pool will remain under regulatory control until the activation levels are lower than the limits for free release.

  19. Certifying the decommissioned Shippingport reactor vessel for transport

    International Nuclear Information System (INIS)

    Towell, R.H.

    1990-01-01

    The decommissioned Shippingport reactor pressure vessel with its concentric neutron shield tank was shipped to Hanford, WA as part of the effort to restore the Shippingport Station to its original condition. The metal walls of the reactor vessel had become radioactive from neutron bombardment while the reactor was operating so it had to be shipped under the regulations for transporting radioactive material. Because of the large amount of radioactivity in the walls, 16,467 Curies, and because the potentially dispersible corrosion layer on the inner walls of both tanks was also radioactive, the Shippingport reactor vessel was transported under the most stringent of the regulations, those for a type B package. Compliance with the packaging regulations was confirmed via independent analysis by the staff of the Department of Energy certifying official and the Shippingport reactor vessel was shipped under DOE Certificate of Compliance USA/9515/B(U)

  20. SAVANNAH RIVER SITE R REACTOR DISASSEMBLY BASIN IN SITU DECOMMISSIONING

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C.; Blankenship, J.; Griffin, W.; Serrato, M.

    2009-12-03

    The US DOE concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate in tact, structurally sound facilities that are no longer needed for their original purpose of, i.e., generating (reactor facilities), processing(isotope separation facilities) or storing radioactive materials. The 105-R Disassembly Basin is the first SRS reactor facility to undergo the in-situ decommissioning (ISD) process. This ISD process complies with the105-R Disassembly Basin project strategy as outlined in the Engineering Evaluation/Cost Analysis for the Grouting of the R-Reactor Disassembly Basin at the Savannah River Site and includes: (1) Managing residual water by solidification in-place or evaporation at another facility; (2) Filling the below grade portion of the basin with cementitious materials to physically stabilize the basin and prevent collapse of the final cap - Sludge and debris in the bottom few feet of the basin will be encapsulated between the basin floor and overlying fill material to isolate if from the environment; (3) Demolishing the above grade portion of the structure and relocating the resulting debris to another location or disposing of the debris in-place; and (4) Capping the basin area with a concrete slab which is part of an engineered cap to prevent inadvertent intrusion. The estimated total grout volume to fill the 105-R Reactor Disassembly Basin is 24,424 cubic meters or 31,945 cubic yards. Portland cement-based structural fill materials were design and tested for the reactor ISD project and a placement strategy for stabilizing the basin was developed. Based on structural engineering analyses and work flow considerations, the recommended maximum lift height is 5 feet with 24 hours between lifts. Pertinent data and information related to the SRS 105-R-Reactor Disassembly Basin in-situ decommissioning include: regulatory documentation, residual water management, area preparation activities, technology needs, fill material designs

  1. Regulatory trends and practices related to nuclear reactor decommissioning

    International Nuclear Information System (INIS)

    Cantor, R.A.

    1984-01-01

    In the next several decades, the electric utility industry will be faced with the retirement of 50,000 megawatts (mW) of nuclear capacity. Responsibility for the financial and technical burdens this activity entails has been delegated to the utilities operating the reactors. However, the operators will have to perform the tasks of reactor decommissioning within the regulatory environment dictated by federal, state and local regulations. The purpose of this paper is to highlight some of the current and likely trends in regulations and regulatory practices that will significantly affect the costs, technical alternatives and financing schemes encountered by the electric utilities and their customers

  2. Level 3 decommissioning of Triton - Nereide research reactor

    International Nuclear Information System (INIS)

    Lopes, E.; Pillette-Cousin, L.

    2002-01-01

    The French Atomic Energy Commission Center located at Fontenay-Aux-Roses has launched an extensive programme of site cleanup and decommissioning of nuclear facilities. This programme includes the level 3 decommissioning of the Triton and Nereide piles. These pool type research reactors were constructed in the late 1950's, primarily for R and D activities related to neutron physics studies, radiological shielding experiments and radioelement production. As of 1982, a level 2 decommissioning was achieved and over the the last twenty years, no activities were carried out in the facility. During 2001, there has been extensive investigation work carried out to acquire a better knowledge of the radiological status of the facility, in order to set up dismantling scenarios and to reduce the volume of generated radioactive waste. Indeed, one of the first and main operations to be carried out for dismantling Triton and Nereide piles is waste zoning, by using the facility layout, operating conditions and history, as well as the present radiological inventory. The paper describes the investigations and studies carried out to implement waste zoning. The paper also describes the preliminary dismantling operations undertaken on equipment and studies conducted to optimize the dismantling and cleanup of the facility. Finally, the paper presents the outline of the preferred dismantling and decommissioning options and the progress of the work to date. (author)

  3. Technology, safety and costs of decommissioning a reference pressurized water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

    International Nuclear Information System (INIS)

    Konzek, G.J.; Smith, R.I.

    1988-07-01

    Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies on conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference pressurized water reactor (PWR) described in the earlier study; defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs; and completing a study of recent PWR steam generator replacements to determine realistic estimates for time, costs and doses associated with steam generator removal during decommissioning. This report presents the results of recent PNL studies to provide supporting information in four areas concerning decommissioning of the reference PWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; assessing the cost and dose impacts of recent steam generator replacements; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation

  4. Present status of decommissioning in the Musashi Reactor Facility (4)

    International Nuclear Information System (INIS)

    Uchiyama, Takafumi; Tanzawa, Tomio; Mitsuhashi, Ishi; Morishima, Kayoko; Matsumoto, Tetsuo

    2012-01-01

    The decommissioning of the Musashi reactor was decided in 2003. Permanent shutdown of the reactor and stopping the operational functions were conducted in 2004. Transportation of the spent fuels was finished in 2006. After 2007, the system and equipment stopping the functions were stored as installed in the reactor facility as radioactive wastes. After separating nonradioactive wastes such as concretes from radioactive wastes with a contamination test, stopping the functions of liquid waste management facility was performed with newly installed drainage facility for radioisotope use in 2010. Solid waste management facility was also dismantled and removed in the same way as liquid waste management facility in 2011. Radioactive wastes packed in containers were moved and stored in the reactor facility. (T. Tanaka)

  5. Decommissioning and dismantling reactors and managing waste

    International Nuclear Information System (INIS)

    Bensoussan, E.; Reicher-Fournel, N.

    2005-01-01

    In the early forties/fifties, a number of countries launched the first developments in the field of nuclear power. Some of them now have large numbers of nuclear facilities and nuclear power plants which have met, and continue to meet, the objectives for which they were designed and built. Other plants, including nuclear fuel production and enrichment plants, experimental reactors or research reactors, will have to be dismantled and demolished in the near future. These activities are handled differently in different countries as a function of specific energy policies, advanced development plants, current financial resources, the availability of qualified engineers and specialized industries able to handle projects of this kind, as well as other factors. All dismantling and demolition projects serve the purpose of returning the respective sites to green-field conditions. (orig.)

  6. Development programs on decommissioning technology for reactors and fuel cycle facilities in Japan

    International Nuclear Information System (INIS)

    Fujiki, K.

    1992-01-01

    The Science and Technology Agency (STA) of Japan is promoting technology development for decommissioning of nuclear facilities by entrusting various research programs to concerned research organisations: JAERI, PNC and RANDEC, including first full scale reactor decommissioning of JPDR. According to the results of these programs, significant improvement on dismantling techniques, decontamination, measurement etc. has been achieved. Further development of advanced decommissioning technology has been started in order to achieve reduction of duration of decommissioning work and occupational exposures in consideration of the decommissioning of reactors and fuel cycle facilities. (author) 5 refs.; 7 figs.; 1 tab

  7. Nuclear reactor decommissioning: an analysis of the regulatory environments

    International Nuclear Information System (INIS)

    Cantor, R.

    1986-08-01

    In the next several decades, the electric utility industry will be faced withthe retirement of 50,000 megawatts (mW) of nuclear capacity. Responsibility for the financial and technical burdens this activity entails has been delegated to the utilities operating the reactors. However, the operators will have to perform the tasks of reactor decommissioning within the regulatory environment dictated by federal, state and local regulations. The purpose of this study was to highlight some of the current and likely trends in regulations and regulatory practices that will significantly affect the costs, technical alternatives and financing schemes encountered by the electric utilities and their customers. To identify significant trends and practices among regulatory bodies and utilities, a reviw of these factors was undertaken at various levels in the regulatory hierarchy. The technical policies were examined in reference to their treatment of allowed technical modes, restoration of the plant site including any specific recognition of the residual radioactivity levels, and planning requirements. The financial policies were examined for specification of acceptable financing arrangements, mechanisms which adjust for changes in the important parameters used to establish the fund, tax and rate-base treatments of the payments to and earnings on the fund, and whether or not escalation and/or discounting were considered in the estimates of decommissioning costs. The attitudes of regulators toward financial risk, the tax treatment of the decommissioning fund, and the time distribution of the technical mode were found to have the greatest effect on the discounted revenue requirements. Under plausible assumptions, the cost of a highly restricted environment is about seven times that of the minimum revenue requirement environment for the plants that must be decommissioned in the next three decades

  8. Robotics take heat out of reactor. [Windscale AGR decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    Rufford, N

    1986-12-04

    The Windscale prototype reactor is being decommissioned and dismantled. The stages are outlined. The first phase began in 1985 and included construction of a waste packaging plant annexed to the steel dome. The boilers will be cut up and, once decontaminated, probably sold for scrap. The second phase involves dismantling the reactor itself. Much of this will be done by a semi-automatic robot which is being specially built and tested. The robot will have an extendable arm with a manipulator which will be equipped with bolt croppers, shears, a saw and oxypropane cutter. This robot will cut up the pressure vessel in sections ready for encasing in concrete. Lessons learnt from the dismantling will be used in future reactor designs and specifications (eg the need to use steels with fewer impurities, especially cobalt). Ultimate disposal of the concrete waste blocks is undecided. (U.K.).

  9. Technology, safety and costs of decommissioning a refernce boiling water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

    International Nuclear Information System (INIS)

    Konzek, G.J.; Smith, R.I.

    1988-07-01

    Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies of conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference boiling water reactor (BWR) described in the earlier study; and defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs. This report presents the results of recent PNL studies to provide supporting information in three areas concerning decommissioning of the reference BWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation

  10. Example of End States of Decommissioning Phases from the Decommissioning of the Multipurpose Research Reactor MZFR, Karlsruhe, Germany

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-06-15

    The multipurpose reactor MZFR was a pressurized water reactor, cooled and moderated with heavy water. It was built from 1961 to 1966, and went critical for the first time on 29 September 1965. After 19 years of successful operation, the reactor was shut down on 3 May 1984. The reactor had a thermal output of 200 MW, and an electrical output of 50 MW. In addition to generating electricity, the MZFR had the following functions: - Testing fuel assemblies and various materials for reactor construction; - Gaining experience in the design, erection and operation of heavy water reactor systems; - Training scientific and technical reactor personnel; - Providing heat (first nuclear combined heat and power system (1979-1984)). In 1989, it was decided to dismantle the reactor completely, step by step. The decommissioning concept for the plant, down to a greenfield site, provides for eight distinct decommissioning steps (phases). A separate decommissioning licence was required for each step. The decommissioning work was carried out according to pre-approved work schedules. About 72 000 t of concrete and 7200 t of metal were to be removed. About 1000 t of concrete (500 t biological shield) and 1680 t of metal were to be classified as radioactive waste.

  11. The Role of Stakeholders in the Decommissioning of Salaspils Research Reactor

    International Nuclear Information System (INIS)

    Abramenkovs, A.

    2009-01-01

    The paper describes the role of different stakeholders in the decommissioning of the Salaspils Research Reactor. Decommissioning was a large challenge for the Latvia, since the country in this moment had no decommissioning experience and necessary technologies for the implementation of the defined goals by the Government. In this case for facilitation of the decommissioning of Salaspils Research Reactor (SRR), the significant role plays the local and international stakeholders. The paper deals with information on the basic stages of decommissioning of SRR and the role of the wide spectrum of stakeholders in preparation, upgrade and implementation of the decommissioning plan. The role of governmental institutions in the decommissioning of Salaspils research reactor is discussed. It was shown, that local municipalities are very important stakeholders, which significantly influence the decommissioning of SRR. The Salaspils municipalities positive impact on the decommissioning processes are discussed. Basic problems with the Baldone municipality in context of radioactive wastes management are indicated. The role of international stakeholders in decommissioning of Salaspils research reactor is discussed. It was shown, that the support from International Atomic Energy Agency significantly promotes the decommissioning of SRR. The main issues were expert support for solution of different technical problems in radioactive wastes management, area monitoring, and verification of decommissioning plans, training of staff and technical expertise during whole process of decommissioning. It was shown, that technical and economical support from DOE, USA provides the possibility to solve the fuel problem during decommissioning of SRR, as well as, to increase the physical safety of SRR and repository 'Radons'. It was shown, that a proper coordination of all activities and using the services from stakeholders can significantly reduce the total project expenses. The cooperation between

  12. Decommissioning of the AVR reactor, concept for the total dismantling

    International Nuclear Information System (INIS)

    Marnet, C.; Wimmers, M.; Birkhold, U.

    1998-01-01

    After more than 21 years of operation, the 15 MWe AVR experimental nuclear power plant with pebble bed high temperature gas-cooled reactor was shout down in 1988. Safestore decommissioning began in 1994. In order to completely dismantle the plant, a concept for Continued dismantling was developed according to which the plant could be dismantled in a step-wise procedure. After each step, there is the possibility to transform the plant into a new state of safe enclosure. The continued dismantling comprises three further steps following Safestore decommissioning: 1. Dismantling the reactor vessels with internals; 2. Dismantling the containment and the auxiliary units; 3. Gauging the buildings to radiation limit, release from the validity range of the AtG (Nuclear Act), and demolition of the buildings. For these steps, various technical procedures and concepts were developed, resulting in a reference concept in which the containment will essentially remain intact (in-situ concept). Over the top of the outer reactor vessel a disassembling area for remotely controlled tools will be erected that tightens on that vessel and can move down on the vessel according to the dismantling progress. (author)

  13. Decommissioning of the High Flux Beam Reactor at Brookhaven Lab

    Energy Technology Data Exchange (ETDEWEB)

    Hu, J. P. [Brookhaven National Lab. (BNL), Upton, NY (United States); Reciniello, R. N. [Brookhaven National Lab. (BNL), Upton, NY (United States); Holden, N. E. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2011-05-27

    The High Flux Beam Reactor at the Brookhaven National Laboratory was a heavy water cooled and moderated reactor that achieved criticality on October 31, 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shutdown in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor’s spent fuel pool. The reactor remained shutdown for almost three years for safety and environmental reviews. In November 1999 the United States Department of Energy decided to permanently shutdown the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR cleanup conducted during 1999-2009 will be described in the paper.

  14. Radiochemical analysis of concrete samples for decommission of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zapata-Garcia, Daniel; Wershofen, Herbert [Physikalisch-Technische Bundesanstalt (PTB), Bundesallee 100 38116, Braunschweig (Germany); Larijani, Cyrus; Sobrino-Petrirena, Maitane; Garcia-Miranda, Maria; Jerome, Simon M. [National Physical Laboratory (NPL), Hampton Road, Teddington, Middlesex, TW11 0LW (United Kingdom)

    2014-07-01

    Decommissioning of the oldest nuclear power reactors are some of the most challenging technological legacy issues many countries will face in forthcoming years, as many power reactors reach the end of their design lives. Decommissioning of nuclear reactors generates large amounts of waste that need to be classified according to their radioactive content. Approximately 10 % of the contaminated material ends up in different repositories (depending on their level of contamination) while the rest is decontaminated, measured and released into the environment or sent for recycling. Such classification needs to be done accurately in order to ensure that both the personnel involved in decommissioning and the population at large are not needlessly exposed to radiation or radioactive material and to minimise the environmental impact of such work. However, too conservative classification strategies should not be applied, in order to make proper use of radioactive waste repositories since space is limited and the full process must be cost-effective. Implicit in decommissioning and classification of waste is the need to analyse large amounts of material which usually combine a complex matrix with a non-homogeneous distribution of the radionuclides. Because the costs involved are large, it is possible to make great savings by the adoption of best available practices, such as the use of validated methods for on-site measurements and simultaneous determination of more than one radionuclide whenever possible. The work we present deals with the development and the validation of a procedure for the simultaneous determination of {sup 241}Am, plutonium isotopes, uranium isotopes and {sup 90}Sr in concrete samples. Samples are firstly ground and fused with LiBO{sub 2} and Li{sub 2}B{sub 4}O{sub 7}. After dissolution of the fused sample, silicate and alkaline elements are removed followed by radiochemical separation of the target radionuclides using extraction chromatography. Measurement

  15. Safety case methodology for decommissioning of research reactors. Assessment of the long term impact of a flooding scenario

    International Nuclear Information System (INIS)

    Vladescu, G.; Banciu, O.

    1999-01-01

    The paper contains the assessment methodology of a Safety Case fuel decommissioning of research reactors, taking into account the international approach principles. The paper also includes the assessment of a flooding scenario for a decommissioned research reactor (stage 1 of decommissioning). The scenario presents the flooding of reactor basement, radionuclide migration through environment and long term radiological impact for public. (authors)

  16. General aspects to be considered in a research reactor decommissioning plan

    International Nuclear Information System (INIS)

    Grossi, Pablo Andrade; Tello, Cledola Cassia Oliveira de

    2009-01-01

    There are more than 200 research reactors that have either not operated for a considerable period of time and may never return to operation or, are close to permanent shutdown. Many countries do not have a decommissioning policy, and like Brazil not all installations have their decommissioning plan as part of the licensing documentation. Brazil is signatory of Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management, but until now there is no decommissioning policy, and specifically for research reactor there is no decommissioning guidelines in the standards. The Nuclear Technology Development Centre (CDTN/CNEN-MG) has a TRIGA Mark I Research Reactor IPR-R1 in operation for 47 years with 3.6% average fuel burn-up. The original power was 100 kW and it is being licensed for 250 kW, and it needs the decommissioning plan as part of the licensing requirements. In the paper it is presented general aspects and contents of a Research Reactors Decommissioning Plan. As the Brazilian regulatory body so far does not have a decommissioning policy established neither a regulatory framework in this issue, individual efforts are being integrated to establish a National Decommissioning Group (matrix structure) to perform the decommissioning planning and activities. The approach used for IPR-R1 is presented as suggestions to develop the national regulatory standards on this issue and applied to Brazilian Research Reactors and other nuclear facilities. (author)

  17. Role of decommissioning plan and its progress for the PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Zakaria, Norasalwa; Mustafa, Muhammad Khairul Ariff; Anuar, Abul Adli; Idris, Hairul Nizam; Ba'an, Rohyiza

    2014-01-01

    Malaysian nuclear research reactor, the PUSPATI TRIGA Reactor, reached its first criticality in 1982, and since then, it has been serving for more than 30 years for training, radioisotope production and research purposes. Realizing the age and the need for its decommissioning sometime in the future, a ground basis of assessment and an elaborative project management need to be established, covering the entire process from termination of reactor operation to the establishment of final status, documented as the Decommissioning Plan. At international level, IAEA recognizes the absence of Decommissioning Plan as one of the factors hampering progress in decommissioning of nuclear facilities in the world. Throughout the years, IAEA has taken initiatives and drawn out projects in promoting progress in decommissioning programmes, like CIDER, DACCORD and R2D2P, for which Malaysia is participating in these projects. This paper highlights the concept of Decommissioning plan and its significances to the Agency. It will also address the progress, way forward and challenges faced in developing the Decommissioning Plan for the PUSPATI TRIGA Reactor. The efforts in the establishment of this plan helps to provide continual national contribution at the international level, as well as meeting the regulatory requirement, if need be. The existing license for the operation of PUSPATI TRIGA Reactor does not impose a requirement for a decommissioning plan; however, the renewal of license may call for a decommissioning plan to be submitted for approval in future

  18. Role of decommissioning plan and its progress for the PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Norasalwa Zakaria; Muhammad Khairul Ariff Mustafa; Abul Adli Anuar; Hairul Nizam Idris; Rohyiza Baan

    2013-01-01

    Full-text: Malaysian nuclear research reactor, the PUSPATI TRIGA Reactor, reached its first criticality in 1982, and since then, it has been serving for more than 30 years for training, radioisotope production and research purposes. Realizing the age and the need for its decommissioning sometime in the future, a ground basis of assessment and an elaborative project management need to be established, covering the entire process from termination of reactor operation to the establishment of final status, documented as the Decommissioning Plan. At international level, IAEA recognizes the absence of Decommissioning Plan as one of the factors hampering progress in decommissioning of nuclear facilities in the world. Throughout the years, IAEA has taken initiatives and drawn out projects in promoting progress in decommissioning programmes, like CIDER, DACCORD and R2D2P, for which Malaysia is participating in these projects. This paper highlights the concept of Decommissioning plan and its significances to the Agency. It will also address the progress, way forward and challenges faced in developing the Decommissioning Plan for the PUSPATI TRIGA Reactor. The efforts in the establishment of this plan helps to provide continual national contribution at the international level, as well as meeting the regulatory requirement, if need be. The existing license for the operation of PUSPATI TRIGA Reactor does not impose a requirement for a decommissioning plan; however, the renewal of license may call for a decommissioning plan to be submitted for approval in future. (author)

  19. Role of decommissioning plan and its progress for the PUSPATI TRIGA Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zakaria, Norasalwa, E-mail: norasalwa@nuclearmalaysia.gov.my; Mustafa, Muhammad Khairul Ariff, E-mail: norasalwa@nuclearmalaysia.gov.my; Anuar, Abul Adli, E-mail: norasalwa@nuclearmalaysia.gov.my; Idris, Hairul Nizam, E-mail: norasalwa@nuclearmalaysia.gov.my; Ba' an, Rohyiza, E-mail: norasalwa@nuclearmalaysia.gov.my [Malaysian Nuclear Agency, 43000 Kajang, Selangor (Malaysia)

    2014-02-12

    Malaysian nuclear research reactor, the PUSPATI TRIGA Reactor, reached its first criticality in 1982, and since then, it has been serving for more than 30 years for training, radioisotope production and research purposes. Realizing the age and the need for its decommissioning sometime in the future, a ground basis of assessment and an elaborative project management need to be established, covering the entire process from termination of reactor operation to the establishment of final status, documented as the Decommissioning Plan. At international level, IAEA recognizes the absence of Decommissioning Plan as one of the factors hampering progress in decommissioning of nuclear facilities in the world. Throughout the years, IAEA has taken initiatives and drawn out projects in promoting progress in decommissioning programmes, like CIDER, DACCORD and R2D2P, for which Malaysia is participating in these projects. This paper highlights the concept of Decommissioning plan and its significances to the Agency. It will also address the progress, way forward and challenges faced in developing the Decommissioning Plan for the PUSPATI TRIGA Reactor. The efforts in the establishment of this plan helps to provide continual national contribution at the international level, as well as meeting the regulatory requirement, if need be. The existing license for the operation of PUSPATI TRIGA Reactor does not impose a requirement for a decommissioning plan; however, the renewal of license may call for a decommissioning plan to be submitted for approval in future.

  20. Development of the Decommissioning Planning System for the WWR-M Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lobach, Y. [Institute for Nuclear Research, Kiev (Ukraine)

    2013-08-15

    Kiev's research reactor WWR-M is in operation for more than 50 years and its continued operation is planned. At the same time the development of a decommissioning plan is a mandatory requirement of the national legislation and it must be performed at the operational stage of nuclear installation as early as possible. Recently, the Decommissioning Programme for the WWR-M reactor has been developed. The programme covers the whole decommissioning process and represents the main guiding document during the whole decommissioning period, which determines and substantiates the principal technical and organizational activities on the preparation and implementation of the reactor decommissioning, the consequence of the decommissioning stages, the sequence of planned works and measures as well as the necessary conditions and infrastructure for the provision and safe implementation. The programme contains the basic directions of further decommissioning planning aimed on the timely preparation for the reactor decommissioning. This paper describes the status of the WWR-M reactor decommissioning planning attained by the middle of 2011. (author)

  1. Radiation protection planning for decommissioning of research reactor facilities

    International Nuclear Information System (INIS)

    Jackson, Roger; Harman, Neil; Craig, David; Fecitt, Lorna; Lobach, Yuri; Gorlinskij, Juri; Kolyadin, Vyacheslav; Pavlenko, Vytali

    2008-01-01

    The MR reactor at the Russian Research Centre Kurchatov Institute (RRCKI), Moscow was a 50 MW multipurpose material testing and research reactor equipped with nine experimental loop facilities to test prototype fuel for various nuclear power reactors being developed. The reactor was shut down in 1993 and de-fuelled. The experimental loops are located in basement rooms around the reactor. The nature of the research into the characteristics of fuel design and coolant chemistry resulted in fission products and activation products in the test loop equipment. Decommissioning of the loops therefore presents a number of challenges. In addition the city of Moscow has expanded such that the RRC KI is now surrounded by housing which had to be taken into account in the radiological protection planning. This paper describes the techniques proposed to undertake the dismantling operations in order to minimise the radiation exposure to workers and members of the public. Estimates have been made of the worker doses which could be incurred during the dismantling process and the environmental impacts which could occur. These are demonstrated to be as low as reasonably achievable. The work was funded by the UK Department of Business Enterprise and Regulatory Reform (DBERR) (formerly the Department of Trade and Industry) under the Nuclear Safety Programme (NSP) set up to address nuclear safety issues in the Former Soviet Union. (author)

  2. Economic analysis vs. capital-recovery requirements of power reactor decommissioning

    International Nuclear Information System (INIS)

    Ferguson, J.S.

    1980-01-01

    As a consultant to electric utilities the author often becomes involved in the development of policy for capital recovery and in the determination of depreciation rates that will implement the policy. Utility capital recovery is controlled by generally accepted depreciation accounting practices and by regulatory commission accounting rules and, as a result, can differ significantly from engineering economics. Those involved with decommissioning of power reactors should be aware of the depreciation accounting and regulatory framework that dictates capital recovery requirements, whether their involvement is related to engineering economics or capital recovery. This presentation defines that framework, points out several significant implications (particularly tax), describes several conforming capital-recovery methods, describes several techniques that have been used with the decommissioning component in economic analysis of alternative energy sources, and discusses why those involved in economic analysis should learn the accounting and regulatory framework for capital recovery

  3. Shippingport Station Decommissioning Project: Removal of piping and equipment and removal of primary system components

    International Nuclear Information System (INIS)

    1989-01-01

    This report is a technical synopsis of the removal of contaminated and non-contaminated piping and equipment from the Shippingport Station Decommissioning Project (SSDP). The information is provided as a part of the Technology Transfer Program to document dismantling activities in support of reactor decommissioning. 5 refs., 29 figs., 4 tabs

  4. Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Perry, E.; Chrzanowski, J.; Rule, K.; Viola, M.; Williams, M.; Strykowsky, R.

    1999-01-01

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. The Decontamination and Decommissioning (D and D) of the TFTR is scheduled to occur over a period of three years beginning in October 1999. This is not a typical Department of Energy D and D Project where a facility is isolated and cleaned up by ''bulldozing'' all facility and hardware systems to a greenfield condition. The mission of TFTR D and D is to: (a) surgically remove items which can be re-used within the DOE complex, (b) remove tritium contaminated and activated systems for disposal, (c) clear the test cell of hardware for future reuse, (d) reclassify the D-site complex as a non-nuclear facility as defined in DOE Order 420.1 (Facility Safety) and (e) provide data on the D and D of a large magnetic fusion facility. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The record-breaking deuterium-tritium experiments performed on TFTR resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 75 mRem/hr. These radiological hazards along with the size and shape of the Tokamak present a unique and challenging task for dismantling

  5. Estimated long lived isotope activities in ET-RR-1 reactor structural materials for decommissioning study

    International Nuclear Information System (INIS)

    Ashoub, N.; Saleh, H.

    1995-01-01

    The first Egyptian research reactor, ET-RR-1 is tank type with light water as a moderator, coolant and reflector. Its nominal power is 2MWt and the average thermal neutron flux is 10 13 n/cm 2 sec -1 . Its criticality was on the fall of 1961. The reactor went through several modifications and updating and is still utilized for experimental research. A plan for decommissioning of ET-RR-1 reactor should include estimation of radioactivity in structural materials. The inventory will help in assessing the radiological consequences of decommissioning. This paper presents a conservative calculation to estimate the activity of the long lived isotopes which can be produced by neutron activation. The materials which are presented in significant quantities in the reactor structural materials are aluminum, cast iron, graphite, ordinary and iron shot concrete. The radioactivity of each component is dependent not only upon the major elements, but also on the concentration of the trace elements. The main radioactive inventory are expected to be from 60 Co and 55 Fe which are presented in aluminium as trace elements and in large quantities in other construction materials. (author)

  6. RA research reactor in 'Vinca' Institute-approach to the decommissioning

    International Nuclear Information System (INIS)

    Ljubenov, V.Lj.; Pesic, M.P.; Sotic, O.

    2002-01-01

    In this paper short overview of decommissioning process of research reactors according to IAEA standards and world practice is given. Basic technical characteristics and details of operational history of the RA research reactor in Vinca Institute of Nuclear Sciences are present. The main nuclear and radiation safety problems related to the RA reactor facility are defined and the outlines of the future decommissioning project are proposed. (author)

  7. Decommissioning planning and the assessment of alternatives for the Hanford production reactors

    International Nuclear Information System (INIS)

    Miller, C.E. Jr.; Potter, R.F.

    1985-01-01

    Several years ago, the US Department of Energy began assessing alternatives and planning the decommissioning of eight shut-down plutonium production reactors located on the DOE Hanford Site in Washington State. The first of these graphite-moderated, water-cooled, reactors was built and started up in 1944 as part of the World War II Manhattan Project. The last of them started up in 1955. The eight reactors each operated for 12 to 24 years, with all eight operating simultaneously for about 10 years. In the 1960's, production needs declined and the reactors were one-by-one permanently shut down, the last of them in 1971. (A ninth Hanford production reactor, N Reactor, was started up in 1963; it is still operating and is not within the scope of the decommissioning planning and alternatives assessment work reported in this paper). This paper provides an overview description of the decommissioning plan for the eight shut-down Hanford production reactors and their associated fuel storage basins. Included are descriptions of the decommissioning alternatives considered for the facilities, along with discussions of National Environmental Policy Act (NEPA) process activities applicable to the Hanford decommissioning work. The criteria used in assessing decommissioning alternatives and the assumptions used in the decommissioning planning are identified. 4 refs., 8 figs., 3 tabs

  8. Argentina: Disposal aspects of RA-1 research reactor decommissioning waste

    Energy Technology Data Exchange (ETDEWEB)

    Harriague, S; Barberis, C; Cinat, E; Grizutti, C; Scolari, H [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    2007-12-15

    The objective of the project is to analyze disposal aspects of waste from total dismantling of Argentinean research reactors, starting with the oldest one, 48 years old RA-1. In order to estimate decommissioning waste, data was collected from files, area monitoring, measurements, sampling to measure activity and composition, operational history and tracing of operational incidents. Measurements were complemented with neutron activation calculations. Decommissioning waste for RA-1 is estimated to be 71.5 metric tons, most of it concrete (57 tons), the rest being steels, lead and reflector graphite (4.8 tons). Due to their low specific activities, no disposal problems are foreseen in the case of metals and concrete. Disposal of aluminium, steel, lead and concrete is analyzed. On the contrary, as the country has no experience in managing graphite radioactive waste, work was concentrated on that material. Stored (Wigner) energy may exist in RA-1 graphite reflectors irradiated at room temperature. Evaluation of stored energy by calorimetric methods is proposed, and its annealing by inductive heating; HEPA filters should be used to deal with gaseous activity emissions, mainly Cl-36 and C-14. Galvanic corrosion, dust explosion, ignition and oxidation can be addressed and should not become disposal problems. Care must be taken with graphite dust generation and disposal, due to wetting and flotation problems. Lessons learned from the project are presented, and the benefits of sharing international experience are stressed. (author)

  9. Decommissioning of nuclear power plants and research reactors. Safety guide

    International Nuclear Information System (INIS)

    1999-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of nuclear power plants and research reactors. It is intended to provide guidance to national authorities and operating organizations for the planning and safe management of the decommissioning of such installations. This Safety Guide has been prepared through a series of Consultants and Technical Committee meetings. It supersedes former Safety Series publications Nos 52, 74 and 105

  10. Decommissioning of nuclear power plants and research reactors. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of nuclear power plants and research reactors. It is intended to provide guidance to national authorities and operating organizations for the planning and safe management of the decommissioning of such installations. This Safety Guide has been prepared through a series of Consultants and Technical Committee meetings. It supersedes former Safety Series publications Nos 52, 74 and 105

  11. Decommissioning of nuclear power plants and research reactors. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of nuclear power plants and research reactors. It is intended to provide guidance to national authorities and operating organizations for the planning and safe management of the decommissioning of such installations. This Safety Guide has been prepared through a series of Consultants and Technical Committee meetings. It supersedes former Safety Series publications Nos 52, 74 and 105

  12. Manipulator and materials handling systems for reactor decommissioning -Cooperation between the university and the plant operator

    International Nuclear Information System (INIS)

    Schreck, G.; Bach, F. W.; Haferkamp, H.

    1995-01-01

    Nuclear reactor dismantling requires suitable handling systems for tools and disassembled components, as well as qualified and reliable disassembly and cutting techniques. From the angle of radiation protection, remote-controlled handling techniques and underwater techniques are the methods of choice, the latter particularly in continuation of plant operating conditions, and this all the more the more disassembly work proceeds towards the reactor core. With the experience accumulated for 20 years now by the Institut fuer Werkstoffkunde (materials science) of Hannover University by basic research and application-oriented development work in the field of thermal cutting technology, especially plasma arc cutting techniques, as well as development work in the field of remote-controlled materials handling systems, the institute is the cut-out partner for disassembly tasks in reactor decommissioning. (Orig./DG) [de

  13. Development of telerobotic systems for reactor decommissioning, (2)

    International Nuclear Information System (INIS)

    Fujii, Yoshio; Usui, Hozumi; Shinohara, Yoshikuni

    1991-01-01

    This paper describes the prototype heavy-duty telerobotic system constructed as a cold test facility for the development of robotic remote handling system technology in reactor decommissioning. The total system is built up with a multi-functional electrical manipulator system, a manipulator transporter system equipped with a tripedal support mechanism, a monitoring system comprising a 3-D TV monitor, and a computer control system for overall system operation. The manipulator system consists of two master manipulators and two corresponding amphibious slave manipulators with load capacities of 100 and 25 daN, respectively. Valuable engineering experiences for developing more advanced heavy-duty telerobotic system have been gained through designing, constructing and testing the system. (author)

  14. Brazilian nuclear power plants decommissioning plan for a multiple reactor site

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Deiglys B.; Moreira, Joao M.L.; Maiorino, Jose R., E-mail: deiglys.monteiro@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Aplicadas. Programa de Pos-Graduacao em Energia e Engenharia da Energia

    2015-07-01

    Actually, Brazil has two operating Nuclear Power Plants and a third one under construction, all at Central Nuclear Almirante Alvaro Alberto - CNAAA. To comply with regulatory aspects the power plants operator, Eletronuclear, must present to Brazilian Nuclear Regulatory Agency, CNEN, a decommissioning plan. Brazilian experience with decommissioning is limited because none of any nuclear reactor at the country was decommissioned. In literature, decommissioning process is well described despite few nuclear power reactors have been decommissioned around the world. Some different approach is desirable for multiple reactors sites, case of CNAAA site. During the decommissioning, a great amount of wastes will be produced and have to be properly managed. Particularly, the construction of Auxiliary Services on the site could be a good choice due to the possibility of reducing costs. The present work intends to present to the Eletronuclear some aspects of the decommissioning concept and decommissioning management, storage and disposal de wastes, based on the available literature, regulatory standards of CNEN and international experience as well as to suggest some solutions to be implemented at CNAAA site before starts the decommissioning project in order to maximize the benefits. (author)

  15. Brazilian nuclear power plants decommissioning plan for a multiple reactor site

    International Nuclear Information System (INIS)

    Monteiro, Deiglys B.; Moreira, Joao M.L.; Maiorino, Jose R.

    2015-01-01

    Actually, Brazil has two operating Nuclear Power Plants and a third one under construction, all at Central Nuclear Almirante Alvaro Alberto - CNAAA. To comply with regulatory aspects the power plants operator, Eletronuclear, must present to Brazilian Nuclear Regulatory Agency, CNEN, a decommissioning plan. Brazilian experience with decommissioning is limited because none of any nuclear reactor at the country was decommissioned. In literature, decommissioning process is well described despite few nuclear power reactors have been decommissioned around the world. Some different approach is desirable for multiple reactors sites, case of CNAAA site. During the decommissioning, a great amount of wastes will be produced and have to be properly managed. Particularly, the construction of Auxiliary Services on the site could be a good choice due to the possibility of reducing costs. The present work intends to present to the Eletronuclear some aspects of the decommissioning concept and decommissioning management, storage and disposal de wastes, based on the available literature, regulatory standards of CNEN and international experience as well as to suggest some solutions to be implemented at CNAAA site before starts the decommissioning project in order to maximize the benefits. (author)

  16. Decommissioning and decontamination of licensed reactor facilities and demonstration nuclear power plants

    International Nuclear Information System (INIS)

    Lear, G.; Erickson, P.B.

    1975-01-01

    Decommissioning of licensed reactors and demonstration nuclear power plants has been accomplished by mothballing (protective storage), entombment, and dismantling or a combination of these three. The alternative selected by a licensee seems to be primarily based on cost. A licensee must, however, show that the decommissioning process provides adequate protection of the health and safety of the public and no adverse impact on the environment. To date the NRC has approved each of the alternatives in the decommissioning of different facilities. The decommissioning of small research reactors has been accomplished primarily by dismantling. Licensed nuclear power plants, however, have been decommissioned primarily by being placed in a mothballed state in which they continue to retain a reactor license and the associated licensee responsibilities

  17. Decommissioning strategy and schedule for a multiple reactor nuclear power plant site

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Deiglys Borges; Moreira, Joao M.L.; Maiorino, Jose Rubens, E-mail: deiglys.monteiro@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Aplicadas

    2015-07-01

    The decommissioning is an important part of every Nuclear Power Plant life cycle gaining importance when there are more than one plant at the same site due to interactions that can arise from the operational ones and a decommissioning plant. In order to prevent undesirable problems, a suitable strategy and a very rigorous schedule should implemented and carried. In this way, decommissioning tasks such as fully decontamination and dismantling of activated and contaminated systems, rooms and structures could be delayed, posing as an interesting option to multiple reactor sites. The present work aims to purpose a strategy and a schedule for the decommissioning of a multiple reactor site highlighting the benefits of delay operational tasks and constructs some auxiliary services in the site during the stand by period of the shutdown plants. As a case study, will be presented a three-reactor site which the decommissioning process actually is in planning stage and that should start in the next decade. (author)

  18. Decommissioning strategy and schedule for a multiple reactor nuclear power plant site

    International Nuclear Information System (INIS)

    Monteiro, Deiglys Borges; Moreira, Joao M.L.; Maiorino, Jose Rubens

    2015-01-01

    The decommissioning is an important part of every Nuclear Power Plant life cycle gaining importance when there are more than one plant at the same site due to interactions that can arise from the operational ones and a decommissioning plant. In order to prevent undesirable problems, a suitable strategy and a very rigorous schedule should implemented and carried. In this way, decommissioning tasks such as fully decontamination and dismantling of activated and contaminated systems, rooms and structures could be delayed, posing as an interesting option to multiple reactor sites. The present work aims to purpose a strategy and a schedule for the decommissioning of a multiple reactor site highlighting the benefits of delay operational tasks and constructs some auxiliary services in the site during the stand by period of the shutdown plants. As a case study, will be presented a three-reactor site which the decommissioning process actually is in planning stage and that should start in the next decade. (author)

  19. Management of the radioactive waste resulting from the Romanian VVR-S research reactor decommissioning

    International Nuclear Information System (INIS)

    Ene, D.; Cepraga, D.G.

    2002-01-01

    The paper consists in a waste study of the Romanian VVR-S reactor which will be prepared for decommissioning operations after the permanent shutdown (23.12.1997). Calculations were carried out to determine the activity arising from neutron activation of structural materials inside the reactor, considering the design of the facility and its operating rules. To this end, the following method was used: i) Neutron flux distribution within the reactor was calculated using the DORT transport code, based on DLC23 shielding library relating to three cylindrical reference systems of the reactor structure: reactor core, horizontal tube and thermal column; ii) Calculation of the activity of each reactor component at different cooling times was performed by the ANITA2000 code, using the neutron flux, compositional data for each material and the power history of the reactor; iii) Unconditional clearance indexes for all material at various cooling times were calculated using the clearance levels defined in IAEA-TECDOC-855; iv) Total activities and masses by material type, within the waste category and for each decay time were calculated by summation of the data previously classified for each reactor component. The resulting activation inventory and waste masses, falling in IAEA defined waste categories are presented in the paper at periods of 100 days, and 6, 10, 25, and 50 years after reactor the shutdown. For some components of the reactor as: aluminum central vessel, the central iron shielding ring, the time behaviour of both the fin spatial activity distribution and the radionuclide contributions to the total activity are plotted in the paper. (author)

  20. Integration of improved decontamination and characterization technologies in the decommissioning of the CP-5 research reactor, United States of America

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L E; Bhattacharyya, S K [Technology Development Division, Decommissioning Program, Argonne National Laboratory, Argonne, IL (United States)

    2002-02-01

    The aging of research reactors worldwide has resulted in a heightened awareness in the international decommissioning community of the timeliness to review and address the needs of research reactor operators in planning for and eventually performing the decommissioning of these types of facilities. Many reactors already undergoing decommissioning can be used as test beds for evaluating enhanced or new/innovative technologies for decommissioning; it is possible that new techniques could be made available for future research reactor-decommissioning projects. Potentially, the new technologies will result in: reduced radiation doses to the work force, larger safety margins in performing decommissioning and cost and schedule savings to the decommissioners in performing the decommissioning of these facilities. Testing of these enhanced technologies for decontamination, dismantling, characterization, remote operations and worker protection are critical to furthering advancements in the technical specialty of decommissioning. Furthermore, regulatory acceptance and routine utilization for future research decommissioning will be assured by testing and developing these technologies in realistically contaminated environments prior to their use in actual research reactor decommissioning. The decommissioning of the CP-5 Research Reactor located at the ANL-East Site has been completed. In this paper we present results of work performed at Argonne National Laboratory (ANL) in the development, testing and deployment of innovative and/or enhanced technologies for the decommissioning of research reactors. In addition, details are provided on other related U.S. D and D activities, which may be useful to the international research reactor D and D community. (author)

  1. Integration of improved decontamination and characterization technologies in the decommissioning of the CP-5 research reactor, United States of America

    International Nuclear Information System (INIS)

    Boing, L.E.; Bhattacharyya, S.K.

    2002-01-01

    The aging of research reactors worldwide has resulted in a heightened awareness in the international decommissioning community of the timeliness to review and address the needs of research reactor operators in planning for and eventually performing the decommissioning of these types of facilities. Many reactors already undergoing decommissioning can be used as test beds for evaluating enhanced or new/innovative technologies for decommissioning; it is possible that new techniques could be made available for future research reactor-decommissioning projects. Potentially, the new technologies will result in: reduced radiation doses to the work force, larger safety margins in performing decommissioning and cost and schedule savings to the decommissioners in performing the decommissioning of these facilities. Testing of these enhanced technologies for decontamination, dismantling, characterization, remote operations and worker protection are critical to furthering advancements in the technical specialty of decommissioning. Furthermore, regulatory acceptance and routine utilization for future research decommissioning will be assured by testing and developing these technologies in realistically contaminated environments prior to their use in actual research reactor decommissioning. The decommissioning of the CP-5 Research Reactor located at the ANL-East Site has been completed. In this paper we present results of work performed at Argonne National Laboratory (ANL) in the development, testing and deployment of innovative and/or enhanced technologies for the decommissioning of research reactors. In addition, details are provided on other related U.S. D and D activities, which may be useful to the international research reactor D and D community. (author)

  2. General principles of nuclear safety management related to research reactor decommissioning

    International Nuclear Information System (INIS)

    Banciu, Ortenzia; Vladescu, Gabriela

    2003-01-01

    The paper contents the general principles applicable to the decommissioning of research reactors to ensure a proper nuclear safety management, during both decommissioning activities and post decommissioning period. The main objective of decommissioning is to ensure the protection of workers, population and environment against all radiological and non-radiological hazards that could result after a reactor shutdown and dismantling. In the same time, it is necessary, by some proper provisions, to limit the effect of decommissioning for the future generation, according to the new Romanian, IAEA and EU Norms and Regulations. Assurance of nuclear safety during decommissioning process involves, in the first step, to establish of some safety principles and requirements to be taken into account during whole process. In the same time, it is necessary to perform a series of analyses to ensure that the whole process is conducted in a planned and safe manner. The general principles proposed for a proper management of safety during research reactor decommissioning are as follows: - Set-up of all operations included in a Decommissioning Plan; - Set-up and qualitative evaluation of safety problems, which could appear during normal decommissioning process, both radiological and nonradiological risks for workers and public; - Set-up of accident list related to decommissioning process the events that could appear both due to some abnormal working conditions and to some on-site and off-site events like fires, explosions, flooding, earthquake, etc.); - Development and qualitative/ quantitative evaluation of scenarios for each incidents; - Development (and evaluation) of safety indicator system. The safety indicators are the most important tools used to assess the level of nuclear safety during decommissioning process, to discover the weak points and to establish safety measures. The paper contains also, a safety case evaluation (description of facility according to the decommissioning

  3. Action Memorandum for Decommissioning the Engineering Test Reactor Complex under the Idaho Cleanup Project

    International Nuclear Information System (INIS)

    A. B. Culp

    2007-01-01

    This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared and released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessel. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface

  4. Application of clearance principles to radioactive waste from the decommissioning of nuclear reactors

    International Nuclear Information System (INIS)

    Lin Xiaoling; Feng Dingsheng; Dong Yonghe

    2010-01-01

    The definition of clearance is introduced. The principles and dose criterion of clearance are also clarified. The main radionuclides in radioactivity waste and the radioactivity waste which can be cleared are investigated. The techniques for the measurement of radioactivity waste from the decommissioning of nuclear reactors are summarized. This paper provides the scientific criterion and methods for the management of radioactive waste, and lays the foundation for the treatment of radioactive waste from the decommissioning of nuclear reactor. (authors)

  5. Technology, safety and costs of decommissioning a reference boiling water reactor power station: Comparison of two decommissioning cost estimates developed for the same commercial nuclear reactor power station

    International Nuclear Information System (INIS)

    Konzek, G.J.; Smith, R.I.

    1990-12-01

    This study presents the results of a comparison of a previous decommissioning cost study by Pacific Northwest Laboratory (PNL) and a recent decommissioning cost study of TLG Engineering, Inc., for the same commercial nuclear power reactor station. The purpose of this comparative analysis on the same plant is to determine the reasons why subsequent estimates for similar plants by others were significantly higher in cost and external occupational radiation exposure (ORE) than the PNL study. The primary purpose of the original study by PNL (NUREG/CR-0672) was to provide information on the available technology, the safety considerations, and the probable costs and ORE for the decommissioning of a large boiling water reactor (BWR) power station at the end of its operating life. This information was intended for use as background data and bases in the modification of existing regulations and in the development of new regulations pertaining to decommissioning activities. It was also intended for use by utilities in planning for the decommissioning of their nuclear power stations. The TLG study, initiated in 1987 and completed in 1989, was for the same plant, Washington Public Supply System's Unit 2 (WNP-2), that PNL used as its reference plant in its 1980 decommissioning study. Areas of agreement and disagreement are identified, and reasons for the areas of disagreement are discussed. 31 refs., 3 figs., 22 tabs

  6. An analysis of decommissioning costs for the AFRRI TRIGA reactor facility

    International Nuclear Information System (INIS)

    Forsbacka, Matt

    1990-01-01

    A decommissioning cost analysis for the AFRRI TRIGA Reactor Facility was made. AFRRI is not at this time suggesting that the AFRRI TRIGA Reactor Facility be decommissioned. This report was prepared to be in compliance with paragraph 50.33 of Title 10, Code of Federal Regulations which requires the assurance of availability of future decommissioning funding. The planned method of decommissioning is the immediate decontamination of the AFRRI TRIGA Reactor site to allow for restoration of the site to full public access - this is called DECON. The cost of DECON for the AFRRI TRIGA Reactor Facility in 1990 dollars is estimated to be $3,200,000. The anticipated ancillary costs of facility site demobilization and spent fuel shipment is an additional $600,000. Thus the total cost of terminating reactor operations at AFRRI will be about $3,800,000. The primary basis for this cost estimate is a study of the decommissioning costs of a similar reactor facility that was performed by Battelle Pacific Northwest Laboratory (PNL) as provided in USNRC publication NUREG/CR-1756. The data in this study were adapted to reflect the decommissioning requirements of the AFRRI TRIGA. (author)

  7. Operating experience with decommissioning of underground components, USA

    International Nuclear Information System (INIS)

    2006-01-01

    In the USA there has over the years been widespread use of underground piping and tank storage systems, as well as use of other storage vaults and miscellaneous underground storage systems. In most instances these systems are associated with other operating facilities such as large nuclear facilities. There has been considerable experience over the last 60 years in the design, installation, operation and decommissioning of these radioactive waste handling and storage structures and systems. Legislation was enacted in the 1980s that required extensive upgrades to existing systems, with newly installed systems being required to meet stringent installation, operation and removal requirements. This has had an impact on the use of some of these underground systems. One major problem with many of these systems is to gauge the integrity of the systems after they have been in operation for many years and to accurately determine whether they have leaked or not. Depending on the results of these investigations, the next question is whether these systems even need to be excavated or if they can be dispositioned in place rather than expending a large amount of effort to excavate them and remove the materials in question to a disposal site for final dispositioning. In some areas this is possible while in others excavation is required. Over the years that nuclear facilities have been operational in the USA, advantage has been taken of the fact that the earth serves as a good location for emplacement of otherwise obtrusive (or even less aesthetically pleasing) structures. These structures are also still clearly able to support the useful mission they were always intended to fulfil. This arrangement also serves as a structural feature favourable to minimizing radiation exposure levels emanating from systems and components containing radioactive materials. In some cases, embedding of various components such as tanks, pits, vaults, pipes and ducting was an easy way to avoid these

  8. Prediction of Decommissioning Cost for Kijang Research Reactor Using Power Data of DACCORD

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Yun Jeong; Jin, Hyung Gon; Park, Hee Seong; Park, Seung Kook [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    There are 3 types of cost estimate that can be used, and each have a different level of accuracy: (i) Order of magnitude estimate: One without detailed engineering data, where an estimate is prepared using scale-up or -down factors and approximate ratios. It is likely that the overall scope of the project has not been well defined. The level of accuracy expected is -30% to +50%. The cost plans to predict referring to abroad examples as decommissioning cost estimation has still not developed and been commercial method for Kijang research reactor. In Kijang research reactor case, overall scope of business isn't yet decided. Then it is supposed to estimate cost with type (i). The IAEA project, entitled 'DACCORD' (Data Analysis and Collection for Costing of Research Reactor Decommissioning) performs decommissioning costing after collecting and analyzing the information related to research reactors around the world for several years. Also decommissioning costing method development tends to increase in the each country. This paper aims to estimate preliminary decommissioning cost based on total decommissioning cost per thermal power rate of research reactor presented in DACCORD project' data which is collected by member state. In this paper, preliminary decommissioning cost is estimated based on total decommissioning cost per thermal power rate of research reactor presented in DACCORD data which is collected by member state. Although there exists a general tendency for costs to increase with increasing thermal power, the limited data available show that decommissioning costs at any given power level can vary widely, with increased variability at higher power levels. Variations in decommissioning cost for the research reactors of the same or similar thermal power are caused by differences in reactor types and design, decommissioning project scopes, country- specific unit workforce costs, and other reactor or project factors. An important factor for the

  9. The application of VR-GIS to decommissioning decision support system (DDSS) of nuclear reactor

    International Nuclear Information System (INIS)

    Zhu Bo

    2005-01-01

    Advanced management technique and Decision Support System (DSS) are needed to solve the problems of the nuclear reactor decommissioning decision-making. In this study, a kind of new DSS technique for nuclear reactor decommissioning is introduced. It is based on the Virtual Reality (VR) and Geography Information System (GIS), which combine with the scientific management method, operational research, cybernetics and behavior science. The proposed DDSS (Decommissioning Decision Support System) can provide decision-maker the real time 3-D virtual Environment, GIS information and background material of the decommissioning reactor, help to ascertain the decision-making target, modify the decision module and optimize the dismantling plan. The data from three modules (VR Environment Module, VR-DOSE Management Module and Route Layout GIS Module) are used to continuously update and show the statistic at the same time, and the final advice will be given to decision-maker. (authors)

  10. Research reactor utilization, safety, decommissioning, fuel and waste management. Posters of an international conference

    International Nuclear Information System (INIS)

    2005-01-01

    For more than 50 years research reactors have played an important role in the development of nuclear science and technology. They have made significant contributions to a large number of disciplines as well as to the educational and research programmes of about 70 countries world wide. About 675 research reactors have been built to date, of which some 278 are now operating in 59 countries (86 of them in 38 developing Member States). Altogether over 13,000 reactor-years of cumulative operational experience has been gained during this remarkable period. The objective of this conference was to foster the exchange of information on current research reactor concerns related to safety, operation, utilization, decommissioning and to provide a forum for reactor operators, designers, managers, users and regulators to share experience, exchange opinions and to discuss options and priorities. The topical areas covered were: a) Utilization, including new trends and directions for utilization of research reactors. Effective management of research reactors and associated facilities. Engineering considerations and experience related to refurbishment and modifications. Strategic planning and marketing. Classical applications (nuclear activation analysis, isotope production, neutron beam applications, industrial irradiations, medical applications). Training for operators. Educational programmes using a reactor. Current developments in design and fabrication of experimental facilities. Irradiation facilities. Projects for regional uses of facilities. Core management and calculation tools. Future trends for reactors. Use of simulators for training and educational programmes. b) Safety, including experience with the preparation and review of safety analysis reports. Human factors in safety analysis. Management of extended shutdown periods. Modifications: safety analysis, regulatory aspects, commissioning programmes. Engineering safety features. Safety culture. Safety peer reviews and

  11. Decommissioning strategy for the 'RA' research nuclear reactor at the 'Vinca' Institute

    International Nuclear Information System (INIS)

    Matausek, M.V.

    2000-01-01

    Adopting the global strategy for decommissioning of the research reactor RA at the Vinca Institute and preliminary planning of particular activities is necessary independently on the decision of the future status of this reactor, namely even in the case that it is decided to complete the modernization and to use the reactor again. In this paper the global decommissioning strategy for the RA reactor is proposed, as well as the optimal time schedule of particular activities, based on the relevant experiences from other countries (author) [sr

  12. Development of a preliminary decommissioning plan of the reactor IPEN/MB-01

    International Nuclear Information System (INIS)

    Vivas, Ary de Souza; Carneiro, Alvaro Luiz Guimaraes

    2013-01-01

    Around the world, many nuclear plants were built and need to be turned off at a certain time because they are close to their recommended time of use is approximately 50 years. So the IAEA (International Atomic Energy Agency), seeks to guide and recommend, through publications, guidelines for the conduct of activities both for decommissioning nuclear power plants and for research reactors, with special attention to countries that do not have a framework regulatory Legal that sustain the activities of decommissioning. Brazil, so far, does not have a specific standard to guide the steps of the guidelines regarding decommissioning research reactors, having only a standard applied to decommissioning power plants which was published in November 2012. The Nuclear and Energy Research Institute (IPEN) has two research reactors one being the reactor IPEN/MB-01. The aim of this work is to develop a preliminary plan for decommissioning of nuclear reactor research, considering the technical documentation of the system (RAS-Safety Analysis Report), the existing rules of CNEN (National Nuclear Energy Commission), as well as regulatory instructions and recommendations of the IAEA. The preliminary decommissioning plan consists of the presentation of actions and steps required as well as the strategies to be adopted for the shutdown of the facility under the technical and administrative, seeking the safety, health workers and the general public, minimizing environmental impacts. (author)

  13. Technology, safety and costs of decommissioning a reference pressurized water reactor power station. Classification of decommissioning wastes. Addendum 3

    International Nuclear Information System (INIS)

    Murphy, E.S.

    1984-09-01

    The radioactive wastes expected to result from decommissioning of the reference pressurized water reactor power station are reviewed and classified in accordance with 10 CFR 61. The 17,885 cubic meters of waste from DECON are classified as follows: Class A, 98.0%; Class B, 1.2%; Class C, 0.1%. About 0.7% (133 cubic meters) of the waste would be generally unacceptable for disposal using near-surface disposal methods

  14. Technology, safety, and costs of decommissioning reference nuclear research and test reactors: sensitivity of decommissioning radiation exposure and costs to selected parameters

    International Nuclear Information System (INIS)

    Konzek, G.J.

    1983-07-01

    Additional analyses of decommissioning at the reference research and test (R and T) reactors and analyses of five recent reactor decommissionings are made that examine some parameters not covered in the initial study report (NUREG/CR-1756). The parameters examined for decommissioning are: (1) the effect on costs and radiation exposure of plant size and/or type; (2) the effects on costs of increasing disposal charges and of unavailability of waste disposal capacity at licensed waste disposal facilities; and (3) the costs of and the available alternatives for the disposal of nuclear R and T reactor fuel assemblies

  15. Nonlinear Control of Hydraulic Manipulator for Decommissioning Nuclear Reactor

    International Nuclear Information System (INIS)

    Kim, Myoung-Ho; Lee, Sung-Uk; Kim, Chang-Hoi; Choi, Byung-Seon; Moon, Jei-Kwon

    2016-01-01

    Robot technique is need to decommission nuclear reactor because of high radiation environment. Especially, Manipulator systems are useful for dismantling complex structure in a nuclear facility. In addition, Hydraulic system is applied to handle heavy duty object. Since hydraulic system can demonstrate high power. The manipulator with hydraulic power is already developed. To solve this problem, various nonlinear control method includes acceleration control. But, it is difficult because acceleration value is highly noisy. In this paper, the nonlinear control algorithm without acceleration control is studied. To verify, the hydraulic manipulator model had been developed. Furthermore, the numerical simulation is carried out. The nonlinear control without acceleration parameter method is developed for hydraulic manipulator. To verify control algorithm, the manipulator is modeled by MBD and the hydraulic servo system is also derived. In addition, the numerical simulation is also carried out. Especially, PID gain is determined though TDC algorithm. In the result of numerical simulation, tracking performance is good without acceleration control. Thus, the PID though TDC with SMC is good for hydraulic manipulator control

  16. Assessment of management modes for graphite from reactor decommissioning

    International Nuclear Information System (INIS)

    White, I.F.; Smith, G.M.; Saunders, L.J.; Kaye, C.J.; Martin, T.J.; Clarke, G.H.; Wakerley, M.W.

    1984-01-01

    A technological and radiological assessment has been made of the management options for irradiated graphite wastes from the decommissioning of Magnox and advanced gas-cooled reactors. Detailed radionuclide inventories have been estimated, the main contribution being from activation of the graphite and its stable impurities. Three different packaging methods for graphite have been described; each could be used for either sea or land disposal, is logistically feasible and could be achieved at reasonable cost. Leaching tests have been carried out on small samples of irradiated graphite under a variety of conditions including those of the deep ocean bed; the different conditions had little effect on the observed leach rates of radiologically significant radionuclides. Radiological assessments were made of four generic options for disposal of packaged graphite: on the deep ocean bed, in deep geologic repositories at two different types of site, and by shallow land burial. Incineration of graphite was also considered, though this option presents logistical problems. With appropriate precautions during the lifetime of the Cobalt-60 content of the graphite, any of the options considered could give acceptably low doses to individuals, and all would merit further investigation in site-specific contexts

  17. Nonlinear Control of Hydraulic Manipulator for Decommissioning Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myoung-Ho; Lee, Sung-Uk; Kim, Chang-Hoi; Choi, Byung-Seon; Moon, Jei-Kwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Robot technique is need to decommission nuclear reactor because of high radiation environment. Especially, Manipulator systems are useful for dismantling complex structure in a nuclear facility. In addition, Hydraulic system is applied to handle heavy duty object. Since hydraulic system can demonstrate high power. The manipulator with hydraulic power is already developed. To solve this problem, various nonlinear control method includes acceleration control. But, it is difficult because acceleration value is highly noisy. In this paper, the nonlinear control algorithm without acceleration control is studied. To verify, the hydraulic manipulator model had been developed. Furthermore, the numerical simulation is carried out. The nonlinear control without acceleration parameter method is developed for hydraulic manipulator. To verify control algorithm, the manipulator is modeled by MBD and the hydraulic servo system is also derived. In addition, the numerical simulation is also carried out. Especially, PID gain is determined though TDC algorithm. In the result of numerical simulation, tracking performance is good without acceleration control. Thus, the PID though TDC with SMC is good for hydraulic manipulator control.

  18. Management of Materials from the Decommissioning of Nuclear Reactors

    International Nuclear Information System (INIS)

    Braehler, Georg

    2014-01-01

    Georg Braehler of the World Nuclear Association (WNA) gave an insightful presentation on what can be done with materials from the decommissioning of nuclear reactors. The presentation showed that, although the volumes of waste generated seem large, they are in fact small compared to the conventional recycling market and should not have much impact on operations. The main issue surrounding the recycling of these materials is acceptance, both from a public and a legal perspective which are needed to promote a sustainable route for the recovered materials. Georg concluded that recycling is the most practical and affordable process to minimise the environmental impact. Several questions were raised following the presentation about the issue of public acceptance in Germany of recycling metal that has been cleared for release. The main reason for the current public acceptance is that nothing has happened to generate distrust. A comment was also raised about the limited scale of materials from the nuclear industry. The small volumes of metal generated could deter the conventional waste market from accepting the perceived risk of recycling cleared metals from the nuclear industry

  19. Radioactive inventory in structural materials of ET-R R-1 reactor and its implication on decommissioning.

    Energy Technology Data Exchange (ETDEWEB)

    Elkady, A; Amin, E [National center for nuclear safety and radiation control, atomic energy authority, Cairo, (Egypt)

    1995-10-01

    A plan for decommissioning of ET-R R-1 reactor should include estimation of radioactivity in structural materials. The inventory will help in assessing the radiological consequences decommissioning. Conservative calculations have been made to evaluate the activity of the long lived isotopes which can be produced by neutron activation. The materials which are present in significant quantities in the reactor structural materials are aluminium, cast iron, graphite, ordinary and iron shot concrete. The radioactivity of each component is dependent not only upon the major elements, but also on the concentration of the trace elements. The main radioactive inventory are expected to be from Co-60 and Fe-55 which are present in aluminium as trace elements in larger quantities in other construction materials. 2 figs., 4 tabs.

  20. Decommissioning experience of the Japan power demonstration reactor

    International Nuclear Information System (INIS)

    Hoshi, T.; Yanagihara, S.; Tachibana, M.; Momma, T.

    1992-01-01

    Actual dismantling of the Japan Power Demonstration Reactor (JPDR) has been progressing since 1986 aiming to make stage 3 condition as the final goal. Such highly activated components as the reactor pressure vessel (RPV) and the inner portion of biological shield concrete close to the RPV have removed using the remotely operated cutting machines. Useful data on the dismantling techniques and their safety as well as the manpower expenditure and radiation exposure of workers have been obtained. Experiences gained through the dismantling works are described in this paper. (author)

  1. The First Decommissioning of a Fusion Reactor Fueled by Deuterium-Tritium

    International Nuclear Information System (INIS)

    Gentile, Charles A.; Perry, Erik; Rule, Keith; Williams, Michael; Parsells, Robert; Viola, Michael; Chrzanowski, James

    2003-01-01

    The Tokamak Fusion Test Reactor (TFTR) at the Plasma Physics Laboratory of Princeton University (PPPL) was the first fusion reactor fueled by a mixture of deuterium and tritium (D-T) to be decommissioned in the world. The decommissioning was performed over a period of three years and was completed safely, on schedule, and under budget. Provided is an overview of the project and detail of various factors which led to the success of the project. Discussion will cover management of the project, engineering planning before the project started and during the field work as it was being performed, training of workers in the field, the novel adaptation of tools from other industry, and the development of an innovative process for the use of diamond wire to segment the activated/contaminated vacuum vessel. The success of the TFTR decommissioning provides a viable model for the decommissioning of D-T burning fusion devices in the future

  2. Research and development activities for reactor decommissioning. Developing technology of Fuji Electric Co., Ltd

    International Nuclear Information System (INIS)

    Shirakawa, Masahiro; Takaya, Jyunichi; Mizukoshi, Seiji; Hosoda, Hiroshi; Tomizuka, Chiaki; Funaguchi, Susumu; Ito, Katsuhito

    1997-01-01

    Fuji Electric Co., Ltd. is conducting decommissioning R and D for commercial reactor, especially for gas cooled reactor since the construction of the Tokai-1 power station of JAPCO, in the field of system engineering, residual radioactivity evaluation, dismantling of core internals, remote handling, treatment and disposal of radioactive waste, and radioactivity measurement. These R and D have been performed mainly under contract of JAPCO and JAERI. This paper gives a summary of the present status and future plan concerning technical development for decommissioning of nuclear reactor by Fuji Electric Co., Ltd. (author)

  3. Decommissioning of research reactors: Evolution, state of the art, open issues

    International Nuclear Information System (INIS)

    2006-01-01

    Many research reactors throughout the world date from the original nuclear research programmes in Member States. Consequently, dozens of old research reactors are candidates for near term decommissioning in parallel with progressive ageing and technical and economic obsolescence. Many of them are located in countries/institutions that, although familiar with the operation and management of their reactors, do not necessarily have adequate expertise and technologies for planning and implementing state of the art decommissioning projects. It is felt that IAEA reports may contribute to the awareness of technologies and know-how already tested successfully elsewhere. This report addresses a subject area that was dealt with earlier by two IAEA publications, namely, Planning and Management for the Decommissioning of Research Reactors and Other Small Nuclear Facilities (Technical Reports Series No. 351) and Decommissioning Techniques for Research Reactors (Technical Reports Series No. 373). This publication updates those reports in view of the technological progress, experience gained and the progressive ageing of research reactors, many of which have already reached the permanent shutdown stage and should be decommissioned soon. It is intended to contribute to the systematic coverage of the entire range of activities that have been addressed by the IAEA's decommissioning work in past years. The perspective of the report is historical, in that relevant issues are identified as solved, pending, or emerging. Much of the information provided in this report will also be of use for the decommissioning of nuclear power plants and other nuclear facilities. A Technical Committee Meeting on this subject was held in Vienna from 17 to 21 May 2004, at which the participants reviewed a draft report written by consultants from Canada, Germany, Israel, the Russian Federation and the United Kingdom

  4. Identification and evaluation of facilitation techniques for decommissioning light water power reactors

    International Nuclear Information System (INIS)

    LaGuardia, T.S.; Risley, J.F.

    1986-06-01

    This report describes a study sponsored by the US Nuclear Regulatory Commission to identify practical techniques to facilitate the decommissioning of nuclear power generating facilities. The objective of these ''facilitation techniques'' is to reduce the radioactive exposures and/or volumes of waste generated during the decommissioning process. The report presents the possible facilitation techniques identified during the study and discusses the corresponding facilitation of the decommissioning process. Techniques are categorized by their applicability of being implemented during the three stages of power reactor life: design/construction, operation, or decommissioning. Detailed cost-benefit analyses were performed for each technique to determine the anticipated exposure and/or radioactive waste reduction; the estimated costs for implementing each technique were then calculated. Finally, these techniques were ranked by their effectiveness in facilitating the decommissioning process. This study is a part of the Nuclear Regulatory Commission's evaluation of decommissioning policy and its modification of regulations pertaining to the decommissioning process. The findings can be used by the utilities in the planning and establishment of activities to ensure that all objectives of decommissioning will be achieved

  5. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report

    International Nuclear Information System (INIS)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form

  6. Final project report: TA-35 Los Alamos Power Reactor Experiment No. II (LAPRE II) decommissioning project

    International Nuclear Information System (INIS)

    Montoya, G.M.

    1993-02-01

    This final report addresses the decommissioning of the LAPRE II Reactor, safety enclosure, fuel reservoir tanks, emergency fuel recovery system, primary pump pit, secondary loop, associated piping, and the post-remediation activities. Post-remedial action measurements are also included. The cost of the project including, Phase I assessment and Phase II remediation was approximately $496K. The decommissioning operation produced 533 M 3 of mixed waste

  7. Final project report, TA-35 Los Alamos Power Reactor Experiment No. II (LAPRE II) decommissioning project

    International Nuclear Information System (INIS)

    Montoya, G.M.

    1992-01-01

    This final report addresses the decommissioning of the LAPRE II Reactor, safety enclosure, fuel reservoir tanks, emergency fuel recovery system, primary pump pit, secondary loop, associated piping, and the post-remediation activities. Post-remedial action measurements are also included. The cost of the project, including Phase I assessment and Phase II remediation was approximately $496K. The decommissioning operation produced 533 m 3 of low-level solid radioactive waste and 5 m 3 of mixed waste

  8. Project WAGR: the UK demonstration project for power reactor decommissioning - a review of the tools used to dismantle the reactor core

    International Nuclear Information System (INIS)

    Benest, T.G.

    2008-01-01

    The United Kingdom Atomic Energy Authority (UKAEA) has built and operated a wide range of nuclear facilities since the late 1940. UKAEA mission is to restore the environment of its sites in a safe and secure manner. This restoration includes the decommissioning of a number of redundant research and power reactors. The Windscale Advanced Gas-cooled Reactor (WAGR) was the UK prototype Advanced gas cooled reactor and became the forerunner of a family of 14 reactors built to generate cheaper and more efficient electricity in the UK. WAGR was constructed between 1957 and 1961 and was a carbon dioxide cooled, graphite moderated reactor using uranium oxide fuel in stainless steel cans. The reactor consisted of a graphite moderator housed in a cylindrical reactor vessel with hemispherical ends. The reactor and associated heat exchangers were enclosed in the iconic spherical containment building regularly used by the media in the UK as an illustration of the nuclear industry. The reactor first produced power in August 1962 and achieved full design output in 1963. It operated at an electrical output of 33 MW (E) for 18 years (average load factor of 75%). In 1981 the reactor was shut down after satisfactory completion of all the research and development objectives. In anticipation of the UK likely nuclear decommissioning needs the UKAEA decided to decommission WAGR to the International Atomic Energy Agency (IAEA) stage 3 (restoration of the area occupied by the facility to a condition of unrestricted re-usability) as the national demonstration exercise for power reactor decommissioning. Since 1998 the UKAEA and its contractors have been undertaking the dismantling of the reactor core components and pressure vessel in a series of 10 campaigns. These contain neutron activated components expected to produce dose rates well in excess of 1 Sv/hr. To carry out the work UKAEA installed an 8M remote dismantling machine (RDM) a waste recovery and transport system and a shielded waste

  9. TEM [transmission electron microscopy], APFIM [atom-probe field ion microscopy], and SANS [small-angle neutron scattering] examination of aged duplex stainless steel components from some decommissioned reactors

    International Nuclear Information System (INIS)

    Chung, H.M.; Chopra, O.K.

    1987-12-01

    The present investigation indicates that the primary embrittlement processes of the CF-8 grade cast stainless steel components during extended reactor service are spinodal decomposition of the ferrite phase and M 23 C 6 carbide precipitation on the austenite-ferrite boundaries. The ferrite hardness measured for the Shippingport reactor valves appears to reflect the different extent of spinodal decomposition among the different valves which contain slightly different Cr contents. G-phase precipitation was minimal compared to that during accelerated aging of CF-8 steel in the laboratory (i.e., near 400/degree/C). This indicates that the activation energy may be strongly influenced by the synergism among the G-phase precipitation, carbide formation, and spinodal decomposition. 13 refs., 2 figs

  10. Radioactive Wastes Cementation during Decommissioning Of Salaspils Research Reactor

    International Nuclear Information System (INIS)

    Abramenkova, G.; Klavins, M.; Abramenkovs, A.

    2009-01-01

    This paper deals with information on the radioactive wastes cementation technology for decommissioning of Salaspils Research Reactor (SRR). Dismantled radioactive materials were cemented in concrete containers using tritiated water-cement mortar. The laboratory tests system was developed to meet the waste acceptance criteria for disposal of containers with cemented radioactive wastes in near-surface repository 'Radons'. The viscosity of water-cement mortar, mechanical tests of solidified mortar's samples, change of temperature of the samples during solidification time and leakage of Cs-137 and T-3 radionuclides was studied for different water-cement compositions with different additives. The pH and electro conductivity of the solutions during leakage tests were controlled. It was shown, that water/cement ratio significantly influences on water-cement mortar's viscosity and solidified samples mechanical stability. Increasing of water ratio from 0.45 up to 0.62 decreases water-cement mortar's viscosity from 1100 mPas up to 90 mPas and decreases mechanical stability of water-cement samples from 23 N/mm 2 to the 12 N/mm 2 . The role of additives - fly ash and Penetron admix in reduction of solidification temperature is discussed. It was found, that addition of fly ash to the cement-water mortar can reduce the solidification temperature from 81 deg. C up to 62 deg. C. The optimal interval of water ratio in cement mortar is discussed. Tritium and Cs-137 leakage tests show, that radionuclides release curves has a complicate structure. The possible radionuclides release mechanisms are discussed. Experimental results indicated that addition of fly ash result in facilitation of tritium and cesium leakage in water phase. Further directions of investigations are drafted. (authors)

  11. An overview of reactor vessel internals segmentation for nuclear plant decommissioning

    International Nuclear Information System (INIS)

    Litka, T.J.

    1994-01-01

    Several nuclear plants have undergone reactor vessel (RV) internals segmentation as part of or in preparation for decommissioning the plant. In addition, several other nuclear facilities are planning for similar work efforts. The primary technology used for segmentation of RV internals, whether in-air or underwater is Plasma Arc Cutting (PAC). Metal Disintegration Machining (MDM) is also used for difficult to make cuts. PAC and MDM are deployed by various means including Long Handled Tools (LHTs), fixtures, tracks, and multi-axis manipulators. These enable remote cutting due to the radiation and/or underwater environment. A Boiling Water Reactor (BWR), a Pressurized Water Reactor (PWR), and a High Temperature Gas Reactor (HTGR) have had their internals removed and segmented using PAC and MDM. The cutting technology used for each component, location of cut, cut geometry and environment had to be determined well before the actual cutting operations. This allowed for the design, fabrication, and testing of the delivery systems. The technologies, selection process, and methodology for RV internals segmentation will be discussed in this paper

  12. Experimental Boiling Water Reactor decontamination and decommissioning project

    International Nuclear Information System (INIS)

    Fellhauer, C.

    1995-01-01

    The author begins by discussing the problems encountered during decontamination and decommissioning. Next, he discusses waste packaging and recycling. His last topic of lessons learned is subdivided into prevention and early detection, recovery issues, management issues, and noteworthy practices

  13. Operational experience of decommissioning techniques for research reactors in the United Kingdom

    International Nuclear Information System (INIS)

    England, M.R.; McCool, T.M.

    2002-01-01

    In previous co-ordinated research projects (CRP) conducted by the IAEA no distinction was made between decommissioning activities carried out at nuclear power plants, research reactors or nuclear fuel cycle facilities. As experience was gained and technology advanced it became clear that decommissioning of research reactors had certain specific characteristics which needed a dedicated approach. It was within this context that a CRP on Decommissioning Techniques for Research Reactors was launched and conducted by the IAEA from 1997 to 2001. This paper considers the experience gained from the decommissioning of two research reactors during the course of the CRP namely: (a) the ICI Triga Mk I reactor at Billingham UK which was largely complete by the end of the research project and (b) the Argonaut 100 reactor at the Scottish Universities Research and Reactor centre at East Kilbride in Scotland which is currently is the early stages of dismantling/site operations. It is the intention of this paper with reference to the two case studies outlined above to compare the actual implementation of these works against the original proposals and identify areas that were found to be problematical and/or identify any lessons learnt. (author)

  14. Factors relevant to the decommissioning of land-based nuclear reactor plants

    International Nuclear Information System (INIS)

    1980-01-01

    This document applies to all classes of land-based nuclear fission reactors, including those reactors used for the production of electricity or heat, for testing, for research, and for the production of radionuclides. The document covers the technical and administrative aspects related to the conduct of decommissioning, and to the associated radiation protection of man and his environment both during and after decommissioning. The document is intended to provide assistance to those responsible for planning or implementing the decommissioning of a land-based nuclear reactor. The user of this report is further encouraged to review past experience gained with nuclear facilities and the published technical data cited in the section entitled Bibliography

  15. Release from control of inactive material from decommissioning the ASTRA research reactor

    International Nuclear Information System (INIS)

    Brandl, A.; Hrnecek, E.; Steger, F.; Kurz, H.; Meyer, F.; Karacson, P.

    2003-01-01

    The Austrian Research Centers Seibersdorf have been operating a 10 MW ASTRA research reactor from 1960 until 1999. After that date, a submission of the intention to decommission the reactor has been provided to the Competent Authorities. After completion of an Environmental Impact Study by the Competent Authorities and modification of the Permissions for Site Use, the reactor finally entered the decommissioning phase in 2003. Inactive materials from the decommissioning site are expected to be released from control. The procedure for such a release from control agreed upon between the Competent Authorities and ARC Seibersdorf involves a four-step measurement, verification, and certification process detailed in this paper. By September 2003, this four-step procedure has been completed for 16500 kg of steel re-enforced concrete and for 5500 kg of other materials; the release from control of 3000 kg of paraffin and 10000 kg of graphite from the thermal column are planned for the near future. (author)

  16. An evaluation of the worker exposure during the RA reactor decommissioning due to presence of contamination

    International Nuclear Information System (INIS)

    Ljubenov, V.; Simovic, R.

    2005-01-01

    In this paper a possibility to establish the relationship between the presence of the surface contamination in the RA reactor rooms and the worker exposure during the decommissioning activities is discussed. RESRAD-BUILD code and the models typical for the RA reactor has been used to determine annual doses from the supposed activities of the main expected RA reactor contaminants. Different room air exchange rates have been modelled and analysed. (author) [sr

  17. Characterization of decommissioned reactor internals: Monte Carlo analysis technique

    International Nuclear Information System (INIS)

    Reid, B.D.; Love, E.F.; Luksic, A.T.

    1993-03-01

    This study discusses computer analysis techniques for determining activation levels of irradiated reactor component hardware to yield data for the Department of Energy's Greater-Than-Class C Low-Level Radioactive Waste Program. The study recommends the Monte Carlo Neutron/Photon (MCNP) computer code as the best analysis tool for this application and compares the technique to direct sampling methodology. To implement the MCNP analysis, a computer model would be developed to reflect the geometry, material composition, and power history of an existing shutdown reactor. MCNP analysis would then be performed using the computer model, and the results would be validated by comparison to laboratory analysis results from samples taken from the shutdown reactor. The report estimates uncertainties for each step of the computational and laboratory analyses; the overall uncertainty of the MCNP results is projected to be ±35%. The primary source of uncertainty is identified as the material composition of the components, and research is suggested to address that uncertainty

  18. Gas cooled reactor decommissioning. Packaging of waste for disposal in the United Kingdom deep repository

    International Nuclear Information System (INIS)

    Barlow, S.V.; Wisbey, S.J.; Wood, P.

    1998-01-01

    United Kingdom Nirex Limited has been established to develop and operate a deep underground repository for the disposal of the UK's intermediate and certain low level radioactive waste. The UK has a significant Gas Cooled Reactor (GCR) programme, including both Magnox and AGR (Advanced Gas-cooled Reactor) capacity, amounting to 26 Magnox reactors, 15 AGR reactors as well as research and prototype reactor units such as the Windscale AGR and the Windscale Piles. Some of these units are already undergoing decommissioning and Nirex has estimated that some 15,000 m 3 (conditioned volume) will come forward for disposal from GCR decommissioning before 2060. This volume does not include final stage (Stage 3) decommissioning arisings from commercial reactors since the generating utilities in the UK are proposing to adopt a deferred safe store strategy for these units. Intermediate level wastes arising from GCR decommissioning needs to be packaged in a form suitable for on-site interim storage and eventual deep disposal in the planned repository. In the absence of Conditions for Acceptance for a repository in the UK, the dimensions, key features and minimum performance requirements for waste packages are defined in Waste Package Specifications. These form the basis for all assessments of the suitability of wastes for disposal, including GCR wastes. This paper will describe the nature and characteristics of GCR decommissioning wastes which are intended for disposal in a UK repository. The Nirex Waste Package Specifications and the key technical issues, which have been identified when considering GCR decommissioning waste against the performance requirements within the specifications, are discussed. (author)

  19. Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

    International Nuclear Information System (INIS)

    1994-10-01

    The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition

  20. Comparison of standardised decommissioning costing tools on pilot Vienna TRIGA MARK-II research reactor

    International Nuclear Information System (INIS)

    Hornacek, M.; Kristofova, K.; Slugen, V.; Zachar, M.; Stummer, T.

    2017-01-01

    The main purpose of the paper is to compare decommissioning costing code CERREX (Cost Estimation for Research Reactors in Excel) with advanced calculation methodology applied in eOMEGA-RR code. CERREX code was developed in line with the IAEA recommendations for decommissioning costing of research facilities and fully implements the ISDC (International Structure for Decommissioning Costing of Nuclear Installations) structure and costing methodology. In comparison with CERREX, usually applied in preliminary costing, the code eOMEGA-RR incorporates the realistic activity and material flow during decommissioning process (e.g. decontamination, dismantling and waste management). This advanced approach enables to carry out the decommissioning planning and costing more effectively. Moreover, the user-friendly interface helps to perform wide range of sensitivity analyses. In order to meet the above mentioned objectives, the model calculation costing case for TRIGA MARK-II research reactor in Vienna was developed in both calculation codes. The whole process covered four step-by-step procedures to be implemented. At first, inventory database taking into account physical as well as radiological parameters (e.g.: contamination, dose rates, nuclide vectors, limits and conditions) was developed. At second, advanced decommissioning costing case using CERREX and eOMEGA-RR code was created. At third, sensitivity analyses to estimate the impact of changing input parameters on calculated results were performed. Finally, costing results obtained from both cost calculation codes are compared and discussed. (authors)

  1. Los Alamos National Laboratory case studies on decommissioning of research reactors and a small nuclear facility

    International Nuclear Information System (INIS)

    Salazar, M.D.

    1998-01-01

    Approximately 200 contaminated surplus structures require decommissioning at Los Alamos National Laboratory. During the last 10 years, 50 of these structures have undergone decommissioning. These facilities vary from experimental research reactors to process/research facilities contaminated with plutonium-enriched uranium, tritium, and high explosives. Three case studies are presented: (1) a filter building contaminated with transuranic radionuclides; (2) a historical water boiler that operated with a uranyl-nitrate solution; and (3) the ultra-high-temperature reactor experiment, which used enriched uranium as fuel

  2. Los Alamos National Laboratory case studies on decommissioning of research reactors and a small nuclear facility

    Energy Technology Data Exchange (ETDEWEB)

    Salazar, M.D.

    1998-12-01

    Approximately 200 contaminated surplus structures require decommissioning at Los Alamos National Laboratory. During the last 10 years, 50 of these structures have undergone decommissioning. These facilities vary from experimental research reactors to process/research facilities contaminated with plutonium-enriched uranium, tritium, and high explosives. Three case studies are presented: (1) a filter building contaminated with transuranic radionuclides; (2) a historical water boiler that operated with a uranyl-nitrate solution; and (3) the ultra-high-temperature reactor experiment, which used enriched uranium as fuel.

  3. Waste minimization value engineering workshop for the Los Alamos National Laboratory Omega West Reactor Decommissioning Project

    International Nuclear Information System (INIS)

    Hartnett, S.; Seguin, N.; Burns, M.

    1995-01-01

    The Los Alamos National Laboratory Pollution Prevention Program Office sponsored a Value Engineering (VE) Workshop to evaluate recycling options and other pollution prevention and waste minimization (PP/WMin) practices to incorporate into the decommissioning of the Omega West Reactor (OWR) at the laboratory. The VE process is an organized, systematic approach for evaluating a process or design to identify cost saving opportunities, or in this application, waste reduction opportunities. This VE Workshop was a facilitated process that included a team of specialists in the areas of decontamination, decommissioning, PP/WMin, cost estimating, construction, waste management, recycling, Department of Energy representatives, and others. The uniqueness of this VE Workshop was that it used an interdisciplinary approach to focus on PP/WMin practices that could be included in the OWR Decommissioning Project Plans and specifications to provide waste reduction. This report discusses the VE workshop objectives, summarizes the OWR decommissioning project, and describes the VE workshop activities, results, and lessons learned

  4. Waste minimization value engineering workshop for the Los Alamos National Laboratory Omega West Reactor Decommissioning Project

    Energy Technology Data Exchange (ETDEWEB)

    Hartnett, S.; Seguin, N. [Benchmark Environmental Corp., Albuquerque, NM (United States); Burns, M. [Los Alamos National Lab., NM (United States)

    1995-12-31

    The Los Alamos National Laboratory Pollution Prevention Program Office sponsored a Value Engineering (VE) Workshop to evaluate recycling options and other pollution prevention and waste minimization (PP/WMin) practices to incorporate into the decommissioning of the Omega West Reactor (OWR) at the laboratory. The VE process is an organized, systematic approach for evaluating a process or design to identify cost saving opportunities, or in this application, waste reduction opportunities. This VE Workshop was a facilitated process that included a team of specialists in the areas of decontamination, decommissioning, PP/WMin, cost estimating, construction, waste management, recycling, Department of Energy representatives, and others. The uniqueness of this VE Workshop was that it used an interdisciplinary approach to focus on PP/WMin practices that could be included in the OWR Decommissioning Project Plans and specifications to provide waste reduction. This report discusses the VE workshop objectives, summarizes the OWR decommissioning project, and describes the VE workshop activities, results, and lessons learned.

  5. Decommissioning the WAGR

    Energy Technology Data Exchange (ETDEWEB)

    Lawton, H. (UKAEA Windscale Nuclear Power Development Labs.)

    1982-11-01

    The planned decommissioning of the Windscale Advanced Gas-cooled Reactor, which will take about ten years, is discussed with especial reference to the radioactive decay of the reactor components, the problems of disposal of the resulting radioactive waste, and the planning of the necessary engineering works.

  6. Decommissioning the WAGR

    International Nuclear Information System (INIS)

    Lawton, H.

    1982-01-01

    The planned decommissioning of the Windscale Advanced Gas-cooled Reactor, which will take about ten years, is discussed with especial reference to the radioactive decay of the reactor components, the problems of disposal of the resulting radioactive waste, and the planning of the necessary engineering works. (U.K.)

  7. Transforming criticality control methods for EBR-II fuel handling during reactor decommissioning

    International Nuclear Information System (INIS)

    Eberle, C.S.; Dean, E.M.; Angelo, P.L.

    1995-01-01

    A review of the Department of Energy (DOE) request to decommission the Experimental Breeder Reactor-II (EBR-II) was conducted in order to develop a scope of work and analysis method for performing the safety review of the facility. Evaluation of the current national standards, DOE orders, EBR-II nuclear safeguards and criticality control practices showed that a decommissioning policy for maintaining criticality safety during a long term fuel transfer process did not exist. The purpose of this research was to provide a technical basis for transforming the reactor from an instrumentation and measurement controlled system to a system that provides both physical constraint and administrative controls to prevent criticality accidents. Essentially, this was done by modifying the reactor core configuration, reactor operations procedures and system instrumentation to meet the safety practices of ANS-8.1-1983. Subcritical limits were determined by applying established liquid metal reactor methods for both the experimental and computational validations

  8. The decommissioning of commercial magnox gas cooled reactor power stations in the United Kingdom

    International Nuclear Information System (INIS)

    Holt, G.

    1998-01-01

    There are nine commercial Magnox gas-cooled reactor power stations in the United Kingdom. Three of these stations have been shutdown and are being decommissioning, and plans have also been prepared for the eventual decommissioning of the remaining operational stations. The preferred strategy for the decommissioning of the Magnox power stations has been identified as 'Safestore' in which the decommissioning activities are carried out in a number of steps separated by quiescent periods of care and maintenance. The final clearance of the site could be deferred for up to 135 years following station shutdown so as to obtain maximum benefit from radioactive decay. The first step in the decommissioning strategy is to defuel the reactors and transport all spent and new fuel off the site. This work has been completed at all three shutdown stations. Decommissioning work is continuing on the three sites and has involved activities such as dismantling, decontamination, recycling and disposal of some plant and structures, and the preparation of others for retention on the site for a period of care and maintenance. Significant experience has been gained in the practical application of decommissioning, with successful technologies and processes being identified for a wide range of activities. For example, large and small metallic and concrete structures, some with complex geometries, have been successfully decontaminated. Also, the reactors have been prepared for a long period of care and maintenance, with instrumentation and sampling systems having been installed to monitor their continuing integrity. All of this work has been done under careful safety, technical, and financial control. (author)

  9. Radiation protection in the decommissioning of a post accident reactor

    International Nuclear Information System (INIS)

    Rankine, A.; Wilkinson, J.L.; Dalton, J.

    1996-01-01

    This paper describes the control and limitation of dose uptake to operators during the early stages of decommissioning of the Windscale Piles. This was achieved by careful planning, the use of inactive trials. thoughtful use of remote handling techniques and review and feedback of information. Built between 1947 and 1950, the Windscale Piles were shut down following the Windscale Incident in 1957. UKAEA Government Division are now undertaking the early stages of decommissioning of these facilities, removing material from the air and water ducts and preparing for subsequent core removal. As part of the overall strategy of UKAEA GD, this work is being carried out using contract staff including the use of a Managing Agency, W S Atkins (Northern). Decommissioning utilizes the same means of dose reduction and control as any other nuclear operation although sometimes in novel ways. In the Windscale Piles, fully remote operations have been used to remove fuel and debris from the environs of the core which was damaged during the 1957 incident. Much use has also been made of training in mock-up facilities allowing manual techniques to be used for some jobs. The implications of using various different contractors rather than an in-house team is also discussed. It is concluded that decommissioning of major facilities can be carried out within acceptable dose uptake criteria by utilising both novel and adaptations of traditional, active handling techniques. (author)

  10. Activation calculation for the dismantling and decommissioning of a light water reactor using MCNP™ with ADVANTG and ORIGEN-S

    Science.gov (United States)

    Schlömer, Luc; Phlippen, Peter-W.; Lukas, Bernard

    2017-09-01

    The decommissioning of a light water reactor (LWR), which is licensed under § 7 of the German Atomic Energy Act, following the post-operational phase requires a comprehensive licensing procedure including in particular radiation protection aspects and possible impacts to the environment. Decommissioning includes essential changes in requirements for the systems and components and will mainly lead to the direct dismantling. In this context, neutron induced activation calculations for the structural components have to be carried out to predict activities in structures and to estimate future costs for conditioning and packaging. To avoid an overestimation of the radioactive inventory and to calculate the expenses for decommissioning as accurate as possible, modern state-of-the-art Monte-Carlo-Techniques (MCNP™) are applied and coupled with present-day activation and decay codes (ORIGEN-S). In this context ADVANTG is used as weight window generator for MCNP™ i. e. as variance reduction tool to speed up the calculation in deep penetration problems. In this paper the calculation procedure is described and the obtained results are presented with a validation along with measured activities and photon dose rates measured in the post-operational phase. The validation shows that the applied calculation procedure is suitable for the determination of the radioactive inventory of a nuclear power plant. Even the measured gamma dose rates in the post-operational phase at different positions in the reactor building agree within a factor of 2 to 3 with the calculation results. The obtained results are accurate and suitable to support effectively the decommissioning planning process.

  11. Activation calculation for the dismantling and decommissioning of a light water reactor using MCNP™ with ADVANTG and ORIGEN-S

    Directory of Open Access Journals (Sweden)

    Schlömer Luc

    2017-01-01

    Full Text Available The decommissioning of a light water reactor (LWR, which is licensed under § 7 of the German Atomic Energy Act, following the post-operational phase requires a comprehensive licensing procedure including in particular radiation protection aspects and possible impacts to the environment. Decommissioning includes essential changes in requirements for the systems and components and will mainly lead to the direct dismantling. In this context, neutron induced activation calculations for the structural components have to be carried out to predict activities in structures and to estimate future costs for conditioning and packaging. To avoid an overestimation of the radioactive inventory and to calculate the expenses for decommissioning as accurate as possible, modern state-of-the-art Monte-Carlo-Techniques (MCNP™ are applied and coupled with present-day activation and decay codes (ORIGEN-S. In this context ADVANTG is used as weight window generator for MCNP™ i. e. as variance reduction tool to speed up the calculation in deep penetration problems. In this paper the calculation procedure is described and the obtained results are presented with a validation along with measured activities and photon dose rates measured in the post-operational phase. The validation shows that the applied calculation procedure is suitable for the determination of the radioactive inventory of a nuclear power plant. Even the measured gamma dose rates in the post-operational phase at different positions in the reactor building agree within a factor of 2 to 3 with the calculation results. The obtained results are accurate and suitable to support effectively the decommissioning planning process.

  12. The Waste Management Plan integration into Decommissioning Plan of the WWR-S research reactor from Romania

    International Nuclear Information System (INIS)

    Barariu, Gheorghe; Oprescu, Theodor; Filip, Mihaela; Sociu, Florin

    2008-01-01

    The paper presents the progress of the Radioactive Waste Management Plan which accompanies the Decommissioning Plan for research reactor WWR-S located in Magurele, Ilfov, near Bucharest, Romania. The new variant of the Decommissioning Plan was elaborated taking into account the IAEA recommendation concerning radioactive waste management. A new feasibility study for WWR-S decommissioning was also developed. The preferred safe management strategy for radioactive wastes produced by reactor decommissioning is outlined. The strategy must account for reactor decommissioning, as well as rehabilitation of the existing Radioactive Waste Treatment Plant and the upgrade of the Radioactive Waste Disposal Facility at Baita-Bihor. Furthermore, the final rehabilitation of the laboratories and reusing of cleaned reactor building is envisaged. An inventory of each type of radioactive waste is presented. The proposed waste management strategy is selected in accordance with the IAEA assistance. Environmental concerns are part of the radioactive waste management strategy. (authors)

  13. Decontamination and decommissioning project status of the TRIGA mark-2±3 research reactors

    International Nuclear Information System (INIS)

    Jung, K. J.; Baek, S. T.; Jung, W. S.; Park, S. K.; Jung, K. H.

    1999-01-01

    TRIGA Mark-II, the first research reactor in Korea, has operated since 1962, and the second one, TRIGA Mark-III since 1972. Both of them had their operation phased out in 1995 due to their lives and operation of the new research reactor, HANARO at the Korea Atomic Energy Research Institute (KAERI) in Taejeon. Decontamination and decommissioning (D and D) project of the TRIGA Mark-II and Mark-III was started in January 1997 and will be completed in December 2002. In the first year of the project, work was performed in preparation of the decommissioning plan, start of the environmental impact assessment and setup licensing procedure and documentation for the project with cooperation of Korea Institute of Nuclear Safety (KINS). In 1998, Hyundai Engineering Company (HEC) is the main contractor to do design and licensing documentation for the D and D of both reactors. British Nuclear Fuels plc (BNFL) is technical assisting partner of HEC. The decommissioning plan document was submitted to the Ministry of Science and Technology (MOST) for the decommissioning license in December 1998, and it expecting to be issued a license at the end of September 1999. The goal of this project is to release the reactor site and buildings as an unrestricted area. This paper summarizes current status and future plan for the D and D project

  14. Criteria and measurement techniques applicable to residual radioactivity on a decommissioned reactor site

    International Nuclear Information System (INIS)

    Woollam, P.B.

    1988-12-01

    This document summarises the radiological criteria which might be developed to cover the release of a partly decommissioned nuclear reactor site, then looks at the techniques available by which the site could be monitored to assure compliance with these criteria. In particular, the implications of existing levels of radioactive contamination resulting from airburst nuclear weapons tests and the Chernobyl accident are discussed. (author)

  15. Simulation studies for quantification of solid waste during decommissioning of nuclear reactors

    International Nuclear Information System (INIS)

    Sobhan Babu, K.; Gopalakrishnan, R.K.; Gupta, P.C.

    2007-01-01

    Decommissioning is the final phase in the lifecycle of a nuclear installation and in the area of occupational radiation protection, decommissioning constitute a challenge mainly due to the huge and complex radioactive waste generation. In the context of management and disposal of waste and reuse/recycle of usable materials during decommissioning of reactors, clearance levels for relevant radionuclides are of vital importance. During the process of decommissioning radionuclide-specific clearance levels allow the release of a major quantity of materials to the environment, without regulatory considerations. These levels may also be used to declare the usable materials for reuse or recycle. Assessment of activity concentration in huge quantities of material, for the purpose of clearance, is a challenge in decommissioning process. This paper describes the simulation studies being carried out for the design of a monitoring system for the estimation of activity concentration of the decommissioned materials, especially rubbles/concrete, using mathematical models. Several designs were studied using simulation and it was observed that for the estimation of very low levels of activity concentration, to satisfy the conditions of unrestricted releases, detection system using the principle of Emission Computed Tomography (ECT) is the best suitable method. (author)

  16. Post-accident cleanup and decommissioning of a reference pressurized water reactor

    International Nuclear Information System (INIS)

    Murphy, E.S.; Holter, G.M.

    1982-01-01

    This paper summarizes the results of a conceptual study to evaluate the technical requirements, costs, and safety impacts of the cleanup and decommissioning of a large pressurized water reactor (PWR) involved in an accident. The costs and occupational doses for post-accident cleanup and decommissioning are estimated to be substantially higher than those for decommissioning following the orderly shutdown of a reactor. A major factor in these cost and occupational dose increases is the high radiation environment that exists in the containment building following an accident which restricts worker access and increases the difficulty of performing certain tasks. Other factors which influence accident cleanup and decommissioning costs are requirements for the design and construction of special tools and equipment, increased requirements for regulatory approvals, and special waste management needs. Radiation doses to the public from routine accident cleanup and decommissioning operations are estimated to be below permissible radiation dose levels in unrestricted areas and within the range of annual doses from normal background. 6 references, 1 figure, 7 tables

  17. Chemical decontamination of reactor components

    International Nuclear Information System (INIS)

    Riess, R.; Berthold, H.O.

    1977-08-01

    A solution for the decontamination of reactor components of the primary system was developed. This solution is a modification of the APAC- (Alkaline Permanganate Ammonium Citrate) system described in the literature. The most important advantage of the present solution over the APAC-method is that it does not induce any selective corrosion attack on materials like stainless steel (austenitic), Inconel 600 and Incoloy 800. (orig.) [de

  18. Technology development and demonstration for TRIGA research reactor decontamination, decommissioning and site restoration

    International Nuclear Information System (INIS)

    Oh, Won Zin; Jung, Ki Jung; Lee, Byung Jik

    1997-01-01

    This paper describes the introduction to research reactor decommissioning plan at KAERI, the background of technology development and demonstration, and the current status of the system decontamination technology for TRIGA reactors, concrete decontamination and dust treatment technologies, wall ranging robot and graphic simulation of dismantling processes, soil decontamination and restoration technology, recycling or reuse technologies for radioactive metallic wastes, and incineration technology demonstration for combustible wastes. 9 figs

  19. Behavior of generated aerosols in decommissioning of reactor

    International Nuclear Information System (INIS)

    Tomii, H.; Nakamura, K.

    1999-01-01

    Generated aerosols in dismantling of the JPDR were investigated for making an estimation of air contamination. The maximum dose equivalent rate at the surface of each reactor component was 9.4 Sv/h for core shroud, 80 mSv/h for pressure vessel, 2.0 mSv/h for biological shield, respectively. An under-water cutting method with remote handling plasma torch was used for dismantling of the core shroud and the pressure vessel. The biological shield was dismantled by an in-air cutting method and a controlled blasting method. Pipes connected to recirculation system were dismounted by a conventional mechanical and thermal cutting machine in the air. Generated radioactive aerosols were collected in the exhaust air of green house which enclosed the upper part of the reactor room to control the air contamination. An Andersen sampler was used for the measurement of particle distribution in the aerosols. Most of the particle size was below 0.1 μm in the under-water cutting method. The particle size distribution in the in-air cutting method, however, was divided into two parts at 0.1 μm and 0.3 μm. Dispersion rate of aerosol into the atmosphere was decreased exponentially with the depth of water. The dispersion rate and the size distribution of aerosol generated during cutting of the stainless steel pipes and blasting of the biological shield are also reported in the paper. (Suetake, M.)

  20. Manufacture of components for Canadian reactor programs

    International Nuclear Information System (INIS)

    Perry, L.P.

    Design features, especially those relating to calandrias, are pointed out for many CANDU-type reactors and the Taiwan research reactor. The special requirements shouldered by the Canadian suppliers of heavy reactor components are analyzed. (E.C.B.)

  1. A survey of commercially available manipulators, end-effectors, and delivery systems for reactor decommissioning activities

    International Nuclear Information System (INIS)

    Henley, D.R.; Litka, T.J.

    1996-05-01

    Numerous nuclear facilities owned by the U.S. Department of Energy (DOE) are under consideration for decommissioning. Currently, there are no standardized, automated, remote systems designed to dismantle and thereby reduce the size of activated reactor components and vessels so that they can be packaged and shipped to disposal sites. Existing dismantling systems usually consist of customized, facility-specific tooling that has been developed to dismantle a specific reactor system. Such systems have a number of drawbacks. Generally, current systems cannot be disassembled, moved, and reused. Developing and deploying the tooling for current systems is expensive and time-consuming. In addition, the amount of manual work is significant because long-handled tools must be used; as a result, personnel are exposed to excessive radiation. A standardized, automated, remote system is therefore needed to deliver the tooling necessary to dismantle nuclear facilities at different locations. Because this system would be reusable, it would produce less waste. The system would also save money because of its universal design, and it would be more reliable than current systems

  2. Project management system for the decommissioning of research reactors

    International Nuclear Information System (INIS)

    Park, J. H.

    2006-01-01

    KAERI has developed a computer information system, named DECOMMIS, for the project management with the increased effectiveness of the decommissioning projects and the record keeping for a next decommissioning project. The management system consists of three parts, code management system, data input system (DDIS) and data processing and output system (DDPS). Through the DDIS, the data can be directly inputted at sites and the system can play roles of daily work reports to minimize the time gap between the dismantling activities and the evaluation of the data for project management. The DDPS provides useful information to the staff for more effective project management and this information include several fields, such as project progress management, man power management, waste management, radiation dose of workers and so on. It is expected that the system would enable to maintain the decommissioning data, to prepare the source data for the R and D for development of planning tools and to give information to the staff for the decision on the progress of the projects. In this paper, the overall system will be briefly explained and several examples of the utilization, focused on the waste and manpower control, for the project management will be introduced

  3. Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

    International Nuclear Information System (INIS)

    Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

    2005-01-01

    G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

  4. Technology, safety and costs of decommissioning a Reference Boiling Water Reactor Power Station. Main report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWe.

  5. Experience of partial dismantling and large component removal of light water reactors

    International Nuclear Information System (INIS)

    Dubourg, M.

    1987-01-01

    Not any of the French PWR reactors need to be decommissioned before the next decade or early 2000. However, feasibility studies of decommissioning have been undertaken and several dismantling scenarios have been considered including the dismantling of four PWR units and the on-site entombment of the active components into a reactor building for interim disposal. In addition to theoretical evaluation of radwaste volume and activity, several operations of partial dismantling of active components and decontamination activities have been conducted in view of dismantling for both PWR and BWR units. By analyzing the concept of both 900 and 1300 MWe PWR's, it appears that the design improvements taken into account for reducing occupational dose exposure of maintenance personnel and the development of automated tools for performing maintenance and repairs of major components, contribute to facilitate future dismantling and decommissioning operations

  6. UK contributions to the decommissioning of the BN-350 reactor in Kazakhstan: 2002 – 2011

    International Nuclear Information System (INIS)

    Wells, D.

    2011-01-01

    UK assistance with the decommissioning of BN-350 has cost ~£8.9 million over ten years, ~£4 million spent directly in Kazakhstan. The Programme has immobilised key wastes, contributed to irreversible shutdown of the reactor and addressed issues associated with sodium coolant processing. The Programme funded the operations to load spent fuel canisters into casks at BN-350, together with their despatch from site and receipt at the secure storage facility. The Programme also delivered technical and project management training, assisted in the production of the BN-350 Decommissioning Plan and contributed to the radiation survey effort in the STS

  7. Technologies for gas cooled reactor decommissioning, fuel storage and waste disposal. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1998-09-01

    Gas cooled reactors (GCRs) and other graphite moderated reactors have been important part of the world's nuclear programme for the past four decades. The wide diversity in status of this very wide spectrum of plants from initial design to decommissioning was a major consideration of the International Working group on Gas Cooled Reactors which recommended IAEA to convene a Technical Committee Meeting dealing with GCR decommissioning, including spent fuel storage and radiological waste disposal. This Proceedings includes papers 25 papers presented at the Meeting in three sessions entitled: Status of Plant Decommissioning Programmes; Fuels Storage Status and Programmes; waste Disposal and decontamination Practices. Each paper is described here by a separate abstract

  8. Use Of Cementitious Materials For SRS Reactor Facility In-Situ Decommissioning - 11620

    International Nuclear Information System (INIS)

    Langton, C.; Stefanko, D.; Serrato, M.; Blankenship, J.; Griffin, W.; Waymer, J.; Matheny, D.; Singh, D.

    2010-01-01

    The United States Department of Energy (US DOE) concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate in tact, structurally sound facilities that are no longer needed for their original purpose of, i.e., producing (reactor facilities), processing (isotope separation facilities) or storing radioactive materials. The Savannah River Site 105-P and 105-R Reactor Facility ISD requires about 250,000 cubic yards of grout to fill the below grade structure. The fills are designed to prevent subsidence, reduce water infiltration, and isolate contaminated materials. This work is being performed as a Comprehensive Environmental Response, Compensations and Liability Act (CERCLA) action and is part of the overall soil and groundwater completion projects for P- and R-Areas. Cementitious materials were designed for the following applications: (1) Below grade massive voids/rooms: Portland cement-based structural flowable fills for - Bulk filling, Restricted placement and Underwater placement. (2) Special below grade applications for reduced load bearing capacity needs: Cellular portland cement lightweight fill (3) Reactor vessel fills that are compatible with reactive metal (aluminum metal) components in the reactor vessels: Calcium sulfoaluminate flowable fill, and Magnesium potassium phosphate flowable fill. (4) Caps to prevent water infiltration and intrusion into areas with the highest levels of radionuclides: Portland cement based shrinkage compensating concrete. A system engineering approach was used to identify functions and requirements of the fill and capping materials. Laboratory testing was performed to identify candidate formulations and develop final design mixes. Scale-up testing was performed to verify material production and placement as well as fresh and cured properties. The 105-P and 105-R ISD projects are currently in progress and are expected to be complete in 2012. The focus of this paper is to describe the (1) grout mixes

  9. USE OF CEMENTITIOUS MATERIALS FOR SRS REACTOR FACILITY IN-SITU DECOMMISSIONING - 11620

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C.; Stefanko, D.; Serrato, M.; Blankenship, J.; Griffin, W.; Waymer, J.; Matheny, D.; Singh, D.

    2010-12-07

    The United States Department of Energy (US DOE) concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate in tact, structurally sound facilities that are no longer needed for their original purpose of, i.e., producing (reactor facilities), processing (isotope separation facilities) or storing radioactive materials. The Savannah River Site 105-P and 105-R Reactor Facility ISD requires about 250,000 cubic yards of grout to fill the below grade structure. The fills are designed to prevent subsidence, reduce water infiltration, and isolate contaminated materials. This work is being performed as a Comprehensive Environmental Response, Compensations and Liability Act (CERCLA) action and is part of the overall soil and groundwater completion projects for P- and R-Areas. Cementitious materials were designed for the following applications: (1) Below grade massive voids/rooms: Portland cement-based structural flowable fills for - Bulk filling, Restricted placement and Underwater placement. (2) Special below grade applications for reduced load bearing capacity needs: Cellular portland cement lightweight fill (3) Reactor vessel fills that are compatible with reactive metal (aluminum metal) components in the reactor vessels: Calcium sulfoaluminate flowable fill, and Magnesium potassium phosphate flowable fill. (4) Caps to prevent water infiltration and intrusion into areas with the highest levels of radionuclides: Portland cement based shrinkage compensating concrete. A system engineering approach was used to identify functions and requirements of the fill and capping materials. Laboratory testing was performed to identify candidate formulations and develop final design mixes. Scale-up testing was performed to verify material production and placement as well as fresh and cured properties. The 105-P and 105-R ISD projects are currently in progress and are expected to be complete in 2012. The focus of this paper is to describe the (1) grout mixes

  10. Decommissioning the Jason Argonaut research reactor at a world heritage site

    Energy Technology Data Exchange (ETDEWEB)

    Lockwood, R.J.S.; Beeley, P.A. [HMS Sultan, Gosport, Hampshire (United Kingdom)

    2001-07-01

    The Jason low power Argonaut type, water and graphite moderated reactor was located in King William Building, which is a Grade 1 listed building within the Royal Naval College, Greenwich, London. The College itself is a Scheduled Ancient Monument with World Heritage Site status and is situated about a mile from the Greenwich Dome. The decision to decommission Jason to International Atomic Energy Agency Stage 3 status (unrestricted site use) was taken in 1996. All physical decommissioning work was completed by October 1999, site radiological clearance was obtained in November 1999, the site license was withdrawn and the site was handed over for future unrestricted use on 9 December 1999. The Jason decommissioning project was safely completed to time, cost and quality by the Millennium [2] without any adverse effects on World Heritage aspects of the site. In this paper details are provided about the Jason fuel removal phase and an outline of the other phases of the project.

  11. Irradiated concrete maze is confronted by robotics. [Uncertainties of nuclear reactor decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    Holmes, A

    1984-09-01

    Nuclear reactor decommissioning and demolition are discussed. Three stages of the process are defined, and three options are described, depending on the rate at which the stages of the process are carried out. The options are: immediate decommissioning and demolition within 10 to 15 years of shutdown; partial deferment, the final stage being deferred for 10 to 100 years; total deferment, the second and third stages being deferred for 50 years or more. The possibilities and problems of designing a task-specific robot to carry out decommissioning are discussed. It is pointed out that specialist demolition will be needed. The problem of massive amounts of radioactive waste disposal is considered. The large unknown cost of the operation, and the desirability of getting experience in the problems involved, are discussed.

  12. Decommissioning of the research nuclear reactor WWR-S Magurele - Bucharest. General presentation of the project

    International Nuclear Information System (INIS)

    Dragulescu, Emilian; Dragusin, Mitica; Popa, Victor; Boicu, Alin; Tuca, Carmen; Iorga, Ioan; Vrabie, Ionut; Mustata, Carmen

    2003-01-01

    A decommissioning project was worked out concerning the nuclear facility research reactor WWR-S Magurele-Bucharest to remove the radioactive and hazardous materials and so to exclude any risk for human health and environment. The project involves the four phases named assessment, development, operations and closeout. There are two major parts to the assesment phase: preliminary characterisation and the review and decision-making process. Characterisation is needed to develop project baseline data, which should include sufficient chemical, physical, and radiological characterisation to meet planning needs. Based on the conclusions of these studies, possible decommissioning alternative will be analyzed and: the best alternative chosen, final goal identified, risk assessments are evaluated. Also, taken into account are: regulations supporting assessment, land use considerations, financial concerns, disposal availability, public involvement, technology developments. After a decommissioning alternative was chosen, detailed engineering will begin following appropriate regulatory guidance. The plan will include characterisation information, namely: review of decommissioning alternatives; justification for the selected alternative; provision for regulatory compliance; predictions of personnel exposure, radioactive waste volume, and cost. Other activities are: scheduling, preparation for decommissioning operations; coordination, documentation, characterization report, feasibility studies, Decommissioning Plan, project daily report, radiological survey, airborne sampling records, termination survey of the site. The operations imply: identification and sequencing the operations on contaminated materials, storing on site the wastes, awaiting processing or disposal, and packaging of materials for transport to processing or disposal facilities.The key operations are: worker protection, health and safety program, review of planing work, work area assessment, work area controls

  13. Nuclear power plant decommissioning. The nature of problems

    Energy Technology Data Exchange (ETDEWEB)

    Yunus, Yaziz

    1986-04-01

    A number of issues have to be taken into account before the introduction of any nuclear power plant in any country. These issues include reactor safety (site and operational), waste disposal and, lastly, the decommissioning of the reactor inself. Because of the radioactive nature of the components, nuclear power plants require a different approach to decommission compared to other plants. Until recently, issues on reactor safety and waste disposal were the main topics discussed. As for reactor decommissioning, the debates have been academic until now. Although reactors have operated for 25 years, decommissioning of retired reactors has simply not been fully planned. But the Shippingport Atomic Power Plant in Pennysylvania, the first large-scale power reactor to be retired, is now being decommissioned. The work has rekindled the debate in the light of reality. Outside the United States, decommissioning is also being confronted on a new plane.

  14. IEA-R1 research reactor: operational life extension and considerations regarding future decommissioning

    International Nuclear Information System (INIS)

    Frajndlich, Roberto

    2009-01-01

    The IEA-R1 reactor is a pool type research reactor moderated and cooled by light water and uses graphite and beryllium reflectors. The reactor is located at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), in the city of Sao Paulo, Brazil. It is the oldest research reactor in the southern hemisphere and one of the oldest of this kind in the world. The first criticality of the reactor was obtained on September 16, 1957. Given the fact that Brazil does not have yet a definitive radioactive waste repository and a national policy establishing rules for the spent fuel storage, the institutions which operate the research reactors for more than 50 years in the country have searched internal solutions for continued operation. This paper describes the spent fuel assemblies and radioactive waste management process for the IEA-R1 reactor and the refurbishment and modernization program adopted to extend its lifetime. Some considerations about the future decommissioning of the reactor are also discussed which, in my opinion, might help the operating organization to make decisions about financial, legal and technical aspects of the decommissioning procedures in a time frame of 10-15 years(author)

  15. Development of computer systems for planning and management of reactor decommissioning

    International Nuclear Information System (INIS)

    Yanagihara, Satoshi; Sukegawa, Takenori; Shiraishi, Kunio

    2001-01-01

    The computer systems for planning and management of reactor decommissioning were developed for effective implementation of a decommissioning project. The systems are intended to be applied to construction of work breakdown structures and estimation of manpower needs, worker doses, etc. based on the unit productivity and work difficulty factors, which were developed by analyzing the actual data on the JPDR dismantling activities. In addition, information necessary for project planning can be effectively integrated as a graphical form on a computer screen by transferring the data produced by subprograms such as radioactive inventory and dose rate calculation routines among the systems. Expert systems were adopted for modeling a new decommissioning project using production rules by reconstructing work breakdown structures and work specifications. As the results, the systems were characterized by effective modeling of a decommissioning project, project management data estimation based on feedback of past experience, and information integration through the graphical user interface. On the other hands, the systems were validated by comparing the calculated results with the actual manpower needs of the JPDR dismantling activities; it is expected that the systems will be applicable to planning and evaluation of other decommissioning projects. (author)

  16. Analysis of Removal Alternatives for the Heavy Water Components Test Reactor at the Savannah River Site

    International Nuclear Information System (INIS)

    Owen, M.B.

    1996-08-01

    This engineering study was developed to evaluate different options for decommissioning of the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site. This document will be placed in the DOE-SRS Area reading rooms for a period of 30 days in order to obtain public input to plans for the demolition of HWCTR

  17. Decommissioning of the research reactors at the Russian Research Centre Kurchatov Institute

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoy, N.N.; Ryantsev, E.P.; Kolyadin, V.I.; Kucharkin, N.E.; Melkov, E.S.; Gorlinsky, Yu.E.; Kyznetsova, T.I.; Bulkin, B.K.

    2002-01-01

    The Kurchatov Institute is the largest research center of Russia in the field of nuclear science and engineering. It comprises more than 10 research institutes and scientific-technological complexes carrying out research work in the field of safe development of atomic engineering, controlled thermonuclear fusion, and plasma physics, nuclear physics and elementary particle physics, research reactors, radiation materials technology, solid state physics and superconductivity, molecular and chemical physics, and also perspective know-how's, information science and ecology. This report is basically devoted to the decommissioning of the research reactor installations, in particular to the reactor MR because of the volume and complexity of actions involved. (author)

  18. Decontamination and decommissioning of the SPERT-I Reactor Building at the Idaho National Engineering Laboratory. Final report

    International Nuclear Information System (INIS)

    Dolenc, M.R.

    1986-02-01

    This final report documents the decontamination and decommissioning of the SPERT-I Reactor Building. This 20- by 40-ft galvanized steel building was dismantled; and the resultant contaminated sludge, liquid, and carbon steel were disposed of at the Radioactive Waste Management Complex of the Idaho National Engineering Laboratory. This report presents the results of the characterization, decision analysis, planning, and decommissioning of the facility. The total cost of these activities was $139,500. Of this total, $103,500 was required for decommissioning operations. (This latter figure represents a 20% savings over the estimated costs generated during the planning effort.) The objectives of decommissioning this facility were to stabilize the seepage pit area and remove the reactor building. The D and D work was divided into two parts; the seepage pit was decommissioned in 1984, and the reactor building in 1985. The entire area was backfilled with radiologically clean soil, graded, and seeded. Two markers were installed to identify the locations of the pit and reactor building. The only isotopes found in either decommissioning operation were cesium-137 and uranium-235 in very low concentrations. Decommissioning operations of the reactor building were carried out during August 1985. The project generate 297 ft 3 of radioactive waste. No personnel radiation exposure above background was received by D and D workers

  19. Use of Cementitious Materials for SRS Reactor Facility In-Situ Decommissioning

    International Nuclear Information System (INIS)

    Langton, C.A.; Stefanko, D.B.; Serrato, M.G.; Blankenship, J.K.; Griffin, W.G.; Long, J.T.

    2013-01-01

    The United States Department of Energy (US DOE) concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate intact, structurally sound facilities that are no longer needed for their original purpose of producing (reactor facilities), processing (isotope separation facilities) or storing radioactive materials. The Savannah River Site 105-P and 105-R Reactor Facility ISD project requires approximately 250000 cubic yards of cementitious materials to fill the below-grade structure. The fills are designed to prevent subsidence, reduce water infiltration, and isolate contaminated materials. This work is being performed as a Comprehensive Environmental Response, Compensations and Liability Act (CERCLA) action and is part of the overall soil and groundwater completion projects for P- and R-Areas. Funding is being provided under the American Recovery and Reinvestment Act (ARRA). Cementitious materials were designed for the following applications: (A) Below-grade massive voids / rooms: Portland cement-based structural flowable fills for: (A.1) Bulk filling; (A.2) Restricted placement and (A.3) Underwater placement. (B) Special below-grade applications for reduced load bearing capacity needs: (B.1) Cellular portland cement lightweight fill. (C) Reactor vessel fills that are compatible with reactive metal (aluminum metal) components in the reactor vessels (C.1) Blended calcium aluminate - calcium sulfate based flowable fill; (C.2) Magnesium potassium phosphate flowable fill. (D) Caps to prevent water infiltration and intrusion into areas with the highest levels of radionuclides: (D.1) Portland cement based shrinkage compensating concrete. A system engineering approach was used to identify functions and requirements of the fill and capping materials. Laboratory testing was performed to identify candidate formulations and develop final design mixes. Scale-up testing was performed to verify material production and placement as well as fresh and cured

  20. Tritium in fusion reactor components

    International Nuclear Information System (INIS)

    Watson, J.S.; Fisher, P.W.; Talbot, J.B.

    1980-01-01

    When tritium is used in a fusion energy experiment or reactor, several implications affect and usually restrict the design and operation of the system and involve questions of containment, inventory, and radiation damage. Containment is expected to be particularly important both for high-temperature components and for those components that are prone to require frequent maintenance. Inventory is currently of major significance in cases where safety and environmental considerations limit the experiments to very low levels of tritium. Fewer inventory restrictions are expected as fusion experiments are placed in more-remote locations and as the fusion community gains experience with the use of tritium. However, the advent of power-producing experiments with high-duty cycle will again lead to serious difficulties based principally on tritium availability; cyclic operations with significant regeneration times are the principal problems

  1. Post-accident cleanup and decommissioning of a reference pressurized-water reactor

    International Nuclear Information System (INIS)

    Murphy, E.S.; Holter, G.M.

    1982-10-01

    This paper summarizes the results of a conceptual study to evaluate the technical requirements, costs, and safety impacts of the cleanup and decommissioning of a large pressurized water reactor (PWR) involved in an accident. The costs and occupational doses for post-accident cleanup and dcommissioning are estimated to be substantially higher than those for decommissioning following the orderly shutdown of a reactor. A major factor in these cost and occupational dose increases is the high radiation environment that exists in the containment building following an accident which restricts worker access and increases the difficulty of performing certain tasks. Other factors which influence accident cleanup and decommissioning costs are requirements for the design and construction of special tools and equipment, increased requirements for regulatory approvals, and special waste management needs. Radiation doses to the public from routine accident cleanup and decommissioning operations are estimated to be below permissible radiation dose levels in unrestricted areas and within the range of annual doses from normal background

  2. Planning and management for the decommissioning of research reactors and other small nuclear facilities

    International Nuclear Information System (INIS)

    1993-01-01

    Many research reactors and other small nuclear facilities throughout the world date from the original nuclear research programmes in the Member States. Consequently, a large number of these plants have either been retired from service or will soon reach the end of their useful lives and are likely to become significant decommissioning tasks for those Members States. In recognition of this situation and in response to considerable interest shown by Member States, the IAEA has produced this document on planning and management for the decommissioning of research reactors and other small nuclear facilities. While not directed specifically at large nuclear installations, it is likely that much of the information presented will also be of interest to those involved in the decommissioning of such facilities. Current views, information and experience on the planning and management of decommissioning projects in Member States were collected and assessed during a Technical Committee Meeting held by the IAEA in Vienna from 29 July to 2 August 1991. It was attended by 22 participants from 14 Member States and one international organization. 28 refs, 2 figs, 3 tabs

  3. Development of code system for management of reactor decommissioning (COSMARD), 1

    International Nuclear Information System (INIS)

    Yanagihara, Satoshi; Ogihara, Hirohito

    1994-02-01

    The Code System for Management of Reactor Decommissioning (COSMARD) was developed for use in the effective planning and management of reactor decommissioning. The decommissioning management data evaluation facility (DMAF) which is the main part of COSMARD has functions to evaluate various project management data such as manpower needs, radiation exposure of workers, amount of waste arisings necessary for each activity in a project using input data and calculation models consisting of simple arithmetic formulas and unit factors in the database. Using a set of command descriptors developed in COSMARD, work conditions and procedures for decommissioning a nuclear facility are describes as input data. The management data are evaluated by adopting the calculation models, which are placed in the activities at the lowest level of the work breakdown structure (WBS). The management data evaluated by the models are summed up in the ascending direction of WBS to obtain necessary data for the activities at any levels of WBS. In addition, scheduling calculations are conducted to obtain scheduling bar chart and histograms of the management data, on the basis of the work precedence conditions attached at certain activities. This report describes the outline of DMAF and user's manual of the sets of command descriptors. (author)

  4. Lesson Learned in Preparation for Decommissioning of Three Canadian Prototype Power Reactors

    International Nuclear Information System (INIS)

    Vickerd, Meggan; Kenny, Stephen

    2016-01-01

    Lesson learned by Canadian Nuclear Laboratories (CNL)(former AECL) in preparation for decommissioning of three Prototype Reactors is a result of various strategies used for each site. CNL is responsible for the eventual decommissioning of three prototype power reactors; Nuclear Power Demonstration (NPD), Gentilly-1 and Douglas Point. Each of the Canadian prototype power reactor sites shutdown using different strategies. Depending on the site location, configuration, and intended designation of the respective sites, the individual facility systems (ventilation, electrical system, fire detection etc.) were also shut down using different strategies and operating objectives. As CNL embarks on decommissioning the first Canadian prototype reactor, this paper will reflect on the lessons learned over the past thirty years and what CNL is adjusting in the decommissioning strategy to prepare better plans for the future. The Nuclear Power Demonstration Nuclear Generating Station (NPDNGS) was constructed in late 1950's and operated from 1962 to 1987 when it was permanently shutdown after exceeding its operational goals. The NPD reactor was the first Canadian nuclear power reactor and it consisted of a single 20 MWe pressurized heavy water reactor located on a single facility site in Rolphton, Ontario. The NPD facility was shutdown to a 'Cold, Dark and Quiet' state and is maintained using an unmanned strategy by managing the site remotely with active fire detection and security surveillance systems, minimal electrical supply and an active ventilation system which is operated periodically to allow for intermittent inspections. The Douglas Point Nuclear Generating Station (DPNGS) was constructed in the early 1960's and operated from 1968 to 1984 when it was permanently shutdown. It consisted of a 200 MW prototype Canada Deuterium Uranium (CANDU) reactor and is embedded on the Bruce Power site near Kincardine, Ontario. The Douglas Point site is maintained in a

  5. Decommissioning an Active Historical Reactor Facility at the Savannah River Site - 13453

    Energy Technology Data Exchange (ETDEWEB)

    Bergren, Christopher L.; Long, J. Tony; Blankenship, John K. [Savannah River Nuclear Solutions, LLC, Bldg. 730-4B, Aiken, SC 29808 (United States); Adams, Karen M. [United States Department of Energy, Bldg. 730-B, Aiken, SC 29808 (United States)

    2013-07-01

    The Savannah River Site (SRS) is an 802 square-kilometer United States Department of Energy (US DOE) nuclear facility located along the Savannah River near Aiken, South Carolina, where Management and Operations are performed by Savannah River Nuclear Solutions (SRNS). In 2004, DOE recognized SRS as structure within the Cold War Historic District of national, state and local significance composed of the first generation of facilities constructed and operated from 1950 through 1989 to produce plutonium and tritium for our nation's defense. DOE agreed to manage the SRS 105-C Reactor Facility as a potentially historic property due to its significance in supporting the U.S. Cold War Mission and for potential for future interpretation. This reactor has five primary areas within it, including a Disassembly Basin (DB) that received irradiated materials from the reactor, cooled them and prepared the components for loading and transport to a Separation Canyon for processing. The 6,317 square meter area was divided into numerous work/storage areas. The walls between the individual basin compartments have narrow vertical openings called 'slots' that permit the transfer of material from one section to another. Data indicated there was over 830 curies of radioactivity associated with the basin sediments and approximately 9.1 M liters of contaminated water, not including a large quantity of activated reactor equipment, scrap metal, and debris on the basin floor. The need for an action was identified in 2010 to reduce risks to personnel in the facility and to eliminate the possible release of contaminants into the environment. The release of DB water could potentially migrate to the aquifer and contaminate groundwater. DOE, its regulators [U. S. Environmental Protection Agency (USEPA)-Region 4 and the South Carolina Department of Health and Environmental Control (SCDHEC)] and the SC Historical Preservation Office (SHPO) agreed/concurred to perform a non

  6. Decommissioning an Active Historical Reactor Facility at the Savannah River Site - 13453

    International Nuclear Information System (INIS)

    Bergren, Christopher L.; Long, J. Tony; Blankenship, John K.; Adams, Karen M.

    2013-01-01

    The Savannah River Site (SRS) is an 802 square-kilometer United States Department of Energy (US DOE) nuclear facility located along the Savannah River near Aiken, South Carolina, where Management and Operations are performed by Savannah River Nuclear Solutions (SRNS). In 2004, DOE recognized SRS as structure within the Cold War Historic District of national, state and local significance composed of the first generation of facilities constructed and operated from 1950 through 1989 to produce plutonium and tritium for our nation's defense. DOE agreed to manage the SRS 105-C Reactor Facility as a potentially historic property due to its significance in supporting the U.S. Cold War Mission and for potential for future interpretation. This reactor has five primary areas within it, including a Disassembly Basin (DB) that received irradiated materials from the reactor, cooled them and prepared the components for loading and transport to a Separation Canyon for processing. The 6,317 square meter area was divided into numerous work/storage areas. The walls between the individual basin compartments have narrow vertical openings called 'slots' that permit the transfer of material from one section to another. Data indicated there was over 830 curies of radioactivity associated with the basin sediments and approximately 9.1 M liters of contaminated water, not including a large quantity of activated reactor equipment, scrap metal, and debris on the basin floor. The need for an action was identified in 2010 to reduce risks to personnel in the facility and to eliminate the possible release of contaminants into the environment. The release of DB water could potentially migrate to the aquifer and contaminate groundwater. DOE, its regulators [U. S. Environmental Protection Agency (USEPA)-Region 4 and the South Carolina Department of Health and Environmental Control (SCDHEC)] and the SC Historical Preservation Office (SHPO) agreed/concurred to perform a non-time critical removal

  7. Beneficial Re-Use of Metal from Decommissioning of Power Reactors

    International Nuclear Information System (INIS)

    Eshleman, Troy; Raw, Graham; Moloney, Barry

    2014-01-01

    Utilities and contractors decommissioning nuclear power reactors can recycle a high proportion of the scrap metal generated during dismantling either by free release for general re-use directly from the point of generation, or by recycling off-site at facilities specifically licensed for radioactive material. The worldwide commercial vendors operate different commercial models of volumetric decontamination of ferrous metals by thermal treatment. Some aim to achieve free release of output metals for general use, while others accept higher activity metals as feedstock for the manufacture of steel products which contain residual radioactivity, which we term 'Beneficial Re-use'. It is estimated that 10-30% of metals from light water reactor decommissioning have been exposed to neutron radiation (activated) and/or are contaminated to such an extent that free release is not achievable. This paper outlines a cost-effective alternative to managed storage or disposal for lightly activated or contaminated metal, utilising a 'Beneficial Re-Use' programme which has been in routine operation in the United States for over 20 years. 'Beneficial Re-Use' describes the manufacture of products such as radiation shielding from radioactive scrap metal. Unlike recycling practised in Europe, such products remain under control in licensed facilities and the metal does not find its way into general circulation or consumer products. Since 1992, EnergySolutions and its predecessor Duratek has been melting scrap at their Bear Creek, Tennessee facility to produce shield blocks for use in high energy research facilities. Over 62,300 t of scrap steel have been re-used, and the demand for shielding products continues long into the future. 3,000 t of this feedstock originated outside the US. This paper proposes the potential for activated steel that will not be acceptable at European recycling facilities to enter the Beneficial Re-use programme. Acceptance criteria

  8. Project WAGR: The UK demonstration project for power reactor decommissioning - removing the core and looking to completion

    International Nuclear Information System (INIS)

    Benest, T. G.

    2003-01-01

    The United Kingdom Atomic Energy Authority (UKAEA) has built and operated a wide range of nuclear facilities since the late 1940's. UKAEA's present mission is to restore the environment of these facilities in a safe and environmentally responsible manner. This restoration includes the decommissioning of a number of redundant research and power reactors, one of which is the Windscale Advanced Gas-cooled Reactor (WAGR). Following shut down, UKAEA decided to continue the prototype function of the reactor into the decommissioning phase to develop dismantling techniques and establish waste routes. The reactor core and pressure vessel are now being dismantled in a programme of 10 campaigns, seven of which have been completed since 1998. It is anticipated that the current programme will be completed by summer 2005. This paper outlines the history of the reactor, the operation of the waste-processing route, the installed dismantling equipment and the successful completion of the first seven campaigns. This earlier work has been described in a number of publications and conferences, so this paper concentrates on recent work to select and develop cutting equipment to dismantle the core support structures and the pressure vessel. The decommissioning of the Windscale Advance Gas-cooled reactor is being undertaken to demonstrate that a power reactor can be decommissioned shortly after shutdown. The removal of the core and pressure vessel has been broken down into a series of 10 campaigns associated with particular core components. The first 7 campaigns have been successfully completed and the 8., is expected to commence in September 2003 17 months earlier than planned. Dismantling methodologies and tools have been developed specifically for each of these campaigns. Full-scale mock-ups have been used to test the tools, train the operators and assess the duration of operations. However, despite successful trials, operational experience has shown that some of these tools have not

  9. Decommissioning of the secondary containment of the steam generating heavy water reactor at UKAEA-Winfrith

    International Nuclear Information System (INIS)

    Miller, Keith; Cornell, Rowland; Parkinson, Steve; McIntyre, Kevin; Staples, Andy

    2007-01-01

    Available in abstract form only. Full text of publication follows: The Winfrith SGHWR was a prototype nuclear power plant operated for 23 years by the United Kingdom Atomic Energy Authority (UKAEA) until 1990 when it was shut down permanently. The current Stage 1 decommissioning contract is part of a multi-stage strategy. It involves the removal of all the ancillary plant and equipment in the secondary containment and non-containment areas ahead of a series of contracts for the decommissioning of the primary containment, the reactor core and demolition of the building and all remaining facilities. As an outcome of a competitive tending process, the Stage 1 decommissioning contract was awarded to NUKEM with operations commencing in April 2005. The decommissioning processes involved with these plant items will be described with some emphasis of the establishment of multiple work-fronts for the production, recovery, treatment and disposal of mainly tritium-contaminated waste arising from its contact with the direct cycle reactor coolant. The means of size reduction of a variety of large, heavy and complex items of plant made from a range of materials will also be described with some emphasis on the control of fumes during hot cutting operations and establishing effective containments within a larger secondary containment structure. Disposal of these wastes in a timely and cost-effective manner is a major challenge facing the decommissioning team and has required the development of a highly efficient means of packing the resultant materials into mainly one-third height ISO containers for disposal as LLW. Details of the quantities of LLW and exempt wastes handled during this process will be given with a commentary about the difficulty in segregating these two waste streams efficiently. (authors)

  10. Lubrication of nuclear reactor components

    International Nuclear Information System (INIS)

    Wild, E.; Mack, K.J.

    1978-01-01

    Safe operation of liquid metal cooled nuclear reactors requires a knowledge of the tribological behaviour of contacting components at high temperatures with slow relative movement at high frictional loads in a chemically aggressive environment. Experiments have been performed on various material combinations in liquid sodium and argon. Because of the small sliding movements, hydrodynamic lubrication is not expected and thus surface finish is an important factor. Tests have been performed on brushed, ground and lapped surfaces. Among the material combinations tested a CrC-coating on a 1.4961 stainless steel substrate performed well. Friction coefficients of 0.35-0.5 in argon and 0.1-1.2 in liquid sodium were recorded. (author)

  11. Progress of the decommissioning process of Musashi Institute of Technology reactor (4)

    International Nuclear Information System (INIS)

    Uchiyama, Takafumi; Tanzawa, Tomio; Mitsuhashi, Ishi; Morishima, Kayoko; Matsumoto, Tetsuo

    2012-01-01

    The research reactor of Tokyo City University Atomic Energy Research Laboratory (Musashi Institute of Technology reactor) is zirconium-moderated water-cooled solid homogeneous type (TRIGA-II type), and its maximum heat output is 100 kW. It got into the first critical state in January 1963, and since then, it has mainly contributed to education and training for upgrading nuclear engineers, radioactivation analysis and reactor physics, and medical researches, as the joint usage research facilities across Japan. Then, after a long-term suspension, the university submitted the file in 2004 to the Ministry of Education, Culture, Sports, Science and Technology on the dismantling for the purpose of facility abolishment. Through the procedure of submitting a decommissioning plan, it was approved. Furthermore, in order to perform the function stop of the disposal facilities of liquid waste, application for change authorization for the decommissioning plan was submitted and approved. Regarding the progress of the decommissioning plan, the dismantling and removal of waste facilities for liquid waste and solid waste was carried out in FY2011 without any trouble. This paper explains this progress and future work plans. (A.O.)

  12. Technology, safety, and costs of decommissioning a reference pressurized water reactor power station

    International Nuclear Information System (INIS)

    Smith, R.I.; Konzek, G.J.; Kennedy, W.E. Jr.

    1978-05-01

    Safety and cost information was developed for the conceptual decommissioning of a large [1175 MW(e)] pressurized water reactor (PWR) power station. Two approaches to decommissioning, Immediate Dismantlement and Safe Storage with Deferred Dismantlement, were studied to obtain comparisons between costs, occupational radiation doses, potential radiation dose to the public, and other safety impacts. Immediate Dismantlement was estimated to require about six years to complete, including two years of planning and preparation prior to final reactor shutdown, at a cost of $42 million, and accumulated occupational radiation dose, excluding transport operations, of about 1200 man-rem. Preparations for Safe Storage were estimated to require about three years to complete, including 1 1 / 2 years for planning and preparation prior to final reactor shutdown, at a cost of $13 million and an accumulated occupational radiation dose of about 420 man-rem. The cost of continuing care during the Safe Storage period was estimated to be about $80 thousand annually. Accumulated occupational radiation dose during the Safe Storage period was estimated to range from about 10 man-rem for the first 10 years to about 14 man-rem after 30 years or more. The cost of decommissioning by Safe Storage with Deferred Dismantlement was estimated to be slightly higher than Immediate Dismantlement. Cost reductions resulting from reduced volumes of radioactive material for disposal, due to the decay of the radioactive containments during the deferment period, are offset by the accumulated costs of surveillance and maintenance during the Safe Storage period

  13. The choice of cement for the manufacture of concrete to be activated: the potential for reducing the radiological consequences of reactor decommissioning

    International Nuclear Information System (INIS)

    Woollam, P.B.

    1985-05-01

    This report presents trace element analyses of some candidate cements which might be used in the manufacture of Sizewell 'B' concrete. It completes a programme of work whose aim was to investigate the potential for reducing the radiological consequences of reactor decommissioning through selection of construction materials for activated components. In particular, consideration has been given to the potential for reducing the concentration of elements known to activate to long lived daughters. (U.K.)

  14. Gamma dose rate estimation and operation management suggestions for decommissioning the reactor pressure vessel of HTR-PM

    Energy Technology Data Exchange (ETDEWEB)

    Sheng Fang; Hong Li; Jianzhu Cao; Wenqian Li; Feng Xie; Jiejuan Tong [Institute of Nuclear and New Energy Technology, Tsinghua, University, Beijing (China)

    2013-07-01

    China is now designing and constructing a high temperature gas cooled reactor-pebble bed module (HTR-PM). In order to investigate the future decommissioning approach and evaluate possible radiation dose, gamma dose rate near the reactor pressure vessel was calculated for different cooling durations using QAD-CGA program. The source term of this calculation was provided by KORIGEN program. Based on the calculated results, the spatial distribution and temporal changes of gamma dose rate near reactor pressure vessel was systematically analyzed. A suggestion on planning decommissioning operation of reactor pressure vessel of HTRPM was given based on calculated dose rate and the Chinese Standard GB18871-2002. (authors)

  15. Studies on decommissioning of TRIGA reactors and site restoration technologies in the Republic of Korea

    International Nuclear Information System (INIS)

    Oh, Won-Zin; Kim, Gye-Nam; Won, Hui-Jun

    2002-01-01

    Research and development on research reactor decommissioning and environmental restoration has been carried out at KAERI since 1997 to prepare for the decommissioning of KAERI's two TRIGA-type research reactors, which had been shut down since 1995. A 3-D graphic model of the TRIGA research reactor was built using IGRIP. The dismantling process was simulated in the graphic environment to verify the feasibility of individual operations before the execution of the remote dismantling process. An under-water wall-climbing robot, moving by propeller injection, and identifying its coordinates by using a laser sensor, was developed and tested in the TRIGA reactor pool by measuring a radioactive contamination map of the reactor surface. Using MODFLOW and TRIGA site geological data, a computer simulation of the underground migration of residual radionuclides, after the TRIGA reactor decommissioning, was carried out. It was found that the underground migration rate was very slow such that, when radionuclide decay and dilution are considered, the residual radionuclides will not have a significant environmental impact. The soil decontamination R and D, using soil washing, solvent flushing and electro-decontamination technologies, was carried out to determine the best method for decontaminating the soil waste accumulated in KAERI. The decontamination results indicated that, using the soil washing method, more than 80% of the soil wastes could be decontaminated well enough to discharge them to the environment. It was also determined that the control of solution pH and temperature in the soil washing process is important for the reduction of decontamination waste. Further decontamination, using an electro-kinetic decontamination method, was considered necessary for the residual soil waste, which consisted mainly of fine soil particles. (author)

  16. A study on source term assessment and waste disposal requirement of decontamination and decommissioning for the TRIGA research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Whang, Joo Ho; Lee, Kyung JIn; Lee, Jae Min; Choi, Gyu Seup; Shin, Byoung Sun [Kyunghee Univ., Seoul (Korea, Republic of)

    1999-08-15

    The objective and necessity of the project : TRIGA is the first nuclear facility that decide to decommission and decontamination in our nation. As we estimate the expected life of nuclear power generation at 30 or 40 years, the decommissioning business should be conducted around 2010, and the development of regulatory technique supporting it should be developed previously. From a view of decommissioning and decontamination, the research reactor is just small in scale but it include all decommissioning and decontamination conditions. So, the rules by regulatory authority with decommissioning will be a guide for nuclear power plant in the future. The basis of regulatory technique required when decommissioning the research reactor are the radiological safety security and the data for it. The source term is very important condition not only for security of worker but for evaluating how we dispose the waste is appropriate for conducting the middle store and the procedure after it when the final disposal is considered. The content and the scope in this report contain the procedure of conducting the assessment of the source term which is most important in understanding the general concept of the decommissioning procedure of the decommissioning and decontamination of TRIGA research reactor. That is, the sampling and measuring method is presented as how to measure the volume of the radioactivity of the nuclear facilities. And also, the criterion of classifying the waste occurred in other countries and the site release criteria which is the final step of decommissioning and decontamination presented through MARSSIM. Finally, the program to be applicable through comparing the methods of our nation and other countries ones is presented as plan for disposal of the waste in the decommissioning.

  17. SAVANNAH RIVER SITE R-REACTOR DISASSEMBLY BASIN IN-SITU DECOMMISSIONING -10499

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C.; Serrato, M.; Blankenship, J.; Griffin, W.

    2010-01-04

    The US DOE concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate intact, structurally sound facilities that are no longer needed for their original purpose, i.e., generating (reactor facilities), processing(isotope separation facilities) or storing radioactive materials. The 105-R Disassembly Basin is the first SRS reactor facility to undergo the in-situ decommissioning (ISD) process. This ISD process complies with the 105-R Disassembly Basin project strategy as outlined in the Engineering Evaluation/Cost Analysis for the Grouting of the R-Reactor Disassembly Basin at the Savannah River Site and includes: (1) Managing residual water by solidification in-place or evaporation at another facility; (2) Filling the below grade portion of the basin with cementitious materials to physically stabilize the basin and prevent collapse of the final cap - Sludge and debris in the bottom few feet of the basin will be encapsulated between the basin floor and overlying fill material to isolate it from the environment; (3) Demolishing the above grade portion of the structure and relocating the resulting debris to another location or disposing of the debris in-place; and (4) Capping the basin area with a concrete slab which is part of an engineered cap to prevent inadvertent intrusion. The estimated total grout volume to fill the 105-R Reactor Disassembly Basin is 24,384 cubic meters or 31,894 cubic yards. Portland cement-based structural fill materials were designed and tested for the reactor ISD project, and a placement strategy for stabilizing the basin was developed. Based on structural engineering analyses and material flow considerations, maximum lift heights and differential height requirements were determined. Pertinent data and information related to the SRS 105-R Reactor Disassembly Basin in-situ decommissioning include: regulatory documentation, residual water management, area preparation activities, technology needs, fill material

  18. Use of MCNP for characterization of reactor vessel internals waste from decommissioned nuclear reactors

    International Nuclear Information System (INIS)

    Love, E.F.; Pauley, K.A.; Reid, B.D.

    1995-09-01

    This study describes the use of the Monte Carlo Neutron-Photon (MCNP) code for determining activation levels of irradiated reactor vessel internals hardware. The purpose of the analysis is to produce data for the Department of Energy's Greater-Than-Class C Low-Level Radioactive Waste Program. An MCNP model was developed to analyze the Yankee Rowe reactor facility. The model incorporates reactor geometry, material compositions, and operating history data acquired from Yankee Atomic Electric Company. In addition to the base activation analysis, parametric studies were performed to determine the sensitivity of activation to specific parameters. A component sampling plan was also developed to validate the model results, although the plan was not implemented. The calculations for the Yankee Rowe reactor predict that only the core baffle and the core support plates will be activated to levels above the Class C limits. The parametric calculations show, however, that the large uncertainties in the material compositions could cause errors in the estimates that could also increase the estimated activation level of the core barrel to above the Class C limits. Extrapolation of the results to other reactor facilities indicates that in addition to the baffle and support plates, core barrels may also be activated to above Class C limits; however the classification will depend on the specific operating conditions of the reactor and the specific material compositions of the metal, as well as the use of allowable concentration averaging practices in packaging and classifying the waste

  19. Component failure data base of TRIGA reactors

    International Nuclear Information System (INIS)

    Djuricic, M.

    2004-10-01

    This compilation provides failure data such as first criticality, component type description (reactor component, population, cumulative calendar time, cumulative operating time, demands, failure mode, failures, failure rate, failure probability) and specific information on each type of component of TRIGA Mark-II reactors in Austria, Bangladesh, Germany, Finland, Indonesia, Italy, Indonesia, Slovenia and Romania. (nevyjel)

  20. Nuclear decommissioning

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    The paper on nuclear decommissioning was presented by Dr H. Lawton to a meeting of the British Nuclear Energy Society and Institution of Nuclear Engineers, 1986. The decommissioning work currently being undertaken on the Windscale advanced gas cooled reactor (WAGR) is briefly described, along with projects in other countries, development work associated with the WAGR operation and costs. (U.K.)

  1. Experience relevant to safety obtained from reactor decommissioning operations in the French Atomic Energy Commission

    International Nuclear Information System (INIS)

    Giraudel, B.; Langlois, G.

    1979-01-01

    From among the nuclear facilities constructed in France the authors cite eight large reactors, ranging from critical assemblies to power reactors, that have been finally shut-down since 1965. A brief account is given of the way in which the various operations were carried out after the final control rod drop, a distinction being drawn between the shut-down proper and the containment and dismantling work. A description is also given, from the technical and regulatory standpoint, of the final stage attained, and mention is made of French safety arrangements and of the part played by the safety services during decommissioning operations. Among the lessons derived from French experience, the authors mention the completion of operations without any serious safety problems, and with guarantees for the protection of personnel and the population as a whole, by conventional techniques; the advantage of planning decommissioning operations from the very beginning of construction of the facilities; and the importance of filing descriptive documents. In view of the experience gained, the French Atomic Energy Commission has devised internal procedures for facilitating the application of regulations governing the shut-down and decommissioning phases, which are aimed at preserving surveillance procedures similar to those in force during normal operation. (author)

  2. The radioactive inventory of a decommissioned magnox power station structure. 1. Measurements of neutron induced activity in samples from the reactor island

    International Nuclear Information System (INIS)

    Woollam, P.B.

    1978-09-01

    This report is the first of a series which, together, aim to produce an accurate assessment of neutron induced activation levels in the fixed structural components of a reactor of the steel pressure vessel Magnox type. It describes the measurements made of induced activation, necessary in order to establish credibility in the complex calculations described in the subsequent reports. The report also attempts systematically to assess the potential contributions to the dose and disposal problem from all isotopes with a half-life in excess of 5 years. This is necessary in order to ensure that no isotope has been overlooked which could limit any part of the plan for the decommissioning of a Magnox reactor. In addition the report aims to determine concentrations, in each major material type, of trace elements which lead to the isotopes limiting in decommissioning. (author)

  3. Decommissioning techniques for research reactors. Final report of a co-ordinated research project 1997-2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-02-01

    In its international role, the IAEA is faced with a wide variety of national situations and different availability of technical, human and financial resources. While it is recognised that nuclear decommissioning is a mature industry in some developed countries, and may soon become a routine activity, the situation is by no means so clear in other countries. In addition, transfer of technologies and know-how from developed to developing countries is not a spontaneous, straightforward process, and will take time and considerable effort. As mandated by its own statute and Member States' requests, the IAEA continues to respond to its Member States by monitoring technological progress, ensuring development of safer and more efficient strategies and fostering international information exchange. Previous co-ordinated research projects (CRP) conducted respectively from 1984 to 1987, and from 1989 to 1993, investigated the overall domain of decommissioning. In those CRPs no distinction was made between decommissioning activities carried out at nuclear power plants, research reactors or nuclear fuel cycle facilities. With technological progress and experience gained, it became clear that decommissioning of research reactors had certain specific characteristics which needed a dedicated approach. In addition, a large number of research reactors reached a state of permanent shutdown in the 1990s and were candidates for prompt decommissioning. With the progressive ageing of research reactors, many more of these units will soon become redundant worldwide and require decommissioning. Within this context, a CRP on Decommissioning Techniques for Research Reactors was launched and conducted by the IAEA from 1997 to 2001 in order to prepare for eventual decommissioning. Concluding reports that summarized the work undertaken under the aegis of the CRP were presented at the third and final Research Co-ordination Meeting held in Kendal, United Kingdom, 14-18 May 2001, and are collected

  4. Decommissioning techniques for research reactors. Final report of a co-ordinated research project 1997-2001

    International Nuclear Information System (INIS)

    2002-02-01

    In its international role, the IAEA is faced with a wide variety of national situations and different availability of technical, human and financial resources. While it is recognised that nuclear decommissioning is a mature industry in some developed countries, and may soon become a routine activity, the situation is by no means so clear in other countries. In addition, transfer of technologies and know-how from developed to developing countries is not a spontaneous, straightforward process, and will take time and considerable effort. As mandated by its own statute and Member States' requests, the IAEA continues to respond to its Member States by monitoring technological progress, ensuring development of safer and more efficient strategies and fostering international information exchange. Previous co-ordinated research projects (CRP) conducted respectively from 1984 to 1987, and from 1989 to 1993, investigated the overall domain of decommissioning. In those CRPs no distinction was made between decommissioning activities carried out at nuclear power plants, research reactors or nuclear fuel cycle facilities. With technological progress and experience gained, it became clear that decommissioning of research reactors had certain specific characteristics which needed a dedicated approach. In addition, a large number of research reactors reached a state of permanent shutdown in the 1990s and were candidates for prompt decommissioning. With the progressive ageing of research reactors, many more of these units will soon become redundant worldwide and require decommissioning. Within this context, a CRP on Decommissioning Techniques for Research Reactors was launched and conducted by the IAEA from 1997 to 2001 in order to prepare for eventual decommissioning. Concluding reports that summarized the work undertaken under the aegis of the CRP were presented at the third and final Research Co-ordination Meeting held in Kendal, United Kingdom, 14-18 May 2001, and are collected

  5. 2016 Annual Inspection and Radiological Survey Results for the Piqua, Ohio, Decommissioned Reactor Site, July 2016

    Energy Technology Data Exchange (ETDEWEB)

    Zimmerman, Brian [USDOE Office of Legacy Management, Washington, DC (United States); Miller, Michele [Navarro Research and Engineering, Oak Ridge, TN (United States)

    2016-07-01

    This report presents the findings of the annual inspection and radiological survey of the Piqua, Ohio, Decommissioned Reactor Site (site). The decommissioned nuclear power demonstration facility was inspected and surveyed on April 15, 2016. The site, located on the east bank of the Great Miami River in Piqua, Ohio, was in fair physical condition. There is no requirement for a follow-up inspection, partly because City of Piqua (City) personnel participated in a March 2016 meeting to address reoccurring safety concerns. Radiological survey results from 104 locations revealed no removable contamination. One direct beta activity reading in a floor drain on the 56-foot level (1674 disintegrations per minute [dpm]/100 square centimeters [cm2]) exceeded the minimum detectable activity (MDA). Beta activity has been detected in the past at this floor drain. The reading was well below the action level of 5000 dpm/100 cm2.

  6. Decontamination and decommissioning of the MTR [Materials Testing Reactor]-603 HB-2 cubicle

    International Nuclear Information System (INIS)

    Smith, D.L.

    1987-10-01

    This paper describes the decontamination and decommissioning (D and D) of the MTR-603 HB-2 cubicle located at the Idaho National Engineering Laboratory (INEL). The HB-2 cubicle became radioactively contaminated during out-of-pile circulating water loop experiments conducted in the Materials Testing Reactor in the 1950s and 1960s. This paper describes work performed to accomplish the D and D objectives of reducing the high radiation fields caused by contamination inside the cubicle, preventing future contamination spread, and making about 1400 ft 2 of floor space available for reuse. Decommissioning of the HB-2 cubicle consisted of total dismantlement of the cubicle and its contents and was performed without disrupting ongoing laboratory work being conducted in areas surrounding the HB-2 cubicle. 3 refs., 7 figs., 4 tabs

  7. What will we do with the low level waste from reactor decommissioning?

    International Nuclear Information System (INIS)

    Meehan, A. R.; Wilmott, S.; Crockett, G.; Watt, N. R.

    2008-01-01

    The decommissioning of the UK's Magnox reactor sites will produce large volumes of low level waste (LLW) arisings. The vast majority of this waste takes the form of concrete, building rubble and redundant plant containing relatively low levels of radioactivity. Magnox Electric Ltd (Magnox) is leading a strategic initiative funded by the Nuclear Decommissioning Authority (NDA) to explore opportunities for the disposal of such waste to suitably engineered facilities that might be located on or adjacent to the site of waste arising, if appropriate and subject to regulatory acceptance and stakeholder views. The strategic issues surrounding this initiative are described along with an update of progress with stakeholder consultations in relation to the proposed licensing of the first such facility at Hinkley Point A, which could be viewed as a test case for the development of similar disposal facilities at other nuclear sites in England and Wales. (authors)

  8. Decommissioning of eight surplus production reactors at the Hanford Site, Richland, Washington

    International Nuclear Information System (INIS)

    1992-12-01

    The first section of this volume summarizes the content of the draft environmental impact statement (DEIS) and this Addendum, which together constitute the final environmental impact statement (FEIS) prepared on the decommissioning of eight surplus plutonium production reactors at Hanford. The FEIS consists of two volumes. The first volume is the DEIS as written. The second volume (this Addendum) consists of a summary; Chapter 9, which contains comments on the DEIS and provides DOE's responses to the comments; Appendix F, which provides additional health effects information; Appendix K, which contains costs of decommissioning in 1990 dollars; Appendix L, which contains additional graphite leaching data; Appendix M, which contains a discussion of accident scenarios; Appendix N, which contains errata; and Appendix 0, which contains reproductions of the letters, transcripts, and exhibits that constitute the record for the public comment period

  9. Preparation for Future Defuelling and Decommissioning Works on EDF Energy's UK Fleet of Advanced Gas Cooled Reactors

    International Nuclear Information System (INIS)

    Bryers, John; Ashmead, Simon

    2016-01-01

    EDF Energy/Nuclear Generation is the owner and operator of 14 Advanced Gas cooled Reactors (AGR) and one Pressurised Water Reactor (PWR), on 8 nuclear stations in the UK. EDF Energy/Nuclear Generation is responsible for all the activities associated with the end of life of its nuclear installations: de-fuelling, decommissioning and waste management. As the first AGR is forecast to cease generation within 10 years, EDF Energy has started planning for the decommissioning. This paper covers: - broad outline of the technical strategy and arrangements for future de-fuelling and decommissioning works on the UK AGR fleet, - high level strategic drivers and alignment with wider UK nuclear policy, - overall programme of preparation and initial works, - technical approaches to be adopted during decommissioning. (authors)

  10. A method for evaluation the activity of the reactor components

    International Nuclear Information System (INIS)

    Gugiu, E.D.; Roth, Cs.

    2003-01-01

    The ability to predict the radioactivity levels of the reactor components is an important aspect from waste management point of view, as well as from radioprotection purposes. A special case is represented by the research reactors where, one of the major contributions to the radioactivity inventory is due to the experimental devices involved in various research works during reactor life. Generally, aluminum and aluminum alloys are used in manufacturing these devices; as a result, the work presented in this paper is focused on the qualitative and quantitative analysis of the radioactive isotopes contained in these materials. A device used for silicon doping by neutron transmutation that was placed near TRIGA reactor core is investigated. The isotopic content of various samplings drawn from various points of the device was analyzed by gamma spectrometry using a HPGe detector. Computations, using the MCNP5 code, are also performed in order to evaluate the reaction rates for all the isotopes and their reactions. The Monte Carlo simulations are performed for a detailed geometry and material composition of the reactor core and the device. The Origen-S code is also used in order to evaluate the isotopic inventory and the activity values. A detailed analysis regarding the possibility to estimate by computations and/or by gamma spectrometry the activity values of the isotopes which are of interest for decommissioning is presented in the paper. (authors)

  11. Certification of the decommissioning project for the PROTEUS research reactor at the Paul Scherrer Institute; Gutachten zum Stilllegungsprojekt der Kernanlage PROTEUS am Paul Scherrer Institut

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2015-12-15

    The Paul Scherrer Institute (PSI) revised the documents concerning the decommissioning of the PROTEUS research reactor. This report presents the results of the evaluation by the Swiss Federal Nuclear Safety Inspectorate (ENSI). PSI considered all relevant stipulations of nuclear energy legislation, the law on radiation protection, as well as ENSI directives. Moreover, ENSI made sure that the PROTEUS decommissioning project corresponds to the IAEA, WENRA and OECD/NEA international requirements, and corresponds to current state of science and technology. ENSI ascertained some facts that have to be looked at more deeply. Before beginning with the decommissioning work, all the fuel must be taken out of the PROTEUS facility. For each step an authorization has to be requested from ENSI with a detailed description of the work foreseen. Personal dosimetry has to be performed with calibrated dosimeters. By the use of mechanical, thermal or chemical methods to partition radioactive components, the air on the working place has to be continuously checked for radioactive aerosols. The dose limit of 0.3 mSv per year must be respected. The surveillance of the release of radioactive materials has to be done according to the PSI release regulations. By large material quantities like barite concrete blocks, graphite reflector and steel components, PSI has to describe the process used to declare the materials as inactive. For the radioactive materials expected, the containers specified have to be approved by ENSI. Before the first dismantling phase, the organization plan for all participating persons and their responsibilities have to be presented to ENSI. In its request for the decommissioning of the PROTEUS research reactor, PSI consistently demonstrates that the protection of persons and environment against radioactive radiation can be guaranteed during the dismantling of the facility and that the wastes produced can be safely managed. In consequence, all required conditions for

  12. Decommissioning of the High Flux Beam Reactor at Brookhaven National Laboratory.

    Science.gov (United States)

    Hu, Jih-Perng; Reciniello, Richard N; Holden, Norman E

    2012-08-01

    The High Flux Beam Reactor (HFBR) at the Brookhaven National Laboratory was a heavy-water cooled and moderated reactor that achieved criticality on 31 October 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shut down in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor's spent fuel pool. The reactor remained shut down for almost 3 y for safety and environmental reviews. In November 1999, the United States Department of Energy decided to permanently shut down the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome, which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR's cleanup performed during 1999-2009, to allow the BNL facilities to be re-accessed by the public, will be described in the paper.

  13. Project plan for the decontamination and decommissioning of the Argonne National Laboratory Experimental Boiling Water Reactor

    International Nuclear Information System (INIS)

    Boing, L.E.

    1989-12-01

    In 1956, the Experimental Boiling Water Reactor (EBWR) Facility was first operated at Argonne National Laboratory (ANL) as a test reactor to demonstrate the feasibility of operating an integrated power plant using a direct cycle boiling water reactor as a heat source. In 1967, ANL permanently shut down the EBWR and placed it in dry lay-up. This project plan presents the schedule and organization for the decontamination and decommissioning of the EBWR Facility which will allow it to be reused by other ANL scientific research programs. The project total estimated cost is $14.3M and is projected to generate 22,000 cubic feet of low-level radioactive waste which will be disposed of at an approved DOE burial ground. 18 figs., 3 tabs

  14. Radiological survey support activities for the decommissioning of the Ames Laboratory Research Reactor Facility, Ames, Iowa

    Energy Technology Data Exchange (ETDEWEB)

    Wynveen, R.A.; Smith, W.H.; Sholeen, C.M.; Justus, A.L.; Flynn, K.F.

    1984-09-01

    At the request of the Engineering Support Division of the US Department of Energy-Chicago Operations Office and in accordance with the programmatic overview/certification responsibilities of the Department of Energy Environmental and Safety Engineering Division, the Argonne National Laboratory Radiological Survey Group conducted a series of radiological measurements and tests at the Ames Laboratory Research Reactor located in Ames, Iowa. These measurements and tests were conducted during 1980 and 1981 while the reactor building was being decontaminated and decommissioned for the purpose of returning the building to general use. The results of these evaluations are included in this report. Although the surface contamination within the reactor building could presumably be reduced to negligible levels, the potential for airborne contamination from tritiated water vapor remains. This vapor emmanates from contamination within the concrete of the building and should be monitored until such time as it is reduced to background levels. 2 references, 8 figures, 6 tables.

  15. Project plan for the decontamination and decommissioning of the Argonne National Laboratory Experimental Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L.E.

    1989-12-01

    In 1956, the Experimental Boiling Water Reactor (EBWR) Facility was first operated at Argonne National Laboratory (ANL) as a test reactor to demonstrate the feasibility of operating an integrated power plant using a direct cycle boiling water reactor as a heat source. In 1967, ANL permanently shut down the EBWR and placed it in dry lay-up. This project plan presents the schedule and organization for the decontamination and decommissioning of the EBWR Facility which will allow it to be reused by other ANL scientific research programs. The project total estimated cost is $14.3M and is projected to generate 22,000 cubic feet of low-level radioactive waste which will be disposed of at an approved DOE burial ground. 18 figs., 3 tabs.

  16. Decontamination and decommissioning project of the TRIGA Mark-2 and 3 research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jung, K J; Baik, S T; Chung, U S; Jung, K H; Park, S K; Lee, B J; Kim, J K; Yang, S H

    2000-01-01

    During the review on the decommissioning plan and environmental impact assessment report by the KINS, the number of the inquired items were two hundred and fifty one, and the answers were made and sent until September 10, 1999, as the screened review results were reported to Ministry of Science and Technology(MOST) in December 14, 1999, all the reviews on the licence were over. Radioactive liquid wastes of 400 tons generated during the operation of the research reactors including reactor vessels are stored in the facility of the research reactor 1 and 2. Those liquid wastes have the low-level-radioactivity which can be discharged to the surroundings, but was wholly treated to be vaporized naturally by means of the increased numbers of the natural vaporization disposal facilities with the annual capacity of 200 tons for the purpose of the minimized environmental contamination.

  17. Preparatory activities of the Fugen decommissioning

    International Nuclear Information System (INIS)

    Iguchi, Y.; Tajiri, T.; Kiyota, S.

    2004-01-01

    The Advanced Thermal Reactor Fugen is a 165 MWe, heavy water moderated, light-water cooled, pressure-tube type reactor. In February 1998, the Atomic Energy Commission of Japan introduced a new policy that development and research of decommissioning of Fugen should be promoted in order to carry out the decommissioning smoothly after the shutdown. The Fugen reactor was shut down definitely in March 2003, and Fugen has been preparing for the project, including necessary development of technologies. The development of decommissioning for Fugen is divided into two areas. One area is the development of unique technology for dismantling special components such as the reactor core and the heavy water system. Another area is the improvement and enhancement of existing technologies. Especially the former area requires effort and comprises development of the reactor dismantlement, tritium decontamination of heavy water system and engineering support systems. The activities are as follows: the density and amount of radioactive nuclides in all equipment or concrete including the reactor core need to be evaluated for the decommissioning. To prepare for decommissioning, analysis, measurement and evaluation of the neutron flux density have been executed during reactor operation. Special dismantling process is necessary for the heavy water system and the reactor that are unique to Fugen. Some studies and tests are going on for the safe dismantling based on existing technologies and their combination. Systems engineering approach is necessary in order to optimize the work load, exposure dose, waste mass and cost by selecting appropriate dismantling process at the planning stage of the decommissioning. For this reason, in order to make a decommissioning plan efficiently, we have been developing an Engineering Support System for decommissioning by adopting new information technologies such as three-dimensional computer-aided design system and virtual reality system. Moreover, the

  18. Remote handling equipment for the decommissioning of the Windscale Advanced Gas Cooled Reactor

    International Nuclear Information System (INIS)

    Barker, A.; Birss, I.R.; Fish, G.

    1984-01-01

    A decision to decommission the Windscale Advanced Gas Cooled Reactor was taken shortly after reactor shutdown in 1981. The fuel has now been discharged and the decommissioning programme will last about 10-12 years. The paper describes the programme and objectives and deals with methods of handling and disposing of the radioactive waste material. The main new facility required is a Waste Packaging Building adjacent to the existing reactor in which the waste boxes will be filled, active waste encapsulated in concrete and the boxes cleaned, swabbed and monitored to comply with IAEA transport regulations. The handling machine concept and features are described. The assaying and packaging of the waste material, the control of box movement and the process of concrete encapsulation is described. The paper concludes with a description of the development programme to support the Project. The tasks include a study of cutting techniques, production and control of dust and smoke, viewing and lighting methods, filtration, decontamination and fixing of contamination

  19. Implementation of the long term stewardship model of decommissioning power reactors. Update on the Zion project

    Energy Technology Data Exchange (ETDEWEB)

    Christian, J.; Hess, J.; Moloney, B.P. [EnergySolutions EU, Swindon (United Kingdom)

    2012-11-01

    Several countries have announced programmes to phase out nuclear power. Many NPPs built in the 1960s-80s are in any case reaching the end of their planned operating lives. Over the next decade, approximately 60-80 reactors worldwide will reach end of useful life and become candidates for decontamination and dismantling (D and D). Utilities will therefore commission over the coming decade a much larger number of decommissioning programmes to discharge their license responsibilities for reactor dismantling and site remediation. One major strategic question for the utilities is whether they regard decommissioning reactors as part of their core business or whether they wish to transfer this burden and risk in part or in whole to a specialised contractor. This paper reviews progress at the first programme in the US where a non-utility company has taken on the full license responsibility from the utility to undertake site remediation and license termination. We call this model ''Long Term Stewardship'' and it is now fully underway at the Zion NPP near Chicago, Illinois. (orig.)

  20. Development and optimisation of generic decommissioning strategies for civil Magnox reactors

    International Nuclear Information System (INIS)

    Carpenter, G.; Hebditch, D.; Meek, N.; Patel, A.; Reeve, P.

    2004-01-01

    BNFL Environmental Services has formulated updated proposals for the use of decision analysis in the development of decommissioning strategy. The proposals are based on the Department of Transport, Local Government and the Regions manual for practitioners on multi-criteria analysis, specifically multi-criteria decision analysis, as suited to complex problems with a mixture of monetary and non-monetary objectives. They take account of up-to-date academic methodology, the newly issued BNFL decision analysis framework for environmental decisions and a wide variety of other engineering, optioneering and optimisation processes. The paper also summarises legislative and company policy areas of importance to decommissioning strategy development. Higher-level generic reactor and site remediation strategies already exist. At the lower level, various generic decommissioning reference processes and project options need development. For the past year, Environmental Services has held responsibility to respond to the Nuclear Installations Inspectorates' quinquennial review, develop and maintain up-to-date strategies, institute the review of a selected number of key strategies, and respond to changing circumstances including stakeholder views. Environmental Services is performing a range of generic studies for selection of strategies and end-points as used for a variety of waste management and site care and maintenance preparations. (author)

  1. Decommissioning of the ASTRA research reactor - planning, executing and summarizing the project

    International Nuclear Information System (INIS)

    Meyer, F.

    2010-01-01

    The decommissioning of the ASTRA research reactor at the Austrian Research Centres Seibersdorf was described within three technical papers already released in Nuclear Technology and Radiation Protection throughout the years 2003, 2006, and 2008. Following a suggestion from IAEA the project was investigated well after the files were closed regarding rather administrative than technical matters starting with the project mission, explaining the project structure and identifying the key factors and the key performance indicators. The continuous documentary and reporting system as implemented to fulfil the informational needs of stakeholders, management, and project staff alike is described. Finally the project is summarized in relationship to the performance indicators. (author)

  2. Management routes for materials arising from the decommissioning of a PWR reactor

    International Nuclear Information System (INIS)

    Klein, M.; Demeulemeester, Y.; Moers, S.; Ponnet, M.

    2001-01-01

    The management of wastes from decommissioning is described for the on-going dismantling of the BR3 PWR small reactor. The incentive is put on the radionuclides characterization, the description of the various waste streams, the conditioning techniques for low radioactive waste (LAW) to high radioactive waste (RAW), the alternative evacuation routes (recycling in the nuclear, free release by decontamination) and the minimization of secondary wastes during dismantling. Finally, some considerations are given on the overall dismantling cost and on the relative costs of the various evacuation routes. (author)

  3. Cutting Technology for Decommissioning of the Reactor Pressure Vessels in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Jeong, Kwan Seong; Kim, Geun Ho; Moon, Jei Kwon; Choi, Byung Seon

    2012-01-01

    Lots of nuclear power plants have been decommissioned during the last 2 decades. An essential part of this work is the dismantling of the Reactor Pressure Vessel and its Internals. For this purpose a wide variety of different cutting technologies have been developed, adapted and applied. A detailed introduction to Plasma Arc cutting, Contact Arc Metal cutting and Abrasive Water Suspension Jet cutting is given, as it turned out that these cutting technologies are particularly suitable for these type of segmentation work. A comparison of these technologies including gaseous emissions, cutting power, manipulator requirements as well as selected design approaches are given. Process limits as well as actual limits of application are presented

  4. Decontamination and decommissioning project of the TRIGA mark - 2 and 3 research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jung, K. J.; Baik, S. T.; Chung, U. S.; Jung, K. H.; Park, S. K.; Kim, J. K.; Lee, D. G.; Kim, H. R.; Lee, B. J.; Yang, S. H.

    2001-01-15

    The decommissioning license for KRR (Korea Research Reactor) 1 and 2 was issued Nov. 23, 2000. The atmospheric stability on the KRR site was evaluated using the meteorological data measured at the site. From the results of this evaluation, the population dose was evaluated for the public who lives at the periphery of the site. The Radiation Safety Management Guideline was developed and it will be used as a base line making Radiation Safety Management Procedure. The container was specially designed and manufactured for the storing of low level radioactive solid waste arising from the D and D activities. Firstly, the 50 containers were completely manufactured.

  5. Generation of an activation map for decommissioning planning of the Berlin Experimental Reactor-II

    Science.gov (United States)

    Lapins, Janis; Guilliard, Nicole; Bernnat, Wolfgang

    2017-09-01

    The BER-II is an experimental facility with 10 MW that was operated since 1974. Its planned operation will end in 2019. To support the decommissioning planning, a map with the overall distribution of relevant radionuclides has to be created according to the state of the art. In this paper, a procedure to create these 3-d maps using a combination of MCNP and deterministic methods is presented. With this approach, an activation analysis is performed for the whole reactor geometry including the most remote parts of the concrete shielding.

  6. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    M.E. Lumia; C.A. Gentile

    2002-01-01

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed

  7. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    CERN Document Server

    Lumia, M E

    2002-01-01

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed.

  8. Reuse of waste water from high pressure water jet decontamination for reactor decommissioning scrap metal

    International Nuclear Information System (INIS)

    Deng Junxian; Li Xin; Hou Huijuan

    2011-01-01

    For recycle and reuse of reactor decommissioning scrap metal by high pressure water jet decontamination, large quantity of radioactive waste water will be generated. To save the cost of radioactive waste water treatment and to reduce the cost of the scrap decontamination, this part of radioactive waste water should be reused. Most of the radioactivities in the decontamination waste water come from the solid particle in the water. Thus to reuse the waste water, the solid particle in the waster should be removed. Different possible treatment technologies have been investigated. By cost benefit analysis the centrifugal separation technology is selected. (authors)

  9. Decommissioning: A critical component of the design for uranium tailings management facilities

    International Nuclear Information System (INIS)

    Clifton, W.A.; Barsi, R.G.; Misfeldt, G.A.

    2000-01-01

    Uranium was discovered in the Beaverlodge area of northern Saskatchewan in 1934 with the first major mill beginning operation in 1953. Little attention was paid to tailings quality or tailings management practices. With the onset of the modem uranium operations beginning in the late 1970's, it was repeatedly evident, that the public had significant concerns, particularly with respect to tailings management, that must be addressed if the developments were to be allowed to proceed. Primary considerations related to environmental protection, public safety and an assurance of the ongoing sustainable development of the region. Integrating the decommissioning of a mine/mill site into development planning from the very outset has proven to be a critical component that has contributed to the ongoing success of the Saskatchewan uranium operations. This paper will provide a case study of the evolution of the uranium tailings management technology utilized in Saskatchewan. It documents the evolution of tailings management processes and the characteristics of tailings produced by successive mines in northern Saskatchewan. It also discusses the evolution of technologies applied to management of uranium mill tailings and demonstrates how progressively increasing levels of environmental protection have been achieved during the last 47 years of uranium mill operation. The paper also shows that the planned and progressive decommissioning of an operational site is the key to: Minimizing environmental impacts; Satisfying public and regulatory concerns; Minimizing operational and decommissioning costs; Minimizing corporate liability; and Shifting public resistance to public support. (author)

  10. Decommissioning: A critical component of the design for uranium tailings management facilities

    International Nuclear Information System (INIS)

    Clifton, A.W.; Barsi, R.G.; Misfeldt, G.A.

    2002-01-01

    Uranium was discovered in the Beaverlodge area of northern Saskatchewan in 1934 with the first major mill beginning operation in 1953. Little attention was paid to tailings quality or tailings management practices. With the onset of the modern uranium operations beginning in the late 1970's, it was repeatedly evident, that the public had significant concerns, particularly with respect to tailings management, that must be addressed if the developments were to be allowed to proceed. Primary considerations related to environmental protection, public safety and an assurance of the ongoing sustainable development of the region. Integrating the decommissioning of a mine/mill site into development planning from the very outset has proven to be a critical component that has contributed to the ongoing success of the Saskatchewan uranium operations. This paper will provide a case study of the evolution of the uranium tailings management technology utilized in Saskatchewan. It documents the evolution of tailings management processes and the characteristics of tailings produced by successive mines in northern Saskatchewan. It also discusses the evolution of technologies applied to management of uranium mill tailings and demonstrates how progressively increasing levels of environmental protection have been achieved during the last 47 years of uranium mill operation. The paper also shows that the planned and progressive decommissioning of an operational site is the key to: Minimizing environmental impacts; Satisfying public and regulatory concerns; Minimizing operational and decommissioning costs; Minimizing corporate liability; and Shifting public resistance to public support. (author)

  11. Research reactor decommissioning planning - It is never too early to start

    International Nuclear Information System (INIS)

    Eby, R.S.; Buttram, C.; Ervin, P.; Lundberg, L.; Hertel, N.; Marske, S.G.

    2003-01-01

    Whether an organization is in the process of designing, constructing or operating nuclear research reactors, past experiences prove it is never too early to start planning for the eventual decontamination, dismantlement and decommissioning (DD and D) of the reactor. If one waits until writing the Decommissioning Plan to seriously think about the DD and D activities, they have lost a key opportunity to be able to efficiently and effectively carry out the DD and D activities and will end up spending large sums of unnecessary funds during the DD and D. This paper will review all phases of research reactor decommissioning from characterization through planning, to eventual DD and D and license termination and highlight areas where early planning can significantly reduce the financial, safety and schedule risks associated with the DD and D activities. CH2M HILL served as the Executive Engineer for the Georgia Institute of Technology and the State of Georgia to oversee the successful DD and D of their tank type research reactor. CH2M HILL is currently serving as the DD and D contractor for the University of Virginia pool type UVAR and the low power CAVALIER research reactors and as the characterization and Decommissioning Planning contractor for the University of Michigan Ford Nuclear Reactor. Through these activities, an array of lessons learned have been compiled that will prove invaluable to the research reactor owner when they eventually face the DD and D challenge. As an example, in almost every case CH2M HILL has been involved in reactor DD and D, less than adequate up-front characterization has significantly impacted the ultimate DD and D process cost and schedule. Due to regulatory reasons, intrusive characterization may not always be possible prior to DD and D. However, a thorough understanding of the materials of construction and the quantities of additives or impurities present in those materials; e.g., cobalt in stainless steel, rare earth elements or

  12. Energy deposition in STARFIRE reactor components

    International Nuclear Information System (INIS)

    Gohar, Y.; Brooks, J.N.

    1985-04-01

    The energy deposition in the STARFIRE commercial tokamak reactor was calculated based on detailed models for the different reactor components. The heat deposition and the 14 MeV neutron flux poloidal distributions in the first wall were obtained. The poloidal surface heat load distribution in the first wall was calculated from the plasma radiation. The Monte Carlo method was used for the calculation to allow an accurate modeling for the reactor geometry

  13. Flexible solution of linear program with an application to decommissioning planning of nuclear reactor

    International Nuclear Information System (INIS)

    Shimizu, Yoshiaki

    1988-01-01

    Due to the simplicity and effectiveness, linear program has been popular in the actual optimization in various fields. In the previous study, the uncertainty involved in the model at the different stage of optimization was dealt with by post-optimizing analysis. But it often becomes insufficient to make a decision how to deal with an uncertain system especially suffering large parameter deviation. Recently in the field of processing systems, it is desired to obtain a flexible solution which can present the counterplan to a deviating system from a practical viewpoint. The scope of this preliminary note presents how to apply a methodology development to obtain the flexible solution of a linear program. For this purpose, a simple example associated with nuclear reactor decommissioning is shown. The problem to maximize a system performance given as an objective function under the constraint of the static behavior of the system is considered, and the flexible solution is determined. In Japan, the decommissioning of commercial nuclear power plants will being in near future, and the study using the retired research reactor JPDR is in progress. The planning of decontamination and the reuse of wastes is taken as the example. (Kako, I.)

  14. Cleanup and decommissioning of a nuclear reactor after a severe accident

    International Nuclear Information System (INIS)

    1992-01-01

    Although the development of commercial nuclear power plants has in general been associated with an excellent record of nuclear safety, the possibility of a severe accident resulting in major fuel and core damage cannot be excluded and such accidents have in fact already occurred. For over a decade, IAEA publications have provided technical guidance and recommendations for post-accident planning to be considered by appropriate authorities. Guidance and recommendations have recently been published on the management of damaged nuclear fuel, sealing of the reactor building and related safety and performance assessment aspects. The present technical report on the cleanup and decommissioning of reactors which have undergone a severe accident represents a further publication in the series. Refs, figs and tabs.

  15. Decontamination and decommissioning of the Organic Moderated Reactor Experiment facility (OMRE)

    International Nuclear Information System (INIS)

    Hine, R.E.

    1980-09-01

    This report describes the decontamination and decommissioning (D and D) of the Organic Moderated Reactor Experiment (OMRE) facility performed from October 1977 through September 1979. This D and D project included removal of all the facilities and as much contaminated soil and rock as practical. Removal of the reactor pressure vessel was an unusually difficult problem, and an extraordinary, unexpected amount of activated rock and soil was removed. After removal of all significantly contaminated material, the site consisted of a 20-ft deep excavation surrounded by backfill material. Before this excavation was backfilled, it and the backfill material were radiologically surveyed and detailed records made of these surveys. After the excavation was backfilled and graded, the site surface was surveyed again and found to be essentially uncontaminated

  16. Travelling cranes for heavy reactor component handling

    International Nuclear Information System (INIS)

    Champeil, M.

    1977-01-01

    Structure and operating machinery of two travelling cranes (600 t and 450 t) used in the Framatome factory for handling heavy reactor components are described. When coupled, these cranes can lift loads up to 1000 t [fr

  17. Status report on the Experimental Boiling Water Reactor (EBWR) Decontamination and Decommissioning (D ampersand D) Project

    International Nuclear Information System (INIS)

    Sears, L.; Garlock, G.; Mencarelli, R.; Fellhauer, C.

    1994-01-01

    ALARON Corporation is under contract, to Argonne National Laboratory - East (ANL-E), to complete the decontamination and decommissioning of the Experimental Boiling Water Reactor (EBWR). The project, begun, in 1986 by ANL-E personnel, is projected to be completed by the end of 1994. The final phase of work was awarded to ALARON in December 1993 with the scope of work including the disassembly and removal of all remaining reactor internals, the reactor vessel, the lead bio-shield, the core liner, and the activated portion of the concrete bio-shield. This paper discusses the work undertaken beginning in January 1994 and continuing through July 1994. During this period the required pre-mobilization documentation was prepared and approved, mobilization was completed, and the reactor internals, reactor vessel, lead bio-shield and core liner were removed. The paper will compare the planned schedule to the actual schedule, discuss problems encountered, review volume reduction techniques and health and safety issues including radiological aspects of the project

  18. Component failures that lead to reactor scrams

    International Nuclear Information System (INIS)

    Burns, E.T.; Wilson, R.J.; Lim, E.Y.

    1980-04-01

    This report summarizes the operating experience scram data compiled from 35 operating US light water reactors (LWRs) to identify the principal components/systems related to reactor scrams. The data base utilized to identify the scram causes is developed from a EPRI-utility sponsored survey conducted by SAI coupled with recent data from the USNRC Gray Books. The reactor population considered in this evaluation is limited to 23 PWRs and 12 BWRs because of the limited scope of the program. The population includes all the US NSSS vendors. It is judged that this population accurately characterizes the component-related scrams in LWRs over the first 10 years of plant operation

  19. Two Approaches to Reactor Decommissioning: 10 CFR Part 50 License Termination and License Amendment, Lessons Learned from the Regulatory Perspective

    International Nuclear Information System (INIS)

    Watson, B.A.; Buckley, J.T.; Craig, C.M.

    2006-01-01

    Trojan Nuclear Plant (Trojan) and Maine Yankee Nuclear Plant (Maine Yankee) were the first two power reactors to complete decommissioning under the U. S. Nuclear Regulatory Commission's (NRC's) License Termination Rule (LTR), 10 CFR Part 20, Subpart E. The respective owners' decisions to decommission the sites resulted in different approaches to both the physical aspects of the decommissioning, and the approach for obtaining approval for completing the decommissioning in accordance with regulations. Being in different States, the two single-unit pressurized water reactor sites had different State requirements and levels of public interest that impacted the decommissioning approaches. This resulted in significant differences in decommissioning planning, conduct of decommissioning operations, volumes of low- level radioactive waste disposed, and the final status survey (FSS) program. While both licensees have Independent Spent Fuel Storage Installations (ISFSIs), Trojan obtained a separate license for the ISFSI in accordance with the requirements of 10 CFR Part 72 and terminated their 10 CFR Part 50 license. Maine Yankee elected to obtain a general license under 10 CFR Part 50 for the ISFSI and reduce the physical site footprint to the ISFSI through a series of license amendments. While the NRC regulations are flexible and allow different approaches to ISFSI licensing there are separate licensing requirements that must be addressed. In 10 CFR 50.82, the NRC mandates public participation in the decommissioning process. For Maine Yankee, public input resulted in the licensee entering into an agreement with a concerned citizen group and resulted in State legislation that significantly lowered the dose limit below the NRC radiological criteria of 25 mrem (0.25 mSv) per year (yr) in 10 CFR 20.1402 for unrestricted use. The lowering of the radiological criteria resulted in a significant dose modeling effort using site-specific Derived Concentrations Guideline Levels (DCGLs

  20. New techniques for cutting and decontamination for decommissioning of nuclear research reactors with consideration of cost reduction. Final report

    International Nuclear Information System (INIS)

    Bach, W.; Redeker, C.F.; Versemann, R.

    2003-06-01

    In decommissioning of research reactors, specific boundary conditions, such as special materials, nuclides, geometries, spatial circumstances exist. This project aims on the development and adaptation of progressive procedures for decommissioning tasks with respect to economical aspects (cost reduction). Explored are Laser Cutting techniques, especially the cutting of aluminium in atmosphere and under water, remote- and manually controlled; decontamination and removal of concrete and ceramics by a combined diodelaser / cryogenic carbon dioxide thermal shock treatment (dry ice blasting) as well as cutting with the water-abrasive-suspension-jet and the plasma beam. The project will lead to application in the decommissioning of PTB's FRMB (research reactor of Physikalisch Technische Bundesanstalt, Braunschweig, Germany). (orig.) [de

  1. Cooling system for auxiliary reactor component

    International Nuclear Information System (INIS)

    Fujihira, Tomoko.

    1991-01-01

    A cooling system for auxiliary reactor components comprises three systems, that is, two systems of reactor component cooling water systems (RCCW systems) and a high pressure component cooling water system (HPCCW system). Connecting pipelines having partition valves are intervened each in a cooling water supply pipeline to an emmergency component of each of the RCCW systems, a cooling water return pipeline from the emmergency component of each of the RCCW systems, a cooling water supply pipeline to each of the emmergency components of one of the RCCW system and the HPCCW system and a cooling water return pipeline from each of the emmergency components of one of the RCCW system and the HPCCW system. With such constitution, cooling water can be supplied also to the emmergency components in the stand-by system upon periodical inspection or ISI, thereby enabling to improve the backup performance of the emmergency cooling system. (I.N.)

  2. Factors relevant to the recycling or reuse of components arising from the decommissioning and refurbishment of nuclear facilities

    International Nuclear Information System (INIS)

    1988-01-01

    The decommissioning and decontamination of nuclear facilities is a topic of great interest to many Member States of the International Atomic Energy Agency (IAEA) because of the large number of older nuclear facilities which are or soon will be retired from service. To assist in the development of the required decommissioning expertise, the IAEA is developing reports and recommendations which will eventually form an integrated information base covering in a systematic way the wide range of topics associated with decommissioning. This information is required so that Member States can decommission their nuclear facilities in a safe, timely and cost effective manner and the IAEA can effectively respond to requests for assistance. One area which warrants more detailed analyses is an assessment of the factors important to the recycling or reuse of components arising from the refurbishment or decommissioning of nuclear plants, the topic of the present report. The document provides an up to date review of the engineering, social, scientific and administrative factors relevant to the safe recycling or reuse of components arising from decommissioning or refurbishment of nuclear facilities. This report should be of interest to owners, operators, policy makers and regulators involved with nuclear facilities, especially those in developing countries. Refs, figs and tabs

  3. Nuclear decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    Lawton, H.

    1987-02-01

    Sufficient work has now been done, on a world-wide basis, to justify confidence that full decommissioning of nuclear installations, both plant and reactors, can be carried out safely and efficiently. Projects in several countries should confirm this in the next few years. In the UK, good progress has been made with the WAGR and supporting development work is finding solutions to resolve uncertainties. Estimates from several sources suggest that decommissioning costs can be kept to an acceptable level.

  4. Nuclear decommissioning

    International Nuclear Information System (INIS)

    Lawton, H.

    1987-01-01

    Sufficient work has now been done, on a world-wide basis, to justify confidence that full decommissioning of nuclear installations, both plant and reactors, can be carried out safely and efficiently. Projects in several countries should confirm this in the next few years. In the UK, good progress has been made with the WAGR and supporting development work is finding solutions to resolve uncertainties. Estimates from several sources suggest that decommissioning costs can be kept to an acceptable level. (author)

  5. Some aspects related to radioprotection during decommissioning of the WWR-S research reactor

    International Nuclear Information System (INIS)

    Pantazi, Doina; Stan, Camelia

    2007-01-01

    Radiological safety management ensures protection of personnel, public and environment. During decommissioning of a WWR-S type research reactor, besides other specific industrial problems, radiation and/or contamination sources will be produced and their effects have to be kept under control. In any decommissioning operation that implies working in a radioactive environment, the main concern being the minimization of the total dose received by the workers. To minimize the possible dose that an individual could receive, prior entering the working area, a definite set of stages of a radiation protection plan, developed according to ALARA principle, should be implemented. Of major interest is estimation the effective dose which operators will receive during a year, considering all operations in that he is involved and all the different possible paths of irradiation or contamination (inhalation, skin penetration, injury, etc.). The estimation of doses received by operating personnel will take into consideration the following steps: - the determination of jobs and events which could involve a significant radiation dose exposure; - whole body and extremities exposure doses should be assessed taking into consideration that the likelihood of contact with radiation and/or contamination sources is higher for hands and legs; - all possible paths of exposure will be identified (external irradiation is the most expected while the internal exposure due to intake could happen following an accidental inhalation of radionuclides or an injury in contaminated medium); - technological controls and administrative measures for exposure minimization will be rigorously implemented; - estimated doses will be compared with maximum permissible levels. The paper describes some general methodologies for computing the total effective doses received by workers involved in decommissioning operations, as well as their application for few special situations, that could contribute significantly to

  6. Shippingport: A relevant decommissioning project

    International Nuclear Information System (INIS)

    Crimi, F.P.

    1988-01-01

    Because of Shippingport's low electrical power rating (72 MWe), there has been some misunderstanding on the relevancy of the Shippingport Station Decommissioning Project (SSDP) to a modern 1175 MWe commercial pressurized water reactor (PWR) power station. This paper provides a comparison of the major components of the reactor plant of the 72 MWe Shippingport Atomic Power Station and an 1175 MWe nuclear plant and the relevancy of the Shippingport decommissioning as a demonstration project for the nuclear industry. For the purpose of this comparison, Portland General Electric Company's 1175 MWe Trojan Nuclear Plant at Rainier, Oregon, has been used as the reference nuclear power plant. 2 refs., 2 figs., 1 tab

  7. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 1. Main report. Technical report, September 1977-October 1979

    International Nuclear Information System (INIS)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE

  8. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 1. Main report. Technical report, September 1977-October 1979

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE.

  9. Safety and dose management during decommissioning of a fire damaged nuclear reactor

    International Nuclear Information System (INIS)

    Pomfret, D.G.

    2000-01-01

    Windscale Piles 1 and 2 in Cumbria in the UK were constructed in the early 1950s. They were not intended to produce electricity but were for military purposes only. They were graphite-moderated, air-cooled reactors with horizontal fuel channels fuelled with uranium rods clad in finned aluminium. In October 1957, Windscale Pile 1 suffered a core fire during a planned release of Wigner energy and both Piles were subsequently closed down. Following the fire, Pile 2 was defuelled entirely and as much fuel as was possible was removed from Pile 1. However, it is estimated that up to 15 Te of fuel remains in the core, a large proportion of which is located in the central Fire Affected Zone (FAZ). The condition of the fuel and graphite moderator in the FAZ is not known. It is possible that the moderator could contain voidage, greatly reduced density graphite and fused materials in a disordered matrix. It is probable that there is residual Wigner energy in the graphite. Use of water as part of the attempt to extinguish the fire, together with the possibility that the solidification of molten materials or other local sealing mechanisms could have excluded air mean that uranium hydrides, carbides and other pyrophorics may be present within the core. Pile 1 is now being decommissioned and a considerable amount of preparatory work has already been carried out during Phase 1. Phase 2 decommissioning, which will remove the residual fuel, the moderator and associated steelwork, condition the wastes and place them in a purpose built store is now underway. This work will be carried out for the site licensee, the United Kingdom Energy Authority, by a consortium of British Nuclear Fuels plc, Rolls Royce and NUKEM. This paper will describe the methods to be used to decommission Pile 1 and the systems and procedures which will be used to ensure that it is done safely and with the lowest reasonably practicable environmental impact. It will also describe the methods which will be used to

  10. Estimating pressurized water reactor decommissioning costs: A user's manual for the PWR Cost Estimating Computer Program (CECP) software

    International Nuclear Information System (INIS)

    Bierschbach, M.C.; Mencinsky, G.J.

    1993-10-01

    With the issuance of the Decommissioning Rule (July 27, 1988), nuclear power plant licensees are required to submit to the US Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. This user's manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personnel computer, provides estimates for the cost of decommissioning PWR plant stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning

  11. Mechanical development for reliable reactor components

    International Nuclear Information System (INIS)

    Ross-Ross, P.A.; Metcalfe, R.

    1983-09-01

    The CANDU reactor has achieved worldwide distinction because of its reliable performance. To achieve this, special attention was given to the reliability and maintainability of components in the heavy water circuits. Development programs were initiated early in the history of the CANDU reactor to improve the effectiveness of pump seals, valves, and static seals because of unacceptable performance of the commercial equipment then available. As a result, pump seals with a five year life now appear achievable, and valves and static seals are no longer a significant concern in CANDU reactors. Increasing effort is being given remotely operated tools and fabrication systems for radioactive environments

  12. Technology, safety and costs of decommissioning reference light water reactors following postulated accidents

    International Nuclear Information System (INIS)

    Konzek, G.J.; Smith, R.I.

    1990-12-01

    The estimated costs for post-accident cleanup at the reference BWR (developed previously in NUREG/CR-2601, Technology, Safety and Costs of Decommissioning Reference Light Water Reactors Following Postulated Accidents) are updated to January 1989 dollars in this report. A simple formula for escalating post-accident cleanup costs is also presented. Accident cleanup following the most severe accident described in NUREG/CR-2601 (i.e., the Scenario 3 accident) is estimated to cost from $1.22 to 1.44 billion, in 1989 dollars, for assumed escalation rates of 4% or 8% in the years following 1989. The time to accomplish cleanup remained unchanged from the 8.3 years originally estimated. No reanalysis of current information on the technical aspects of TMI-2 cleanup has been performed. Only the cost of inflation has been evaluated since the original PNL analysis was completed. 32 refs., 12 tabs

  13. Business operations and decommissioning strategy for imperial college London research reactor 'Consort' - A financial risk management approach

    International Nuclear Information System (INIS)

    Franklin, S.J.; Gardner, D.; Mumford, J.; Lea, R.; Knight, J.

    2005-01-01

    Imperial College London (IC) operates commercially a 100 kW research reactor, and as site licensee is responsible for funding both operations and eventual decommissioning. With long lead times ahead urgent decisions on the future business options have had to be made in 2004/5 including choices on whether to move to early decommissioning, recognising the high costs entailed, or to pursue continuing operations involving life extension measures such as refuelling. To develop a coherent overall approach strategy a financial risk driven programme was initiated to help define a robust transparent business and termination case for the reactor. This study was carried out in collaboration with a UK firm of financial risk experts, PURE Risk Management Ltd (PURE), working within a dedicated IC London reactor project team. This work evaluated immediate closure options due to financial constraints or life limiting failures, and options for continuing operation extending to 2028. Decommissioning and clean up were reviewed. Bespoke financial models created single value cost outputs and ranges of probabilistic net present values (NPV) for decommissioning costs and financial provisions to meet those costs at various levels of risk acceptance and regulatory compliance. (author)

  14. Decommissioning in the United States - Past, present and future - 16318

    International Nuclear Information System (INIS)

    Devgun, Jas S.

    2009-01-01

    The experience related to decommissioning of nuclear facilities in the United States is very substantial and covers power reactors, research reactors, and many facilities in the Department of Energy complex. The focus of this paper however is on the commercial power plants. With 104 operating reactors, the U.S. fleet of civilian reactors is still the largest in the world. Nuclear power industry in the United States has undergone a dramatic upturn after decades of stalemate. One effect of this nuclear renaissance has been that the plans have changed for several reactors that were initially destined for decommissioning. Instead, the focus now is on re-licensing of the reactors and on power up-rates. In fact, after the peak period between 1987 and 1998, no additional power reactors have been shutdown. On the contrary, power up-rates in the past twenty years have added a cumulative capacity equivalent to five new reactors. Almost all the operating reactors plan to have license extensions, thus postponing the eventual decommissioning. Nevertheless, in addition to the 9 reactors where licenses have been terminated following decommissioning, 12 power and early demonstration reactors and 14 test and research reactors are permanently shutdown and are in decommissioning phase. Substantial experience and lessons learned are available from the U.S. projects that are of value to the international decommissioning projects, especially where such projects are in early stages. These lessons cover a wide array of areas from decommissioning plans, technology applications, large component removal, regulatory and public interface, decommissioning funding and costs, clean up criteria, surveys of the decommissioned site, and license termination. Additionally, because of the unavailability of a national spent fuel disposition facility, most decommissioning sites are constructing above ground interim storage facilities for the spent nuclear fuel. The U.S. nuclear power projects are also

  15. Licensing activities for the partial decommissioning of IRT-2000 research reactor in Sofia

    International Nuclear Information System (INIS)

    Apostolov, T.; Ilieva, Kr.; Papukchiev, A.; Kalchev, B.

    2001-01-01

    The project for refurbishment of IRT-2000 research reactor in Sofia into low-power reactor (200 kW) is based on the retention of some IRT-2000 buildings, facilities and equipment. The activities, which determine the partial decommissioning should be realized in accordance with preliminary developed licensing documents as General Plan, Safety Analysis Report and Environment Impact assessment Report. The goal of these documents is to provide and guarantee safe and effective activities with radioactive materials, to define strictly the dismantling procedures, and in the same time to minimize their influence on the environment. The Technical Tasks for General Plan, Safety Analysis Report and Environment Impact Assessment Report have been prepared and will be presented as preliminary licensing documents to the National Regulatory Body for approval before their application. A Quality Management system is being developed nowadays at INRNE. After its certification some requirements of the regulatory body will be completed. This certified QA system is a major part of the licensing procedure for the reconstruction of IRT-2000 research reactor. (author)

  16. Conditioning for definitive storage of radioactive graphite bricks from reactor decommissioning

    International Nuclear Information System (INIS)

    Costes, J.R.; Koch, C.; Tassigny, C. de; Vidal, H.; Raymond, A.

    1990-01-01

    The decommissioning of gas-graphite reactors in the EC (e.g. French UNGGs, British Magnox reactors and AGRs, and reactors in Spain and in Italy) will produce large amounts of graphite bricks. This graphite cannot be accepted without particular conditioning by the existing shallow land disposal sites. The aim of the study is to examine the behaviour of graphite waste and to develop a conditioning technique which makes this waste acceptable for shallow land disposal sites. 18 kg of graphite core samples with an outside diameter of 74 mm were removed from the G2 gas-cooled reactor at Marcoule. Their radioactivity is highly dependent on the position of the graphite bricks inside the reactor. Measured results indicate an activity range of 100-400 MBq/kg with 90% Tritium, 5% 14 C, 3% 60 Co, 1.5% 63 Ni. Repeated porosity analyses showed that open porosity ranging from 0 to 100 μm exceeded 23 vol% in the graphite. Water penetration kinetics were investigated in unimpregnated graphite and resulted in impregnation by water of 50-90% of the open porosity. Preliminary lixiviation tests on the crude samples showed quick lixidegree of Cs (several per cent) and of 60 Co, and 133 Ba at a lesser degree. The proposed conditioning technique does not involve a simple coating but true impregnation by a tar-epoxy mixture. The bricks recovered intact from the core by robot services will be placed one by one inside a cylindrical metallic container. But this container may corrode and the bricks may become fragmented in the future, the normally porous graphite will be unaffected by leaching since it is proved that all pores larger than 0.1 μm will be filled with the tar-epoxy mixture. This is a true long-term waste packaging concept. The very simple technology required for industrial implementation is discussed

  17. Technology, safety, and costs of decommissioning reference light-water reactors following postulated accidents. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, E S; Holter, G M

    1982-11-01

    Appendices contain information concerning the reference site description; reference PWR facility description; details of reference accident scenarios and resultant contamination levels; generic cleanup and decommissioning information; details of activities and manpower requirements for accident cleanup at a reference PWR; activities and manpower requirements for decommissioning at a reference PWR; costs of decommissioning at a reference PWR; cost estimating bases; safety assessment details; and details of post-accident cleanup and decommissioning at a reference BWR.

  18. Nuclear decommissioning

    International Nuclear Information System (INIS)

    Lawton, H.

    1987-01-01

    Sufficient work has now been done, on a world-wide basis, to justify confidence that full decommissioning of nuclear installations, both plant and reactors, can be carried out safely and efficiently. Projects in several countries should confirm this in the next few years. In the United Kingdom, good progress has been made with the Windscale Advanced Gas-cooled Reactor and supporting development work is finding solutions to resolve uncertainties. Estimates from several sources suggest that decommissioning costs can be kept to an acceptable level. (author)

  19. Characterization, treatment and conditioning of radioactive graphite from decommissioning of nuclear reactors

    International Nuclear Information System (INIS)

    2006-09-01

    Graphite has been used as a moderator and reflector of neutrons in more than 100 nuclear power plants and in many research and plutonium-production reactors. It is used primarily as a neutron reflector or neutron moderator, although graphite is also used for other features of reactor cores, such as fuel sleeves. Many of the graphite-moderated reactors are now quite old, with some already shutdown. Therefore radioactive graphite dismantling and the management of radioactive graphite waste are becoming an increasingly important issue for a number of IAEA Member States. Worldwide, there are more than 230 000 tonnes of radioactive graphite which will eventually need to be managed as radioactive waste. Proper management of radioactive graphite waste requires complex planning and the implementation of several interrelated operations. There are two basic options for graphite waste management: (1) packaging of non-conditioned graphite waste with subsequent direct disposal of the waste packages, and (2) conditioning of graphite waste (principally either by incineration or calcination) with separate disposal of any waste products produced, such as incinerator ash. In both cases, the specific properties of graphite - such as Wigner energy, graphite dust explosibility, and radioactive gases released from waste graphite - have a potential impact on the safety of radioactive graphite waste management and need to be carefully considered. Radioactive graphite waste management is not specifically addressed in IAEA publications. Only general and limited information is available in publications dealing with decommissioning of nuclear reactors. This report provides a comprehensive discussion of radioactive graphite waste characterization, handling, conditioning and disposal throughout the operating and decommissioning life cycle. The first draft report was prepared at a meeting on 23-27 February 1998. A technical meeting (TM) was held in October 1999 in coincidence with the Seminar on

  20. Considerations about decommissioning of the IEA-R1 research reactor and the future of its installations after shutdown

    International Nuclear Information System (INIS)

    Frajndlich, Roberto

    2014-01-01

    The IEA-R1 Nuclear Research Reactor, in operation since 1957, in the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), is one of the oldest research reactors in the world. However at some point in time in the future, as example of the other reactors, it will be shutdown definitively. Before that time actually arrives, the operational organization needs to plan the future of its installations and define the final destination of equipment and radioactive as well as non-radioactive material contained inside the installations. These and other questions should be addressed in the so called Preliminary decommissioning plan of the installation, which is the subject of this work. The work initially presents an over view about the theme and defines the general and specific objectives describing, in succession, the directions that the operating organization should consider for the formulation of a decommissioning plan. The present structure of the Brazilian nuclear sector emphasizing principally the norms utilized in the management of radioactive waste is also presented. A description of principle equipment of the IEA-R1 reactor which constitutes its inventory of radioactive and non-radioactive material is given. The work emphasizes the experience of the reactor technicians, acquired during several reforms and modifications of the reactor installations realized during its useful life time. This experience may be of great help for the decommissioning in the future. An experiment using the high resolution gamma spectrometric method and computer calculation using Monte Carlo theory were performed with the objective of obtaining an estimate of the radioactive waste produced from dismantling of the reactor pool walls. The cost of reactor decommissioning for different choices of strategies was determined using the CERREX code. Finally, a discussion about different strategies is presented. On the basis of these discussions it is concluded that the most advantageous

  1. 'Experience with decommissioning of research and test reactors at Argonne National Laboratory'

    International Nuclear Information System (INIS)

    Bhattacharyya, S.K.; Yule, T.J.; Fellhauer, C.R.; Boing, L.E.

    2002-01-01

    A large number of research reactors around the world have reached the end of their useful operational life. Many of these are kept in a controlled storage mode awaiting decontamination and decommissioning (D and D). At Argonne National Laboratory located near Chicago in the United States of America, significant experience has been gained in the D and D of research and test reactors. These experiences span the entire range of activities in D and D - from planning and characterization of the facilities to the eventual disposition of all waste. A multifaceted D nd D program has been in progress at the Argonne National Laboratory - East site for nearly a decade. The program consists of three elements: - D and D of nuclear facilities on the site that have reached the end of their useful life; - Development and demonstrations of technologies that help in safe and cost effective D and D; - Presentation of training courses in D and D practices. Nuclear reactor facilities have been constructed and operated at the ANL-E site since the earliest days of nuclear power. As a result, a number of these early reactors reached end-of-life long before reactors on other sites and were ready for D and D earlier. They presented an excellent set of test beds on which D and D practices and technologies could be demonstrated in environments that were similar to commercial reactors, but considerably less hazardous. As shown, four reactor facilities, plutonium contaminated glove boxes and hot cells, a cyclotron facility and assorted other nuclear related facilities have been decommissioned in this program. The overall cost of the program has been modest relative to the cost of comparable projects undertaken both in the U.S. and abroad. The safety record throughout the program was excellent. Complementing the actual operations, a set of D and D technologies are being developed. These include robotic methods of tool handling and operation, chemical and laser decontamination techniques, sensors

  2. External vibrations measurement of reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Rogers, S A [Nuclear Electric plc, Barnwood (United Kingdom); Sugden, J [Magnox Electric, Berkeley (United Kingdom)

    1997-12-31

    The paper outlines the use of External Vibration Monitoring for remote vibration assessment of internal reactor components. The main features of the technique are illustrated by a detailed examination of the specific application to the problem of Heysham 2 Fuel Plug Unit monitoring. (author). 6 figs.

  3. Decommissioning the Romanian Water-Cooled Water-Moderated Research Reactor: New Environmental Perspective on the Management of Radioactive Waste

    International Nuclear Information System (INIS)

    Barariu, G.; Giumanca, R.

    2006-01-01

    Pre-feasibility and feasibility studies were performed for decommissioning of the water-cooled water-moderated research reactor (WWER) located in Bucharest - Magurele, Romania. Using these studies as a starting point, the preferred safe management strategy for radioactive wastes produced by reactor decommissioning is outlined. The strategy must account for reactor decommissioning, as well as for the rehabilitation of the existing Radioactive Waste Treatment Plant and for the upgrade of the Radioactive Waste Disposal Facility at Baita-Bihor. Furthermore, the final rehabilitation of the laboratories and ecological reconstruction of the grounds need to be provided for, in accordance with national and international regulations. In accordance with IAEA recommendations at the time, the pre-feasibility study proposed three stages of decommissioning. However, since then new ideas have surfaced with regard to decommissioning. Thus, taking into account the current IAEA ideology, the feasibility study proposes that decommissioning of the WWER be done in one stage to an unrestricted clearance level of the reactor building in an Immediate Dismantling option. Different options and the corresponding derived preferred option for waste management are discussed taking into account safety measures, but also considering technical, logistical and economic factors. For this purpose, possible types of waste created during each decommissioning stage are reviewed. An approximate inventory of each type of radioactive waste is presented. The proposed waste management strategy is selected in accordance with the recommended international basic safety standards identified in the previous phase of the project. The existing Radioactive Waste Treatment Plant (RWTP) from the Horia Hulubei Institute for Nuclear Physics and Engineering (IFIN-HH), which has been in service with no significant upgrade since 1974, will need refurbishing due to deterioration, as well as upgrading in order to ensure the

  4. Plan for fully decontaminating and decommissioning of the Westinghouse Advanced Reactors Division Fuel Laboratories at Cheswick, Revision 3

    International Nuclear Information System (INIS)

    1982-01-01

    The project scope of work included the complete decontamination and decommissioning (D and D) of the Westinghouse ARD Fuel Laboratories at the Cheswick Site in the shortest possible time. This has been accomplished in the following four phases: (1) preparation of documents and necessary paperwork; packaging and shipping of all special nuclear materials in an acceptable form to a reprocessing agency; (2) decontamination of all facilities, glove boxes and equipment; loading of generated waste into bins, barrels and strong wooden boxes; (3) shipping of all bins, barrels and boxes containing waste to the designated burial site; removal of all utility services from the laboratories; (4) final survey of remaining facilities and certification for nonrestricted use; preparation of final report. This volume contains the following 3 attachments: (1) Plan for Fully Decontamination and Decommissioning of the Westinghouse Advanced Reactors Division Fuel Laboratories at Cheswick; (2) Environmental Assessment for Decontamination and Decommissioning the Westinghouse Advanced Reactors Division Plutonium Fuel Laboratories, Cheswick, PA; and (3) WARD-386, Quality Assurance Program Description for Decontamination and Decommissioning Activities

  5. Decontamination and Decommissioning Project of the TRIGA Mark - 2 and 3 research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jung, K. J.; Baik, S. T.; Chung, U. S.; Park, S. K.; Moon, J. S.; Jung, K. H.; Lee, B. J.; Kim, J. K.; Kim, K. H

    1999-02-01

    Design work for the D and D began in 1998, and expected to be finished at the end of February 1999. Base on the D and D design, the decommissioning plan and the environmental impact assessment report have been completed and submitted to the MOST for licensing at the end of 1998. These documents are being reviewing at the KINS. It is expected that the licensing from authority will be come out at the end of September 1999. Then D and D practical work will be started in the latter half of the year 1999 and completed by 2002. The first practical work of the D and D will be the KRR-2 reactor hall to transform the hall as a temporary storage of radioactive waste produced during the D and D work. Meanwhile, all the spent fuel from KRR-1 and 2 were safely transported to us at the middle of 1998. The D and D project was originally planned, to be finished by the end of the year 1999. This project were lately modified and extended until the end of the year 2002 because of the interim storage of radioactive wastes arising from the D and D work. These radioactive wastes will be stored at the KRR-2 reactor hall until a disposal facility is operational. (author)

  6. Modular core component support for nuclear reactor

    International Nuclear Information System (INIS)

    Finch, L.M.; Anthony, A.J.

    1975-01-01

    The core of a nuclear reactor is made up of a plurality of support modules for containing components such as fuel elements, reflectors and control rods. Each module includes a component support portion located above a grid plate in a low-pressure coolant zone and a coolant inlet portion disposed within a module receptacle which depends from the grid plate into a zone of high-pressure coolant. Coolant enters the module through aligned openings within the receptacle and module inlet portion and flows upward into contact with the core components. The modules are hydraulically balanced within the receptacles to prevent expulsion by the upward coolant forces. (U.S.)

  7. The decision on the application to carry out a decommissioning project at Hinkley Point A Power Station under the Nuclear Reactors (Environmental Impact Assessment for Decommissioning) Regulations 1999

    International Nuclear Information System (INIS)

    2003-01-01

    European Council Directive 85/337/EEC, as amended by Council Directive 97/1 I/EC, sets out a framework on the assessment of the effects of certain public and private projects on the environment. The Directive is implemented in Great Britain for decommissioning nuclear reactor projects by the Nuclear Reactors (Environmental Impact Assessment for Decommissioning) Regulations 1999. The intention of the Directive and Regulations is to involve the public through consultation in considering the potential environmental impacts of a decommissioning project, and to make the decision-making process on granting consent open and transparent. The Regulations require the licensee to undertake an environmental impact assessment, prepare an environmental statement that summarises the environmental effects of the project, and apply to the Health and Safety Executive (HSE) for consent to carry out a decommissioning project. There is an optional stage where the licensee may request from HSE an opinion on what the environmental statement should contain (called a pre-application opinion). The licensee of Hinkley Point A Power Station, Magnox Electric pie, requested a pre-application opinion and provided information in a scoping report in December 2000. HSE undertook a public consultation on the scoping report and provided its pre- application opinion in April 2001. The licensee applied to HSE for consent to carry out a decommissioning project and provided an environmental statement in December 2001. Following a public consultation on the environmental statement, HSE requested further information that was subsequently provided by the licensee. A further public consultation was undertaken on the further information that ended in March 2003. All these public consultations involved around 60 organisations. HSE granted consent to carry out a decommissioning project at Hinkley Point A Power Station under the Regulations in July 2003, and attached conditions to the Consent. HSE took relevant

  8. An overview of the U.S. Department of Energy Experimental Boiling Water Reactor Decontamination and Decommissioning Project

    International Nuclear Information System (INIS)

    Murphie, W.E.; Mckernan, M.L.

    1991-01-01

    This paper provides an overview of the U.S. Department of Energy's (DOE) Experimental Boiling Water Reactor (EBWR) Decontamination and Decommissioning (D and D) Project. Physical decommissioning work started in 1986 and is scheduled for completion in 1994. The project total estimated cost is 14.3 million (1990, U.S.) dollars. The reactor pressure vessel will be removed by segmentation. Another notable project feature is that D and D operations were planned for and carried out with a small work force comprised of four to six D and D laborers, one or two health physics technicians, an engineer, and a project manager. When the D and D work is completed the facility will be recycled for other productive uses. (author)

  9. Envisioning Communications with Future Stakeholders - A Case Study Using the In-Situ Decommissioning of P-Reactor

    International Nuclear Information System (INIS)

    Antonucci, D.L.

    2009-01-01

    This paper will explore opportunities to expand the CAB's public outreach by the incorporation of technologies typically used in social networks and distance learning. Envisioning opportunities to engage next generation CAB members in public involvement will be delineated by retracing the decision process used with the in-situ decommissioning of P-Reactor at the Savannah River Site (SRS). This paper will discuss existing opportunities to enable another group of stakeholders to take part in the environmental policy decision making process regarding the inclusion of some very long lived radioactive constituents. The aim of the paper will be to locate places in the current process where alternate or parallel informational dissemination pathways could exist. These alternatives will incorporate the next generation's expectation for instantaneous information and universal ownership of hand-held communication devices. The goal of this paper is to use the present framework of CAB communications and add the components of virtual networking and distance learning in hopes of bridging the generational technology gap and extending the dialog to future stakeholders. (authors)

  10. Decommissioning handbook

    Energy Technology Data Exchange (ETDEWEB)

    Manion, W.J.; LaGuardia, T.S.

    1980-11-01

    This document is a compilation of information pertinent to the decommissioning of surplus nuclear facilities. This handbook is intended to describe all stages of the decommissioning process including selection of the end product, estimation of the radioactive inventory, estimation of occupational exposures, description of the state-of-the-art in re decontamination, remote csposition of wastes, and estimation of program costs. Presentation of state-of-the-art technology and data related to decommissioning will aid in consistent and efficient program planning and performance. Particular attention is focused on available technology applicable to those decommissioning activities that have not been accomplished before, such as remote segmenting and handling of highly activated 1100 MW(e) light water reactor vessel internals and thick-walled reactor vessels. A summary of available information associated with the planning and estimating of a decommissioning program is also presented. Summarized in particular are the methodologies associated with the calculation and measurement of activated material inventory, distribution, and surface dose level, system contamination inventory and distribution, and work area dose levels. Cost estimating techniques are also presented and the manner in which to account for variations in labor costs as impacting labor-intensive work activities is explained.

  11. Decommissioning handbook

    International Nuclear Information System (INIS)

    Manion, W.J.; LaGuardia, T.S.

    1980-11-01

    This document is a compilation of information pertinent to the decommissioning of surplus nuclear facilities. This handbook is intended to describe all stages of the decommissioning process including selection of the end product, estimation of the radioactive inventory, estimation of occupational exposures, description of the state-of-the-art in re decontamination, remote csposition of wastes, and estimation of program costs. Presentation of state-of-the-art technology and data related to decommissioning will aid in consistent and efficient program planning and performance. Particular attention is focused on available technology applicable to those decommissioning activities that have not been accomplished before, such as remote segmenting and handling of highly activated 1100 MW(e) light water reactor vessel internals and thick-walled reactor vessels. A summary of available information associated with the planning and estimating of a decommissioning program is also presented. Summarized in particular are the methodologies associated with the calculation and measurement of activated material inventory, distribution, and surface dose level, system contamination inventory and distribution, and work area dose levels. Cost estimating techniques are also presented and the manner in which to account for variations in labor costs as impacting labor-intensive work activities is explained

  12. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Appendices. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Appendices are presented concerning the evaluations of decommissioning financing alternatives; reference site description; reference BWR facility description; radiation dose rate and concrete surface contamination data; radionuclide inventories; public radiation dose models and calculated maximum annual doses; decommissioning methods; generic decommissioning information; immediate dismantlement details; passive safe storage, continuing care, and deferred dismantlement details; entombment details; demolition and site restoration details; cost estimating bases; public radiological safety assessment details; and details of alternate study bases.

  13. Decommissioning of DR 1, Final report

    International Nuclear Information System (INIS)

    Lauridsen, Kurt

    2006-01-01

    The report describes the decommissioning activities carried out at the 2kW homogeneous reactor DR 1 at Risoe National Laboratory. The decommissioning work took place from summer 2004 until late autumn 2005. The components with the highest activity, the core vessel the recombiner and the piping and valves connected to these, were dismantled first by Danish Decommissioning's own technicians. Demolition of the control rod house and the biological shield as well as the removal of the floor in the reactor hall was carried out by an external demolition contractor. The building was emptied and left for other use. Clearance measurements of the building showed that radionuclide concentrations were everywhere below the clearance limit set by the Danish nuclear regulatory authorities. Furthermore, measurements on the surrounding area showed that there was no contamination that could be attributed to the operation and decommissioning of DR 1. (au)

  14. Decommissioning of DR 1, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lauridsen, Kurt

    2006-01-15

    The report describes the decommissioning activities carried out at the 2kW homogeneous reactor DR 1 at Risoe National Laboratory. The decommissioning work took place from summer 2004 until late autumn 2005. The components with the highest activity, the core vessel the recombiner and the piping and valves connected to these, were dismantled first by Danish Decommissioning's own technicians. Demolition of the control rod house and the biological shield as well as the removal of the floor in the reactor hall was carried out by an external demolition contractor. The building was emptied and left for other use. Clearance measurements of the building showed that radionuclide concentrations were everywhere below the clearance limit set by the Danish nuclear regulatory authorities. Furthermore, measurements on the surrounding area showed that there was no contamination that could be attributed to the operation and decommissioning of DR 1. (au)

  15. Thermal analysis to support decommissioning of the molten salt reactor experiment

    International Nuclear Information System (INIS)

    Sulfredge, C.D.; Morris, D.G.; Park, J.E.; Williams, P.T.

    1996-06-01

    As part of the decommissioning process for the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory, several thermal-sciences issues were addressed. Apparently a mixture of UF 6 and F 2 had diffused into the upper portion of one charcoal column in the MSRE auxiliary charcoal bed (ACB), leading to radiative decay heating and possible chemical reaction sources. A proposed interim corrective action was planned to remove the water from the ACB cell to reduce criticality and reactivity concerns and then fill the ACB cell with an inert material. This report describes design of a thermocouple probe to obtain temperature measurements for mapping the uranium deposit, as well as development of steady-state and transient numerical models for the heat transfer inside the charcoal column. Additional numerical modeling was done to support filling of the ACB cell. Results from this work were used to develop procedures for meeting the goals of the MSRE Remediation Project without exceeding appropriate thermal limits

  16. Decontamination and decommissioning of the Experimental Boiling Water Reactor (EBWR): Project final report, Argonne National Laboratory

    International Nuclear Information System (INIS)

    Fellhauer, C.R.; Boing, L.E.; Aldana, J.

    1997-03-01

    The Final Report for the Decontamination and Decommissioning (D ampersand D) of the Argonne National Laboratory - East (ANL-E) Experimental Boiling Water Reactor (EBWR) facility contains the descriptions and evaluations of the activities and the results of the EBWR D ampersand D project. It provides the following information: (1) An overall description of the ANL-E site and EBWR facility. (2) The history of the EBWR facility. (3) A description of the D ampersand D activities conducted during the EBWR project. (4) A summary of the final status of the facility, including the final and confirmation surveys. (5) A summary of the final cost, schedule, and personnel exposure associated with the project, including a summary of the total waste generated. This project report covers the entire EBWR D ampersand D project, from the initiation of Phase I activities to final project closeout. After the confirmation survey, the EBWR facility was released as a open-quotes Radiologically Controlled Area,close quotes noting residual elevated activity remains in inaccessible areas. However, exposure levels in accessible areas are at background levels. Personnel working in accessible areas do not need Radiation Work Permits, radiation monitors, or other radiological controls. Planned use for the containment structure is as an interim transuranic waste storage facility (after conversion)

  17. Thermal analysis to support decommissioning of the molten salt reactor experiment

    Energy Technology Data Exchange (ETDEWEB)

    Sulfredge, C.D.; Morris, D.G.; Park, J.E.; Williams, P.T.

    1996-06-01

    As part of the decommissioning process for the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory, several thermal-sciences issues were addressed. Apparently a mixture of UF{sub 6} and F{sub 2} had diffused into the upper portion of one charcoal column in the MSRE auxiliary charcoal bed (ACB), leading to radiative decay heating and possible chemical reaction sources. A proposed interim corrective action was planned to remove the water from the ACB cell to reduce criticality and reactivity concerns and then fill the ACB cell with an inert material. This report describes design of a thermocouple probe to obtain temperature measurements for mapping the uranium deposit, as well as development of steady-state and transient numerical models for the heat transfer inside the charcoal column. Additional numerical modeling was done to support filling of the ACB cell. Results from this work were used to develop procedures for meeting the goals of the MSRE Remediation Project without exceeding appropriate thermal limits.

  18. Safety aspects in decontamination operations: Lessons learned during the decommissioning of a small PWR reactor

    International Nuclear Information System (INIS)

    Klein, M.; Ponnet, M.; Emond, O.

    2002-01-01

    Decontamination operations are generally executed during the decommissioning of nuclear installations for different objectives: decontamination of loops or large pieces to reduce the dose rate inside a contaminated plant or decontamination to minimize the amount of radioactive waste. These decontamination operations raise safety issues such as radiological exposure, classical safety, environmental releases, production and management of secondary waste, management of primary resources, etc. This paper presents the return of experience from decontamination operations performed during the dismantling of the BR3 PWR reactor. The safety issues are discussed for 3 types of decontamination operations: full system decontamination of the primary loop with a chemical process to reduce the dose rate by a factor of 10; thorough decontamination with an aggressive chemical process of dismantled pieces to reach the unconditional clearance values; and thorough decontamination processes with physical processes of metals and of concrete to reach the unconditional clearance values. For the protection of the workers, we must consider the ALARA aspects and the classical safety issues. During the progress of our dismantling operations, the dose rate issue was becoming less important but the classical safety issues were becoming preponderant due to the use of very aggressive techniques. For the protection of the environment, we must take all the precautions to avoid any leakages from the plant and we must use processes which minimize the use of toxic products and which minimize the production of secondary wastes. We therefore promote the use of regenerative processes. (author)

  19. Nuclide Inventory Calculation Using MCNPX for Wolsung Unit 1 Reactor Decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    Rabir, Mohamad Hairie; Noh, Kyoung Ho; Hah, Chang Joo [KEPCO International Nuclear Graduate School, Daejeon (Korea, Republic of)

    2014-05-15

    The CINDER90 computation process involves utilizing linear Markovian chains to determine the time dependent nuclide densities. The CINDER90 depletion algorithm is implemented the MCNPX code package. The coupled depletion process involves a Monte-Carlo steady-state reaction rate calculation linked to a deterministic depletion calculation. The process is shown in Fig.1. MCNPX runs a steady state calculation to determine the system eigenvalue collision densities, recoverable energies from fission and neutrons per fission events. In order to generate number densities for the next time step, the CINDER90 code takes the MCNPX generated values and performs a depletion calculation. MCNPX then takes the new number densities and caries out a new steady-stated calculation. The process repeats itself until the final time step. This paper describe the preliminary source term and nuclide inventory calculation of Candu single fuel channel using MCNPX, as a part of the activities to support the equilibrium core model development and decommissioning evaluation process of a Candu reactor. The aim of this study was to apply the MCNPX code for source term and nuclide inventory calculation of Candu single fuel channel. Nuclide inventories as a function of burnup will be used to model an equilibrium core for Candu reactor. The core lifetime neutron fluence obtained from the model is used to estimate radioactivity at the stage of decommisioning. In general, as expected, the actinides and fission products build up increase with increasing burnup. Despite the fact that the MCNPX code is still in development we can conclude that the code is capable of obtaining relevant results in burnup and source term calculation. It is recommended that in the future work, the calculation has to be verified on the basis of experimental data or comparison with other codes.

  20. Waste from decommissioning of research reactors and other small nuclear facilities

    International Nuclear Information System (INIS)

    Massaut, V.

    2001-01-01

    Full text: Small nuclear facilities were often built for research or pilot purposes. It includes the research reactors of various types and various aims (physics research, nuclear research, nuclear weapons development, materials testing reactor, isotope production, pilot plant, etc.) as well as laboratories, hot cells and accelerators used for a broad spectrum of research or production purposes. These installations are characterized not only by their size (reduced footprint) but also, and even mostly, by the very diversified type of materials, products and isotopes handled within these facilities. This large variety can sometimes enhance the difficulties encountered for the dismantling of such facilities. The presence of materials like beryllium, graphite, lead, PCBs, sodium, sometimes in relatively large quantities, are also challenges to be faced by the dismantlers of such facilities, because these types of waste are either toxic or no solutions are readily available for their conditioning or long term disposal. The paper will review what is currently done in different small nuclear facilities, and what are the remaining problems and challenges for future dismantling and waste management. The question of whether Research and Development for waste handling methods and processes is needed is still pending. Even for the dismantling operation itself, important improvements can be brought in the fields of characterization, decontamination, remote handling, etc. by further developments and innovative systems. The way of funding such facilities decommissioning will be reviewed as well as the very difficult cost estimation for such facilities, often one-of-a-kind. The aspects of radioprotection optimization (ALARA principle) and classical operators safety will also be highlighted, as well as the potential solutions or improvements. In fact, small nuclear facilities encounter often, when dismantling, the same problems as the large nuclear power plants, but have in

  1. A study of implementing In-Cycle-Shuffle strategy to a decommissioning boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Chung-Yuan, E-mail: tuckjason@iner.gov.tw; Tung, Wu-Hsiung; Yaur, Shyun-Jung

    2017-06-15

    Highlights: • A loading pattern strategy ICS (In-Cycle-Shuffle) was implemented to the last cycle of the boiling water reactor. • The best power sharing distribution and ICS timing was found. • A new parameter “Burnup sharing” is presented to evaluate ICS strategy. - Abstract: In this paper, a loading pattern strategy In-Cycle-Shuffle (ICS) is implemented to the last cycle of the boiling water reactor (BWR) before decommissioning to save the fuel cycle cost. This method needs a core shutdown during the operation of a cycle to change the loading pattern to gain more reactivity. The reactivity model is used to model the ICS strategy in order to find out the best ICS timing and the optimum power sharing distribution before ICS and after ICS. Several parameters of reactivity model are modified and the effect of burnable poison, gadolinium (Gd), is considered in this research. Three cases are presented and it is found that the best ICS timing is at about two-thirds of total cycle length no matter the poisoning effect of Gd is considered or not. According to the optimum power sharing distribution result, it is suggested to decrease the once burnt power and increase the thrice burnt fuel power as much as possible before ICS. After ICS, it is suggested to increase the positive reactivity fuel power and decrease the thrice burnt fuel power as much as possible. A new parameter “Burnup sharing” is presented to evaluate the special case whose EOC power weighting factor and the burnup accumulation factor in the reactivity model are quite different.

  2. A study of implementing In-Cycle-Shuffle strategy to a decommissioning boiling water reactor

    International Nuclear Information System (INIS)

    Chen, Chung-Yuan; Tung, Wu-Hsiung; Yaur, Shyun-Jung

    2017-01-01

    Highlights: • A loading pattern strategy ICS (In-Cycle-Shuffle) was implemented to the last cycle of the boiling water reactor. • The best power sharing distribution and ICS timing was found. • A new parameter “Burnup sharing” is presented to evaluate ICS strategy. - Abstract: In this paper, a loading pattern strategy In-Cycle-Shuffle (ICS) is implemented to the last cycle of the boiling water reactor (BWR) before decommissioning to save the fuel cycle cost. This method needs a core shutdown during the operation of a cycle to change the loading pattern to gain more reactivity. The reactivity model is used to model the ICS strategy in order to find out the best ICS timing and the optimum power sharing distribution before ICS and after ICS. Several parameters of reactivity model are modified and the effect of burnable poison, gadolinium (Gd), is considered in this research. Three cases are presented and it is found that the best ICS timing is at about two-thirds of total cycle length no matter the poisoning effect of Gd is considered or not. According to the optimum power sharing distribution result, it is suggested to decrease the once burnt power and increase the thrice burnt fuel power as much as possible before ICS. After ICS, it is suggested to increase the positive reactivity fuel power and decrease the thrice burnt fuel power as much as possible. A new parameter “Burnup sharing” is presented to evaluate the special case whose EOC power weighting factor and the burnup accumulation factor in the reactivity model are quite different.

  3. Technology, safety, and costs of decommissioning a reference pressurized water reactor power station. Appendices

    International Nuclear Information System (INIS)

    Smith, R.I.; Konzek, G.J.; Kennedy, W.E. Jr.

    1978-05-01

    Detailed appendices are presented under the following headings: reference PWR facility description, reference PWR site description, estimates of residual radioactivity, alternative methods for financing decommissioning, radiation dose methodology, generic decommissioning activities, intermediate dismantlement activities, safe storage and deferred dismantlement activities, compilation of unit cost factors, and safety assessment details

  4. Design codes for gas cooled reactor components

    International Nuclear Information System (INIS)

    1990-12-01

    High-temperature gas-cooled reactor (HTGR) plants have been under development for about 30 years and experimental and prototype plants have been operated. The main line of development has been electricity generation based on the steam cycle. In addition the potential for high primary coolant temperature has resulted in research and development programmes for advanced applications including the direct cycle gas turbine and process heat applications. In order to compare results of the design techniques of various countries for high temperature reactor components, the IAEA established a Co-ordinated Research Programme (CRP) on Design Codes for Gas-Cooled Reactor Components. The Federal Republic of Germany, Japan, Switzerland and the USSR participated in this Co-ordinated Research Programme. Within the frame of this CRP a benchmark problem was established for the design of the hot steam header of the steam generator of an HTGR for electricity generation. This report presents the results of that effort. The publication also contains 5 reports presented by the participants. A separate abstract was prepared for each of these reports. Refs, figs and tabs

  5. Lessons learned from the shut down, planning, and the preparatory activities of decommissioning the research reactor VVR-S Magurele, Bucharest

    International Nuclear Information System (INIS)

    Dragusin, M.; Copaciu, V.

    2006-01-01

    The nuclear research reactor type VVR was shut down in December 1997 after forty years of operation. The main characteristics of this reactor are: Thermal power 2 MW, Thermal energy - 9.59 GWhd, Average flux of thermal neutrons-10 13 n/cm 2 .s, nine horizontal channels, sixteen vertical exposure channels, three biological channels, reactor type tank, water used as a moderator, coolant and reflector. The reactor was used in research and radioisotope production. The reactor has been permanently shut down since April 2002, when the decommissioning was officially announced. Discussions regarding funding mechanisms for the conservation phase, and decommissioning (planning, preparatory activities, spent nuclear fuel management), have taken place since five years ago when the final decision of permanent shut down was taken. Quality management includes procedures for recording and archiving the lessons learned. The planning of decommissioning started in 1990 when the reactor was still operational. After fifteen years the regulatory body has not yet approved the decommissioning plan for the reactor. In this paper the following aspects are discussed: decommissioning strategy from safe enclosure to immediate dismantling, specific features of the site (treatment of radioactive waste near reactor) and state of decommissioning, use of the lessons learned in the planning of decommissioning for the other two small nuclear facilities situated in the same area with VVR-reactor: Sub critical Assembly 'HELEN' and Zero Power Critical Reactor RP-0, AFR ponds for spent nuclear fuel, other radiological facilities for radioisotopes production facilities radiation processing and accelerators. Preparatory activities for decommissioning have included: elaboration of a plan (inter alia, justification of the selected strategy, management of the radioactive waste in accordance with the waste acceptance criteria), reactor storage in parallel with the removal of the equipment and materials used in

  6. Hydrodynamic impact of reactor components - a review

    International Nuclear Information System (INIS)

    Krajcinovic, D.

    1977-01-01

    A variety of components belonging to a nuclear reactor are by virtue of their design exposed to a mass of fluid which is either in motion or can be set into motion under certain conditions. While the reactor is in its operational mode, the excitations of the structure by the fluid are generally of moderate intensities. In the case of a well designed component, these pressure fluctuations should not cause the failure of the structure. Problems of this type, generally known as vibrations of structures immersed into fluid (under either periodic or random excitations) have been studied in the past rather extensively. In an upset or emergency condition, a pressure pulse is usually generated and propagated through the fluid. While this hypothetical event is an occurrence of low probability the associated pressures are, as a rule, of intensities sufficiently large to cause extensive damage or even the failure of the component. This type of transient interaction problem is much less studied and the aim of this review is to offer a brief discussion of some of the more interesting results. (Auth.)

  7. Decommissioning of nuclear power plants

    International Nuclear Information System (INIS)

    Friske, A.; Thiele, D.

    1988-01-01

    The IAEA classification of decommissioning stages is outlined. The international development hitherto observed in decommissioning of nuclear reactors and nuclear power stations is presented. The dismantling, cutting and decontamination methods used in the decommissioning process are mentioned. The radioactive wastes from decommissioning are characterized, the state of the art of their treatment and disposal is given. The radiation burdens and the decommissioning cost in a decommissioning process are estimated. Finally, some evaluation of the trends in the decommissioning process of nuclear power plants is given. 54 refs. (author)

  8. Decommissioning of eight surplus production reactors at the Hanford Site, Richland, Washington. Addendum (Final Environmental Impact Statement)

    Energy Technology Data Exchange (ETDEWEB)

    1992-12-01

    The first section of this volume summarizes the content of the draft environmental impact statement (DEIS) and this Addendum, which together constitute the final environmental impact statement (FEIS) prepared on the decommissioning of eight surplus plutonium production reactors at Hanford. The FEIS consists of two volumes. The first volume is the DEIS as written. The second volume (this Addendum) consists of a summary; Chapter 9, which contains comments on the DEIS and provides DOE`s responses to the comments; Appendix F, which provides additional health effects information; Appendix K, which contains costs of decommissioning in 1990 dollars; Appendix L, which contains additional graphite leaching data; Appendix M, which contains a discussion of accident scenarios; Appendix N, which contains errata; and Appendix 0, which contains reproductions of the letters, transcripts, and exhibits that constitute the record for the public comment period.

  9. Selection of hardfacing material for components of the Indian Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Bhaduri, A.K.; Indira, R.; Albert, S.K.; Rao, B.P.S.; Jain, S.C.; Asokkumar, S.

    2004-01-01

    Nickel-base hardfacing alloys have been chosen to replace cobalt-base alloys as hardfacing material for components of the Indian Prototype Fast Breeder Reactor, for minimising the dose rate to personnel during maintenance and decommissioning, and to reduce the shielding thickness required for component handling. Induced activity, dose rate and shielding computations showed that replacing cobalt-base alloys with nickel-base alloys for hardfacing of components would result in a marked reduction in both the dose rate from the components and the thickness of lead handling flasks. Long-term ageing studies on the nickel-base hardface deposits on austenitic stainless steel showed that the hardface deposit would retain adequate hardness at the end of the components' design service-life of 40 years of exposure at 823 K

  10. In-core power sharing and fuel requirement study for a decommissioning Boiling Water Reactor using the linear reactivity model

    International Nuclear Information System (INIS)

    Chen, Chung-Yuan; Tung, Wu-Hsiung; Yaur, Shung-Jung; Kuo, Weng-Sheng

    2014-01-01

    Highlights: • Linear reactivity model (LRM) was modified and applied to Boiling Water Reactor. • The power sharing and fuel requirement study of the last cycle and two cycles before decommissioning was implemented. • The loading pattern design concept for the cycles before decommissioning is carried out. - Abstract: A study of in-core power sharing and fuel requirement for a decommissioning BWR (Boiling Water Reactor) was carried out using the linear reactivity model (LRM). The power sharing of each fuel batch was taken as an independent variable, and the related parameters were set and modified to simulate actual cases. Optimizations of the last cycle and two cycles before decommissioning were both implemented; in the last-one-cycle optimization, a single cycle optimization was carried out with different upper limits of fuel batch power, whereas, in the two-cycle optimization, two cycles were optimized with different cycle lengths, along with two different optimization approaches which are the simultaneous optimization of two cycles (MO) and two successive single-cycle optimizations (SO). The results of the last-one-cycle optimization show that it is better to increase the fresh fuel power and decrease the thrice-burnt fuel power as much as possible. It also shows that relaxing the power limit is good to the fresh fuel requirement which will be reduced under lower power limit. On the other hand, the results of the last-two-cycle (cycle N-1 and N) optimization show that the MO is better than SO, and the power of fresh fuel batch should be decreased in cycle N-1 to save its energy for the next cycle. The results of the single-cycle optimization are found to be the same as that in cycle N of the multi-cycle optimization. Besides that, under the same total energy requirement of two cycles, a long-short distribution of cycle length design can save more fresh fuel

  11. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 2. Appendices. Technical report, September 1977-October 1979

    International Nuclear Information System (INIS)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE. This volume contains the appendices

  12. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 2. Appendices. Technical report, September 1977-October 1979

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE. This volume contains the appendices.

  13. Methodology for assessing suitable systems for management of reactor decommissioning wastes

    International Nuclear Information System (INIS)

    Davis, J.P.; Barraclough, I.M.; Mobbs, S.F.

    1990-01-01

    This report demonstrates a methodology for comparing quantitatively the options open to decision-makers at various stages of the decommissioning process, taking into account radiological protection and other factors considered to be relevant. In order to compare different decommissioning options, various impacts associated with decommissioning operations and waste disposal were assessed, namely, costs and radiological, environmental and socio-political impact. The post-disposal radiological impact was calculated for four generic concepts (near surface, deep geological, coastal tunnel and deep ocean bed)

  14. The optimum shielding for a power reactor using local components

    International Nuclear Information System (INIS)

    AlHajali, S.; Kharita, M. H.; Yousef, S.; Naoom, B.; Al-Nassar, M.

    2009-07-01

    Some local concrete mixtures have been picked out (selected) to be studied as shielding concrete for prospective nuclear power reactor in Syria. This research has interested in the attenuation of gamma radiation and neutron fluxes by these local concretes in the ordinary conditions. In addition to the heat effect on the shielding and physical properties of local concrete. Furthermore the neutron activation of the elements of the local concrete mixtures have been studied that for selection the low-activation materials (low dose rate and short half life radioisotopes). In this way biological shielding for nuclear reactor can be safe during operation of nuclear power reactor, in addition to be low radioactive waste after decommissioning the reactor. (author)

  15. Shippingport Station Decommissioning Project (SSDP). A progress report

    International Nuclear Information System (INIS)

    Mullee, G.R.; Usher, J.M.

    1986-01-01

    The Shippingport Atomic Power Station was shutdown in October, 1982 by the Plant Operator, Duquesne Light Company, for decommissioning by the US Department of Energy. The planning for decommissioning was completed in September, 1983. In September, 1984 operational responsibility for the station was transferred to the DOE's Decommissioning Operations Contractor - the General Electric Company (assisted by an integrated subcontractor, MK Ferguson Company). Significant accomplishments to date include the completion of all prerequisites for decommissioning, the removal of asbestos from plant systems, loading of irradiated reactor components into the reactor vessel for shipment, the commencement of electrical deactivations and the commencement of piping/component removal. Decontamination and waste processing are progressing in support of the project schedule. The reactor vessel will be shipped as one piece on a barge for burial at Hanford, Washington. The final release of the site is scheduled for April, 1990. A technology transfer program is being utilized to disseminate information about the project

  16. Omega West Reactor program management and communication key to successful Decontamination and Decommissioning (D and D)

    Energy Technology Data Exchange (ETDEWEB)

    Mee, Stephen F.; Rendell, Keith R.; Peifer, Martin J. [Los Alamos National Laboratory, PO Box 1663, Los Alamos, NM 87545 (United States); Gallegos, John A. [National Nuclear Security Administration, P.O. Box 5400, Albuquerque, NM 87185 (United States); Straehr, James P.; Stringer, Joe B. [Framatome ANP, Tour AREVA, 92084 - Paris la Defense (France)

    2003-07-01

    This paper describes what differentiates the Omega West Reactor (OWR) Decommissioning and Decontamination (D and D) Project from other projects with similar scope and how the project was successfully completed ahead of schedule. With less than 26 months to scope, schedule, advertise, select a contractor and complete the actual D and D, Los Alamos National Laboratory (LANL) needed a new approach to form the foundation for the project's success and ensure that the project was completed on time and within the original contract value. This paper describes the three key elements of this new approach - including team building, strong project management and technical innovation. LANL and WD3, a joint venture between Framatome ANP, Inc. and Washington Group Inc., teamed through a fixed price best value contract to perform the D and D of the OWR. The project was initiated in an effort to reduce the risk to LANL facilities identified in the aftermath of the Cerro Grande fires. Between May 4 and June 10, 2000, a devastating wildfire swept across the Bandelier National Monument in the Jemez Mountains of northern New Mexico and onto the Department of Energy's (DOE's) LANL. The Cerro Grande fire burned about 43,000 acres, including 7,500 acres of LANL property. Large areas of vegetation in the Jemez Mountains surrounding LANL were destroyed. The DOE, LANL, other federal agencies, and the State of New Mexico initiated prompt actions to identify and mitigate the risks from the fire aftermath. Assessments conducted after the fire determined that serious environmental and safety problems associated with flash floods, erosion, and contaminant run-off would persist at LANL for a number of years. Since the OWR was located in a potential flash flood area it was decided to accelerate the D and D of the facility. (authors)

  17. Omega West Reactor program management and communication key to successful Decontamination and Decommissioning (D and D)

    International Nuclear Information System (INIS)

    Mee, Stephen F.; Rendell, Keith R.; Peifer, Martin J.; Gallegos, John A.; Straehr, James P.; Stringer, Joe B.

    2003-01-01

    This paper describes what differentiates the Omega West Reactor (OWR) Decommissioning and Decontamination (D and D) Project from other projects with similar scope and how the project was successfully completed ahead of schedule. With less than 26 months to scope, schedule, advertise, select a contractor and complete the actual D and D, Los Alamos National Laboratory (LANL) needed a new approach to form the foundation for the project's success and ensure that the project was completed on time and within the original contract value. This paper describes the three key elements of this new approach - including team building, strong project management and technical innovation. LANL and WD3, a joint venture between Framatome ANP, Inc. and Washington Group Inc., teamed through a fixed price best value contract to perform the D and D of the OWR. The project was initiated in an effort to reduce the risk to LANL facilities identified in the aftermath of the Cerro Grande fires. Between May 4 and June 10, 2000, a devastating wildfire swept across the Bandelier National Monument in the Jemez Mountains of northern New Mexico and onto the Department of Energy's (DOE's) LANL. The Cerro Grande fire burned about 43,000 acres, including 7,500 acres of LANL property. Large areas of vegetation in the Jemez Mountains surrounding LANL were destroyed. The DOE, LANL, other federal agencies, and the State of New Mexico initiated prompt actions to identify and mitigate the risks from the fire aftermath. Assessments conducted after the fire determined that serious environmental and safety problems associated with flash floods, erosion, and contaminant run-off would persist at LANL for a number of years. Since the OWR was located in a potential flash flood area it was decided to accelerate the D and D of the facility. (authors)

  18. Influence of accounting concepts and regulatory rules on the funding of power reactor decommissioning costs

    International Nuclear Information System (INIS)

    Ferguson, J.S.

    1985-01-01

    Under normal circumstances, an evaluation of nuclear plant decommissioning costs by an engineering analyst will not produce the same results as an evaluation by a financial analyst. These analysts should understand evaluations based on each other's bases to ensure that their evaluation techniques are appropriate for the circumstances. The intent of this discussion is to enhance that understanding by describing the accounting and regulatory framework that is applicable to the decommissioning costs of U.S. nuclear power plants, and by explaining why evaluations of decommissioning costs prepared by engineering analysts often look different from evaluations prepared by financial analysts. Of major importance are the financial implications of several methods of funding the decommissioning costs. Since many owners of nuclear plants are subject to revenue rate regulation, financial implications often translate directly to regulatory implications

  19. Work Breakdown Structure and Work Packages for Decommissioning the Nuclear Research Reactor VVR-S Magurele-Bucharest

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-06-15

    The research reactor type VVR-S (tank type, water cooled, moderator and reflector, thermal power 2 MW, thermal energy 9.52 GWd) was put into service in July 1957, and in December 1997, was shut down. In 2002, the Romanian Government decided to put the research reactor into a permanent shutdown condition in order to start decommissioning. This nuclear facility had been used in nuclear research and radioisotope production for 40 years without any events, incidents or accidents. At the same site, in the immediate vicinity of the research reactor, there are many other nuclear facilities: a radioactive waste treatment plant, a tandem Van de Graaff heavy ion accelerator, a cyclotron, an industrial irradiator and a radioisotope production centre.

  20. Engineering Evaluation/Cost Analysis for Decommissioning of the Engineering Test Reactor Complex

    Energy Technology Data Exchange (ETDEWEB)

    A. B. Culp

    2006-10-01

    Preparation of this Engineering Evaluation/Cost Analysis is consistent with the joint U.S. Department of Energy and U.S. Environmental Protection Agency Policy on Decommissioning of Department of Energy Facilities Under the Comprehensive Environmental Response, Compensation, and Liability Act, which establishes the Comprehensive Environmental Response, Compensation, and Liability Act non-time-critical removal action (NTCRA) process as an approach for decommissioning.

  1. Influence of design features on decommissioning of a large fast breeder reactor

    International Nuclear Information System (INIS)

    Fournie, J.L.; Alary, C.; Jean, M.; Maire, D.; Peyrard, G.

    1990-01-01

    In the conception and realization studies of a LMFBR, the decommissioning aspects are not much taken in account. It appears that low cost and unsophisticated dispositions can notably facilitate these operations of decommissioning. The objective of this research consists in the identification of the conception and construction measures that can facilitate these operations. By priority we considerate the measures requiring low cost and low development study [fr

  2. Engineering Evaluation/Cost Analysis for Decommissioning of the Engineering Test Reactor Complex

    International Nuclear Information System (INIS)

    A. B. Culp

    2006-01-01

    Preparation of this Engineering Evaluation/Cost Analysis is consistent with the joint U.S. Department of Energy and U.S. Environmental Protection Agency Policy on Decommissioning of Department of Energy Facilities Under the Comprehensive Environmental Response, Compensation, and Liability Act, which establishes the Comprehensive Environmental Response, Compensation, and Liability Act non-time-critical removal action (NTCRA) process as an approach for decommissioning

  3. CESAR5.3: Isotopic depletion for Research and Testing Reactor decommissioning

    Science.gov (United States)

    Ritter, Guillaume; Eschbach, Romain; Girieud, Richard; Soulard, Maxime

    2018-05-01

    , CESAR includes a portable Graphical User Interface which can be broadly deployed in R&D or industrial facilities. Aging facilities currently face decommissioning and dismantling issues. This way to the end of the nuclear fuel cycle requires a careful assessment of source terms in the fuel, core structures and all parts of a facility that must be disposed of with "industrial nuclear" constraints. In that perspective, several CESAR cross section libraries were constructed for early CEA Research and Testing Reactors (RTR's). The aim of this paper is to describe how CESAR operates and how it can be used to help these facilities care for waste disposal, nuclear materials transport or basic safety cases. The test case will be based on the PHEBUS Facility located at CEA - Cadarache.

  4. Decommissioning nuclear facilities

    International Nuclear Information System (INIS)

    Harmon, K.M.; Jenkins, C.E.; Waite, D.A.; Brooksbank, R.E.; Lunis, B.C.; Nemec, J.F.

    1976-01-01

    This paper describes the currently accepted alternatives for decommissioning retired light water reactor fuel cycle facilities and the current state of decommissioning technology. Three alternatives are recognized: Protective Storage; Entombment; and Dismantling. Application of these alternatives to the following types of facilities is briefly described: light water reactors; fuel reprocessing plants, and mixed oxide fuel fabrication plants. Brief descriptions are given of decommissioning operations and results at a number of sites, and recent studies of the future decommissioning of prototype fuel cycle facilities are reviewed. An overview is provided of the types of operations performed and tools used in common decontamination and decommissioning techniques and needs for improved technology are suggested. Planning for decommissioning a nuclear facility is dependent upon the maximum permitted levels of residual radioactive contamination. Proposed guides and recently developed methodology for development of site release criteria are reviewed. 21 fig, 32 references

  5. Seismic behaviour of gas cooled reactor components

    International Nuclear Information System (INIS)

    1990-08-01

    On invitation of the French Government the Specialists' Meeting on the Seismic Behaviour of Gas-Cooled Reactor Components was held at Gif-sur-Yvette, 14-16 November 1989. This was the second Specialists' Meeting on the general subject of gas-cooled reactor seismic design. There were 27 participants from France, the Federal Republic of Germany, Israel, Japan, Spain, Switzerland, the United Kingdom, the Soviet Union, the United States, the CEC and IAEA took the opportunity to present and discuss a total of 16 papers reflecting the state of the art of gained experiences in the field of their seismic qualification approach, seismic analysis methods and of the capabilities of various facilities used to qualify components and verify analytical methods. Since the first meeting, the sophistication and expanded capabilities of both the seismic analytical methods and the test facilities are apparent. The two main methods for seismic analysis, the impedance method and the finite element method, have been computer-programmed in several countries with the capability of each of the codes dependent on the computer capability. The correlations between calculation and tests are dependent on input assumptions such as boundary conditions, soil parameters and various interactions between the soil, the buildings and the contained equipment. The ability to adjust these parameters and match experimental results with calculations was displayed in several of the papers. The expanded capability of some of the new test facilities was graphically displayed by the description of the SAMSON vibration test facility at Juelich, FRG, capable of dynamically testing specimens weighing up to 25 tonnes, and the TAMARIS facility at the CEA laboratories in Gif-sur-Yvette where the largest table is capable of testing specimens weighing up to 100 tonnes. The proceedings of this meeting contain all 16 presented papers. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  6. Establishing the user requirements for the research reactor decommissioning database system

    International Nuclear Information System (INIS)

    Park, S. K.; Park, H. S.; Lee, G. W.; Park, J. H.

    2002-01-01

    In generally, so much information and data will be raised during the decommissioning activities. It is need a systematical electric system for the management of that. A database system for the decommissioning information and data management from the KRR-1 and 2 decommissioning project is developing now. All information and data will be put into this database system and retrieval also. For the developing the DB system, the basic concept, user requirements were established the then set up the system for categorizing the information and data. The entities of tables for input the data was raised and categorized and then converted the code. The ERD (Entity Relation Diagram) was also set up to show their relation. In need of the developing the user interface system for retrieval the data, is should be studied the analyzing on the relation between the input and output the data. Through this study, as results, the items of output tables are established and categorized according to the requirement of the user interface system for the decommissioning information and data. These tables will be used for designing the prototype and be set up by several feeds back for establishing the decommissioning database system

  7. Decommissioning plan - decommissioning project for KRR 1 and 2 (revised)

    International Nuclear Information System (INIS)

    Jung, K. J.; Paik, S. T.; Chung, U. S.; Jung, K. H.; Park, S. K.; Lee, D. G.; Kim, H. R.; Kim, J. K.; Yang, S. H.; Lee, B. J.

    2000-10-01

    This report is the revised Decommissioning Plan for the license of TRIGA research reactor decommissioning project according to Atomic Energy Act No. 31 and No. 36. The decommissioning plan includes the TRIGA reactor facilities, project management, decommissioning method, decontamination and dismantling activity, treatment, packaging, transportation and disposal of radioactive wastes. the report also explained the radiation protection plan and radiation safety management during the decommissioning period, and expressed the quality assurance system during the period and the site restoration after decommissioning. The first decommissioning plan was made by Hyundai Engineering Co, who is the design service company, was submitted to the Ministry of Science and Technology, and then was reviewed by the Korea Institute of Nuclear Safety. The first decommissioning plan was revised including answers for the questions arising from review process

  8. Decommissioning plan - decommissioning project for KRR 1 and 2 (revised)

    Energy Technology Data Exchange (ETDEWEB)

    Jung, K. J.; Paik, S. T.; Chung, U. S.; Jung, K. H.; Park, S. K.; Lee, D. G.; Kim, H. R.; Kim, J. K.; Yang, S. H.; Lee, B. J

    2000-10-01

    This report is the revised Decommissioning Plan for the license of TRIGA research reactor decommissioning project according to Atomic Energy Act No. 31 and No. 36. The decommissioning plan includes the TRIGA reactor facilities, project management, decommissioning method, decontamination and dismantling activity, treatment, packaging, transportation and disposal of radioactive wastes. the report also explained the radiation protection plan and radiation safety management during the decommissioning period, and expressed the quality assurance system during the period and the site restoration after decommissioning. The first decommissioning plan was made by Hyundai Engineering Co, who is the design service company, was submitted to the Ministry of Science and Technology, and then was reviewed by the Korea Institute of Nuclear Safety. The first decommissioning plan was revised including answers for the questions arising from review process.

  9. Creep buckling problems in fast reactor components

    International Nuclear Information System (INIS)

    Ramesh, R.; Damodaran, S.P.; Chellapandi, P.; Chetal, S.C.; Bhoje, S.B.

    1995-01-01

    Creep buckling analyses for two important components of 500 M We Prototype Fast Breeder Reactor (PFBR), viz. Intermediate Heat Exchanger (IHX) and Inner Vessel (IV), are reported. The INCA code of CASTEM system is used for the large displacement elasto-plastic-creep analysis of IHX shell. As a first step, INCA is validated for a typical benchmark problem dealing with the creep buckling of a tube under external pressure. Prediction of INCA is also compared with the results obtained using Hoff's theory. For IV, considering the prohibitively high computational cost for the actual analysis, a simplified analysis which involves only large displacement elastoplastic buckling analysis is performed using isochronous stress strain curve approach. From both of these analysis is performed using isochronous stress strain curve approach. From both of these analysis, it has been inferred that creep buckling failure mode is not of great concern in the design of PFBR components. It has also been concluded from the analysis that Creep Cross Over Curve given in RCC-MR is applicable for creep buckling failure mode also. (author). 8 refs., 9 figs., 1 tab

  10. Feedback experience from the decommissioning of Spanish nuclear facilities

    International Nuclear Information System (INIS)

    Santiago, J.L.

    2008-01-01

    The Spain has accumulated significant experience in the field of decommissioning of nuclear and radioactive facilities. Relevant projects include the remediation of uranium mills and mines, the decommissioning of research reactors and nuclear research facilities and the decommissioning of gas-graphite nuclear power plants. The decommissioning of nuclear facilities in Spain is undertaken by ENRESA, who is also responsible for the management of radioactive wastes. The two most notable projects are the decommissioning of the Vandellos I nuclear power plant and the decommissioning of the CIEMAT nuclear research centre. The Vandellos I power plant was decommissioned in about five years to what is known as level 2. During this period, the reactor vessel was confined, most plant systems and components were dismantled, the facility was prepared for a period of latency and a large part of the site was restored for subsequent release. In 2005 the facility entered into the phase of dormancy, with minimum operating requirements. Only surveillance and maintenance activities are performed, among which special mention should be made to the five-year check of the leak tightness of the reactor vessel. After the dormancy period (25 - 30 years), level 3 of decommissioning will be initiated including the total dismantling of the remaining parts of the plant and the release of the whole site for subsequent uses. The decommissioning of the CIEMAT Research Centre includes the dismantling of obsolete facilities such as the research reactor JEN-1, a pilot reprocessing plant, a fuel fabrication facility, a conditioning plant for liquid and a liquid waste storage facility which were shutdown in the early eighties. Dismantling works have started in 2006 and will be completed by 2009. On the basis of the experience gained in the above mentioned sites, this paper describes the approaches adopted by ENRESA for large decommissioning projects. (author)

  11. Check experiment of the high pressure water washing technology used to the decommissioning of reactor

    International Nuclear Information System (INIS)

    Han Jianping; Hou Yongming; Fu Yunshan

    2004-01-01

    High pressure water washing technology has been widely applied in the field of the decommissioning of nuclear facilities, and it is used to wash the sump for craft conveyance, the craft workshop, the hermetic sump, and some other nuclear equipment as well. The authors have got a set of technical data correlated with high pressure water washing technology by comparing the situations between the test before and after the washing work. At the same time, authors also improve the technique on some special cases, which made the high pressure water washing technology more perfect in the field of the decommissioning of nuclear facilities. (authors)

  12. Decommissioning of naval nuclear ships

    International Nuclear Information System (INIS)

    Oelgaard, P.L.

    1993-10-01

    During the next decade the two major nuclear powers will each have to decommission more than 100 naval nuclear vessels, in particular submarines. The problems connected with this task is considered in this report. Firstly the size of the task is considered, i.e. the number of nuclear vessels that has to be decommissioned. Secondly the reactors of these vessels, their fuel elements, their power level, the number of reactors per vessel and the amount of radioactivity to be handled are discussed. Thirdly the decommissioning procedures, i.e. The removal of fuel from the vessels, the temporary storage of the reactor fuel near the base, and the cleaning and disposal of the reactor and the primary circuit components are reviewed. Finally alternative uses of the newer submarines are briefly considered. It should be emphasizes that much of the detailed information on which this report is based, may be of dubious nature, and that may to some extent affect the validity of the conclusions of the report. (au)

  13. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

    Bierschbach, M.C.

    1997-01-01

    1 - Description of program or function: The Cost Estimating Computer Program CECP, designed for use on an IBM personal computer or equivalent, was developed for estimating the cost of decommissioning boiling water reactor (BWR) and light-water reactor (PWR) power stations to the point of license termination. 2 - Method of solution: Cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial volume and costs; and manpower staffing costs. Using equipment and consumables costs and inventory data supplied by the user, CECP calculates unit cost factors and then combines these factors with transportation and burial cost algorithms to produce a complete report of decommissioning costs. In addition to costs, CECP also calculates person-hours, crew-hours, and exposure person-hours associated with decommissioning. 3 - Restrictions on the complexity of the problem: The program is designed for a specific waste charge structure. The waste cost data structure cannot handle intermediate waste handlers or changes in the charge rate structures. The decommissioning of a reactor can be divided into 5 periods. 200 different items for special equipment costs are possible. The maximum amount for each special equipment item is 99,999,999$. You can support data for 10 buildings, 100 components each; ESTS1071/01: There are 65 components for 28 systems available to specify the contaminated systems costs (BWR). ESTS1071/02: There are 75 components for 25 systems available to specify the contaminated systems costs (PWR)

  14. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR PART 5 - FUEL CHANEL DECOMMISSIONING

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2014-05-01

    Full Text Available As many nuclear power plants are reaching their end of lifecycle, the decommissioning of these installations has become one of the 21st century’s great challenges. Each project may be managed differently, depending on the country, development policies, financial considerations, and the availability of qualified engineers or specialized companies to handle such projects. The principle objective of decommissioning is to place a facility into such a condition that there is no unacceptable risk from the decommissioned facility to public health and safety of the environment. In order to ensure that at the end of its life the risk from a facility is within acceptable bounds, action is normally required. The overall decommissioning strategy is to deliver a timely, cost-effective program while maintaining high standards of safety, security and environmental protection. If facilities were not decommissioned, they could degrade and potentially present an environmental radiological hazard in the future. Simply abandoning or leaving a facility after ceasing operations is not considered to be an acceptable alternative to decommissioning. The final aim of decommissioning is to recover the geographic site to its original condition.

  15. Revised analyses of decommissioning for the reference pressurized Water Reactor Power Station. Volume 2, Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure: Appendices, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.; Smith, R.I.; Bierschbach, M.C.; McDuffie, P.N.

    1995-11-01

    With the issuance of the final Decommissioning Rule (July 27, 1998), owners and operators of licensed nuclear power plants are required to prepare, and submit to the US Nuclear Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. The NRC staff is in need of bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to provide some of the needed bases documentation. This report contains the results of a review and reevaluation of the 1978 PNL decommissioning study of the Trojan nuclear power plant (NUREG/CR-0130), including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the nuclear power plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives. These alternatives now include an initial 5--7 year period during which time the spent fuel is stored in the spent fuel pool, prior to beginning major disassembly or extended safe storage of the plant. Included for information (but not presently part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clean structures on the site and to restore the site to a ``green field`` condition. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low-level waste (i.e., Greater-Than-Class C), and reflects 1993 costs for labor, materials, transport, and disposal activities.

  16. Revised analyses of decommissioning for the reference pressurized Water Reactor Power Station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure, Volume 1, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.; Smith, R.I.; Bierschbach, M.C.; McDuffie, P.N. [Pacific Northwest Lab., Richland, WA (United States)

    1995-11-01

    With the issuance of the final Decommissioning Rule (July 27, 1988), owners and operators of licensed nuclear power plants are required to prepare, and submit to the US Nuclear Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. The NRC staff is in need of bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to provide some of the needed bases documentation. This report contains the results of a review and reevaluation of the {prime}978 PNL decommissioning study of the Trojan nuclear power plant (NUREG/CR-0130), including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the nuclear power plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives. These alternatives now include an initial 5--7 year period during which time the spent fuel is stored in the spent fuel pool, prior to beginning major disassembly or extended safe storage of the plant. Included for information (but not presently part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clean structures on the site and to restore the site to a ``green field`` condition. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low-level waste (i.e., Greater-Than-Class C), and reflects 1993 costs for labor, materials, transport, and disposal activities.

  17. Revised analyses of decommissioning for the reference pressurized Water Reactor Power Station. Volume 2, Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure: Appendices, Final report

    International Nuclear Information System (INIS)

    Konzek, G.J.; Smith, R.I.; Bierschbach, M.C.; McDuffie, P.N.

    1995-11-01

    With the issuance of the final Decommissioning Rule (July 27, 1998), owners and operators of licensed nuclear power plants are required to prepare, and submit to the US Nuclear Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. The NRC staff is in need of bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to provide some of the needed bases documentation. This report contains the results of a review and reevaluation of the 1978 PNL decommissioning study of the Trojan nuclear power plant (NUREG/CR-0130), including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the nuclear power plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives. These alternatives now include an initial 5--7 year period during which time the spent fuel is stored in the spent fuel pool, prior to beginning major disassembly or extended safe storage of the plant. Included for information (but not presently part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clean structures on the site and to restore the site to a ''green field'' condition. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low-level waste (i.e., Greater-Than-Class C), and reflects 1993 costs for labor, materials, transport, and disposal activities

  18. Revised analyses of decommissioning for the reference pressurized Water Reactor Power Station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure, Volume 1, Final report

    International Nuclear Information System (INIS)

    Konzek, G.J.; Smith, R.I.; Bierschbach, M.C.; McDuffie, P.N.

    1995-11-01

    With the issuance of the final Decommissioning Rule (July 27, 1988), owners and operators of licensed nuclear power plants are required to prepare, and submit to the US Nuclear Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. The NRC staff is in need of bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to provide some of the needed bases documentation. This report contains the results of a review and reevaluation of the '978 PNL decommissioning study of the Trojan nuclear power plant (NUREG/CR-0130), including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the nuclear power plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives. These alternatives now include an initial 5--7 year period during which time the spent fuel is stored in the spent fuel pool, prior to beginning major disassembly or extended safe storage of the plant. Included for information (but not presently part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clean structures on the site and to restore the site to a ''green field'' condition. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low-level waste (i.e., Greater-Than-Class C), and reflects 1993 costs for labor, materials, transport, and disposal activities

  19. Alternate form and placement of short lived reactor waste and associated fuel hardware for decommissioning of EBR-II

    Energy Technology Data Exchange (ETDEWEB)

    Planchon, H.P.; Singleterry, R.C. Jr.

    1995-12-01

    Upon the termination of EBR-II operation in 1994, the mission has progressed to decommissioning and waste cleanup of the facility. The simplest method to achieve this goal is to bury the raw fuel and activated steel in an approved burial ground or deep geologic repository. While this might be simple, it could be very expensive, consume much needed burial space for other materials, and leave large amounts of fissile easily available to future generations. Also, as with any operation, an associated risk to personnel and the public from the buried waste exists. To try and reduce these costs and risks, alternatives to burial are sought. One alternative explored here for EBR-II is to condition the fuel and store the fission products and steel either permanently or temporarily in the sealed primary boundary of the decommissioned reactor. The first problem is to identify which subassemblies are going to be conditioned and their current composition and decay time. The next problem is to identify the conditioning process and determine the composition and form of the waste streams. The volume, mass, heat, and curie load of the waste streams needs to be determined so a waste-assembly can be designed. The reactor vessel and internals need to be analyzed to determine if they can handle these loads. If permanent storage is the goal, then mechanisms for placing the waste-assembly in the reactor vessel and sealing the vessel are needed. If temporary storage is the goal, then mechanisms for waste-assembly placement and retrieval are needed. This paper answers the technical questions of volume, mass, heat, and curie loads while just addressing the other questions found in a safety analysis. The final conclusion will compare estimated risks from the burial option and this option.

  20. Canisters and nonfuel components at commercial nuclear reactors

    International Nuclear Information System (INIS)

    Gibbard, K.; Thorpe, J.; Moore, R.S.

    1995-01-01

    The Energy Information Administration of the U.S. Department of Energy (DOE) collects data annually from commercial nuclear power reactors via the Nuclear Fuel Data survey, Form RW-859. Over the past three years, the survey has collected data on the quantities and types of nonfuel components and on the quantities and contents of canisters in storage at reactor sites. This paper focuses on the annual changes in the data, specific implications of these changes, and lessons that should be applied to future revisions of the study. The total number of canisters reported by utilities for each year from 1986 to 1993 is listed. Changes in the quantities of nonfuel components report by General Reactors from 1992 to 1993 are also provided. Comparisons of canister and nonfuel components components data from year to year and from reactor to reactor point out that survey questions on these topics have been interpreted differently by reactor personnel

  1. Aging of metal components in US nuclear reactors

    International Nuclear Information System (INIS)

    Mayfield, M.E.; Strosnider, J.R.

    1998-01-01

    This paper presents an overview of the aging of metal components in U.S. Light Water Reactors. The types of degradation being experienced in components such as the pressure vessel, piping, reactor internals, and steam generators, and the programs being implemented to manage the degradation are discussed. (author)

  2. The possibility of creating a new low power nuclear facility with slightly enriched nuclear fuel on the basis of the decommissioned IRT-M reactor intended for applied purposes

    International Nuclear Information System (INIS)

    Abramidze, Sh.P.; Katamadze, N.M.; Kiknadze, G.G.; Rostomashvili, Z.I.; Saralidze, Z.K.

    2002-01-01

    Nearly 50 years have passed since the appearance of the first nuclear research reactors. Most of them have completed their operating life and must be dismantled. But it is known that the dismantling of permanently shut down nuclear reactors is a very complex process, full realization that it generates a lot of radioactive waste (both solid and liquid), it is connected with high financial expenditures, and its solution is apparently beyond the possibilities of many countries, including Georgia In the given paper we consider a radiologically safe, ecologically clean and economically beneficial version of the decommissioning of the IRT-M nuclear research reactor and the stages of its implementation that are not connected with the dismantling of its highly radioactive technological components. We justify the possibility of creating a new Low Power Nuclear Facility on the basis of the decommissioned IRT-M reactor to solve the problems of applied nature in different fields of science and technology being very important for Georgia. (author)

  3. Summary record of the topical session at WPDD-10: Management of large components from decommissioning to storage and disposal, 18-19 November 2009

    International Nuclear Information System (INIS)

    Dutzer, Michel

    2010-01-01

    At its tenth meeting, the WPDD held a topical session on Management of Large Components from Decommissioning to Storage and Disposal. The topical session was organised by a new task group of the WPDD that recently began work on this topic. The group is aiming to prepare a technical guide that provides a methodology to assess different management options and facilitates involvement of the different interested parties in the process of selecting the preferred management option. This report is made of 3 parts: Part 1 presents the Main Outcomes of Topical Session on Management of Large Components from Decommissioning to Storage and Disposal (Summary of Presentations and Discussions and Rapporteurs Report); Part 2 presents the Agenda of the Topical Session on Management of Large Components from Decommissioning to Storage and Disposal; and Part 3 is the List of Participants

  4. Overview moderator material for nuclear reactor components

    International Nuclear Information System (INIS)

    Mairing Manutu Pongtuluran; Hendra Prihatnadi

    2009-01-01

    In order for a reactor design is considered acceptable absolute technical requirement is fulfilled because the most important part of a reactor design. Safety considerations emphasis on the handling of radioactive substances emitted during the operation of a reactor and radioactive waste handling. Moderator material is a layer that interacts directly with neutrons split the nuclear fuel that will lead to changes in physical properties, nuclear properties, mechanical properties and chemical properties. Reviews moderator of this time is of the types of moderator is often used to meet the requirements as nuclear material. (author)

  5. Estimating boiling water reactor decommissioning costs. A user's manual for the BWR Cost Estimating Computer Program (CECP) software: Draft report for comment

    International Nuclear Information System (INIS)

    Bierschbach, M.C.

    1994-12-01

    With the issuance of the Decommissioning Rule (July 27, 1988), nuclear power plant licensees are required to submit to the U.S. Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. This user's manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning BWR power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning

  6. State of decommissioning process in Romania

    International Nuclear Information System (INIS)

    Ciuculescu, C.

    2002-01-01

    In Romania, there are several installations that arrived at the decommissioning stage. These installations are: VVR-S research reactor, Sub critical Assembly HELEN, and Zero Power Reactor (RP-0). In this paper, the methods the Romanian Regulatory Body is developing the legal framework for decommissioning process of nuclear installations are described. There is a draft of decommissioning norms for research reactors. This regulation provides each stage of decommissioning and requirements for decommissioning plan. Also, CNCAN has evaluated and made requirements for completion of a VVR-S research reactor decommissioning plan submitted by IFIN-HH. Further, the reasons for which the decommissioning plan was rejected and requirements that the owner of VVR-S research reactor must fulfil in order to receive decommissioning licence are presented. (author)

  7. Management of solid wastes during decommissioning of research reactors. Evaluation of gross clearance levels and mathematical simulation of solid waste assay techniques

    International Nuclear Information System (INIS)

    Gopalakrishnan, R.K.; Sobhan Babu, K.; Sharma, D.N.

    2008-01-01

    Full text: Decommissioning of nuclear facilities constitute a challenge mainly due to the huge and complex nature of radioactive waste generated during this process. In the context of management and disposal of waste and reuse/recycle of usable materials during decommissioning of reactors, clearance levels for relevant radio nuclides are of vital importance. Radionuclide specific clearance levels are developed by IAEA and such levels allow the facility for free release of materials to the environment without further regulatory consideration. An effort has been made in this paper to establish clearance levels for radionuclides associated with various system and structural components of a research reactor and rather than radionuclide specific clearance levels, these values are derived for gross activity concentration, which is more practical for radioactive waste categorization, disposal and reuse or recycle of usable materials. The first step towards the derivation of clearance levels is the calculation of annual doses relating to unit activity concentration for each nuclide using various enveloping scenarios. After the estimation of doses, the limiting enveloping scenario (the one that gives the highest dose) is identified. The clearance levels are then derived by dividing the reference dose level (10 μSv/y) by the annual dose calculated per unit activity concentration for the limiting enveloping scenario The clearance level for gross beta-gamma activity concentration is then evaluated as the product of the limiting clearance level and the number of radionuclides characterized for the structural components. Simulation studies were also carried out for the design of a monitoring system for estimation of activity concentration of the decommissioned materials, especially rubbles/ concrete, using mathematical models. Conventional solid waste assay techniques would not suffice to the requirement of decommissioning waste categorization since very low level activity

  8. Decommissioning of the research nuclear reactor IRT-M and problems connected with radioactive waste

    International Nuclear Information System (INIS)

    Abramidze, S.P.; Katamadze, N.M.; Kiknadze, G.G.; Saralidze, Z.K.

    2000-01-01

    The nuclear research reactor IRT-2000 is described, along with modifications and upgrades made over the past three decades. Considerations are outlined which followed a decision to shut-down the reactor and to dismantle it. (author)

  9. The decommissioning of a research reactor in the USA: A case study from planning to site release

    International Nuclear Information System (INIS)

    Boing, L.E.

    1997-01-01

    Argonne National Laboratory (ANL) has completed the D ampersand D of the Experimental Boiling Water Reactor (EBWR). The project consisted of the decontamination and/or packaging as radioactive waste the reactor vessel and internals, contaminated piping systems, miscellaneous tanks, pumps, and associated equipment. The dismantling process involved the removal and size reduction of equipment and associated plumbing, ductwork, drain lines, etc. Size reduction of reactor vessel internals was performed in the fuel pool. All radioactive and mixed waste was packaged and manifested. A thorough survey of the facility was performed after the removal of contaminated and activated material. Non-radioactive waste was disposed of in the ANL landfill or recycled as appropriate. The EBWR D ampersand D project was divided into four phases. Phases I and II were completed by ANL personnel, while Phases III and IV were done by a contractor under ANL management. The final survey was performed by the contractor, while the verification survey was performed by ANL. The project lasted 118 months. Phase I was initiated in April 1986 and the final report was issued February 1996. The duration of the project was driven by the availability of funding for decommissioning. Total exposure to project personnel was 208.7 person-mSv (20.87 person-rem), with no personnel exceeding the EBWR project dose limit of 15 mSv (1.5 rem)

  10. An evaluation of cost estimates of nuclear power reactor decommissioning in Sweden, Germany and the United States

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, S O; Varley, G; Heibel, R; Rusch, C [NAC International, Zurich (Switzerland)

    1995-11-01

    Nominal base decommissioning cost estimates in Sweden, Germany and the US differ by large amounts. Even after adjustments to normalize the work scopes, significant cost differences remain. Variations in national cost structures, achievable productivity, the extent of preexisting infrastructure and institutional factors all contribute to make up the differences. Exchange rate aberrations are a complication for which appropriate adjustments have to be made in order to achieve a meaningful comparison. Our analyses demonstrate that virtually all these differences between the Swedish, German and US estimates can be explained by these factors. In terms of the overall reasonableness of the Swedish estimate as a basis for making financial provisions, there remain some issues that may warrant further investigation. One is the potential for and financial consequences of a serious interruption to the proposed sea transportation system. Secondly, the limited number of individual system analyses we have performed indicated some significant potential underestimates. For example, dismantling of the reactor pressure vessel costs appear to be underestimated by up to 70 MSEK (about 10 MUSD) per reactor, or up to 900 MSEK for the whole Swedish program of 12 reactors. Overall, the Swedish estimates appear to be built up in a logical and reasonable way. Our analyses indicate that some internal inconsistencies exist and that some specific input data assumptions may not be valid. In summary, the credibility of the estimates would benefit from further refinement of the scenarios and assumptions. 21 refs., 15 figs., 42 tabs.

  11. An evaluation of cost estimates of nuclear power reactor decommissioning in Sweden, Germany and the United States

    International Nuclear Information System (INIS)

    Andersson, S.O.; Varley, G.; Heibel, R.; Rusch, C.

    1995-11-01

    Nominal base decommissioning cost estimates in Sweden, Germany and the US differ by large amounts. Even after adjustments to normalize the work scopes, significant cost differences remain. Variations in national cost structures, achievable productivity, the extent of preexisting infrastructure and institutional factors all contribute to make up the differences. Exchange rate aberrations are a complication for which appropriate adjustments have to be made in order to achieve a meaningful comparison. Our analyses demonstrate that virtually all these differences between the Swedish, German and US estimates can be explained by these factors. In terms of the overall reasonableness of the Swedish estimate as a basis for making financial provisions, there remain some issues that may warrant further investigation. One is the potential for and financial consequences of a serious interruption to the proposed sea transportation system. Secondly, the limited number of individual system analyses we have performed indicated some significant potential underestimates. For example, dismantling of the reactor pressure vessel costs appear to be underestimated by up to 70 MSEK (about 10 MUSD) per reactor, or up to 900 MSEK for the whole Swedish program of 12 reactors. Overall, the Swedish estimates appear to be built up in a logical and reasonable way. Our analyses indicate that some internal inconsistencies exist and that some specific input data assumptions may not be valid. In summary, the credibility of the estimates would benefit from further refinement of the scenarios and assumptions. 21 refs., 15 figs., 42 tabs

  12. Fukushima Daiichi Unit 1 Accident Progression Uncertainty Analysis and Implications for Decommissioning of Fukushima Reactors - Volume I.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mattie, Patrick D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-01-01

    Sandia National Laboratories (SNL) has conducted an uncertainty analysis (UA) on the Fukushima Daiichi unit (1F1) accident progression with the MELCOR code. The model used was developed for a previous accident reconstruction investigation jointly sponsored by the US Department of Energy (DOE) and Nuclear Regulatory Commission (NRC). That study focused on reconstructing the accident progressions, as postulated by the limited plant data. This work was focused evaluation of uncertainty in core damage progression behavior and its effect on key figures-of-merit (e.g., hydrogen production, reactor damage state, fraction of intact fuel, vessel lower head failure). The primary intent of this study was to characterize the range of predicted damage states in the 1F1 reactor considering state of knowledge uncertainties associated with MELCOR modeling of core damage progression and to generate information that may be useful in informing the decommissioning activities that will be employed to defuel the damaged reactors at the Fukushima Daiichi Nuclear Power Plant. Additionally, core damage progression variability inherent in MELCOR modeling numerics is investigated.

  13. BLENDED CALCIUM ALUMINATE-CALCIUM SULFATE CEMENT-BASED GROUT FOR P-REACTOR VESSEL IN-SITU DECOMMISSIONING

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C.; Stefanko, D.

    2011-03-10

    The objective of this report is to document laboratory testing of blended calcium aluminate - calcium hemihydrate grouts for P-Reactor vessel in-situ decommissioning. Blended calcium aluminate - calcium hemihydrate cement-based grout was identified as candidate material for filling (physically stabilizing) the 105-P Reactor vessel (RV) because it is less alkaline than portland cement-based grout which has a pH greater than 12.4. In addition, blended calcium aluminate - calcium hemihydrate cement compositions can be formulated such that the primary cementitious phase is a stable crystalline material. A less alkaline material (pH {<=} 10.5) was desired to address a potential materials compatibility issue caused by corrosion of aluminum metal in highly alkaline environments such as that encountered in portland cement grouts [Wiersma, 2009a and b, Wiersma, 2010, and Serrato and Langton, 2010]. Information concerning access points into the P-Reactor vessel and amount of aluminum metal in the vessel is provided elsewhere [Griffin, 2010, Stefanko, 2009 and Wiersma, 2009 and 2010, Bobbitt, 2010, respectively]. Radiolysis calculations are also provided in a separate document [Reyes-Jimenez, 2010].

  14. Management of waste associated with the decommissioning of the JASON research reactor and the nuclear laboratories at the Royal Naval College Greenwich

    International Nuclear Information System (INIS)

    Beeley, P.A.; Lockwood, R.J.S.; Hoult, D.; Major, R.

    2001-01-01

    In 1996 the UK Government announced that the Royal Naval College, Greenwich would pass to non-defence use by the millennium. As a consequence of this decision, the decommissioning of the JASON 10 kW Argonaut research reactor and the relocation of the Department of Nuclear Science and Technology (DNST) were approved by the Ministry of Defence. The decommissioning of the reactor commenced in November 1997 while DNST remained operational until October 1998. The DNST was responsible for education and training in support of the UK Naval Nuclear Propulsion Programme and operated academic laboratories for atomic and nuclear physics, health physics, instrument calibration and radiochemistry. Therefore, besides the nuclear reactor, open and sealed sources (alpha, beta and gamma), intense x-ray (sealed tube) and gamma-ray ( 60 CO and 137 Cs) sources and small 241 Am/Be neutron sources had been used in the Department for over 35 years. Decommissioning of all facilities was therefore a relatively complex task and the management of waste streams was challenging. All facilities were successfully decommissioned for unrestricted site release by December 1999 and this paper will describe the methodology used for preparation, storage, characterisation and disposal of all waste streams. The most significant waste management task during this decommissioning programme was that associated with the JASON reactor. It should be noted that the JASON reactor fuel was not designated as nuclear waste, the fuel removal and storage were covered under separate contracts and therefore no high level waste was generated. With respect to other waste streams, a combination of Monte Carlo modelling and selective sampling and analysis of the reactor materials was used to estimate the quantities of waste as follows: LLW - 76 tonnes packed in 4 half height ISO containers; LLW - 6 Tonnes packed in 200litre drums in 1 full height ISO container; ILW - 60 kg packed in approved shielded containers; FRW -121

  15. The Management of Large Components from Decommissioning to Storage and Disposal. A Report of the Task Group on Large Components of the NEA Working Party on Decommissioning and Dismantling (WPDD)

    International Nuclear Information System (INIS)

    2012-01-01

    During the decommissioning operations of a nuclear facility or during the maintenance activities of an operating nuclear facility, dismantled components may either be segmented (i.e., reduced in size) in order to put the pieces in standardised containers or removed as single or multiple large pieces for treatment or transportation to a disposal facility. Both options have been used and experience exists in most countries with mature nuclear programmes. The considerations and criteria leading to such decisions are multifaceted and include both legal and regulatory aspects dealing notably with transportation, disposal and waste-acceptance criteria (WAC), as well as the availability and acceptance of sized transport containers and dose limits. The final management option may not be selected solely by the decommissioning entity. In any case, radioactive waste will be generated and will need to be processed through a sequence of operations, including deconstruction, conditioning, extraction, short- or long-term storage on site, transport conditioning, transportation to a waste management facility that may be a storage facility, a treatment facility or a disposal facility. The different actors involved play a significant role in choosing the most relevant management option throughout the entire process, since a single option may not be the most appropriate for every stage of the process. If the removal of large pieces raises very complex transportation issues, for instance, it may be preferable to cut components into smaller pieces; in some cases, the disposal of large components may also generate an unacceptable or less than optimal capacity of the repository. For such reasons, integration is required throughout the entire project. The proposed management option for large components should be described in the facility's decommissioning plan and should explain why a particular option was selected. In choosing the proposed option, the decommissioning organisation needs to

  16. New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

    International Nuclear Information System (INIS)

    Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

    1996-01-01

    The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

  17. Revised analyses of decommissioning for the reference boiling water reactor power station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure - appendices. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.I.; Bierschbach, M.C.; Konzek, G.J.; McDuffie, P.N.

    1996-07-01

    The NRC staff is in need of decommissioning bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to update the needed bases documentation. This report presents the results of a review and reevaluation of the PNL 1980 decommissioning study of the Washington Public Power Supply System`s Washington Nuclear Plant Two (WNP-2) located at Richland, Washington, including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives. These alternatives now include an initial 5-7 year period during which time the spent fuel is stored in the spent fuel pool prior to beginning major disassembly or extended safe storage of the plant. Included for information (but not presently part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clear structures on the site and to restore the site to a {open_quotes}green field{close_quotes} condition. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low-level waste (i.e., Greater-Than-Class C), and reflects 1993 costs for labor, materials, transport, and disposal activities. Sensitivity of the total license termination cost to the disposal costs at different low-level radioactive waste disposal sites, to different depths of contaminated concrete surface removal within the facilities, and to different transport distances is also examined.

  18. Revised analyses of decommissioning for the reference boiling water reactor power station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure - appendices. Final report

    International Nuclear Information System (INIS)

    Smith, R.I.; Bierschbach, M.C.; Konzek, G.J.; McDuffie, P.N.

    1996-07-01

    The NRC staff is in need of decommissioning bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to update the needed bases documentation. This report presents the results of a review and reevaluation of the PNL 1980 decommissioning study of the Washington Public Power Supply System's Washington Nuclear Plant Two (WNP-2) located at Richland, Washington, including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives. These alternatives now include an initial 5-7 year period during which time the spent fuel is stored in the spent fuel pool prior to beginning major disassembly or extended safe storage of the plant. Included for information (but not presently part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clear structures on the site and to restore the site to a open-quotes green fieldclose quotes condition. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low-level waste (i.e., Greater-Than-Class C), and reflects 1993 costs for labor, materials, transport, and disposal activities. Sensitivity of the total license termination cost to the disposal costs at different low-level radioactive waste disposal sites, to different depths of contaminated concrete surface removal within the facilities, and to different transport distances is also examined

  19. Coatings for fast breeder reactor components

    International Nuclear Information System (INIS)

    Johnson, R.N.

    1984-04-01

    Several types of metallurgical coatings are used in the unique environments of the fast breeder reactor. Most of the coatings have been developed for tribological applications, but some also serve as corrosion barriers, diffusion barriers, or radionuclide traps. The materials that have consistently given the best performance as tribological coatings in the breeder reactor environments have been coatings based on chromium carbide, nickel aluminide, or Tribaloy 700 (a nickel-base hard-facing alloy). Other coatings that have been qualified for limited applications include chromium plating for low temperature galling protection and nickel plating for radionuclide trapping

  20. Numerical analysis of magnetoelastic coupled buckling of fusion reactor components

    International Nuclear Information System (INIS)

    Demachi, K.; Yoshida, Y.; Miya, K.

    1994-01-01

    For a tokamak fusion reactor, it is one of the most important subjects to establish the structural design in which its components can stand for strong magnetic force induced by plasma disruption. A number of magnetostructural analysis of the fusion reactor components were done recently. However, in these researches the structural behavior was calculated based on the small deformation theory where the nonlinearity was neglected. But it is known that some kinds of structures easily exceed the geometrical nonlinearity. In this paper, the deflection and the magnetoelastic buckling load of fusion reactor components during plasma disruption were calculated

  1. Decommissioning of the Nuclear Reactors R2 and R2-0 at Studsvik, Sweden. General Data as called for under Article 37 of the Euratom Treaty

    International Nuclear Information System (INIS)

    2009-01-01

    This document describes the plans for decommissioning of the nuclear research and material test reactors R2 and R2-0, situated at the Studsvik site close to the city of Nykoeping, Sweden. The purpose of the document is to serve as information for the European Commission, and to fulfil the requirements of Article 37 of the Euratom Treaty. Studsvik is situated on the Baltic coast, about 20 km east of Nykoeping and 80 km southwest of Stockholm. The site comprises the reactors R2 and R2-0 and several facilities for material investigation and radioactive waste treatment and storage. The reactors were used for a number of different purposes from 1960 until June 2005, when they were shut down following a decision by the operator. Decommissioning of the reactor facility is planned to be completed in 2016 after dismantling and conditioning of radioactive parts and demolition of the facility. Solid and liquid radioactive wastes from the dismantling activities will be treated and stored on-site awaiting final disposal. The waste treatment facilities, which are situated in other buildings at the Studsvik site, are planned to continue operation during and after the decommissioning of the reactor facility. All nuclear fuel has been transferred to a separate storage facility and is being shipped to the US according to existing agreements. The objective of the planned dismantling activities is to achieve clearance of the facility to make it possible to either demolish the buildings or use them for other purposes. The operator has divided the planning for dismantling and demolition of the facility into three phases [1]: Dismantling 1, including primary system decontamination, dismantling of the reactors with systems in the reactor pool, draining, cleaning and temporary covering of the reactor pool. This phase has begun and is due to last till approximately December 2009. Dismantling 2, including dismantling of systems in the reactor facility, removal of equipment, radiological

  2. New Materials Developed To Meet Regulatory And Technical Requirements Associated With In-Situ Decommissioning Of Nuclear Reactors And Associated Facilities

    International Nuclear Information System (INIS)

    Blankenship, J.; Langton, C.; Musall, J.; Griffin, W.

    2012-01-01

    For the 2010 ANS Embedded Topical Meeting on Decommissioning, Decontamination and Reutilization and Technology, Savannah River National Laboratory's Mike Serrato reported initial information on the newly developed specialty grout materials necessary to satisfy all requirements associated with in-situ decommissioning of P-Reactor and R-Reactor at the U.S. Department of Energy's Savannah River Site. Since that report, both projects have been successfully completed and extensive test data on both fresh properties and cured properties has been gathered and analyzed for a total of almost 191,150 m 3 (250,000 yd 3 ) of new materials placed. The focus of this paper is to describe the (1) special grout mix for filling the P-Reactor vessel (RV) and (2) the new flowable structural fill materials used to fill the below grade portions of the facilities. With a wealth of data now in hand, this paper also captures the test results and reports on the performance of these new materials. Both reactors were constructed and entered service in the early 1950s, producing weapons grade materials for the nation's defense nuclear program. R-Reactor was shut down in 1964 and the P-Reactor in 1991. In-situ decommissioning (ISD) was selected for both facilities and performed as Comprehensive Environmental Response, Compensations and Liability Act actions (an early action for P-Reactor and a removal action for R-Reactor), beginning in October 2009. The U.S. Department of Energy concept for ISD is to physically stabilize and isolate intact, structurally robust facilities that are no longer needed for their original purpose of producing (reactor facilities), processing (isotope separation facilities), or storing radioactive materials. Funding for accelerated decommissioning was provided under the American Recovery and Reinvestment Act. Decommissioning of both facilities was completed in September 2011. ISD objectives for these CERCLA actions included: (1) Prevent industrial worker exposure to

  3. The decommissioning of nuclear facilities

    International Nuclear Information System (INIS)

    Niel, J.Ch.; Rieu, J.; Lareynie, O.; Delrive, L.; Vallet, J.; Girard, A.; Duthe, M.; Lecomte, C.; Rozain, J.P.; Nokhamzon, J.G.; Davoust, M.; Eyraud, J.L.; Bernet, Ph.; Velon, M.; Gay, A.; Charles, Th.; Leschaeva, M.; Dutzer, M.; Maocec, Ch.; Gillet, G.; Brut, F.; Dieulot, M.; Thuillier, D.; Tournebize, F.; Fontaine, V.; Goursaud, V.; Birot, M.; Le Bourdonnec, Th.; Batandjieva, B.; Theis, St.; Walker, St.; Rosett, M.; Cameron, C.; Boyd, A.; Aguilar, M.; Brownell, H.; Manson, P.; Walthery, R.; Wan Laer, W.; Lewandowski, P.; Dorms, B.; Reusen, N.; Bardelay, J.; Damette, G.; Francois, P.; Eimer, M.; Tadjeddine, A.; Sene, M.; Sene, R.

    2008-01-01

    This file includes five parts: the first part is devoted to the strategies of the different operators and includes the following files: the decommissioning of nuclear facilities Asn point of view, decommissioning of secret nuclear facilities, decommissioning at the civil Cea strategy and programs, EDF de-construction strategy, Areva strategy for decommissioning of nuclear facilities; the second one concerns the stakes of dismantling and includes the articles as follow: complete cleanup of buildings structures in nuclear facilities, decommissioning of nuclear facilities and safety assessment, decommissioning wastes management issues, securing the financing of long-term decommissioning and waste management costs, organizational and human factors in decommissioning projects, training for the decommissioning professions: the example of the Grenoble University master degree; the third part is devoted to the management of dismantling work sites and includes the different articles as follow: decommissioning progress at S.I.C.N. plant, example of decommissioning work site in Cea Grenoble: Siloette reactor decommissioning, matters related to decommissioning sites, decommissioning of french nuclear installations: the viewpoint of a specialist company, specificities of inspections during decommissioning: the Asn inspector point of view; the fourth part is in relation with the international approach and includes as follow: IAEA role in establishing a global safety regime on decommissioning, towards harmonization of nuclear safety practices in Europe: W.E.N.R.A. and the decommissioning of nuclear facilities, EPA superfund program policy for decontamination and decommissioning, progress with remediation at Sellafield, progress and experiences from the decommissioning of the Eurochemic reprocessing plant in Belgium, activities of I.R.S.N. and its daughter company Risk-audit I.r.s.n./G.r.s. international in the field of decommissioning of nuclear facilities in eastern countries

  4. Revised analyses of decommissioning for the reference boiling water reactor power station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure - main report. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.I.; Bierschbach, M.C.; Konzek, G.J.; McDuffie, P.N.

    1996-07-01

    The NRC staff is in need of updated bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to update the needed bases documentation. This report presents the results of a review and reevaluation of the PNL 1980 decommissioning study of the Washington Public Power Supply System`s Washington Nuclear Plant Two (WNP-2), which is a boiling water reactor (BWR), located at Richland, Washington, including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives. These alternatives now include an initial 5-7 year period during which time the spent fuel is stored in the spent fuel pool prior to beginning major disassembly or extended safe storage of the plant. Included for information (but not part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clean structures on the site and to restore the site to a {open_quotes}green field{close_quotes} condition. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low- level waste (i.e., Greater-Than-Class C), and reflects 1993 costs for labor, materials, transport, and disposal activities. Sensitivity of the total license termination cost to the disposal costs at different low-level radioactive waste disposal sites, to different depths of contaminated concrete surface removal within the facilities, and to different transport distances is also examined.

  5. Revised analyses of decommissioning for the reference boiling water reactor power station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure - main report. Final report

    International Nuclear Information System (INIS)

    Smith, R.I.; Bierschbach, M.C.; Konzek, G.J.; McDuffie, P.N.

    1996-07-01

    The NRC staff is in need of updated bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to update the needed bases documentation. This report presents the results of a review and reevaluation of the PNL 1980 decommissioning study of the Washington Public Power Supply System's Washington Nuclear Plant Two (WNP-2), which is a boiling water reactor (BWR), located at Richland, Washington, including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives. These alternatives now include an initial 5-7 year period during which time the spent fuel is stored in the spent fuel pool prior to beginning major disassembly or extended safe storage of the plant. Included for information (but not part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clean structures on the site and to restore the site to a open-quotes green fieldclose quotes condition. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low- level waste (i.e., Greater-Than-Class C), and reflects 1993 costs for labor, materials, transport, and disposal activities. Sensitivity of the total license termination cost to the disposal costs at different low-level radioactive waste disposal sites, to different depths of contaminated concrete surface removal within the facilities, and to different transport distances is also examined

  6. Abrasive blasting, a technique for the industrial decontamination of metal components and concrete blocks from decommissioning to unconditional release levels

    International Nuclear Information System (INIS)

    Gills, R.; Lewandowski, P.; Ooms, B.; Reusen, N.; Van Laer, W.; Walthery, R.

    2007-01-01

    When decommissioning nuclear installations, large quantities of metal components are produced as well as significant amounts of other radioactive materials, which mostly show low surface contamination. Having been used or having been brought for a while in a controlled area marks them as 'suspected material'. In view of the very high costs for radioactive waste processing and disposal, alternatives have been considered, and much effort has gone to recycling through decontamination, melting and unconditional release of metals. In a broader context, recycling of materials can considered to be a first order ecological priority in order to limit the quantities of radioactive wastes for final disposal and to reduce the technical and economic problems involved with the management of radioactive wastes. It will help as well to make economic use of primary material and to conserve natural resources of basic material for future generations. In a demonstration programme, Belgoprocess has shown that it is economically interesting to decontaminate metal components to unconditional release levels using dry abrasive blasting techniques, the unit cost for decontamination being only 30 % of the global cost for radioactive waste treatment, conditioning, storage and disposal. As a result, an industrial dry abrasive blasting unit was installed in the Belgoprocess central decontamination infrastructure. At the end of December 2006, more than 1,128 Mg of contaminated metal has been treated as well as 313 Mg of concrete blocks. The paper gives an overview of the experience relating to the decontamination of metal material and concrete blocks at the decommissioning of the Eurochemic reprocessing plant in Dessel, Belgium as well from the decontamination of concrete containers by abrasive blasting. (authors)

  7. Mechanical components: fabrication of major reactor structures

    International Nuclear Information System (INIS)

    Nicholson, S.

    1985-01-01

    The paper examines the validity of criticisms of quality assurance of mechanical plant and welded products within major reactor structures, taking into account experience gained on the AGR's. Various constructive recommendations are made aimed at furthering the objectives of quality assurance in the nuclear industry and making it more cost-effective. Current levels of quality related costs in the fabrication industry are provided as a basis for discussion. (U.K.)

  8. Guidelines of Decommissioning Schedule Establishment

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jae Yong; Yun, Taesik; Kim, Younggook; Kim, Hee-Geun [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    Decommissioning has recently become an issue highlighted in Korea due to the Permanent Shutdown (PS) of Kori-1 plant. Since Korea Hydro and Nuclear Power (KHNP) Company decided the PS of Kori-1 instead of further continued operation, Kori-1 will be the first decommissioning plant of the commercial reactors in Korea. Korean regulatory authority demands Initial Decommissioning Plan (IDP) for all the plants in operation and under construction. In addition, decommissioning should be considered for the completion of the life cycle of NPPs. To date, Korea has no experience regarding decommissioning of the commercial reactor and a lot of uncertainties will be expected due to its site-specific factors. However, optimized decommissioning process schedule must be indispensable in the safety and economic efficiency of the project. Differed from USA, Korea has no experience and know-hows of the operation and site management for decommissioning. Hence, in Korea, establishment of decommissioning schedule has to give more weight to safety than precedent cases. More economical and rational schedule will be composed by collecting and analyzing the experience data and site-specific data and information as the decommissioning progresses. In a long-range outlook, KHNP having capability of NPP decommissioning will try to decommissioning business in Korea and foreign countries.

  9. Study on decommissioning

    International Nuclear Information System (INIS)

    2012-01-01

    This project consists of researches on (1) establishment of review plan on application of decommissioning, (2) establishment of specific method to confirm decommissioning completion, of decommissioning and (3) establishment of radioactive waste management guideline during dismantling and (4) development of the regulatory system on decommissioning in response to Fukushima Daiichi NPP accident. About researches on establishment of review plan on application of decommissioning. 'Planning of the Commercial Power Reactor Decommissioning:2001' which was published by Atomic Energy Society of Japan, was evaluated whether it suited the requirement for the decommissioning stipulated in the law, and the draft evaluation report was prepared. About researches on establishment of specific method to confirm decommissioning completion, technical information of practical procedures on the confirmation in U.S.A. were organized based on MARSSIM (Multi-Agency Radiation Survey and Site Investigation Manual, NUREG-1575) and applicability of MARSSIM on the confirmation in Japan was examined. Exposed doses for public during decommissioning period were estimated to study dose criterion of the confirmation. Radioactive concentrations in the soil of Tokai and Hamaoka NPP caused by the Fukushima Daiichi NPP accident were also investigated. About researches on establishment of radioactive waste management guideline during dismantling, one concrete core was sampled in biological shield of the Tokai NPP and radioactive concentrations were investigated. About researches on development of the regulatory system on decommissioning in response to Fukushima Daiichi NPP accident, present status of Three Mile Island Unit 2 and Chernobyl NPP Unit 4 were investigated. Present status of regulatory systems for decommissioning in foreign countries taken in consideration of the accident was also researched. (author)

  10. Rancho Seco--Decommissioning Update

    International Nuclear Information System (INIS)

    Newey, J. M.; Ronningen, E. T.; Snyder, M. W.

    2003-01-01

    The Rancho Seco Nuclear Generating Station ceased operation in June of 1989 and entered an extended period of SAFSTOR to allow funds to accumulate for dismantlement. Incremental dismantlement was begun in 1997 of steam systems and based on the successful completion of work, the Sacramento Municipal Utility District (SMUD) board of directors approved full decommissioning in July 1999. A schedule has been developed for completion of decommissioning by 2008, allowing decommissioning funds to accumulate until they are needed. Systems removal began in the Auxiliary Building in October of 1999 and in the Reactor Building in January of 2000. Systems dismantlement continues in the Reactor Building and should be completed by the end of 2003. System removal is near completion in the Auxiliary Building with removal of the final liquid waste tanks in progress. The spent fuel has been moved to dry storage in an onsite ISFSI, with completion on August 21, 2002. The spent fuel racks are currently being removed from the pool, packaged and shipped, and then the pool will be cleaned. Also in the last year the reactor coolant pumps and primary piping were removed and shipped. Characterization and planning work for the reactor vessel and internals is also in progress with various cut-up and/or disposal options being evaluated. In the year ahead the remaining systems in the Reactor Building will be removed, packaged and sent for disposal, including the pressurizer. Work will be started on embedded and underground piping and the large outdoor tanks. Building survey and decontamination will begin. RFP's for removal of the vessel and internals and the steam generators are planned to fix the cost of those components. If the costs are consistent with current estimates the work will go forward. If they are not, hardened SAFSTOR/entombment may be considered

  11. Nuclear decommissioning and society

    International Nuclear Information System (INIS)

    Pasqualetti, M.J.

    1990-01-01

    Links between decommissioning in general, reactor decommissioning in particular, and the public are indexed. The established links are recognised and others, such as jobs, are discussed. Finally the links with policy, such as political geography, and wider issues of the environment and public concern over waste disposal are considered. Decommissioning is a relatively new field where public opinion must now be considered but it has implications both for existing nuclear power plants and those planned for the future, especially in their siting. This book looks especially at the situation in the United Kingdom. There are twelve papers, all indexed separately. (UK)

  12. Decision-making process to shut down, refurbish/modify, or decommission research reactors

    International Nuclear Information System (INIS)

    Stover, R.L.; Murphie, W.E.

    1992-01-01

    Most US research reactors were built more than 20 years ago and some more than 40 years ago. Many have undergone refurbishments and modifications to update their safety systems and experimental capabilities. But changing safety bases, social concerns, and budget constraints have required research reactor operators to continually make decisions to shut down or refurbish/modify their facilities. These decisions involve potential replacement of reactor equipment that has reached its lifetime limits. Changes in philosophy and operation of the reactors are also factors to be considered. In this paper, each of the four factors involved in the decision-making process are discussed in detail. Then, several examples from DOE research reactors in the United States are discussed. Finally, some general conclusions are given to aid in the decision-making process

  13. Demonstration test on manufacturing steel bars for concrete reinforcement for recycling of reactor decommissioning metal scrap

    International Nuclear Information System (INIS)

    Sakurai, D.; Anabuki, Y.

    1993-01-01

    To prove the possibility of recycling the steel scrap resulting from decommissioning of a nuclear power plant, this salvaged steel would be formed into steel bars for concrete reinforcement, as the restricted use and limited use at nuclear plants. The shifting behavior of radioactive isotopes (RI) in the melting process was confirmed through the laboratory hot test using the RI. Then, the demonstration cold test for steel bars for reinforcement using the nonradioactive isotope was conducted in on-line production facilities. In this test the quality of steel bars and uniform distribution of RI were proven and material balance and operational data were obtained. These data show the recycling to steel bars for concrete reinforcement is applicable from economical and safety aspects

  14. Reactor component inventory system at FFTF

    International Nuclear Information System (INIS)

    Ordonez, C.R.; Redekopp, R.D.; Reed, E.A.

    1985-02-01

    A reliable inventory control system was developed at the Fast Flux Test Facility (FFTF) to keep track of the occupancy of 900 refueling facility locations, to compile historical data on the movement of each reactor assembly, and to simulate assembly moves. The simulate capability is valuable because it allows verification of documents before they are issued for use in the plant, and eliminates the possibility of planning illegal or impossible moves. The system is installed on a UNIVAC 1100 computer and is maintained using a data base management system by Sperry Univac called MAPPER

  15. HEU Measurements of Holdup and Recovered Residue in the Deactivation and Decommissioning Activities of the 321-M Reactor Fuel Fabrication Facility at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    DEWBERRY, RAYMOND; SALAYMEH, SALEEM R.; CASELLA, VITO R.; MOORE, FRANK S.

    2005-03-11

    This paper contains a summary of the holdup and material control and accountability (MC&A) assays conducted for the determination of highly enriched uranium (HEU) in the deactivation and decommissioning (D&D) of Building 321-M at the Savannah River Site (SRS). The 321-M facility was the Reactor Fuel Fabrication Facility at SRS and was used to fabricate HEU fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the SRS production reactors. The facility operated for more than 35 years. During this time thousands of uranium-aluminum-alloy (U-Al) production reactor fuel tubes were produced. After the facility ceased operations in 1995, all of the easily accessible U-Al was removed from the building, and only residual amounts remained. The bulk of this residue was located in the equipment that generated and handled small U-Al particles and in the exhaust systems for this equipment (e.g., Chip compactor, casting furnaces, log saw, lathes A & B, cyclone separator, Freon{trademark} cart, riser crusher, ...etc). The D&D project is likely to represent an important example for D&D activities across SRS and across the Department of Energy weapons complex. The Savannah River National Laboratory was tasked to conduct holdup assays to quantify the amount of HEU on all components removed from the facility prior to placing in solid waste containers. The U-235 holdup in any single component of process equipment must not exceed 50 g in order to meet the container limit. This limit was imposed to meet criticality requirements of the low level solid waste storage vaults. Thus the holdup measurements were used as guidance to determine if further decontamination of equipment was needed to ensure that the quantity of U-235 did not exceed the 50 g limit and to ensure that the waste met the Waste Acceptance Criteria (WAC) of the solid waste storage vaults. Since HEU is an accountable nuclear material, the holdup assays and assays of recovered

  16. Common cause failures of reactor pressure components

    International Nuclear Information System (INIS)

    Mankamo, T.

    1978-01-01

    The common cause failure is defined as a multiple failure event due to a common cause. The existence of common failure causes may ruin the potential advantages of applying redundancy for reliability improvement. Examples relevant to large mechanical components are presented. Preventive measures against common cause failures, such as physical separation, equipment diversity, quality assurance, and feedback from experience are discussed. Despite the large number of potential interdependencies, the analysis of common cause failures can be done within the framework of conventional reliability analysis, utilizing, for example, the method of deriving minimal cut sets from a system fault tree. Tools for the description and evaluation of dependencies between components are discussed: these include the model of conditional failure causes that are common to many components, and evaluation of the reliability of redundant components subjected to a common load. (author)

  17. Fatigue evaluation in reactor vessel components

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel; Miranda, Carlos A. de J.

    1994-01-01

    This paper presents a sequence of increasing complexity forms of evaluating fatigue damage of nuclear pressure vessel components caused by cycling loadings. Examples are included in order to illustrate such procedures. (author)

  18. Conference on decontamination and decommissioning of nuclear facilities

    International Nuclear Information System (INIS)

    Meservey, R.H.

    1979-01-01

    A brief history of Decontamination and Decommissioning (D and D) experience at the Idaho National Engineering Laboratory is presented as an introduction to the status of current projects. Details are then presented as an introduction to the status of current projects. Details are then presented on a project to remove sodium from some major components of the Hallam reactor and on the Organic Moderated Reactor Experiment (OMRE) decommissioning project. Cost, schedule, waste volume, and other technical data from these projects are presented. In addition, a brief summary of the future INEL D and D program is presented

  19. Fort St. Vrain decommissioning project

    International Nuclear Information System (INIS)

    Fisher, M.

    1998-01-01

    Public Service Company of Colorado (PSCo), owner of the Fort St. Vrain nuclear generating station, achieved its final decommissioning goal on August 5, 1997 when the Nuclear Regulatory Commission terminated the Part 50 reactor license. PSCo pioneered and completed the world's first successful decommissioning of a commercial nuclear power plant after many years of operation. In August 1989, PSCo decided to permanently shutdown the reactor and proceed with its decommissioning. The decision to proceed with early dismantlement as the appropriate decommissioning method proved wise for all stake holders - present and future - by mitigating potential environmental impacts and reducing financial risks to company shareholders, customers, employees, neighboring communities and regulators. We believe that PSCo's decommissioning process set an exemplary standard for the world's nuclear industry and provided leadership, innovation, advancement and distinguished contributions to other decommissioning efforts throughout the world. (author)

  20. Lifetime analysis of fusion-reactor components

    International Nuclear Information System (INIS)

    Mattas, R.F.

    1983-01-01

    A one-dimensional computer code has been developed to examine the lifetime of first-wall and impurity-control components. The code incorporates the operating and design parameters, the material characteristics, and the appropriate failure criteria for the individual components. The major emphasis of the modelling effort has been to calculate the temperature-stress-strain-radiation effects history of a component so that the synergystic effects between sputtering erosion, swelling, creep, fatigue, and crack growth can be examined. The general forms of the property equations are the same for all materials in order to provide the greatest flexibility for materials selection in the code. The code is capable of determining the behavior of a plate, composed of either a single or dual material structure, that is either totally constrained or constrained from bending but not from expansion. The code has been utilized to analyze the first walls for FED/INTOR and DEMO

  1. Decommissioning of the Dragon High Temperature Reactor (HTR) Located at the Former United Kingdom Atomic Energy Authority (UKAEA) Research Site at Winfrith - 13180

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Anthony A. [Research Sites Restoration Ltd, Winfrith, Dorset (United Kingdom)

    2013-07-01

    The Dragon Reactor was constructed at the United Kingdom Atomic Energy Research Establishment at Winfrith in Dorset through the late 1950's and into the early 1960's. It was a High Temperature Gas Cooled Reactor (HTR) with helium gas coolant and graphite moderation. It operated as a fuel testing and demonstration reactor at up to 20 MW (Thermal) from 1964 until 1975, when international funding for this project was terminated. The fuel was removed from the core in 1976 and the reactor was put into Safestore. To meet the UK's Nuclear Decommissioning Authority (NDA) objective to 'drive hazard reduction' [1] it is necessary to decommission and remediate all the Research Sites Restoration Ltd (RSRL) facilities. This includes the Dragon Reactor where the activated core, pressure vessel and control rods and the contaminated primary circuit (including a {sup 90}Sr source) still remain. It is essential to remove these hazards at the appropriate time and return the area occupied by the reactor to a safe condition. (author)

  2. Method of decommissioning nuclear reactor building by utilizing sea water buyoancy

    International Nuclear Information System (INIS)

    Iwashima, Sumio; Ogoshi, Shigeru; Kobari, Shin-ichi.

    1989-01-01

    Upon dismantling nuclear reactor buildings, peripheral yards are excavated and channels leading to sea shore are formed. Since the outer walls of the reactor buildings are made of iron-reinforced concretes, the opening poritons are grouted with concretes to attain a tightly such closed structure that radioactive wastes, etc. in the inside are not flown out upon reactor discommisioning. Peripheral buildings at relatively low level of radiation contaminations are dismantled and withdrawn. The fundations of the nuclear reactor buildings were dug out and jacked to separate base rocks and the reactor buildings. Then, sea water is introduced into the water channels to entirely float up the buildings. A water gate is disposed in the water channel on the side of sea shore to control the level of sea water. The buildings are moved and guided to the sea shore and towed to a site optimum as a permanent storage area and then burried in that place. The operation period for the discommissioning work can greatly be shortened and the radiation dose and the amount of the wastes can be reduced. (T.M.)

  3. Data Validation Package, June 2016 Groundwater Sampling at the Hallam, Nebraska, Decommissioned Reactor Site, August 2016

    Energy Technology Data Exchange (ETDEWEB)

    Surovchak, Scott [USDOE Office of Legacy Management, Washington, DC (United States); Miller, Michele [Navarro Research and Engineering, Oak Ridge, TN (United States)

    2016-08-01

    The 2008 Long-Term Surveillance Plan [LTSP] for the Decommissioned Hallam Nuclear Power Facility, Hallam, Nebraska (http://www.lm.doe.gov/Hallam/Documents.aspx) requires groundwater monitoring once every 2 years. Seventeen monitoring wells at the Hallam site were sampled during this event as specified in the plan. Planned monitoring locations are shown in Attachment 1, Sampling and Analysis Work Order. Water levels were measured at all sampled wells and at two additional wells (6A and 6B) prior to the start of sampling. Additionally, water levels of each sampled well were measured at the beginning of sampling. See Attachment 2, Trip Report, for additional details. Sampling and analysis were conducted as specified in Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated, http://energy.gov/lm/downloads/sampling-and-analysis-plan-us-department- energy-office-legacy-management-sites). Gross alpha and gross beta are the only parameters that were detected at statistically significant concentrations. Time/concentration graphs of the gross alpha and gross beta data are included in Attachment 3, Data Presentation. The gross alpha and gross beta activity concentrations observed are consistent with values previously observed and are attributed to naturally occurring radionuclides (e.g., uranium and uranium decay chain products) in the groundwater.

  4. Nuclear archaeology: the influence of decommissioning on future reactor siting in the UK

    International Nuclear Information System (INIS)

    Openshaw, S.

    1990-01-01

    Whereas considerations of waste, cost, opinion, and the law are often abstract elements of nuclear power, the siting of generating stations produces concrete land use and aesthetic reminders of its presence. What will future generations find as a result of the nuclear era? Will the power plant sites be restored to greenfield conditions? When will this occur? How will such decisions affect surrounding land use and societal characteristics such as employment opportunities and quality of life? These and other questions revolve around the land use and siting links to decommissioning decisions. The two prime determinants of the land use legacy will be the timing of stage 3 dismantlement and the degree to which the sites can be restored to pre-use conditions. In this chapter it is argued that the sites are unlikely ever to be released for unrestricted use, but rather that the use and the visual impacts of the sites will increase over time, eventually becoming the objects which, perhaps, future archaeologists will study. (author)

  5. Development of plasma arc cutting technique for dismantlement of reactor internals in JPDR decommissioning program

    International Nuclear Information System (INIS)

    Yanagihara, Satoshi; Tanaka, Mitsugu; Ujihara, Norio.

    1988-01-01

    The decommissioning program for JPDR has been conducted by JAERI since 1981 under contact with the Science and Technology Agency of Japan. The development of cutting tools for dismantling the JPDR is one of the important items in the program. An underwater plasma arc cutting technique was selected for dismantling the JPDR core internals. The study was concentrated on improving the cutting ability in water. Various cutting tests were conducted changing the parameters such as arc current, supply gas and cutting speed to evaluate the most effective cutting condition. Through the study, it has been achieved to be able to cut a 130 mm thick stainless steel plate in water. In addition, the amount and the characteristics of by-products were measured during the cutting tests for the safety evaluation of the dismantling activities. Final cutting tests and checkout of whole plasma arc cutting system were conducted using a mockup water pool and test pieces simulating the JPDR core internals. It was proved from the tests that the cutting system developed in the program will be applicable for the JPDR core internals dismantlement. (author)

  6. In Situ Decommissioning (ISD) Concepts and Approaches for Excess Nuclear Facilities Decommissioning End State - 13367

    International Nuclear Information System (INIS)

    Serrato, Michael G.; Musall, John C.; Bergren, Christopher L.

    2013-01-01

    The United States Department of Energy (DOE) currently has numerous radiologically contaminated excess nuclear facilities waiting decommissioning throughout the Complex. The traditional decommissioning end state is complete removal. This commonly involves demolishing the facility, often segregating various components and building materials and disposing of the highly contaminated, massive structures containing tons of highly contaminated equipment and piping in a (controlled and approved) landfill, at times hundreds of miles from the facility location. Traditional demolition is costly, and results in significant risks to workers, as well as risks and costs associated with transporting the materials to a disposal site. In situ decommissioning (ISD or entombment) is a viable alternative to demolition, offering comparable and potentially more protective protection of human health and the environment, but at a significantly reduced cost and worker risk. The Savannah River Site (SRS) has completed the initial ISD deployment for radiologically contaminated facilities. Two reactor (P and R Reactors) facilities were decommissioned in 2011 using the ISD approach through the American Recovery and Reinvestment Act. The SRS ISD approach resolved programmatic, regulatory and technical/engineering issues associated with avoiding the potential hazards and cost associated with generating and disposing of an estimated 124,300 metric tons (153,000 m 3 ) of contaminated debris per reactor. The DOE Environmental Management Office of Deactivation and Decommissioning and Facility Engineering, through the Savannah River National Laboratory, is currently investigating potential monitoring techniques and strategies to assess ISD effectiveness. As part of SRS's strategic planning, the site is seeking to leverage in situ decommissioning concepts, approaches and facilities to conduct research, design end states, and assist in regulatory interactions in broad national and international

  7. In Situ Decommissioning (ISD) Concepts and Approaches for Excess Nuclear Facilities Decommissioning End State - 13367

    Energy Technology Data Exchange (ETDEWEB)

    Serrato, Michael G. [Savannah River National Laboratory, Savannah River Nuclear Solutions, Aiken, SC 29808 (United States); Musall, John C.; Bergren, Christopher L. [Savannah River Nuclear Solutions, Aiken, SC 29808 (United States)

    2013-07-01

    The United States Department of Energy (DOE) currently has numerous radiologically contaminated excess nuclear facilities waiting decommissioning throughout the Complex. The traditional decommissioning end state is complete removal. This commonly involves demolishing the facility, often segregating various components and building materials and disposing of the highly contaminated, massive structures containing tons of highly contaminated equipment and piping in a (controlled and approved) landfill, at times hundreds of miles from the facility location. Traditional demolition is costly, and results in significant risks to workers, as well as risks and costs associated with transporting the materials to a disposal site. In situ decommissioning (ISD or entombment) is a viable alternative to demolition, offering comparable and potentially more protective protection of human health and the environment, but at a significantly reduced cost and worker risk. The Savannah River Site (SRS) has completed the initial ISD deployment for radiologically contaminated facilities. Two reactor (P and R Reactors) facilities were decommissioned in 2011 using the ISD approach through the American Recovery and Reinvestment Act. The SRS ISD approach resolved programmatic, regulatory and technical/engineering issues associated with avoiding the potential hazards and cost associated with generating and disposing of an estimated 124,300 metric tons (153,000 m{sup 3}) of contaminated debris per reactor. The DOE Environmental Management Office of Deactivation and Decommissioning and Facility Engineering, through the Savannah River National Laboratory, is currently investigating potential monitoring techniques and strategies to assess ISD effectiveness. As part of SRS's strategic planning, the site is seeking to leverage in situ decommissioning concepts, approaches and facilities to conduct research, design end states, and assist in regulatory interactions in broad national and

  8. Method to chemically decontaminate nuclear reactor components

    International Nuclear Information System (INIS)

    Bertholdt, H.O.

    1984-01-01

    The large decontamination of components of the primary circuit of activated corrosion products in the oxide layer of the structure materials firstly involves an approx. 1 hour oxidation treatment with alkali permanganate solution. Following intermediate rinsing with deionate, they are etched with an inhibited citrate-oxalate solution for 5-20 hours. This is followed by post-treatment with a citric acid/H2O2 solution containing suspended fiber particles. (orig./PW)

  9. Decommissioning of the Shippingport Atomic Power Station

    International Nuclear Information System (INIS)

    LaGuardia, T.S.

    1988-01-01

    The Shippingport reactor was originally designed as a pressurized water reactor and operated for approximately 10 years in that mode. Later, in 1967 it was converted to a light water breeder reactor and continued its operation until 1985, when the reactor was shut down. However, the decommissioning planning for Shippingport was begun in 1979. Detailed engineering and planning was undertaken to look at alternatives for disposal of the reactor vessel, the overall detailed estimated costs, the exposure to the workers and the waste volume generated and to prepare activity specifications for performance of the work. The program scope and component removal are detailed. The scarification of contaminated concrete, building demolition, special tools and equipment needed and work performance data are described. The successful removal of the primary system components and piping has been completed. (author)

  10. Proceedings of the US Nuclear Regulatory Commission fifteenth water reactor safety information meeting: Volume 6, Decontamination and decommissioning, accident management, TMI-2

    International Nuclear Information System (INIS)

    Weiss, A.J.

    1988-02-01

    This six-volume report contains 140 papers out of the 164 that were presented at the Fifteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 26-29, 1987. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. This report, Volume 6, discusses decontamination and decommissioning, accident management, and the Three Mile Island-2 reactor accident. Thirteen reports have been cataloged separately

  11. Sodium components cleaning status in the Italian fast reactor program

    Energy Technology Data Exchange (ETDEWEB)

    De Luca, B [CNEN-RIT/MAT - Laboratorio Sviluppo Processi - C.S.N. Cassacia, Rome (Italy); Labanti, V [CNEN-DRV, Bologna (Italy); Mennucci, M [NIRA, Genoa (Italy)

    1978-08-01

    As a consequence of the Italian Fast Reactor Development, mainly aimed to the PEC project and to the participation in the French Superphenix project, it is of increasing importance to set up a reliable method for specific reactor components and related test loops. The first problem was the cleaning of the PEC fuelling machine. In order to perform the routine maintenance of the machine an alcohol cleaning method based on the use of 2-butoxyethanol-NN dimethylformamide mixture has been proposed.

  12. Nondestructive testing of nuclear reactor components integrity

    International Nuclear Information System (INIS)

    Mala, M.; Miklos, M.

    2011-01-01

    Nuclear energy must respond to current challenges in the energy market. The significant parameters are increase of the nuclear fuel price, closed fuel cycle, reduction and safe and the final disposal of high level radioactive waste. Nowadays, the discussions on suitable energy mix are taking place not only here in Czech Republic, but also in many other European countries. It is necessary to establish an appropriate ratio among the production of electricity from conventional, nuclear and renewable energy sources. Also, it is necessary to find ways how to streamline the economy, central part of the nuclear fuel cycle and thereby to increase the competitiveness of nuclear energy. This streamlining can be carried out by improving utilization of existing nuclear fuel with maintaining a high degree of nuclear facilities safety. Increasing operational reliability and safety together with increasing utilization of nuclear fuel place increasing demands on monitoring of changes during fuel burnup. The potential fuel assembly damages in light water reactors are prevented by the introduction of new procedures and programs of the fuel assembly monitoring. One of them is the Post Irradiation Inspection Program (PIIP) which is a good tool for monitoring of chemical regime impact on the fuel assembly cladding behavior. Main nondestructive techniques that are used at nuclear power plants for the fuel assembly integrity evaluation are ultrasonic measurements, eddy current measurements, radiographic testing, acoustic techniques and others. Ultrasonic system is usual tool for leak fuel rod evaluation and it is also used at Temelin NPP. Since 2009, Temelin NPP has cooperated with Research Center Rez Ltd in frame of PIIP program at both units WWER 1000. This program was established for US VVantage6 fuel assemblies and also it continues for Russian TVSA-T fuel assemblies. (author)

  13. Radiological impact on the workers, members of the public, and environment from the partial decommissioning of Pakistan Research Reactor-I and its associated radioactive residues.

    Science.gov (United States)

    Ali, A; Orfi, S D; Manzur, H; Aslam, M

    2001-05-01

    The Pakistan Research Reactor-I (PARR-I) is a swimming pool type research reactor originally designed and built for a thermal power of 5 MW using High Enriched Uranium (HEU) fuel. In 1990-1991 the reactor was redesigned, partially decommissioned and recommissioned to operate with Low Enriched Uranium (LEU) fuel at a thermal power of 10 MW. An essential requirement, construction and commissioning of a wet spent fuel storage bay and fabrication of an irradiated fuel transfer cask were completed before actual dismantling of the reactor core. During the partial decommissioning operations, radioactive waste generated included 600 m3 low-level liquid radioactive waste and 14 m3 of solid radioactive waste with an average specific activity of 4.52 Bq ml(-1) and 2.22 kBq g(-1), respectively. External radiation doses of the workers were determined using TLD (NG 6,7) and direct reading dosimeters. The maximum individual external radiation dose received by any worker during this practice was 5 mSv, which was 25% of the annual dose limit of 20 mSv. Detection and measurement of internal contamination was carried out using bioassay techniques. During the whole operation, not a single case of internal contamination was detected. The ambient radiation levels around waste seepage pits are periodically monitored using TLD (G-2 cards) and G. M. radiation survey meters. Underground migration of radioactivity is checked by analyzing seepage water samples taken from boreholes that have been dug at different locations in the vicinity of the radioactive residues. The monitoring around disposal sites containing radioactive residues has been continued during the last 9 y and will be continued in the future. So far, no rise in the environmental gamma radiation dose level and migration of underground radionuclides has been found in the vicinity of these disposal sites. Working personal during the decommissioning of PARR-I have been found to be radiologically safe. Adherence to the ALARA

  14. Overview of decommissioning activities in the US

    International Nuclear Information System (INIS)

    LaGuardia, T.S.

    2006-01-01

    The U.S. has been involved the successful decontamination, decommissioning and reutilization of nuclear facilities for over 20 years. A number of commercial power plants in the United States have either completed their decommissioning, or will be in the next few years. In addition, the U.S. DOE has taken an aggressive approach to site remediation focusing on site closures so as to better utilize its financial resources. The U.S. initiative to deregulate the electrical generation industry to promote competition and presumably to reduce electricity prices to the consumer, is again in flux. Some utilities, facing the real or perceived threat of competition in its markets decided to shut down the 'costly' nuclear plants to alleviate the drain on their financial reserves. The older nuclear units experienced serious mechanical problems, entailing expensive repairs and replacements. Such difficulties have caused owner-operator utilities to decide to decommission these facilities rather than incur the expense of upgrading or repairing the plants to meet current regulatory and design criteria. Plants that were marginally cost-competitive, or not at all competitive, were shut down and decommissioned. Other utilities have bought some of the older nuclear plants in the Northeast (a high power demand region) to operate them and to extend their licenses for continued life. This paper will discuss the decommissioning lessons learned, management approaches, site characterization and challenges faced in disposition of radioactive waste and large components, contracting practice, and the status of several of these shut down reactor-decommissioning programs. The industry has proven that nuclear power plants can be cost effectively and safely decommissioned. (author)

  15. Decommissioning in western Europe

    International Nuclear Information System (INIS)

    Lundqvist, K.

    1999-12-01

    This report gives an overview of the situation in Western Europe. The original aim was to focus on organisational and human issues with regard to nuclear reactor decommissioning, but very few articles were found. This is in sharp contrast to the substantial literature on technical issues. While most of the reports on decommissioning have a technical focus, several provide information on regulatory issues, strategies and 'state of the art'. The importance of the human and organizational perspective is however discovered, when reading between the lines of the technical publications, and especially when project managers summarize lessons learned. The results are to a large extent based on studies of articles and reports, mainly collected from the INIS database. Decommissioning of nuclear facilities started already in the sixties, but then mainly research and experimental facilities were concerned. Until now about 70 reactors have been shutdown world-wide. Over the years there have been plenty of conferences for exchanging experiences mostly about technical matters. Waste Management is a big issue. In the 2000s there will be a wave of decommissioning when an increasing amount of reactors will reach the end of their calculated lifetime (40 years, a figure now being challenged by both life-extension and pre-shutdown projects). Several reactors have been shut-down for economical reasons. Shutdown and decommissioning is however not identical. A long period of time can sometimes pass before an owner decides to decommission and dismantle a facility. The conditions will also differ depending on the strategy, 'immediate dismantling' or 'safe enclosure'. If immediate dismantling is chosen the site can reach 'green-field status' in less than ten years. 'Safe enclosure', however, seems to be the most common strategy. There are several pathways, but in general a safe store is constructed, enabling the active parts to remain in safe and waterproof conditions for a longer period of

  16. Decommissioning nuclear installations

    International Nuclear Information System (INIS)

    Dadoumont, J.

    2010-01-01

    When a nuclear installation is permanently shut down, it is crucial to completely dismantle and decontaminate it on account of radiological safety. The expertise that SCK-CEN has built up in the decommissioning operation of its own BR3 reactor is now available nationally and internationally. Last year SCK-CEN played an important role in the newly started dismantling and decontamination of the MOX plant (Mixed Oxide) of Belgonucleaire in Dessel, and the decommissioning of the university research reactor Thetis in Ghent.

  17. Safety problems in decommissioning nuclear power plants

    International Nuclear Information System (INIS)

    Auler, I.; Bardtenschlager, R.; Gasch, A.; Majohr, N.

    1975-12-01

    The safety problems at decommissioning are illustrated by the example of a LWR with 1300 MW electric power after 40 years of specified normal operation. For such a facility the radioactivity in the form of activation and contamination one year after being finally taken out of service is in the order of magnitude of 10 7 Ci, not counting the fuel assemblies. The dose rates occurring during work on the reactor vessel at nozzle level may amount to some 10 4 rem/h. After a rough estimation the accumulated dose for the decommissioning personnel during total dismantling will be about 1200 rem. During performance of the decommissioning activities the problems are mainly caused by direct radiation of the active components and systems and by the release of radioactive particles, aerosols and liquids if these components are crushed. The extent of later dismantling problems may be reduced by selecting appropriate materials as well as considering the requirements for dismantling in design and arrangement of the components already in the design stage of new facilities. Apart from plant design also the concept for the disposal of the radioactive waste from decommissioning will provide important boundary conditions. E.g. the maximum size of the pieces to be stored in the ultimate storage place will very much influence the dose expenditure for handling these parts. For complete dismantling of nuclear power plants an ultimate store must be available where large amounts of bulky decommissioning waste, containing relatively low activity, can be stored. The problems and also the cost for decommissioning may be considerably reduced by delaying complete disposal of the radioactive material >= 40 years and during this period, keeping the radioactivity enclosed within the plant in the form of a safe containment. (orig./HP) [de

  18. BNFL decommissioning strategy and techniques

    International Nuclear Information System (INIS)

    Taylor, D.

    2002-01-01

    This paper provides an overview of the range of reactor decommissioning projects being managed by BNFL, both on its own sites and for other client organizations in the UK and abroad. It also describes the decommissioning strategies and techniques that have been developed by BNFL and adopted in order to carry out this work

  19. UK fast reactor components - sodium removal decontamination and requalification

    International Nuclear Information System (INIS)

    Donaldson, D.M.; Bray, J.A.; Newson, I.H.

    1978-01-01

    Over the past two decades extensive experience on sodium removal techniques has been gained at the UKAEA's Dounreay Nuclear Establishment from both the Dounreay Fact Reactor (DFR) and the Prototype Fast Reactor (PFR). This experience has created confidence that complex components can be cleaned of sodium, maintenance or repair operations carried out, and the components successfully re-used. Part 2 of the paper, which describes recent operations associated with the PFR, demonstrates the background to these views. This past and continuing experience is being used in forming the basis of the plant to be provided for sodium removal, decontamination and requalification of components in the UK's future commercial fast reactors. Further improvements in techniques and in component designs can be expected in the course of the next few years. Consequently UK philosophy and approach with respect to maintenance and repair operations is sufficiently flexible to enable relevant improvements to be incorporated into the next scheduled fast reactor - the Commercial Demonstration Fast Reactor (CUR). This paper summarises the factors which are being taken into consideration in this continuously advancing field

  20. UK fast reactor components - sodium removal decontamination and requalification

    Energy Technology Data Exchange (ETDEWEB)

    Donaldson, D M [FRDD, UKAEA, Risley (United Kingdom); Bray, J A; Newson, I H [UKAEA, Dounreay Nuclear Power Establishment, Thurso (United Kingdom)

    1978-08-01

    Over the past two decades extensive experience on sodium removal techniques has been gained at the UKAEA's Dounreay Nuclear Establishment from both the Dounreay Fact Reactor (DFR) and the Prototype Fast Reactor (PFR). This experience has created confidence that complex components can be cleaned of sodium, maintenance or repair operations carried out, and the components successfully re-used. Part 2 of the paper, which describes recent operations associated with the PFR, demonstrates the background to these views. This past and continuing experience is being used in forming the basis of the plant to be provided for sodium removal, decontamination and requalification of components in the UK's future commercial fast reactors. Further improvements in techniques and in component designs can be expected in the course of the next few years. Consequently UK philosophy and approach with respect to maintenance and repair operations is sufficiently flexible to enable relevant improvements to be incorporated into the next scheduled fast reactor - the Commercial Demonstration Fast Reactor (CUR). This paper summarises the factors which are being taken into consideration in this continuously advancing field.

  1. Decommissioning Work Modeling System for Nuclear Facility Decommissioning Design

    International Nuclear Information System (INIS)

    Park, S. K.; Cho, W. H.; Choi, Y. D.; Moon, J. K.

    2012-01-01

    During the decommissioning activities of the KRR-1 and 2 (Korea Research Reactor 1 and 2) and UCP (Uranium Conversion Plant), all information and data, which generated from the decommissioning project, were record, input and managed at the DECOMMIS (DECOMMissioning Information management System). This system was developed for the inputting and management of the data and information of the man-power consumption, operation time of the dismantling equipment, the activities of the radiation control, dismantled waste management and Q/A activities. When a decommissioning is planed for a nuclear facility, an investigation into the characterization of the nuclear facility is first required. The results of such an investigation are used for calculating the quantities of dismantled waste volume and estimating the cost of the decommissioning project. That is why, the DEFACS (DEcommissioning FAcility Characterization DB System) was established for the management of the facility characterization data. The DEWOCS (DEcommissioning WOrk-unit productivity Calculation System) was developed for the calculation of the workability on the decommissioning activities. The work-unit productivities are calculated through this system using the data from the two systems, DECOMMIS and DEFACS. This result, the factors of the decommissioning work-unit productivities, will be useful for the other nuclear facility decommissioning planning and engineering. For this, to set up the items and plan for the decommissioning of the new objective facility, the DEMOS (DEcommissioning work Modeling System) was developed. This system is for the evaluation the cost, man-power consumption of workers and project staffs and technology application time. The factor of the work-unit productivities from the DEWOCS and governmental labor cost DB and equipment rental fee DB were used for the calculation the result of the DEMOS. And also, for the total system, DES (Decommissioning Engineering System), which is now

  2. Current status of Chernobyl NPP decommissioning

    International Nuclear Information System (INIS)

    2009-01-01

    Strategy of Chernobyl NPP decommissioning with the decommissioning license 2002-2064 is presented. The main activities at the stage of ChNPP units shutdown (2002 - 2012) are: units maintenance in safe state; decommissioning infrastructure construction; unloading of SNF – main activity determining the stage duration; systems and elements final shutdown; decommissioning life-support systems reconstruction; Comprehensive engineering and radiation survey (CERS); dismantling of the reactor facilities external equipment; removal of RAW from units; decommissioning documentation development. The decommissioning activities main results are presented

  3. The decommissioning of Berkeley II

    International Nuclear Information System (INIS)

    Hannan, A.

    2002-01-01

    This paper describes the decommissioning progress at the Magnox site at Berkeley in Gloucestershire.Throughout the work at Berkeley the emphasis has been on conducting decommissioning safely. This has been reflected in the progress of decommissioning starting with removal of the fuel from site and thus much greater than 99% of the radioactive inventory. The major radioactive hazard is the Intermediate Level Waste in the form of fuel element debris (graphite struts and extraneous magnox components removed to increase the packing density of fuel elements in flasks going to Sellafield), miscellaneous activated components, sludges and resins. Approximately 1500 m 3 of such material exists and is stored in underground waste vaults on site. Work is underway to recover and encapsulate the waste in cement so rendering it 'passively safe'. All work on site is covered by a nuclear safety case which has a key objective of minimising the radiological exposures that could accrue to workers. Reflecting this an early decision has been taken to leave work on the Reactor Pressure Vessels themselves for several decades. Also important in protection of the workforce has been control of asbestos.Much material has been removed with redundant plant and equipment, but a programme of remediation in line with government legislation has been required to ensure personnel safety throughout the decommissioning period and into Care and Maintenance.In addition to health and safety matters the site approach to environmental issues has been consistent. Formally such standards as ISO 14001 have been adhered to and the appropriate certification maintained. At a working level the principles of reduce, reuse and recycle have been inculcated

  4. Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Owen, M.B.

    1997-04-01

    This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future.

  5. Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1

    International Nuclear Information System (INIS)

    Owen, M.B.

    1997-04-01

    This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future

  6. Decommissioning, Dismantling and Disarming: a Unique Information Showroom Inside the G2 Reactor at Marcoule Centre (France) - 12068

    Energy Technology Data Exchange (ETDEWEB)

    Volant, Emmanuelle [CEA DAM, Bruyeres-le-Chatel (France); Garnier, Cedric [CEA DEN, Marcoule (France)

    2012-07-01

    The paper aims at presenting the new information showroom called 'Escom G2' (for 'Espace Communication') inaugurated by the French Atomic Energy and Alternative Energies Commission (CEA) in spring 2011. This showroom is settled directly inside the main building of the G2 nuclear reactor: a facility formerly dedicated to weapon-grade plutonium production since the late 1950's at the Marcoule nuclear centre, in south of France. After its shutdown, and reprocessing of the last spent fuels, a first dismantling step was successfully completed from 1986 to 1996. Unique in France and in Europe, Escom G2 is focused on France dismantling expertise and its action for disarmament. This showroom comprises of a 300-square meters permanent exhibition, organized around four themes: France strategy for disarmament, decommissioning and dismantling technical aspects, uranium and plutonium production cycles. Each of these topics is illustrated with posters, photos, models and technical pieces from the dismantled plants. It is now used to present France's action in disarmament to highly ranked audiences such as: state representatives, diplomats, journalists... The paper explains the background story of this original project. As a matter of fact, in 1996 France was the first nuclear state to decide to shut down and dismantle its fissile material production facilities for nuclear weapons. First, the paper presents the history of the G2 reactor in the early ages of Marcoule site, its operating highlights as well as its main dismantling operations, are presented. In Marcoule, where the three industrial-scale reactors G1, G2 and G3 used to be operated for plutonium production (to be then reprocessed in the nearby UP1 plant), the initial dismantling phase has now been completed (in 1980's for G1 and in 1996 for G2 and G3). The second phase, aimed at completely dismantling these three reactors, will restart in 2020, and is directly linked to the opening of

  7. Assessment of foreign decommissioning technology with potential application to US decommissioning needs

    International Nuclear Information System (INIS)

    Allen, R.P.; Konzek, G.J.; Schneider, K.J.; Smith, R.I.

    1987-09-01

    This study was conducted by the Pacific Northwest Laboratory (PNL) for the US Department of Energy (DOE) to identify and technically assess foreign decommissioning technology developments that may represent significant improvements over decommissioning technology currently available or under development in the United States. Technology need areas for nuclear power reactor decommissioning operations were identified and prioritized using the results of past light water reactor (LWR) decommissioning studies to quantitatively evaluate the potential for reducing cost and decommissioning worker radiation dose for each major decommissioning activity. Based on these identified needs, current foreign decommissioning technologies of potential interest to the US were identified through personal contacts and the collection and review of an extensive body of decommissioning literature. These technologies were then assessed qualitatively to evaluate their uniqueness, potential for a significant reduction in decommissioning costs and/or worker radiation dose, development status, and other factors affecting their value and applicability to US needs

  8. Safety problems in decommissioning of nuclear power plants

    International Nuclear Information System (INIS)

    1975-12-01

    The safety problems in decommissioning are presented by the example of light water reactors with an electric power of 1300 MW and 40 years of preceding specified operation. In such a plant the radioactivity in the form of activation and contamination is of the order of 10 7 Ci one year after final shut-down. The fuel elements are not taken into account. During the work at the reactor vessel dose rates of some 10 4 rem/h may occur at the flange level. According to a rough estimation the dose accumulated by the decommissioning personnel during dismantling of the radioactive components amounts to 1200 rem. During the decommissioning work the problems are caused predominantly by the direct radiation from the radioactive components and systems as well as from the release of radioactive particles, aerosols and liquids on cutting them up. In designing new plants the extent of later decommissioning problems can be reduced above all by selection of suitable materials and by decommissioning-minded design and arrangement of the components and parts of the plant. (orig./RW) [de

  9. Characteristics of outage radiation fields around various reactor components

    International Nuclear Information System (INIS)

    Verzilov, Y.; Husain, A.; Corbin, G.

    2008-01-01

    Full text: Activity monitoring surveys, consisting of gamma spectroscopy and dose rate measurements, of various CANDU station components such as the reactor face, feeder cabinet, steam generators and moderator heat exchangers are often performed during shutdown in order to trend the transport of activity around the primary heat transport and moderator systems. Recently, the increased dose expenditure for work such as feeder inspection and replacement in the reactor vault has also spurred interest in improved characterization of the reactor face fields to facilitate better ALARA decision making and hence a reduction in future dose expenditures. At present, planning for reactor face work is hampered by insufficient understanding of the relative contribution of the various components to the overall dose. In addition to the increased dose expenditure for work at the reactor face, maintenance work associated with horizontal flux detectors and liquid injection systems has also resulted in elevated dose expenditures. For instance at Darlington, radiation fields in the vicinity of horizontal flux detectors (HFD) and Liquid Injection Shutdown System (LISS) nozzle bellows are trending upwards with present contact fields being in the range 16-70 rem/h and working distance fields being in the range 100-500 mrem/h. This paper presents findings based on work currently being funded by the CANDU Owners Group. Measurements were performed at Ontario Power Generation's Pickering and Darlington nuclear stations. Specifically, the following are addressed: Characteristics of Reactor Vault Fields; Characteristics of Steam Generator Fields; Characteristics of Moderator Heat Exchanger Fields. Measurements in the reactor vault were performed at the reactor face, along the length of end fittings, along the length of feeders, at the bleed condenser and at the HFD and LISS nozzle bellows. Steam generator fields were characterized at various elevations above the tube sheet, with and without the

  10. Shielding requirements for the transport of nuclear warhead components under decommissioning

    International Nuclear Information System (INIS)

    Hansen, L.F.

    1994-09-01

    The requirements to carry out accurate shielding calculations involved with the safe off-site transportation of packages containing nuclear warhead components, special assemblies and radioactive materials are discussed. The need for (a) detailed information on the geometry and material composition of the packaging and radioactive load, (b) accurate representation of the differential energy spectra (dN/dE) for the neutron and gamma spectra emitted by the radioactive materials enclosed in the packaging, (c) well-tested neutron and photon cross section libraries, (d) and accurate three-dimensional Monte Carlo transport codes are illustrated. A brief discussion of the need for reliable dose measurements is presented

  11. Replacement of core components in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Durney, J.L.; Croucher, D.W.

    1990-01-01

    The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. The methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and is expected to continue operation for at least and additional 25 years. Aging evaluations are in progress to address additional replacements that may be needed during this period

  12. Replacement of core components in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Durney, J.L.; Croucher, D.W.

    1989-01-01

    The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. Methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and will continue operation for perhaps another 20 years. Aging evaluations are in program to address additional replacements that may be needed during this extended time period. 3 figs

  13. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FORTHE DECOMMISSIONING OF THE FUEL CHANNELS IN THECANDU NUCLEAR REACTOR

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2013-05-01

    Full Text Available As many nuclear power plants are reaching their end of lifecycle, the decommissioning of theseinstallations has become one of the 21stcentury’s great challenges. Each project may be managed differently,depending on the country, development policies, financial considerations, and the availability of qualifiedengineers or specialized companies to handle such projects. The principle objective of decommissioning is toplace a facility into such a condition that there is no unacceptable risk from the decommissioned facility topublic health and safety of the environment. In order to ensure that at the end of its life the risk from a facility iswithin acceptable bounds, action is normally required. The overall decommissioning strategy is todeliver a timely, cost-effective program while maintaining high standards of safety, security and environmentalprotection. If facilities were not decommissioned, they could degrade and potentially present an environmentalradiological hazard in the future. Simply abandoning or leaving a facility after ceasing operations is notconsidered to be an acceptable alternative to decommissioning. The final aim of decommissioning is torecover the geographic site to its original condition.

  14. Ageing management program for reactor components in HANARO

    International Nuclear Information System (INIS)

    Cho, Yeong-Garp; Wu, Sang-Ik; Lee, Jung-Hee; Ryu, Jeong-Soo; Park, Yong-Chul; Wu, Jong-Sup; Jun, Byung Jin

    2003-01-01

    The HANARO, an open-tank-in-pool type research reactor of 30MWth power in Korea, has operated for 8 years since its initial criticality in February of 1995. The reactor power has been gradually increased to 24 MWth through the service period. Therefore the reactor age is very young from the viewpoint of the ageing effect on the reactor structure and components by neutron irradiation considering the expected reactor lifetime. But, we have a few programs to manage the ageing from the aspect of design lifetime of reactor components. This paper summarizes the overall progress and plan for the ageing management for the reactor components including lifetime extension and design improvement, remote measurements and in-service inspections. The shutoff units and control absorber units have aged more rapidly than other structures or components because the number of rod drop cycles was higher than expected at the design stage. The system commissioning tests, periodic performance tests, and weekly operation for the stable supply of medical radioisotopes overriding the normal cycle operation have contributed to the high frequency of rod drop. Therefore, we have instituted a program to extend the lifetime of the shutoff units and the control absorber units. This program includes an endurance test to verify the performance for the extended number of drops and the management of shutdown methods to minimize the drop cycles for both the shutoff units and the control absorber units. The program also includes the design improvement of the damper mechanism of the control absorber units to reduce the impact force caused by rod drop. The inner shell of the reflector vessel surrounding the core is the most critical part from the viewpoint of neutron irradiation. The periodic measurement of the dimensional change in the vertical straightness of the inner shell is considered as one of the in-service inspections. We developed a few tools and verified the performance to measure the

  15. The main objectives of lifetime management of reactor unit components

    International Nuclear Information System (INIS)

    Dragunov, Y.; Kurakov, Y.

    1998-01-01

    The main objectives of the work concerned with life management of reactor components in Russian Federation are as follows: development of regulations in the field of NPP components ageing and lifetime management; investigations of ageing processes; residual life evaluation taking into account the actual state of NPP systems, real loading conditions and number of load cycles, results of in-service inspections; development and implementation of measures for maintaining/enhancing the NPP safety

  16. Experimental Facilities for Performance Evaluation of Fast Reactor Components

    International Nuclear Information System (INIS)

    Chandramouli, S.; Kumar, V.A. Suresh; Shanmugavel, M.; Vijayakumar, G.; Vinod, V.; Noushad, I.B.; Babu, B.; Kumar, G. Padma; Nashine, B.K.; Rajan, K.K.

    2013-01-01

    Brief details about various experimental facilities catering to the testing and performance evaluation requirements of fast reactor components have been brought out. These facilities have been found to be immensely useful to continue research and development activities in the areas of component development and testing, sodium technology, thermal hydraulics and sodium instrumentation for the SFR’s. In addition new facilities which have been planned will be of great importance for the developmental activities related to future SFR’s

  17. Design of sodium cooled reactor systems and components for maintainability

    International Nuclear Information System (INIS)

    Carr, R.W.; Charnock, H.O.; McBride, J.P.

    1978-09-01

    Special maintenability problems associated with the design and operation of sodium cooled reactor plants are discussed. Some examples of both good and bad design practice are introduced from the design of the FFTF plant and other plants. Subjects include design for drainage, cleaning, decontamination, access, component removal, component disassembly and reassembly, remote tooling, jigs, fixtures, and design for minimizing radiation exposure of maintenance personnel. Check lists are included

  18. Metal plutonium conversion to components of nuclear reactor fuel

    International Nuclear Information System (INIS)

    Subbotin, V.G.; Panov, A.V.; Mashirev, V.P.

    2000-01-01

    Capabilities of different technologies for plutonium conversion to the fuel components of nuclear reactors are studied. Advantages and shortcomings of aqueous and nonaqueous methods of plutonium treatment are shown. Proposals to combine and coordinate efforts of world scientific and technological community in solving problems concerning plutonium of energetic and weapon origin treatment were put forward. (authors)

  19. Chemical immobilization of fission products reactive with nuclear reactor components

    International Nuclear Information System (INIS)

    Grossman, L.N.; Kaznoff, A.I.; Clukey, H.V.

    1975-01-01

    This invention teaches a method of immobilizing deleterious fission products produced in nuclear fuel materials during nuclear fission chain reactions through the use of additives. The additives are disposed with the nuclear fuel materials in controlled quantities to form new compositions preventing attack of reactor components, especially nuclear fuel cld, by the deleterious fission products. (Patent Office Record)

  20. Fabrication of high performance components for Indian nuclear reactors

    International Nuclear Information System (INIS)

    Jayaraj, R.N.

    2011-01-01

    Nuclear Fuel Complex (NFC), a Unit of the Department of Atomic Energy (DAE) has been engaged for well over three-and-half decades in the manufacture of fuels for Pressurized Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs). All the fuel assembly components, like, fuel clad tubes, end plugs, spacers, spacer grids etc. are also being manufactured at NFC in Zirconium alloy material. Apart from the regular production of these components and finished fuel assemblies, NFC has also been engaged in the production of Zirconium alloy reactor core structurals, like, pressure tubes, calandria tubes, garter springs and reactivity control mechanisms for PHWRs and square channels for BWRs. While all these structural components are produced through standardized flow sheets, there have been continuous innovations carried out in the processes to meet the ever increasing end-use characteristics laid down by the utilities. The paper enumerates various aspects of different technologies developed at NFC for the manufacture of high performance components for reactor applications

  1. Metal plutonium conversion to components of nuclear reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Subbotin, V.G.; Panov, A.V. [Russian Federal Nuclear Center, ALL-Russian Science and Research, Institute of Technical Physics, Snezhinsk (Russian Federation); Mashirev, V.P. [ALL-Russian Science and Research Institute of Chemical Technology, Moscow (Russian Federation)

    2000-07-01

    Capabilities of different technologies for plutonium conversion to the fuel components of nuclear reactors are studied. Advantages and shortcomings of aqueous and nonaqueous methods of plutonium treatment are shown. Proposals to combine and coordinate efforts of world scientific and technological community in solving problems concerning plutonium of energetic and weapon origin treatment were put forward. (authors)

  2. Decommissioning: the final folly

    International Nuclear Information System (INIS)

    Dibdin, T.

    1990-01-01

    The Second International Seminar on Decommissioning of Nuclear Facilities held in London is reviewed. Various solutions to the reactor decommissioning, including isolating the reactor core, and turning the surrounding buildings into a theme park, are mentioned. The International Atomic Energy Agency identifies three decommissioning stages. Stage 1, defuelling; Stage 2 dismounting of non-radioactive plant with isolation of the nuclear island and Stage 3, return to a 'green field' site. The real debate is about waste management and timing of the stages - whether to defer Stage 3 for a century or so, or even whether to attempt Stage 3 at all. Cost estimation is also discussed. In the United Kingdom, the timing of completion of the deep repository for high level waste will affect the timing. (UK)

  3. Influence of design features on decommissioning of a large fast breeder reactor

    International Nuclear Information System (INIS)

    Fournie, J.-L.; Alary, C.; Maire, D.; Seroux, N. de; Peyrard, G.

    1990-01-01

    The evolution of FBR design in Europe shows that pool-type design will become the reference design for future FBR and the projected European Fast Reactor (EFR) is based on this concept. The identification of design features shows that the main contributors of the sodium and structures activity are the Co60 for gamma radiation source and low decay, Ni63, Nb94 and Ni59 for long time decay. So, the technical benefits of a Co content reduction are interesting for the high activated structures and for diagrid thimbles coating and we made proposals to lower Co content in steels or alloys and to substitute coatings. We identify measures which must facilitate both the sodium draining and the reactor block and internal cleaning: all which improve the gravity draining and the downing of the sodium flow make easier the penetration of cleaning products. The features, connected with the dismantling of the very activated internal structures, of the roof and of the lay-out, are mentioned. (author)

  4. Decommissioning of nuclear facilities: 'it can and has been done'

    International Nuclear Information System (INIS)

    2009-01-01

    Considerable international experience gained over the last 20 years demonstrates that nuclear facilities can be safely dismantled and decommissioned once a decision is made to cease operations and permanently shut them down. The term decommissioning is used to describe all the management and technical actions associated with ceasing operation of a nuclear installation and its subsequent dismantling to facilitate its removal from regulatory control (de-licensing). These actions involve decontamination of structures and components, dismantling of components and demolition of buildings, remediation of any contaminated ground and removal of the resulting waste. Worldwide, of the more than 560 commercial nuclear power plants that are or have been in operation, about 120 plants have been permanently shut down and are at some stage of decommissioning. About 10% of all shutdown plants have been fully decommissioned, including eight reactors of more than 100 MWe. A larger number of various types of fuel cycle and research facilities have also been shut down and decommissioned, including: facilities for the extraction and enrichment of uranium, facilities for fuel fabrication and reprocessing, laboratories, isotope production facilities and particle accelerators. This brochure looks at decommissioning across a spectrum of nuclear facilities and shows worldwide examples of successful projects. Further information can be found in NEA publications and on a number of web-sites

  5. Generic component reliability data for research reactor PSA

    International Nuclear Information System (INIS)

    1997-02-01

    The purpose of this document is to provide reference generic component-reliability information for a variety of research reactor types. As noted in Section 2 and Table IV, component data accumulated over many years is in the database. It is expected that the report should provide representative data which will remain valid for a number of years. The database provides component failure rates on a time and/or demand related basis according to the operational modes of the components. No update of the database is presently planned. As a result of the implementation of data collection systems in the research reactors represented in these studies, updating of data from individual facilities could be made available by the contributing research reactor facilities themselves. As noted in Section 1.1, the report does not include a detailed discussion of information regarding component classification and reliability parameter definitions. The report does provide some insights and discussions regarding the practicalities of the data collection process and some guidelines for database usage. 9 refs, 7 tabs

  6. Examination of core components removed from CANDU reactors

    International Nuclear Information System (INIS)

    Cheadle, B.A.; Coleman, C.E.; Rodgers, D.K.; Davies, P.H.; Chow, C.K.; Griffiths, M.

    1988-11-01

    Components in the core of a nuclear reactor degrade because the environment is severe. For example, in CANDU reactors the pressure tubes must contend with the effects of hot pressurised water and damage by a flux of fast neutrons. To evaluate any deterioration of components and determine the cause of the occasional failure, we have developed a wide range of remote-handling techniques to examine radioactive materials. As well as pressure tubes, we have examined calandria tubes, garter springs, end fittings, liquid-zone control units and flux detectors. The results from these examinations have produced solutions to problems and continually provide information to help understand the processes that may limit the lifetime of a component

  7. Non-aqueous removal of sodium from reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Welch, F H; Steele, O P [Rockwell International, Atomics International Division, Canoga Park (United States)

    1978-08-01

    Reactor components from sodium-cooled systems. whether radioactive or not, must have the sodium removed before they can be safely handled for 1) disposal, 2) examination and test, or 3) decontamination, repair, and requalification. In the latter two cases, the sodium must be removed in a manner which will not harm the component. and prevent future use. Two methods for sodium removal using non-aqueous techniques have been studied extensively in the U.S.A. in the past few years: the Alcohol Process, which uses a fully denatured ethanol to react away the sodium; and the Evaporative Process, which uses heat and vacuum to evaporate the sodium from the component.

  8. Non-aqueous removal of sodium from reactor components

    International Nuclear Information System (INIS)

    Welch, F.H.; Steele, O.P.

    1978-01-01

    Reactor components from sodium-cooled systems. whether radioactive or not, must have the sodium removed before they can be safely handled for 1) disposal, 2) examination and test, or 3) decontamination, repair, and requalification. In the latter two cases, the sodium must be removed in a manner which will not harm the component. and prevent future use. Two methods for sodium removal using non-aqueous techniques have been studied extensively in the U.S.A. in the past few years: the Alcohol Process, which uses a fully denatured ethanol to react away the sodium; and the Evaporative Process, which uses heat and vacuum to evaporate the sodium from the component

  9. Studies on components for a molten salt reactor

    International Nuclear Information System (INIS)

    Nejedly, M.; Matal, O.

    2003-01-01

    The aim is contribute to a design of selected components of molten salt reactors with fuel in the molten fluoride salt matrix. Molten salt reactors (MSRs) permit the utilization of plutonium and minor actinides as new nuclear fuel from a traditional nuclear power station with production of electric energy. Results of preliminary feasibility studies of an intermediate heat exchanger, a small power molten salt pump and a modular conception of a steam generator for a demonstration unit of the MSR (30 MW) are summarized. (author)

  10. Solutions against PWSCC in dissimilar welds cracks of reactor components

    International Nuclear Information System (INIS)

    Schlader, D.; Michaut, B.; Knapp, M

    2005-01-01

    This article provides a brief overview of the experience accumulated by Framatome ANP in the development and use of repair and mitigation techniques of the PWSCC in dissimilar welds cracks of reactor components. A brief description of the alternatives available to the industry for the solution of this problem for both PWR and BWR reactor types is also included. These solutions have been implemented many times by Framatome ANP in Europe and the US. The article also describes the way the know-how is shared among the different regions of the company in order to offer customer specific solutions. (Author)

  11. A plan for safety and integrity of research reactor components

    International Nuclear Information System (INIS)

    Moatty, Mona S. Abdel; Khattab, M.S.

    2013-01-01

    Highlights: ► A plan for in-service inspection of research reactor components is put. ► Section XI of the ASME Code requirements is applied. ► Components subjected to inspection and their classes are defined. ► Flaw evaluation and its acceptance–rejection criteria are reviewed. ► A plan of repair or replacement is prepared. -- Abstract: Safety and integrity of a research reactor that has been operated over 40 years requires frequent and thorough inspection of all the safety-related components of the facility. The need of increasing the safety is the need of improving the reliability of its systems. Diligent and extensive planning of in-service inspection (ISI) of all reactor components has been imposed for satisfying the most stringent safety requirements. The Safeguards Officer's responsibilities of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code ASME Code have been applied. These represent the most extensive and time-consuming part of ISI program, and identify the components subjected to inspection and testing, methods of component classification, inspection and testing techniques, acceptance/rejection criteria, and the responsibilities. The paper focuses on ISI planning requirements for welded systems such as vessels, piping, valve bodies, pump casings, and control rod-housing parts. The weld in integral attachments for piping, pumps, and valves are considered too. These are taken in consideration of safety class (1, 2, 3, etc.), reactor age, and weld type. The parts involve in the frequency of inspection, the examination requirements for each inspection, the examination method are included. Moreover the flaw evaluation, the plan of repair or replacement, and the qualification of nondestructive examination personnel are considered

  12. Residual radioactivity guidelines for the heavy water components test reactor at the Savannah River Site

    International Nuclear Information System (INIS)

    Owen, M.B. Smith, R.; McNeil, J.

    1997-04-01

    Guidelines were developed for acceptable levels of residual radioactivity in the Heavy Water Components Test Reactor (HWCTR) facility at the conclusion of its decommissioning. Using source terms developed from data generated in a detailed characterization study, the RESRAD and RASRAD-BUILD computer codes were used to calculate derived concentration guideline levels (DCGLs) for the radionuclides that will remain in the facility. The calculated DCGLs, when compared to existing concentrations of radionuclides measured during a 1996 characterization program, indicate that no decontamination of concrete surfaces will be necessary. Also, based on the results of the calculations, activated concrete in the reactor biological shield does not have to be removed, and imbedded radioactive piping in the facility can remain in place. Viewed in another way, the results of the calculations showed that the present inventory of residual radioactivity in the facility (not including that associated with the reactor vessel and steam generators) would produce less than one millirem per year above background to a hypothetical individual on the property. The residual radioactivity is estimated to be approximately 0.04 percent of the total inventory in the facility as of March, 1997. According to the results, the only radionuclides that would produce greater than 0.0.1-millirem per year are Am-241 (0.013 mrem/yr at 300 years), C-14 (0.022 mrem/yr at 1000 years) and U-238 (0.034 mrem/yr at 6000 years). Human exposure would occur only through the groundwater pathways, that is, from water drawn from, a well on the property. The maximum exposure would be approximately one percent of the 4 millirem per year ground water exposure limit established by the U.S. Environmental Protection Agency. 11 refs., 13 figs., 15 tabs

  13. Assessment of derived emission limits for radioactive effluents resulted from the decommissioning activities of the VVR-S nuclear research reactor

    International Nuclear Information System (INIS)

    Tuca, C.; Stochioiu, A.; Sahagia, M.; Gurau, D.; Dragusin, M.

    2015-01-01

    This paper presents complex studies on establishment of derived emission limits for potential radionuclides emitted as gaseous and liquid effluents, during the decommissioning activities (2nd and 3rd phases) of a nuclear research reactor, cooled and moderated with distilled water, type VVR-S, owned by the IFIN-HH. In the present paper there are described: the analysis methods and equipment used, the methodologies for calculating doses and the Derived Emission Limits (DEL), the experimentally measured activities of the representative radionuclides found in gaseous and liquid effluents resulted from decommissioning activities, as well as the effective derived limits of liquid and gaseous effluents, applying the calculation methodologies, specific to critical categories of exposed subjects. A constraint effective dose limit for a person from the critical group of 50 μSv/year was considered in calculations. From the comparison of the two series of values, measured released activities and DELs, there has been concluded that for the gaseous effluents they comply with the DELs, while in the case of liquid effluents they are higher and consequently they must be treated as liquid radioactive wastes. - Highlights: • The DELs for gaseous and liquid effluents in decommissioning were studied. • The impact on the environment and critical group was assessed. • Gamma-ray spectrometry was used to determine the radionuclide composition. • Based on the dosimetry models, the values of conservative DELs were calculated. • The DELs are compliant for gases, not for liquids; require the treatment as rad-waste

  14. Decommissioning Handbook

    International Nuclear Information System (INIS)

    Cusack, J.G.; Dalfonso, P.H.; Lenyk, R.G.

    1994-01-01

    The Decommissioning Handbook provides technical guidance on conducting decommissioning projects. Information presented ranges from planning logic, regulations affecting decommissioning, technology discussion, health and safety requirements, an developing a cost estimate. The major focus of the handbook are the technologies -- decontamination technologies, waste treatment, dismantling/segmenting/demolition, and remote operations. Over 90 technologies are discussed in the handbook providing descriptions, applications, and advantages/disadvantages. The handbook was prepared to provide a compendium of available or potentially available technologies in order to aid the planner in meeting the specific needs of each decommissioning project. Other subjects presented in the Decommissioning Handbook include the decommissioning plan, characterization, final project configuration based planning, environmental protection, and packaging/transportation. These discussions are presented to complement the technologies presented in the handbook

  15. Decommissioning project of commercial nuclear power plant

    International Nuclear Information System (INIS)

    Karigome, S.

    2008-01-01

    Decommissioning project of commercial nuclear power plant in Japan was outlined. It is expected that the land, after the decommissioning of commercial nuclear power plants, will serve as sites for new plants. Steps will be taken to reduce the amount of wastes generated and to recycle/reuse them. Wastes with a radioactivity concentration below the 'clearance level' need not be dealt with as radioactive material, and may be handled in the same way as conventional wastes. The Tokai-1 power station, a 166 MWe carbon dioxide cooled reactor which closed down in 1998, is being decommissioned and the first ten years as 'safe storage' to allow radioactivity to decay. Non-reactor grade components such as turbines were already removed, heat exchanger dismantling started and the reactor will be dismantled, the buildings demolished and the site left ready for reuse. All radioactive wastes will be classified as low-level wastes in three categories and will be buried under the ground. The total cost will be 88.5 billion yen -34.7 billion for dismantling and 53.8 billion for waste treatment including the graphite moderator. (T. Tanaka)

  16. Challenges in design of zirconium alloy reactor components

    International Nuclear Information System (INIS)

    Kakodkar, Anil; Sinha, R.K.

    1992-01-01

    Zirconium alloy components used in core-internal assemblies of heavy water reactors have to be designed under constraints imposed by need to have minimum mass, limitations of fabrication, welding and joining techniques with this material, and unique mechanisms for degradation of the operating performance of these components. These constraints manifest as challenges for design and development when the size, shape and dimensions of the components and assemblies are unconventional or untried, or when one is aiming for maximization of service life of these components under severe operating conditions. A number of such challenges were successfully met during the development of core-internal components and assemblies of Dhruva reactor. Some of the then untried ideas which were developed and successfully implemented include use of electron beam welding, cold forming of hemispherical ends of reentrant cans, and a large variety of rolled joints of innovative designs. This experience provided the foundation for taking up and successfully completing several tasks relating to coolant channels, liquid poison channels and sparger channels for PHWRs and test sections for the in-pile loops of Dhruva reactor. For life prediction and safety assessment of coolant channels of PHWRs some analytical tools, notably, a computer code for prediction of creep limited life of coolant channels has been developed. Some of the future challenges include the development of easily replaceable coolant channels and also large diameter coolant channels for Advanced Heavy Water Reactor, and development of solutions to overcome deterioration of service life of coolant channels due to hydriding. (author). 5 refs., 13 figs., 1 tab

  17. Blue Ribbon Commission, Yucca Mountain Closure, Court Actions - Future of Decommissioned Reactors, Operating Reactors and Nuclear Power - 13249

    International Nuclear Information System (INIS)

    Devgun, Jas S.

    2013-01-01

    Issues related to back-end of the nuclear fuel cycle continue to be difficult for the commercial nuclear power industry and for the decision makers at the national and international level. In the US, the 1982 NWPA required DOE to develop geological repositories for SNF and HLW but in spite of extensive site characterization efforts and over ten billion dollars spent, a repository opening is nowhere in sight. There has been constant litigation against the DOE by the nuclear utilities for breach of the 'standard contract' they signed with the DOE under the NWPA. The SNF inventory continues to rise both in the US and globally and the nuclear industry has turned to dry storage facilities at reactor locations. In US, the Blue Ribbon Commission on America's Nuclear Future issued its report in January 2012 and among other items, it recommends a new, consent-based approach to siting of facilities, prompt efforts to develop one or more geologic disposal facilities, and prompt efforts to develop one or more consolidated storage facilities. In addition, the March 2011 Fukushima Daiichi accident had a severe impact on the future growth of nuclear power. The nuclear industry is focusing on mitigation strategies for beyond design basis events and in the US, the industry is in the process of implementing the recommendations from NRC's Near Term Task Force. (authors)

  18. Decommissioning of the Nuclear Reactors R2 and R2-0 at Studsvik, Sweden. General Data as called for under Article 37 of the Euratom Treaty

    Energy Technology Data Exchange (ETDEWEB)

    2009-01-15

    This document describes the plans for decommissioning of the nuclear research and material test reactors R2 and R2-0, situated at the Studsvik site close to the city of Nykoeping, Sweden. The purpose of the document is to serve as information for the European Commission, and to fulfil the requirements of Article 37 of the Euratom Treaty. Studsvik is situated on the Baltic coast, about 20 km east of Nykoeping and 80 km southwest of Stockholm. The site comprises the reactors R2 and R2-0 and several facilities for material investigation and radioactive waste treatment and storage. The reactors were used for a number of different purposes from 1960 until June 2005, when they were shut down following a decision by the operator. Decommissioning of the reactor facility is planned to be completed in 2016 after dismantling and conditioning of radioactive parts and demolition of the facility. Solid and liquid radioactive wastes from the dismantling activities will be treated and stored on-site awaiting final disposal. The waste treatment facilities, which are situated in other buildings at the Studsvik site, are planned to continue operation during and after the decommissioning of the reactor facility. All nuclear fuel has been transferred to a separate storage facility and is being shipped to the US according to existing agreements. The objective of the planned dismantling activities is to achieve clearance of the facility to make it possible to either demolish the buildings or use them for other purposes. The operator has divided the planning for dismantling and demolition of the facility into three phases [1]: Dismantling 1, including primary system decontamination, dismantling of the reactors with systems in the reactor pool, draining, cleaning and temporary covering of the reactor pool. This phase has begun and is due to last till approximately December 2009. Dismantling 2, including dismantling of systems in the reactor facility, removal of equipment, radiological

  19. Statistical techniques for the identification of reactor component structural vibrations

    International Nuclear Information System (INIS)

    Kemeny, L.G.

    1975-01-01

    The identification, on-line and in near real-time, of the vibration frequencies, modes and amplitudes of selected key reactor structural components and the visual monitoring of these phenomena by nuclear power plant operating staff will serve to further the safety and control philosophy of nuclear systems and lead to design optimisation. The School of Nuclear Engineering has developed a data acquisition system for vibration detection and identification. The system is interfaced with the HIFAR research reactor of the Australian Atomic Energy Commission. The reactor serves to simulate noise and vibrational phenomena which might be pertinent in power reactor situations. The data acquisition system consists of a small computer interfaced with a digital correlator and a Fourier transform unit. An incremental tape recorder is utilised as a backing store and as a means of communication with other computers. A small analogue computer and an analogue statistical analyzer can be used in the pre and post computational analysis of signals which are received from neutron and gamma detectors, thermocouples, accelerometers, hydrophones and strain gauges. Investigations carried out to date include a study of the role of local and global pressure fields due to turbulence in coolant flow and pump impeller induced perturbations on (a) control absorbers, (B) fuel element and (c) coolant external circuit and core tank structure component vibrations. (Auth.)

  20. Decommissioning Handbook

    Energy Technology Data Exchange (ETDEWEB)

    1994-03-01

    The Decommissioning Handbook is a technical guide for the decommissioning of nuclear facilities. The decommissioning of a nuclear facility involves the removal of the radioactive and, for practical reasons, hazardous materials to enable the facility to be released and not represent a further risk to human health and the environment. This handbook identifies and technologies and techniques that will accomplish these objectives. The emphasis in this handbook is on characterization; waste treatment; decontamination; dismantling, segmenting, demolition; and remote technologies. Other aspects that are discussed in some detail include the regulations governing decommissioning, worker and environmental protection, and packaging and transportation of the waste materials. The handbook describes in general terms the overall decommissioning project, including planning, cost estimating, and operating practices that would ease preparation of the Decommissioning Plan and the decommissioning itself. The reader is referred to other documents for more detailed information. This Decommissioning Handbook has been prepared by Enserch Environmental Corporation for the US Department of Energy and is a complete restructuring of the original handbook developed in 1980 by Nuclear Energy Services. The significant changes between the two documents are the addition of current and the deletion of obsolete technologies and the addition of chapters on project planning and the Decommissioning Plan, regulatory requirements, characterization, remote technology, and packaging and transportation of the waste materials.

  1. Decommissioning of fuel PIE caves at Berkeley Nuclear Laboratories

    International Nuclear Information System (INIS)

    Brant, A.W.

    1990-01-01

    This paper describes the first major contract awarded to private industry to carry out decommissioning of a facility with significant radiation levels. The work required operatives to work in pressurised suits, entry times were significantly affected by sources of radiation in the Caves, being as low as thirty minutes per day initially. The Caves at Berkeley Nuclear Laboratories carry out post irradiation examination of fuel elements support units and reactor core components from CEGB power stations. The decommissioning work is part of an overall refurbishment of the facility to allow the receipt of AGR Fuel Stringer Component direct from power stations. The paper describes the decommissioning and decontamination of the facility from the remote removal and clean up work carried out by the client to the hands-on work. It includes reference to entry times, work patterns, interfaces with the client and the operations of the laboratory. Details of a specially adapted size reduction method are given. (Author)

  2. Scale modeling flow-induced vibrations of reactor components

    International Nuclear Information System (INIS)

    Mulcahy, T.M.

    1982-06-01

    Similitude relationships currently employed in the design of flow-induced vibration scale-model tests of nuclear reactor components are reviewed. Emphasis is given to understanding the origins of the similitude parameters as a basis for discussion of the inevitable distortions which occur in design verification testing of entire reactor systems and in feature testing of individual component designs for the existence of detrimental flow-induced vibration mechanisms. Distortions of similitude parameters made in current test practice are enumerated and selected example tests are described. Also, limitations in the use of specific distortions in model designs are evaluated based on the current understanding of flow-induced vibration mechanisms and structural response

  3. Vinca nuclear decommissioning program

    International Nuclear Information System (INIS)

    Pesic, M.; Subotic, K.; Sotic, O.; Plecas, I.; Ljubenov, V.; Peric, A.

    2002-01-01

    In this paper a preliminary program for the nuclear decommissioning in The Vinca Institute of Nuclear Sciences is presented. Proposed Projects and Activities, planned to be done in the next 10 years within the frames of the Program, should improve nuclear and radiation safety and should solve the main problems that have arisen in the previous period. Project of removal of irradiated spent nuclear fuel from the RA reactor, as a first step in all possible