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Sample records for handling offsite nuclear

  1. Online Decision Support System (IRODOS) - an emergency preparedness tool for handling offsite nuclear emergency

    International Nuclear Information System (INIS)

    Vinod Kumar, A.; Oza, R.B.; Chaudhury, P.; Suri, M.; Saindane, S.; Singh, K.D.; Bhargava, P.; Sharma, V.K.

    2009-01-01

    A real time online decision support system as a nuclear emergency response system for handling offsite nuclear emergency at the Nuclear Power Plants (NPPs) has been developed by Health, Safety and Environment Group, Bhabha Atomic Research Centre (BARC), Department of Atomic Energy (DAE) under the frame work of 'Indian Real time Online Decision Support System 'IRODOS'. (author)

  2. Operating experience feedback on lose of offsite power supply for nuclear power plant

    International Nuclear Information System (INIS)

    Jiao Feng; Hou Qinmai; Che Shuwei

    2013-01-01

    The function of the service power system of a nuclear power plant is to provide safe and reliable power supply for the nuclear power plant facilities. The safety of nuclear power plant power supply is essential for nuclear safety. The serious accident of Fukushima Daiichi nuclear power plant occurred due to loss of service power and the ultimate heat sink. The service power system has two independent offsite power supplies as working power and auxiliary power. This article collected events of loss of offsite power supply in operating nuclear power plants at home and abroad, and analyzed the plant status and cause of loss of offsite power supply events, and proposed improvement measures for dealing with loss of offsite power supply. (authors)

  3. Off-site nuclear emergency exercises in Japan

    International Nuclear Information System (INIS)

    Eiji, U.; Kiyoshi, T.; Masao, O.; Shigeru, F.

    1993-01-01

    Nuclear emergency planning and preparedness in Japan have been organized by both national and local governments based on the Disaster Countermeasures Basic Act. Off-site nuclear emergency exercises are classified into two types: national-government level exercises and local-government level exercises. National-government level exercises are carried out once a year by the competent national authorities. Among these authorities, the Science and Technology Agency (STA) fills a leading position in the Japanese nuclear emergency planning and preparedness. Local-government level exercises are carried out once a year or once in a few years by the local governments of the prefectures where nuclear facilities are located. Most of the off-site nuclear emergency exercises in Japan are performed by local-governments. The aim of these exercises is to reinforce the skills of the emergency staff. The national government (STA etc.) provides advices and assistance including financial support to the local-governments. Emergency exercises with the participation of residents have been carried out in some local-governments. As an example of local-government level exercises, an experience in Shizuoka prefecture (central part of Japan) is presented

  4. The knowledge-based off-site emergency response system for a nuclear power plant

    International Nuclear Information System (INIS)

    Ho, L.W.; Loa, W.W.; Wang, C.L.

    1987-01-01

    A knowledge-based expert system for a nuclear power plant off-site emergency response system is described. The system incorporates the knowledge about the nuclear power plant behaviours, site environment and site geographic factors, etc. The system is developed using Chinshan nuclear power station of Taipower Company, Taiwan, ROC as a representative model. The objectives of developing this system are to provide an automated intelligent system with functions of accident simulation, prediction and with learning capabilities to supplement the actions of the emergency planners and accident managers in order to protect the plant personnel and the surrounding population, and prevent or mitigate property damages resulting from the plant accident. The system is capable of providing local and national authorities with rapid retrieval data from the site characteristics and accident progression. The system can also provide the framework for allocation of available resources and can handle the uncertainties in data and models

  5. Principles of off-site nuclear emergency exercises

    International Nuclear Information System (INIS)

    Miska, H.

    2011-01-01

    Due to high safety standards at nuclear power plants, no experience exits with nuclear emergencies in Western Europe. Thus, emergency exercises are the only possibility to assure effective protective measures should the very unlikely severe accident occur. The main objectives of exercises are generally the check of response plans for suitability, the test of the equipment's applicability and training of personnel for the unusual task to manage a nuclear emergency. After an introduction into the different types of exercises, this contribution focuses on offsite nuclear emergency exercises, explaining frame conditions to ensure good practice and, finally, reports some experience from exercises. (orig.)

  6. Losses of off-site power at U.S. nuclear power plants -- through 1995. Final report

    International Nuclear Information System (INIS)

    Wyckoff, H.

    1996-04-01

    This report provides a database and summary analysis of losses of off-site power at US nuclear generating units. It includes the 16 years 1980 through 1995. This is the twelfth update of this database and analysis. During 1994 there were no losses of all off-site power and in 1995 only two short losses. Both the short term and long term US loss of all off-site power experience is extremely favorable. The frequency of losing all off-site power is an important input to many nuclear plant safety assessments. The industry's loss of all off-site power experience that is set forth in this report can provide perspective to plant specific probabilistic safety assessments

  7. A re-evaluation of nuclear plant offsite power supplies

    International Nuclear Information System (INIS)

    William E Berger; Robert E Henry

    2005-01-01

    Full text of publication follows: De-regulation of the electric power industry has resulted in separate ownership of the transmission and power generation facilities as well as a revised format for operating the transmission facilities. Currently we see the transfer of large blocks of bulk power between markets which can impact the voltage regulation at the offsite power supply. Where Nuclear Plant operations once knew with a large degree of certainty the operating range of the system supplying the offsite power supply, this may no longer be the case and more challenges to the safety systems could result. These challenges may manifest themselves as either a loss of offsite power or voltage levels approaching the degraded level setpoints. In this paper we will first explore what challenges are caused by deregulation and how they impact offsite power supply operations. Next we will incorporate the knowledge grained regarding accidents and consequences from the Individual Plant Evaluations (IPE's) to see how the offsite power supply could be operated to mitigate the challenges and extend the capacity of the auxiliary power system. Various scenarios will be examined using the Modular Accident Analysis Program (MAAP) as an integral plant model. MAAP simulations that include both the plant thermal hydraulic responses and corresponding electric power demand are presented to demonstrate the impact of alternate approaches to offsite power system operation. The original design phase of the offsite and onsite power distribution system was based on a criterion relating to the starting of all safety loads if a safety injection signal was present independent of the accident or its progression. The IPE and risk informed insights that are readily available today will be applied in the re-analyses of the offsite distribution system response. (authors)

  8. Receiving Basin for Offsite Fuels and the Resin Regeneration Facility Safety Analysis Report, Executive Summary

    International Nuclear Information System (INIS)

    Shedrow, C.B.

    1999-01-01

    The Safety Analysis Report documents the safety authorization basis for the Receiving Basin for Offsite Fuels (RBOF) and the Resin Regeneration Facility (RRF) at the Savannah River Site (SRS). The present mission of the RBOF and RRF is to continue in providing a facility for the safe receipt, storage, handling, and shipping of spent nuclear fuel assemblies from power and research reactors in the United States, fuel from SRS and other Department of Energy (DOE) reactors, and foreign research reactors fuel, in support of the nonproliferation policy. The RBOF and RRF provide the capability to handle, separate, and transfer wastes generated from nuclear fuel element storage. The DOE and Westinghouse Savannah River Company, the prime operating contractor, are committed to managing these activities in such a manner that the health and safety of the offsite general public, the site worker, the facility worker, and the environment are protected

  9. Receiving Basin for Offsite Fuels and the Resin Regeneration Facility Safety Analysis Report, Executive Summary

    Energy Technology Data Exchange (ETDEWEB)

    Shedrow, C.B.

    1999-11-29

    The Safety Analysis Report documents the safety authorization basis for the Receiving Basin for Offsite Fuels (RBOF) and the Resin Regeneration Facility (RRF) at the Savannah River Site (SRS). The present mission of the RBOF and RRF is to continue in providing a facility for the safe receipt, storage, handling, and shipping of spent nuclear fuel assemblies from power and research reactors in the United States, fuel from SRS and other Department of Energy (DOE) reactors, and foreign research reactors fuel, in support of the nonproliferation policy. The RBOF and RRF provide the capability to handle, separate, and transfer wastes generated from nuclear fuel element storage. The DOE and Westinghouse Savannah River Company, the prime operating contractor, are committed to managing these activities in such a manner that the health and safety of the offsite general public, the site worker, the facility worker, and the environment are protected.

  10. Alternative off-site power supply improves nuclear power plant safety

    International Nuclear Information System (INIS)

    Gjorgiev, Blaže; Volkanovski, Andrija; Kančev, Duško; Čepin, Marko

    2014-01-01

    Highlights: • Additional power supply for mitigation of the station blackout event in NPP is used. • A hydro power plant is considered as an off-site alternative power supply. • An upgrade of the probabilistic safety assessment from its traditional use is made. • The obtained results show improvement of nuclear power plant safety. - Abstract: A reliable power system is important for safe operation of the nuclear power plants. The station blackout event is of great importance for nuclear power plant safety. This event is caused by the loss of all alternating current power supply to the safety and non-safety buses of the nuclear power plant. In this study an independent electrical connection between a pumped-storage hydro power plant and a nuclear power plant is assumed as a standpoint for safety and reliability analysis. The pumped-storage hydro power plant is considered as an alternative power supply. The connection with conventional accumulation type of hydro power plant is analysed in addition. The objective of this paper is to investigate the improvement of nuclear power plant safety resulting from the consideration of the alternative power supplies. The safety of the nuclear power plant is analysed through the core damage frequency, a risk measure assess by the probabilistic safety assessment. The presented method upgrades the probabilistic safety assessment from its common traditional use in sense that it considers non-plant sited systems. The obtained results show significant decrease of the core damage frequency, indicating improvement of nuclear safety if hydro power plant is introduced as an alternative off-site power source

  11. Collection and evaluation of complete and partial losses of off-site power at nuclear power plants

    International Nuclear Information System (INIS)

    Battle, R.E.

    1985-02-01

    Events involving loss of off-site power that have occurred at nuclear power plants through 1983 are described and categorized as complete or partial losses. The events were identified as plant-centered or grid-related failures. In addition, the causes of the failures were classified as weather, human error, design error, or hardware failure. The plant-centered failures were usually of shorter duration than the weather-related grid failures. For this reason, the weather-related events were reviewed in detail. Design features that may be important factors affecting off-site power system reliability were tabulated for most of the operating nuclear power plants. The tabulated information was provided to NRC for a statistical analysis to determine the importance of these design features for losses of off-site power. The frequency of losses of off-site power versus duration was estimated for three time periods. The frequency of loss of off-site power was estimated to be 0.09/reactor-year based on industry-wide data for the years 1959 through 1983

  12. Evolution of Onsite and Offsite Power Systems in US Nuclear Power Plants

    International Nuclear Information System (INIS)

    Mathew, Roy K.

    2015-01-01

    The AC electric power system is the source of power for station auxiliaries during normal operation and for the reactor protection system and emergency safety features during abnormal and accident conditions. Since the construction of early plants in US, the functional adequacy and requirements of the offsite power systems, safety and non safety related onsite electric power systems have changed considerably to ensure that these systems have adequate redundancy, independence, quality, maintenance and testability to support safe shutdown of the nuclear plant. The design of AC systems has evolved from a single train to multiple (up to four) redundant trains in the current evolutionary designs coupled with other auxiliary AC systems. The early plants were designed to cope with a Loss of Offsite Power (LOOP) event through the use of onsite power supplies only. However operating experience has indicated that onsite and offsite power AC power systems can fail due to natural phenomena (earthquakes, lightning strikes, fires, geomagnetic storms, tsunamis, etc.) or operational abnormalities such as loss of a single phase, switching surges or human error. The onsite DC systems may not be adequately sized to support plant safe shutdown over an extended period if AC power cannot be restored within a reasonable time. This paper will discuss the requirements to improve availability and reliability of offsite and onsite alternating current (AC) power sources to U.S. Nuclear Power Plants. In addition, the paper will discuss the requirements and guidance beyond design basis events. (author)

  13. Assessment of off-site consequences of nuclear accidents (MARIA)

    International Nuclear Information System (INIS)

    Haywood, S.M.

    1985-01-01

    A brief report is given of a workshop held in Luxembourg in 1985 on methods for assessing the off-site radiological consequences of nuclear accidents (MARIA). The sessions included topics such as atmospheric dispersion; foodchain transfer; urban contamination; demographic and land use data; dosimetry, health effects, economic and countermeasures models; uncertainty analysis; and application of probabilistic risk assessment results as input to decision aids. (U.K.)

  14. Planning and exercise experiences related to an off-site nuclear emergency in Canada: the federal component

    International Nuclear Information System (INIS)

    Eaton, R.S.

    1986-01-01

    The Canadian Government's Federal Nuclear Emergency Response Plan (off-site) (FNERP) was issued in 1984. In this plan, a nuclear emergency is defined as an emergency involving the release of radionuclides but does not include the use of nuclear weapons against North America. Because of the federal nature of Canada and its large area, special considerations are required for the plan to cover both the response to nuclear emergencies where the national government has primary responsibility and to provincial requests for assistance where the federal response becomes secondary to the provincial. The nuclear emergencies requiring the implementation of this plan are: (a) an accident in the nuclear energy cycle in Canada with off-site implications; (b) an accident in the nuclear energy cycle in another country which may affect Canada; (c) nuclear weapons testing with off-site implications which may affect Canada; and (d) nuclear-powered devices impacting on Canadian territory. Each emergency requires a separate sub-plan and usually requires different organizations to respond. Some scenarios are described. The Department of National Health and Welfare has established a Federal Nuclear Emergency Control Centre (FNECC). The FNECC participated in September 1985 in an exercise involving a nuclear reactor facility in the Province of Ontario and the experience gained from this activity is presented. The FNECC co-operates with its counterparts in the United States of America through a nuclear emergency information system and this network is also described. (author)

  15. Project Execution Plan for the Remote Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    Danny Anderson

    2014-07-01

    As part of ongoing cleanup activities at the Idaho National Laboratory (INL), closure of the Radioactive Waste Management Complex (RWMC) is proceeding under the Comprehensive Environmental Response, Compensation, and Liability Act (42 USC 9601 et seq. 1980). INL-generated radioactive waste has been disposed of at RWMC since 1952. The Subsurface Disposal Area (SDA) at RWMC accepted the bulk of INL’s contact and remote-handled low-level waste (LLW) for disposal. Disposal of contact-handled LLW and remote-handled LLW ion-exchange resins from the Advanced Test Reactor in the open pit of the SDA ceased September 30, 2008. Disposal of remote-handled LLW in concrete disposal vaults at RWMC will continue until the facility is full or until it must be closed in preparation for final remediation of the SDA (approximately at the end of fiscal year FY 2017). The continuing nuclear mission of INL, associated ongoing and planned operations, and Naval spent fuel activities at the Naval Reactors Facility (NRF) require continued capability to appropriately dispose of contact and remote handled LLW. A programmatic analysis of disposal alternatives for contact and remote-handled LLW generated at INL was conducted by the INL contractor in Fiscal Year 2006; subsequent evaluations were completed in Fiscal Year 2007. The result of these analyses was a recommendation to the Department of Energy (DOE) that all contact-handled LLW generated after September 30, 2008, be disposed offsite, and that DOE proceed with a capital project to establish replacement remote-handled LLW disposal capability. An analysis of the alternatives for providing replacement remote-handled LLW disposal capability has been performed to support Critical Decision-1. The highest ranked alternative to provide this required capability has been determined to be the development of a new onsite remote-handled LLW disposal facility to replace the existing remote-handled LLW disposal vaults at the SDA. Several offsite DOE

  16. Effects of spent fuel types on offsite consequences of hypothetical accidents

    International Nuclear Information System (INIS)

    Courtney, J. C.; Dwight, C. C.; Lehto, M. A.

    2000-01-01

    Argonne National Laboratory (ANL) conducts experimental work on the development of waste forms suitable for several types of spent fuel at its facility on the Idaho National Engineering and Environmental Laboratory (INEEL) located 48 km West of Idaho Falls, ID. The objective of this paper is to compare the offsite radiological consequences of hypothetical accidents involving the various types of spent nuclear fuel handled in nonreactor nuclear facilities. The highest offsite total effective dose equivalents (TEDEs) are estimated at a receptor located about 5 km SSE of ANL facilities. Criticality safety considerations limit the amount of enriched uranium and plutonium that could be at risk in any given scenario. Heat generated by decay of fission products and actinides does not limit the masses of spent fuel within any given operation because the minimum time elapsed since fissions occurred in any form is at least five years. At cooling times of this magnitude, fewer than ten radionuclides account for 99% of the projected TEDE at offsite receptors for any credible accident. Elimination of all but the most important nuclides allows rapid assessments of offsite doses with little loss of accuracy. Since the ARF (airborne release fraction), RF (respirable fraction), LPF (leak path fraction) and atmospheric dilution factor (χ/Q) can vary by orders of magnitude, it is not productive to consider nuclides that contribute less than a few percent of the total dose. Therefore, only 134 Cs, 137 Cs- 137m Ba, and the actinides significantly influence the offsite radiological consequences of severe accidents. Even using highly conservative assumptions in estimating radiological consequences, they remain well below current Department of Energy guidelines for highly unlikely accidents

  17. Core management and fuel handling for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    requirements and safety precautions for transport beyond the site, off-site storage and ultimate disposal of irradiated fuel and core components are beyond the scope of this publication. Aspects of fuel accounting not directly related to nuclear safety are not considered in this publication. Thus, safeguards are beyond the scope of this publication. Section 2 provides guidance on the core management programme. Section 3 identifies the main aspects of handling and storage of fresh fuel. Section 4 gives guidance on the implementation of the refuelling programme. Section 5 explains the different aspects of handling, storage and inspection of irradiated fuel. Section 6 deals with the handling of core components, in particular those that have been irradiated. Section 7 provides general recommendations on the preparatory arrangements in connection with fuel dispatch from the site. Sections 8 and 9 give guidance on administrative and organizational arrangements for fuel management, including general aspects of documentation

  18. Core management and fuel handling for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    requirements and safety precautions for transport beyond the site, off-site storage and ultimate disposal of irradiated fuel and core components are beyond the scope of this publication. Aspects of fuel accounting not directly related to nuclear safety are not considered in this publication; thus, safeguards are beyond the scope of this publication. Section 2 provides guidance on the core management programme. Section 3 identifies the main aspects of handling and storage of fresh fuel. Section 4 gives guidance on the implementation of the refuelling programme. Section 5 explains the different aspects of handling, storage and inspection of irradiated fuel. Section 6 deals with the handling of core components, in particular those that have been irradiated. Section 7 provides general recommendations on the preparatory arrangements in connection with fuel dispatch from the site. Sections 8 and 9 give guidance on administrative and organizational arrangements for fuel management, including general aspects of documentation

  19. Nuclear power plants in Germany. Recent developments in off-site nuclear emergency preparedness and response; Kernkraftwerke in Deutschland. Neue Entwicklungen im anlagenexternen Notfallschutz

    Energy Technology Data Exchange (ETDEWEB)

    Gering, Florian [Bundesamt fuer Strahlenschutz, Oberschleissheim/Neuherberg (Germany). Abt. SW 2.2 Entscheidungshilfesysteme, Lageermittlung und Kommunikation

    2014-10-15

    The reactor accident in Fukushima, Japan, in 2011 triggered a thorough review of the off-site emergency preparedness and response for nuclear power plants in Germany. ''Off-site emergency preparedness and response'' includes all actions to protect the public outside the fence of a nuclear power plant. This review resulted in several changes in off-site emergency preparedness and response, which are briefly described in this article. Additionally, several recent activities are described which may influence emergency preparedness and response in the future.

  20. Analysis of offsite dose calculation methodology for a nuclear power reactor

    International Nuclear Information System (INIS)

    Moser, D.M.

    1995-01-01

    This technical study reviews the methodology for calculating offsite dose estimates as described in the offsite dose calculation manual (ODCM) for Pennsylvania Power and Light - Susquehanna Steam Electric Station (SSES). An evaluation of the SSES ODCM dose assessment methodology indicates that it conforms with methodology accepted by the US Nuclear Regulatory Commission (NRC). Using 1993 SSES effluent data, dose estimates are calculated according to SSES ODCM methodology and compared to the dose estimates calculated according to SSES ODCM and the computer model used to produce the reported 1993 dose estimates. The 1993 SSES dose estimates are based on the axioms of Publication 2 of the International Commission of Radiological Protection (ICRP). SSES Dose estimates based on the axioms of ICRP Publication 26 and 30 reveal the total body estimates to be the most affected

  1. Benchmarking of RESRAD-OFFSITE : transition from RESRAD (onsite) to RESRAD-OFFSITE and comparison of the RESRAD-OFFSITE predictions with peercodes

    International Nuclear Information System (INIS)

    Yu, C.; Gnanapragasam, E.; Cheng, J.-J.; Biwer, B.

    2006-01-01

    The main purpose of this report is to document the benchmarking results and verification of the RESRAD-OFFSITE code as part of the quality assurance requirements of the RESRAD development program. This documentation will enable the U.S. Department of Energy (DOE) and its contractors, and the U.S. Nuclear Regulatory Commission (NRC) and its licensees and other stakeholders to use the quality-assured version of the code to perform dose analysis in a risk-informed and technically defensible manner to demonstrate compliance with the NRC's License Termination Rule, Title 10, Part 20, Subpart E, of the Code of Federal Regulations (10 CFR Part 20, Subpart E); DOE's 10 CFR Part 834, Order 5400.5, ''Radiation Protection of the Public and the Environment''; and other Federal and State regulatory requirements as appropriate. The other purpose of this report is to document the differences and similarities between the RESRAD (onsite) and RESRAD-OFFSITE codes so that users (dose analysts and risk assessors) can make a smooth transition from use of the RESRAD (onsite) code to use of the RESRAD-OFFSITE code for performing both onsite and offsite dose analyses. The evolution of the RESRAD-OFFSITE code from the RESRAD (onsite) code is described in Chapter 1 to help the dose analyst and risk assessor make a smooth conceptual transition from the use of one code to that of the other. Chapter 2 provides a comparison of the predictions of RESRAD (onsite) and RESRAD-OFFSITE for an onsite exposure scenario. Chapter 3 documents the results of benchmarking RESRAD-OFFSITE's atmospheric transport and dispersion submodel against the U.S. Environmental Protection Agency's (EPA's) CAP88-PC (Clean Air Act Assessment Package-1988) and ISCLT3 (Industrial Source Complex-Long Term) models. Chapter 4 documents the comparison results of the predictions of the RESRAD-OFFSITE code and its submodels with the predictions of peer models. This report was prepared by Argonne National Laboratory's (Argonne

  2. Development of a decision support system for off-site emergency management in the early phase of a nuclear accident

    International Nuclear Information System (INIS)

    Datta, D.; Sharma, R.M.

    2002-01-01

    Full text: Experience gained after the Chernobyl accident clearly demonstrated the importance of improving administrative, organizational and technical emergency management arrangements in India. The more important areas where technical improvements were needed were early warning monitoring, communication networks for the rapid and reliable exchange of radiological and other information and decision support systems for off-site emergency management. A PC based artificial intelligent software has been developed to have a decision support system that can easily implement to manage off-site nuclear emergency and subsequently analyze the off-site consequences of the nuclear accident. A decision support tool, STEPS (source term estimate based on plant status), that provides desired input to the present software was developed. The tool STEPS facilitates meta knowledge of the system. The paper describes the details of the design of the software, functions of various modules, tuning of respective knowledge base and overall its scope in real sense in nuclear emergency preparedness and response

  3. Techniques and decision making in the assessment of off-site consequences of an accident in a nuclear facility

    International Nuclear Information System (INIS)

    1987-01-01

    This Guide is intended to complement the IAEA's existing technical guidance on emergency planning and preparedness by providing information and practical guidance related to the assessment of off-site consequences of an accident in a nuclear or radioactive materials installation and to the decision making process in implementing protective measures. This Guide contains information on emergency response philosophy, fundamental factors affecting accident consequences, principles of accident assessment, data acquisition and handling, systems, techniques and decision making principles. Many of the accident assessment concepts presented are considerably more advanced than some of those that now pertain in most countries. They could, if properly interpreted, developed and applied, significantly improve emergency response in the early and intermediate phases of an accident. Furthermore, they are considered to be applicable to a broad range of serious nuclear accidents and radiological emergencies. The extent of their application is governed by both the scale of the accident and by the availability of preplanned resources for accident assessment and emergency response. 68 refs, 28 figs, 14 tabs

  4. The modelling of off-site economic consequences of nuclear accidents

    International Nuclear Information System (INIS)

    Alonso, A.; Gallego, E.; Martin, J.E.

    1991-01-01

    The paper presents a computer model for the probabilistic assessment of the off-site economic risk derived from nuclear accidents. The model is called MECA (Model for Economic Consequence Assessment) and takes into consideration the direct costs caused, following an accident, by the different countermeasures adopted to prevent both the early and chronic exposure of the population to the radionuclides released, as well as the direct costs derived from health damage to the affected population. The model uses site-specific data that are organized in a socio-economic data base; detailed distributions of population, livestock census, agricultural production and farmland use, as well as of employment, salaries, and added value for different economic sectors are included. This data base has been completed for Spain, based on available official statistics. The new code, coupled to a general ACA code, provides capability to complete probabilistic risk assessments from the point of view of the off-site economic consequences, and also to perform cost-effectiveness analysis of the different countermeasures in the field of emergency preparedness

  5. Offsite environmental monitoring report; radiation monitoring around United States nuclear test areas, Calendar Year 1996

    International Nuclear Information System (INIS)

    Davis, M.G.; Flotard, R.D.; Fontana, C.A.; Huff, P.A.; Maunu, H.K.; Mouck, T.L.; Mullen, A.A.; Sells, M.D.

    1997-08-01

    This report describes the Offsite Radiation Safety Program. This laboratory operated an environmental radiation monitoring program in the region surrounding the Nevada Test Site (NTS) and at former test sites in Alaska, Colorado, Mississippi, Nevada, and New Mexico. The surveillance program is designed to measure levels and trends of radioactivity, if present, in the environment surrounding testing areas to ascertain whether current radiation levels and associated doses to the general public are in compliance with existing radiation protection standards. The surveillance program additionally has the responsibility to take action to protect the health and well being of the public in the event of any accidental release of radioactive contaminants. Offsite levels of radiation and radioactivity are assessed by sampling milk, water, and air; by deploying thermoluminescent dosimeters (TLDs); and using pressurized ionization chambers (PICs). No nuclear weapons testing was conducted in 1996 due to the continuing nuclear test moratorium. During this period, R and IE personnel maintained readiness capability to provide direct monitoring support if testing were to be resumed and ascertained compliance with applicable EPA, DOE, state, and federal regulations and guidelines. Comparison of the measurements and sample analysis results with background levels and with appropriate standards and regulations indicated that there was no airborne radioactivity from diffusion or resuspension detected by the various EPA monitoring networks surrounding the NTS. There was no indication of potential migration of radioactivity to the offsite area through groundwater and no radiation exposure above natural background was received by the offsite population. All evaluated data were consistent with previous data history

  6. The sensitivity of calculated doses to critical assumptions for the offsite consequences of nuclear power reactor accidents

    International Nuclear Information System (INIS)

    Moeller, M.P.; Scherpelz, R.I.; Desrosiers, A.E.

    1982-01-01

    This work analyzes the sensitivity of calculated doses to critical assumptions for offsite consequences following a PWR-2 accident at a nuclear power reactor. The calculations include three radiation dose pathways: internal dose resulting from inhalation, external doses from exposure to the plume, and external doses from exposure to contaminated ground. The critical parameters are the time period of integration for internal dose commitment and the duration of residence on contaminated ground. The data indicate the calculated offsite whole body dose will vary by as much as 600% depending upon the parameters assumed. When offsite radiation doses determine the size of emergency planning zones, this uncertainty has significant effect upon the resources allocated to emergency preparedness

  7. Off-site emergency preparedness activities within the European Commission

    International Nuclear Information System (INIS)

    Kelly, G.N.

    1998-01-01

    Increasing attention is being given by the European Commission to off-site emergency preparedness as part of its broader contribution to improving nuclear safety in Eastern Europe. The main initiatives being taken or planned by the Commission in this area are summarised. Particular attention is given to two topics: Firstly, the development of the RODOS (Real-time On-line DecisiOn Support) system for supporting off-site emergency management in the event of a nuclear accident; and, secondly, the work of an Inter-Service Group on nuclear Off-Site Emergency Preparedness (OSEP) in Eastern Europe that has been established within the Commission. The contribution that each is making to improving emergency preparedness, both in Eastern Europe and in Europe more widely, is described. (orig.)

  8. European commission contribution to improving off-site emergency preparedness

    International Nuclear Information System (INIS)

    Kelly, G.N.

    1996-01-01

    Increasing attention is being given by the European Commission to off-site emergency preparedness as part of its broader contribution to improving nuclear safety in Eastern Europe. The main initiatives being taken or planned by the Commission in this area are summarized. Particular attention is given to two topics: firstly, the development of the RODOS (Real-time On-line Decision Support) system for supporting off-site emergency management in the event of a nuclear accident; and, secondly, the work of an Inter-Service Group on nuclear Off-Site Emergency Preparedness (OSEP) in Eastern Europe that has recently been established within the Commission. The contribution that each is making to improving emergency preparedness, both in Eastern Europe and in Europe more widely, is described

  9. Study on the code system for the off-site consequences assessment of severe nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sora; Mn, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents.

  10. Study on the code system for the off-site consequences assessment of severe nuclear accident

    International Nuclear Information System (INIS)

    Kim, Sora; Mn, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk

    2016-01-01

    The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents

  11. Nuclear fuel handling apparatus

    International Nuclear Information System (INIS)

    Andrea, C.; Dupen, C.F.G.; Noyes, R.C.

    1977-01-01

    A fuel handling machine for a liquid metal cooled nuclear reactor in which a retractable handling tube and gripper are lowered into the reactor to withdraw a spent fuel assembly into the handling tube. The handling tube containing the fuel assembly immersed in liquid sodium is then withdrawn completely from the reactor into the outer barrel of the handling machine. The machine is then used to transport the spent fuel assembly directly to a remotely located decay tank. The fuel handling machine includes a decay heat removal system which continuously removes heat from the interior of the handling tube and which is capable of operating at its full cooling capacity at all times. The handling tube is supported in the machine from an articulated joint which enables it to readily align itself with the correct position in the core. An emergency sodium supply is carried directly by the machine to provide make up in the event of a loss of sodium from the handling tube during transport to the decay tank. 5 claims, 32 drawing figures

  12. Preparation of off-site emergency preparedness plans for nuclear installations

    International Nuclear Information System (INIS)

    1999-10-01

    Safety of public, occupational workers and the protection of environment should be assured while activities for economic and social progress are pursued. These activities include the establishment and utilisation of nuclear facilities and use of radioactive sources. This document is issued as a lead document to facilitate preparation of specific site manuals by the Responsible Organisation for emergency response plans at each site to ensure their preparedness to meet any eventuality due to site emergency in order to mitigate its consequences on the health and safety of site personnel. It takes cognizance of an earlier AERB publication on the subject: Safety Manual on Off-Site Emergency Plan for Nuclear Installations, AERB/SM/NISD-2, 1988 and also takes into consideration the urgent need for promoting public awareness and drawing up revised emergency response plans, which has come out in a significant manner after the accidents at Chernobyl and Bhopal

  13. Off-site nuclear emergency management

    International Nuclear Information System (INIS)

    Miska, H.

    2003-01-01

    Full text: Urgent protective measures for the possibly affected population are the main items to be addressed here, that means actions to be planned and taken in the pre-release and release phase of a nuclear accident. Since we will focus an off-site nuclear emergency management, the utility or licensee only plays a subordinate role, but nevertheless may be the potential cause of all actions. At the other end, there is the possible affected population, the environment, and also economic values. Emergency preparedness and response aims at minimizing adverse effects from the power plant to the values to protect. In the early phase of an accident under consideration here, prompt and sharp actions are necessary to ensure efficacy. On the other hand, the available information on the situation is most limited in this phase such that pre-determined actions based on simple criteria are indispensable. The responsibility for early response actions normally rest with a regional authority which may have some county administrations at subordinate level. The leader of the regional staff has to decide upon protective measures to be implemented at county or municipal level; thus, coherence of the response is ensured at least at a regional level. The decision will be governed at the one side by the existing or predicted radiological situation, on the other side an practical limitations like availability of teams and means. The radiological situation has to be assessed by an advisory team that compiles all information from the utility, the weather conditions, and monitoring results. While the staff leader is experienced through response to major non-nuclear events, the advisors mainly come from the environmental side, having no experience in taking swift decisions in an emergency, but are used to control and prevent. This might be the source of conflicts as observed in several exercises. The radiation protection advisors collect information from the utility, especially about time

  14. Off-site environmental monitoring report: Radiation monitoring around United States Nuclear Test areas, Calendar year 1986

    International Nuclear Information System (INIS)

    Patzer, R.G.; Fontana, C.A.; Grossman, R.F.; Black, S.C.; Dye, R.E.; Smith, D.D.; Thome', D.J.; Mullen, A.A.

    1987-05-01

    The principal activity at the NTS is testing of nuclear devices, though other related projects are also conducted. The principal activities of the Off-Site Radiological Safety Program are routine environmental monitoring for radioactive materials in various media and for radiation in areas which may be affected by nuclear tests; and protective actions in support of the nuclear testing program. These are conducted to document compliance with standards, to identify trends, and to provide information to the public. 28 refs., 37 figs., 30 tabs

  15. Offsite environmental monitoring report. Radiation monitoring around United States nuclear test areas, calendar year 1982

    International Nuclear Information System (INIS)

    Black, S.C.; Grossman, R.F.; Mullen, A.A.; Potter, G.D.; Smith, D.D.

    1983-07-01

    A principal activity of the Offsite Radiological Safety Program is routine environmental monitoring for radioactive materials in various media and for radiation in areas which may be affected by nuclear tests. It is conducted to document compliance with standards, to identify trends, and to provide information to the public. This report summarizes these activities for CY 1982

  16. Remote handling technology for nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Sakai, Akira; Maekawa, Hiromichi; Ohmura, Yutaka

    1997-01-01

    Design and R and D on nuclear fuel cycle facilities has intended development of remote handling and maintenance technology since 1977. IHI has completed the design and construction of several facilities with remote handling systems for Power Reactor and Nuclear Fuel Development Corporation (PNC), Japan Atomic Energy Research Institute (JAERI), and Japan Nuclear Fuel Ltd. (JNFL). Based on the above experiences, IHI is now undertaking integration of specific technology and remote handling technology for application to new fields such as fusion reactor facilities, decommissioning of nuclear reactors, accelerator testing facilities, and robot simulator-aided remote operation systems in the future. (author)

  17. Overview of the spent nuclear fuel storage facilities at the Savannah River Site

    International Nuclear Information System (INIS)

    Conatser, E.R.; Thomas, J.E.

    2000-01-01

    The May 1996 Record of Decision on a Proposed Nuclear Weapons Nonproliferation Policy concerning Foreign Research Reactor Spent Nuclear Fuel initiated a 13 year campaign renewing a policy to support the return of spent nuclear fuel containing uranium of U.S. origin from foreign research reactors to the United States. As of December 1999, over 22% of the approximately 13,000 spent nuclear fuel assemblies from participating countries have been returned to the Savannah River Site (SRS). These ∼2650 assemblies are currently stored in two dedicated SRS wet storage facilities. One is the Receiving Basin for Off-site Fuels (RBOF) and the other as L-Basin. RBOF, built in the early 60's to support the 'Atoms for Peace' program, has been receiving off-site fuel for over 35 years. RBOF has received approximately 1950 casks since startup and has the capability of handling all of the casks currently used in the FRR program. However, RBOF is 90% filled to capacity and is not capable of storing all of the fuel to be received in the program. L-Basin was originally used as temporary storage for materials irradiated in SRS's L-Reactor. New storage racks and other modifications were completed in 1996 that improved water quality and allowed the L-Basin to receive, handle and store spent nuclear fuel assemblies and components from off-site. The first foreign cask was received into the L-Area in April 1997 and approximately 105 foreign and domestic casks have been received since that time. This paper provides an overview of activities related to fuel receipt and storage in both the Receiving Basin for Off-site Fuels (RBOF) and L-Basin facilities. It will illustrate each step of the fuel receipt program from arrival of casks at SRS through cask unloading and decontamination. It will follow the fuel handling process, from fuel unloading, through the cropping and bundling stages, and final placement in the wet storage rack. Decontamination methods and equipment will be explained to show

  18. Overview of the spent nuclear fuel storage facilities at the Savannah River Site

    International Nuclear Information System (INIS)

    Thomas, Jay

    1999-01-01

    The May 1996 Record of Decision on a Proposed Nuclear Weapons Nonproliferation Policy concerning Foreign Research Reactor Spent Nuclear Fuel initiated a 13 year campaign renewing a policy to support the return of spent nuclear fuel containing uranium of U.S.-origin from foreign research reactors to the United States. As of July 1999, over 18% of the approximately 13,000 spent nuclear fuel assemblies from participating countries have been returned to the Savannah River Site (SRS). These 2400 assemblies are currently stored in two dedicated SRS wet storage facilities. One is the Receiving Basin for Off-site Fuels (RBOF) and the other as L-Basin. RBOF, built in the early 60's to support the 'Atoms for Peace' program, has been receiving off-site fuel for over 35 years. RBOF has received approximately 1950 casks since startup and has the capability of handling all of the casks currently used in the FRR program. However, RBOF is 90% filled to capacity and is not capable of storing all of the fuel to be received in the program. L-Basin was originally used as temporary storage for materials irradiated in SRS's L-Reactor. New storage racks and other modifications were completed in 1996 that improved water quality and allowed L-Basin to receive, handle and store spent nuclear fuel assemblies and components from off-site. The first foreign cask was received into L-Area in April 1997 and approximately 86 foreign and domestic casks have been received since that time. This paper provides an overview of activities related to fuel receipt and storage in both the Receiving Basin for Off-site Fuels (RBOF) and L-Basin facilities. It will illustrate each step of the fuel receipt program from arrival of casks at SRS through cask unloading and decontamination. It will follow the fuel handling process, from fuel unloading, through the cropping and bundling stages, and final placement in the wet storage rack. Decontamination methods and equipment will be explained to show how the empty

  19. Overview of the spent nuclear fuel storage facilities at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Conatser, E.R.; Thomas, J.E. [Westinghouse Savannah River Company, Aiken, SC 29808 (United States)

    2000-07-01

    The May 1996 Record of Decision on a Proposed Nuclear Weapons Nonproliferation Policy concerning Foreign Research Reactor Spent Nuclear Fuel initiated a 13 year campaign renewing a policy to support the return of spent nuclear fuel containing uranium of U.S. origin from foreign research reactors to the United States. As of December 1999, over 22% of the approximately 13,000 spent nuclear fuel assemblies from participating countries have been returned to the Savannah River Site (SRS). These {approx}2650 assemblies are currently stored in two dedicated SRS wet storage facilities. One is the Receiving Basin for Off-site Fuels (RBOF) and the other as L-Basin. RBOF, built in the early 60's to support the 'Atoms for Peace' program, has been receiving off-site fuel for over 35 years. RBOF has received approximately 1950 casks since startup and has the capability of handling all of the casks currently used in the FRR program. However, RBOF is 90% filled to capacity and is not capable of storing all of the fuel to be received in the program. L-Basin was originally used as temporary storage for materials irradiated in SRS's L-Reactor. New storage racks and other modifications were completed in 1996 that improved water quality and allowed the L-Basin to receive, handle and store spent nuclear fuel assemblies and components from off-site. The first foreign cask was received into the L-Area in April 1997 and approximately 105 foreign and domestic casks have been received since that time. This paper provides an overview of activities related to fuel receipt and storage in both the Receiving Basin for Off-site Fuels (RBOF) and L-Basin facilities. It will illustrate each step of the fuel receipt program from arrival of casks at SRS through cask unloading and decontamination. It will follow the fuel handling process, from fuel unloading, through the cropping and bundling stages, and final placement in the wet storage rack. Decontamination methods and equipment

  20. Offsite environmental monitoring report: Radiation monitoring around United States nuclear test areas

    International Nuclear Information System (INIS)

    1990-01-01

    This report describes the Offsite Radiation Safety Program conducted during 1990 by the Environmental Protection Agency's (EPA's) Environmental Monitoring Systems Laboratory -- Las Vegas. This laboratory operates an environmental radiation monitoring program in the region surrounding the Nevada Test Site (NTS) and at former test sites in Alaska, Colorado, Mississippi, Nevada, and New Mexico. The surveillance program is designed to measure levels and trends of radioactivity, if present, in the environment surrounding testing areas to ascertain whether current radiation levels and associated doses to the general public are in compliance with existing radiation protection standards. The surveillance program additionally has the responsibility to take action to protect the health and well being of the public in the event of any accidental release of radioactive contaminants. Offsite levels of radiation and radioactivity are assessed by sampling milk, water, and air; by deploying thermoluminescent dosimeters (TLDs) and using pressurized ion chambers (PICs); and by biological monitoring of animals, food crops, and humans. Personnel with mobile monitoring equipment are placed in areas downwind from the test site prior to each nuclear weapons test to implement protective actions, provide immediate radiation monitoring, and obtain environmental samples rapidly after any occurrence of radioactivity release

  1. Monitoring the performance of off-site processors

    International Nuclear Information System (INIS)

    Miller, C.C.

    1995-01-01

    Commercial nuclear power plants have been able to utilize the latest technologies and achieve large volume reduction by obtaining off-site waste processor services. Although the use of such services reduce the burden of waste processing it also reduces the utility's control over the process. Monitoring the performance of off-site processors is important so that the utility is cognizant of the waste disposition for required regulatory reporting. In addition to obtaining data for Reg Guide 1.21 reporting, Performance monitoring is important to determine which vendor and which services to utilize. Off-site processor services were initially offered for the decontamination of metallic waste. Since that time the list of services has expanded to include supercompaction, survey for release, incineration and metal melting. The number of vendors offering off-site services has increased and the services they offer vary. processing rates vary between vendors and have different charge bases. Determining which vendor to use for what service can be complicated and confusing

  2. Developments in the JRodos decision support system for off-site nuclear emergency management and rehabilitation

    Energy Technology Data Exchange (ETDEWEB)

    Landman, Claudia [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany); Pro-Science GmbH, Ettlingen (Germany); Raskob, Wolfgang; Trybushnyi, Dmytro [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany)

    2016-07-01

    JRodos is a non-commercial computer-based decision support system for nuclear accidents. The simulation models for assessing radiological and other consequences and the system features and components allow real-time operation for off-site emergency management as well as the use as a tool for preparing exercises and pre-plannng of countermeasures. There is an active user community that takes influence on further developments.

  3. Hoisting appliances and fuel handling equipment at nuclear facilities

    International Nuclear Information System (INIS)

    1987-01-01

    The guide is followed by the Finnish Centre for Radiation and Nuclear Safety (STUK) in regulating hoisting and handling equipment Class 3 at nuclear facilities. The guide is applied e.g. to the following equipment: reactor building overhead cranes, hoisting appliances at nuclear fuel storages, fuel handling machines, other hoisting appliances, which because of nuclear safety aspects are classified in Safety Class 3, and load-bearing devices connected with the above equipment, such as replaceable hoisting tools and auxiliary lifting devices. The regulating of hoisting and handling equipment comprises the following stages: handling of preliminary and final safety analysis reports, inspection of the construction plan, supervision of fabrication and construction inspection, and supervision of initial start-up and commissioning inspection

  4. Importance ranking of various aspects of offsite radiological emergency preparedness

    International Nuclear Information System (INIS)

    Hockert, J.W.; Carter, T.F.

    1987-01-01

    Under contract to the Edison Electric Institute, IEAL developed a method to assess the relative importance of various aspects of offsite radiological emergency preparedness. The basic approach involved structuring the 35 objectives that the Federal Emergency Management Agency expects offsite emergency planners to demonstrate during nuclear power plant emergency preparedness exercises into a hierarchy based upon the emergency response capabilities they support. The analytical hierarchy process (AHP) was employed to derive the quantitative relative importance of each of the 35 objectives based upon its contribution to the overall capability of offsite agencies to assist in protecting public health and safety in the event of an emergency at a nuclear power plant. The judgments of a cross-section of state and local emergency planners, federal regulators, and intervenors were solicited to rank the 35 objectives

  5. German offsite accident consequence model for nuclear facilities: further development and application

    International Nuclear Information System (INIS)

    Bayer, A.

    1985-01-01

    The German Offsite Accident Consequence Model - first applied in the German Risk Study for nuclear power plants with light water reactors - has been further developed with the improvement of several important submodels in the areas of atmospheric dispersion, shielding effects of houses, and the foodchains. To aid interpretation, the presentation of results has been extended with special emphasis on the presentation of the loss of life expectancy. The accident consequence model has been further developed for application to risk assessments for other nuclear facilities, e.g., the liquid metal fast breeder reactor (SNR-300) and the high temperature gas cooled reactor. Moreover the model have been further developed in the area of optimal countermeasure strategies (sheltering, evacuation, etc.) in the case of the Central European conditions. Preliminary considerations has been performed in connection with safety goals on the basis of doses

  6. Environmental Assessment Offsite Thermal Treatment of Low-Level Mixed Waste

    Energy Technology Data Exchange (ETDEWEB)

    N/A

    1999-05-06

    The U.S. Department of Energy (DOE), Richland Operations Office (RL) needs to demonstrate the economics and feasibility of offsite commercial treatment of contact-handled low-level mixed waste (LLMW), containing polychlorinated biphenyls (PCBS) and other organics, to meet existing regulatory standards for eventual disposal.

  7. A simplified model for calculating early offsite consequences from nuclear reactor accidents

    International Nuclear Information System (INIS)

    Madni, I.K.; Cazzoli, E.G.; Khatib-Rahbar, M.

    1988-07-01

    A personal computer-based model, SMART, has been developed that uses an integral approach for calculating early offsite consequences from nuclear reactor accidents. The solution procedure uses simplified meteorology and involves direct analytic integration of air concentration equations over time and position. This is different from the discretization approach currently used in the CRAC2 and MACCS codes. The SMART code is fast-running, thereby providing a valuable tool for sensitivity and uncertainty studies. The code was benchmarked against both MACCS version 1.4 and CRAC2. Results of benchmarking and detailed sensitivity/uncertainty analyses using SMART are presented. 34 refs., 21 figs., 24 tabs

  8. Preliminary results on food consumption rates for off-site dose calculation of nuclear power plants

    International Nuclear Information System (INIS)

    Lee, Gab Bock; Chung, Yang Geun; Bang, Sun Young; Kang, Duk Won

    2005-01-01

    The Internal dose by food consumption mostly account for radiological dose of public around nuclear power plants(NPP). But, food consumption rate applied to off-site dose calculation in Korea which is the result of field investigation around Kori NPP by the KAERI in 1988. is not reflected of the latest dietary characteristics. The Ministry of Health and Welfare Affairs has investigated the food and nutrition of nations every 3 years based on the Law of National Health Improvement. To update the food consumption rates of the maximum individual, the analysis of the national food investigation results and field surveys around nuclear power plant sites have been carried out

  9. Hoisting appliances and fuel handling equipment at nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-12-31

    The guide is followed by the Finnish Centre for Radiation and Nuclear Safety (STUK) in regulating hoisting and handling equipment Class 3 at nuclear facilities. The guide is applied e.g. to the following equipment: reactor building overhead cranes, hoisting appliances at nuclear fuel storages, fuel handling machines, other hoisting appliances, which because of nuclear safety aspects are classified in Safety Class 3, and load-bearing devices connected with the above equipment, such as replaceable hoisting tools and auxiliary lifting devices. The regulating of hoisting and handling equipment comprises the following stages: handling of preliminary and final safety analysis reports, inspection of the construction plan, supervision of fabrication and construction inspection, and supervision of initial start-up and commissioning inspection. 36 refs. Translation. The original text is published under the same guide number. The guide is valid from 5 January 1987 and will be in force until further notice.

  10. Robotics and remote handling in the nuclear industry

    Energy Technology Data Exchange (ETDEWEB)

    1984-01-01

    This book presents the papers given at a conference on the use of remote handling equipment in nuclear facilities. Topics considered at the conference included dose reduction, artificial intelligence in nuclear plant maintenance, robotic welding, uncertainty covariances, reactor operation and inspection, reactor maintenance and repair, uranium mining, fuel fabrication, reactor component manufacture, irradiated fuel and radioactive waste management, and radioisotope handling.

  11. Handling and transfer operations for partially-spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ibrahim, J K [PUSPATI, Kuala Lumpur (Malaysia)

    1983-12-01

    This project involved the handling and transfer of partially-spent reactor fuel from the Oregon State University TRIGA Reactor in Corvallis, Oregon to Hanford Engineering Development Laboratory in Richland, Washington. The method of handling is dependent upon the burn-up history of the fuel elements. Legal constraints imposed by standing U.S. nuclear regulations determine the selection of transport containers, transportation procedures, physical security arrangements in transit and nuclear material accountability documentation. Results of in-house safety evaluations of the project determine the extent of involvement of pertinent nuclear regulatory authorities. The actual handling activities and actual radiation dose rates are also presented.

  12. Severe accident modeling and offsite dose consequence evaluations for nuclear power plant emergency planning

    Energy Technology Data Exchange (ETDEWEB)

    Chen, S.H.; Feng, T.S.; Huang, K.C. [National Tsing-Hua Univ., Hsinchu, Taiwan (China); Wang, J.R. [Inst. of Nuclear Energy Research, Longtan, Taiwan (China); Cheng, Y.H. [Industrial Tech. Res. Inst., Hsinchu, Taiwan (China); Shih, C., E-mail: ckshih@ess.nthu.edu.tw [National Tsing-Hua Univ., Hsinchu, Taiwan (China)

    2011-07-01

    We have investigated the roles of Firewater Addition System and Passive Flooder in ABWR severe accidents, such as LOCA and SBO. The results are apparent that Firewater System is vital in the highly unlikely situation where all AC are lost. Also in this paper, we present EPZDose, an effective and faster-than-real time code for offsite dose consequences predictions and evaluations. Illustrations with the release from our severe accident scenario show friendly and informative user's interface for supporting decision makings in nuclear emergency situations. (author)

  13. Off-site nuclear emergency management in Germany under the auspices of the federal structure

    International Nuclear Information System (INIS)

    Bayer, A.; Bittner, S.; Korn, H.

    1998-01-01

    Both the individual states (Laender) and the federation (Bund) are involved in off-site emergency management in Germany. The states operate site-related Remote Monitoring Systems for Nuclear Power Plants, while the federation operates a nationwide Integrated Measurement and Information System. The states are responsible for accident response, the federation is responsible for radiation precaution measures. In the event of an accident, the state and federal authorities make their decisions and implement the corresponding emergency measures within their responsibility. Exchange of information exists between the two levels. (P.A.)

  14. Handling and transfer operations for partially-spent nuclear fuel

    International Nuclear Information System (INIS)

    Ibrahim, J.K.

    1983-01-01

    This project involved the handling and transfer of partially-spent reactor fuel from the Oregon State University TRIGA Reactor in Corvallis, Oregon to Hanford Engineering Development Laboratory in Richland, Washington. The method of handling is dependent upon the burn-up history of the fuel elements. Legal constraints imposed by standing U.S. nuclear regulations determine the selection of transport containers, transportation procedures, physical security arrangements in transit and nuclear material accountability documentation. Results of in-house safety evaluations of the project determine the extent of involvement of pertinent nuclear regulatory authorities. The actual handling activities and actual radiation dose rates are also presented (author)

  15. Emergency management in nuclear power plants: a regulatory view

    International Nuclear Information System (INIS)

    Shukla, Vikas; Chander, Vipin; Vijayan, P.; Nair, P.S.; Krishnamurthy, P.R.

    2011-01-01

    The nuclear power plants in India adopts a high level of defence in depth concept in design and operates at highest degree of safety, however the possibility of nuclear accidents cannot be ruled out. The safety and regulatory review of Nuclear Power Plants (NPPs) in India are carried out by Atomic Energy Regulatory Board (AERB). Section 33 of Atomic Energy (Radiation Protection) Rules-2004 provides the basic requirements of emergency preparedness aspects for a nuclear facility. Prior to the issuance of a license for the operation of NPPs, AERB ensures that the site specific emergency response manuals are in place and tested. The emergency response plan includes the emergency response organization, their responsibilities, the detailed scheme of emergency preparedness, response, facilities, equipments, coordination and support of various organizations and other technical aspects. These emergency preparedness plans are tested at periodic interval to check the overall effectiveness. The plant and site emergency exercise is handled by the plant authorities as per the site emergency plan. The events with off-site consequences are handled by the district authorities according to the off-site emergency plan. In off-site emergency exercises, observers from AERB and other associated organizations participate. Observations of the participants are discussed in the feedback session of the exercise for their disposition. This paper reviews the current level of emergency planning and preparedness, statistics of emergency exercises conducted and their salient findings. The paper highlights improvement in the emergency management programme over the years including development of advance technical support systems. The major challenges in off-site emergency management programme such as industrial growth and increase in population within the sterilized zone, frequent transfer of district officials and the floating population around the NPPs are outlined. The areas for improvement in

  16. Off-site preparedness and nuclear-power-plant licensing

    International Nuclear Information System (INIS)

    Perry, S.W.

    1983-01-01

    The first year and a half in which off-site emergency preparedness issues have been litigated before the Atomic Safety and Licensing Boards of the NRC have surfaced unique problems of proof for the applicant as well as the staff. These problems seem to be abating as the boards and the parties become more comfortable with the field and its issues, and as FEMA-NRC emergency management expertise gains credibility. Emergency preparedness presentations have also improved as the parties have become more sensitive to the seasonality of the preparedness case, and have increasingly attempted to raise it at a time when a fully developed set of facts is available for the record. Off-site preparedness issues are only now beginning to be raised on appeal to the NRC appeals board, the full commission, and the courts. Helpful guidance on what constitutes an adequate record in this area will undoubtedly be forthcoming in decisions handed down by these bodies in the months ahead

  17. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt

    International Nuclear Information System (INIS)

    Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

    1978-06-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects

  18. Control panel handling of a nuclear simulator

    International Nuclear Information System (INIS)

    Martin Polo, F.; Jimenez Fraustro, L.A.; Banuelos Galindo, A.; Diamant Rubinstein, A.

    1985-01-01

    The handling of the control panels for a Nuclear Simulator for operating training is described. The control panels are handled by a set of intelligent controllers, each with at least two processors (8035 - Communications Controller and a 8085 - Master processor). The Controllers are connected to the main computers (Two dual processor Gould concept 32/6780 and a single processor Gould concept 32/6705) via serial asynchronous channels in a multidrop, star-like architecture. The controllers transmit to the main computers only the changes detected and receive the whole set of output variables as computed by the mathematical models of the Nuclear Plant

  19. Fuel handling and storage systems in nuclear power plants

    International Nuclear Information System (INIS)

    1984-01-01

    The scope of this Guide includes the design of handling and storage facilities for fuel assemblies from the receipt of fuel into the nuclear power plant until the fuel departs from that plant. The unirradiated fuel considered in this Guide is assumed not to exhibit any significant level of radiation so that it can be handled without shielding or cooling. This Guide also gives limited consideration to the handling and storage of certain core components. While the general design and safety principles are discussed in Section 2 of this Guide, more specific design requirements for the handling and storage of fuel are given in detailed sections which follow the general design and safety principles. Further useful information is to be found in the IAEA Technical Reports Series No. 189 ''Storage, Handling and Movement of Fuel and Related Components at Nuclear Power Plants'' and No. 198 ''Guide to the Safe Handling of Radioactive Wastes at Nuclear Power Plants''. However, the scope of the Guide does not include consideration of the following: (1) The various reactor physics questions associated with fuel and absorber loading and unloading into the core; (2) The design aspects of preparation of the reactor for fuel loading (such as the removal of the pressure vessel head for a light water reactor) and restoration after loading; (3) The design of shipping casks; (4) Fuel storage of a long-term nature exceeding the design lifetime of the nuclear power plant; (5) Unirradiated fuel containing plutonium

  20. Spent fuel cask handling at an operating nuclear power plant

    International Nuclear Information System (INIS)

    Pal, A.C.

    1988-01-01

    The importance of spent fuel handling at operating nuclear power plants cannot be overstated. Because of its highly radioactive nature, however, spent fuel must be handled in thick, lead-lined containers or casks. Thus, all casks for spent fuel handling are heavy loads by the US Nuclear Regulatory Commission's definition, and any load-drop must be evaluated for its potential to damage safety-related equipment. Nuclear Regulatory Guide NUREG-0612 prescribes the regulatory requirements of alternative heavy-load-handling methodologies such as (a) by providing cranes that meet the requirements of NUREG-0554, which shall be called the soft path, or (b) by providing protective devices at all postulated load-drop areas to prevent any damage to safety-related equipment, which shall be called the hard path. The work reported in this paper relates to cask handling at New York Power Authority's James A. FitzPatrick (JAF) plant

  1. Cost-effectiveness of reduction of off-site dose

    International Nuclear Information System (INIS)

    McGrath, J.J.; Macphee, R.; Arbeau, N.; Miskin, J.; Scott, C.K.; Winters, E.

    1988-03-01

    Since the early 1970's, nuclear power plants have been designed and operated with a target of not releasing more than one percent of the licensed limits (derived emission limits) in liquid and gaseous effluents. The AECB initiated this study of the cost-effectiveness of the reduction of off-site doses as part of a review to determine if further measures to reduce off-site doses might be reasonably achievable. Atlantic Nuclear has estimated the cost of existing technology options that can be applied for a further reduction of radioactive effluents from future CANDU nuclear power plants. Detritiation, filtration, ion exchange and evaporation are included in the assessment. The costs are presented in 1987 Canadian dollars, and include capital and operating costs for a reference 50 year plant life. Darlington NGS and Point Lepreau NGS are the reference nuclear power plant types and locations. The effect resulting from the hypothetical application of each technology has been calculated as the resulting reduction in world collective radiation dose detriment. The CSA N288.1 procedure was used for local pathway analysis and the global dispersion model developed by the NEA (OECD) group of experts was used for dose calculations. The reduction in the 'collective effective dose equivalent commitment' was assumed to exist for 10,000 years, the expected life-span of solid waste repositories. No attempt was made to model world population dynamics. The collective dose reductions were calculated for a nominal world population of 10 billion persons. The estimated cost and effect of applying the technology options are summarized in a tabular form for input to further consideration of 'reasonably achievable off-site dose levels'

  2. Dungeness Power Station off-site emergency plan

    International Nuclear Information System (INIS)

    1993-01-01

    This off-site Emergency Plan in the event of an accidental release of radioactivity at the Dungeness Nuclear power station sets out the necessary management and coordination processes between Nuclear Electric, operators of the site, the emergency services and relevant local authorities. The objectives promoting the aim are identified and the activities which will be undertaken to protect the public and the environment in the event of an emergency are outlined. (UK)

  3. Health effects models for off-site radiological consequence analysis on nuclear reactor accidents (II)

    Energy Technology Data Exchange (ETDEWEB)

    Homma, Toshimitsu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Takahashi, Tomoyuki [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst; Yonehara, Hidenori [National Inst. of Radiological Sciences, Chiba (Japan)] [eds.

    2000-12-01

    This report is a revision of JAERI-M 91-005, 'Health Effects Models for Off-Site Radiological Consequence Analysis of Nuclear Reactor Accidents'. This revision provides a review of two revisions of NUREG/CR-4214 reports by the U.S. Nuclear Regulatory Commission which is the basis of the JAERI health effects models and other several recent reports that may impact the health effects models by international organizations. The major changes to the first version of the JAERI health effects models and the recommended parameters in this report are for late somatic effects. These changes reflect recent changes in cancer risk factors that have come from longer followup and revised dosimetry in major studies on the Japanese A-bomb survivors. This report also provides suggestions about future revisions of computational aspects on health effects models. (author)

  4. Health effects models for off-site radiological consequence analysis on nuclear reactor accidents (II)

    International Nuclear Information System (INIS)

    Homma, Toshimitsu

    2000-12-01

    This report is a revision of JAERI-M 91-005, 'Health Effects Models for Off-Site Radiological Consequence Analysis of Nuclear Reactor Accidents'. This revision provides a review of two revisions of NUREG/CR-4214 reports by the U.S. Nuclear Regulatory Commission which is the basis of the JAERI health effects models and other several recent reports that may impact the health effects models by international organizations. The major changes to the first version of the JAERI health effects models and the recommended parameters in this report are for late somatic effects. These changes reflect recent changes in cancer risk factors that have come from longer followup and revised dosimetry in major studies on the Japanese A-bomb survivors. This report also provides suggestions about future revisions of computational aspects on health effects models. (author)

  5. A report on anti-earthquake countermeasures of the switchyard etc. with a view to securing reliability of offsite power supply for Nuclear Power Stations etc. (Attachment)

    International Nuclear Information System (INIS)

    2011-01-01

    Submission of a report on anti-earthquake countermeasures of the switchyard etc. with a view to securing reliability of offsite power supply for Nuclear Power Stations etc. to Nuclear and Industrial Safety Agency of Ministry of Economy, Trade and Industry

  6. Improvement of Off-site Dose Assessment Code for Operating Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Juyub; Kim, Juyoul; Shin, Kwangyoung [FNC Technology Co. Ltd., Yongin (Korea, Republic of); You, Songjae; Moon, Jongyi [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-05-15

    XOQDOQ code which calculates atmospheric Dispersion factor was included into INDAC also. A research on the improvement of off-site dose assessment system for an operating nuclear power plant was performed by KINS in 2011. As a result, following improvements were derived: - Separation of dose assessment for new and existing facilities - Update of food ingestion data - Consideration of multi-unit operation and so on In order to reflect the results, INDAC is under modification. INDAC is an integrated dose assessment code for an operating nuclear power plant and consists of three main modules: XOQDOQ, GASDOS and LIQDOS. The modules are under modification in order to improve the accuracy of assessment and usability. Assessment points for multi-unit release can be calculated through the improved code and the method on dose assessment for multi-unit release has been modified, so that the dose assessment result of multi-unit site becomes more realistic by relieving excessive conservatism. Finally, as the accuracy of calculation modules has been improved, the reliability of dose assessment result has been strengthened.

  7. Offsite environmental monitoring report. Radiation monitoring around United States nuclear test areas, calendar year 1980

    International Nuclear Information System (INIS)

    Smith, D.D.; Grossman, R.F.; Corkern, W.D.; Thome, D.J.; Patzer, R.G.; Hopper, J.L.

    1981-06-01

    The US Environmental Protection Agency's (EPA) Environmental Monitoring Systems Laboratory in Las Vegas (EMSL-LV) continued its Offsite Radiological Safety Program for the Nevada Test Site (NTS) and other sites of past underground nuclear tests. For each test, the Laboratory provided airborne meteorological measurements, ground and airborne radiation monitoring teams, and special briefings to the Test Controller's Advisory Panel. Test-related radioactivity from the NTS was detected offsite following the Riola Test conducted on September 25, 1980. This consisted of xenon-133 (3.4 x 10 -11 μCi/m1) and xenon-135 (3.6 x 10 -10 μCi/m1) in a compressed air sample collected at Lathrop Wells, Nevada. The estimated dose equivalent to the whole body of a hypothetical receptor at Lathrop Wells from exposure to the radioxenon was 0.011 mrem, which is 0.006 percent of the radiation protection guide for a suitable sample of the general population. Whole-body counts of individuals residing in the environs of the NTS showed no manmade radionuclides attributable to the testing program. The only radioactivity from non-NTS sites of past underground nuclear tests was due to tritium in water samples collected from the Project Dribble Site near Hattiesburg, Mississippi, and the Project Long Shot Site on Amchitka Island, Alaska. The maximum concentrations measured at these locations were 1 and 0.1 percent of the Concentration Guide for drinking water, respectively. A small amount of airborne radioactivity originating from nuclear tests carried out by the People's Republic of China was detected during 1980 at some stations scattered throughout the Air Surveillance Network. The Laboratory's Animal Investigation Program sampled tissues from wildlife and domestic animals on and around the NTS. Data from analysis of these tissues are published separately in an annual report

  8. Handling effluent from nuclear thermal propulsion system ground tests

    International Nuclear Information System (INIS)

    Shipers, L.R.; Allen, G.C.

    1992-01-01

    A variety of approaches for handling effluent from nuclear thermal propulsion system ground tests in an environmentally acceptable manner are discussed. The functional requirements of effluent treatment are defined and concept options are presented within the framework of these requirements. System concepts differ primarily in the choice of fission-product retention and waste handling concepts. The concept options considered range from closed cycle (venting the exhaust to a closed volume or recirculating the hydrogen in a closed loop) to open cycle (real time processing and venting of the effluent). This paper reviews the different methods to handle effluent from nuclear thermal propulsion system ground tests

  9. The training for nuclear fuel handling at EDF

    International Nuclear Information System (INIS)

    Marion, J.P.

    1999-01-01

    The handling of fuel assemblies in a nuclear power plant presents 3 types of work: the taking delivery of fresh fuel, the refueling and the disposal of spent fuel. These operations are realized by teams made up of 3 handling operators and a supervisor. The refueling is made by 3*8-hour teams. These handling operations are important for the nuclear safety, a mishandling can damage the fuel cladding which is the first containment barrier, so a training center (CETIC) has been created. This center was founded in 1986 by EDF and Framatome, the purpose was to validate maintenance procedures, to test handling equipment and to train the teams which work on site. Various training programmes have been set up and a system of qualification degrees has been organized. The CETIC is fitted up with equipment that are full-sized mockups of real installations. Fuel assemblies don't react in a similar way to the different mechanical and neutronic stresses they undergo while they are in the core, they get deformed and the handling operations become more delicate. The mockup fuel assemblies are quite deformed to train the teams and prepare them to face any real situation. (A.C.)

  10. Offsite environmental monitoring report: Radiation monitoring around United States nuclear test areas, calendar year 1991

    International Nuclear Information System (INIS)

    Chaloud, D.J.; Dicey, B.B.; Mullen, A.A.; Neale, A.C.; Sparks, A.R.; Fontana, C.A.; Carroll, L.D.; Phillips, W.G.; Smith, D.D.; Thome, D.J.

    1992-01-01

    This report describes the Offsite Radiation Safety Program conducted during 1991 by the Environmental Protection Agency's (EPA's) Environmental Monitoring Systems Laboratory-Las Vegas. This laboratory operates an environmental radiation monitoring program in the region surrounding the Nevada Test Site (NTS) and at former test sites in Alaska, Colorado, Mississippi, Nevada, and New Mexico. The surveillance program is designed to measure levels and trends of radioactivity, if present, in the environment surrounding testing areas to ascertain whether current radiation levels and associated doses to the general public are in compliance with existing radiation protection standards. The surveillance program additionally has the responsibility to take action to protect the health and well being of the public in the event of any accidental release of radioactive contaminants. Offsite levels of radiation and radioactivity are assessed by sampling milk, water, and air; by deploying thermoluminescent dosimeters (TLDs) and using pressurized ion chambers (PICs); and by biological monitoring of animals, food crops, and humans. Personnel with mobile monitoring equipment are placed in areas downwind from the test site prior to each nuclear weapons test to implement protective actions, provide immediate radiation monitoring, and obtain environmental samples rapidly after any occurrence of radioactivity release. Comparison of the measurements and sample analysis results with background levels and with appropriate standards and regulations indicated that there was no radioactivity detected offsite by the various EPA monitoring networks and no exposure above natural background to the population living in the vicinity of the NTS that could be attributed to current NTS activities. Annual and long-term trends were evaluated in the Noble Gas, Tritium, Milk Surveillance, Biomonitoring, TLD, PIC networks, and the Long-Term Hydrological Monitoring Program

  11. Off-site emergency planning

    International Nuclear Information System (INIS)

    Narrog, J.

    1980-01-01

    In the event of a nuclear accident, the actions taken to protect the public from off-site consequences must be effective. An effective organization of emergency actions is based on two components: the actions of the operator of the nuclear facility and the actions of the competent authorities. The measures of the operator are of special importance in the first hours after the beginning of the nuclear accident, because there is no other help. Therefore the operator of a nuclear facility shall be obliged under the nuclear licensing procedure to make provisions of his own and carry out protective measures which should be compiled in a so-called 'alarm-plan'. On the other hand the means of the operator are too small in many cases and there is a need for actions by the responsible authorities. The actions of the authorities should be compiled in a so-called 'emergency response plan'. The emergency response plan shall apply to all cases in which, as a result of occurrences in or at a nuclear facility, a damaging impact on the environment is expected or has occurred requiring the authorities in charge to intervene for its prevention or limitation. (orig./RW)

  12. Fuel handling grapple for nuclear reactor plants

    International Nuclear Information System (INIS)

    Rousar, D.L.

    1992-01-01

    This patent describes a fuel handling system for nuclear reactor plants. It comprises: a reactor vessel having an openable top and removable cover and containing therein, submerged in water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units, the fuel handling system consisting essentially of the combination of: a fuel bundle handling platform movable over the open top of the reactor vessel; a fuel bundle handling mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grapple means comprising complementary hooks which pivot inward toward each other to securely grasp a bail handle of a nuclear reactor fuel bundle and pivot backward away from each other to release a bail handle; the grapple means having a hollow cylindrical support shaft fixed within the grapple head with hollow cylindrical sleeves rotatably mounted and fixed in longitudinal axial position on the support shaft and each sleeve having complementary hooks secured thereto whereby each hook pivots with the rotation of the sleeve secured thereto; and the hollow cylindrical support shaft being provided with complementary orifices on opposite sides of its hollow cylindrical and intermediate to the sleeves mounted thereon whereby the orifices on both sides of the hollow cylindrical support shaft are vertically aligned providing a direct in-line optical viewing path downward there-through and a remote operator positioned above the grapple means can observe from overhead the area immediately below the grapple hooks

  13. Offsite environmental monitoring report: Radiation monitoring around United States nuclear test areas, calendar year 1991

    Energy Technology Data Exchange (ETDEWEB)

    Chaloud, D.J.; Dicey, B.B.; Mullen, A.A.; Neale, A.C.; Sparks, A.R.; Fontana, C.A.; Carroll, L.D.; Phillips, W.G.; Smith, D.D.; Thome, D.J.

    1992-01-01

    This report describes the Offsite Radiation Safety Program conducted during 1991 by the Environmental Protection Agency`s (EPA`s) Environmental Monitoring Systems Laboratory-Las Vegas. This laboratory operates an environmental radiation monitoring program in the region surrounding the Nevada Test Site (NTS) and at former test sites in Alaska, Colorado, Mississippi, Nevada, and New Mexico. The surveillance program is designed to measure levels and trends of radioactivity, if present, in the environment surrounding testing areas to ascertain whether current radiation levels and associated doses to the general public are in compliance with existing radiation protection standards. The surveillance program additionally has the responsibility to take action to protect the health and well being of the public in the event of any accidental release of radioactive contaminants. Offsite levels of radiation and radioactivity are assessed by sampling milk, water, and air; by deploying thermoluminescent dosimeters (TLDs) and using pressurized ion chambers (PICs); and by biological monitoring of animals, food crops, and humans. Personnel with mobile monitoring equipment are placed in areas downwind from the test site prior to each nuclear weapons test to implement protective actions, provide immediate radiation monitoring, and obtain environmental samples rapidly after any occurrence of radioactivity release. Comparison of the measurements and sample analysis results with background levels and with appropriate standards and regulations indicated that there was no radioactivity detected offsite by the various EPA monitoring networks and no exposure above natural background to the population living in the vicinity of the NTS that could be attributed to current NTS activities. Annual and long-term trends were evaluated in the Noble Gas, Tritium, Milk Surveillance, Biomonitoring, TLD, PIC networks, and the Long-Term Hydrological Monitoring Program.

  14. Offsite environmental monitoring report. Radiation monitoring around United States nuclear test areas, calendar year 1983

    International Nuclear Information System (INIS)

    Patzer, R.G.; Black, S.C.; Grossman, R.F.; Smith, D.D.

    1984-07-01

    This report covers the routine radiation monitoring activities conducted by the Environmental Monitoring Systems Laboratory-Las Vegas in areas which may be affected by nuclear testing programs of the Department of Energy. This monitoring is conducted to document compliance with standards, to identify trends in environmental radiation, and to provide such information to the public. It summarizes these activities for calendar year 1983. No radioactivity attributable to NTS activities was detectable offsite by the monitoring networks. Using recorded wind data and Pasquill stability categories, atmospheric dispersion calculations based on reported radionuclide releases yield an estimated dose of 5 x 10 -5 man-rem to the population within 80 km of the Nevada Test Site during 1983. World-wide fallout of Kr-85, Sr-90, Cs-137, and Pu-239 detected by the monitoring networks would cause maximum exposure to an individual of less than 0.2 mrem per year. Plutonium and krypton in air were similar to 1982 levels while cesium and strontium in other samples were near the detection limits. An occasional net exposure to offsite residents has been detected by the TLD network. On investigation, the cause of such net exposures has been due to personal habits or occupational activities, not to NTS activities. 29 references, 35 figures, 30 tables

  15. CEC workshop on methods for assessing the offsite radiological consequences of nuclear accidents

    International Nuclear Information System (INIS)

    Luykx, F.; Sinnaeve, J.

    1986-01-01

    On Apr 15-19, 1985, in Luxembourg, the Commission of the European Communities (CEC), in collaboration with the Kernforschungszentrum Karlsruhe (KfK), Federal Republic of Germany, and the National Radiological Protection Board (NRPB), United Kingdom, presented a workshop on methods for assessing the offsite radiological consequences of nuclear accidents. The program consisted of eight sessions. The main conclusions, which were presented in the Round Table Session by the individual Session Chairmen, are summarized. Session topics are as follows: Session I: international developments in the field of accident consequence assessment (ACA); Session II: atmospheric dispersion; Session III: food chain models; Session IV: urban contamination; Session V: demographic and land use data; Session VI: dosimetry, health effects, economic and counter measure models; Session VII: uncertainty analysis; and Session VIII: application of probabilistic consequence models as decision aids

  16. Modeling time to recovery and initiating event frequency for loss of off-site power incidents at nuclear power plants

    International Nuclear Information System (INIS)

    Iman, R.L.; Hora, S.C.

    1988-01-01

    Industry data representing the time to recovery of loss of off-site power at nuclear power plants for 63 incidents caused by plant-centered losses, grid losses, or severe weather losses are fit with exponential, lognormal, gamma and Weibull probability models. A Bayesian analysis is used to compare the adequacy of each of these models and to provide uncertainty bounds on each of the fitted models. A composite model that combines the probability models fitted to each of the three sources of data is presented as a method for predicting the time to recovery of loss of off-site power. The composite model is very general and can be made site specific by making adjustments on the models used, such as might occur due to the type of switchyard configuration or type of grid, and by adjusting the weights on the individual models, such as might occur with weather conditions existing at a particular plant. Adjustments in the composite model are shown for different models used for switchyard configuration and for different weights due to weather. Bayesian approaches are also presented for modeling the frequency of initiating events leading to loss of off-site power. One Bayesian model assumes that all plants share a common incidence rate for loss of off-site power, while the other Bayesian approach models the incidence rate for each plant relative to the incidence rates of all other plants. Combining the Bayesian models for the frequency of the initiating events with the composite Bayesian model for recovery provides the necessary vehicle for a complete model that incorporates uncertainty into a probabilistic risk assessment

  17. Rethinking the Business Model in Construction by the Use of Off-Site System Deliverance: Case of the Shaft Project

    OpenAIRE

    Thuesen, Christian; Hvam, Lars

    2013-01-01

    This paper presents a set of insights to be used in the development of business models for off-site system deliveries contributing to the development of Off-Site Manufacturing practices (OSM). The theoretical offset for discussing the development of business models is the blue ocean strategy literature combined with theories on mass-customization and platform development identifying the optimization of cost and value through the handling of complexity as the central process. This framework is...

  18. State fund of decommissioning of nuclear installations and handling of spent nuclear fuels and nuclear wastes (Slovak Republic)

    International Nuclear Information System (INIS)

    Kozma, Milos

    2006-01-01

    State Fund for Decommissioning of Nuclear Installations and Handling of Spent Nuclear Fuels and Nuclear Wastes was established by the Act 254/1994 of the National Council of the Slovak Republic as a special-purpose fund which concentrates financial resources intended for decommissioning of nuclear installations and for handling of spent nuclear fuels and radioactive wastes. The Act was amended in 2000, 2001 and 2002. The Fund is legal entity and independent from operator of nuclear installations Slovak Power Facilities Inc. The Fund is headed by Director, who is appointed and recalled by Minister of Economy of the Slovak Republic. Sources of the Fund are generated from: a) contributions by nuclear installation operators; b) penalties imposed by Nuclear Regulatory Authority of the Slovak Republic upon natural persons and legal entities pursuant to separate regulation; c) bank credits; d) interest on Fund deposits in banks; e) grants from State Budget; f) other sources as provided by special regulation. Fund resources may be used for the following purposes: a) decommissioning of nuclear installations; b) handling of spent nuclear fuels and radioactive wastes after the termination of nuclear installation operation; c) handling of radioactive wastes whose originator is not known, including occasionally seized radioactive wastes and radioactive materials stemming from criminal activities whose originator is not known, as confirmed by Police Corps investigator or Ministry of Health of the Slovak Republic; d) purchase of land for the establishment of nuclear fuel and nuclear waste repositories; e) research and development in the areas of decommissioning of nuclear installations and handling of nuclear fuels and radioactive wastes after the termination of the operation of nuclear installations; f) selection of localities, geological survey, preparation, design, construction, commissioning, operation and closure of repositories of spent nuclear fuels and radioactive wastes

  19. Off-site protective action selection for nuclear reactor accidents

    International Nuclear Information System (INIS)

    Weerakkody, S.D.

    1986-01-01

    A computer program based upon a model using a rational theoretical basis was developed to select appropriate off-site protective actions during nuclear reactor accidents. The special features of this program include (a) introduction of a precursor concept that uses the history of the accident progression to determine the spectrum of potential accident scenarios and estimates of the likelihoods of each accident scenario; (b) use of statistical decision theory and the concept of entropy of a spectrum to select the appropriate protective actions using either the minimax principle or the Bayes action method; and (c) introduction of methods to quantify evacuation travel risks. In order to illustrate the usefulness of the computer program, it was applied at three stages of the Three Mile Island accident scenario. Quantified non-radiological risks of evaluation have been used to establish dose thresholds below which evacuations are not justified. Using the Poisson analysis for evacuation risks and the absolute L-L BEIR model for radiation risk suggests 330 mrems as a reasonable value for this threshold. The usefulness of the program in developing a technical basis to select the size of the plume exposure pathway emergency planning zone (EPZ) is discussed. Quantified evacuation risks, cost, and the current rationale upon which the EPZ is based, justify an EPZ between 5-10 miles for WASH-1400 source-terms

  20. Offsite emergency radiological monitoring system and technology

    International Nuclear Information System (INIS)

    Mao Yongze

    1994-01-01

    The study and advance of the offsite radiological monitoring system and technology which is an important branch in the field of nuclear monitoring technology are described. The author suggests that the predicting and measuring system should be involved in the monitoring system. The measuring system can further be divided into four sub-systems, namely plume exposure pathway, emergency worker, ingestion exposure pathway and post accident recovery measuring sub-systems. The main facilities for the monitoring system are concluded as one station, one helicopter, one laboratory and two vehicles. The instrumentation for complement of the facilities and their good performance characteristics, up-to-date technology are also introduced in brief. The offsite emergency radiation monitoring system and technology are compared in detail with those recommended by FEMA U.S.A.. Finally the paper discusses some trends in development of emergency radiation monitoring system and technology in the developed countries

  1. Potential application of nuclear remote-handling technology to underwater inspection and maintenance

    International Nuclear Information System (INIS)

    Eccleston, M.J.

    1990-01-01

    Examples are given of remote handling equipment developed within the nuclear industry and employing telemanipulative or telerobotic principles. In telerobotics the nuclear industry has been following a trend towards increased levels of autonomy, delegating operator control to a computer, for example, in resolved rate manipulator tip control, teach-and-repeat control and collision avoidance. Illustrations are presented of remote-handling techniques from the nuclear industry which may be carried over into undersea remote inspection, maintenance and repair systems. (author)

  2. Cable handling system for use in a nuclear reactor

    International Nuclear Information System (INIS)

    Crosgrove, R.O.; Larson, E.M.; Moody, E.

    1982-01-01

    A cable handling system for use in an installation such as a nuclear reactor is disclosed herein along with relevant portions of the reactor which, in a preferred embodiment, is a liquid metal fast breeder reactor. The cable handling system provides a specific way of interconnecting certain internal reactor components with certain external components, through an assembly of rotatable plugs. Moreover, this is done without having to disconnect these components from one another during rotation of the plugs and yet without interfering with other reactor components in the vicinity of the rotating plugs and cable handling system

  3. Off-site environmental monitoring report: radiation monitoring around United States nuclear test areas, calendar year 1984

    International Nuclear Information System (INIS)

    Potter, G.D.; Black, S.C.; Grossman, R.F.; Patzer, R.G.; Smith, D.D.

    1985-04-01

    This report covers the routine radiation monitoring activities conducted by the Environmental Monitoring Systems Laboratory-Las Vegas in areas which may be affected by nuclear testing programs of the Department of Energy. This monitoring is conducted to document compliance with standards, to identify trends in environmental radiation, and to provide such information to the public. It summarizes these activities for calendar year 1984. No radioactivity attributable to NTS activities was detectable offsite by the monitoring networks. Using recorded wind data and Pasquill stability categories, atmospheric dispersion calculations based on reported radionuclides releases yield an estimated dose of 1 x 10 -3 person-rem to the population within 80 km of the Nevada Test Site during 1983. World-wide fallout of Kr-85, Sr-90, Cs-137, and Pu-239 detected by the monitoring networks would cause maximum exposure to an individual of less than 0.6 mrem per year. Plutonium in air was still detectable along with krypton-85, which continued its gradual increase, as has been reported previously. Cesium and strontium in air were near their detection limits. An occasional net exposure to offsite residents has been detected by the TLD network. On investigation, the cause of such net exposures has been due to personal habits or occupational activities, not to NTS activities. 32 refs., 36 figs., 27 tabs

  4. Material handling for the Los Alamos National Laboratory Nuclear Storage Facility

    International Nuclear Information System (INIS)

    Pittman, P.; Roybal, J.; Durrer, R.; Gordon, D.

    1999-01-01

    This paper will present the design and application of material handling and automation systems currently being developed for the Los Alamos National Laboratory (LANL) Nuclear Material Storage Facility (NMSF) renovation project. The NMSF is a long-term storage facility for nuclear material in various forms. The material is stored within tubes in a rack called a basket. The material handling equipment range from simple lift assist devices to more sophisticated fully automated robots, and are split into three basic systems: a Vault Automation System, an NDA automation System, and a Drum handling System. The Vault Automation system provides a mechanism to handle a basket of material cans and to load/unload storage tubes within the material vault. In addition, another robot is provided to load/unload material cans within the baskets. The NDA Automation System provides a mechanism to move material within the small canister NDA laboratory and to load/unload the NDA instruments. The Drum Handling System consists of a series of off the shelf components used to assist in lifting heavy objects such as pallets of material or drums and barrels

  5. 340 waste handling facility interim safety basis

    Energy Technology Data Exchange (ETDEWEB)

    VAIL, T.S.

    1999-04-01

    This document presents an interim safety basis for the 340 Waste Handling Facility classifying the 340 Facility as a Hazard Category 3 facility. The hazard analysis quantifies the operating safety envelop for this facility and demonstrates that the facility can be operated without a significant threat to onsite or offsite people.

  6. 340 waste handling facility interim safety basis

    International Nuclear Information System (INIS)

    VAIL, T.S.

    1999-01-01

    This document presents an interim safety basis for the 340 Waste Handling Facility classifying the 340 Facility as a Hazard Category 3 facility. The hazard analysis quantifies the operating safety envelop for this facility and demonstrates that the facility can be operated without a significant threat to onsite or offsite people

  7. Adequate technologies for wireless real-time dose rate monitoring for off-site emergency management

    International Nuclear Information System (INIS)

    Dielmann, R.; Buerkin, W.

    2003-01-01

    Full text: What are the requirements for off-site gamma dose rate monitoring systems? What are the pros and cons of available communication technologies? This report gives an overview of modern communication techniques and their applicability for reliable real-time data acquisition as basis for off-site nuclear emergency management. The results of three years operating experience with a wireless gamma dose rate monitoring system, installed around the NPPs of KURSK, KALININ and BALAKOVA (Russia) in the year 2000, are shown. (author)

  8. Dry cask handling system for shipping nuclear fuel

    International Nuclear Information System (INIS)

    Jones, C.R.

    1975-01-01

    A nuclear facility is described for improved handling of a shipping cask for nuclear fuel. After being brought into the building, the cask is lowered into a tank mounted on a transporter, which then carries the tank into a position under an auxiliary well to which it is sealed. Fuel can then be loaded into or unloaded from the cask via the auxiliary well which is flooded. Throughout the procedure, the cask surface remains dry. (U.S.)

  9. Analysis of operational possibilities and conditions of remote handling systems in nuclear facilities

    International Nuclear Information System (INIS)

    Hourfar, D.

    1989-01-01

    Accepting the development of the occupational radiation exposure in nuclear facilities, it will be showing possibilities of cost effective reduction of the dose rate through the application of robots and manipulators for the maintenance of nuclear power plants, fuel reprocessing plants, decommissioning and dismantling of the mentioned plants. Based on the experiences about industrial robot applications by manufacturing and manipulator applications by the handling of radioactive materials as well as analysis of the handling procedures and estimation of the dose intensity, it will be defining task-orientated requirements for the conceptual design of the remote handling systems. Furthermore the manifold applications of stationary and mobil arranged handling systems in temporary or permanent operation are described. (orig.) [de

  10. Off-site response for radiological emergencies

    International Nuclear Information System (INIS)

    Eldridge, J.S.; Oakes, T.W.; Hubbard, H.M.; Hibbitts, H.W.

    1982-01-01

    Environmental radiological surveillance under emergency conditions at off-site locations is one of the advisory functions provided by DOE within the ORO jurisdiction. The Department of Environmental Management of ORNL has been requested to provide sampling and analytical assistance at such emergency response activities. We have assembled and identified specific individuals and equipment to provide a rapid response force to perform field measurements for environmental radioactivity releases as a consequence of nuclear accidents. Survey teams for sample collection and field measurements are provided along with analytical assistance to operate the radioactivity measuring equipment in the DOE emergency van

  11. Robotics and remote handling concepts for disposal of high-level nuclear waste

    International Nuclear Information System (INIS)

    McAffee, Douglas; Raczka, Norman; Schwartztrauber, Keith

    1997-01-01

    This paper summarizes preliminary remote handling and robotic concepts being developed as part of the US Department of Energy's (DOE) Yucca Mountain Project. The DOE is currently evaluating the Yucca Mountain Nevada site for suitability as a possible underground geologic repository for the disposal of high level nuclear waste. The current advanced conceptual design calls for the disposal of more than 12,000 high level nuclear waste packages within a 225 km underground network of tunnels and emplacement drifts. Many of the waste packages may weigh as much as 66 tonnes and measure 1.8 m in diameter and 5.6 m long. The waste packages will emit significant levels of radiation and heat. Therefore, remote handling is a cornerstone of the repository design and operating concepts. This paper discusses potential applications areas for robotics and remote handling technologies within the subsurface repository. It also summarizes the findings of a preliminary technology survey which reviewed available robotic and remote handling technologies developed within the nuclear, mining, rail and industrial robotics and automation industries, and at national laboratories, universities, and related research institutions and government agencies

  12. Extensions to the Joshua GDMS to support environmental science and analysis data handling requirements

    International Nuclear Information System (INIS)

    Suich, J.E.; Honeck, H.C.

    1978-01-01

    For the past ten years, a generalized data management system (GDMS) called JOSHUA has been in use at the Savannah River Laboratory. Originally designed and implemented to support nuclear reactor physics and safety computational applications, the system is now also supporting environmental science modeling and impact assessment. Extensions to the original system are being developed to meet neet new data handling requirements, which include more general owner-member record relationships occurring in geographically encoded data sets, unstructured (relational) inquiry capability, cartographic analysis and display, and offsite data exchange. This paper discusses the need for these capabilities, places them in perspective as generic scientific data management activities, and presents the planned context-free extensions to the basic JOSHUA GDMS

  13. Extensions to the Joshua GDMS to support environmental science and analysis data handling requirements

    International Nuclear Information System (INIS)

    Suich, J.E.; Honeck, H.C.

    1977-01-01

    For the past ten years, a generalized data management system (GDMS) called JOSHUA has been in use at the Savannah River Laboratory. Originally designed and implemented to support nuclear reactor physics and safety computational applications, the system is now also supporting environmental science modeling and impact assessment. Extensions to the original system are being developed to meet new data handling requirements, which include more general owner-member record relationships occurring in geographically encoded data sets, unstructured (relational) inquiry capability, cartographic analysis and display, and offsite data exchange. This paper discusses the need for these capabilities, places them in perspective as generic scientific data management activities, and presents the planned context-free extensions to the basic JOSHUA GDMS

  14. Criteria for preparation and evaluation of radiological emergency response plans and preparedness in support of nuclear power plants: Criteria for utility offsite planning and preparedness: Final report

    International Nuclear Information System (INIS)

    Podolak, E.M. Jr.; Sanders, M.E.; Wingert, V.L.; Donovan, R.W.

    1988-09-01

    The Nuclear Regulatory Commission (NRC) and the Federal Emergency Management Agency (FEMA) have added a supplement to NUREG-0654/FEMA-REP-1, Rev. 1 that provides guidance for the development, review, and evaluation of utility offsite radiological emergency response planning and preparedness for those situations in which state and/or local governments decline to participate in emergency planning. While this guidance primarily applies to plants that do not have full-power operating licenses, it does have relevance to operating nuclear power plants

  15. Storage, handling and movement of fuel and related components at nuclear power plants

    International Nuclear Information System (INIS)

    1979-01-01

    The report describes in general terms the various operations involved in the handling of fresh fuel, irradiated fuel, and core components such as control rods, neutron sources, burnable poisons and removable instruments. It outlines the principal safety problems in these operations and provides the broad safety criteria which must be observed in the design, operation and maintenance of equipment and facilities for handling, transferring, and storing nuclear fuel and core components at nuclear power reactor sites

  16. Handling and carrying head for nuclear fuel assemblies and installation including this head

    International Nuclear Information System (INIS)

    Artaud, R.; Cransac, J.P.; Jogand, P.

    1986-01-01

    The present invention proposes a handling and carrying head ensuring efficiently the cooling of the nuclear fuel asemblies it transports so that any storage in liquid metal in a drum within or adjacent the reactor vessel is suppressed. The invention claims also a nuclear fuel handling installation including the head; it allows a longer time between loading and unloading campaigns and the space surrounding the reactor vessel keeps free without occupying a storage zone within the vessel [fr

  17. 40 CFR 68.22 - Offsite consequence analysis parameters.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 15 2010-07-01 2010-07-01 false Offsite consequence analysis... PROGRAMS (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Hazard Assessment § 68.22 Offsite consequence analysis parameters. (a) Endpoints. For analyses of offsite consequences, the following endpoints shall be...

  18. Losses of off-site power at U.S. nuclear power plants - all years through 1985. Final report, May 1986

    International Nuclear Information System (INIS)

    1986-05-01

    This report provides a database describing losses of all off-site power at U.S. nuclear plants. It includes all years through 1985. During 1985 there were 2 losses lasting longer than 30 minutes for 0.031 losses per site year and 4 shorter losses for 0.063 losses per site year, giving a total of 0.094 losses per site year. This is in line with the 1984 total of 0.089 losses per site year, and with the comparable numbers for all years through 1985 of 0.038 losses per site year of longer than 30 minutes, 0.045 for shorter losses and a total of 0.083. (author)

  19. Spent nuclear fuel characterization for a bounding reference assembly for the receiving basin for off-site fuel

    International Nuclear Information System (INIS)

    Kahook, S.D.; Garrett, R.L.; Canas, L.R.

    1995-01-01

    A basis for interim operation 1 (BIO) for the receiving basin for off-site fuel (RBOF) facility at the U.S. Department of Energy's (DOE) Savannah River site nuclear materials production complex has been developed in accordance to draft DOE-STD-0019-93 (Ref. 2). The latter document requires a hazard categorization per DOE-STD-1027-92 (Ref. 3) for the safety analysis portion of the BIO. This classification places the facility in one of three categories as defined in DOE 5480.23 (Ref. 4) per the total radioactivity, which can be released during an accident. The diversity of spent nuclear fuels stored in the RBOF made an exacting assessment of the total radioactive inventory virtually impossible. This restriction led to a conservative calculation based on the concept of a hypothetical bounding reference fuel assembly (RFA) integrated over the total capacity of the facility. The RFA is derived from a systematic ranking of the real assemblies (current and expected) according to a maximum burnup criterion. The indicated scheme is not only simple but precluded a potential delay in the completion of the BIO

  20. Development of web-based Off-Site Consequence Analysis Program (OSCAP) for extending ILRT intervals and its application

    International Nuclear Information System (INIS)

    Jeon, Ho-Jun; Hwang, Seok-Won; Oh, Ji-Yong

    2012-01-01

    Highlights: ► We develop web-based offsite consequence analysis program based on MACCS II code. ► The program has an automatic processing module to make the main input data. ► It is effective in conducting risk assessments according to extending ILRT intervals. ► Even a beginner can perform offsite consequence analysis with the program. - Abstract: For an offsite consequence analysis, MELCOR Accident Consequence Code System (MACCS) II code is widely used as a tool. In this study, the algorithm of web-based Off-Site Consequence Analysis Program (OSCAP) using the MACCS II code was developed for an integrated leak rate test (ILRT) interval extension and Level 3 probabilistic safety assessment (PSA), and verification and validation (V and V) of the program was performed. The main input data of the MACCS II code are meteorological data, population distribution data and source term data. However, it requires lots of time and efforts to generate the main input data for an offsite consequence analysis using the MACCS II code. For example, the meteorological data are collected from each nuclear power site in real time, but the formats of the raw data collected are different from each other as a site. To reduce efforts and time for risk assessments, the web-based OSCAP has an automatic processing module which converts the format of the raw data collected from each site in Korea to the input data format of the MACCS II code. The program also provides an automatic function of converting the latest population data from Statistics Korea, the National Statistical Office, to the population distribution input data format of the MACCS II code. In case of the source term data, the program includes the release fraction of each source term category resulting from Modular Accident Analysis Program (MAAP) code analysis and the core inventory data from ORIGEN code analysis. These analysis results of each plant in Korea are stored in a database module of the web-based OSCAP, so a

  1. Benchmarking the New RESRAD-OFFSITE Source Term Model with DUST-MS and GoldSim - 13377

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, J.J.; Kamboj, S.; Gnanapragasam, E.; Yu, C. [Argonne National Laboratory, Argonne, IL 60439 (United States)

    2013-07-01

    RESRAD-OFFSITE is a computer code developed by Argonne National Laboratory under the sponsorship of U.S. Department of Energy (DOE) and U.S. Nuclear Regulatory Commission (NRC). It is designed on the basis of RESRAD (onsite) code, a computer code designated by DOE and NRC for evaluating soil-contaminated sites for compliance with human health protection requirements pertaining to license termination or environmental remediation. RESRAD-OFFSITE has enhanced capabilities of modeling radionuclide transport to offsite locations and calculating potential radiation exposure to offsite receptors. Recently, a new source term model was incorporated into RESRAD-OFFSITE to enhance its capability further. This new source term model allows simulation of radionuclide releases from different waste forms, in addition to the soil sources originally considered in RESRAD (onsite) and RESRAD-OFFSITE codes. With this new source term model, a variety of applications can be achieved by using RESRAD-OFFSITE, including but not limited to, assessing the performance of radioactive waste disposal facilities. This paper presents the comparison of radionuclide release rates calculated by the new source term model of RESRAD-OFFSITE versus those calculated by DUST-MS and GoldSim, respectively. The focus of comparison is on the release rates of radionuclides from the bottom of the contaminated zone that was assumed to contain radioactive source materials buried in soil. The transport of released contaminants outside of the primary contaminated zone is beyond the scope of this paper. Overall, the agreement between the RESRAD-OFFSITE results and the DUST-MS and GoldSim results is fairly good, with all three codes predicting identical or similar radionuclide release profiles over time. Numerical dispersion in the DUST-MS and GoldSim results was identified as potentially contributing to the disagreement in the release rates. In general, greater discrepancy in the release rates was found for short

  2. Benchmarking the New RESRAD-OFFSITE Source Term Model with DUST-MS and GoldSim - 13377

    International Nuclear Information System (INIS)

    Cheng, J.J.; Kamboj, S.; Gnanapragasam, E.; Yu, C.

    2013-01-01

    RESRAD-OFFSITE is a computer code developed by Argonne National Laboratory under the sponsorship of U.S. Department of Energy (DOE) and U.S. Nuclear Regulatory Commission (NRC). It is designed on the basis of RESRAD (onsite) code, a computer code designated by DOE and NRC for evaluating soil-contaminated sites for compliance with human health protection requirements pertaining to license termination or environmental remediation. RESRAD-OFFSITE has enhanced capabilities of modeling radionuclide transport to offsite locations and calculating potential radiation exposure to offsite receptors. Recently, a new source term model was incorporated into RESRAD-OFFSITE to enhance its capability further. This new source term model allows simulation of radionuclide releases from different waste forms, in addition to the soil sources originally considered in RESRAD (onsite) and RESRAD-OFFSITE codes. With this new source term model, a variety of applications can be achieved by using RESRAD-OFFSITE, including but not limited to, assessing the performance of radioactive waste disposal facilities. This paper presents the comparison of radionuclide release rates calculated by the new source term model of RESRAD-OFFSITE versus those calculated by DUST-MS and GoldSim, respectively. The focus of comparison is on the release rates of radionuclides from the bottom of the contaminated zone that was assumed to contain radioactive source materials buried in soil. The transport of released contaminants outside of the primary contaminated zone is beyond the scope of this paper. Overall, the agreement between the RESRAD-OFFSITE results and the DUST-MS and GoldSim results is fairly good, with all three codes predicting identical or similar radionuclide release profiles over time. Numerical dispersion in the DUST-MS and GoldSim results was identified as potentially contributing to the disagreement in the release rates. In general, greater discrepancy in the release rates was found for short

  3. Analysis of loss of offsite power events reported in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Volkanovski, Andrija, E-mail: Andrija.VOLKANOVSKI@ec.europa.eu [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Ballesteros Avila, Antonio; Peinador Veira, Miguel [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Kančev, Duško [Kernkraftwerk Goesgen-Daeniken AG, CH-4658 Daeniken (Switzerland); Maqua, Michael [Gesellschaft für Anlagen-und-Reaktorsicherheit (GRS) gGmbH, Schwertnergasse 1, 50667 Köln (Germany); Stephan, Jean-Luc [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), BP 17 – 92262 Fontenay-aux-Roses Cedex (France)

    2016-10-15

    Highlights: • Loss of offsite power events were identified in four databases. • Engineering analysis of relevant events was done. • The dominant root cause for LOOP are human failures. • Improved maintenance procedures can decrease the number of LOOP events. - Abstract: This paper presents the results of analysis of the loss of offsite power events (LOOP) in four databases of operational events. The screened databases include: the Gesellschaft für Anlagen und Reaktorsicherheit mbH (GRS) and Institut de Radioprotection et de Sûreté Nucléaire (IRSN) databases, the IAEA International Reporting System for Operating Experience (IRS) and the U.S. Licensee Event Reports (LER). In total 228 relevant loss of offsite power events were identified in the IRSN database, 190 in the GRS database, 120 in U.S. LER and 52 in IRS database. Identified events were classified in predefined categories. Obtained results show that the largest percentage of LOOP events is registered during On power operational mode and lasted for two minutes or more. The plant centered events is the main contributor to LOOP events identified in IRSN, GRS and IAEA IRS database. The switchyard centered events are the main contributor in events registered in the NRC LER database. The main type of failed equipment is switchyard failures in IRSN and IAEA IRS, main or secondary lines in NRC LER and busbar failures in GRS database. The dominant root cause for the LOOP events are human failures during test, inspection and maintenance followed by human failures due to the insufficient or wrong procedures. The largest number of LOOP events resulted in reactor trip followed by EDG start. The actions that can result in reduction of the number of LOOP events and minimize consequences on plant safety are identified and presented.

  4. Survey of technology for decommissioning of nuclear fuel cycle facilities. 8. Remote handling and cutting techniques

    Energy Technology Data Exchange (ETDEWEB)

    Ogawa, Ryuichiro; Ishijima, Noboru [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1999-03-01

    In nuclear fuel cycle facility decommissioning and refurbishment, the remote handling techniques such as dismantling, waste handling and decontamination are needed to reduce personnel radiation exposure. The survey research for the status of R and D activities on remote handling tools suitable for nuclear facilities in the world and domestic existing commercial cutting tools applicable to decommissioning of the facilities was conducted. In addition, the drive mechanism, sensing element and control system applicable to the remote handling devices were also surveyed. This report presents brief surveyed summaries. (H. Itami)

  5. The handling of nuclear emergencies in Argentina

    International Nuclear Information System (INIS)

    Hernandez, Daniel; Jordan, Osvaldo; Kunst, Juan; Bruno, Hector

    2008-01-01

    Full text: In 1998, the Executive signed the decree 1390, which defined the scope and the procedures corresponding to the Nuclear Activity Law. In this decree, the new functions of the Nuclear Regulatory Authority (ARN) are described, being the most important related to preparation and response for a nuclear emergency the following ones: 1) ARN must provide protection from harmful effects of ionizing radiations under normal conditions and emergency situations; 2) ARN must advise the Executive in case of radiological and nuclear emergencies; 3) ARN shall establish the criteria for the emergency plans of the facilities and train the members of neighbor public to the facilities in case of nuclear emergencies; 4) The emergency plans developed by local, provincial and national authorities must be approved by the ARN; 5) ARN shall lead the actions within the area covered by the emergency plans of the facilities. Security Forces and the Representatives of Civil Institutions shall report the designated ARN officer. The ARN recognized immediately the responsibility imposed by this law and, at the same time, the opportunity of improving the handling of emergencies through a centralized direction of the operations. Under this frame, ARN created the Radiological Emergencies Intervention System (SIER) with the goal of taking charge of the preparation and the handling of emergency situations. From the beginning, the purpose of the SIER was to improve the preparation and response to nuclear emergencies in a regular form, bearing in mind the cultural and socioeconomic situation of the country, as well as the local peculiarities. The first steep to achieve such a target was to gain the confidence of other organizations included in the response on the ARN technical and operational aptitude to lead the actions inside the emergency area and, later, to establish the pertinent arrangements. The strategy chosen by ARN to respond to nuclear emergencies consists in establishing an expert

  6. Handling final storage of unreprocessed spent nuclear fuel

    International Nuclear Information System (INIS)

    1978-01-01

    The present second report from KBS describes how the safe final storage of spent unreprocessed nuclear fuel can be implemented. According to the Swedish Stipulation Law, the owner must specify in which form the waste is to be stored, how final storage is to be effected, how the waste is to be transported and all other aspects of fuel handling and storage which must be taken into consideration in judging whether the proposed final storage method can be considered to be absolutely safe and feasible. Thus, the description must go beyond general plans and sketches. The description is therefore relatively detailed, even concerning those parts which are less essential for evaluating the safety of the waste storage method. For those parts of the handling chain which are the same for both alternatives of the Stipulation Law, the reader is referred in some cases to the first report. Both of the alternatives of the Stipulation Law may be used in the future. Handling equipment and facilities for the two storage methods are so designed that a combination in the desired proportions is practically feasible. In this first part of the report are presented: premises and data, a description of the various steps of the handling procedure, a summary of dispersal processes and a safety analysis. (author)

  7. 18 years experience on UF{sub 6} handling at Japanese nuclear fuel manufacturer

    Energy Technology Data Exchange (ETDEWEB)

    Fujinaga, H.; Yamazaki, N.; Takebe, N. [Japan Nucelar Fuel Conversion Co., Ltd., Ibaraki (Japan)

    1991-12-31

    In the spring of 1991, a leading nuclear fuel manufacturing company in Japan, celebrated its 18th anniversary. Since 1973, the company has produced over 5000 metric ton of ceramic grade UO{sub 2} powder to supply to Japanese fabricators, without major accident/incident and especially with a successful safety record on UF{sub 6} handling. The company`s 18 years experience on nuclear fuel manufacturing reveals that key factors for the safe handling of UF{sub 6} are (1) installing adequate facilities, equipped with safety devices, (2) providing UF{sub 6} handling manuals and executing them strictly, and (3) repeating on and off the job training for operators. In this paper, equipment and the operation mode for UF{sub 6} processing at their facility are discussed.

  8. Nuclear robotics and remote handling at Harwell Laboratory

    International Nuclear Information System (INIS)

    Abel, E.; Brown, M.H.; Fischer, P.J.; Garlick, D.R.; Hanna, T.T.; Siva, K.V.

    1988-01-01

    After reviewing robotics technology and its possible application in nuclear remote handling systems of the future, six main research topics were identified where particular effort should be made. The Harwell Nuclear Robotics Programme is currently establishing sets of demonstration hardware which will allow generic research to be carried out on telerobotics, systems integration, the man machine interface, communications, servo systems and radiation tolerance. The objectives of the demonstrators are to allow validation of the techniques required for successful active facility applications such as decommissioning, decontamination, refurbishment, maintenance and repair, and to act as training aids to encourage plant designers and operators to adopt developments in new technology. (author)

  9. Development of nuclear fuel microsphere handling techniques and equipment

    International Nuclear Information System (INIS)

    Mack, J.E.; Suchomel, R.R.; Angelini, P.

    1979-01-01

    Considerable progress has been made in the development of microsphere handling techniques and equipment for nuclear applications. Work at Oak Ridge National Laboratory with microspherical fuel forms dates back to the early sixties with the development of the sol-gel process. Since that time a number of equipment items and systems specifically related to microsphere handling and characterization have been identified and developed for eventual application in a remote recycle facility. These include positive and negative pressure transfer systems, samplers, weighers, a blender-dispenser, and automated devices for particle size distribution and crushing strength analysis. The current status of these and other components and systems is discussed

  10. On-site and off-site emergency planning at Cekmece Nuclear Research and Training Centre

    International Nuclear Information System (INIS)

    Soyberk, O.A.

    1986-01-01

    An emergency plan was prepared for minimizing the consequences of any unforeseen radiation accident in Cekmece Nuclear Research and Training Centre (CNAEM) in Istanbul, Turkey. CNAEM is situated near Kucukcekmece Lake, which is about 30 km to the west of Istanbul. It includes two pool-type research reactors of 1 MW(th) and 5 MW(th). The population in the nearest inhabited areas varies from 1000 to 50,000. Accidents are classified, according to their severity, into three categories at CNAEM: (a) local emergency, (b) on-site emergency, (c) off-site emergency. During local emergency situations evacuation is not necessary. An on-site emergency situation requires the evacuation of personnel from the plant. Personnel hearing the emergency alarm should move directly to the preselected place as soon as possible. An off-site emergency is any accident that leads to widespread contamination outside the boundary. In this situation the Turkish Atomic Energy Authority and governmental authorities are notified immediately. The emergency organization group consists of: (a) Plant Superintendent, (b) Emergency Director, (c) Reactor Supervisor, (d) Senior Health Physicist, (e) Reactor Shift Operator, (f) Health Physicists. The administration building will be used as the Emergency Control Centre. The emergency teams working under the direction of the Emergency Director consist of: (a) Health Physics, (b) Fire and Rescue, (c) First Aid and Decontamination, (d) Transportation, (e) Security and Patrol. The emergency situation is evaluated in three phases at CNAEM. The first phase is the first few hours after the beginning of the accident. The second phase is between 8-10 hours or more following the first phase. The third phase is the recovery phase. The integrated doses over periods of two hours and two days are calculated according to the situation of the core, i.e. total or partial melting, and weather conditions. The results of the calculated parameters can be adapted to possible

  11. Generalized Nuclear Data: A New Structure (with Supporting Infrastructure) for Handling Nuclear Data

    International Nuclear Information System (INIS)

    Mattoon, C.M.; Beck, B.R.; Patel, N.R.; Summers, N.C.; Hedstrom, G.W.; Brown, D.A.

    2012-01-01

    The Evaluated Nuclear Data File (ENDF) format was designed in the 1960s to accommodate neutron reaction data to support nuclear engineering applications in power, national security and criticality safety. Over the years, the scope of the format has been extended to handle many other kinds of data including charged particle, decay, atomic, photo-nuclear and thermal neutron scattering. Although ENDF has wide acceptance and support for many data types, its limited support for correlated particle emission, limited numeric precision, and general lack of extensibility mean that the nuclear data community cannot take advantage of many emerging opportunities. More generally, the ENDF format provides an unfriendly environment that makes it difficult for new data evaluators and users to create and access nuclear data. The Cross Section Evaluation Working Group (CSEWG) has begun the design of a new Generalized Nuclear Data (or 'GND') structure, meant to replace older formats with a hierarchy that mirrors the underlying physics, and is aligned with modern coding and database practices. In support of this new structure, Lawrence Livermore National Laboratory (LLNL) has updated its nuclear data/reactions management package Fudge to handle GND structured nuclear data. Fudge provides tools for converting both the latest ENDF format (ENDF-6) and the LLNL Evaluated Nuclear Data Library (ENDL) format to and from GND, as well as for visualizing, modifying and processing (i.e., converting evaluated nuclear data into a form more suitable to transport codes) GND structured nuclear data. GND defines the structure needed for storing nuclear data evaluations and the type of data that needs to be stored. But unlike ENDF and ENDL, GND does not define how the data are to be stored in a file. Currently, Fudge writes the structured GND data to a file using the eXtensible Markup Language (XML), as it is ASCII based and can be viewed with any text editor. XML is a meta-language, meaning that it

  12. Generalized Nuclear Data: A New Structure (with Supporting Infrastructure) for Handling Nuclear Data

    Energy Technology Data Exchange (ETDEWEB)

    Mattoon, C.M. [Lawrence Livermore National Laboratory, 7000 East Avenue, Livermore CA (United States); Beck, B.R.; Patel, N.R.; Summers, N.C.; Hedstrom, G.W. [Lawrence Livermore National Laboratory, 7000 East Avenue, Livermore CA (United States); Brown, D.A. [National Nuclear Data Center, Upton NY (United States)

    2012-12-15

    The Evaluated Nuclear Data File (ENDF) format was designed in the 1960s to accommodate neutron reaction data to support nuclear engineering applications in power, national security and criticality safety. Over the years, the scope of the format has been extended to handle many other kinds of data including charged particle, decay, atomic, photo-nuclear and thermal neutron scattering. Although ENDF has wide acceptance and support for many data types, its limited support for correlated particle emission, limited numeric precision, and general lack of extensibility mean that the nuclear data community cannot take advantage of many emerging opportunities. More generally, the ENDF format provides an unfriendly environment that makes it difficult for new data evaluators and users to create and access nuclear data. The Cross Section Evaluation Working Group (CSEWG) has begun the design of a new Generalized Nuclear Data (or 'GND') structure, meant to replace older formats with a hierarchy that mirrors the underlying physics, and is aligned with modern coding and database practices. In support of this new structure, Lawrence Livermore National Laboratory (LLNL) has updated its nuclear data/reactions management package Fudge to handle GND structured nuclear data. Fudge provides tools for converting both the latest ENDF format (ENDF-6) and the LLNL Evaluated Nuclear Data Library (ENDL) format to and from GND, as well as for visualizing, modifying and processing (i.e., converting evaluated nuclear data into a form more suitable to transport codes) GND structured nuclear data. GND defines the structure needed for storing nuclear data evaluations and the type of data that needs to be stored. But unlike ENDF and ENDL, GND does not define how the data are to be stored in a file. Currently, Fudge writes the structured GND data to a file using the eXtensible Markup Language (XML), as it is ASCII based and can be viewed with any text editor. XML is a meta-language, meaning that it

  13. The design of in-cell crane handling systems for nuclear plants

    International Nuclear Information System (INIS)

    Hansford, S.M.; Scott, R.

    1992-01-01

    The reprocessing and waste management facilities at (BNFL's) British Nuclear Fuels Limited's Sellafield site make extensive use of crane handling systems. These range from conventional mechanical handling operations as used generally in industry to high integrity applications through to remote robotic handling operations in radiation environments. This paper describes the design methodologies developed for the design of crane systems for remote handling operations - in-cell crane systems. In most applications the in-cell crane systems are an integral part of the plant process equipment and reliable and safe operations are a key design parameter. Outlined are the techniques developed to achieve high levels of crane system availability for operations in hazardous radiation environments. These techniques are now well established and proven through many years of successful plant operation. A recent application of in-cell crane handling systems design for process duty application is described. The benefits of a systematic design approach and a functionally-based engineering organization are also highlighted. (author)

  14. 40 CFR 273.55 - Off-site shipments.

    Science.gov (United States)

    2010-07-01

    ....55 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) STANDARDS FOR UNIVERSAL WASTE MANAGEMENT Standards for Universal Waste Transporters § 273.55 Off-site... universal waste being shipped off-site meets the Department of Transportation's definition of hazardous...

  15. 40 CFR 68.165 - Offsite consequence analysis.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 15 2010-07-01 2010-07-01 false Offsite consequence analysis. 68.165 Section 68.165 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Risk Management Plan § 68.165 Offsite consequence...

  16. Autonomous underwater handling system for service, measurement and cutting tasks for the decommissioning of nuclear facilities

    International Nuclear Information System (INIS)

    Hahn, M.; Haferkamp, H.; Bach, W.; Rose, N.

    1992-01-01

    For about 10 years the Institute for Material Science at the Hanover University has worked on projects of underwater cutting and welding. Increasing tasks to be done in nuclear facilities led to the development of special handling systems to support and handle the cutting tools. Also sensors and computers for extensive and complex tasks were integrated. A small sized freediving handling system, equipped with 2 video cameras, ultrasonic and radiation sensors and a plasma cutting torch for inspection and decommissioning tasks in nuclear facilities is described in this paper. (Author)

  17. A model for the calculation of the off-site economic consequences of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Gallego, E.; Alonso, A.

    1988-01-01

    The off-site economic cost of nuclear reactor accidents will depend on the countermeasures adopted to reduce its radiological impact. The assessment of the direct costs of emergency countermeasures (evacuation, early relocation and food disposal) as well as those of long-term protective actions (food disposal, decontamination or interdiction) is the objective of a model under development, with the sponsorship of the CEC Radiation Protection Programme, called MECA (Model for assessing the Economic Consequences of Accidents). The meteorological and socio-economical peculiarities of each site studied will be taken into account, by means of a flexible meteorological sampling scheme, which considers the geographical distribution of population and economic centers, and a data-base, compatible with the existing European grid, that contains the population distribution and the economic characteristics of the environs of the site to be studied with more detail near the reactor. The paper summarizes the particular models which will be included in MECA and shows the importance of site-specific adaptable modelling for economic risk evaluation

  18. Transient evaluation using EMTP at single open phase with the offsite power transformer for the emergency power supply systems of nuclear power plants

    International Nuclear Information System (INIS)

    Shimada, Yoshio

    2017-01-01

    The emergency power supply systems of nuclear power plants, as the objects of this research, are critical in supplying stable electric power to such systems as the emergency core cooling system (ECCS), and in maintaining safety of the nuclear power reactor; this was apparent in the accident at the Fukushima Daiichi Nuclear Power Station. The Nuclear Regulatory Commission (USNRC) issued regulatory documents (BL 2012-011, IN 2012-032), and has commenced evaluations on newly discovered vulnerability in the design of power supply systems which cannot be detected with under-voltage protection relays, with certain kinds of configuration of coils and iron core structures, such as when the offsite power supply side is a Y-connection and the load side is a ⊿-connection etc., when the detection of single open phase fault with the circuit of a transformer which is without a ground fault connected to the offsite power supply system. This report uses simulation by the electro magnetic transients program (EMTP) and clearly describe the response at the time of the power supply single open phase without ground fault for various configuration of coils and various iron core structures of the three-phase transformer, and identify the important issues in the response of emergency power supply systems and the safety related components of representative domestic PWR plants when the single open phase fault occurred without ground fault. This report describes the results of the simulations of operations of the protection relays of the emergency power supply systems and the safety related components of representative a domestic PWR plant with EMTP. This report explains the method to detect open-phase when the transformer is no-load which United States Electric Power Research Institute (EPRI) developed. As the detailed analyses data from EPRI related to the detection method concerned have not been disclosed officially yet, in this paper, the quantitative and detailed verification results

  19. State of offsite construction in India-Drivers and barriers

    Science.gov (United States)

    Arif, M.; Bendi, D.; Sawhney, A.; Iyer, K. C.

    2012-05-01

    The rapid growth of the construction industry in India has influenced key players in the industry to adopt alternative technologies addressing time, cost and quality. The rising demand in housing, infrastructure and other facilities have further highlighted the need for the construction industry to look at adopting alternate building technologies. Offsite construction has evolved as a panacea to dealing with the under-supply and poor quality in the current age construction industry. Several offsite techniques have been adopted by the construction sector. Although, different forms of offsite techniques have been around for a while but their uptake has been low in the Indian context. This paper presents the perceptions about offsite construction in India and highlights some of the barriers and drivers facing the Indian construction industry. The data was gathered through a survey of 17 high level managers from some of the largest stakeholder organizations of the construction sector in India. The influence of time and cost has been highlighted as a major factor fuelling the adoption of offsite construction. However, the influence of current planning systems and the need for a paradigm shift are some of the prominent barriers towards the adoption of offsite techniques.

  20. Handling nuclear waste over long periods

    International Nuclear Information System (INIS)

    Ancelin, B.; Chenevier, E.

    1983-01-01

    The handling of nuclear waste over long periods throws up new problems, such as the safety for a very long term and the employment of economic logic in order to justify choices involving extended time scales. The result is a very great difficulty of apprehension of the problem by the specialists as well as by the public. A clear policy decision, associated with a coherent administrative organization, will therefore have to make up for an impossible technical-economical optimization of the various possible options. The difficulty of simple technical choices is only going to reinforce this wish; the absence of a global and comparative measuring system is responsible for the fact that in this field the passions often override many of the scientific truths [fr

  1. Determinations of TSD facility acceptability under the CERCLA Off-Site Rule

    International Nuclear Information System (INIS)

    1997-06-01

    On September 22, 1993, the US Environmental Protection Agency (EPA) published the ''Off-Site Rule'' to implement section 121(d)(3) of the Comprehensive Environmental Response Compensation and Liability Act (CERCLA). CERCLA section 121(d)(3) requires that wastes generated as a result of remediation activities taken under CERCLA authority and transferred off-site be managed only at facilities that comply with the Resource Conservation and Recovery Act. In 1994, the DOE's Office of Environmental Policy and Assistance (OEPA), RCRA/CERCLA Division (EH-413) published a CERCLA Information Brief titled ''The Off-Site Rule'' which describes the content of the Off-Site Rule and clarifies some of its implications for DOE remedial actions under CERCLA. Additionally, EH-413 published the Guide on Selecting Compliant Off-Site Hazardous Waste Treatment, Storage and Disposal Facilities which provides a regulatory roadmap for accomplishing off-site transfers of environmental restoration and process hazardous waste at DOE facilities in a manner compliant with the Off-Site Rule and other relevant Federal regulations. Those guidance documents concentrate primarily on DOE's perspective as a hazardous waste generator. The purpose of this Information Brief is to address the implications of the Off-Site Rule for DOE-owned hazardous waste treatment, storage or disposal facilities that accept CERCLA remediation wastes from off-site locations

  2. Nuclear fuel handling grapple carriage with self-lubricating bearing

    International Nuclear Information System (INIS)

    Wade, E.E.

    1978-01-01

    Disclosed is a nuclear fuel handling grapple carriage having a bearing with a lubricant reservoir that is capable of being refilled when the bearing and reservoir are submerged in a lubricant pool. The lubricant reservoir supplies lubricant to the bearing while the bearing allows a small amount of lubricant to leak passed appropriately placed seals creating a positive out flow of lubricant thereby preventing foreign material from entering the bearing

  3. On-Site or Off-Site Renewable Energy Supply Options?

    DEFF Research Database (Denmark)

    Marszal, Anna Joanna; Heiselberg, Per; Jensen, Rasmus Lund

    2012-01-01

    The concept of a Net Zero Energy Building (Net ZEB) encompasses two options of supplying renewable energy, which can offset energy use of a building, in particular on-site or off-site renewable energy supply. Currently, the on-site options are much more popular than the off-site; however, taking...... into consideration the limited area of roof and/or façade, primarily in the dense city areas, the Danish weather conditions, the growing interest and number of wind turbine co-ops, the off-site renewable energy supply options could become a meaningful solution for reaching ‘zero’ energy goal in the Danish context...... five technologies, i.e., two on-site options: (1) photovoltaic, (2) micro combined heat and power, and three off-site options: (1) off-site windmill, (2) share of a windmill farm and (3) purchase of green energy from the 100% renewable utility grid. The results indicate that in case of the on...

  4. An examination of the proposals for the off-site electrical power sources at the Sizewell B PWR

    Energy Technology Data Exchange (ETDEWEB)

    Woodhouse, P. A. [HM Nuclear Installations Inspectorate, London (United Kingdom)

    1986-02-15

    Over the past few years there has been an increase in the attention being given to the adequacy and reliability of alternative sources of power provided to supply safety equipment should off-site electrical sources fail. This paper discusses the rationale of HM Nuclear Installations Inspectorates assessment of the electrical systems proposed for the UK's first Pressurized water Reactor, Sizewell 3. The requirements for on-site sources are given, and a discussion is provided of the NII's Assessment Principles including common mode failure, single failure criterion and reliability targets. Where the assessment has resulted in notifications to the original design the reasons are given. The UK's large interconnected Grid System makes complete losses of off-site power comparatively rare. The potential exists however and this paper shows how the current approach ensures that not only are adequate on-site sources available but also that their siting, maintenance and testing are such that loss of off-site power will not cause an unacceptable risk to the public. (author)

  5. Towards a better mastery of risks in the handling of nuclear fuel: the contributions of ergonomics

    International Nuclear Information System (INIS)

    Samson, L.

    1999-01-01

    Nuclear fuel is handled under water in the reactor pool using procedures that have yet to be automated. The knowledge and skill of the operators is therefore of prime importance. Ergonomic consultants have prepared a report on the problems facing the operators when handling nuclear fuel? These problems have been addressed by the installation of a new system to detect and prevent incorrect operator commands and to provide software assistance in planning movements together with diagnostic functions. The new system has resulted in considerable time savings and a reduction in the risk of error. However, it has been necessary to modify the control software in the light of the handling strategies traditionally used by the operators. (author)

  6. Man/machine interface for a nuclear cask remote handling control station: system design requirements

    International Nuclear Information System (INIS)

    Clarke, M.M.; Kreifeldt, J.G.; Draper, J.V.

    1984-01-01

    Design requirements are presented for a control station of a proposed semi-automated facility for remote handling of nuclear waste casks. Functional and operational man/machine interface: controls, displays, software format, station architecture, and work environment. In addition, some input is given to the design of remote sensing systems in the cask handling areas. 18 references, 9 figures, 12 tables

  7. Off-site relations and emergency planning or the importance of being earnest

    International Nuclear Information System (INIS)

    Dunkle, M.K.

    1987-01-01

    Emergency planning is and will continue to be a vulnerable spot for the nuclear industry. Emergency planning issues can be reopened at any time during the life of the plant and this represents a threat that continues for the life of the plant. The area of planning in which utilities find themselves most vulnerable is off-site relations with the state and local government officials, the public, and even the news media. Utilities face two very basic challenges in developing and maintaining good off-site relations for emergency preparedness: (1) utility managers must understand and be capable of working with the myriad of personalities and dynamics in the emergency preparedness arena. (2) Emergency preparedness is an emotional issue and a technical subject not well understood by the average citizen. The public looks to well-founded emergency plans and strong leaders to effect them. With these, a sound communications strategy, and a good plant record, a utility stands a chance of achieving the real key to success, credibility

  8. User's Guide for RESRAD-OFFSITE

    Energy Technology Data Exchange (ETDEWEB)

    Gnanapragasam, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, C. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-04-01

    The RESRAD-OFFSITE code can be used to model the radiological dose or risk to an offsite receptor. This User’s Guide for RESRAD-OFFSITE Version 3.1 is an update of the User’s Guide for RESRAD-OFFSITE Version 2 contained in the Appendix A of the User’s Manual for RESRAD-OFFSITE Version 2 (ANL/EVS/TM/07-1, DOE/HS-0005, NUREG/CR-6937). This user’s guide presents the basic information necessary to use Version 3.1 of the code. It also points to the help file and other documents that provide more detailed information about the inputs, the input forms and features/tools in the code; two of the features (overriding the source term and computing area factors) are discussed in the appendices to this guide. Section 2 describes how to download and install the code and then verify the installation of the code. Section 3 shows ways to navigate through the input screens to simulate various exposure scenarios and to view the results in graphics and text reports. Section 4 has screen shots of each input form in the code and provides basic information about each parameter to increase the user’s understanding of the code. Section 5 outlines the contents of all the text reports and the graphical output. It also describes the commands in the two output viewers. Section 6 deals with the probabilistic and sensitivity analysis tools available in the code. Section 7 details the various ways of obtaining help in the code.

  9. User's Guide for RESRAD-OFFSITE

    International Nuclear Information System (INIS)

    Gnanapragasam, E.; Yu, C.

    2015-01-01

    The RESRAD-OFFSITE code can be used to model the radiological dose or risk to an offsite receptor. This User's Guide for RESRAD-OFFSITE Version 3.1 is an update of the User's Guide for RESRAD-OFFSITE Version 2 contained in the Appendix A of the User's Manual for RESRAD-OFFSITE Version 2 (ANL/EVS/TM/07-1, DOE/HS-0005, NUREG/CR-6937). This user's guide presents the basic information necessary to use Version 3.1 of the code. It also points to the help file and other documents that provide more detailed information about the inputs, the input forms and features/tools in the code; two of the features (overriding the source term and computing area factors) are discussed in the appendices to this guide. Section 2 describes how to download and install the code and then verify the installation of the code. Section 3 shows ways to navigate through the input screens to simulate various exposure scenarios and to view the results in graphics and text reports. Section 4 has screen shots of each input form in the code and provides basic information about each parameter to increase the user's understanding of the code. Section 5 outlines the contents of all the text reports and the graphical output. It also describes the commands in the two output viewers. Section 6 deals with the probabilistic and sensitivity analysis tools available in the code. Section 7 details the various ways of obtaining help in the code.

  10. Transport, acceptance, storage and handling of the itens of nuclear power plants

    International Nuclear Information System (INIS)

    1989-01-01

    The norm aiming to establish the requirements applied to workers or organizations which participate of the activities of transport, acceptance, storage and handling of important itens for safety of nuclear power plants, is presented. The established requirements treat of protection and control necessary to assure that the quality of important itens for safety be it preserved from the end of fabrication until their incorporation to nuclear power plant. (M.C.K.) [pt

  11. Effects of secondary containment air cleanup system leakage on the accident offsite dose as determined during preop tests of the Sequoyah Nuclear Plant

    International Nuclear Information System (INIS)

    Klaes, L.J.; Nass, S.A.; Proctor, L.D.

    1981-01-01

    The Sequoyah Nuclear Plant has two secondary containments. One is the annular region between the primary containment and the shield building surrounding the primary containment. The second is the auxiliary building secondary containment enclosure which is potentially subject to direct airborne radioactivity. Two air cleanup systems are provided to serve these areas. The emergency gas treatment system (EGTS) serves the annulus between the primary containment and the shield building, and the auxiliary building gas treatment system (ABGTS) serves the area inside of the auxiliary building secondary containment enclosure. The major function served by these air cleanup systems is that of controlling and processing airborne contamination released in these areas during any accident up to a design basis accident. This is accomplished by (1) creating a negative pressure in the areas served to ensure that no unprocessed air is released to the atmosphere, (2) providing filtration units to process all air exhausted from the secondary containment spaces, and (3) providing a low-leakage enclosure to limit exhaust flows. Offsite dose effects due to secondary containment release rates, bypass leakage, and duct and damper leakages are presented and parameter variations are considered. For the EGTS, a recirculation system, the most important parameter is the total inleakage of the system which causes an increase in both whole body (gamma) and thyroid (iodine) doses. For the ABGTS, a once-through system, the most important paramter is the inleakage which bypasses the filters resulting in an increase in the thyroid dose only. Actual preoperational test data are utilized. Problems encountered during the preop test are summarized. Solutions incorporated to bring the EGTS and ABGTS air cleanup systems within the test acceptance criteria required to meet offsite dose limitations are discussed and the resultant calculated offsite dose is presented

  12. Application of GIS in prediction and assessment system of off-site accident consequence for NPP

    International Nuclear Information System (INIS)

    Wang Xingyu; Shi Zhongqi

    2002-01-01

    The assessment and prediction software system of off-site accident consequence for Guangdong Nuclear Power Plant (GNARD2.0) is a GIS-based software system. The spatial analysis of radioactive materials and doses with geographic information is available in this system. The structure and functions of the GNARD system and the method of applying ArcView GIS are presented

  13. A study on the food consumption rates for off-site radiological dose assessment around Korean Nuclear Power Plants

    International Nuclear Information System (INIS)

    Lee, Gab Bock; Chung, Yang Geun

    2008-01-01

    The internal dose by food consumption mostly accounts for radiological dose of public around Nuclear Power Plants (NPPs). But, food consumption rates applied to off-site dose calculation in Korea which are the result of field investigation around Kori NPP by the KAERI (Korea Atomic Energy Research Institute) in 1988, are not able to reflect the latest dietary characteristics of Korean. The food consumption rates to be used for radiological dose assessment in Korea are based on the maximum individual of US NRC (Nuclear Regulatory Commission) Regulatory Guide 1.109. However, the representative individual of the critical group is considered in the recent ICRP (International Commission on Radiological Protection) recommendation and European nations' practice. Therefore, the study on the re-establishment of the food consumption rates for individual around nuclear power plant sites in Korea was carried out to reflect on the recent change of the Korean dietary characteristics and to apply the representative individual of critical group to domestic regulations. The ministry of Health and Welfare Affairs has investigated the food and nutrition of nations every 3 years based on the Law of National Health Improvement. The statistical data such as mean, standard deviation, various percentile values about food consumption rates to be used for the representative individual of the critical group were analyzed by using the raw data of the national food consumption survey in 2001∼2002. Also, the food consumption rates for maximum individual are re-estimated

  14. Offsite transportation hazards assessment

    International Nuclear Information System (INIS)

    Burnside, M.E.

    1997-01-01

    This report documents the emergency preparedness Hazards Assessment for the offsite transportation of hazardous material from the Hanford Site. The assessment is required by the US Department of Energy (DOE) Order 151.1. Offsite transportation accidents are categorized using the DOE system to assist communication within the DOE and assure that appropriate assistance is provided to the people in charge at the scene. The assistance will initially include information about the load and the potential hazards. Local authorities will use the information to protect the public following a transportation accident. This Hazards Assessment will focus on the material being transported from the Hanford Site. Shipments coming to Hanford are the responsibility of the shipper and the carrier and, therefore, are not included in this Hazards Assessment, unless the DOE elects to be the shipper of record

  15. Robotic control architecture development for automated nuclear material handling systems

    International Nuclear Information System (INIS)

    Merrill, R.D.; Hurd, R.; Couture, S.; Wilhelmsen, K.

    1995-02-01

    Lawrence Livermore National Laboratory (LLNL) is engaged in developing automated systems for handling materials for mixed waste treatment, nuclear pyrochemical processing, and weapon components disassembly. In support of these application areas there is an extensive robotic development program. This paper will describe the portion of this effort at LLNL devoted to control system architecture development, and review two applications currently being implemented which incorporate these technologies

  16. RODOS: decision support system for off-site emergency management in Europe

    International Nuclear Information System (INIS)

    Ehrhardt, J.; Shershakov, V.; Zheleznyak, M.; Mikhalevich, A.

    1996-01-01

    The integrated and comprehensive real-time on-line decision support system, RODOS, for off-site emergency management of nuclear accidents is being developed under the auspices of the European Commission's Radiation Protection Research Action. A large number of both West and East European institutes are involved in the further development of the existing prototype versions to operational use with significant contributions coming from the partner institutes in the CIS Republics. This paper summarizes the structure, the main functions and the status of the RODOS system

  17. In-plant considerations for optimal offsite response to reactor accidents

    International Nuclear Information System (INIS)

    Burke, R.P.; Heising, C.D.; Aldrich, D.C.

    1982-11-01

    Offsite response decision-making methods based on in-plant conditions are developed for use during severe reactor-accident situations. Dose projections are used to eliminate all LWR plant systems except the reactor core and the spent-fuel storage pool from consideration for immediate offsite emergency response during accident situations. A simple plant information-management scheme is developed for use in offsite response decision-making. Detailed consequence calculations performed with the CRAC2 model are used to determine the appropriate timing of offsite-response implementation for a range of PWR accidents involving the reactor core. In-plant decision criteria for offsite-response implementation are defined. The definition of decision criteria is based on consideration of core-accident physical processes, in-plant accident monitoring information, and results of consequence calculations performed to determine the effectiveness of various public-protective measures. The benefits and negative aspects of the proposed response-implementation criteria are detailed

  18. Analysis of Loss-of-Offsite-Power Events 1997-2015

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, Nancy Ellen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-01

    Loss of offsite power (LOOP) can have a major negative impact on a power plant’s ability to achieve and maintain safe shutdown conditions. LOOP event frequencies and times required for subsequent restoration of offsite power are important inputs to plant probabilistic risk assessments. This report presents a statistical and engineering analysis of LOOP frequencies and durations at U.S. commercial nuclear power plants. The data used in this study are based on the operating experience during calendar years 1997 through 2015. LOOP events during critical operation that do not result in a reactor trip, are not included. Frequencies and durations were determined for four event categories: plant-centered, switchyard-centered, grid-related, and weather-related. Emergency diesel generator reliability is also considered (failure to start, failure to load and run, and failure to run more than 1 hour). There is an adverse trend in LOOP durations. The previously reported adverse trend in LOOP frequency was not statistically significant for 2006-2015. Grid-related LOOPs happen predominantly in the summer. Switchyard-centered LOOPs happen predominantly in winter and spring. Plant-centered and weather-related LOOPs do not show statistically significant seasonality. The engineering analysis of LOOP data shows that human errors have been much less frequent since 1997 than in the 1986 -1996 time period.

  19. General principles of the nuclear criticality safety for handling, processing and transportation fissile materials in the USSR

    International Nuclear Information System (INIS)

    Vnukov, V.S.; Rjazanov, B.G.; Sviridov, V.I.; Frolov, V.V.; Zubkov, Y.N.

    1991-01-01

    The paper describes the general principles of nuclear criticality safety for handling, processing, transportation and fissile materials storing. Measures to limit the consequences of critical accidents are discussed for the fuel processing plants and fissile materials storage. The system of scientific and technical measures on nuclear criticality safety as well as the system of control and state supervision based on the rules, limits and requirements are described. The criticality safety aspects for various stages of handling nuclear materials are considered. The paper gives descriptions of the methods and approaches for critical risk assessments for the processing facilities, plants and storages. (Author)

  20. Computer control of fuel handling activities at FFTF

    International Nuclear Information System (INIS)

    Romrell, D.M.

    1985-03-01

    The Fast Flux Test Facility near Richland, Washington, utilizes computer control for reactor refueling and other related core component handling and processing tasks. The computer controlled tasks described in this paper include core component transfers within the reactor vessel, core component transfers into and out of the reactor vessel, remote duct measurements of irradiated core components, remote duct cutting, and finally, transferring irradiated components out of the reactor containment building for off-site shipments or to long term storage. 3 refs., 16 figs

  1. The remote handling of canisters containing nuclear waste in glass at the Savannah River Plant

    International Nuclear Information System (INIS)

    Callan, J.E.

    1986-01-01

    The Defense Waste Processing Facility (DWPF) is a complete production area being constructed at the Savannah River Plant for the immobilization of nuclear waste in glass. The remote handling of canisters filled with nuclear waste in glass is an essential part of the process of the DWPF at the Savannah River Plant. The canisters are filled with nuclear waste containing up to 235,000 curies of radioactivity. Handling and movement of these canisters must be accomplished remotely since they radiate up to 5000 R/h. Within the Vitrification Building during filling, cleaning, and sealing, canisters are moved using standard cranes and trolleys and a specially designed grapple. During transportation to the Glass Waste Storage Building, a one-of-a-kind, specially designed Shielded Canister Transporter (SCT) is used. 8 figs

  2. LHCb: The LHCb off-Site HLT Farm Demonstration

    CERN Multimedia

    Liu, Guoming

    2012-01-01

    The LHCb High Level Trigger (HLT) farm consists of about 1300 nodes, which are housed in the underground server room of the experiment point. Due to the constraints of the power supply and cooling system, it is difficult to install more servers in this room for the future. Off-site computing farm is a solution to enlarge the computing capacity. In this paper, we will demonstrate the LHCb off-site HLT farm which locate in the CERN computing center. Since we use private IP addresses for the HLT farm, we would need virtual private network (VPN) to bridge both sites. There are two kinds of traffic in the event builder: control traffic for the control and monitoring of the farm and the Data Acquisition (DAQ) traffic. We adopt IP tunnel for the control traffic and Network Address Translate (NAT) for the DAQ traffic. The performance of the off-site farm have been tested and compared with the on-site farm. The effect of the network latency has been studied. To employ a large off-site farm, one of the potential bottle...

  3. RELIABILITY ANALYSIS OF THE ELECTRICAL POWER DISTRIBUTION SYSTEM TO SELECTED PORTIONS OF THE NUCLEAR HVAC SYSTEM

    International Nuclear Information System (INIS)

    Ramirez, N.

    2004-01-01

    A design requirement probability of 0.01 or less in a 4-hour period ensures that the nuclear heating, ventilation, and air-conditioning (HVAC) system in the primary confinement areas of the Dry Transfer Facilities (DTFs) and Fuel Handling Facility (FHF) is working during a Category 1 drop event involving commercial spent nuclear fuel (CSNF) assemblies (BSC 2004a , Section 5.1.1.48). This corresponds to an hourly HVAC failure rate of 2.5E-3 per hour or less, which is contributed to by two dominant causes: equipment failure and loss of electrical power. Meeting this minimum threshold ensures that a Category 1 initiating event followed by the failure of HVAC is a Category 2 event sequence. The two causes for the loss of electrical power include the loss of offsite power and the loss of onsite power distribution. Thus, in order to meet the threshold requirement aforementioned, the failure rate of mechanical equipment, loss of offsite power, and loss of onsite power distribution must be less than or equal to 2.5E-3 per hour for the nuclear HVAC system in the primary confinement areas of the DTFs and FHF. The loss of offsite power occurs at a frequency of 1.1E-5 per hour (BSC 2004a, Section 5.1.1.48). The purpose of this analysis is to determine the probability of occurrence of the unavailability of the nuclear HVAC system in the primary confinement areas of the DTFs and FHF due to loss of electrical power. In addition, this analysis provides insights on the contribution to the unavailability of the HVAC system due to equipment failure. The scope of this analysis is limited to finding the frequency of loss of electrical power to the nuclear HVAC system in the primary confinement areas of the DTFs and FHF

  4. Spent nuclear fuel retrieval system fuel handling development testing. Final report

    International Nuclear Information System (INIS)

    Jackson, D.R.; Meeuwsen, P.V.

    1997-09-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin, clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge), remove the contents from the canisters and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. This report describes fuel handling development testing performed from May 1, 1997 through the end of August 1997. Testing during this period was mainly focused on performance of a Schilling Robotic Systems' Conan manipulator used to simulate a custom designed version, labeled Konan, being fabricated for K-Basin deployment. In addition to the manipulator, the camera viewing system, process table layout, and fuel handling processes were evaluated. The Conan test manipulator was installed and fully functional for testing in early 1997. Formal testing began May 1. The purposes of fuel handling development testing were to provide proof of concept and criteria, optimize equipment layout, initialize the process definition, and identify special needs/tools and required design changes to support development of the performance specification. The test program was set up to accomplish these objectives through cold (non-radiological) development testing using simulated and prototype equipment

  5. Offsite environmental monitoring report. Radiation monitoring around United States nuclear test areas, calendar year 1997

    International Nuclear Information System (INIS)

    Davis, M.G.; Flotard, R.D.; Fontana, C.A.; Hennessey, P.A.; Maunu, H.K.; Mouck, T.L.; Mullen, A.A.; Sells, M.D.

    1999-01-01

    This report describes the Offsite Radiological Environmental Monitoring Program (OREMP) conducted during 1997 by the US Environmental Protection Agency's (EPAs), Radiation and Indoor Environments National Laboratory, Las Vegas, Nevada. This laboratory operated an environmental radiation monitoring program in the region surrounding the Nevada Test Site (NTS) and at former test sites in Alaska, Colorado, Mississippi, Nevada, and New Mexico. The surveillance program is designed to measure levels and trends of radioactivity, if present, in the environment surrounding testing areas to ascertain whether current radiation levels and associated doses to the general public are in compliance with existing radiation protection standards. The surveillance program additionally has the responsibility to take action to protect the health and well being of the public in the event of any accidental release of radioactive contaminants. Offsite levels of radiation and radioactivity are assessed by sampling and analyzing milk, water, and air; by deploying and reading thermoluminescent dosimeters (TLDs); and using pressurized ionization chambers (PICs) to measure ambient gamma exposure rates with a sensitivity capable of detecting low level exposures not detected by other monitoring methods

  6. Offsite environmental monitoring report: Radiation monitoring around United States nuclear test areas, calendar year 1997

    Energy Technology Data Exchange (ETDEWEB)

    Davis, M.G.; Flotard, R.D.; Fontana, C.A.; Hennessey, P.A.; Maunu, H.K.; Mouck, T.L.; Mullen, A.A.; Sells, M.D.

    1999-01-01

    This report describes the Offsite Radiological Environmental Monitoring Program (OREMP) conducted during 1997 by the US Environmental Protection Agency`s (EPAs), Radiation and Indoor Environments National Laboratory, Las Vegas, Nevada. This laboratory operated an environmental radiation monitoring program in the region surrounding the Nevada Test Site (NTS) and at former test sites in Alaska, Colorado, Mississippi, Nevada, and New Mexico. The surveillance program is designed to measure levels and trends of radioactivity, if present, in the environment surrounding testing areas to ascertain whether current radiation levels and associated doses to the general public are in compliance with existing radiation protection standards. The surveillance program additionally has the responsibility to take action to protect the health and well being of the public in the event of any accidental release of radioactive contaminants. Offsite levels of radiation and radioactivity are assessed by sampling and analyzing milk, water, and air; by deploying and reading thermoluminescent dosimeters (TLDs); and using pressurized ionization chambers (PICs) to measure ambient gamma exposure rates with a sensitivity capable of detecting low level exposures not detected by other monitoring methods.

  7. Offsite environmental monitoring report: Radiation monitoring around United States nuclear test areas, calendar year 1993

    International Nuclear Information System (INIS)

    Chaloud, D.J; Daigler, D.M.; Davis, M.G.

    1996-06-01

    This report describes the Offsite Radiation Safety Program conducted during 1993 by the Environmental Protection Agency's (EPA's) Environmental Monitoring Systems Laboratory - Las Vegas (EMSL-LV). This laboratory operates an environmental radiation monitoring program in the region surrounding the Nevada Test Site (NTS) and at former test sites in Alaska, Colorado, Mississippi, Nevada, and New Mexico. The surveillance program is designed to measure levels and trends of radioactivity, if present, in the environment surrounding testing areas to ascertain whether current radiation levels and associated doses to the general public are in compliance with existing radiation protection standards. The surveillance program additionally has the responsibility to take action to protect the health and well being of the public in the event of any accidental release of radioactive contaminants. Offsite levels of radiation and radioactivity are assessed by sampling milk, water, and air; by deploying thermoluminescent dosimeters (TLDs) and using pressurized ionization chambers (PICs); by biological monitoring of foodstuffs including animal tissues and food crops; and by measurement of radioactive material deposited in humans

  8. Offsite environmental monitoring report: Radiation monitoring around United States nuclear test areas, calendar year 1993

    Energy Technology Data Exchange (ETDEWEB)

    Chaloud, D.J; Daigler, D.M.; Davis, M.G. [and others

    1996-06-01

    This report describes the Offsite Radiation Safety Program conducted during 1993 by the Environmental Protection Agency`s (EPA`s) Environmental Monitoring Systems Laboratory - Las Vegas (EMSL-LV). This laboratory operates an environmental radiation monitoring program in the region surrounding the Nevada Test Site (NTS) and at former test sites in Alaska, Colorado, Mississippi, Nevada, and New Mexico. The surveillance program is designed to measure levels and trends of radioactivity, if present, in the environment surrounding testing areas to ascertain whether current radiation levels and associated doses to the general public are in compliance with existing radiation protection standards. The surveillance program additionally has the responsibility to take action to protect the health and well being of the public in the event of any accidental release of radioactive contaminants. Offsite levels of radiation and radioactivity are assessed by sampling milk, water, and air; by deploying thermoluminescent dosimeters (TLDs) and using pressurized ionization chambers (PICs); by biological monitoring of foodstuffs including animal tissues and food crops; and by measurement of radioactive material deposited in humans.

  9. Nuclear insurance fire risk

    International Nuclear Information System (INIS)

    Dressler, E.G.

    2001-01-01

    Nuclear facilities operate under the constant risk that radioactive materials could be accidentally released off-site and cause injuries to people or damages to the property of others. Management of this nuclear risk, therefore, is very important to nuclear operators, financial stakeholders and the general public. Operators of these facilities normally retain a portion of this risk and transfer the remainder to others through an insurance mechanism. Since the nuclear loss exposure could be very high, insurers usually assess their risk first-hand by sending insurance engineers to conduct a nuclear insurance inspection. Because a serious fire can greatly increase the probability of an off-site release of radiation, fire safety should be included in the nuclear insurance inspection. This paper reviews essential elements of a facility's fire safety program as a key factor in underwriting nuclear third-party liability insurance. (author)

  10. Offsite demonstrations for MWLID technologies

    International Nuclear Information System (INIS)

    Williams, C.; Gruebel, R.

    1995-01-01

    The goal of the Offsite Demonstration Project for Mixed Waste Landfill Integrated Demonstration (MWLID)-developed environmental site characterization and remediation technologies is to facilitate the transfer, use, and commercialization of these technologies to the public and private sector. The meet this goal, the project identified environmental restoration needs of mixed waste and/or hazardous waste landfill owners (Native American, municipal, DOE, and DoD); documenting potential demonstration sites and the contaminants present at each site; assessing the environmental regulations that would effect demonstration activities; and evaluating site suitability for demonstrating MWLID technologies at the tribal and municipal sites identified. Eighteen landfill sites within a 40.2-km radius of Sandia National Laboratories are listed on the CERCLIS Site/Event Listing for the state of New Mexico. Seventeen are not located within DOE or DoD facilities and are potential offsite MWLID technology demonstration sites. Two of the seventeen CERCLIS sites, one on Native American land and one on municipal land, were evaluated and identified as potential candidates for off-site demonstrations of MWLID-developed technologies. Contaminants potentially present on site include chromium waste, household/commercial hazardous waste, volatile organic compounds, and petroleum products. MWLID characterization technologies applicable to these sites include Magnetometer Towed Array, Cross-borehole Electromagnetic Imaging, SitePlanner trademark/PLUME, Hybrid Directional Drilling, Seamist trademark/Vadose Zone Monitoring, Stripping Analyses, and x-ray Fluorescence Spectroscopy for Heavy Metals

  11. Classification and handling of non-conformance item of nuclear class equipment during manufacture phase

    International Nuclear Information System (INIS)

    Wang Ruiping

    2001-01-01

    Based on inspection experiences in years on nuclear class equipment manufacturing, the author discusses the classification and handling of non-conformance items occurred during equipment manufacturing, and certain technical considerations are presented

  12. Offsite doses from SRP radioactive releases - 1985

    International Nuclear Information System (INIS)

    Marter, W.L.

    1986-01-01

    This memorandum summarizes the offsite doses from releases of radioactive materials to the environment from SRP operations in 1985. These doses were calculated for inclusion in the environmental report for 1985 to be issued by the Health Protection Department (DPSPU-86-30-1). The environmental report is prepared annually for distribution to state environmental agencies, the news media, and interested members of the public. More detailed data on offsite exposures by radionuclide and exposure pathway will be included in the environmental report

  13. Offsite Shipment Campaign Readiness Assessment (OSCRA): A tool for offsite shipment campaigns

    Energy Technology Data Exchange (ETDEWEB)

    Michelhaugh, R.D.; Pope, R.B. [Oak Ridge National Lab., TN (United States); Bisaria, A. [Science Applications International Corp., Oak Ridge, TN (United States)

    1995-12-31

    The Offsite Shipment Campaign Readiness Assessment (OSCRA) tool is designed to assist program managers in identifying, implementing, and verifying applicable transportation and disposal regulatory requirements for specific shipment campaigns. OSCRA addresses these issues and provides the program manager with a tool to support planning for safe and compliant transportation of waste and other regulated materials. Waste transportation and disposal requirements must be identified and addressed in the planning phase of a waste management project. In the past, in some cases, transportation and disposal requirements have not been included in overall project plans. These planning deficiencies have led to substantial delays and cost impacts. Additionally, some transportation regulatory requirements have not been properly implemented, resulting in substantial fines and public embarrassment for the U.S. Department of Energy (DOE). If a material has been processed and packaged for onsite storage (prior to offsite disposal) in a package that does not meet transportation requirements, it must be repackaged in U.S. Department of Transportation (DOT)-compliant packaging for transport. This repackaging can result in additional cost, time, and personnel radiation exposure. The original OSCRA concept was developed during the Pond Waste Project at the K-25 Site in Oak Ridge, Tennessee. The continued development of OSCRA as a user-friendly tool was funded in 1995 by the DOE Office of Environmental Management, Transportation Management Division (TMD). OSCRA is designed to support waste management managers, site remediation managers, and transportation personnel in defining applicable regulatory transportation and disposal requirements for offsite shipment of hazardous waste and other regulated materials. The need for this tool stems from increasing demands imposed on DOE and the need to demonstrate and document safe and compliant packaging and shipment of wastes from various DOE sites.

  14. Development and implementation of automated radioactive materials handling systems

    International Nuclear Information System (INIS)

    Jacoboski, D.L.

    1992-12-01

    Material handling of radioactive and hazardous materials has forced the need to pursue remotely operated and robotic systems in light of operational safety concerns. Manual maneuvering, repackaging, overpacking and inspecting of containers which store radioactive and hazardous materials is the present mode of operation at the Department of Energy (DOE) Fernald Environmental Management Project (FEMP) in Fernald Ohio. The manual methods are unacceptable in the eyes of concerned site workers and influential community oversight committees. As an example to respond to the FEMP material handling needs, design efforts have been initiated to provide a remotely operated system to repackage thousands of degradated drums containing radioactive Thorium: Later, the repackaged Thorium will be shipped offsite to a predesignated repository again requiring remote operation

  15. Risk methodologies for offsite hazardous materials

    International Nuclear Information System (INIS)

    Kot, C.A.; Eichler, T.V.; Wiedermann, A.H.

    1983-01-01

    A number of suggestions have been advanced in recent years concerning the risks posed to nuclear power plants by offsite hazardous materials relative to (1) the regulatory approach including considerations of minimum and safe standoff distances, exclusion distances, site acceptance ceilings and floors, screening distances and screening probabilities, plant design, etc., and (2) the analysis and evaluation procedures such as material screening criteria, plant vulnerability, standarized physical models, etc. An evaluation of current analyses and approaches indicates that this complex problem, variety of approaches, and safety concerns may be better accommodated by developing criteria and treatments along the lines of a so-called conditional risk approach. Specifically, the probability (P) of some ultimate consequence (C) occurring from an accident (A) involving hazardous materials is given as P(C) = P(C/A) x P(A). Assuming that the plant to accident site standoff distance is the fundamental independent variable of the risk methodology, certain conditional risk designations and conditions can be made and are presented

  16. 13 CFR 120.1025 - Off-site reviews and monitoring.

    Science.gov (United States)

    2010-01-01

    ... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Off-site reviews and monitoring. 120.1025 Section 120.1025 Business Credit and Assistance SMALL BUSINESS ADMINISTRATION BUSINESS LOANS Risk-Based Lender Oversight Supervision § 120.1025 Off-site reviews and monitoring. SBA may conduct off...

  17. Integrate offsites management with information systems

    Energy Technology Data Exchange (ETDEWEB)

    Valleur, M. (TECHNIP, Paris (France))

    1993-11-01

    Computerized offsites management systems in oil refineries offer a unique opportunity to integrate advanced technology into a coherent refinery information system that contributes to benefits-driven optimal operations: from long-term, multirefinery linear programming (LP) models to sequential control of transfer lineups in the tank farm. There are strong incentives to automate and optimize the offsites operations, and benefits can be quantified to justify properly sized projects. The paper discusses the following: business opportunities, oil movement and advanced technology, project scoping and sizing, review of functional requirements, transfer automation, blending optimal control, on-line analyzers, oil movement and scheduling, organizational issues, and investment and benefits analysis.

  18. Spent nuclear fuel shipping cask handling capabilities of commercial light water reactors

    International Nuclear Information System (INIS)

    Daling, P.M.; Konzek, G.J.; Lezberg, A.J.; Votaw, E.F.; Collingham, M.I.

    1985-04-01

    This report describes an evaluation of the cask handling capabilities of those reactors which are operating or under construction. A computerized data base that includes cask handling information was developed with information from the literature and utility-supplied data. The capability of each plant to receive and handle existing spent fuel shipping casks was then evaluated. Modal fractions were then calculated based on the results of these evaluations and the quantities of spent fuel projected to be generated by commercial nuclear power plants through 1998. The results indicated that all plants are capable of receiving and handling truck casks. Up to 118 out of 130 reactors (91%) could potentially handle the larger and heavier rail casks if the maximum capability of each facility is utilized. Design and analysis efforts and physical modifications to some plants would be needed to achieve this high rail percentage. These modifications would be needed to satisfy regulatory requirements, increase lifting capabilities, develop rail access, or improve other deficiencies. The remaining 12 reactors were determined to be capable of handling only the smaller truck casks. The percentage of plants that could receive and handle rail casks in the near-term would be reduced to 64%. The primary reason for a plant to be judged incapable of handling rail casks in the near-term was a lack of rail access. The remaining 36% of the plants would be limited to truck shipments. The modal fraction calculations indicated that up to 93% of the spent fuel accumulated by 1998 could be received at federal storage or disposal facilities via rail (based on each plant's maximum capabilities). If the near-term cask handling capabilities are considered, the rail percentage is reduced to 62%

  19. Planning, Coordinating, and Managing Off-Site Storage is an Area of Increasing, Professional Responsibility for Special Collections Departments

    Directory of Open Access Journals (Sweden)

    Melissa Goertzen

    2016-03-01

    two locations instead of one. Also, the integration of new workflows required additional oversight to ensure adequate control at all points of process. Static staffing levels and increased levels of responsibility impacted preservation and conservation activities as well. A central concern was the handling of materials by facility staff not trained as special collections professionals. In regard to the facilities themselves, a general concern was that commercial warehouses do not always provide the kind of environmental control systems recommended for storage of special collections materials. Of the total sample group, 12 participants (19% said their institution does not use off-site storage for special collections. When asked if this may occur in the future, four directors (33% said they anticipate off-site storage use within the next five years. Lack of space was listed as the primary motivation. Conclusion – Study findings provide evidence for what was previously known anecdotally: planning, coordinating, and managing off-site storage is a significant professional responsibility that will only grow in the future. As primary resources are integrated into research, teaching, and learning activities, the acquisition of special collections materials will continue to grow. Discussions regarding off-site storage workflows and strategic planning will continue as professionals seek compromises that meet the unique needs of acquisition, preservation, and public service.

  20. Nuclear liability claims handling and costs - Germany and some comparative solutions

    International Nuclear Information System (INIS)

    Harbruecker, D.

    2000-01-01

    Comparison of legal status in Central Europe: coverage by insurance and State intervention, coverage of legal expenses and interests on awards technical problems of claims handing after a nuclear incident: guidelines to be prepared by insurer before and not after an incident occurred, demands on provider of financial security claims handling for part guaranteed by State to be transferred to insurer, necessary regulations of such arrangements (author)

  1. User's Manual for RESRAD-OFFSITE Version 2.

    Energy Technology Data Exchange (ETDEWEB)

    Yu, C.; Gnanapragasam, E.; Biwer, B. M.; Kamboj, S.; Cheng, J. -J.; Klett, T.; LePoire, D.; Zielen, A. J.; Chen, S. Y.; Williams, W. A.; Wallo, A.; Domotor, S.; Mo, T.; Schwartzman, A.; Environmental Science Division; DOE; NRC

    2007-09-05

    The RESRAD-OFFSITE code is an extension of the RESRAD (onsite) code, which has been widely used for calculating doses and risks from exposure to radioactively contaminated soils. The development of RESRAD-OFFSITE started more than 10 years ago, but new models and methodologies have been developed, tested, and incorporated since then. Some of the new models have been benchmarked against other independently developed (international) models. The databases used have also expanded to include all the radionuclides (more than 830) contained in the International Commission on Radiological Protection (ICRP) 38 database. This manual provides detailed information on the design and application of the RESRAD-OFFSITE code. It describes in detail the new models used in the code, such as the three-dimensional dispersion groundwater flow and radionuclide transport model, the Gaussian plume model for atmospheric dispersion, and the deposition model used to estimate the accumulation of radionuclides in offsite locations and in foods. Potential exposure pathways and exposure scenarios that can be modeled by the RESRAD-OFFSITE code are also discussed. A user's guide is included in Appendix A of this manual. The default parameter values and parameter distributions are presented in Appendix B, along with a discussion on the statistical distributions for probabilistic analysis. A detailed discussion on how to reduce run time, especially when conducting probabilistic (uncertainty) analysis, is presented in Appendix C of this manual.

  2. Standard Guide for Preparing Waste Management Plans for Decommissioning Nuclear Facilities

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide addresses the development of waste management plans for potential waste streams resulting from decommissioning activities at nuclear facilities, including identifying, categorizing, and handling the waste from generation to final disposal. 1.2 This guide is applicable to potential waste streams anticipated from decommissioning activities of nuclear facilities whose operations were governed by the Nuclear Regulatory Commission (NRC) or Agreement State license, under Department of Energy (DOE) Orders, or Department of Defense (DoD) regulations. 1.3 This guide provides a description of the key elements of waste management plans that if followed will successfully allow for the characterization, packaging, transportation, and off-site treatment or disposal, or both, of conventional, hazardous, and radioactive waste streams. 1.4 This guide does not address the on-site treatment, long term storage, or on-site disposal of these potential waste streams. 1.5 This standard does not purport to address ...

  3. Nuclear hydrogen production and its safe handling

    International Nuclear Information System (INIS)

    Chung, Hongsuk; Paek, Seungwoo; Kim, Kwang-Rag; Ahn, Do-Hee; Lee, Minsoo; Chang, Jong Hwa

    2003-01-01

    An overview of the hydrogen related research presently undertaken at the Korea Atomic Energy Research Institute are presented. These encompass nuclear hydrogen production, hydrogen storage, and the safe handling of hydrogen, High temperature gas-cooled reactors can play a significant role, with respect to large-scale hydrogen production, if used as the provider of high temperature heat in fossil fuel conversion or thermochemical cycles. A variety of potential hydrogen production methods for high temperature gas-cooled reactors were analyzed. They are steam reforming of natural gas, thermochemical cycles, etc. The produced hydrogen should be stored safely. Titanium metal was tested primarily because its hydride has very low dissociation pressures at normal storage temperatures and a high capacity for hydrogen, it is easy to prepare and is non-reactive with air in the expected storage conditions. There could be a number of potential sources of hydrogen evolution risk in a nuclear hydrogen production facility. In order to reduce the deflagration detonation it is necessary to develop hydrogen control methods that are capable of dealing with the hydrogen release rate. A series of experiments were conducted to assess the catalytic recombination characteristics of hydrogen in an air stream using palladium catalysts. (author)

  4. A fiber optic link for the remote handling in nuclear environment

    International Nuclear Information System (INIS)

    Breuze, G.; Carnet, B.; Friant, A.; Blanc, F.; Lordet, J.; Boisde, G.

    1988-01-01

    At CEA a R/D program is running to improve performances of servomanipulators used in nuclear fuel reprocessing plants. Present work gives the main environmental parameters (gamma rays exposition, temperature) and shows the basis of the digital link designed to remote-handle such a manipulator. Up to 10 5 Gy behavior of optical fibers and electronic components was studied. Two different optical cables were built, one for the long link (100 m), the second to set in an especially designed winding unwinding wheel. Six way permanent or remote-handle connectors were developed to connect optical interfaces and a leaktight penetration. Measured budget of the link taking into account efficient photoblesching of the pure silica core fiber and influence of gamma rays on the slave interface is presented [fr

  5. Off-line programming and simulation in handling nuclear components

    International Nuclear Information System (INIS)

    Baker, C.P.

    1993-10-01

    IGRIP was used to create a simulation of the robotic workcell design for handling components at the PANTEX nuclear arms facility. This initial simulation identified problems with the customer's proposed worker layout, and allowed a correction to be proposed. Refinement of the IGRIP simulation allowed the design and construction of a workcell mock-up and accurate off-line programming of the system. IGRIP's off-line programming capabilities are being used to develop the motion control code for the workcell. PNLs success in this area suggests that simulation and off-line programming may be valuable tools for developing robotics in some automation resistant industries

  6. Assessment of RELAP5/MOD2 code using loss of offsite power transient data of KNU [Korea Nuclear Unit] No. 1 Plant

    International Nuclear Information System (INIS)

    Chung, Bud-Dong; Kim, Hho-Jung

    1990-04-01

    This report presents a code assessment study based on a real plant transient that occurred on June 9, 1981 at the KNU number-sign 1 (Korea Nuclear Unit Number 1). KNU number-sign 1 is a two-loop Westinghouse PWR plant of 587 Mwe. The loss of offsite power transient occurred at the 77.5% reactor power with 0.5%/hr power ramp. The real plant data were collected from available on-line plant records and computer diagnostics. The transient was simulated by RELAP5/MOD2/36.05 and the results were compared with the plant data to assess the code weaknesses and strengths. Some nodalization studies were performed to contribute to developing a guideline for PWR nodalization for the transient analysis. 5 refs., 18 figs., 3 tabs

  7. On-site and off-site forensic analysis capabilities for proliferation and terrorism prevention

    International Nuclear Information System (INIS)

    Whipple, R.E.; Nunes, P.J.; Reynolds, J.G.; Alcaraz, A.; Hart, B.R.

    2004-01-01

    Full text: We will present current on-site technologies that can be utilized for the screening of explosives, chemical agents, and environmental contaminants. These techniques must have the capability to detect various hazardous materials at very low levels, since they pose a major challenge for first responders. Specifically, the technology must detect concealed explosives or chemical agents on-site rapidly. Lawrence Livermore National Laboratory (LLNL) is currently developing several different high explosive screening and detection technologies for field use. Two technologies that have demonstrated an ability to screen for explosives at low levels are colorimetric spot tests and thin layer chromatography (TLC). Another technology that has demonstrated usefulness for the on-site analysis of unknowns is portable gas chromatography-mass spectrometry (GC-MS) with solid phase micro-extraction (SPME) sampling. Several examples utilizing these technologies and their usefulness will be presented. In addition to developing on-site screening methods, LLNL is an Organization for the Prohibition of Chemical Weapons (OPCW) designated laboratory and is certified to accept CW suspect samples. Currently, LLNL is expanding its ISO-17025 certification to include nuclear forensics and explosives. These off-site forensic analysis capabilities and certified procedures will support the needs for homeland security. We will highlight some of the ISO-17025 requirements to accredit procedures, handle samples, and reports. Work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under contract W-7405-ENG-48

  8. On-site and off-site forensic analysis capabilities for proliferation and terrorism prevention

    International Nuclear Information System (INIS)

    Hart, B.R.; Whipple, R.E.; Nunes, P.J.; Reynolds, J.G.; Alcaraz, A.

    2004-01-01

    We present current on-site technologies that can be utilized for the screening of explosives, chemical agents, and environmental contaminants. These techniques must have the capability to detect various hazardous materials at very low levels, since they pose a major challenge for first responders. Specifically, the technology must detect concealed explosives or chemical agents on-site rapidly. Lawrence Livermore National Laboratory (LLNL) is currently developing several different high explosive screening and detection technologies for field use. Two technologies that have demonstrated an ability to screen for explosives at low levels are colorimetric spot tests and thin layer chromatography (TLC). Another technology that has demonstrated usefulness for the on-site analysis of unknowns is portable gas chromatography-mass spectrometry (GC-MS) with solid phase micro-extraction (SPME) sampling. Several examples utilizing these technologies and their usefulness will be presented. In addition to developing on-site screening methods, LLNL is an Organization for the Prohibition of Chemical Weapons (OPCW) designated laboratory and is certified to accept CW suspect samples. Currently, LLNL is expanding its ISO-17025 certification to include nuclear forensics and explosives. These off-site forensic analysis capabilities and certified procedures will support the needs for homeland security. We will highlight some of the ISO-17025 requirements to accredit procedures, handle samples, and reports. Work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under contract W-7405-ENG-48. (author)

  9. Development of remote handling technology for nuclear fuel cycle facilities in Japan

    International Nuclear Information System (INIS)

    Maekawa, Hiromichi; Sakai, Akira; Miura, Noriaki; Kozaka, Tetsuo; Hamada, Takashi

    2015-01-01

    Remote handling technology has been systematically developed for nuclear fuel cycle facilities in Japan since 1970s, primarily in parallel with the development of reprocessing and HLLW (High Level Liquid Waste) vitrification process. In case of reprocessing and vitrification process to handle highly radioactive and hazardous materials, the most of components are installed in the radiation shielded hot cells and operators are not allowed to enter the work area in the cells for operation and maintenance. Therefore, a completely remote handling system is adopted for the cells to reduce radiation doses of operators and increase the availability of the facility. The hot cells are generally designed considering the scale of components (laboratory, demonstration, or full-scale), the function of the systems (chemical process, material handling, dismantling, decontamination, or chemical analysis), and the environmental conditions (radiation dose rate, airborne concentration, surface contamination, or fume/mist/dust). Throughout our domestic development work for remote handling technology, the concept of the large scale integrated cell has been adopted rather than a number of small scale separated cells, for the reasons to reduce the total installation space and the number of remote handling equipment required for the each cell as much as possible. In our domestic remote maintenance design, several new concepts have been developed, tested, and demonstrated in the Tokai Virtrification Facility (TVF) and the Rokkasho HLLW Vitrification and Storage Facility (K-facility). Layout in the hot cells, the performance of remote handling equipment, and the structure of the in-cell components are important factors for remote maintenance design. In case of TVF (hot tests started in 1995), piping and vessels are prefabricated in the rack modules and installed in two lines on both sides of the cell. These modules are designed to be remotely replaced in the whole rack. Two overhead cranes

  10. How problems of storing waste nuclear fuel are handled in some countries

    International Nuclear Information System (INIS)

    Langhe, R.

    1983-01-01

    This report is a survey of the situation in a number of European countries, in the United States and the Soviet Union as well. In all democratic countries, the nuclear power issue is controversial. Everywhere it has met with opposition and criticism, even in countries where nuclear power is officially promoted. Which of the elements comprised in the nuclear power issue is regarded as most controversial varies from country to country. In some countries, final storage and handling of waste nuclear fuel are referred to this category, in others nuclear power plant safety is claimed to be of greater importance. In the last few months, some public opinion has been coupling the peaceful use of nuclear power with nuclear weapons, thereby deeming the greatest danger to be the risk of unwanted distribution of nuclear weapons. Technical difficulties as well as public opinion have indefinitely adjourned the final solution of the disposal of waste nuclear fuel. This problem is of such magnitude that a final solution is urgently needed. Apart from opinions, the existence of waste nuclear power fuel emitting dangerous radiation for over 40 generations to come, makes it a moral obligation to find a way to spare future generations that heritage. (author)

  11. 40 CFR 1400.8 - Access to off-site consequence analysis information by Federal government officials.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Access to off-site consequence... MANAGEMENT PROGRAMS UNDER THE CLEAN AIR ACT SECTION 112(r)(7); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Access to Off-Site Consequence Analysis...

  12. Examination of offsite emergency protective measures for core melt accidents

    International Nuclear Information System (INIS)

    Aldrich, D.C.; McGrath, P.E.; Ericson, D.M. Jr.; Jones, R.B.; Rasmussen, N.C.

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to potential nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each protective measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment falure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects

  13. Preparation for commissioning of nuclear plant with reference to British Nuclear Fuels Plc fuel handling plant project

    International Nuclear Information System (INIS)

    Bamber, D.R.

    1987-01-01

    The new Fuel Handling Plant at British Nuclear Fuels Sellafield is part of a Pound 550M complex which provides facilities for the receipt, storage and mechanical preparation of both Magnox and A.G.R. fuel. The plant is very large and complex with considerable use of computer based process control systems, providing for physical and nuclear safety. The preparation of such plant for active commissioning necessitates a great many physical checks and technical evaluations in support of its safety case. This paper describes arrangements for plant commissioning checks, against the regulatory framework and explains the physical preparations necessary for their timely accomplishment. (author)

  14. Review of off-site emergency preparedness and response plan of Indian NPPs based on experience of Fukushima nuclear accident

    International Nuclear Information System (INIS)

    Singh, Hukum; Dash, M.; Shukla, Vikas; Vijayan, P.; Krishnamurthy, P.R.

    2012-01-01

    Nuclear power plants in India are designed, constructed and operated based on the principle of the highest priority to nuclear safety. To deal with any unlikely situation of radiological emergency, the emergency preparedness and response plans are ensured to be in place at all NPPs prior to their commissioning. These plans are periodically reviewed and tested by conducting emergency exercise with the participation of various agencies such as Nuclear Power Corporation of India Limited, NDMA, district authorities, regulatory body and general public. On March 11, 2011 an earthquake of magnitude 9.0 hit the Fukushima Dai-ichi and Dai-ni followed by tsunami waves of height 15 meters above reference sea level. This resulted in large scale release of radioactive material from Fukushima Dai-ichi NPS. This led to the evacuation of a large number of people from the areas surrounding the affected nuclear power plants. The event was rated as level 7 event in International Nuclear Event Scale (INES). The event also revealed the challenges in handling radiological emergency situation in adverse environmental conditions, The experience of managing radiological emergency situation during Fukushima nuclear accident provides opportunities to review and improve emergency preparedness and response programme. The present paper presents the chronology of the emergency situation, challenges faced and handled in Fukushima. Even though the possibility of a Fukushima type nuclear accident in India is very remote due to the low probability of a high intensity earthquake followed by tsunami at NPP sites, the efforts needs to be initiated from the regulatory point of view for an effective Nuclear and Radiological Emergency Preparedness and Response Plans. The Emergency Preparedness and Response Plans of NPP sites were reviewed in the light of unique challenges of accident at Fukushima. It is realized that multi unit events are the realities that must be addressed as part of Emergency

  15. Review of off-site emergency preparedness and response plan of Indian NPPs based on experience of Fukushima nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Singh, Hukum; Dash, M.; Shukla, Vikas; Vijayan, P.; Krishnamurthy, P.R., E-mail: vshukla@aerb.gov.in [Operating Plants Safety Division, Atomic Energy Regulatory Board, Mumbai (India)

    2012-07-01

    Nuclear power plants in India are designed, constructed and operated based on the principle of the highest priority to nuclear safety. To deal with any unlikely situation of radiological emergency, the emergency preparedness and response plans are ensured to be in place at all NPPs prior to their commissioning. These plans are periodically reviewed and tested by conducting emergency exercise with the participation of various agencies such as Nuclear Power Corporation of India Limited, NDMA, district authorities, regulatory body and general public. On March 11, 2011 an earthquake of magnitude 9.0 hit the Fukushima Dai-ichi and Dai-ni followed by tsunami waves of height 15 meters above reference sea level. This resulted in large scale release of radioactive material from Fukushima Dai-ichi NPS. This led to the evacuation of a large number of people from the areas surrounding the affected nuclear power plants. The event was rated as level 7 event in International Nuclear Event Scale (INES). The event also revealed the challenges in handling radiological emergency situation in adverse environmental conditions, The experience of managing radiological emergency situation during Fukushima nuclear accident provides opportunities to review and improve emergency preparedness and response programme. The present paper presents the chronology of the emergency situation, challenges faced and handled in Fukushima. Even though the possibility of a Fukushima type nuclear accident in India is very remote due to the low probability of a high intensity earthquake followed by tsunami at NPP sites, the efforts needs to be initiated from the regulatory point of view for an effective Nuclear and Radiological Emergency Preparedness and Response Plans. The Emergency Preparedness and Response Plans of NPP sites were reviewed in the light of unique challenges of accident at Fukushima. It is realized that multi unit events are the realities that must be addressed as part of Emergency

  16. Pathways for Off-site Corporate PV Procurement

    Energy Technology Data Exchange (ETDEWEB)

    Heeter, Jenny S [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

    2017-09-06

    Through July 2017, corporate customers contracted for more than 2,300 MW of utility-scale solar. This paper examines the benefits, challenges, and outlooks for large-scale off-site solar purchasing through four pathways: power purchase agreements, retail choice, utility partnerships (green tariffs and bilateral contracts with utilities), and by becoming a licensed wholesale seller of electricity. Each pathway differs based on where in the United States it is available, the value provided to a corporate off-taker, and the ease of implementation. The paper concludes with a discussion of future pathway comparison, noting that to deploy more corporate off-site solar, new procurement pathways are needed.

  17. Safe handling and monitoring of tritium in research on nuclear fusion

    International Nuclear Information System (INIS)

    Yoshida, Yoshikazu; Naruse, Yuji

    1978-01-01

    An actual condition of technique of safety handling and monitoring of tritium in the laboratory which treated a great quantity of tritium in relation to nuclear fusion, was described. With respect to the technique of safety handling of tritium, an actual condition of the technique in the U.S.A. which had wide experience in treating a great quantity of 3 H was mainly introduced, and it was helpful to a safety measure and a reduction of tritium effluence. Glovebox, hood, and other component machinery and tools for treating 3 H were also introduced briefly. As a monitoring technique, monitoring of indoor air and air exhaust by ionization chamber-type monitor for continuous monitoring of a great quantity of gaseous tritium was mentioned. Next, monitoring of a room, the surfaces of equipments, and draining, internal exposure of the individual, and monitoring of the environment were introduced. (Kanao, N.)

  18. Management of radioactive waste in nuclear power: handling of irradiated graphite from water-cooled graphite reactors

    International Nuclear Information System (INIS)

    Anfimov, S.S.

    2001-01-01

    In this paper an radioactive waste processing of graphite from graphite moderated nuclear reactors at its decommissioning is discussed. Methods of processing of irradiated graphite are presented. It can be concluded that advanced methods for graphite radioactive waste handling are available nowadays. Implementation of these methods will allow to enhance environmental safety of nuclear power that will benefit its progress in the future

  19. Illegal handling of radioactive and nuclear materials. Threats and suggestions for measures

    International Nuclear Information System (INIS)

    Oliver, Lena; Melin, Lena; Prawitz, Jan; Ringbom, Anders; Sandstroem, Bjoern; Wigg, Lars; Wirstam, Jens

    2004-01-01

    This project deals with threats from smuggling or other illegal transportation of radioactive or nuclear materials across the borders to Sweden, and with the security of handling such materials in Sweden. The project has included studies of relevant documentation; visits and interviews at industries, hospitals, research institutes and military institutions in Sweden that handle radioactive materials; a pilot study at the Stockholm freeport, where equipment for detection of radioactive materials has been tested for six months; an analysis of incidents reported to the IAEA database; and an analysis of Swedish incidents. The following conclusions are drawn: Stricter rules regarding the physical protection of radiation sources and radioactive materials should be implemented in Sweden. The recommendations recently issued by IAEA should serve as a point of departure for working out such rules

  20. Claims Handling Co-operation between Nuclear Insurance Pools in a Case of Transboundary Damage - Multilateral and Bilateral Agreements in Progress

    International Nuclear Information System (INIS)

    Zaruba, P.

    2008-01-01

    The paper is a short progress report on matters concerning the core reason for insurance of nuclear third party liability - registration, handling, organizing and settling of claims in case of a major nuclear incident, underlining claims handling co-operation between national nuclear insurance pools when damage to health or property becomes international. The contents of this paper is in close relation to information provided on this subject during the 6th International Conference in 2006. Commercial insurance companies have gained extensive experience with handling large scale claims (e.g. after floods and other natural disasters) and are capable in gathering and organizing a high number of professional loss surveyors and adjusters in a very short period of time. In case of nuclear insurance pools co-operation between members (commercial insurance companies) is an added value and can be used practically all over the country bringing into action the network of branches and offices of all the pool members. This advantage is also used in case of cross border claims when it is necessary to gather information and claims advises from a large number of subjects and from many countries, sometimes very far apart. The international network of nuclear insurance pools is an ideal tool for this task and can be mobilized practically at once. Operators of nuclear installations, especially nuclear power plants, do not have the possibility to put aside hundreds of workers to handle claims and are also usually not sufficiently equipped with the necessary know-how. The same goes for governments and government agencies which in many countries guarantee the payments of claims to victims. National nuclear insurance pools are on the other hand well equipped for this task which usually has to be in place for many years after a nuclear incident. Multilateral and bilateral agreements between national nuclear insurance pools and other institutions should be prepared and signed before any

  1. MHSS: a material handling system simulator

    Energy Technology Data Exchange (ETDEWEB)

    Pomernacki, L.; Hollstien, R.B.

    1976-04-07

    A Material Handling System Simulator (MHSS) program is described that provides specialized functional blocks for modeling and simulation of nuclear material handling systems. Models of nuclear fuel fabrication plants may be built using functional blocks that simulate material receiving, storage, transport, inventory, processing, and shipping operations as well as the control and reporting tasks of operators or on-line computers. Blocks are also provided that allow the user to observe and gather statistical information on the dynamic behavior of simulated plants over single or replicated runs. Although it is currently being developed for the nuclear materials handling application, MHSS can be adapted to other industries in which material accountability is important. In this paper, emphasis is on the simulation methodology of the MHSS program with application to the nuclear material safeguards problem. (auth)

  2. Ocular organ dose assessment of nuclear medicine workers handling diagnostic radionuclides

    International Nuclear Information System (INIS)

    Cho, Yong In; Kim, Ja Mee; Kim, Jung Hoon

    2017-01-01

    The dose distribution in the ocular organs of nuclear medicine workers during the handling of diagnostic radionuclides was assessed via simulation in virtual space. The cornea and lenses received the highest dose, and the dose distribution tended to be proportional to the gamma-ray energy emitted from the radiation source being handled. Moreover, calculations on the dose-reduction effects of eye-wear protectors for the eyes of the workers showed that the effects were inversely proportional to the emitted gamma-ray energy, with the dose-reduction effect decreasing in the order of "2"0"1Tl, "1"2"3I, "9"9mTc, "6"7Ga, "1"1"1In and "1"8F. Among the considered sources, the dose-reduction effect was significant for sources that emit relatively less energy, namely "1"2"3I, "2"0"1Tl and "9"9mTc, while it was lower for the remaining sources, namely "1"8F, "1"1"1In and "6"7Ga. (authors)

  3. Impact of Uncertainty on Calculations for Recovery from Loss of Offsite Power

    International Nuclear Information System (INIS)

    Kelly, Dana L.

    2010-01-01

    Uncertainty, both aleatory and epistemic, can have a significant impact on estimated probabilities of recovering from loss of offsite power within a specified time window, and such probabilities are an input to risk-informed decisions as to the significance of inspection findings in the U.S. Nuclear Regulatory Commission's Reactor Oversight Process. In particular, the choice of aleatory model for offsite power recovery time can have a significant impact on the estimated nonrecovery probability, especially if epistemic uncertainty regarding parameters in the aleatory model is accounted for properly. In past and current analyses, such uncertainty has largely been ignored. This paper examines the impact of both aleatory and epistemic uncertainty on the results, using modern open-source Bayesian inference software, which implements Markov chain Monte Carlo sampling. It includes examples of time-dependent convolution calculations to show the impact that uncertainty can have on this increasingly frequent type of calculation, also. The results show that the 'point estimate' result, which is an input to risk-informed decisions, can easily be uncertain by a factor of 10 if both aleatory and epistemic uncertainties are considered. The paper also illustrates the use of Bayesian model selection criteria to aid in the choice of aleatory model.

  4. System design for safe robotic handling of nuclear materials

    International Nuclear Information System (INIS)

    Drotning, W.; Wapman, W.; Fahrenholtz, J.; Kimberly, H.; Kuhlmann, J.

    1996-01-01

    Robotic systems are being developed by the Intelligent Systems and Robotics Center at Sandia National Laboratories to perform automated handling tasks with radioactive nuclear materials. These systems will reduce the occupational radiation exposure to workers by automating operations which are currently performed manually. Because the robotic systems will handle material that is both hazardous and valuable, the safety of the operations is of utmost importance; assurance must be given that personnel will not be harmed and that the materials and environment will be protected. These safety requirements are met by designing safety features into the system using a layered approach. Several levels of mechanical, electrical and software safety prevent unsafe conditions from generating a hazard, and bring the system to a safe state should an unexpected situation arise. The system safety features include the use of industrial robot standards, commercial robot systems, commercial and custom tooling, mechanical safety interlocks, advanced sensor systems, control and configuration checks, and redundant control schemes. The effectiveness of the safety features in satisfying the safety requirements is verified using a Failure Modes and Effects Analysis. This technique can point out areas of weakness in the safety design as well as areas where unnecessary redundancy may reduce the system reliability

  5. Handling and disposal of SP-100 ground test nuclear fuel and equipment

    International Nuclear Information System (INIS)

    Wilson, C.E.; Potter, J.D.; Hodgson, R.D.

    1990-05-01

    The post SP-100 reactor testing period will focus on defueling the reactor, packaging the various radioactive waste forms, and shipping this material to the appropriate locations. Remote-handling techniques will be developed to defuel the reactor. Packaging the spent fuel and activated reactor components is a challenge in itself. This paper presents an overview of the strategy, methods, and equipment that will be used during the closeout phase of nuclear testing

  6. Handling and disposal of SP-100 ground test nuclear fuel and equipment

    International Nuclear Information System (INIS)

    Wilson, C.E.; Potter, J.D.; Hodgson, R.D.

    1991-01-01

    The post SP-100 reactor testing period will focus on defueling the reactor, packaging the various radiactive waste forms, and shipping this material to the appropriate locations. Remote-handling techniques will be developed to defuel the reactor. Packaging the spent fuel and activated reactor components is a challenge in itself. This paper presents an overview of the strategy, methods, and equipment that will be used during the closeout phase of nuclear testing

  7. 40 CFR 1400.9 - Access to off-site consequence analysis information by State and local government officials.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Access to off-site consequence... CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Access to Off-Site Consequence Analysis Information by Government Officials. § 1400.9 Access to off-site consequence analysis...

  8. Incorporation of safety interlocks in commercial robotics for handling of nuclear materials

    International Nuclear Information System (INIS)

    Moore, F.W.

    1986-01-01

    Current robotic systems have been developed primarily for the automotive and electronic industry. The adaptation of these commercial robotic systems to applications in the manufacturing of nuclear fuel requires the addition of safety interlocks as to the handling and accountability of nuclear materials. Also, additional safety interlocks are required when the robots are operated in containment enclosures that are environmentally sealed. Interlocks have been incorporated into a commercial robot. The robotic system has been installed in the containment enclosure as part of the pellet storage subsystem into the Secure Automated Fabrication (SAF) facility currently being built by Westinghouse Hanford Company (WHC) for the US Department of Energy (DOE). The system has been installed in the Fuel Cycle Plant and is scheduled for initial operational testing in 1986

  9. Incorporation of safety interlocks in commercial robotics for handling of nuclear materials

    International Nuclear Information System (INIS)

    Moore, F.W.

    1986-01-01

    The adaptation of commercial robotic systems to applications in the manufacturing of nuclear fuel has required the addition of safety interlocks as to the handling and accountability of nuclear materials. Also, additional safety interlocks are required when the robots are operated in containment enclosures that are environmentally sealed. Interlocks have been incorporated in a commercial robot which was modified and with additional interlocks into the existing robotic control system. The robotic system has been installed in the containment enclosure as part of the pellet storage subsystem in the Secure Automated Fabrication facility currently being built by Westinghouse Hanford Company for the US Department of Energy. The system has been installed in the Fuel Cycle Plant and is scheduled for initial operational testing in 1986

  10. Critical examination of emergency plans for nuclear accidents

    International Nuclear Information System (INIS)

    Catsaros, Nicolas.

    1986-08-01

    An analysis of emergency plans of various countries for nuclear installations on- and off-site emergency preparedness is presented. The analysis is focused on the off-site organization and countermeasures to protect public health and safety. A critical examination of the different approaches is performed and recommendations for effectiveness improvement and optimization are formulated. (author)

  11. 29 CFR 1908.4 - Offsite consultation.

    Science.gov (United States)

    2010-07-01

    ... on occupational safety and health issues by telephone and correspondence, and at locations other than... 29 Labor 5 2010-07-01 2010-07-01 false Offsite consultation. 1908.4 Section 1908.4 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR...

  12. Millstone nuclear power plant emergency system assessment

    International Nuclear Information System (INIS)

    Akhmad Khusyairi

    2011-01-01

    U.S.NRC determined an obligation to build a nuclear power plant emergency response organization for both on-site and off-site. Millstone Nuclear Power Plants have 3 nuclear reactors and 2 of 3 still in commercial operation. Reactor unit 1, BWR type has been permanently shut down in 1998, while the two others, units 2 and 3 obtain the extended operating license respectively until 2035 and 2045. As a nuclear installation has the high potential radiological impact, Millstone nuclear power plant emergency response organization must establish both on-site or off-site. Emergency response organization that is formed must involve several state agencies, both state agencies and municipality. They have specific duties and functions in a state of emergency, so that protective measures can be undertaken in accordance with the community that has been planned. Meanwhile, NRC conduct their own independent assessment of nuclear power plant emergencies. (author)

  13. Getting to grips with remote handling and robotics

    Energy Technology Data Exchange (ETDEWEB)

    Mosey, D [Ontario Hydro, Toronto (Canada)

    1984-12-01

    A report on the Canadian Nuclear Society Conference on robotics and remote handling in the nuclear industry, September 1984. Remote handling in reactor operations, particularly in the Candu reactors is discussed, and the costs and benefits of use of remote handling equipment are considered. Steam generator inspection and repair is an area in which practical application of robotic technology has made a major advance.

  14. Offsite testing in support of the Salt Repository Project

    International Nuclear Information System (INIS)

    Kalia, H.N.

    1987-04-01

    This report presents a rationale and recommendation to perform an offsite testing program in support of the Salt Repository Project. The investigation to be performed primarily consists of qualifying test methods and procedures, qualifying personnel-training procedures, evaluating test instruments and selected equipment, and obtaining mining and production equipment performance-related information. The key objective of these activities is to develop capabilities to be used at the exploratory shaft facility (ESF). The ESF is to be excavated at the Deaf Smith County site to characterize the salt site for the construction of a repository used to isolate radioactive waste from the biosphere. The bulk of the offsite testing work will be performed at Avery Island Salt Mine at New Iberia, Lousiana. Additional knowledge will be obtained by exchanging technical information either as participants or as observers at the Waste Isolation Pilot Plant (WIPP) site and the Asse Mine in the Federal Republic of Germany (FRG). It is estimated that the offsite testing program will cost approximately $9.3 million over 4 fiscal years. 14 refs., 1 fig., 8 tabs

  15. Study on the operational guides of the off-site emergency management center

    International Nuclear Information System (INIS)

    Park, Won Jong; Han, S. J.; Oh, K. H.

    2005-01-01

    The emergency response organizational groups and roles of Off-site Emergency Management Center was proposed to respond in case of radiological emergency. Development of implementing procedures of Off-site Emergency Management Center in case of radiological emergency to improve effective co-operation and rapid response in radiological emergency. Establishment of 'The Ordinance of Operation of residence radiological emergency office of the Minister of Science and Technology' and announced by the Minister of Science and Technology. The Implementing procedures of Off-site Emergency Management Center and 'The Ordinance of Operation of residence radiological emergency office of the Minister of Science and Technology' can be provide guidelines in case of emergency

  16. Interim report spent nuclear fuel retrieval system fuel handling development testing

    Energy Technology Data Exchange (ETDEWEB)

    Ketner, G.L.; Meeuwsen, P.V.; Potter, J.D.; Smalley, J.T.; Baker, C.P.; Jaquish, W.R.

    1997-06-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project at the Hanford Site. The project will retrieve spent nuclear fuel, clean and remove fuel from canisters, repackage fuel into baskets, and load fuel into a multi-canister overpack (MCO) for vacuum drying and interim dry storage. The FRS is required to retrieve basin fuel canisters, clean fuel elements sufficiently of uranium corrosion products (or sludge), empty fuel from canisters, sort debris and scrap from whole elements, and repackage fuel in baskets in preparation for MCO loading. The purpose of fuel handling development testing was to examine the systems ability to accomplish mission activities, optimization of equipment layouts for initial process definition, identification of special needs/tools, verification of required design changes to support performance specification development, and validation of estimated activity times/throughput. The test program was set up to accomplish this purpose through cold development testing using simulated and prototype equipment; cold demonstration testing using vendor expertise and systems; and graphical computer modeling to confirm feasibility and throughput. To test the fuel handling process, a test mockup that represented the process table was fabricated and installed. The test mockup included a Schilling HV series manipulator that was prototypic of the Schilling Hydra manipulator. The process table mockup included the tipping station, sorting area, disassembly and inspection zones, fuel staging areas, and basket loading stations. The test results clearly indicate that the Schilling Hydra arm cannot effectively perform the fuel handling tasks required unless it is attached to some device that can impart vertical translation, azimuth rotation, and X-Y translation. Other test results indicate the importance of camera locations and capabilities, and of the jaw and end effector tool design. 5 refs., 35 figs., 3 tabs.

  17. Emergency planning and preparedness for a nuclear accident

    International Nuclear Information System (INIS)

    Rahe, E.P.

    1985-01-01

    Based on current regulations, FEMA approves each site-specific plan of state and local governments for each power reactor site after 1) formal review offsite preparedness, 2) holding a public meeting at which the preparedness status has been reviewed, and 3) a satisfactory joint exercise has been conducted with both utility and local participation. Annually, each state, within any position of the 10-mile emergency planning zone, must conduct a joint exercise with the utility to demonstrate its preparedness for a nuclear accident. While it is unlikely that these extreme measures will be needed as a result of an accident at a nuclear power station, the fact that these plans have been well thought out and implemented have already proven their benefit to society. The preparedness for a nuclear accident can be of great advantage in other types of emergencies. For example, on December 11, 1982, a non-nuclear chemical storage tank exploded at a Union Carbide plant in Louisiana shortly after midnight. More than 20,000 people were evacuated from their homes. They were evacuated under the emergency response plan formulated for use in the event of a nuclear accident at the nearby Waterford Nuclear plants. Clearly, this illustrates how a plan conceived for one purpose is appropriate to handle other types of accidents that occur in a modern industrial society

  18. The RODOS system: decision support for nuclear off-site emergency management in Europe

    International Nuclear Information System (INIS)

    Raskob, W.; Ehrhardt, J.

    2000-01-01

    The integrated and comprehensive real-time on-line decision support system, RODOS, for off-site emergency management of nuclear accidents has been developed with support of the European Commission and the German Ministry of Environment. About 40 West and East European institutes have been involved in the development of the existing version for (pre-) operational use. This paper gives an overview of the structure, the content, the main functions and the development status of the RODOS system. It describes how the system has been and is being installed in emergency centres of a number of European counties. Designed as a generic tool, the RODOS system is applicable from the very early stages of an accident up to many year after the release and from the vicinity of a site to far distant areas, unperturbed by national boundaries. Decision support is provided by the system at various levels, ranging from the largely descriptive with information on the present and future radiological situation, to an evaluation of the benefits and disadvantages of different countermeasures' options and their feasibility. This includes ranking them according to the decision-makers' expressed preferences and weights with due consideration of subjective arguments on socio-psychological and political influences. The capability of the RODOS software framework for integrating models, methods and database in a modular way and the flexibility of the user interface will be addressed in the paper. Their functionalities offer the possibility of adapting RODOS to local, regional and national conditions, in particular to the corresponding meteorological and radiological monitoring networks, the geographical and economic structures, different plant types and accident conditions. A hierarchy of user interfaces allows adaptation of the system to the needs and qualifications of users in real emergencies and in training and exercises. The potential role of RODOS for improving emergency response in Europe

  19. Domestic round robin exercise on analysis of uranium for nuclear material handling facilities in Japan

    International Nuclear Information System (INIS)

    Kato, Yoshiyasu; Nagai, Kohta; Handa, Takamitsu; Inoue, Shin-ichi; Sato, Yoshihiro

    2016-01-01

    Interlaboratory comparison programme as well as internal quality control system is an effective tool for an analytical laboratory responsible to nuclear material accountancy of a nuclear facility to maintain and enhance its capability for analysis. However, it is a burden on nuclear material handling facilities in Japan to attend interlaboratory comparison programme run by overseas institutions because of high costs and complicated procedure for importing nuclear materials, and therefore facilities which can participate in such international programme would be limited. Nuclear Material Control Center has hence started and organised an annual domestic round robin exercise on analysis of uranium standard materials, funded by the Japan Safeguards Office of the Nuclear Regulation Authority, since 2008 to enhance analytical capability of Japanese Facilities. The outline of the round robin exercise will be given and the results of uranium isotopic and concentration analysis reported by participant facilities from 2008 to 2015 will be summarised in the presentation. (author)

  20. Conditioning and handling of tritiated wastes at Canadian nuclear power facilities

    International Nuclear Information System (INIS)

    Krochmalnek, L.S.; Krasznai, J.P.; Carney, M.

    1987-04-01

    Ontario Hydro operates a 10,000 MW capacity nuclear power system utilizing the CANDU pressurized heavy water reactor design. The use of D 2 O as moderator and coolant results in the production of about 2400 Ci of tritium per MWe-yr. As a result, there is significant Canadian experience in the treatment, handling, transport and storage of tritiated wastes. Ontario Hydro operates its own reactor waste storage site which includes systems for volume reduction, immobilization and packaging of wastes. In addition, a facility to remove tritium from heavy water is presently being commissioned at the Darlington nuclear site. This facility will generate tritiated liquid and solid waste that will have to be properly conditioned prior to storage or disposal. The nature of these various wastes and the processes/packaging required to meet storage/disposal criteria are judged to have relevance to investigations in fusion facility waste arisings. Experience to date, planned operational procedures and ongoing R and D in this area are described

  1. Using RFID to Enhance Security in Off-Site Data Storage

    Directory of Open Access Journals (Sweden)

    Enrique de la Hoz

    2010-08-01

    Full Text Available Off-site data storage is one of the most widely used strategies in enterprises of all sizes to improve business continuity. In medium-to-large size enterprises, the off-site data storage processes are usually outsourced to specialized providers. However, outsourcing the storage of critical business information assets raises serious security considerations, some of which are usually either disregarded or incorrectly addressed by service providers. This article reviews these security considerations and presents a radio frequency identification (RFID-based, off-site, data storage management system specifically designed to address security issues. The system relies on a set of security mechanisms or controls that are arranged in security layers or tiers to balance security requirements with usability and costs. The system has been successfully implemented, deployed and put into production. In addition, an experimental comparison with classical bar-code-based systems is provided, demonstrating the system’s benefits in terms of efficiency and failure prevention.

  2. Using RFID to Enhance Security in Off-Site Data Storage

    Science.gov (United States)

    Lopez-Carmona, Miguel A.; Marsa-Maestre, Ivan; de la Hoz, Enrique; Velasco, Juan R.

    2010-01-01

    Off-site data storage is one of the most widely used strategies in enterprises of all sizes to improve business continuity. In medium-to-large size enterprises, the off-site data storage processes are usually outsourced to specialized providers. However, outsourcing the storage of critical business information assets raises serious security considerations, some of which are usually either disregarded or incorrectly addressed by service providers. This article reviews these security considerations and presents a radio frequency identification (RFID)-based, off-site, data storage management system specifically designed to address security issues. The system relies on a set of security mechanisms or controls that are arranged in security layers or tiers to balance security requirements with usability and costs. The system has been successfully implemented, deployed and put into production. In addition, an experimental comparison with classical bar-code-based systems is provided, demonstrating the system’s benefits in terms of efficiency and failure prevention. PMID:22163638

  3. Using RFID to enhance security in off-site data storage.

    Science.gov (United States)

    Lopez-Carmona, Miguel A; Marsa-Maestre, Ivan; de la Hoz, Enrique; Velasco, Juan R

    2010-01-01

    Off-site data storage is one of the most widely used strategies in enterprises of all sizes to improve business continuity. In medium-to-large size enterprises, the off-site data storage processes are usually outsourced to specialized providers. However, outsourcing the storage of critical business information assets raises serious security considerations, some of which are usually either disregarded or incorrectly addressed by service providers. This article reviews these security considerations and presents a radio frequency identification (RFID)-based, off-site, data storage management system specifically designed to address security issues. The system relies on a set of security mechanisms or controls that are arranged in security layers or tiers to balance security requirements with usability and costs. The system has been successfully implemented, deployed and put into production. In addition, an experimental comparison with classical bar-code-based systems is provided, demonstrating the system's benefits in terms of efficiency and failure prevention.

  4. NDMA guidelines on handling of nuclear and radiological emergencies

    Energy Technology Data Exchange (ETDEWEB)

    Abani, M C [National Disaster Management Authority, New Delhi (India)

    2010-07-01

    The vulnerability to the disasters is high in India due to the large population density, fast growing urbanization, industrialization and also because of poor economic conditions of people. Natural disasters have been recurring phenomena in India, leading to extensive loss of life, livelihood and property. The primary reason for such heavy losses can be attributed to the reactive and response-centric approach adopted in the past in handling of the disasters. Based on the Guidelines a holistic approach is to be adopted for Nuclear Emergency Management Framework that assigns the highest priority to prevention, mitigation and compliance to regulatory requirements, while strengthening preparedness, capacity development, response etc. It will be implemented through strengthening of the existing action plans or by preparing new action plans at national, state and district levels by the stakeholders at all levels of administration

  5. NDMA guidelines on handling of nuclear and radiological emergencies

    International Nuclear Information System (INIS)

    Abani, M.C.

    2010-01-01

    The vulnerability to the disasters is high in India due to the large population density, fast growing urbanization, industrialization and also because of poor economic conditions of people. Natural disasters have been recurring phenomena in India, leading to extensive loss of life, livelihood and property. The primary reason for such heavy losses can be attributed to the reactive and response-centric approach adopted in the past in handling of the disasters. Based on the Guidelines a holistic approach is to be adopted for Nuclear Emergency Management Framework that assigns the highest priority to prevention, mitigation and compliance to regulatory requirements, while strengthening preparedness, capacity development, response etc. It will be implemented through strengthening of the existing action plans or by preparing new action plans at national, state and district levels by the stakeholders at all levels of administration

  6. An off-site screening process for the public in radiation emergencies and disasters

    International Nuclear Information System (INIS)

    Yoon, Seok Won; Ho, Ha Wi; Jin, Young Woo

    2016-01-01

    A contamination screening process for the local population in radiation emergencies is discussed. We present an overview of the relevant Korean governmental regulations that underpin the development of an effective response system. Moreover, case studies of foreign countries responding to mass casualties are presented, and indicate that responses should be able to handle a large demand for contamination screening of the local public as well as screening of the immediate victims of the incident. We propose operating procedures for an off-site contamination screening post operated by the local government for members of the public who have not been directly harmed in the accident. In order to devise screening categories, sorting strategies assessing contamination and exposure are discussed, as well as a psychological response system. This study will lead to the effective operation of contamination screening clinics if an accident occurs. Furthermore, the role of contamination screening clinics in the overall context of the radiation emergency treatment system should be clearly established

  7. Low-level radioactive waste management: transitioning to off-site disposal at Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Dorries, Alison M.

    2010-01-01

    Facing the closure of nearly all on-site management and disposal capability for low-level radioactive waste (LLW), Los Alamos National Laboratory (LANL) is making ready to ship the majority of LLW off-site. In order to ship off-site, waste must meet the Treatment, Storage, and Disposal Facility's (TSDF) Waste Acceptance Criteria (WAC). In preparation, LANL's waste management organization must ensure LANL waste generators characterize and package waste compliantly and waste characterization documentation is complete and accurate. Key challenges that must be addressed to successfully make the shift to off-site disposal of LLW include improving the detail, accuracy, and quality of process knowledge (PK) and acceptable knowledge (AK) documentation, training waste generators and waste management staff on the higher standard of data quality and expectations, improved WAC compliance for off-site facilities, and enhanced quality assurance throughout the process. Certification of LANL generators will allow direct off-site shipping of LLW from their facilities.

  8. WALS: A sensor-based robotic system for handling nuclear materials

    International Nuclear Information System (INIS)

    Drotning, W.; Kimberly, H.; Wapman, W.

    1997-01-01

    An automated system is being developed for handling large payloads of radioactive nuclear materials in an analytical laboratory. The system uses machine vision and force/torque sensing to provide sensor-based control of the automation system to enhance system safety, flexibility, and robustness and achieve easy remote operation. The automation system also controls the operation of the laboratory measurement systems and the coordination of them with the robotic system. Particular attention has been given to system design features and analytical methods that provide an enhanced level of operational safety. Independent mechanical gripper interlock and too release mechanisms were designed to prevent payload mishandling. An extensive failure modes and effects analysis (FMEA) of the automation system was developed as a safety design analysis tool

  9. 40 CFR 273.18 - Off-site shipments.

    Science.gov (United States)

    2010-07-01

    ....18 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) STANDARDS FOR UNIVERSAL WASTE MANAGEMENT Standards for Small Quantity Handlers of Universal Waste § 273.18... universal waste. (c) If a universal waste being offered for off-site transportation meets the definition of...

  10. 40 CFR 273.38 - Off-site shipments.

    Science.gov (United States)

    2010-07-01

    ....38 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) STANDARDS FOR UNIVERSAL WASTE MANAGEMENT Standards for Large Quantity Handlers of Universal Waste § 273.38... universal waste. (c) If a universal waste being offered for off-site transportation meets the definition of...

  11. Safe handling of tritium

    International Nuclear Information System (INIS)

    1991-01-01

    The main objective of this publication is to provide practical guidance and recommendations on operational radiation protection aspects related to the safe handling of tritium in laboratories, industrial-scale nuclear facilities such as heavy-water reactors, tritium removal plants and fission fuel reprocessing plants, and facilities for manufacturing commercial tritium-containing devices and radiochemicals. The requirements of nuclear fusion reactors are not addressed specifically, since there is as yet no tritium handling experience with them. However, much of the material covered is expected to be relevant to them as well. Annex III briefly addresses problems in the comparatively small-scale use of tritium at universities, medical research centres and similar establishments. However, the main subject of this publication is the handling of larger quantities of tritium. Operational aspects include designing for tritium safety, safe handling practice, the selection of tritium-compatible materials and equipment, exposure assessment, monitoring, contamination control and the design and use of personal protective equipment. This publication does not address the technologies involved in tritium control and cleanup of effluents, tritium removal, or immobilization and disposal of tritium wastes, nor does it address the environmental behaviour of tritium. Refs, figs and tabs

  12. Siting Study for the Remote-Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    Lisa Harvego; Joan Connolly; Lance Peterson; Brennon Orr; Bob Starr

    2010-10-01

    The U.S. Department of Energy has identified a mission need for continued disposal capacity for remote-handled low-level waste (LLW) generated at the Idaho National Laboratory (INL). An alternatives analysis that was conducted to evaluate strategies to achieve this mission need identified two broad options for disposal of INL generated remote-handled LLW: (1) offsite disposal and (2) onsite disposal. The purpose of this study is to identify candidate sites or locations within INL boundaries for the alternative of an onsite remote handled LLW disposal facility and recommend the highest-ranked locations for consideration in the National Environmental Policy Act process. The study implements an evaluation based on consideration of five key elements: (1) regulations, (2) key assumptions, (3) conceptual design, (4) facility performance, and (5) previous INL siting study criteria, and uses a five-step process to identify, screen, evaluate, score, and rank 34 separate sites located across INL. The result of the evaluation is identification of two recommended alternative locations for siting an onsite remote-handled LLW disposal facility. The two alternative locations that best meet the evaluation criteria are (1) near the Advanced Test Reactor Complex and (2) west of the Idaho Comprehensive Environmental Response, Compensation, and Liability Act Disposal Facility.

  13. Considerations for handling failed fuel at the Barnwell Nuclear Fuel Plant

    International Nuclear Information System (INIS)

    Anderson, R.T.; Cholister, R.J.

    1982-05-01

    The impact of failed fuel receipt on reprocessing operations is qualitatively described. It appears that extended storage of fuel, particularly with advanced storage techniques, will increase the quantity of failed fuel, the nature and possibly the configuration of the fuel. The receipt of failed fuel at the BNFP increases handling problems, waste volumes, and operator exposure. If it is necessary to impose special operating precautions to minimize this impact, a loss in plant throughput will result. Hence, ideally, the reprocessing plant operator would take every reasonable precaution so that no failed fuel is received. An alternative policy would be to require that failed fuel be placed in a sealed canister. In the latter case the canister must be compatible with the shipping cask and suitable for in-plant storage. A required inspection of bare fuel would be made at the reactor prior to shipping off-site. This would verify fuel integrity. These requirements are obviously idealistic. Due to the current uncertain status of reprocessing and the need to keep reactors operating, business or governmental policy may be enacted resulting in the receipt of a negotiated quantity of non-standard fuel (including failed fuel). In this situation, BNFP fuel receiving policy based soley on fuel cladding integrity would be difficult to enforce. There are certain areas where process incompatibility does exist and where a compromise would be virtually impossible, e.g., canned fuel for which material or dimensional conflicts exist. This fuel would have to be refused or the fuel would require recanning prior to shipment. In other cases, knowledge of the type and nature of the failure may be acceptable to the operator. A physical inspection of the fuel either before shipment or after the cask unloading operation would be warranted. In this manner, concerns with pool contamination can be identified and the assembly canned if deemed necessary

  14. Impact of hazardous waste handling legislation on nuclear installations and radioactive waste management in the United States

    International Nuclear Information System (INIS)

    Trosten, L.M.

    1988-01-01

    The United States has enacted complex legislation to help assure proper handling of hazardous waste and the availability of funds to cover the expenditures. There are a number of uncertainties concerning the impact of this legislation, and regulations promulgated by the Environmental Protection Agency and the states, upon nuclear installations and radioactive waste management. This report provides an overview of the U.S. hazardous waste legislation and examines the outlook for its application to the nuclear industry (NEA) [fr

  15. Pesticide use and off-site risk assessment

    NARCIS (Netherlands)

    Yang, X.

    2016-01-01

    Pesticide use and off-site risk assessment: a case study of glyphosate fate in Chinese Loess soil

    Xiaomei Yang

    Abstract: Repeated applications of pesticide may contaminate the soil and water, threatening their quality within the

  16. Closure of Off-Site FTP

    CERN Multimedia

    2004-01-01

    To reduce the number of regular break-ins on CERN machines due to passwords exposed on the network in clear text, OFF-SITE FTP ACCESS TO CERN WILL BE BLOCKED in the CERN firewall from: Tuesday 20th January 2004 If you use ftp to access CERN computers from outside CERN then please see the link below for alternative access means and further advice: http://cern.ch/security/ftp Denise Heagerty, CERN Computer Security officer, Computer.Security@cern.ch

  17. Closure of Off-Site Telnet

    CERN Multimedia

    Denise Heagerty

    2003-01-01

    To reduce the number of regular break-ins on CERN machines due to passwords exposed on the network in clear text, OFF-SITE TELNET ACCESS TO CERN WILL BE BLOCKED in the CERN firewall from Tuesday 28th January 2003 If you use telnet to access CERN computers from outside CERN then please see the link below for alternative access means and further advice http://cern.ch/security/telnet Denise Heagerty, CERN Computer Security officer, Computer.Security@cern.ch

  18. Closure of Off-Site FTP

    CERN Multimedia

    2003-01-01

    To reduce the number of regular break-ins on CERN machines due to passwords exposed on the network in clear text, OFF-SITE FTP ACCESS TO CERN WILL BE BLOCKED in the CERN firewall from: Tuesday 20th January 2004 If you use ftp to access CERN computers from outside CERN then please see the link below for alternative access means and further advice: http://cern.ch/security/ftp Denise Heagerty, CERN Computer Security officer, Computer.Security@cern.ch

  19. CLOSURE OF OFF-SITE TELNET

    CERN Multimedia

    Denise Heagerty

    2002-01-01

    To reduce the number of regular break-ins on CERN machines due to passwords exposed on the network in clear text, OFF-SITE TELNET ACCESS TO CERN WILL BE BLOCKED in the CERN firewall from Tuesday 28 January 2003 If you use telnet to access CERN computers from outside CERN then please see the link below for alternative access means and further advice http://cern.ch/security/telnet Denise Heagerty, CERN Computer Security officer, Computer.Security@cern.ch

  20. Application of RASCAL code for multiunit accident in domestic nuclear sites

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Hyun; Jeong, Seung Young [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-10-15

    All of domestic nuclear power plant sites are multiunit site (at least 5 - 6 reactors are operating), so this capability has to be quickly secured for nuclear licensee and institutes responsible for nuclear emergency response. In this study, source term and offsite dose from multiunit event were assessed using a computer code, RASCAL. An emergency exercise scenario was chosen to verify applicability of the codes to domestic nuclear site accident. Employing tools and new features of the code, such as merging more than two individual source terms and source term estimate for long term progression accident, main parameters and information in the scenario, release estimates and dose projections were performed. Radiological releases and offsite doses from multiunit accident were calculated using RASCAL.. A scenario, in which three reactors were damaged coincidently by a great natural disaster, was considered. Surrogate plants were chosen for the code calculation. Source terms of each damaged unit were calculated individually first, and then total source term and integrated offsite dose assessment data was acquired using a source term merge function in the code. Also comparison between LTSBO and LOCA source term estimate options was performed. Differences in offsite doses were caused by release characteristics. From LTSBO option, iodines were released much higher than LOCA. Also LTSBO source term release was delayed and the duration was longer than LOCA. This option would be useful to accidents which progress with much longer time frame than LOCA. RASCAL can be useful tool for radiological consequence assessment in domestic nuclear site accidents.

  1. Human factors issues in fuel handling

    International Nuclear Information System (INIS)

    Beattie, J.D.; Iwasa-Madge, K.M.; Tucker, D.A.

    1994-01-01

    The staff of the Atomic Energy Control Board wish to further their understanding of human factors issues of potential concern associated with fuel handling in CANDU nuclear power stations. This study contributes to that objective by analysing the role of human performance in the overall fuel handling process at Ontario Hydro's Darlington Nuclear Generating Station, and reporting findings in several areas. A number of issues are identified in the areas of design, operating and maintenance practices, and the organizational and management environment

  2. A sensor-based automation system for handling nuclear materials

    International Nuclear Information System (INIS)

    Drotning, W.; Kimberly, H.; Wapman, W.; Darras, D.

    1997-01-01

    An automated system is being developed for handling large payloads of radioactive nuclear materials in an analytical laboratory. The automation system performs unpacking and repacking of payloads from shipping and storage containers, and delivery of the payloads to the stations in the laboratory. The system uses machine vision and force/torque sensing to provide sensor-based control of the automation system in order to enhance system safety, flexibility, and robustness, and achieve easy remote operation. The automation system also controls the operation of the laboratory measurement systems and the coordination of them with the robotic system. Particular attention has been given to system design features and analytical methods that provide an enhanced level of operational safety. Independent mechanical gripper interlock and tool release mechanisms were designed to prevent payload mishandling. An extensive Failure Modes and Effects Analysis of the automation system was developed as a safety design analysis tool

  3. Development of a computer code system for selecting off-site protective action in radiological accidents based on the multiobjective optimization method

    International Nuclear Information System (INIS)

    Ishigami, Tsutomu; Oyama, Kazuo

    1989-09-01

    This report presents a new method to support selection of off-site protective action in nuclear reactor accidents, and provides a user's manual of a computer code system, PRASMA, developed using the method. The PRASMA code system gives several candidates of protective action zones of evacuation, sheltering and no action based on the multiobjective optimization method, which requires objective functions and decision variables. We have assigned population risks of fatality, injury and cost as the objective functions, and distance from a nuclear power plant characterizing the above three protective action zones as the decision variables. (author)

  4. Closeout of IE Bulletin 80-15: Possible loss of Emergency Notification System (ENS) with loss of offsite power

    International Nuclear Information System (INIS)

    Foley, W.J.; Dean, R.S.; Hennick, A.

    1990-12-01

    Documentation is provided in this report for the closeout of IE Bulletin 80-15 for nuclear power reactors. This bulletin pertained to a possible loss of the Emergency Notification System (ENS) upon loss of offsite power. Closeout is based on the implementation and verification of six (6) required actions by licensees of nuclear power reactors in operation or near to receiving an operating license when the bulletin was issued on June 18, 1980. Evaluation of utility responses and NRC/Region inspection reports indicates that the bulletin is closed for all of the 69 nuclear power reactors to which it was issued for action and which were not shut down indefinitely or permanently at the time of issuance of this report. Background information is supplied in the Introduction and Appendix A. Nuclear fuel facilities as well as nuclear power facilities were identified in the enclosures to the bulletin. However, per an NRC memorandum, the closeout of the bulletin for nuclear fuel facilities is not within the scope of this report

  5. Cask system design guidance for robotic handling

    International Nuclear Information System (INIS)

    Griesmeyer, J.M.; Drotning, W.D.; Morimoto, A.K.; Bennett, P.C.

    1990-10-01

    Remote automated cask handling has the potential to reduce both the occupational exposure and the time required to process a nuclear waste transport cask at a handling facility. The ongoing Advanced Handling Technologies Project (AHTP) at Sandia National Laboratories is described. AHTP was initiated to explore the use of advanced robotic systems to perform cask handling operations at handling facilities for radioactive waste, and to provide guidance to cask designers regarding the impact of robotic handling on cask design. The proof-of-concept robotic systems developed in AHTP are intended to extrapolate from currently available commercial systems to the systems that will be available by the time that a repository would be open for operation. The project investigates those cask handling operations that would be performed at a nuclear waste repository facility during cask receiving and handling. The ongoing AHTP indicates that design guidance, rather than design specification, is appropriate, since the requirements for robotic handling do not place severe restrictions on cask design but rather focus on attention to detail and design for limited dexterity. The cask system design features that facilitate robotic handling operations are discussed, and results obtained from AHTP design and operation experience are summarized. The application of these design considerations is illustrated by discussion of the robot systems and their operation on cask feature mock-ups used in the AHTP project. 11 refs., 11 figs

  6. The Chalk River Nuclear Laboratories contingency plan -a brief description

    International Nuclear Information System (INIS)

    White, J.M.

    1982-01-01

    A brief description of the contingency plan which deals with both the on-site and off-site consequences of a serious nuclear accident is given. The off-site consequences of different size releases and the subsequent action taken by employees, radiation protection experts, municipal, Provincial and Federal authorities is described and the interaction of the various groups is discussed. (author)

  7. Storage, handling and internal transport of radioactive materials (fuel elements excepted) in nuclear power plants

    International Nuclear Information System (INIS)

    1983-06-01

    The rule applies to storage and handling as well as to transport within the plant and to the exchange of - solid radioactive wastes, - liquid radioactive wastes, except for those covered by the rule KTA 3603, - radioactive components and parts which are planned to be mounted and dismounted until shutdown of the plant, - radioactive-contaminated tools and appliances, - radioactive preparations. The rule is to be applied within the fenced-in sites of stationary nuclear power plants with LWR or HTR including their transport load halls, as fas as these are situated so as to be approachable from the nuclear power station by local transport systems. (orig./HP) [de

  8. RPV in-situ segmentation combined with off-site treatment for volume reduction and recycling - Proven In-Situ Segmentation Combined with Off-Site Treatment for Volume Reduction and Recycling. RPV case study

    International Nuclear Information System (INIS)

    Larsson, Arne; Lidar, Per; Segerud, Per; Hedin, Gunnar

    2014-01-01

    Decommissioning of nuclear power plants generates large volumes of radioactive or potentially radioactive waste. The proper management of the large components and the dismantling waste are key success factors in a decommissioning project. A large component of major interest is, due to its size and its span in radioactivity content, the RVP, which can be disposed as is or be segmented, treated, partially free released for recycling and conditioned for disposal in licensed packages. To a certain extent the decommissioning program have to be led by the waste management process. The costs for the plant decommissioning can be reduced by the usage of off-site waste treatment facilities as the time needed for performing the decommissioning project will be reduced as well as the waste volumes for disposal. Long execution times and delays due to problems with on-site waste management processes are major cost drivers for decommissioning projects. This involves also the RPV. In Sweden, the extension of the geological repository SFR plans for a potential disposal of whole RPVs. Disposal of whole RPVs is currently the main alternative but other options are considered. The target is to avoid extensive on-site waste management of RPVs to reduce the risk for delays. This paper describes in-situ RPV segmentation followed by off-site treatment aiming for free release for recycling of a substantial amount of the material, and volume efficient conditioning of the remaining parts. Real data from existing LWR RPVs was used for this study. Proven segmentation methods are intended to be used for the in situ segmentation followed by proven methods for packaging, transportation, treatment, recycling and conditioning for disposal. The expected volume reduction for disposal can be about 90% compared to whole RPV disposal. In this respect the in-situ segmentation of the RVPs to large pieces followed by off-site treatment is an interesting alternative that fits very well with the objective

  9. OFFSITE RADIOLOGICAL CONSEQUENCE CALCULATION FOR THE BOUNDING MIXING OF INCOMPATIBLE MATERIALS ACCIDENT

    International Nuclear Information System (INIS)

    SANDGREN, K.R.

    2006-01-01

    This document quantifies the offsite radiological consequence of the bounding mixing of incompatible materials accident for comparison with the 25 rem Evaluation Guideline established in Appendix A of DOE-STD-3009. The bounding accident is an inadvertent addition of acid to a waste tank. The calculated offsite dose does not challenge the Evaluation Guideline. Revision 4 updates the analysis to consider bulk chemical additions to single shell tanks (SSTs)

  10. On-site transportation and handling of uranium-233 special nuclear material: Preliminary hazards and accident analysis. Final

    International Nuclear Information System (INIS)

    Solack, T.; West, D.; Ullman, D.; Coppock, G.; Cox, C.

    1995-01-01

    U-233 Special Nuclear Material (SNM) currently stored at the T-Building Storage Areas A and B must be transported to the SW/R Tritium Complex for repackaging. This SNM is in the form of oxide powder contained in glass jars which in turn are contained in heat sealed double polyethylene bags. These doubled-bagged glass jars have been primarily stored in structural steel casks and birdcages for approximately 20 years. The three casks, eight birdcages, and one pail/pressure vessel will be loaded onto a transport truck and moved over an eight day period. The Preliminary Hazards and Accident Analysis for the on-site transportation and handling of Uranium-233 Special Nuclear Material, documented herein, was performed in accordance with the format and content guidance of DOE-STD-3009-94, Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports, dated July 1994, specifically Chapter Three, Hazard and Accident Analysis. The Preliminary Hazards Analysis involved detailed walkdowns of all areas of the U-233 SNM movement route, including the T-Building Storage Area A and B, T-Building truck tunnel, and the roadway route. Extensive discussions were held with operations personnel from the Nuclear Material Control Group, Nuclear Materials Accountability Group, EG and G Mound Security and the Material Handling Systems Transportation Group. Existing documentation related to the on-site transportation of hazardous materials, T-Building and SW/R Tritium Complex SARs, and emergency preparedness/response documentation were also reviewed and analyzed to identify and develop the complete spectrum of energy source hazards

  11. 40 CFR 1400.5 - Internet access to certain off-site consequence analysis data elements.

    Science.gov (United States)

    2010-07-01

    ... consequence analysis data elements. 1400.5 Section 1400.5 Protection of Environment ENVIRONMENTAL PROTECTION... UNDER THE CLEAN AIR ACT SECTION 112(r)(7); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Public Access § 1400.5 Internet access to certain off...

  12. Incineration of DOE offsite mixed waste at the Idaho National Engineering and Environmental Laboratory

    International Nuclear Information System (INIS)

    Harris, J.D.; Harvego, L.A.; Jacobs, A.M.; Willcox, M.V.

    1998-01-01

    The Waste Experimental Reduction Facility (WERF) incinerator at the Idaho National Engineering and Environmental Laboratory (INEEL) is one of three incinerators in the US Department of Energy (DOE) Complex capable of incinerating mixed low-level waste (MLLW). WERF has received MLLW from offsite generators and is scheduled to receive more. The State of Idaho supports receipt of offsite MLLW waste at the WERF incinerator within the requirements established in the (INEEL) Site Treatment Plan (STP). The incinerator is operating as a Resource Conservation and Recovery Act (RCRA) Interim Status Facility, with a RCRA Part B permit application currently being reviewed by the State of Idaho. Offsite MLLW received from other DOE facilities are currently being incinerated at WERF at no charge to the generator. Residues associated with the incineration of offsite MLLW waste that meet the Envirocare of Utah waste acceptance criteria are sent to that facility for treatment and/or disposal. WERF is contributing to the treatment and reduction of MLLW in the DOE Complex

  13. Transient evaluation using EMTP at one phase opening of the offsite power transformer of the emergency power supply systems for nuclear power plants. This report is a follow-up the last year's

    International Nuclear Information System (INIS)

    Shimada, Yoshio

    2014-01-01

    The emergency power supply systems of nuclear power plants, as the objects of this research, are critical to supplying stable electric power to such systems as the emergency core cooling system (ECCS), and to maintaining the safety of the nuclear power reactor; this was apparent from the accident at the Fukushima Daiichi Nuclear Power Station. The USNRC issued regulatory documents (BL 2012-01, IN 2012-03), and has commenced to review those problems which cannot be detected by degraded-voltage protection relays such as new design vulnerability of the power supply systems that are certain kinds of winding structures and iron core structures when the one-phase open fault occurs without a ground fault of the transformer connected to the offsite power supply system, including when the offsite power supply side is a wye connection and the load side is a delta connection etc. The report of the INSS JOURNAL 2013 used simulation by the electro-magnetic transients program (EMTP) and obtained findings that clearly specified the response at the time of the power supply side one-phase open without ground fault of various winding structures and various iron core structures of the three-phase transformer and identified the important issues for the response of emergency power supply systems and the safety related components of representative domestic PWR plants when the one-phase open fault occurred without ground fault. In a continuation of the previous report, this paper summarizes the previous results, and then presents the principles by which normal voltages are maintained by both the primary and the secondary sides when the one-phase open failure without ground fault occurs on the primary side of the transformer, and the results of the analysis of operations of the protection relays of the emergency power systems and the safety related components of representative a domestic PWR plant by the simulation using EMTP. (author)

  14. Human factors issues in fuel handling

    Energy Technology Data Exchange (ETDEWEB)

    Beattie, J D; Iwasa-Madge, K M; Tucker, D A [Humansystems Inc., Milton, ON (Canada)

    1994-12-31

    The staff of the Atomic Energy Control Board wish to further their understanding of human factors issues of potential concern associated with fuel handling in CANDU nuclear power stations. This study contributes to that objective by analysing the role of human performance in the overall fuel handling process at Ontario Hydro`s Darlington Nuclear Generating Station, and reporting findings in several areas. A number of issues are identified in the areas of design, operating and maintenance practices, and the organizational and management environment. 1 fig., 4 tabs., 19 refs.

  15. Descriptive statistics of occupational employment in nuclear power utilities. Final working paper

    International Nuclear Information System (INIS)

    Little, J.R.; Johnson, R.C.

    1982-10-01

    The Institute of Nuclear Power Operations conducted a survey of its 58 member utilities during the Spring of 1982. This was the second such survey performed to identify employment trends and to project needs for trained personnel in the industry to 1991. The first was performed in 1981. The 1982 employment survey consisted of four questionnaires, asking for information on: (1) on-site employment; (2) on-site turnover; (3) off-site employment; and (4) off-site turnover. The survey instruments were designed to reflect approaches used by the utilities to meet the labor requirements for operation of nuclear power plants through off-site support personnel, contractors, and holding company personnel, as well as utility employees working at the plant site. On-site information was received from all 83 plants at the 58 utilities. However, employment information from Surry of VEPCO arrived too late to be included in the analysis. Therefore, their numbers are reflected in the adjusted totals. Responses to requests for off-site employment information were received from 55 of the 58 utilities

  16. Off-site emergency response plans in case of technological catastrophes: the case Angra dos Reis

    International Nuclear Information System (INIS)

    Souza Junior, M.D. de.

    1990-04-01

    In the first part of the thesis a discussion of the technical, operational and methodological features of the current practices for emergency planning in case of a nuclear fallout. Based on this general reference is possible to evaluate the features in the natural and social environment of Angra dos Reis that probably will obstruct the application of the protective countermeasures to the public. These critical points are enhanced to permit the discussion of a methodological approach that is supposed to be suitable to the reality of Angra dos Reis. The approach was developed specifically to this region and was introduced as a part of the general emergency off-site plan to the Central Nuclear Almirante Alvaro Alberto (CNAAA). Starting from this experience will be possible to enlarge this approach in a further research, in order to study this potential hazards of other industrial plants. (author)

  17. Nukem Nuclear GmbH activity in CIS countries in the sphere of radioactive wastes and nuclear fuel handling

    International Nuclear Information System (INIS)

    Vaihard, A.

    1997-01-01

    NUKEM was founded in 1960 as one of the first nuclear companies in the German Federal Republic. With this work, Nukem developed not only processes for producing fuels and fuel elements, but also the plant and equipment necessary for this production. NUKEM engineers further planned and built the total infrastructure for operation of these manufacturing plants, including the supply and waste plants, the nuclear ventilation technology, the laboratory and the remote handling manipulators. The scope of activities extends from the design to the manufacture and construction of turnkey plants. The points of emphasis are plants and processes for the Treatment of radioactive wastes, storage and container technology, the decommissioning of nuclear plants, the planning and building of nuclear laboratories, the design of fuel elements and safety and monitoring technology. NUKEM Nuclear Technology is an independent division within the plant construction of the NUKEM Group. Additionally, five further subsidiary and partner companies have a spectrum of nuclear technology activities. Altogether, Nukem Nuclear Technology counts around 300 highly qualified engineers, scientists and technicians. Numerous Designs and patents underline the strength of innovative output in this area. The engineering service offered by NUKEM includes the whole spectrum of process and technology as well as construction and start-up as general engineer or general contractor: Basic engineering; Detail engineering; Procurement; Personnel Training; Start-up. Engineering and safety for nuclear technology: Process and plant planing; Media supply and disposal; Building and architecture; Electrical, measurement and control technology; Safety and accident analysis; Licensing procedures. Treatment of Radioactive Wastes: - Volume reduction of soil and liquid wastes: vaporizer plants; evaporator plants; incineration plants; pyrolysis plants; compactors. - Chemical/physical processes for residue treatment: boric acid

  18. Off-site environmental monitoring report. Radiation monitoring around United States nuclear test areas, calendar year 1985

    International Nuclear Information System (INIS)

    Grossman, R.F.; Black, S.C.; Dye, R.E.; Smith, D.D.; Thome, D.J.; Mullen, A.A.

    1986-04-01

    The EMSL-LV operates an Off-Site Radiological Safety Program around the NTS and other sites as requested by the Department of Energy (DOE) under an Interagency Agreement between DOE and EPA. This report, prepared in accordance with DOE guidelines (DOE85a), covers the program activities for calendar year 1985. It contains descriptions of pertinent features of the NTS and its environs, summaries of the EMSL-LV dosimetry and sampling methods, analytical procedures, quality assurance, and the analytical results from environmental measurements. Where applicable, dosimetry and sampling data are compared to appropriate guides for external and internal exposures of humans to ionizing radiation

  19. The FMEA Analysis for Fuel Handling System at Cernavoda Unit 2

    International Nuclear Information System (INIS)

    Park, Jin Hee; Kim, Tae Woon; Rhee, Bo Wook; Yoon, Chul; Kim, Hyeong Tae; Cho, In Gil; Kim, Seong Ho

    2006-01-01

    A Nuclear Safety Evaluation was performed by an independent assessor at the request of the regulatory authority CNCAN (Comisia Nationala pentru Controlul Activitatilor Nucleare. National Committee for Nuclear Activities Control in Romania) to provide an independent overview of all the nuclear safety aspects of Cernavoda Unit 2 under construction and an expert opinion whether the completed Cernavoda Unit-2 Nuclear Power Plant would satisfy current Western European nuclear safety objectives and practices. A report was produced (Cernavoda 2 Nuclear Safety Expert Project, 'Task 10 . Safety Evaluation Report', A.F.Parsons, NNC Limited, December 2001) and contains recommendations either mandatory or advisory. The FMEA study, one of the mandatory recommendations, is performing now for fuel handling system and radioactive waste handling system for Cernavoda unit 2 in Romania sponsored by KHNP. In this paper, only the FMEA study for fuel handling system is presented

  20. Operational analysis and improvement of a spent nuclear fuel handling and treatment facility using discrete event simulation

    International Nuclear Information System (INIS)

    Garcia, H.E.

    2000-01-01

    Spent nuclear fuel handling and treatment often require facilities with a high level of operational complexity. Simulation models can reveal undesirable characteristics and production problems before they become readily apparent during system operations. The value of this approach is illustrated here through an operational study, using discrete event modeling techniques, to analyze the Fuel Conditioning Facility at Argonne National Laboratory and to identify enhanced nuclear waste treatment configurations. The modeling approach and results of what-if studies are discussed. An example on how to improve productivity is presented.

  1. 40 CFR 68.33 - Defining offsite impacts-environment.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 15 2010-07-01 2010-07-01 false Defining offsite impacts-environment. 68.33 Section 68.33 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR... impacts—environment. (a) The owner or operator shall list in the RMP environmental receptors within a...

  2. Improved Safety Margin Characterization of Risk from Loss of Offsite Power

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, Paul [Texas A & M Univ., College Station, TX (United States)

    2017-11-07

    Original intent: The original intent of this task was “support of the Risk-Informed Safety Margin Characteristic (RISMC) methodology in order” “to address … efficiency of computation so that more accurate and cost-effective techniques can be used to address safety margin characterizations” (S. M. Hess et al., “Risk-Informed Safety Margin Characterization,” Procs. ICONE17, Brussels, July 2009, CD format). It was intended that “in Task 1 itself this improvement will be directed toward upon the very important issue of Loss of Offsite Power (LOOP) events,” more specifically toward the challenge of efficient computation of the multidimensional nonrecovery integral that has been discussed by many previous contributors to the theory of nuclear safety. It was further envisioned that “three different computational approaches will be explored,” corresponding to the three subtasks listed below; deliverables were tied to the individual subtasks.

  3. American nuclear insurers

    International Nuclear Information System (INIS)

    Oliveira, R.A.

    1988-01-01

    Nuclear liability insurance covers liability for damages directly caused by the nuclear energy hazard. This coverage includes offsite bodily injury and property damage sustained by members of the general public, and bodily injury to onsite third party personnel. Recent nuclear liability claims allege bodily injury and property damage resulting from releases or radioactive materials to the environmental and occupational radiation worker exposures. Routine reactor operations involving radioactive waste have the potential to result in such claims. The nuclear insurance Pools believe that one way such claims can be minimized is through the implementation of an effective radioactive waste management program

  4. Remote handling machines

    International Nuclear Information System (INIS)

    Sato, Shinri

    1985-01-01

    In nuclear power facilities, the management of radioactive wastes is made with its technology plus the automatic techniques. Under the radiation field, the maintenance or aid of such systems is important. To cope with this situation, MF-2 system, MF-3 system and a manipulator system as remote handling machines are described. MF-2 system consists of an MF-2 carrier truck, a control unit and a command trailer. It is capable of handling heavy-weight objects. The system is not by hydraulic but by electrical means. MF-3 system consists of a four-crawler truck and a manipulator. The truck is versatile in its posture by means of the four independent crawlers. The manipulator system is bilateral in operation, so that the delicate handling is made possible. (Mori, K.)

  5. Highlights of the American Nuclear Society topical meeting on the treatment and handling of radioactive wastes

    International Nuclear Information System (INIS)

    Blasewitz, A.G.; Lerch, R.E.; Richardson, G.L.

    1983-01-01

    The American Nuclear Society Topical Meeting on the Treatment and Handling of Radioactive Wastes was held in Richland, Washington, from 19-22 April 1982. The object of the meeting was to provide a thorough assessment of the status of technology. The response to the meeting was excellent: 123 papers were presented. There were 505 registrations; 83 were from outside the USA, representing 13 countries. The large and diverse attendance provided a broad technological view and perspective. The following major points emerged from the conference: (1) In an extensive world-wide effort, techniques are being developed to cover all phases of radioactive waste management. (2) A broad and deep technological base has been developed. (3) Many adequate processes are ready for actual application while others are ready for demonstration of applicability. These demonstrations are important to further public acceptance of nuclear energy. (4) At the present level of maturity, systematic analyses should be performed to determine actual requirements for the treatment and handling of radioactive wastes. These analyses can be used to focus our research and development, and demonstration activities to achieve treatment and conditioning systems which are both appropriate and cost-effective. (author)

  6. Radiological criteria for underground nuclear tests

    International Nuclear Information System (INIS)

    Malik, J.S.; Brownlee, R.R.; Costa, C.F.; Mueller, H.F.; Newman, R.W.

    1981-04-01

    The radiological criteria for the conduct of nuclear tests have undergone many revisions with the current criteria being 0.17 rad for uncontrolled populations and 0.5 rad for controllable populations. Their effect upon operations at the Nevada Test Site and the current off-site protective plans are reviewed for areas surrounding the Site. The few accidental releases that have occurred are used to establish estimates of probability of release and of hazard to the population. These are then put into context by comparing statistical data on other accidents and cataclysms. The guidelines established by DOE Manual Chapter MC-0524 have never been exceeded during the entire underground nuclear test program. The probability of real hazard to off-site populations appears to be sufficiently low as not to cause undue concern to the citizenry

  7. Radiological criteria for underground nuclear tests

    Energy Technology Data Exchange (ETDEWEB)

    Malik, J.S.; Brownlee, R.R.; Costa, C.F.; Mueller, H.F.; Newman, R.W.

    1981-04-01

    The radiological criteria for the conduct of nuclear tests have undergone many revisions with the current criteria being 0.17 rad for uncontrolled populations and 0.5 rad for controllable populations. Their effect upon operations at the Nevada Test Site and the current off-site protective plans are reviewed for areas surrounding the Site. The few accidental releases that have occurred are used to establish estimates of probability of release and of hazard to the population. These are then put into context by comparing statistical data on other accidents and cataclysms. The guidelines established by DOE Manual Chapter MC-0524 have never been exceeded during the entire underground nuclear test program. The probability of real hazard to off-site populations appears to be sufficiently low as not to cause undue concern to the citizenry.

  8. SRTC criticality technical review: Nuclear Criticality Safety Evaluation 93-18 Uranium Solidification Facility's Waste Handling Facility

    International Nuclear Information System (INIS)

    Rathbun, R.

    1993-01-01

    Separate review of NMP-NCS-930058, open-quotes Nuclear Criticality Safety Evaluation 93-18 Uranium Solidification Facility's Waste Handling Facility (U), August 17, 1993,close quotes was requested of SRTC Applied Physics Group. The NCSE is a criticality assessment to determine waste container uranium limits in the Uranium Solidification Facility's Waste Handling Facility. The NCSE under review concludes that the NDA room remains in a critically safe configuration for all normal and single credible abnormal conditions. The ability to make this conclusion is highly dependent on array limitation and inclusion of physical barriers between 2x2x1 arrays of boxes containing materials contaminated with uranium. After a thorough review of the NCSE and independent calculations, this reviewer agrees with that conclusion

  9. 40 CFR 1400.3 - Public access to paper copies of off-site consequence analysis information.

    Science.gov (United States)

    2010-07-01

    ...-site consequence analysis information. 1400.3 Section 1400.3 Protection of Environment ENVIRONMENTAL... PROGRAMS UNDER THE CLEAN AIR ACT SECTION 112(r)(7); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Public Access § 1400.3 Public access to...

  10. 40 CFR 68.30 - Defining offsite impacts-population.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 15 2010-07-01 2010-07-01 false Defining offsite impacts-population. 68.30 Section 68.30 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR... impacts—population. (a) The owner or operator shall estimate in the RMP the population within a circle...

  11. A versatile data handling system for nuclear physics experiments based on PDP 11/03 micro-computers

    International Nuclear Information System (INIS)

    Raaf, A.J. de

    1979-01-01

    A reliable and low cost data handling system for nuclear physics experiments is described. It is based on two PDP 11/03 micro-computers together with Gec-Elliott CAMAC equipment. For the acquisition of the experimental data a fast system has been designed. It consists of a controller for four ADCs together with an intelligent 38k MOS memory with a word size of 24 bits. (Auth.)

  12. The Follow-up IAEA International Mission on Remediation of Large Contaminated Areas Off-Site the Fukushima Daiichi Nuclear Power Plant, Tokyo and Fukushima Prefecture, Japan, 14-21 October 2013. Final Report

    International Nuclear Information System (INIS)

    2014-01-01

    In October 2011, the IAEA conducted an International Mission to Japan to support the remediation of large contaminated areas off-site TEPCO's Fukushima Daiichi Nuclear Power Plant (NPP). In response to the request made by the Government of Japan, in October 2013, the IAEA organized a follow-up International Mission on remediation of large contaminated areas off-site TEPCO's Fukushima Daiichi NPP (hereinafter referred to as the 'Follow-up Mission' or the 'Mission') with the main purpose of evaluating the progress of the on-going remediation works achieved since the previous mission in October 2011. The Follow-up Mission Team involved 13 international experts. Additionally, 3 experts of the Working Group 5 (Subgroup 5.2, Remediation) in charge of preparing the IAEA Report on TEPCO Fukushima Daiichi Accident accompanied the Mission as observers to obtain first-hand information for the report. The Follow-up Mission had the following three objectives: 1. To provide assistance to Japan in assessing the progress made with the remediation of the Special Decontamination Area (not included in the previous mission of 2011) and the Intensive Contamination Survey Areas; 2. To review remediation strategies, plans and works, in view of the advice provided by the previous mission on remediation of large contaminated off-site areas; and 3. To share its findings with the international community as lessons learned. The Mission was conducted through the assessment of information provided to the Team and by means of professional and open discussions with the relevant institutions in Japan, including national, prefectural and local institutions. The Japanese authorities provided comprehensive information on their remediation programme. The Mission Team visited the affected areas, including several sites where activities on remediation were conducted. The Team also visited some temporary storage sites for radioactive waste and soil generated in the remediation activities, as well as a

  13. Nuclear safety philosophy and its general application to fuel management and handling - a regulator's viewpoint

    International Nuclear Information System (INIS)

    Petty, I.C.

    1995-01-01

    The Nuclear Safety Division (NSD) of the Health and Safety Executive (HSE) informs the UK Nuclear Industry of the principles that it applies in assessing whether licensees have demonstrated that their nuclear plants are as safe as is reasonably practicable. The paper commences with a discussion of the non-prescriptive approach to health and safety regulation which is the basis of the regulatory activities of NSD's operating arm -the Nuclear Installations Inspectorate (NII). It then describes in broad terms the overall approach used by NII for analysing the safety of nuclear plant, including fuel, which will cover both deterministic and probabilistic methodologies. The paper then introduces the sections of the Safety Assessment Principles which apply to nuclear fuel safety (both fuel handling and management). Most of these principles are of a general nature and do not just apply to fuel. The paper explains how safety cases might relate to the SAPs and offers some views on how a licensee might interpret them in developing his safety case. Particular emphasis is placed on the importance of submitting a high quality safety case and the type of information that should be in it. The advantages of the approach proposed, to the licensee as well as to the regulator, are identified. (author)

  14. Status of the Real-time On-line Decision Support (RODOS) system for off-site emergency management after nuclear and radiological accidents

    International Nuclear Information System (INIS)

    Raskov, W.; Ehrhardt, J.; Landman, C.; Pasler-Sauer, J.

    2006-01-01

    Under the auspices of its EURATOM Research Framework Programmes, the European Commission (EC) has supported the development of the comprehensive decision support system RODOS (Real-time On-line Decision Support) for off-site emergency management after nuclear accidents for more than a decade. Many national research programmes, research institutes and industrial collaborators contributed to the project, in particular the German Ministry of Environment, Nature Conservation and Reactor Safety (B MU). The RODOS system can be applied to accidental releases into the atmosphere and various aquatic environments within and across Europe. It provides coherent support before, during and after such a release to assist analysis of the situation and decision making about short and long-term countermeasures for mitigating the consequences with respect to health, the environment, and the economy. Appropriate interfaces exist with local and national radiological monitoring data systems, meteorological measurements and forecasts, and for the adaptation to local, regional and national conditions in Europe. Within the European Integrated Project EURANOS of the sixth Framework Programme, the RODOS system is being enhanced, among others, for radiological emergencies such as dirty bombs attacks, transport accidents and satellite crashes by extensions of the nuclide list, the source term characteristics and the atmospheric dispersion model

  15. Introducing PCTRAN as an evaluation tool for nuclear power plant emergency responses

    International Nuclear Information System (INIS)

    Cheng, Yi-Hsiang; Shih, Chunkuan; Chiang, Show-Chyuan; Weng, Tung-Li

    2012-01-01

    Highlights: ► PCTRAN is integrated with an atmospheric dispersion algorithm. ► The improved PCTRAN acts as an accident/incident simulator and a data exchange system. ► The software helps the responsible organizations decide the rescue and protective actions. ► The evaluation results show the nuclear power plant accident and its off-site dose consequences. ► The software can be used for nuclear power plant emergency responses. - Abstract: Protecting the public from radiation exposure is important if a nuclear power plant (NPP) accident occurs. Deciding appropriate protective actions in a timely and effective manner can be fulfilled by using an effective accident evaluation tool. In our earlier work, we have integrated PCTRAN (Personal Computer Transient Analyzer) with the off-site dose calculation model. In this study, we introduce PCTRAN as an evaluation tool for nuclear power plant emergency responses. If abnormal conditions in the plant are monitored or observed, the plant staffs can distinguish accident/incident initiation events. Thus, the responsible personnel can immediately operate PCTRAN and set up those accident/incident initiation events to simulate the nuclear power plant transient or accident in conjunction with off-site dose distributions. The evaluation results consequently help the responsible organizations decide the rescue and protective actions. In this study, we explain and demonstrate the capabilities of PCTRAN for nuclear emergency responses, through applying it to simulate the postulated nuclear power plant accident scenarios.

  16. Safety issues in robotic handling of nuclear weapon parts

    International Nuclear Information System (INIS)

    Drotning, W.; Wapman, W.; Fahrenholtz, J.

    1993-01-01

    Robotic systems are being developed by the Intelligent Systems and Robotics Center at Sandia National Laboratories to perform automated handling tasks with radioactive weapon parts. These systems will reduce the occupational radiation exposure to workers by automating operations that are currently performed manually. The robotic systems at Sandia incorporate several levels of mechanical, electrical, and software safety for handling hazardous materials. For example, tooling used by the robot to handle radioactive parts has been designed with mechanical features that allow the robot to release its payload only at designated locations in the robotic workspace. In addition, software processes check for expected and unexpected situations throughout the operations. Incorporation of features such as these provides multiple levels of safety for handling hazardous or valuable payloads with automated intelligent systems

  17. Handling missing data in transmission disequilibrium test in nuclear families with one affected offspring.

    Directory of Open Access Journals (Sweden)

    Gulhan Bourget

    Full Text Available The Transmission Disequilibrium Test (TDT compares frequencies of transmission of two alleles from heterozygote parents to an affected offspring. This test requires all genotypes to be known from all members of the nuclear families. However, obtaining all genotypes in a study might not be possible for some families, in which case, a data set results in missing genotypes. There are many techniques of handling missing genotypes in parents but only a few in offspring. The robust TDT (rTDT is one of the methods that handles missing genotypes for all members of nuclear families [with one affected offspring]. Even though all family members can be imputed, the rTDT is a conservative test with low power. We propose a new method, Mendelian Inheritance TDT (MITDT-ONE, that controls type I error and has high power. The MITDT-ONE uses Mendelian Inheritance properties, and takes population frequencies of the disease allele and marker allele into account in the rTDT method. One of the advantages of using the MITDT-ONE is that the MITDT-ONE can identify additional significant genes that are not found by the rTDT. We demonstrate the performances of both tests along with Sib-TDT (S-TDT in Monte Carlo simulation studies. Moreover, we apply our method to the type 1 diabetes data from the Warren families in the United Kingdom to identify significant genes that are related to type 1 diabetes.

  18. Development of commercial robots for radwaste handling

    International Nuclear Information System (INIS)

    Colborn, K.A.

    1988-01-01

    The cost and dose burden associated with low level radwaste handling activities is a matter of increasing concern to the commercial nuclear power industry. This concern is evidenced by the fact that many utilities have begun to revaluate waste generation, handling, and disposal activities at their plants in an effort to improve their overall radwaste handling operations. This paper reports on the project Robots for Radwaste Handling, to identify the potential of robots to improve radwaste handling operations. The project has focussed on the potential of remote or automated technology to improve well defined, recognizable radwaste operations. The project focussed on repetitive, low skill level radwaste handling and decontamination tasks which involve significant radiation exposure

  19. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    C.E. Sanders

    2005-01-01

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for the current design of the CHF and may not reflect the ongoing design evolution of the facility

  20. Hazard classification criteria for non-nuclear facilities

    International Nuclear Information System (INIS)

    Mahn, J.A.; Walker, S.A.

    1997-01-01

    Sandia National Laboratories' Integrated Risk Management Department has developed a process for establishing the appropriate hazard classification of a new facility or operation, and thus the level of rigor required for the associated authorization basis safety documentation. This process is referred to as the Preliminary Hazard Screen. DOE Order 5481.1B contains the following hazard classification for non-nuclear facilities: high--having the potential for onsite or offsite impacts to large numbers of persons or for major impacts to the environment; moderate--having the potential for considerable onsite impacts but only minor offsite impacts to people or the environment; low--having the potential for only minor onsite and negligible offsite impacts to people or the environment. It is apparent that the application of such generic criteria is more than likely to be fraught with subjective judgment. One way to remove the subjectivity is to define health and safety classification thresholds for specific hazards that are based on the magnitude of the hazard, rather than on a qualitative assessment of possible accident consequences. This paper presents the results of such an approach to establishing a readily usable set of non-nuclear facility hazard classifications

  1. Uranium hexafluoride handling

    International Nuclear Information System (INIS)

    1991-01-01

    The United States Department of Energy, Oak Ridge Field Office, and Martin Marietta Energy Systems, Inc., are co-sponsoring this Second International Conference on Uranium Hexafluoride Handling. The conference is offered as a forum for the exchange of information and concepts regarding the technical and regulatory issues and the safety aspects which relate to the handling of uranium hexafluoride. Through the papers presented here, we attempt not only to share technological advances and lessons learned, but also to demonstrate that we are concerned about the health and safety of our workers and the public, and are good stewards of the environment in which we all work and live. These proceedings are a compilation of the work of many experts in that phase of world-wide industry which comprises the nuclear fuel cycle. Their experience spans the entire range over which uranium hexafluoride is involved in the fuel cycle, from the production of UF 6 from the naturally-occurring oxide to its re-conversion to oxide for reactor fuels. The papers furnish insights into the chemical, physical, and nuclear properties of uranium hexafluoride as they influence its transport, storage, and the design and operation of plant-scale facilities for production, processing, and conversion to oxide. The papers demonstrate, in an industry often cited for its excellent safety record, continuing efforts to further improve safety in all areas of handling uranium hexafluoride

  2. Improved Safety Margin Characterization of Risk from Loss of Offsite Power

    International Nuclear Information System (INIS)

    Nelson, Paul

    2017-01-01

    Original intent: The original intent of this task was ''support of the Risk-Informed Safety Margin Characteristic (RISMC) methodology in order'' ''to address ... efficiency of computation so that more accurate and cost-effective techniques can be used to address safety margin characterizations'' (S. M. Hess et al., ''Risk-Informed Safety Margin Characterization,'' Procs. ICONE17, Brussels, July 2009, CD format). It was intended that ''in Task 1 itself this improvement will be directed toward upon the very important issue of Loss of Offsite Power (LOOP) events,'' more specifically toward the challenge of efficient computation of the multidimensional nonrecovery integral that has been discussed by many previous contributors to the theory of nuclear safety. It was further envisioned that ''three different computational approaches will be explored,'' corresponding to the three subtasks listed below; deliverables were tied to the individual subtasks.

  3. Utilities and offsites design baseline. Outside Battery Limits Facility 6000 tpd SRC-I Demonstration Plant. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    None

    1984-05-25

    As part of the overall Solvent Refined Coal (SRC-1) project baseline being prepared by International Coal Refining Company (ICRC), the RUST Engineering Company is providing necessary input for the Outside Battery Limits (OSBL) Facilities. The project baseline is comprised of: design baseline - technical definition of work; schedule baseline - detailed and management level 1 schedules; and cost baseline - estimates and cost/manpower plan. The design baseline (technical definition) for the OSBL Facilities has been completed and is presented in Volumes I, II, III, IV, V and VI. The OSBL technical definition is based on, and compatible with, the ICRC defined statement of work, design basis memorandum, master project procedures, process and mechanical design criteria, and baseline guidance documents. The design basis memorandum is included in Paragraph 1.3 of Volume I. The baseline design data is presented in 6 volumes. Volume I contains the introduction section and utility systems data through steam and feedwater. Volume II continues with utility systems data through fuel system, and contains the interconnecting systems and utility system integration information. Volume III contains the offsites data through water and waste treatment. Volume IV continues with offsites data, including site development and buildings, and contains raw materials and product handling and storage information. Volume V contains wastewater treatment and solid wastes landfill systems developed by Catalytic, Inc. to supplement the information contained in Volume III. Volume VI contains proprietary information of Resources Conservation Company related to the evaporator/crystallizer system of the wastewater treatment area.

  4. Current emergency programs for nuclear installations in Japan

    International Nuclear Information System (INIS)

    Chino, Masamichi

    2007-01-01

    Large effort has been taken for nuclear emergency programs in Japan especially after the JCO accident. A special law for nuclear emergency was established after the accident. The law extended the scope of emergency preparedness to fuel cycle facilities, research reactors, etc. and clarified the roles and responsibilities of the national government, local governments and license holders. For initial responses, the action levels and action procedures are defined based on environmental doses and specific initial events of NPPs. A senior specialist was dispatched to each site for nuclear emergency and a facility 'Off-site center' to be used as the local emergency headquator was designated at each site. This paper describes the structure of emergency program, responsibility of related organizations and the definition of unusual events for notification and emergency. Emergency preparedness, emergency radiation monitoring and computer-based prediction of on- and off-site situation are also addressed. (author)

  5. Safety for fuel assembly handling in the nuclear ship Mutsu

    International Nuclear Information System (INIS)

    Ando, Yoshio

    1978-01-01

    The safety for fuel assembly handling in the nuclear ship Mutsu is deliberated by the committee of general inspection and repair technique examination for Mutsu. The result of deliberation for both cases of removing fuel assemblies and keeping them in the reactor is outlined. The specification of fuel assemblies, and the nuclides and designed radioactivity of fission products of fuel are described. The possibility of shielding repair work and general safety inspection keeping the fuel assemblies in the reactor, the safety consideration when the fuel assemblies are removed at a quay, in a dry dock and on the ocean, the safety of fuel transport in special casks and fuel storage are explained. It is concluded finally that the safety of shielding repair work and general inspection work is secured when the fuel assemblies are kept in the reactor and also when the fuel assemblies are removed from the reactor by cautious working. (Nakai, Y.)

  6. The Off-Site Plowshare and Vela Uniform Programs: Assessing Potential Environmental Liabilities through an Examination of Proposed Nuclear Projects,High Explosive Experiments, and High Explosive Construction Activities Volume 1 of 3

    Energy Technology Data Exchange (ETDEWEB)

    Beck Colleen M,Edwards Susan R.,King Maureen L.

    2011-09-01

    This document presents the results of nearly six years (2002-2008) of historical research and field studies concerned with evaluating potential environmental liabilities associated with U.S. Atomic Energy Commission projects from the Plowshare and Vela Uniform Programs. The Plowshare Program's primary purpose was to develop peaceful uses for nuclear explosives. The Vela Uniform Program focused on improving the capability of detecting, monitoring and identifying underground nuclear detonations. As a result of the Project Chariot site restoration efforts in the early 1990s, there were concerns that there might be other project locations with potential environmental liabilities. The Desert Research Institute conducted archival research to identify projects, an analysis of project field activities, and completed field studies at locations where substantial fieldwork had been undertaken for the projects. Although the Plowshare and Vela Uniform nuclear projects are well known, the projects that are included in this research are relatively unknown. They are proposed nuclear projects that were not executed, proposed and executed high explosive experiments, and proposed and executed high explosive construction activities off the Nevada Test Site. The research identified 170 Plowshare and Vela Uniform off-site projects and many of these had little or no field activity associated with them. However, there were 27 projects that merited further investigation and field studies were conducted at 15 locations.

  7. The Off-Site Plowshare and Vela Uniform Programs: Assessing Potential Environmental Liabilities through an Examination of Proposed Nuclear Projects,High Explosive Experiments, and High Explosive Construction Activities Volume 2 of 3

    Energy Technology Data Exchange (ETDEWEB)

    Beck Colleen M.,Edwards Susan R.,King Maureen L.

    2011-09-01

    This document presents the results of nearly six years (2002-2008) of historical research and field studies concerned with evaluating potential environmental liabilities associated with U.S. Atomic Energy Commission projects from the Plowshare and Vela Uniform Programs. The Plowshare Program's primary purpose was to develop peaceful uses for nuclear explosives. The Vela Uniform Program focused on improving the capability of detecting, monitoring and identifying underground nuclear detonations. As a result of the Project Chariot site restoration efforts in the early 1990s, there were concerns that there might be other project locations with potential environmental liabilities. The Desert Research Institute conducted archival research to identify projects, an analysis of project field activities, and completed field studies at locations where substantial fieldwork had been undertaken for the projects. Although the Plowshare and Vela Uniform nuclear projects are well known, the projects that are included in this research are relatively unknown. They are proposed nuclear projects that were not executed, proposed and executed high explosive experiments, and proposed and executed high explosive construction activities off the Nevada Test Site. The research identified 170 Plowshare and Vela Uniform off-site projects and many of these had little or no field activity associated with them. However, there were 27 projects that merited further investigation and field studies were conducted at 15 locations.

  8. The Off-Site Plowshare and Vela Uniform Programs: Assessing Potential Environmental Liabilities through an Examination of Proposed Nuclear Projects,High Explosive Experiments, and High Explosive Construction Activities Volume 3 of 3

    Energy Technology Data Exchange (ETDEWEB)

    Beck Colleen M.,Edwards Susan R.,King Maureen L.

    2011-09-01

    This document presents the results of nearly six years (2002-2008) of historical research and field studies concerned with evaluating potential environmental liabilities associated with U.S. Atomic Energy Commission projects from the Plowshare and Vela Uniform Programs. The Plowshare Program's primary purpose was to develop peaceful uses for nuclear explosives. The Vela Uniform Program focused on improving the capability of detecting, monitoring and identifying underground nuclear detonations. As a result of the Project Chariot site restoration efforts in the early 1990s, there were concerns that there might be other project locations with potential environmental liabilities. The Desert Research Institute conducted archival research to identify projects, an analysis of project field activities, and completed field studies at locations where substantial fieldwork had been undertaken for the projects. Although the Plowshare and Vela Uniform nuclear projects are well known, the projects that are included in this research are relatively unknown. They are proposed nuclear projects that were not executed, proposed and executed high explosive experiments, and proposed and executed high explosive construction activities off the Nevada Test Site. The research identified 170 Plowshare and Vela Uniform off-site projects and many of these had little or no field activity associated with them. However, there were 27 projects that merited further investigation and field studies were conducted at 15 locations.

  9. Province of Ontario nuclear emergency plan. Pt. 1

    International Nuclear Information System (INIS)

    1986-06-01

    The Province of Ontario Nuclear Emergency Plan has been developed pursuant to Section 8 of the Emergency Plans Act, 1983. This plan replaces the Province of Ontario Nuclear Contingency Off-Site Plan (June 1980) which is no longer applicable. The wastes plan includes planning, preparation, emergency organization and operational responsibilities and policy

  10. Development of operation control expert system for off-site facilities

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, Masaaki

    1988-09-01

    Concerning off-site facilities of oil refinary, changes of facilities and equipment are frequently made in order to cope flexibly with the market trends and changes of the social environment. In addition, it is desirable to introduce computerization into control and manipulation of off-site facilities for its fast, safe and sure operation. In order to achieve the above, against the existing exclusively control-oriented system, it is necessary to add the processing and generating functions to combinations between valves to be shut and piping as well as equipment to be used along the whole extent of the oil flow in the system and to add the function which makes verification of the above functions easy through a dialogue between users and the system. In order to realize the above, Cosmo Oil and Yokokawa Denki developed jointly an operation control expert system for off-site facilities and the system started its actual operation from October 1986. This article is an outline of the system. The result of its actual operation for one and a half years since its inception showed that the system was operated only by the staff responsible for the operation of the facilities, the workload was reduced to 1/3-1/4 of the workload before the adoption of the system and absolutely no omission of work nor mistake was experienced. (2 figs)

  11. Evaluation of Frequency and Restoration time for Loss of Offsite Power events based on domestic operation experience

    International Nuclear Information System (INIS)

    Park, Jin-Hee; Han, Sang-Hoon; Lee, Ho Joong

    2006-01-01

    It is recognized that the availability of AC power to nuclear power plants is essential for safe operation and shutdown and accident recovery of commercial nuclear power plants (NPPs). Unavailability of AC power can be a important adverse impact on a plant's ability to recover accident sequences and maintain safe shutdown. The probabilistic safety assessment (PSA or PRA) performed for Korea NPPs also indicated that a loss of offsite power (LOOP) event and a station blackout (SBO) event can be a important contributors to total risk at nuclear power plant, contributing from 30% to 70% of the total risk at some NPPs in Korea. But, up to now, the LOOP and subsequent restoration time are important inputs to plant probabilistic risk assessment have relied upon foreign data. Therefore, in this paper, the actual LOOP events that have occurred from 1978 to 2004 at commercial nuclear power plants in Korea are collected. A statistical analysis for LOOP frequency and restoration time are performed to apply NPPs's specific and realistic risk model in Korea. Additionally, an engineering analysis is also performed to obtain the insights about the LOOP events

  12. As-Built Verification Plan Spent Nuclear Fuel Canister Storage Building MCO Handling Machine

    International Nuclear Information System (INIS)

    SWENSON, C.E.

    2000-01-01

    This as-built verification plan outlines the methodology and responsibilities that will be implemented during the as-built field verification activity for the Canister Storage Building (CSB) MCO HANDLING MACHINE (MHM). This as-built verification plan covers THE ELECTRICAL PORTION of the CONSTRUCTION PERFORMED BY POWER CITY UNDER CONTRACT TO MOWAT. The as-built verifications will be performed in accordance Administrative Procedure AP 6-012-00, Spent Nuclear Fuel Project As-Built Verification Plan Development Process, revision I. The results of the verification walkdown will be documented in a verification walkdown completion package, approved by the Design Authority (DA), and maintained in the CSB project files

  13. Handling apparatus for a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Shallenberger, J.M.; Hornak, L.P.; Desmarchais, W.E.

    1978-01-01

    An apparatus is disclosed for handling radioactive fuel assembly during transfer operations. The radioactive fuel assembly is drawn up into a shielding sleeve which substantially reduces the level of radioactivity immediately surrounding the sleeve thereby permitting direct access by operating personnel. The lifting assembly which draws the fuel assembly up within the shielding sleeve is mounted to and forms an integral part of the handling apparatus. The shielding sleeve accompanies the fuel assembly during all of the transfer operations

  14. Decision support for off-site emergency preparedness in Europe

    International Nuclear Information System (INIS)

    Kelly, G.N.; Ehrhardt, J.

    1996-01-01

    The decision support system, RODOS, for off-site emergency management in the event of a future accident is being developed with support from the European Commission. The development is being carried out within a large and fully integrated international project involving about forty institutes from sixteen countries in Eastern and Western Europe. RODOS has been designed to provide comprehensive (i.e. applicable at all distances, at all times and to all important countermeasures) decision support and to be applicable throughout Europe. The background to the development of RODOS is described in this paper together with its basic features, its current status and plans for its further development. Given the context of this Special Issue, particular attention is given to the contribution made by institutes in the former Soviet Union to the development of RODOS and plans for its implementation in these countries. The benefits of the system are increasingly being recognised following the completion of the pilot version in 1995. Of particular importance is its potential role as part of a wider European network, the existence of which would promote a more effective and coherent response to any future nuclear accident that might affect Europe. (Author)

  15. Equipment for the handling of thorium materials

    International Nuclear Information System (INIS)

    Heisler, S.W. Jr.; Mihalovich, G.S.

    1988-01-01

    The Feed Materials Production Center (FMPC) is the United States Department of Energy's storage facility for thorium. FMPC thorium handling and overpacking projects ensure the continued safe handling and storage of the thorium inventory until final disposition of the materials is determined and implemented. The handling and overpacking of the thorium materials requires the design of a system that utilizes remote handling and overpacking equipment not currently utilized at the FMPC in the handling of uranium materials. The use of remote equipment significantly reduces radiation exposure to personnel during the handling and overpacking efforts. The design system combines existing technologies from the nuclear industry, the materials processing and handling industry and the mining industry. The designed system consists of a modified fork lift truck for the transport of thorium containers, automated equipment for material identification and inventory control, and remote handling and overpacking equipment for material identification and inventory control, and remote handling and overpacking equipment for repackaging of the thorium materials

  16. Nuclear assay of coal. Volume 6. Mass flow devices for coal handling

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The mass of coal entering the boiler per unit time is an essential parameter for determinig the total rate of heat input. The mass flow rate of coal on a conveyor belt is generally determined as a product of the instantaneous mass of material on a short section of the belt and the belt velocity. Belt loading could be measured by conventional transducers incorporating mechanical or electromechanical weighers or by gamma-ray attenuation gauge. This report reviews the state of the art in mass flow devices for coal handling. The various methods are compared and commented upon. Special design issues are discussed relative to incorporating a mass flow measuring device in a Continuous On-Line Nuclear Analysis of Coal (CONAC) system

  17. Radiological safety aspects of handling plutonium

    International Nuclear Information System (INIS)

    Sundararajan, A.R.

    2016-01-01

    Department of Atomic Energy in its scheme of harnessing the nuclear energy for electrical power generation and strategic applications has given a huge role to utilization of plutonium. In the power production programme, fast reactors with plutonium as fuel are expected to play a major role. This would require establishing fuel reprocessing plants to handle both thermal and fast reactor fuels. So in the nuclear fuel cycle facilities variety of chemical, metallurgical, mechanical operations have to be carried out involving significant inventories of "2"3"9 Pu and associated radionuclides. Plutonium is the most radiotoxic radionuclide and therefore any facility handling it has to be designed and operated with utmost care. Two problems of major concern in the protection of persons working in plutonium handling facilities are the internal exposure to the operating personnel from uptake of plutonium and transplutonic nuclides as they are highly radiotoxic and the radiation exposure of hands and eye lens during fuel fabrication operations especially while handling recycled high burn up plutonium. In view of the fact that annual limit for intake is very small for "2"3"9Pu and its radiation emission characteristics are such that it is a huge challenge for the health physicists to detect Pu in air and in workers. This paper discusses the principles and practices followed in providing radiological surveillance to workers in plutonium handling areas. The challenges in protecting the workers from receiving exposures to hands and eye lens in handling high burn up plutonium are also discussed. The sites having Pu fuel cycle facilities should have trained medical staff to handle cases involving excessive intake of plutonium. (author)

  18. Evaluation on Cooling Performance of Containment Fan Cooler during Design Basis Accident with Loss of Offsite Power for Kori 3 and 4 Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sung Bok; Lee, Sang Won [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Park, Young Chan [Atomic Creative Technology Co., LTD., Daejeon (Korea, Republic of)

    2007-10-15

    The purpose of this study is to evaluate cooling performance of containment fan cooler units and to review a technical background related to Generic Letter 96-06. In case that design basis accident (DBA) and loss of offsite power (LOOP) occurs, component cooling water (CCW) pumps cannot provide the cooling water source to fan cooler units while fan coolers coast down. Fan cooler units and CCW pumps are restarted by emergency diesel generator (EDG) operation and it takes about 30 seconds. In this scenario, before the EDG restarts and CCW flowrate is restored, heated air in the containment passes through coil of fan cooler units without cooling water source. In this situation, the boiling of water in the fan cooler units may occur. Restarting of CCW pumps may bring about condensation by injected cooling water and water hammer may occur. This thermal-hydraulic effect is sensitive to system configuration, i.e system pressure, containment pressure/temperature, EDG restarting time, etc. In this study, the evaluation of containment fan cooler units was performed for Kori 3 and 4 nuclear power plant.

  19. PND fuel handling decontamination: facilities and techniques

    International Nuclear Information System (INIS)

    Pan, R.Y.

    1996-01-01

    The use of various decontamination techniques and equipment has become a critical part of Fuel Handling maintenance work at Ontario Hydro's Pickering Nuclear Division. This paper presents an overview of the set up and techniques used for decontamination in the PND Fuel Handling Maintenance Facility and the effectiveness of each. (author). 1 tab., 9 figs

  20. The real-time on-line decision support system RODOS for off-site emergency management

    International Nuclear Information System (INIS)

    Benz, G.; Ehrhardt, J.; Faude, D.; Fischer, F.; Paesler-Sauer, J.; Rafat, M.; Schichtel, T.; Schuele, O.; Steinhauer, C.

    1994-01-01

    The project RODOS aims at the development of an integrated and comprehensive real-time on-line decision support system for off-site emergency management of nuclear accidents in Europe, applicable to the vicinity of the accident and its early phase up to far distant areas and later stages. As a joint venture of 18 institutions in the European Union, in close cooperation with institutions in Russia, Belarus, Ukraine and east-central Europe, the project is to provide the methodological basis, develop models, collect data, and install the hardware and software framework of the RODOS system. The paper describes the project status, the overall design of the system and its present software structure, and gives a brief overview of important research and development tasks for the next project phases. (orig.) [de

  1. International handling of fissionable material

    International Nuclear Information System (INIS)

    1975-01-01

    The opinion of the ministry for foreign affairs on international handling of fissionable materials is given. As an introduction a survey is given of the possibilities to produce nuclear weapons from materials used in or produced by power reactors. Principles for international control of fissionable materials are given. International agreements against proliferation of nuclear weapons are surveyed and methods to improve them are proposed. (K.K.)

  2. Examination of off-site emergency protective measures for core melt accidents

    International Nuclear Information System (INIS)

    Aldrich, D.C.; Ericson, D.M. Jr.; Jones, R.B.

    1978-01-01

    Results from the Reactor Safety Study (RSS) have shown that to cause significant impacts off-site, i.e., sufficient quantities of biologically important radionuclides released, it is necessary to have a core melt accident. To mitigate the impact of such potential accidents, the design of appropriate emergency response actions requires information as to the relative merit of publicly available protective measures. In order to provide this information, a study using the consequence model developed for the RSS is being conducted to evaluate (in terms of reduced public health effects and dose exposure) potential off-site protective strategies. The paper describes the methods being used in the study as well as the results and conclusions obtained

  3. Off-site intervention plan of the public health authorities for emergencies at the Caorso nuclear power plant

    International Nuclear Information System (INIS)

    Fabbri, S.; Sogni, R.; Boeri, G.; Cencetti, S.; Melandri, G.; Paterlini, L.

    1986-01-01

    The Caorso nuclear power plant, which is near Piacenza and has an 875 MW boiling water reactor, has been generating electricity on a regular basis since 1978. The off-site intervention plan of the public health authorities, based on an analysis of hypothetical accidents, was approved in 1977 and subsequently revised. A study of the radiological consequences of these accidents for man and the environment indicates that the highest doses likely to be received by inhalation of 131 I would be no more than a few rem, whereas the levels of soil contamination, even at a distance of 40 km, could exceed 1 μCi/m 2 . The main problems caused by such accidents are therefore related to environmental contamination. Under the intervention plan, the provincial prefect is responsible for co-ordinating the work of all the civil, military and medical authorities. Teams from ENEL, the fire service and the local health services (USL) monitor the concentration of 131 I in the air and the exposure level, and take samples within a radius of 10 km around the site. The police and army control road traffic and are responsible, if necessary, for the evacuation of the population. A radiometry co-ordination centre (CCRI) is set up to process the readings made by the teams and provides the prefect with the technical information he requires to take decisions The local medical services (USL) run the centre where the population is assembled and monitor superficial contamination, apply initial decontamination measures, and provide medical assistance in general. The Piacenza USL also takes action by sending out its mobile radiometry laboratory, using its measurement equipment and providing logistic support to the CCRI. It is also to play a major role in informing the population in the event of an alarm. (author)

  4. Handling and final disposal of nuclear waste. Programme for research development and other measures

    International Nuclear Information System (INIS)

    1989-09-01

    The report is divided into two parts. Part 1 presents the premises for waste management in Sweden and the waste types that are produced in Sweden. A brief description is then provided of the measures required for the handling and disposal of the various waste forms. An account of measures for decommissioning of nuclear power plants is also included. Part 2 describes the research program for 1990-1995, which includes plans for siting, repository design; studies of rock properties and chemistry, biosphere, technological barriers. Activities within two large projects, the Stripa laboratory and Natural analogues are also described. 240 refs. 40 figs

  5. Management of radioactive waste in nuclear power: handling of irradiated graphite from water-cooled graphite reactors

    International Nuclear Information System (INIS)

    Anfimov, S.S.

    2000-01-01

    As a result of decommissioning of water-cooled graphite-moderated reactors, a large amount of rad-waste in the form of graphite stack fragments is generated (on average 1500-2000 tons per reactor). That is why it is essentially important, although complex from the technical point of view, to develop advanced technologies based on up-to-date remotely-controlled systems for unmanned dismantling of the graphite stack containing highly-active long-lived radionuclides and for conditioning of irradiated graphite (IG) for the purposes of transportation and subsequent long term and ecologically safe storage either on NPP sites or in special-purpose geological repositories. The main characteristics critical for radiation and nuclear hazards of the graphite stack are as follows: the graphite stack is contaminated with nuclear fuel that has gotten there as a result of the accidents; the graphite mass is 992 tons, total activity -6?104 Ci (at the time of unit shutdown); the fuel mass in the reactor stack amounts to 100-140 kg, as estimated by IPPE and RDIPE, respectively; γ-radiation dose rate in the stack cells varies from 4 to 4300 R/h, with the prevailing values being in the range from 50 to 100 R/h. In this paper the traditional methods of rad-waste handling as bituminization technology, cementing technology are discussed. In terms of IG handling technology two lines were identified: long-term storage of conditioned IG and IG disposal by means of incineration. The specific cost of graphite immobilization in a radiation-resistant polymeric matrix amounts to -2600 USD per 1 t of graphite, whereas the specific cost of immobilization in slag-stone containers with an inorganic binder (cement) is -1400 USD per 1 t of graphite. On the other hand, volume of conditioned IG rad-waste subject for disposal, if obtained by means of the first technology, is 2-2.5 times less than the volume of rad-waste generated by means of the second technology. It can be concluded from the above that

  6. Open-Phase Condition Detecting System for Transformers in Nuclear Power Plant

    International Nuclear Information System (INIS)

    Ha, Che-Wung; Lee, Do-Hwan

    2015-01-01

    Recently, several events involving the loss of one of the three phases of the offsite power circuit occurred in the US nuclear power plants (NPPs).. In some cases, the open-phase condition existed undetected for an extended period and in other case, was not properly responded to. Accordingly, the Nuclear Regulatory Commission (NRC) requested all license holders to take corrective actions to address the open-phase condition. It was also requested that all holders or applicant for a standard design certification (DC) include a description of a protection system to detect and separate the open circuit into design control document (DCD). Currently, NPPs including Duke Energy, Exelon, and institutes including Electric Power Research Institute (EPRI) are working together to resolve issues associated with detecting an open-phase condition. This paper, using Electromagnetic Transients Program (EMTP), presents a system to detect and address the loss of one of three phases of the offsite power circuit connected to main, auxiliary and standby transformers, which is hard to be detected in the current protection system. This paper, using EMTP, presents a system to detect and address the loss of one of three phases of the offsite power circuit running to MT, UAT or SAT which is hard to be detected in the current protection system. The system presented in this paper will be useful not only for the KHNP to meet the NRC requirement, but also for nuclear power plants at home and abroad to take corrective actions to provide protection from a single phase open circuit condition for offsite power sources

  7. Open-Phase Condition Detecting System for Transformers in Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Che-Wung; Lee, Do-Hwan [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Recently, several events involving the loss of one of the three phases of the offsite power circuit occurred in the US nuclear power plants (NPPs).. In some cases, the open-phase condition existed undetected for an extended period and in other case, was not properly responded to. Accordingly, the Nuclear Regulatory Commission (NRC) requested all license holders to take corrective actions to address the open-phase condition. It was also requested that all holders or applicant for a standard design certification (DC) include a description of a protection system to detect and separate the open circuit into design control document (DCD). Currently, NPPs including Duke Energy, Exelon, and institutes including Electric Power Research Institute (EPRI) are working together to resolve issues associated with detecting an open-phase condition. This paper, using Electromagnetic Transients Program (EMTP), presents a system to detect and address the loss of one of three phases of the offsite power circuit connected to main, auxiliary and standby transformers, which is hard to be detected in the current protection system. This paper, using EMTP, presents a system to detect and address the loss of one of three phases of the offsite power circuit running to MT, UAT or SAT which is hard to be detected in the current protection system. The system presented in this paper will be useful not only for the KHNP to meet the NRC requirement, but also for nuclear power plants at home and abroad to take corrective actions to provide protection from a single phase open circuit condition for offsite power sources.

  8. Analysis of HFETR shut-down state caused by loss of off-site power supply

    International Nuclear Information System (INIS)

    Wang Jinghu

    1997-01-01

    During the last 15 years, there are more than 40 unplanned shut-downs caused by loss of off-site power in HFETR. Because HFETR is a special research reactor, the author describes the shut-down state as three period. The author also discusses the influence of the number of shut-down due to loss of off-site power supply on the reactor safety, and propose some suggestions and measures to reduce the effects

  9. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2005-04-07

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for

  10. PND fuel handling decontamination: facilities and techniques

    Energy Technology Data Exchange (ETDEWEB)

    Pan, R Y [Ontario Hydro, Toronto, ON (Canada)

    1997-12-31

    The use of various decontamination techniques and equipment has become a critical part of Fuel Handling maintenance work at Ontario Hydro`s Pickering Nuclear Division. This paper presents an overview of the set up and techniques used for decontamination in the PND Fuel Handling Maintenance Facility and the effectiveness of each. (author). 1 tab., 9 figs.

  11. CLASSIFICATION OF THE MGR OFFSITE UTILITIES SYSTEM

    International Nuclear Information System (INIS)

    J.A. Ziegler

    1999-01-01

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) offsite utilities system structures, systems and components (SSCs) performed by the MGR Safety Assurance Department. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 1998). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333P, ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998)

  12. Principles of preparation of traumatological plans for nuclear facilities and workplaces handling ionizing radiation sources. Recommendations

    International Nuclear Information System (INIS)

    1999-01-01

    The publication encompasses the following topics: (i) Czech legislative framework of emergency preparedness; (ii) Medical care during accidents associated with immediate health consequences (Accidents during ionizing radiation source handling; Physician's action during the treatment of excessively exposed persons; Equipment of emergency site providing first aid during contamination with radioactive substances; Guidelines for the collection of stool and urine); (iii) Licensee's procedures during events at workplaces and principles of organization of medical help; and (iv) Provisions to protect the public in the event of accidents associated with radioactivity leaks into the environment. It is envisaged that this document ('Recommendations') will form a basis for the final Safety Guides to be issued by the State Office for Nuclear Safety, the Czech nuclear regulatory authority. (P.A.)

  13. Off-site training of laparoscopic skills, a scoping review using a thematic analysis.

    Science.gov (United States)

    Thinggaard, Ebbe; Kleif, Jakob; Bjerrum, Flemming; Strandbygaard, Jeanett; Gögenur, Ismail; Matthew Ritter, E; Konge, Lars

    2016-11-01

    The focus of research in simulation-based laparoscopic training has changed from examining whether simulation training works to examining how best to implement it. In laparoscopic skills training, portable and affordable box trainers allow for off-site training. Training outside simulation centers and hospitals can increase access to training, but also poses new challenges to implementation. This review aims to guide implementation of off-site training of laparoscopic skills by critically reviewing the existing literature. An iterative systematic search was carried out in MEDLINE, EMBASE, ERIC, Scopus, and PsychINFO, following a scoping review methodology. The included literature was analyzed iteratively using a thematic analysis approach. The study was reported in accordance with the STructured apprOach to the Reporting In healthcare education of Evidence Synthesis statement. From the search, 22 records were identified and included for analysis. A thematic analysis revealed the themes: access to training, protected training time, distribution of training, goal setting and testing, task design, and unsupervised training. The identified themes were based on learning theories including proficiency-based learning, deliberate practice, and self-regulated learning. Methods of instructional design vary widely in off-site training of laparoscopic skills. Implementation can be facilitated by organizing courses and training curricula following sound education theories such as proficiency-based learning and deliberate practice. Directed self-regulated learning has the potential to improve off-site laparoscopic skills training; however, further studies are needed to demonstrate the effect of this type of instructional design.

  14. Offsite commercial disposal of oil and gas exploration and production waste :availability, options, and cost.

    Energy Technology Data Exchange (ETDEWEB)

    Puder, M. G.; Veil, J. A.

    2006-09-05

    A survey conducted in 1995 by the American Petroleum Institute (API) found that the U.S. exploration and production (E&P) segment of the oil and gas industry generated more than 149 million bbl of drilling wastes, almost 18 billion bbl of produced water, and 21 million bbl of associated wastes. The results of that survey, published in 2000, suggested that 3% of drilling wastes, less than 0.5% of produced water, and 15% of associated wastes are sent to offsite commercial facilities for disposal. Argonne National Laboratory (Argonne) collected information on commercial E&P waste disposal companies in different states in 1997. While the information is nearly a decade old, the report has proved useful. In 2005, Argonne began collecting current information to update and expand the data. This report describes the new 2005-2006 database and focuses on the availability of offsite commercial disposal companies, the prevailing disposal methods, and estimated disposal costs. The data were collected in two phases. In the first phase, state oil and gas regulatory officials in 31 states were contacted to determine whether their agency maintained a list of permitted commercial disposal companies dedicated to oil. In the second stage, individual commercial disposal companies were interviewed to determine disposal methods and costs. The availability of offsite commercial disposal companies and facilities falls into three categories. The states with high oil and gas production typically have a dedicated network of offsite commercial disposal companies and facilities in place. In other states, such an infrastructure does not exist and very often, commercial disposal companies focus on produced water services. About half of the states do not have any industry-specific offsite commercial disposal infrastructure. In those states, operators take their wastes to local municipal landfills if permitted or haul the wastes to other states. This report provides state-by-state summaries of the

  15. Nuclear nonproliferation: Concerns with US delays in accepting foregin research reactors' spent fuel

    International Nuclear Information System (INIS)

    1994-01-01

    One key US nonproliferation goal is to discourage use of highly enriched uranium fuel (HEU), which can be used to make nuclear bombs, in civilian nuclear programs worldwide. DOE's Off-Site Fuels Policy for taking back spent HEU from foreign research reactors was allowed to expire due to environmental reasons. This report provides information on the effects of delays in renewing the Off-Site Fuels Policy on US nonproliferation goals and programs (specifically the reduced enrichment program), DOE's efforts to renew the fuels policy, and the price to be charged to the operators of foreign reactors for DOE's activities in taking back spent fuel

  16. Research from Afar: Considerations for Conducting an Off-Site Research Project.

    Science.gov (United States)

    Williams, Reg Arthur; Hagerty, Bonnie M.; Hoyle, Kenneth; Yousha, Steven M.; Abdoo, Yvonne; Andersen, Curt; Engler, Dorothy

    1999-01-01

    Critical elements in the success of off-site research projects include the following: negotiation, attention to personnel issues, communication, participation of research subjects, data management, and concern for privacy issues. (SK)

  17. Analysis of parameters for the off-site dose calculation due to HTO, OBT, and radioactive carbon ingestion

    International Nuclear Information System (INIS)

    Lee, G. B.; Jeung, Y. K.; Bang, S. Y.; Um, H. M.

    2004-01-01

    For assessment of tritium and radiocarbon ingestion dose to off site individuals, water, hydrogen, and carbon content of main farm produce of Korea were investigated to replace the existing data in K-DOSE60, the Offsite Dose Calculation Manual(ODCM) of Korea Hydro and Nuclear Power Co. Ltd. (KHNP). Main items and weighting factors of farm produce were determined with the nationwide food intake data in 2001, 2002. Main farm produce were sampled around Kori, Wolsong, Ulchin, Yonggwang nuclear power sites. Content of each produce was multiplied by weighting factor and summed up to make the weighted mean group value. For grains, water, hydrogen, and carbon content was not much different from the existing data currently used in K-DOSE60, but root vegetables had 3.5 times more hydrogen, and leafy vegetables and fruits had 0.7 - 1.3 times more or less water, hydrogen, and carbon contents than K-DOSE60

  18. Uranium hexafluoride: A manual of good handling practices. Revision 7

    International Nuclear Information System (INIS)

    1995-01-01

    The United States Enrichment Corporation (USEC) is continuing the policy of the US Department of Energy (DOE) and its predecessor agencies in sharing with the nuclear industry their experience in the area of uranium hexafluoride (UF 6 ) shipping containers and handling procedures. The USEC has reviewed Revision 6 or ORO-651 and is issuing this new edition to assure that the document includes the most recent information on UF 6 handling procedures and reflects the policies of the USEC. This manual updates the material contained in earlier issues. It covers the essential aspects of UF 6 handling, cylinder filling and emptying, general principles of weighing and sampling, shipping, and the use of protective overpacks. The physical and chemical properties of UF 6 are also described. The procedures and systems described for safe handling of UF 6 presented in this document have been developed and evaluated during more than 40 years of handling vast quantities of UF 6 . With proper consideration for its nuclear properties, UF 6 may be safely handled in essentially the same manner as any other corrosive and/or toxic chemical

  19. Encapsulation and handling of spent nuclear fuel for final disposal

    International Nuclear Information System (INIS)

    Loennerberg, B.; Larker, H.; Ageskog, L.

    1983-05-01

    The handling and embedding of those metal parts which arrive to the encapsulation station with the fuel is described. For the encapsulation of fuel two alternatives are presented, both with copper canisters but with filling of lead and copper powder respectively. The sealing method in the first case is electron beam welding, in the second case hot isostatic pressing. This has given the headline of the two chapters describing the methods: Welded copper canister and Pressed copper canister. Chapter 1, Welded copper canister, presents the handling of the fuel when it arrives to the encapsulation station, where it is first placed in a buffer pool. From this pool the fuel is transferred to the encapsulation process and thereby separated from fuel boxes and boron glass rod bundles, which are transported together with the fuel. The encapsulation process comprises charging into a copper canister, filling with molten lead, electron beam welding of the lid and final inspection. The transport to and handling in the final repository are described up to the deposition and sealing in the deposition hole. Handling of fuel residues is treated in one of the sections. In chapter 2, Pressed copper canister, only those parts of the handling, which differ from chapter 1 are described. The hot isostatic pressing process is given in the first sections. The handling includes drying, charging into the canister, filling with copper powder, seal lid application and hot isostatic pressing before the final inspection and deposition. In the third chapter, BWR boxes in concrete moulds, the handling of the metal parts, separated from the fuel, are dealt with. After being lifted from the buffer pool they are inserted in a concrete mould, the mould is filled with concrete, covered with a lid and after hardening transferred to its own repository. The deposition in this repository is described. (author)

  20. Annual report of Nuclear Emergency Assistance and Training Center. April 1, 2008 - March 31, 2009

    International Nuclear Information System (INIS)

    Kanamori, Masashi; Hashimoto, Kazuichiro; Terunuma, Hiroshi; Ikeda, Takeshi; Ohmura, Akiko; Terakado, Naoya; Nagakura, Tomohiro; Fukumoto, Masahiro; Watanabe, Fumitaka; Yamamoto, Kazuya; Abe, Minako; Kikuchi, Masayuki; Sumiya, Akihiro; Matsusaka, Masaru

    2009-09-01

    When a nuclear emergency occurs in Japan, the Japan Atomic Energy Agency (JAEA) provides technical support to the National government, local governments, police, fire station and license holder etc. They are Designated Public Organizations conforming to the Basic Law on Emergency Preparedness and the Basic Plan for Disaster Countermeasures. The Nuclear Emergency Assistance and Training Center (NEAT) of JAEA provides a comprehensive range of technical support activities to an Off-Site Center in case of a nuclear emergency. Specifically, NEAT gives technical advice and information, provides for the dispatch of specialist as required, supplies emergency equipments and materials to the Joint Council of Nuclear Disaster Countermeasures, which meets at the Off-Site Center. NEAT provide various lectures and training course concerning nuclear disaster prevention for those personnel taking an active part in emergency response organizations at normal time. And NEAT researches on nuclear disaster prevention and also cooperate with international organizations. This annual report summarized the activities of JAEA/NEAT in the fiscal year 2008. (author)

  1. Planning for off-site response to radiation accidents in nuclear facilities

    International Nuclear Information System (INIS)

    1981-01-01

    The purpose of this publication is to give guidance to those who are responsible for the protection of the public in the event of an accident occurring at a land-based nuclear facility. This guidance should assist in the advance preparation of emergency response plans and implementing procedures. Basic principles of protective measures along with their advantages and disadvantages are discussed. Other principles related to emergency planning and the operational response to an emergency are outlined. Although the guidance is primarily oriented towards land-based nuclear power facilities, the guidance does have general application to other types of nuclear facility

  2. Planning for off-site response to radiation accidents in nuclear facilities

    International Nuclear Information System (INIS)

    1979-01-01

    The purpose of this manual is to give guidance to those who are responsible for the protection of the public in the event of an accident occurring at a land-based nuclear facility. This guidance should assist in the advance preparation of emergency response plans and implementing procedures. Basic principles of protective measures along with their advantages and disadvantages are discussed. Other principles related to emergency planning and the operational response to an emergency are outlined. Although the guidance is primarily oriented toward land-based nuclear power facilities, the guidance does have general application to other types of nuclear facilities

  3. Design of simulation-based medical education and advantages and disadvantages of in situ simulation versus off-site simulation

    NARCIS (Netherlands)

    Sorensen, J.L.; Ostergaard, D.; Leblanc, V.; Ottesen, B.; Konge, L.; Dieckmann, P.; Vleuten, C. van der

    2017-01-01

    BACKGROUND: Simulation-based medical education (SBME) has traditionally been conducted as off-site simulation in simulation centres. Some hospital departments also provide off-site simulation using in-house training room(s) set up for simulation away from the clinical setting, and these activities

  4. A MGy radiation-hardened sensor instrumentation link for nuclear reactor monitoring and remote handling

    Energy Technology Data Exchange (ETDEWEB)

    Verbeeck, Jens; Cao, Ying [KU Leuven - KUL, Div. LRD-MAGyICS, Kasteelpark Arenberg 10, 3001 Heverlee (Belgium); Van Uffelen, Marco; Mont Casellas, Laura; Damiani, Carlo; Morales, Emilio Ruiz; Santana, Roberto Ranz [Fusion for Energy - F4E, c/Josep,n deg. 2, Torres Diagonal Litoral, Ed. B3, 08019 Barcelona (Spain); Meek, Richard; Haist, Bernhard [Oxford Technologies Ltd. OTL, 7 Nuffield Way, Abingdon OX14 1RL (United Kingdom); De Cock, Wouter; Vermeeren, Ludo [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Steyaert, Michiel [KU Leuven, ESAT-MICAS, KasteelparkArenberg 10, 3001 Heverlee (Belgium); Leroux, Paul [KU Leuven, ESAT-MICAS, KasteelparkArenberg 10, 3001 Heverlee (Belgium)

    2015-07-01

    Decommissioning, dismantling and remote handling applications in nuclear facilities all require robotic solutions that are able to survive in radiation environments. Recently raised safety, radiation hardness and cost efficiency demands from both the nuclear regulatory and the society impose severe challenges in traditional methods. For example, in case of the dismantling of the Fukushima sites, solutions that survive accumulated doses higher than 1 MGy are mandatory. To allow remote operation of these tools in nuclear environments, electronics were used to be shielded with several centimeters of lead or even completely banned in these solutions. However, shielding electronics always leads to bulky and heavy solutions, which reduces the flexibility of robotic tools. It also requires longer repair time and produces extra waste further in a dismantling or decommissioning cycle. In addition, often in current reactor designs, due to size restrictions and the need to inspect very tight areas there are limitations to the use of shielding. A MGy radiation-hardened sensor instrumentation link developed by MAGyICS provides a solution to build a flexible, easy removable and small I and C module with MGy radiation tolerance without any shielding. Hereby it removes all these pains to implement electronics in robotic tools. The demonstrated solution in this poster is developed for ITER Remote Handling equipments operating in high radiation environments (>1 MGy) in and around the Tokamak. In order to obtain adequately accurate instrumentation and control information, as well as to ease the umbilical management, there is a need of front-end electronics that will have to be located close to those actuators and sensors on the remote handling tool. In particular, for diverter remote handling, it is estimated that these components will face gamma radiation up to 300 Gy/h (in-vessel) and a total dose of 1 MGy. The radiation-hardened sensor instrumentation link presented here, consists

  5. Potential problems from shipment of high-curie content contact-handled transuranic (CH-TRU) waste to WIPP

    International Nuclear Information System (INIS)

    Neill, R.H.; Channell, J.K.

    1983-08-01

    There are about 1000 drums of contact-handled transuranic (CH-TRU) wastes containing more than 100 Ci/drum of Pu-238 that are stored at the Savannah River Plant and at the Los Alamos National Laboratory. Studies performed at DOE laboratories have shown that large quantities of gases are generated in stored drums containing 100 Ci of 238 Pu. Concentrations of hydrogen gas in the void space of the drums are often found to be high enough to be explosive. None of the analyses in the DOE WIPP Final Environmental Impact Statement, Safety Analysis Report, and Preliminary Transportation Analysis have considered the possibility that the generation of hydrogen gas by radiolysis may create an explosive or flammable hazard that could increase the frequency and severity of accidental releases of radionuclides during transportation or handling. These high 238 Pu concentration containers would also increase the estimated doses received by individuals and populations from transportation, WIPP site operations, and human intrusion scenarios even if the possibility of gas-enhanced releases is ignored. The WIPP Project Office has evaluated this effect on WIPP site operations and is suggesting a maximum limit of 140 239 Pu equivalent curies (P-Ci) per drum so that postulated accidental off-site doses will not be larger than those listed in the FEIS. The TRUPACT container, which is being designed for the transportation of CH-TRU wastes to WIPP, does not appear to meet the Nuclear Regulatory Commission regulations requiring double containment for the transportation of plutonium in quantities >20 Ci. A 20 alpha Ci/shipment limit would require about 200,000 shipments for the 4 million curies of alpha emitters slated for WIPP

  6. Operation of Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Tossavainen, K.

    1992-09-01

    The Finnish nuclear power plant units Loviisa 1 and 2 as well as TVO I and II were in operation for almost all the time in the first quarter of 1992. The load factor average was 99.8%. All events which are classified on the International Nuclear Event Scale were level 0/below scale on the Scale. Occupational radiation doses and releases of radioactive material off-site remained well below authorised limits. Only quantities of radioactive material insignificant to radiation exposure, originating from the nuclear power plants, were detected in samples collected in the vicinity of the nuclear power plants

  7. Tritium handling experience at Atomic Energy of Canada Limited

    Energy Technology Data Exchange (ETDEWEB)

    Suppiah, S.; McCrimmon, K.; Lalonde, S.; Ryland, D.; Boniface, H.; Muirhead, C.; Castillo, I. [Atomic Energy of Canad Limited - AECL, Chalk River Laboratories, Chalk River, ON (Canada)

    2015-03-15

    Canada has been a leader in tritium handling technologies as a result of the successful CANDU reactor technology used for power production. Over the last 50 to 60 years, capabilities have been established in tritium handling and tritium management in CANDU stations, tritium removal processes for heavy and light water, tritium measurement and monitoring, and understanding the effects of tritium on the environment. This paper outlines details of tritium-related work currently being carried out at Atomic Energy of Canada Limited (AECL). It concerns the CECE (Combined Electrolysis and Catalytic Exchange) process for detritiation, tritium-compatible electrolysers, tritium permeation studies, and tritium powered batteries. It is worth noting that AECL offers a Tritium Safe-Handling Course to national and international participants, the course is a mixture of classroom sessions and hands-on practical exercises. The expertise and facilities available at AECL is ready to address technological needs of nuclear fusion and next-generation nuclear fission reactors related to tritium handling and related issues.

  8. 77 FR 58416 - Comparative Environmental Evaluation of Alternatives for Handling Low-Level Radioactive Waste...

    Science.gov (United States)

    2012-09-20

    ... for Handling Low-Level Radioactive Waste Spent Ion Exchange Resins From Commercial Nuclear Power... Radioactive Waste Spent Ion Exchange Resins from Commercial Nuclear Power Reactors. DATES: Please submit... Evaluation of Alternatives for Handling Low-Level Radioactive Waste Spent Ion Exchange Resins from Commercial...

  9. Estimation of the loss of Offsite power frequency for the probabilistic safety assessment of the Juragua NPP

    International Nuclear Information System (INIS)

    Vilaragut Llanes, J.J.; Valhuerdi Debesa, C.

    1996-01-01

    The loss offsite power is defined as the interruption of the preferred power supply to the essential and non essential switchgear buses necessitating or resulting in the use of emergency AC power supply. Because many safety system required for reactor core decay heat removal and containment heat removal depend on AC power, a loss of offsite power, if emergency power supply (diesel generators) fails, could be severe accidents The purpose of this work was to determine, for the Probabilistic Safety Assessment of the Juragua NPP, the causes, frequency and duration relationships of the loss of offsite power. A description is presented of the different factor that determine the occurrence of this event and the characteristics for the Juragua NPP

  10. Environmental monitoring program for radiological emergencies at the Almirante Alvaro Alberto Nuclear Power Plant, Angra dos Reis, Rio de Janeiro, Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, Lilia M.J. Belem; Ramos Junior, Anthenor C.; Gomes, Carlos A.; Carvalho, Zenildo L.; Gouveia, Vandir; Estrada, Julio; Ney, Cezar [Instituto de Radioprotecao e Dosimetria (IRD), Rio de Janeiro, RJ (Brazil)

    1997-12-31

    In order to respond to a major radiological emergency at the Almirante Alvaro Alberto Nuclear Power Plant, located in Angra do Reis, Rio de Janeiro, the Emergency Response Team of the Institute of Radiation Protection and Dosimetry (IRD/CNEN) established a program of environmental monitoring. A monitoring trend to assess the off-site radiological conditions and give support to decision making for implementing protective measure in case of a radiological accident is presented. For the selection of the monitoring points, the program takes into account atmospheric diffusion, population conglomerates and their habits, water and land use; it includes the entire Emergency Planning Zone of 15 km radius. The program has been organized in the form of a handbook to facilitate handling by field teams 2 refs., 1 fig.; e-mail: lilia at ird.gov.br

  11. PCTRAN enhancement for large break loss of coolant accident concurrent with loss of offsite power in VVER-1000 simulation

    Energy Technology Data Exchange (ETDEWEB)

    Hadad, Kamal; Esmaeili-Sanjavanmareh, Mansour [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Dept. of Nuclear Engineering

    2017-05-15

    PCTRAN capability to simulate a large break loss of coolant accident concurrent with the loss of offsite power in Bushehr Nuclear Power Plant is enhanced and investigated. Following the correction of the accident scenario for Bushehr nuclear power plant in PCTRAN, simulation results are compared with the final safety assessment report of that plant. As a result, the primary loop thermal hydraulics parameters including pressure, total flow rates, leakage flow rates and reactor power are in a good agreement with the reference data. Hot and cold leg temperature variations have the same trends as reference data but have a maximum of 80 C disagreement at the transient initiation. The reason for this disagreement is explained and its adjustment is discussed. Improvements of PCTRAN simulator are mainly due to enhancing user control for atmospheric steam dump valve, containment pressure and emergency core cooling systems which are thoroughly described in this paper.

  12. Mission Need Statement for the Idaho National Laboratory Remote-Handled Low-Level Waste Disposal Project

    International Nuclear Information System (INIS)

    Harvego, Lisa

    2009-01-01

    The Idaho National Laboratory proposes to establish replacement remote-handled low-level waste disposal capability to meet Nuclear Energy and Naval Reactors mission-critical, remote-handled low-level waste disposal needs beyond planned cessation of existing disposal capability at the end of Fiscal Year 2015. Remote-handled low-level waste is generated from nuclear programs conducted at the Idaho National Laboratory, including spent nuclear fuel handling and operations at the Naval Reactors Facility and operations at the Advanced Test Reactor. Remote-handled low-level waste also will be generated by new programs and from segregation and treatment (as necessary) of remote-handled scrap and waste currently stored in the Radioactive Scrap and Waste Facility at the Materials and Fuels Complex. Replacement disposal capability must be in place by Fiscal Year 2016 to support uninterrupted Idaho operations. This mission need statement provides the basis for the laboratory's recommendation to the Department of Energy to proceed with establishing the replacement remote-handled low-level waste disposal capability, project assumptions and constraints, and preliminary cost and schedule information for developing the proposed capability. Without continued remote-handled low-level waste disposal capability, Department of Energy missions at the Idaho National Laboratory would be jeopardized, including operations at the Naval Reactors Facility that are critical to effective execution of the Naval Nuclear Propulsion Program and national security. Remote-handled low-level waste disposal capability is also critical to the Department of Energy's ability to meet obligations with the State of Idaho

  13. Uranium hexafluoride: A manual of good handling practices. Revision 7

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-01-01

    The United States Enrichment Corporation (USEC) is continuing the policy of the US Department of Energy (DOE) and its predecessor agencies in sharing with the nuclear industry their experience in the area of uranium hexafluoride (UF{sub 6}) shipping containers and handling procedures. The USEC has reviewed Revision 6 or ORO-651 and is issuing this new edition to assure that the document includes the most recent information on UF{sub 6} handling procedures and reflects the policies of the USEC. This manual updates the material contained in earlier issues. It covers the essential aspects of UF{sub 6} handling, cylinder filling and emptying, general principles of weighing and sampling, shipping, and the use of protective overpacks. The physical and chemical properties of UF{sub 6} are also described. The procedures and systems described for safe handling of UF{sub 6} presented in this document have been developed and evaluated during more than 40 years of handling vast quantities of UF{sub 6}. With proper consideration for its nuclear properties, UF{sub 6} may be safely handled in essentially the same manner as any other corrosive and/or toxic chemical.

  14. Uranium hexafluoride: A manual of good handling practices

    International Nuclear Information System (INIS)

    1991-10-01

    For many years, the US Department of Energy (DOE) and its predecessor agencies have shared with the nuclear industry their experience in the area of uranium hexafluoride (UF 6 ) shipping containers and handling procedures. The information contained in this manual updates information contained in earlier issues. It covers the essential aspects of UF 6 handling, cylinder filling and emptying, general principles of weighing and sampling, shipping, and the use of protective overpacks. The physical and chemical properties of UF 6 are also described and tabulated. The nuclear industry is responsible for furnishing its own shipping cylinders and suitable protective overpacks. A substantial effort has been made by the industry to standardize UF 6 cylinders, samples, and overpacks. The quality of feed materials is important to the safe and efficient operation of the enriching facilities, and the UF 6 product purity from the enriching facilities is equally important to the fuel fabricator, the utilities, the operators of research reactors, and other users. The requirements have been the impetus for an aggressive effort by DOE and its contractors to develop accurate techniques for sampling and for chemical and isotopic analysis. A quality control program is maintained within the DOE enriching facilities to ensure that the proper degree of accuracy and precision are obtained for all the required measurements. The procedures and systems described for safe handling of UF 6 presented in this document have been developed and evaluated in DOE facilities during more than 40 years of handling vast quantities of UF 6 . With proper consideration for its nuclear properties, UF 6 may be safely handled in essentially the same manner as any other corrosive and/or toxic chemical

  15. Uranium hexafluoride handling. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1991-12-31

    The United States Department of Energy, Oak Ridge Field Office, and Martin Marietta Energy Systems, Inc., are co-sponsoring this Second International Conference on Uranium Hexafluoride Handling. The conference is offered as a forum for the exchange of information and concepts regarding the technical and regulatory issues and the safety aspects which relate to the handling of uranium hexafluoride. Through the papers presented here, we attempt not only to share technological advances and lessons learned, but also to demonstrate that we are concerned about the health and safety of our workers and the public, and are good stewards of the environment in which we all work and live. These proceedings are a compilation of the work of many experts in that phase of world-wide industry which comprises the nuclear fuel cycle. Their experience spans the entire range over which uranium hexafluoride is involved in the fuel cycle, from the production of UF{sub 6} from the naturally-occurring oxide to its re-conversion to oxide for reactor fuels. The papers furnish insights into the chemical, physical, and nuclear properties of uranium hexafluoride as they influence its transport, storage, and the design and operation of plant-scale facilities for production, processing, and conversion to oxide. The papers demonstrate, in an industry often cited for its excellent safety record, continuing efforts to further improve safety in all areas of handling uranium hexafluoride. Selected papers were processed separately for inclusion in the Energy Science and Technology Database.

  16. Report to the Nuclear Regulatory Commission from the staff panel on the Commission's determination of an Extraordinary Nuclear Occurrence (ENO)

    International Nuclear Information System (INIS)

    1980-01-01

    The Panel finds that the first criterion, pertaining to whether the accident caused a discharge of radioactive material or levels of radiation offsite as defined in 10 CFR 140.84, has not been met. It further finds that there is presently insufficient information to support any definitive finding as to whether or not the second criterion, relating to damage to persons or property offsite as defined in 10 CFR 140.85, has been met. Since the Panel has not found that both criteria have been met, it recommends that the Commission determine that the accident at Three Mile Island did not constitute an extraordinary nuclear occurrence

  17. Report to the Nuclear Regulatory Commission from the staff panel on the Commission's determination of an Extraordinary Nuclear Occurrence (ENO)

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-01-01

    The Panel finds that the first criterion, pertaining to whether the accident caused a discharge of radioactive material or levels of radiation offsite as defined in 10 CFR 140.84, has not been met. It further finds that there is presently insufficient information to support any definitive finding as to whether or not the second criterion, relating to damage to persons or property offsite as defined in 10 CFR 140.85, has been met. Since the Panel has not found that both criteria have been met, it recommends that the Commission determine that the accident at Three Mile Island did not constitute an extraordinary nuclear occurrence.

  18. 24 CFR 242.47 - Insured advances for building components stored off-site.

    Science.gov (United States)

    2010-04-01

    ... HOUSING ACT AND OTHER AUTHORITIES MORTGAGE INSURANCE FOR HOSPITALS Construction § 242.47 Insured advances... only for components stored off-site in a quantity required to permit uninterrupted installation at the...

  19. Handling system for nuclear fuel pellet inspection

    International Nuclear Information System (INIS)

    Nyman, D.H.; McLemore, D.R.; Sturges, R.H.

    1978-11-01

    HEDL is developing automated fabrication equipment for fast reactor fuel. A major inspection operation in the process is the gaging of fuel pellets. A key element in the system has been the development of a handling system that reliably moves pellets at the rate of three per second without product damage or excessive equipment wear

  20. Situation of nuclear industry in Japan

    International Nuclear Information System (INIS)

    2002-08-01

    This document is a reprint of a note published by the nuclear service of the French embassy in Japan. It evokes the present day situation of nuclear facilities in Japan, the public acceptance and its attitude in front of accidents, the national energy program, the deregulation and competitiveness of nuclear power, the carrying out of the nuclear program, the future reactors, the fast neutron reactors, the dismantling activities, the fuel enrichment and reprocessing of spent fuels, the use of MOX fuel, the off-site storage, the vitrified and radiological wastes, the geological disposal of wastes, the prospects of the nuclear program, the companies involved in the Japan nuclear industry, the French-Japanese bilateral cooperation, and the ITER project in the domain of nuclear fusion. (J.S.)

  1. An analysis of repository waste-handling operations

    International Nuclear Information System (INIS)

    Dennis, A.W.

    1990-09-01

    This report has been prepared to document the operational analysis of waste-handling facilities at a geologic repository for high-level nuclear waste. The site currently under investigation for the geologic repository is located at Yucca Mountain, Nye County, Nevada. The repository waste-handling operations have been identified and analyzed for the year 2011, a steady-state year during which the repository receives spent nuclear fuel containing the equivalent of 3000 metric tons of uranium (MTU) and defense high-level waste containing the equivalent of 400 MTU. As a result of this analysis, it has been determined that the waste-handling facilities are adequate to receive, prepare, store, and emplace the projected quantity of waste on an annual basis. In addition, several areas have been identified where additional work is required. The recommendations for future work have been divided into three categories: items that affect the total waste management system, operations within the repository boundary, and the methodology used to perform operational analyses for repository designs. 7 refs., 48 figs., 11 tabs

  2. Savannah River Site offsite hazardous waste shipment data validation report. Revision 1

    International Nuclear Information System (INIS)

    Casey, C.; Kudera, D.E.; Page, L.A.; Rohe, M.J.

    1995-05-01

    The objective of this data validation is to verify that waste shipments reported in response to the US Department of Energy Headquarters data request are properly categorized according to DOE-HQ definitions. This report documents all findings and actions resulting from the independent review of the Savannah River Site data submittal, and provides a summary of the SRS data submittal and data validation strategy. The overall hazardous waste management and offsite release process from 1987--1991 is documented, along with an identification and description of the hazardous waste generation facilities. SRS did not ship any hazardous waste offsite before 1987. Sampling and analysis and surface surveying procedures and techniques used in determining offsite releasability of the shipments are also described in this report. SRS reported 150 manifested waste shipments from 1984 to 1991 that included 4,755 drums or lab packs and 13 tankers. Of these waste items, this report categorizes 4,251 as clean (including 12 tankers), 326 as likely clean, 138 as likely radioactive, and 55 as radioactive (including one tanker). Although outside the original scope of this report, 14 manifests from 1992 and 1993 are included, covering 393 drums or lab packs and seven tankers. From the 1992--1993 shipments, 58 drums or lab packs are categorized as radioactive and 16 drums are categorized as likely radioactive. The remainder are categorized as clean

  3. Savannah River Site offsite hazardous waste shipment data validation report. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Casey, C.; Kudera, D.E.; Page, L.A.; Rohe, M.J.

    1995-05-01

    The objective of this data validation is to verify that waste shipments reported in response to the US Department of Energy Headquarters data request are properly categorized according to DOE-HQ definitions. This report documents all findings and actions resulting from the independent review of the Savannah River Site data submittal, and provides a summary of the SRS data submittal and data validation strategy. The overall hazardous waste management and offsite release process from 1987--1991 is documented, along with an identification and description of the hazardous waste generation facilities. SRS did not ship any hazardous waste offsite before 1987. Sampling and analysis and surface surveying procedures and techniques used in determining offsite releasability of the shipments are also described in this report. SRS reported 150 manifested waste shipments from 1984 to 1991 that included 4,755 drums or lab packs and 13 tankers. Of these waste items, this report categorizes 4,251 as clean (including 12 tankers), 326 as likely clean, 138 as likely radioactive, and 55 as radioactive (including one tanker). Although outside the original scope of this report, 14 manifests from 1992 and 1993 are included, covering 393 drums or lab packs and seven tankers. From the 1992--1993 shipments, 58 drums or lab packs are categorized as radioactive and 16 drums are categorized as likely radioactive. The remainder are categorized as clean.

  4. The Greenpeace book of the nuclear age

    International Nuclear Information System (INIS)

    May, John

    1989-01-01

    This is a comprehensive popular account of both civil and military nuclear accidents. It documents major reactor accidents in certain categories such as where significant injury of death occurred, where there was considerable radioactivity release offsite or where there was actual or suspected core damage. Military accidents on nuclear submarines or involving nuclear weapons and civil incidents such as those involving radioactive waste are also included. the book is organised chronologically from the 1940s through succeeding decades to the 1980s. A concluding chapter gives Greenpeace's own perspective on the state of the nuclear age. (author)

  5. Mission Need Statement for the Idaho National Laboratory Remote-Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    Lisa Harvego

    2009-06-01

    The Idaho National Laboratory proposes to establish replacement remote-handled low-level waste disposal capability to meet Nuclear Energy and Naval Reactors mission-critical, remote-handled low-level waste disposal needs beyond planned cessation of existing disposal capability at the end of Fiscal Year 2015. Remote-handled low-level waste is generated from nuclear programs conducted at the Idaho National Laboratory, including spent nuclear fuel handling and operations at the Naval Reactors Facility and operations at the Advanced Test Reactor. Remote-handled low-level waste also will be generated by new programs and from segregation and treatment (as necessary) of remote-handled scrap and waste currently stored in the Radioactive Scrap and Waste Facility at the Materials and Fuels Complex. Replacement disposal capability must be in place by Fiscal Year 2016 to support uninterrupted Idaho operations. This mission need statement provides the basis for the laboratory’s recommendation to the Department of Energy to proceed with establishing the replacement remote-handled low-level waste disposal capability, project assumptions and constraints, and preliminary cost and schedule information for developing the proposed capability. Without continued remote-handled low-level waste disposal capability, Department of Energy missions at the Idaho National Laboratory would be jeopardized, including operations at the Naval Reactors Facility that are critical to effective execution of the Naval Nuclear Propulsion Program and national security. Remote-handled low-level waste disposal capability is also critical to the Department of Energy’s ability to meet obligations with the State of Idaho.

  6. Remote handling equipment for laboratory research of fuel reprocessing in Nuclear Research Institute at Rez

    International Nuclear Information System (INIS)

    Fidler, J.; Novy, P.; Kyrs, M.

    1985-04-01

    Laboratory installations were developed for two nuclear fuel reprocessing methods, viz., the solvent extraction process and the fluoride volatility process. The apparatus for solvent extraction reprocessing consists of a pneumatically driven rod-chopper, a dissolver, mixer-settler extractors, an automatic fire extinguishing device and other components and it was tested using irradiated uranium. The technological line for the fluoride volatility process consists of a fluorimater, condensers, sorption columns with NaF pellets and a distillation column for the separation of volatile fluorides from UF 6 . The line has not yet been tested using irradiated fuel. Some features of the remote handling equipment of both installations are briefly described. (author)

  7. Automation and remote handling activities in BARC: an overview

    International Nuclear Information System (INIS)

    Badodkar, D.N.

    2016-01-01

    Division of Remote Handling and Robotics, BARC has been working on design and development of various application specific remote handling and automation systems for nuclear front-end and back-end fuel cycle technologies. Division is also engaged in preservice and in-service inspection of coolant channels for Pressurized Heavy Water Reactors in India. Design and development of Reactor Control Mechanisms for Nuclear Research and Power Reactors (PHWRs and Compact LWRs) is another important activity carried out in this division. Robotic systems for Indoor and Outdoor surveillance in and around nuclear installations have also been developed. A line scan camera based system has been developed for measuring individual PHWR fuel pellet lengths as well as stack length. An industrial robot is used for autonomous exchange of pellets to achieve desired stack length. The system can be extended for active fuel pellets also. An automation system has been conceptualized for remote handling and transfer of spent fuel bundles from storage pool directly to the chopper unit of reprocessing plant. In case of Advanced Heavy Water Reactor which uses mixed oxides of (Th-Pu) and (Th-"2"3"3U ) as fuel, automation system for front-end fuel cycle has been designed, which includes Powder processing and pressing; Pellet handling and inspection; Pin handling and inspection; and Cluster assembly and dis-assembly in shielded facilities. System demonstration through fullscale mock-up facility is nearing completion. Above talk is presented on behalf of all the officers and staff of DRHR. The talk is mainly focused on development of an automated fuel fabrication facility for mixed oxides of (Th- Pu)/(Th-"2"3"3U ) fuel pins. An overview of divisional ongoing activities in the field of remote handling and automation are also covered. (author)

  8. How the NWC handles software as product

    Energy Technology Data Exchange (ETDEWEB)

    Vinson, D.

    1997-11-01

    This tutorial provides a hands-on view of how the Nuclear Weapons Complex project should be handling (or planning to handle) software as a product in response to Engineering Procedure 401099. The SQAS has published the document SQAS96-002, Guidelines for NWC Processes for Handling Software Product, that will be the basis for the tutorial. The primary scope of the tutorial is on software products that result from weapons and weapons-related projects, although the information presented is applicable to many software projects. Processes that involve the exchange, review, or evaluation of software product between or among NWC sites, DOE, and external customers will be described.

  9. Legislation for nuclear disaster

    International Nuclear Information System (INIS)

    Nagata, Shozo

    2012-01-01

    Fukushima nuclear disaster accident clarified problems on nuclear-related legislation and its application. Legislation for nuclear disaster (LNA) could not respond to severe accident because assumed size of accident was not enough. After emergency event corresponding to the article 15 of LNA, was reported by the operator, more than two hours passed by the issuance of Emergency State Declaration. Off-site center could not work at all. This article reviewed outline of LNA and introduced discussion on the reform of legislation and its application. Reform discussion should be focused on swift and effective response readiness to emergency: 1) operator's substantial nuclear emergency drilling, (2) reinforcement of government's headquarters for emergency response, (3) after nuclear emergency, government's headquarters remained to enhance resident's safety from radiation hazard and (4) enactment of nuclear emergency preparedness guidelines for local communities. (T. Tanaka)

  10. Nuclear safety: Impressive and worrisome trends

    International Nuclear Information System (INIS)

    Meserve, R.A.

    2006-01-01

    The safety record for commercial nuclear power has, in the main, been impressive in recent years. Nonetheless, noteworthy events continue to occur around the globe, including events at reactors operating in countries with extensive operational experience and strong regulatory capabilities. None of the recent events has resulted in a substantial off-site release of radioactivity. But these events reinforce how wrong it would be to assume that the safety challenge has been 'solved' and that attention can be focused on other matters. There are other worrisome problems that include: aging nuclear power plants, decay in the nuclear infrastructure and expanding interest in nuclear power among countries with out prior experience of operating nuclear power plants

  11. Annual report of Nuclear Emergency Assistance and Training Center. April 1, 2006 - March 31, 2008

    International Nuclear Information System (INIS)

    2008-12-01

    When a nuclear emergency occurs in Japan, the Japan Atomic Energy Agency (JAEA) provides technical support to the National government, local governments, police, fire station and license holder etc. They are Designated Public Organizations conforming to the Basic Law on Emergency Preparedness and the Basic Plan for Disaster Countermeasures. The Nuclear Emergency Assistance and Training Center (NEAT) of JAEA provides a comprehensive range of technical support activities to an off-site center in case of a nuclear emergency. Specifically, NEAT gives technical advice and information, provides for the dispatch of specialist as required, supplies emergency equipments and materials to the Joint Council of Nuclear Disaster Countermeasures, which meets at the off-site center. NEAT provide various lectures and training course concerning nuclear disaster prevention for those personnel taking an active part in emergency response organizations at normal time. And NEAT researches on nuclear disaster prevention and also cooperate with international organizations. This annual report summarized the activities of JAEA/NEAT in the fiscal year 2006 and 2007. (author)

  12. Methods and codes for assessing the off-site Consequences of nuclear accidents. Volume 2

    International Nuclear Information System (INIS)

    Kelly, G.N.; Luykx, F.

    1991-01-01

    The Commission of the European Communities, within the framework of its 1980-84 radiation protection research programme, initiated a two-year project in 1983 entitled methods for assessing the radiological impact of accidents (Maria). This project was continued in a substantially enlarged form within the 1985-89 research programme. The main objectives of the project were, firstly, to develop a new probabilistic accident consequence code that was modular, incorporated the best features of those codes already in use, could be readily modified to take account of new data and model developments and would be broadly applicable within the EC; secondly, to acquire a better understanding of the limitations of current models and to develop more rigorous approaches where necessary; and, thirdly, to quantify the uncertainties associated with the model predictions. This research led to the development of the accident consequence code Cosyma (COde System from MAria), which will be made generally available later in 1990. The numerous and diverse studies that have been undertaken in support of this development are summarized in this paper, together with indications of where further effort might be most profitably directed. Consideration is also given to related research directed towards the development of real-time decision support systems for use in off-site emergency management

  13. Nuclear assay of coal. Volume 6. Mass flow devices for coal handling. Final report

    International Nuclear Information System (INIS)

    Gozani, T.; Elias, E.; Bevan, R.

    1980-04-01

    The mass of coal entering the boiler per unit time is an essential parameter for determining the total rate of heat input. The mass flow rate of coal on a conveyor belt is generally determined as a product of the instantaneous mass of material on a short section of the belt and the belt velocity. Belt loading could be measured by conventional transducers incorporating mechanical or electromechanical weighers or by gamma-ray attenuation gauge. This report reviews the state of the art in mass flow devices for coal handling. The various methods are compared and commented upon. Special design issues are discussed relative to incorporating a mass flow measuring device in a Continuous On-Line Nuclear Analysis of Coal (CONAC) system

  14. Installation and method for handling fuel assemblies of fast nuclear reactors

    International Nuclear Information System (INIS)

    Aubert, Michel; Renaux, Charley.

    1982-01-01

    This invention concerns an installation and a method for handling the assemblies which makes it possible to have a large revolving plug smaller in diameter than that of the presently known solutions. This large, coaxial to the core, revolving plug has a handling arm enabling a fraction of the assemblies to be reached and deposited in a handling well. Through a small offset revolving plug the remainder of the assemblies can be reached and deposited in a pick-up well accessible to the arm of the large revolving plug [fr

  15. Experience of safety and performance improvement for fuel handling equipment

    International Nuclear Information System (INIS)

    Gyoon Chang, Sang; Hee Lee, Dae

    2014-01-01

    The purpose of this study is to provide experience of safety and performance improvement of fuel handling equipment for nuclear power plants in Korea. The fuel handling equipment, which is used as an important part of critical processes during the refueling outage, has been improved to enhance safety and to optimize fuel handling procedures. Results of data measured during the fuel reloading are incorporated into design changes. The safety and performance improvement for fuel handling equipment could be achieved by simply modifying the components and improving the interlock system. The experience provided in this study can be useful lessons for further improvement of the fuel handling equipment. (authors)

  16. TMI-2 [Three Mile Island Nuclear Power Station] fuel canister and core sample handling equipment used in INEL hot cells

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Shurtliff, W.T.; Lynch, R.J.; Croft, K.M.; Whitmill, L.J.; Allen, S.M.

    1987-01-01

    This paper describes the specialized remote handling equipment developed and used at the Idaho National Engineering Laboratory (INEL) to handle samples obtained from the core of the damaged Unit 2 reactor at Three Mile Island Nuclear Power Station (TM-2). Samples of the core were removed, placed in TMI-2 fuel canisters, and transported to the INEL. Those samples will be examined as part of the analysis of the TMI-2 accident. The equipment described herein was designed for removing sample materials from the fuel canisters, assisting with initial examination, and processing samples in preparation for detailed examinations. The more complex equipment used microprocessor remote controls with electric motor drives providing the required force and motion capabilities. The remaining components were unpowered and manipulator assisted

  17. Safety Analysis of 'Older/Aged' Handling and Transportation Equipment for Heavy Loads, Radioactive Waste and Materials in Accordance with German Nuclear Standards KTA 3902, 3903 and 3905

    International Nuclear Information System (INIS)

    Macias, P.; Prucker, E.; Stang, W.

    2006-01-01

    The purpose of this paper is to present a general safety analysis of important handling and transportation processes and their related equipment ('load chains' consisting of cranes, load-bearing equipment and load-attaching points). This project was arranged by the responsible Bavarian ministry for environment, health and consumer protection (StMUGV) in agreement with the power plant operators of all Bavarian nuclear power plants to work out potential safety improvements. The range of the equipment (e.g. reactor building, crane, refuelling machine, load-bearing equipment and load-attaching points) covers the handling and transportation of fuel elements (e. g. with fuel flasks), heavy loads (e.g. reactor pressure vessel closure head, shielding slabs) and radioactive materials and waste (e.g. waste flasks, control elements, fuel channels, structure elements). The handling equipment was subjected to a general safety analysis taking into account the ageing of the equipment and the progress of standards. Compliance with the current valid requirements of the state of science and technology as required by German Atomic Act and particularly of the nuclear safety KTA-standards (3902, 3903 and 3905) was examined. The higher protection aims 'safe handling and transportation of heavy loads and safe handling of radioactive materials and waste' of the whole analysis are to avoid a criticality accident, the release of radioactivity and inadmissible effects on important technical equipment and buildings. The scope of the analysis was to check whether these protection aims were fulfilled for all important technical handling and transportation processes. In particularly the design and manufacturing of the components and the regulations of the handling itself were examined. (authors)

  18. Preliminary Safety Design Report for Remote Handled Low-Level Waste Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Timothy Solack; Carol Mason

    2012-03-01

    A new onsite, remote-handled low-level waste disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled low-level waste disposal for remote-handled low-level waste from the Idaho National Laboratory and for nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled low-level waste in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This preliminary safety design report supports the design of a proposed onsite remote-handled low-level waste disposal facility by providing an initial nuclear facility hazard categorization, by discussing site characteristics that impact accident analysis, by providing the facility and process information necessary to support the hazard analysis, by identifying and evaluating potential hazards for processes associated with onsite handling and disposal of remote-handled low-level waste, and by discussing the need for safety features that will become part of the facility design.

  19. Radiological assessment of depleted uranium migration offsite from an ordnance range

    International Nuclear Information System (INIS)

    Rynders, D.G.

    1996-01-01

    The military utilizes ordnance loaded with depleted uranium in order to maximize armor penetrating capabilities. These weapons are tested on open ranges where the weapons are fired through a cloth target and impact into the soil. This paper examines the potential environmental impact from use of depleted uranium in an open setting. A preliminary pathway analysis was performed to examine potential routes of exposure to nonhuman species in the vicinity and ultimately to man. Generic data was used in the study to estimate the isotopic mix and weight of the ordnance. Key factors in the analysis included analyzing the physics of weapon impact on soil, chemical changes in material upon impact, and mechanisms of offsite transport (including atmospheric and overland transport). Non-standard exposure scenarios were investigated, including the possibility of offsite contaminant transport due to range grassfires. Two radiological assessment codes, MEPAS (Multi media Environmental Pollutant Assessment System) and RESRAD were used to help analyze the scenarios

  20. Data handling systems and methods of wiring

    International Nuclear Information System (INIS)

    Grant, J.

    1981-01-01

    An improved data handling system, for monitoring and control of nuclear reactor operations, is described in which time delays associated with scanning are reduced and noise and fault signals in the system are resolved. (U.K.)

  1. Analysis of offsite Emergency Planning Zones (EPZs) for the Rocky Flats Plant

    International Nuclear Information System (INIS)

    Petrocchi, A.J.; Zimmerman, G.A.

    1994-01-01

    During Phase 3 of the EPZ project, a sitewide analysis will be performed applying a spectrum-of-accidents approach to both radiological and nonradiological hazardous materials release scenarios. This analysis will include the MCA but will be wider in scope and will produce options for the State of Colorado for establishing a bounding EPZ that is intended to more comprehensively update the interim, preliminary EPZ developed in Phase 2. EG ampersand G will propose use of a hazards assessment methodology that is consistent with the DOE Emergency Management Guide for Hazards Assessments and other methods required by DOE orders. This will include hazards, accident, safety, and risk analyses. Using this methodology, EG ampersand G will develop technical analyses for a spectrum of accidents. The analyses will show the potential effects from the spectrum of accidents on the offsite population together with identification of offsite vulnerable zones and areas of concern. These analyses will incorporate state-of-the-art technology for accident analysis, atmospheric plume dispersion modeling, consequence analysis, and the application of these evaluations to the general public population at risk. The analyses will treat both radiological and nonradiological hazardous materials and mixtures of both released accidentally to the atmosphere. DOE/RFO will submit these results to the State of Colorado for the State's use in determining offsite emergency planning zones for the Rocky Flats Plant. In addition, the results will be used for internal Rocky Flats Plant emergency planning

  2. Data on loss of off-site electric power simulation tests of the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

    2002-07-01

    The high temperature engineering test reactor (HTTR), the first high temperature gas-cooled reactor (HTGR) in Japan, achieved the first full power of 30 MW on December 7 in 2001. In the rise-to-power test of the HTTR, simulation tests on loss of off-site electric power from 15 and 30 MW operations were carried out by manual shutdown of off-site electric power. Because helium circulators and water pumps coasted down immediately after the loss of off-site electric power, flow rates of helium and water decreased to the scram points. To shut down the reactor safely, the subcriticality should be kept by the insertion of control rods and the auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components. About 50 s later from the loss of off-site electric power, the auxiliary cooling system started up by supplying electricity from emergency power feeders. Temperature of hot plenum block among core graphite structures decreased continuously after the startup of the auxiliary cooling system. This report describes sequences of dynamic components and transient behaviors of the reactor and its cooling system during the simulation tests from 15 and 30 MW operations. (author)

  3. PND fuel handling decontamination program: specialized techniques and results

    International Nuclear Information System (INIS)

    Pan, R.; Hobbs, K.; Minnis, M.; Graham, K.

    1995-01-01

    The use of various decontamination techniques and equipment has become a critical part of Fuel Handling maintenance work at the Pickering Nuclear Station, an eight unit CANDU station located about 30 km east of Toronto. This paper presents an overview of the set up and techniques used for cleaning in the PND Fuel Handling Maintenance Facility, and the results achieved. (author)

  4. A perspective on atomspheric nuclear tests in Nevada fact book

    International Nuclear Information System (INIS)

    Friesen, H.N.

    1992-04-01

    This Fact Book provides historical background and perspective on the nuclear testing program at the Nevada Test Site (NTS). Nuclear tests contributing to the off-site deposition of radioactive fallout are identified, and the concept of cumulative estimated exposure is explained. The difficulty of associating health effects with radiation is presented also. The status of litigation against the government and legislation (as of December 1991) are summarized

  5. AMCO Off-Site Air Monitoring Polygons, Oakland CA, 2017, US EPA Region 9

    Data.gov (United States)

    U.S. Environmental Protection Agency — This feature class was developed to support the AMCO Chemical Superfund Site air monitoring process and depicts a single polygon layer, Off-Site Air Monitors,...

  6. Emergency control centers for nuclear power plants

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    Guidance is provided for the development and implementation of emergency control centers for nuclear power plants, including nuclear plant control room, nuclear plant company headquarters, emergency control center, and nuclear plant alternate emergency control center. Requirements and recommendations are presented for the mission, communications, instrumentation and equipment associated with each type of control center. Decisional aids, manning requirements and resources are also given; the decision aids cover both the accident assessment and protective action areas. Both normal and alternate means of communications are considered. Off-site emergency control centers, although not covered in the strict sense by this standard, are considered in an appendix

  7. Site-specific parameter values for the Nuclear Regulatory Commission's food pathway dose model

    International Nuclear Information System (INIS)

    Hamby, D.M.

    1992-01-01

    Routine operations at the Savannah River Site (SRS) in Western South Carolina result in radionuclide releases to the atmosphere and to the Savannah River. The resulting radiation doses to the off-site maximum individual and the off-site population within 80 km of the SRS are estimated on a yearly basis. These estimates are currently generated using dose models prescribed for the commercial nuclear power industry by the Nuclear Regulatory Commission (NRC). The NRC provides default values for dose-model parameters for facilities without resources to develop site-specific values. A survey of land- and water-use characteristics for the Savannah River area has been conducted to determine site-specific values for water recreation, consumption, and agricultural parameters used in the NRC Regulatory Guide 1.109 (1977) dosimetric models. These site parameters include local characteristics of meat, milk, and vegetable production; recreational and commercial activities on the Savannah River; and meat, milk, vegetable, and seafood consumption rates. This paper describes how parameter data were obtained at the Savannah River Site and the impacts of such data on off-site dose. Dose estimates using site-specific parameter values are compared to estimates using the NRC default values

  8. Multiple external hazards compound level 3 PSA methods research of nuclear power plant

    Science.gov (United States)

    Wang, Handing; Liang, Xiaoyu; Zhang, Xiaoming; Yang, Jianfeng; Liu, Weidong; Lei, Dina

    2017-01-01

    2011 Fukushima nuclear power plant severe accident was caused by both earthquake and tsunami, which results in large amount of radioactive nuclides release. That accident has caused the radioactive contamination on the surrounding environment. Although this accident probability is extremely small, once such an accident happens that is likely to release a lot of radioactive materials into the environment, and cause radiation contamination. Therefore, studying accidents consequences is important and essential to improve nuclear power plant design and management. Level 3 PSA methods of nuclear power plant can be used to analyze radiological consequences, and quantify risk to the public health effects around nuclear power plants. Based on multiple external hazards compound level 3 PSA methods studies of nuclear power plant, and the description of the multiple external hazards compound level 3 PSA technology roadmap and important technical elements, as well as taking a coastal nuclear power plant as the reference site, we analyzed the impact of off-site consequences of nuclear power plant severe accidents caused by multiple external hazards. At last we discussed the impact of off-site consequences probabilistic risk studies and its applications under multiple external hazards compound conditions, and explained feasibility and reasonableness of emergency plans implementation.

  9. COCO-1: model for assessing the cost of offsite consequences of accidental releases of radioactivity

    International Nuclear Information System (INIS)

    Haywood, S.M.; Robinson, C.A.; Heady, C.

    1991-09-01

    This report describes a new model, called COCO-1 (Cost Of Consequences Offsite), for assessing the offsite economic consequences of an accident involving the release of radioactive material. The costs calculated are a measure of the benefit foregone as a result of the accident, and in addition to tangible monetary costs the model attempts to include costs arising from the effect of the accident on individuals, for instance the disruption caused by the loss of homes. The approach has limitations, which are discussed, but offers a broadly applicable and robust technique for estimating the economic impact of most accidents. (author)

  10. Application of advanced handling techniques to transportation cask design

    International Nuclear Information System (INIS)

    Bennett, P.C.

    1992-01-01

    Sandia National Laboratories supports the US Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM) applying technology to the safe transport of nuclear waste. Part of that development effort includes investigation of advanced handling technologies for automation of cask operations at nuclear waste receiving facilities. Although low radiation levels are expected near transport cask surfaces, cumulative occupational exposure at a receiving facility can be significant. Remote automated cask handling has the potential to reduce both the occupational exposure and the time necessary to process a cask. Thus, automated handling is consistent with DOE efforts to reduce the lifecycle costs of the waste disposal system and to maintain public and occupational radiological risks as low as reasonably achievable. This paper describes the development of advanced handling laboratory mock-ups and demonstrations for spent fuel casks. Utilizing the control enhancements described below, demonstrations have been carried out including cask location and identification, contact and non-contact surveys, impact limiter removal, tiedown release, uprighting, swing-free movement, gas sampling, and lid removal operations. Manually controlled movement around a cask under off-normal conditions has also been demonstrated

  11. Spray pond design for nuclear power plants

    International Nuclear Information System (INIS)

    Codell, R.B.; Asce, A.M.

    1986-01-01

    This paper presents a complex methodology for assessing the performance of spray ponds in ultimate heat sink service at nuclear power plants. A spray pond performance model, developed in the companion paper, is used in conjunction with on-site and off-site meteorological data to predict the highest temperature and greatest 30 day water loss which can reasonable be expected to occur during the lifetime of the plant. The performance model for heat and mass transfer is used to develop an efficient phenomenological model used to scan the long-term meteorological records. Refined estimates of temperature or water loss may then be based on more complicated models if necessary. Short-term onsite data are correlated to the long-term off-site data to formulate correction factors for the difference in location. Cumulative distribution functions for temperature and water loss are determined from the long-term meteorological records to predict the occurrence of these quantities which are less severe that the peak. The methodology is demonstrated using data and parameters from the Palo Verde nuclear plant as an example

  12. Microcomputer simulation model for facility performance assessment: a case study of nuclear spent fuel handling facility operations

    International Nuclear Information System (INIS)

    Chockie, A.D.; Hostick, C.J.; Otis, P.T.

    1985-10-01

    A microcomputer based simulation model was recently developed at the Pacific Northwest Laboratory (PNL) to assist in the evaluation of design alternatives for a proposed facility to receive, consolidate and store nuclear spent fuel from US commercial power plants. Previous performance assessments were limited to deterministic calculations and Gantt chart representations of the facility operations. To insure that the design of the facility will be adequate to meet the specified throughput requirements, the simulation model was used to analyze such factors as material flow, equipment capability and the interface between the MRS facility and the nuclear waste transportation system. The simulation analysis model was based on commercially available software and application programs designed to represent the MRS waste handling facility operations. The results of the evaluation were used by the design review team at PNL to identify areas where design modifications should be considered. 4 figs

  13. OFF-SITE SMARTPHONE VS. STANDARD WORKSTATION IN THE RADIOGRAPHIC DIAGNOSIS OF SMALL INTESTINAL MECHANICAL OBSTRUCTION IN DOGS AND CATS.

    Science.gov (United States)

    Noel, Peter G; Fischetti, Anthony J; Moore, George E; Le Roux, Alexandre B

    2016-09-01

    Off-site consultations by board-certified veterinary radiologists benefit residents and emergency clinicians by providing immediate feedback and potentially improving patient outcome. Smartphone devices and compressed images transmitted by email or text greatly facilitate availability of these off-site consultations. Criticism of a smartphone interface for off-site consultation is mostly directed at image degradation relative to the standard radiographic viewing room and monitors. The purpose of this retrospective, cross-sectional, methods comparison study was to compare the accuracy of abdominal radiographs in two imaging interfaces (Joint Photographic Experts Group, off-site, smartphone vs. Digital Imaging and Communications in Medicine, on-site, standard workstation) for the diagnosis of small intestinal mechanical obstruction in vomiting dogs and cats. Two board-certified radiologists graded randomized abdominal radiographs using a five-point Likert scale for the presence of mechanical obstruction in 100 dogs or cats presenting for vomiting. The area under the receiver operator characteristic curves for both imaging interfaces was high. The accuracy of the smartphone and traditional workstation was not statistically significantly different for either reviewer (P = 0.384 and P = 0.536). Correlation coefficients were 0.821 and 0.705 for each reviewer when the same radiographic study was viewed in different formats. Accuracy differences between radiologists were potentially related to years of experience. We conclude that off-site expert consultation with a smartphone provides an acceptable interface for accurate diagnosis of small intestinal mechanical obstruction in dogs and cat. © 2016 American College of Veterinary Radiology.

  14. Off-site shipment request development and review plan

    International Nuclear Information System (INIS)

    1992-05-01

    On May 17, 1991, Department of Energy Headquarters (DOE-HQ) imposed a moratorium on the shipment of all Resource Conservation and Recovery Act (RCRA) hazardous and Toxic Substances Control Act (TSCA) waste to commercial treatment, storage and disposal facilities. The moratorium was imposed after it was discovered that some shipments of RCRA and TSCA waste from Department of Energy (DOE) sites contained small quantities of radioactive and special nuclear material (SNM). The shipment of these wastes has been attributed to inconsistent and possibly erroneous interpretation of DOE Orders and guidance. In an effort to clarify existing DOE Orders and guidance and establish throughout the DOE complex, June 21, 1991, DOE-HQ issued in draft the Performance Objective for Certification of Non-Radioactive Hazardous Waste. This Performance Objective was subsequently approved on November 15, 1991. The Performance Objective contains specific requirements that must be net to allow the shipment of RCRA and TSCA waste for commercial treatment, storage and disposal. On July 16, 1991, based on the initial draft of the Performance Objective, Martin Marietta Energy Systems (MMES) issued a directive which applies the Performance Objective requirements to all wastes and materials. In addition, this MMES directive imposed the requirement for a review by a Central Waste Management (CWM) Readiness Review Board (RRB). Additional DOE and MMES guidance and directives have been issued since May 17, 1991. This plan applies to all waste destined for shipment from the Portsmouth Gaseous Diffusion Plant (PORTS) to off-site commercial treatment, storage and disposal facilities, and to all materials destined for recycle, surplus and salvage

  15. Planning and management of outages in nuclear power plants

    International Nuclear Information System (INIS)

    Sica, G.F.; Fusari, W.; Reginelli, A.

    1984-01-01

    At present the Ente Nazionale per l'Energia Elettrica (ENEL) operates three nuclear power plants, only one of which belongs to the new generation, i.e. the Caorso Nuclear Power Plant which has been in commercial operation since December 1981. Outage planning, implementation and analysis are very important in order to minimize the shutdown time and thus improve plant availability, which is of particular importance for a large nuclear power plant. Such activities are very complicated because of the large number of jobs that have to be performed in accordance with detailed written procedures and which have to be properly documented and controlled. Large off-site resources are required which have to be accurately interfaced with on-site staff. The ENEL is making a great effort to define both the administrative and technical aspects of refuelling outages. As outage planning requires the availability and handling of a large amount of data and information, a maintenance information system that has been widely used in conventional plants was applied, with some modifications made especially for the Caorso Nuclear Power Plant. After two years the following results have been achieved: a large number of raw and processed data are now available, the first refuelling outage was carried out with few problems and according to schedule, and the second refuelling outage, based on the experience of the first, required somewhat less preparation and is developing well even though many special activities have had to be scheduled. The ENEL believes that the efforts made in the planning and management areas will pay off in terms of the short duration, smoothness and economy of further outages, both for Caorso and for future plants. (author)

  16. SITE GENERATED RADIOLOGICAL WASTE HANDLING SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    S. C. Khamankar

    2000-06-20

    The Site Generated Radiological Waste Handling System handles radioactive waste products that are generated at the geologic repository operations area. The waste is collected, treated if required, packaged for shipment, and shipped to a disposal site. Waste streams include low-level waste (LLW) in solid and liquid forms, as-well-as mixed waste that contains hazardous and radioactive constituents. Liquid LLW is segregated into two streams, non-recyclable and recyclable. The non-recyclable stream may contain detergents or other non-hazardous cleaning agents and is packaged for shipment. The recyclable stream is treated to recycle a large portion of the water while the remaining concentrated waste is packaged for shipment; this greatly reduces the volume of waste requiring disposal. There will be no liquid LLW discharge. Solid LLW consists of wet solids such as ion exchange resins and filter cartridges, as-well-as dry active waste such as tools, protective clothing, and poly bags. Solids will be sorted, volume reduced, and packaged for shipment. The generation of mixed waste at the Monitored Geologic Repository (MGR) is not planned; however, if it does come into existence, it will be collected and packaged for disposal at its point of occurrence, temporarily staged, then shipped to government-approved off-site facilities for disposal. The Site Generated Radiological Waste Handling System has equipment located in both the Waste Treatment Building (WTB) and in the Waste Handling Building (WHB). All types of liquid and solid LLW are processed in the WTB, while wet solid waste from the Pool Water Treatment and Cooling System is packaged where received in the WHB. There is no installed hardware for mixed waste. The Site Generated Radiological Waste Handling System receives waste from locations where water is used for decontamination functions. In most cases the water is piped back to the WTB for processing. The WTB and WHB provide staging areas for storing and shipping LLW

  17. SITE GENERATED RADIOLOGICAL WASTE HANDLING SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    S. C. Khamankar

    2000-01-01

    The Site Generated Radiological Waste Handling System handles radioactive waste products that are generated at the geologic repository operations area. The waste is collected, treated if required, packaged for shipment, and shipped to a disposal site. Waste streams include low-level waste (LLW) in solid and liquid forms, as-well-as mixed waste that contains hazardous and radioactive constituents. Liquid LLW is segregated into two streams, non-recyclable and recyclable. The non-recyclable stream may contain detergents or other non-hazardous cleaning agents and is packaged for shipment. The recyclable stream is treated to recycle a large portion of the water while the remaining concentrated waste is packaged for shipment; this greatly reduces the volume of waste requiring disposal. There will be no liquid LLW discharge. Solid LLW consists of wet solids such as ion exchange resins and filter cartridges, as-well-as dry active waste such as tools, protective clothing, and poly bags. Solids will be sorted, volume reduced, and packaged for shipment. The generation of mixed waste at the Monitored Geologic Repository (MGR) is not planned; however, if it does come into existence, it will be collected and packaged for disposal at its point of occurrence, temporarily staged, then shipped to government-approved off-site facilities for disposal. The Site Generated Radiological Waste Handling System has equipment located in both the Waste Treatment Building (WTB) and in the Waste Handling Building (WHB). All types of liquid and solid LLW are processed in the WTB, while wet solid waste from the Pool Water Treatment and Cooling System is packaged where received in the WHB. There is no installed hardware for mixed waste. The Site Generated Radiological Waste Handling System receives waste from locations where water is used for decontamination functions. In most cases the water is piped back to the WTB for processing. The WTB and WHB provide staging areas for storing and shipping LLW

  18. Use of PRA in the nuclear regulatory field in South Africa

    International Nuclear Information System (INIS)

    Hill, T.F.

    1994-01-01

    The nuclear regulatory authority in South Africa (since 1988 the Council for Nuclear Safety (CNS)), established in 1973 nuclear safety criteria against which to assess the level of safety of any facility using radioactive material. It is a regulatory requirement in South Africa to develop and maintain a living PRA for each facility and thereby to provide the necessary information to demonstrate compliance against these criteria. All safety submissions to the CNS must include at least a risk statement based on an accepted PRA study. The function of the CNS is to regulate all activities in South Africa involving the use of radioactive material and posing a significant risk to the public or plant personnel. This includes most aspects of the nuclear fuel cycle and the Koeberg NPS (two 2775 MW(th) PWRs). A PRA study including source terms for the two Koeberg units was presented by the contractor in 1979. This included the risk due to power and shutdown states and non reactor related accidents involving spent fuel storage, fuel handling and waste treatment related activities. At least 20 PRA studies have been performed for other nuclear facilities in the country. The CNS maintains an in-house PRA capability to perform independent assessments of licensee submission, to participate in developments of PRA methodology in the regulatory field, to perform pro-active safety work and to assist in regulatory decision making. Present ongoing work includes the development of a risk monitor, a risk management system, improvement in PRA codes, models, data collection and analysis, off-site risk assessment methodology and associated regulatory policy. (author). 1 fig

  19. Evaluating background noise: Assessing off-site data from field surveys around the Italic sanctuary of S. Giovanni in Galdo, Molise, Italy

    NARCIS (Netherlands)

    Waagen, J.

    2014-01-01

    This study demonstrates the potential of intensive sampling and off-site analyses to identify evidence of human activity in the past using the off-site data collected around the Italic sanctuary of S. Giovanni in Galdo, Molise, Italy. Rather than employing general and monocausal explanatory

  20. A simulator-based nuclear reactor emergency response training exercise.

    Science.gov (United States)

    Waller, Edward; Bereznai, George; Shaw, John; Chaput, Joseph; Lafortune, Jean-Francois

    Training offsite emergency response personnel basic awareness of onsite control room operations during nuclear power plant emergency conditions was the primary objective of a week-long workshop conducted on a CANDU® virtual nuclear reactor simulator available at the University of Ontario Institute of Technology, Oshawa, Canada. The workshop was designed to examine both normal and abnormal reactor operating conditions, and to observe the conditions in the control room that may have impact on the subsequent offsite emergency response. The workshop was attended by participants from a number of countries encompassing diverse job functions related to nuclear emergency response. Objectives of the workshop were to provide opportunities for participants to act in the roles of control room personnel under different reactor operating scenarios, providing a unique experience for participants to interact with the simulator in real-time, and providing increased awareness of control room operations during accident conditions. The ability to "pause" the simulator during exercises allowed the instructors to evaluate and critique the performance of participants, and to provide context with respect to potential offsite emergency actions. Feedback from the participants highlighted (i) advantages of observing and participating "hands-on" with operational exercises, (ii) their general unfamiliarity with control room operational procedures and arrangements prior to the workshop, (iii) awareness of the vast quantity of detailed control room procedures for both normal and transient conditions, and (iv) appreciation of the increased workload for the operators in the control room during a transient from normal operations. Based upon participant feedback, it was determined that the objectives of the training had been met, and that future workshops should be conducted.

  1. Computerization of off-site dose calculations at two nuclear power plants

    International Nuclear Information System (INIS)

    Lei, W.; Robertson, C.E.; Moore, G.T.; Rawls, B.E.; Sipp, J.R.

    1988-01-01

    The Brunswick Nuclear Project (BNP) consists of two boiling water reactors designed to generate a total net output of 1642 MWe. Unit 2 achieved commercial production in 1975, and Unit 1 began commercial operation in 1977. The Harris nuclear Project (HNP) is an 860 MWe pressurized water reactor that entered commercial operation in May of 1987. both plant sites are operated by Carolina Power and Light Company (CP and L). During January 1984, BNP replaced its older effluent technical specifications (part of the plant's original license) with the newer generation of Radiological Effluent Technical Specifications (RETS) mandated by the U.S. Nuclear Regulatory Commission. The RETS for HNP were integrated directly into this initial technical specifications. The initial version of the ODCM for BNP was drafted by a vendor and then extensively rewritten by the plant staff. The manual for HNP was drafted by CP and L corporate staff. At the outset, it was realized how impractical it would be to attempt to manually perform all of the dose calculations and keep the necessary records. The alternative-computerization-required extensive in-plant and corporate efforts to identify the computing resources (hardware) needed, create the software for ODCM implementation, and test, verify, validate, and document the software. This paper discusses these efforts

  2. Investigation of off-site airborne transport of lead from a superfund removal action site using lead isotope ratios and concentrations

    Science.gov (United States)

    Pribil, Michael J.; Maddaloni, Mark A.; Staiger, Kimberly; Wilson, Eric; Magriples, Nick; Ali, Mustafa; Santella, Dennis

    2014-01-01

    Lead (Pb) concentration and Pb isotopic composition of surface and subsurface soil samples were used to investigate the potential for off-site air transport of Pb from a former white Pb processing facility to neighboring residential homes in a six block area on Staten Island, NY. Surface and subsurface soil samples collected on the Jewett White Pb site were found to range from 1.122 to 1.138 for 206Pb/207Pb and 2.393 to 2.411 for 208Pb/207Pb. The off-site surface soil samples collected from residential backyards, train trestle, near site grass patches and background areas varied from 1.144 to 1.196 for 206Pb/207Pb and 2.427 to 2.464 for 208Pb/207Pb. Two soil samples collected along Richmond Terrace, where Jewett site soils accumulated after major rain events, varied from 1.136 to 1.147 for 206Pb/207Pb and 2.407 to 2.419 for 208Pb/207Pb. Lead concentration for on-site surface soil samples ranged from 450 to 8000 ug/g, on-site subsurface soil samples ranged from 90,000 to 240,000 ug/g and off-site samples varied from 380 to 3500 ug/g. Lead concentration and isotopic composition for the Staten Island off-site samples were similar to previously published data for other northeastern US cities and reflect re-suspension and re-mobilization of local accumulated Pb. The considerable differences in both the Pb isotopic composition and Pb concentration of on-site and off-site samples resulted in the ability to geochemically trace the transport of particulate Pb. Data in this study indicate minimal off-site surface transport of Pb from the Jewett site into the neighboring residential area.

  3. PREPD O and VE remote handling system

    International Nuclear Information System (INIS)

    Theil, T.N.

    1985-01-01

    The Process Experimental Pilot Plant (PREPP) at the Idaho National Engineering Laboratory is designed for volume reduction and packaging of transuranic (TRU) waste. The PREPP opening and verification enclosure (O and VE) remote handling system, within that facility, is designed to provide examination of the contents of various TRU waste storage containers. This remote handling system will provide the means of performing a hazardous operation that is currently performed manually. The TeleRobot to be used in this system is a concept that will incorporate and develop man in the loop operation (manual mode), standardized automatic sequencing of end effector tools, increased payload and reach over currently available computer-controlled robots, and remote handling of a hazardous waste operation. The system is designed within limited space constraints and an operation that was originally planned, and is currently being manually performed at other plants. The PREPP O and VE remote handling system design incorporates advancing technology to improve the working environment in the nuclear field

  4. Development of the assessment of nuclear accident consequences and decision support system in China: status, requirement and recommendations

    International Nuclear Information System (INIS)

    Shi Zhongqi; Wang Xingyu

    2003-01-01

    This paper introduces the status of nuclear accident consequence assessment/development of decision-making support system in China. The basic functions and roles of the consequence assessment/decision-making support system for three levels of nuclear emergency response organization (i.e. national, local offsite and nuclear power plant operator) in China are presented in the paper

  5. Development of nuclear fuel cycle remote handling technology

    International Nuclear Information System (INIS)

    Kim, K. H.; Park, B. S.; Kim, S. H.

    2012-04-01

    This report presents the development of remote handling systems and remote equipment for use in the pyprocessing verification at the PRIDE (PyRoprocess Integrated inactive Demonstration facility). There are four areas conducted in this work. In first area, the prototypes of an engineering-scale high-throughput decladding voloxidizer which is capable of separating spent fuel rod-cuts into hulls and powder and collecting them separately, and an automatic equipment which is capable of collecting residual powder remaining on separated hulls were developed. In second area, a servo-manipulator system was developed to operate and maintain pyroprocess equipment located at the argon cell of the PRIDE in a remote manner. A servo-manipulator with dual arm that is mounted on the lower part of a bridge transporter will be installed on the ceiling of the in-cell and can travel the length of the ceiling. In third area, a digital mock-up and a remote handling evaluation mock-up were constructed to evaluate the pyroprocess equipments from the in-cell arrangements, remote operability and maintainability viewpoint before they are installed in the PRIDE. In last area, a base technology for remote automation of integrated pyroprocess was developed. The developed decladding voloxidizer and automatic equipment will be utilized in the development of a head-end process for pyroprocessing. In addition, the developed servo-manipulator will be used for remote operation and maintenance of the pyroprocess equipments in the PRIDE. The constructed digital mock-up and remote handling evaluation mock-up will be also used to verify and improve the pyroprocess equipments for the PRIDE application. Moreover, these remote technologies described above can be directly used in the PRIDE and applied for the KAPF (Korea Advanced Pyroprocess Facility) development

  6. Nuclear fuel handling grapple carriage with self-lubricating bearing

    International Nuclear Information System (INIS)

    1977-01-01

    This invention relates to the provision of a fuel handling grapple carriage for a sodium cooled fast breeder reactor with sodium coolant lubricated bearings in which contamination of the bearings is prevented. (UK)

  7. How to handle station black outs

    Energy Technology Data Exchange (ETDEWEB)

    Reisch, Frigyes [Swedish Nuclear Power Inspectorate, S-10252 Stockholm (Sweden)

    1986-02-15

    Station black out is defined as the loss of ail high voltage alternating current at a nuclear power site. An international study was made to survey the practices in the different countries. The best way to handle station black out is to avoid it therefore briefly the normal off site and emergency on site power supplies are discussed. The ways in use to enhance nuclear power plants using Boiling Water Reactors or Pressurized Water Reactors to cope with a station black out are discussed in some detail. (author)

  8. How to handle station black outs

    International Nuclear Information System (INIS)

    Reisch, Frigyes

    1986-01-01

    Station black out is defined as the loss of ail high voltage alternating current at a nuclear power site. An international study was made to survey the practices in the different countries. The best way to handle station black out is to avoid it therefore briefly the normal off site and emergency on site power supplies are discussed. The ways in use to enhance nuclear power plants using Boiling Water Reactors or Pressurized Water Reactors to cope with a station black out are discussed in some detail. (author)

  9. Hazards to nuclear plants from off-site release of toxic vapors

    International Nuclear Information System (INIS)

    Hornyik, K.

    1976-01-01

    A method for the assessment of risk involved in shipping toxic compounds past nuclear power plants uses a postulated chain of events, starting with a traffic accident causing instantaneous release of the compound as vapor, and leading to incapacitation of control operators in the nuclear plant, described by deterministic and statistical models as appropriate to the respective event. Statistical treatment of relevant atmospheric conditions is a major improvement over more conservative assumptions commonly made in current analyses of this problem. Consequently, one obtains a substantial reduction in the estimated risk expressed in usual terms of the annual probability of an unacceptable event, in spite of the fact that no credit is taken for protective measures other than potential control room isolation

  10. Safety related requirements on future nuclear power plants

    International Nuclear Information System (INIS)

    Niehaus, F.

    1991-01-01

    Nuclear power has the potential to significantly contribute to the future energy supply. However, this requires continuous improvements in nuclear safety. Technological advancements and implementation of safety culture will achieve a safety level for future reactors of the present generation of a probability of core-melt of less than 10 -5 per year, and less than 10 -6 per year for large releases of radioactive materials. There are older reactors which do not comply with present safety thinking. The paper reviews findings of a recent design review of WWER 440/230 plants. Advanced evolutionary designs might be capable of reducing the probability of significant off-site releases to less than 10 -7 per year. For such reactors there are inherent limitations to increase safety further due to the human element, complexity of design and capability of the containment function. Therefore, revolutionary designs are being explored with the aim of eliminating the potential for off-site releases. In this context it seems to be advisable to explore concepts where the ultimate safety barrier is the fuel itself. (orig.) [de

  11. Operation of Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Tossavainen, K.

    1992-05-01

    The Finnish nuclear power plant units Loviisa 1 and 2 as well as TVO I and II were in operation for almost the whole fourth quarter of 1991. The load factor average was 94.7 % (the whole year 90.9 %). All the events in the last annual quarter, which are classified on the International Nuclear Event Scale, were below scale/level 0. Also the events which occurred in the other quarters of the year 1991 were rated at the scale's lowest levels. Occupational radiation doses and releases of radioactive material off-site remained well below authorised limits. Only quantities of radioactive material insignificant to radiation exposure originating in nuclear power plants were detected in the samples collected in the vicinity of the nuclear power plants

  12. Comparing different technologies for wireless real-time dose rate monitoring for on- and offsite emergency management

    International Nuclear Information System (INIS)

    Dielmann, R.

    2004-01-01

    At a nuclear disaster the efficiency of emergency management on-site as well as off-site is closely connected to the quality and reliability of the actual status information. Reliability and short response time of the data communication path are important in the early phase. In order to protect investment and minimize TCO (Total Cost of Ownership) the dose rate measurement systems should also be adequate for the later phase emergency management and rehabilitation of contaminated areas. Based on four years experience the pros and cons of available GSM / GPRS / UMTS / TETRA (Terrestrial Trunked Radio) and satellite based technologies are compared with SkyLINK - a proprietary wireless network technology which is fully owned by the supervising authority or a nuclear installation. The European Commission's decision in 1999 to opt for this state-of-the-art technology within the TACIS program fulfills highest standards for emergency management networks featuring especially: independence of public communication lines, good data availability also in rural areas and in emergency scenarios, fast installation, system startup and training, extreme mobility in emergency cases, high flexibility for changing tasks, very high degree of autonomy, low operating and maintenance costs over a ten year lifetime, long-term reliability of probes and data management system. After more than four years of experience with installations around the world especially those installed around three Russian nuclear power plants a comparison of this reliably operating technology can be done. (author)

  13. IAEA Coordinates International Mission on Remediation of Areas Off-site Fukushima Daiichi NPP

    International Nuclear Information System (INIS)

    2011-01-01

    Full text: The International Atomic Energy Agency (IAEA) will dispatch an international expert mission to Japan to assist the country in its planning to remediate the areas off-site from the Fukushima Daiichi Nuclear Power Plant. Following a request by the Government of Japan, the mission, comprising 12 international and IAEA experts from several countries, will visit Japan between 7 and 15 October 2011 under the leadership of Mr. Juan Carlos Lentijo, General Director for Radiation Protection at Spain's nuclear regulatory authority. The team will go to several locations in the Fukushima Prefecture and conduct meetings in Tokyo with Japanese officials to: Provide assistance to Japan in its plans to manage remediation efforts; Review the country's remediation strategies, plans and work; and Share its findings with the international community. The IAEA mission will provide an opportunity for the international experts to exchange views with the Japanese authorities involved in the decontamination effort and other interested parties. It will also provide an opportunity for the IAEA to take stock of lessons learned from this important decontamination initiative. At the end of the mission a preliminary summary report will be provided to the Government of Japan and be made publically available. The team is also planning to hold a press briefing at the end of the mission. The final report of the mission will be presented to the Government in the month following the conclusion of the mission. Background The accident at Fukushima Daiichi Nuclear Power Plant has led to the radiological contamination of large areas. The Government of Japan has been formulating a strategy and plans to implement countermeasures to remediate these areas. The IAEA organized an International Fact Finding Expert Mission Of The Fukushima Daiichi Nuclear Power Plant Accident Following The Great East Japan Earthquake And Tsunami, which was held between 24 May and 2 June 2011. The current mission is a

  14. Nuclear power and the anti-uranium movement

    International Nuclear Information System (INIS)

    Titterton, E.W.

    1979-01-01

    The generation of electricity with nuclear power reactors is the cheapest, safest and cleanest means of power generation yet devised by man. It has been demonstrated to cause no more off-site nuclear radiation problems, and far less environmental damage, than the coal-fired electricity generating industry. With such a favourable situation it is reasonable to ask why, in the Western world, has a campaign been waged against uranium and the nuclear power industry. The author discusses what he believes is a well coordinated, international campaign which has been pursued through the media, and has been high-lighted by all the modern-day avenues open to protesters

  15. Improvements in or relating to handling of flue gas

    International Nuclear Information System (INIS)

    Ingham, R.V.

    1986-01-01

    The patent describes improvements in the method for handling flue gas from the burning of fossil fuels. The method relates to cleaning the flue gas, from which the sulphur compounds are removed. The gas in then heated by heat derived from a nuclear source, which may be nuclear waste. The heat treatment gives efficient atmospheric dispersion from the chimney. (U.K.)

  16. Estimation of the Radiological Consequences of Fukushima Dai-ichi Nuclear Power Plant Accident using MACCS2

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sora; Min, Byung-Il; Park, Kihyun; Yang, Byung-Mo; Suh, Kyung-suk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Three of them have undergone fuel melting and hydrogen explosions. A significant amount of radioactive material was released into the atmosphere from FDNPP and dispersed all over the world. In this study, we assessed the offsite consequences of Fukushima disaster in the region within a 30-km radius of FDNPP using the MELCOR Accident Consequence Code Systems 2(MACCS2) code, which is the Nuclear Regulatory Commission's (NRC's) code. The reflection of the realistic regional characteristics, such as long-term meteorological data, site- and population-specific data, and radiation safety regulatory, is essential to accurately analyze the off-site consequences. The assessment that reflects regional characteristics would contribute to identify main causes of exposure doses and to find the effective countermeasures for minimizing the accidental off-site consequences.

  17. Development of first full scope commercial CANDU-6 fuel handling simulator

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, W., E-mail: BCrawford@atlanticnuclear.ca [Atlantic Nuclear Services Inc., Fredericton, NB (Canada); McInerney, J. M., E-mail: JMcInerney@nbpower.com [Point Lepreau Generating Station, Maces Bay, NB (Canada); Moran, E.S.; Nice, J. W.; Sinclair, D.M.; Somerville, S.; Usalp, E.C.; Usalp, M., E-mail: EMoran@atlanticnuclear.ca, E-mail: JNice@atlanticnuclear.ca, E-mail: DSinclair@atlanticnuclear.ca, E-mail: SSomerville@atlanticnuclear.ca, E-mail: ECUsalp@atlanticnuclear.ca, E-mail: MUsalp@atlanticnuclear.ca [Atlantic Nuclear Services Inc., Fredericton, NB (Canada)

    2015-07-01

    Unique to CANDU reactors is continuous on-power refueling. In the CANDU-6 design, the fuel bundles are contained within 380 pressure tubes. Fuelling machines, one on either side of the reactor face move on a bridge and carriage system to the appointed channel and fuel under computer control. The fuelling machine is an immensely complicated mechanical device. None of the original Canadian full scope simulators incorporated the interaction of the fuel handling system. Traditionally, the final stages of Fuel Handling Operator qualification utilizes on the job training in a production environment carried out in the station main control room. For the purpose of supporting continual improvement in fuel handling training at the Third Qinshan Nuclear Plant Company (TQNPC), Atlantic Nuclear Services in a joint project with New Brunswick Power, developed the first commercial full scope CANDU-6 Fuel Handling simulator, integrated into the existing TQNPC Full Scope Simulator framework. The TQNPC Fuel Handling simulator is capable of supporting all normal on-power and off-power refuelling procedures as well as other abnormal operating conditions, which will allow training to be conducted, based on the plant specific operating procedures. This paper will discuss its development, the importance of this tool and its advantages over past training practices. (author)

  18. Development of first full scope commercial CANDU-6 fuel handling simulator

    International Nuclear Information System (INIS)

    Crawford, W.; McInerney, J. M.; Moran, E.S.; Nice, J. W.; Sinclair, D.M.; Somerville, S.; Usalp, E.C.; Usalp, M.

    2015-01-01

    Unique to CANDU reactors is continuous on-power refueling. In the CANDU-6 design, the fuel bundles are contained within 380 pressure tubes. Fuelling machines, one on either side of the reactor face move on a bridge and carriage system to the appointed channel and fuel under computer control. The fuelling machine is an immensely complicated mechanical device. None of the original Canadian full scope simulators incorporated the interaction of the fuel handling system. Traditionally, the final stages of Fuel Handling Operator qualification utilizes on the job training in a production environment carried out in the station main control room. For the purpose of supporting continual improvement in fuel handling training at the Third Qinshan Nuclear Plant Company (TQNPC), Atlantic Nuclear Services in a joint project with New Brunswick Power, developed the first commercial full scope CANDU-6 Fuel Handling simulator, integrated into the existing TQNPC Full Scope Simulator framework. The TQNPC Fuel Handling simulator is capable of supporting all normal on-power and off-power refuelling procedures as well as other abnormal operating conditions, which will allow training to be conducted, based on the plant specific operating procedures. This paper will discuss its development, the importance of this tool and its advantages over past training practices. (author)

  19. Research on crisis communication of nuclear and radiation safety

    International Nuclear Information System (INIS)

    Cao Yali; Zhang Ying

    2013-01-01

    Insufficient public cognition of nuclear and radiation safety and absence of effective method to handle crisis lead to common crisis events of nuclear and radiation safety, which brings about unfavorable impact on the sound development of nuclear energy exploring and application of nuclear technology. This paper, based on crisis communication theory, analyzed the effect of current situation on nuclear and radiation safety crisis, discussed how to handle crisis, and tried to explore the effective strategies for nuclear and radiation safety crisis handling. (authors)

  20. Rethinking the Business Model in Construction by the Use of Off-Site System Deliverance: Case of the Shaft Project

    DEFF Research Database (Denmark)

    Thuesen, Christian; Hvam, Lars

    2013-01-01

    This paper presents a set of insights to be used in the development of business models for off-site system deliveries contributing to the development of Off-Site Manufacturing practices (OSM). The theoretical offset for discussing the development of business models is the blue ocean strategy...... of installation shafts. Findings from the development and production of the installation shaft show that system deliveries represent a promising strategy for moving from red ocean competitive environment with the predominant cost+ business model, to a blue ocean situation in which the competition emerges...... in the constant pursue of value creation and cost reduction. On the basis of that system deliverances represent a promising strategy in the future development and application of off-site manufacturing practices. The application of system deliveries is however demanding as it represents a fundamental shift...

  1. Experience in handling concentrated tritium

    International Nuclear Information System (INIS)

    Holtslander, W.J.

    1985-12-01

    The notes describe the experience in handling concentrated tritium in the hydrogen form accumulated in the Chalk River Nuclear Laboratories Tritium Laboratory. The techniques of box operation, pumping systems, hydriding and dehydriding operations, and analysis of tritium are discussed. Information on the Chalk River Tritium Extraction Plant is included as a collection of reprints of papers presented at the Dayton Meeting on Tritium Technology, 1985 April 30 - May 2

  2. Resilience improvements to UK nuclear power plants

    International Nuclear Information System (INIS)

    Pepper, Kevin

    2015-01-01

    Following the events at Fukushima, the Office for Nuclear Regulation (ONR), the UK nuclear safety regulator, undertook a series of reviews into the resilience of UK nuclear power plants to severe events. These reviews highlighted a number of areas in relation to electrical infrastructure where it considered licensees should review their arrangements, considering both onsite and offsite infrastructure as well as the ability to recover following a complete loss of site infrastructure. In response, UK licensees have been exploring four parallel approaches to improving the resilience for each of their sites. Firstly, through modifications on-site such as enhancements to the installed diesel generators and related systems. Secondly through improvements to the resilience of essential instrumentation to Station Black Out events. Thirdly, through the provision of off-site backup equipment that can be deployed to any site following a severe event. Finally, the provision of event qualified connection points on site to enable timely restoration of long term essential electrical supplies and cooling to key systems. This last item gives a central focus to the issues of switchboard availability and the resilience of the whole network to large potentially common cause internal and external hazards. This paper will discuss the electrically related findings of the ONR reviews, explore the reasoning behind those decisions, and describe the approaches being taken by UK licensees. (authors)

  3. Economic risks of nuclear power reactor accidents

    International Nuclear Information System (INIS)

    Burke, R.P.; Aldrich, D.C.

    1984-04-01

    Models to be used for analyses of economic risks from events which occur during US LWR plant operation are developed in this study. The models include capabilities to estimate both onsite and offsite costs of LWR events ranging from routine plant forced outages to severe core-melt accidents resulting in large releases of radioactive material to the environment. The models have been developed for potential use by both the nuclear power industry and regulatory agencies in cost/benefit analyses for decision-making purposes. The new onsite cost models estimate societal losses from power production cost increases, plant capital losses, plant decontamination costs, and plant repair costs which may be incurred after LWR operational events. Early decommissioning costs, plant worker health impact costs, electric utility business costs, nuclear power industry costs, and litigation costs are also addressed. The newly developed offsite economic consequence models estimate The costs of post-accident population protective measures and public health impacts. The costs of population evacuation and temporary relocation, agricultural product disposal, land and property decontamination, and land interdiction are included in the economic models for population protective measures. Costs of health impacts and medical care costs are also included in the models

  4. A Bookless Library, Part I: Relocating Print Materials to Off-Site Storage

    Science.gov (United States)

    Sewell, Bethany B.

    2013-01-01

    This article presents an analysis of the feasibility of a bookless library in a research setting. As spaces for collections are being converted for increased study and community spaces, many libraries have been moving low-use collections to off-site storage. Issues regarding the types of storage spaces available are addressed. Concerns and…

  5. Off-site radiation exposure review project: computer-aided surface interpolation and graphical display

    International Nuclear Information System (INIS)

    Foley, T.A. Jr.

    1981-08-01

    This report presents the implementation of an iterative procedure that solves the following bivariate interpolation problem: Given N distinct points in the plane (x/sub i/, y/sub i/) and N real numbers Z/sub i/, construct a function F(x,y) that satisfies F(x/sub i/, y/sub i/) = Z/sub i/, for i = 1, ..., N. This problem can be interpreted as fitting a surface through N points in three dimensional space. The application of primary concern to the Offsite Radiation Exposure Review Project is the characterization of the radionuclide activity resulting from nuclear tests. Samples of activity were measured at various locations. The location of the sample point is represented by (x/sub i/, y/sub i/), and the magnitude of the reading is represented by Z/sub i/. The method presented in this report is constructed to be efficient on large data sets, stable on the large variations of the Z/sub i/ magnitudes, and capable of smoothly filling in areas that are void of data. This globally defined icode was initiateminednitial shock but to two later eriological invaders are Staphylococcus albus, Beta-hemolytic Streptococcus e to the same general semiclassical treatment

  6. Handling of final storage of unreprocessed spent nuclear fuel

    International Nuclear Information System (INIS)

    1978-01-01

    In this report the various facilities incorporated in the proposed handling chain for spent fuel from the power stations to the final repository are discribed. Thus the geological conditions which are essential for a final repository is discussed as well as the buffer and canister materials and how they contribute towards a long-term isolation of the spent fuel. Furthermore one chapter deals with leaching of the deposited fuel in the event that the canister is penetrated as well as the transport mechanisms which determine the migration of the radioactive substances through the buffer material. The dispersal processes in the geosphere and the biosphere are also described together with the transfer mechanisms to the ecological systems as well as radiation doses. Finally a summary is given of the safety analysis of the proposed method for the handling and final storage of the spent fuel. (E.R.)

  7. Insights into Architects’ Future Roles in Off-Site Construction

    Directory of Open Access Journals (Sweden)

    Jianing Luo

    2017-03-01

    Full Text Available Today’s construction industry is overflowing with new ideas about its future. Off-Site Manufacture and Construction (OSCM is at the heart of the modern construction industry. Much has been written about the state and context of OSCM in different countries regarding its perceived benefits and barriers to implementation. Off-site production (OSP plays an important role in improving fragmented construction processes. Although most OSP research targets the attitudes and practices of OSP adoption, there is limited understanding of the philosophical issues underpinning OSP-related architecture. The roles of the architects’ personal philosophies are neglected and this hampers their implementation of OSCM (which has had a largely technical focus. This paper explores the traditional thinking patterns of architects in China and predicts possible future roles for them. It then conceptualizes an “architectural work” mode and a “building product” mode of design and construction and identifies the shortcomings of architects in an OSCM environment. The arguments made are based on practitioners’ perceptions and the first author’s practical experiences of leading several real-life projects in recent years. The findings reveal the implications and significance of the transformation from an “architectural work” mode to a “building product” mode. We foresee a study approach that focuses on the order and rules for OSCM, resulting in architects’ existing mindsets being changed to thinking patterns and design methodologies better suited to OSCM.

  8. AMCO Off-Site Air Monitoring Map Service, Oakland CA, 2017, US EPA Region 9

    Data.gov (United States)

    U.S. Environmental Protection Agency — This map service contains a single layer: Off-Site Air Monitors. The layer draws at all scales. Full FGDC metadata for the layer may be found by clicking the layer...

  9. Suppliers' activities within the controlled zones of licensees handling ionizing radiation sources. Recommendations

    International Nuclear Information System (INIS)

    1999-01-01

    The Recommendations are intended to lay down a unified procedure for preparing licence applications related to ionizing radiation source handling, including the required documentation. The guidelines were set up based on documents of the Dukovany nuclear power plant and adapted to serve the Temelin nuclear power plant and other workplaces handling ionizing radiation sources as well. Selected provisions of applicable legislation are reproduced, and responsibilities are described. The major part of the publication is constituted by model documents, particularly a model Quality Assurance Programme. (P.A.)

  10. Development of remote handling tools for glove box

    International Nuclear Information System (INIS)

    Tomita, Yutaka; Nemoto, Takeshi; Denuma, Akio; Todokoro, Akio

    1996-01-01

    For a part of advanced nuclear fuel recycling technology development, we will separate and recover Americium from the MOX fuel scrap by solvent extraction. When we carry this examination, reduction of exposure from Americium-241 is one of important problems. To solve this problem fundamentally, we studied many joints type of the remote handling tools for glove box and produced a trial production machine. Also, we carried out basic function examinations of it. As a result, we got the prospect of development of the remote handling tools which could treat Americium in glove box. (author)

  11. Fissile materials principles of criticality safety in handling and processing

    International Nuclear Information System (INIS)

    1976-01-01

    This Swedish Standard consists of the English version of the International Standard ISO 1709-1975-Nuclear energy. Fissile materials. Principles of criticality safety in handling and processing. (author)

  12. Conceptual design report for a remotely operated cask handling system

    International Nuclear Information System (INIS)

    Yount, J.A.; Berger, J.D.

    Recent advances in remote handling utilizing commercial robotics are conceptually applied to the problem of lowering operator cumulative dose and increasing throughput during cask handling operations in proposed nuclear waste container shipping and receiving facilities. The functional criteria for each subsystem are defined, and candidate systems are described. The report also contains a generic description of a waste receiving facility, to show possible deployment configurations for the equipment

  13. Nuclear emergencies and protective actions

    International Nuclear Information System (INIS)

    Sjoeblom, Klaus

    1995-01-01

    Although technical improvements have increased the safety of new and old nuclear power plants, many simultaneous component failures and/or human errors are improbable but possible. Both the plant (on-site) and the nearby area (off-site) have emergency plans. Rescue service authorities are responsible of the off-site. The main protective actions are sheltering, evacuation and iodine ingestion. The Loviisa off-site emergency plan assumes that a major part of this population takes care of their own protective actions; Rescue service authorities can then concentrate on the coordination activities and to those people who need help. To be able to carry out the protective actions timely and effectively the people should have information on radiation risk and emergency planning. In case of a potential accident the local population should follow the rescue service information and know how to shelter and how to evacuate themselves. Though there are many stockpiles of iodine pellets in the area the rescue service authorities recommend that each household should purchase iodine pellets for their own need. The utility and the rescue service authorities have distributed information brochures to all homes within 30 km from Loviisa NPP since 1990. This brochure gives information on radiation and protective actions in case of an accident. Because the brochures might not stay available and so also the local telephone book contains this information

  14. Application of GO methodology in reliability analysis of offsite power supply of Daya Bay NPP

    International Nuclear Information System (INIS)

    Shen Zupei; Li Xiaodong; Huang Xiangrui

    2003-01-01

    The author applies the GO methodology to reliability analysis of the offsite power supply system of Daya Bay NPP. The direct quantitative calculation formulas of the stable reliability target of the system with shared signals and the dynamic calculation formulas of the state probability for the unit with two states are derived. The method to solve the fault event sets of the system is also presented and all the fault event sets of the outer power supply system and their failure probability are obtained. The resumption reliability of the offsite power supply system after the stability failure of the power net is also calculated. The result shows that the GO methodology is very simple and useful in the stable and dynamic reliability analysis of the repairable system

  15. Radiation emergency preparedness in nuclear power plants

    International Nuclear Information System (INIS)

    Geetha, P.V.; Ramamirtham, B.; Khot, P.

    2008-01-01

    The purpose of planning for radiation emergency response is to ensure adequate preparedness for protection of the plant personnel and members of the public from significant radiation exposures in the unlikely event of an accident. With a number of safety features in the reactor design and sound operating procedures, the probability of a major accident resulting in the releases of large quantities of radioactivity is extremely small. However, as an abundant cautious approach a comprehensive radiation emergency response preparedness is in place in all the nuclear power plants (NPPs). Radiation Emergency in NPPs is broadly categorized into three types; plant emergency, site emergency and off-site emergency. During off site emergency conditions, based on levels of radiation in the environment, Civil Authorities may impose several counter measures such as sheltering, administering prophylaxis (stable iodine for thyroid blocking) and evacuation of people from the affected area. Environmental Survey Laboratory (ESL) carries out environmental survey extensively in the affected sector identified by the meteorological survey laboratory. To handle emergency situations, Emergency Control Centre with all communication facility and Emergency Equipment Centre having radiation measuring instruments and protective equipment are functional at all NPPs. AERB stipulates certain periodicity for conducting the exercises on plant, site and off site emergency. These exercises are conducted and deficiencies corrected for strengthening the emergency preparedness system. In the case of off site emergency exercise, observers are invited from AERB and Crisis Management Group of Department of Atomic Energy (DAE). The emergency exercises conducted by Nuclear Power Plant Sites have been very satisfactory. (author)

  16. Effect of source term composition on offsite doses

    International Nuclear Information System (INIS)

    Karahalios, P.; Gardner, R.

    1985-01-01

    The development of new realistic accident source terms has identified the need to establish a basis for comparing the impact of such source terms. This paper attempts to develop a generalized basis of comparison by investigating contributions to offsite acute whole body doses from each group of radionuclides being released to the atmosphere, using CRAC2. The paper also investigates the effect of important parameters such as regional meteorology, sheltering, and duration of release. Finally, the paper focuses on significant changes in the relative importance of individual radionuclide groups in PWR2, SST1, and a revision of the Stone and Webster proposed interim source term

  17. Organisation and tasks concerning population protection in nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Nikolov, N [Civil Defence Administration, Sofia (Bulgaria)

    1996-12-31

    The structure and the scope of activities of Civil Defense Administration of Bulgaria are briefly outlined. Protection of general public in the case of nuclear accident will be provided according to a special off-site emergency plan. The plan is based on results of research and development approved by the permanent commission for public protection. 3 refs.

  18. Organisation and tasks concerning population protection in nuclear accident

    International Nuclear Information System (INIS)

    Nikolov, N.

    1995-01-01

    The structure and the scope of activities of Civil Defense Administration of Bulgaria are briefly outlined. Protection of general public in the case of nuclear accident will be provided according to a special off-site emergency plan. The plan is based on results of research and development approved by the permanent commission for public protection. 3 refs

  19. A radiation monitoring system for nuclear power plants

    International Nuclear Information System (INIS)

    Iwai, Masaru; Nakamori, S.; Ikeda, H.; Oda, M.

    1974-01-01

    Safety with respect to radiation is vital factor, particularly in view of the increasing number of nuclear power plants. For this purpose, a radiation monitoring system is provided to perform constant supervision. This article describes the purpose, installation location, specifications and circuitry of a system which is divided into three units: the process monitor, area monitor and off-site monitor. (auth.)

  20. Fluid management plan for the Project Shoal Area Offsites Subproject

    International Nuclear Information System (INIS)

    1996-08-01

    The US Department of Energy, Nevada Operations Office (DOE/NV) has initiated the Offsites Subproject to characterize the hazards posed to human health and the environment as a result of underground nuclear testing activities at facilities other than the Nevada Test Site (NTS). A primary Subproject objective is to gather adequate data to characterize the various Subproject sites through the collection of surface and subsurface soil samples and by drilling several wells for the collection of groundwater data. The Project Shoal Area (PSA) is one of the Subproject's Nevada sites and is subject to the requirements set forth in the Federal Facility Compliance Agreement and Consent Order (FFACO) (DOE, 1996a). In accordance with the FFACO, a Corrective Action Investigation Plan (CAIP) has been developed for work at the PSA (designated as Corrective Action Unit Number 416). This Fluid Management Plan (FMP) provides guidance for the management of fluids generated from wells constructed at the PSA. Long-term monitoring and future activities at the site, if required, will be set forth in additional documents as required by the FFACO. The ultimate method for disposition of fluids generated by site operations depends upon sample analysis and process knowledge in relation to fluid management criteria. Section 2 describes well site operations; Section 3 discusses fluid management criteria; Section 4 includes the fluid monitoring program; Section 5 presents the fluid management strategy; Section 6 provides for fluid management during routine well monitoring; and Section 7 contains reporting criteria

  1. Off-site monitoring for the Mighty Oak nuclear test

    International Nuclear Information System (INIS)

    Black, S.C.; Smith, A.E.; Costa, C.F.

    1986-07-01

    After a nuclear explosives test, code name Mighty Oak, the tunnel leading to the test point became contaminated with radioactive debris. To re-enter and recover valuable equipment and data, the DOE purged the tunnel air using particulate and charcoal filters to minimize discharge of radioactivity to the atmosphere. During this purging, the EPA established special air samples supplementing their routine air monitoring networks. Analysis of the collected samples for radioactive noble gases and for gamma-emitting radionuclides indicated that only low levels of xenon-133 were released in amounts detectable in populated areas near the Nevada Test Site. The maximum dose to an individual was calculated to be 0.36 microrem, assuming that person remained in the open field at the measurement site during the whole period of the purging

  2. 40 CFR 300.440 - Procedures for planning and implementing off-site response actions.

    Science.gov (United States)

    2010-07-01

    ... unless the Remedial Project Manager or OSC assures the proper management of the CERCLA waste samples or... OIL AND HAZARDOUS SUBSTANCES POLLUTION CONTINGENCY PLAN Hazardous Substance Response § 300.440... § 300.440(d). (5) Off-site transfers of those laboratory samples and treatability study CERCLA wastes...

  3. 3D Analysis of Cooling Performance with Loss of Offsite Power Using GOTHIC Code

    International Nuclear Information System (INIS)

    Oh, Kye Min; Heo, Gyun Young; Na, In Sik; Choi, Yu Jung

    2010-01-01

    GOTHIC code enables to analyze one-dimensional or multi-dimensional problems for evaluating the cooling performance of loss of offsite power. The conventional GOTHIC code analysis performs heat transfer between plant containment and the outside of the fan cooler tubes by modeling each of fan cooler part model and component cooling water inside tube each to analyze boiling probability. In this paper, we suggest a way which reduces the multi-procedure of the cooling performance with loss of offsite power or the heat transfer states with complex geometrical structure to a single-procedure and verify the applicability of the heat transfer differences from the containment atmosphere humidity changes by the multi-nodes which component cooling water of tube or air of Reactor Containment Fan Cooler in the containment, otherwise the component model uses only one node

  4. Developments in modelling the economic impact of Off-site accident consequences

    International Nuclear Information System (INIS)

    Haywood, S.M.; Robinson, C.A.; Faude, D.

    1991-01-01

    Models for assessing the economic consequences of accidental releases of radioactivity have application both in accident consequence codes and in decision aiding computer systems for use in emergency response. Such models may be applied in emergency planning, and studies in connection with the siting, design and licensing of nuclear facilities. Several models for predicting economic impact have been developed, in Europe and the US, and these are reviewed. A new model, called COCO-1 (Cost of Consequences Off-site), has been developed under the CEC MARIA programme and the features of the model are summarised. The costs calculated are a measure of the benefit foregone as a result of the accident, and in addition to tangible monetary costs the model attempts to include costs arising from the effect of the accident on individuals, for instance the disruption caused by the loss of homes. COCO-1 includes the cost of countermeasures, namely evacuation, relocation, sheltering, food restrictions and decontamination, and also the cost of health effects in the exposed population. The primary quantity used in COCO-1 to measure the economic value of land subject to restrictions on usage is Gross Domestic Product (GDP). Examples of default data included in the model are presented, as are the results of an illustrative application. The limitations of COCO-1 are discussed, and areas where further data are needed are identified

  5. Nuclear Technology Series. Course 25: Radioactive Material Handling Techniques.

    Science.gov (United States)

    Center for Occupational Research and Development, Inc., Waco, TX.

    This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutions in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

  6. Practical use of offsite atmospheric measurements to enhance profitability of onsite wind prediction

    Energy Technology Data Exchange (ETDEWEB)

    Collier, Craig [GL Garrad Hassan (Canada)

    2011-07-01

    This paper presents the use of offsite atmospheric measurements to improve the profitability of onsite wind prediction. There are two common sensitivities used, intraday and interday. Results from US mid-western sites show that the error associated with wind predictions is large but there are possibilities for improvement. Inter- and intraday can be used traditionally to contribute towards NWP bias correction. Intraday alone can be used with machine learning and NWP. These techniques are compared and given in order of ease of use and potential accuracy gains. Some considerations and differences for all three techniques, namely, traditional, data assimilation and machine learning are also detailed. An offsite selection matrix shows how elements like location, geography and telemetry rate in the 3 techniques. The experimental setup for all 3 techniques over a 3-month period is given and the results are presented. It can be concluded that the results from these simple experiments show promise but vary in method and time scale.

  7. Handling radioactivity: a practical approach for scientists and engineers

    International Nuclear Information System (INIS)

    Stewart, D.C.

    1981-01-01

    The aim of this book is to present an overall view in a descriptive and essentially nonmathematical way of the practicalities of handling radioactivity. It is hoped that the material will be particularly helpful to those entering the nuclear field for the first time and to those working in related areas whose responsibilities require them to have a general knowledge of the subject of radioactivity handling and its vocabulary. The presentation is primarily for bench-scale operations. There is a considerable emphasis on facilities since these are fundamental to the safe handling of active materials. Facility design and detail is also unfortunately an area where the relevant information is largely scattered through literature sources that are not accessible to most readers. Some of the topics surveyed - such as dosimetry, shielding and nuclear criticality - are extremely complex and no pretense is made that the treatment here represents more than bare bone summaries of the fields. A considerable effort has been made to cite the key references in each area where more detailed information can be found. A few additional useful references not cited directly in the text appear in an abbreviated bibliography at the end of the book

  8. Reviewing reactor engineering and fuel handling

    International Nuclear Information System (INIS)

    1991-12-01

    Experience has shown that the better operating nuclear power plants have well defined and effectively administered policies and procedures for governing reactor engineering and fuel handling (RE and FH) activities. This document provides supplementary guidance to OSART experts for evaluating the RE and FH programmes and activities at a nuclear power plant and assessing their effectiveness and adequacy. It is in no way intended to conflict with existing regulations and rules, but rather to exemplify those characteristics and features that are desirable for an effective, well structured RE and FH programme. This supplementary guidance addresses those aspects of RE and FH activities that are required in order to ensure optimum core operation for a nuclear reactor without compromising the limits imposed by the design, safety considerations of the nuclear fuel. In the context of this document, reactor engineering refers to those activities associated with in-core fuel and reactivity management, whereas fuel handling refers to the movement, storage, control and accountability of unirradiated and irradiated fuel. The document comprises five main sections and several appendices. In Section 2 of this guide, the essential aspects of an effective RE and FH programme are discussed. In Section 3, the various types of documents and reference materials needed for the preparatory work and investigation are listed. In Section 4, specific guidelines for investigation of RE and FH programmes are presented. In Section 5, the essential attributes of an excellent RE and FH programme are listed. The supplementary guidance is concluded with a series of appendices exemplifying the various qualities and attributes of a sound, well defined RE and FH programme

  9. The main ecological principles of ensuring safety of man and biosphere in the handling of radioactive wastes

    International Nuclear Information System (INIS)

    Kryshev, I.I.; Sazykina, T.G.

    1999-01-01

    This paper provides an assessment of ecological safety in the handling of radioactive wastes in the territory of Russia. The following problems are considered: the main sources of radioactive wastes and spent nuclear fuel; assessments of collective dose from the enterprises of the nuclear fuel cycle in Russia; and principles and criteria for ensuring ecological safety when handling radioactive wastes

  10. Data handling with SAM and art at the NOνA experiment

    International Nuclear Information System (INIS)

    Aurisano, A; Backhouse, C; Davies, G S; Illingworth, R; Mengel, M; Norman, A; Mayer, N; Rocco, D; Zirnstein, J

    2015-01-01

    During operations, NOvA produces between 5,000 and 7,000 raw files per day with peaks in excess of 12,000. These files must be processed in several stages to produce fully calibrated and reconstructed analysis files. In addition, many simulated neutrino interactions must be produced and processed through the same stages as data. To accommodate the large volume of data and Monte Carlo, production must be possible both on the Fermilab grid and on off-site farms, such as the ones accessible through the Open Science Grid. To handle the challenge of cataloging these files and to facilitate their off-line processing, we have adopted the SAM system developed at Fermilab. SAM indexes files according to metadata, keeps track of each file's physical locations, provides dataset management facilities, and facilitates data transfer to off-site grids. To integrate SAM with Fermilab's art software framework and the NOvA production workflow, we have developed methods to embed metadata into our configuration files, art files, and standalone ROOT files. A module in the art framework propagates the embedded information from configuration files into art files, and from input art files to output art files, allowing us to maintain a complete processing history within our files. Embedding metadata in configuration files also allows configuration files indexed in SAM to be used as inputs to Monte Carlo production jobs. Further, SAM keeps track of the input files used to create each output file. Parentage information enables the construction of self-draining datasets which have become the primary production paradigm used at NOvA. In this paper we will present an overview of SAM at NOvA and how it has transformed the file production framework used by the experiment. (paper)

  11. Developing Off-site Emergency Preparedness and Response Model (OEPRM) for Severe Accident of NPP in a Densely Populated Country Using System Dynamics Approach

    Energy Technology Data Exchange (ETDEWEB)

    Hossena, Muhammed Mufazzal; Kang, Kyoung Ho; Song, Jin Ho [KAERI, Deajeon (Korea, Republic of)

    2016-05-15

    The main objectives of this study are to find the influencing factors of systems and sub-systems of OEPRM, to find the interdependency among the influencing factors, and to develop a conceptual qualitative OEPRM for densely populated NPP country in case of SA using system dynamics (SD). NPP accidents are classified as nuclear accidents and incidents depending on the severity. Severe accident (SA) is certain low probability accident that are beyond design basis accident which may arise due to multiple failures of safety systems leading to significant core degradation and jeopardize the integrity of many or all of the barriers to the release of radioactive material. The weakness to the off-site emergency response in the time of Fukushima accident was observed. So, it is crucial to develop an off-site emergency preparedness and responses model (OEPRM) for radiological emergency in densely populated country from the Fukushima emergency response lesson. In this study, an OEPRM is developed for densely populated NPP country to mitigate radiological effects in case of SA using SD approach. Besides, this study focuses the weakness of emergency response in Fukushima accident and proposed solution approach. The development of OEPRM in case of SA of NPP is very complex because of the involvement of various organization and it requires highly specialized agencies and technical experts. Moreover, if the country is agriculture based, it will make completely sophisticated.

  12. Developing Off-site Emergency Preparedness and Response Model (OEPRM) for Severe Accident of NPP in a Densely Populated Country Using System Dynamics Approach

    International Nuclear Information System (INIS)

    Hossena, Muhammed Mufazzal; Kang, Kyoung Ho; Song, Jin Ho

    2016-01-01

    The main objectives of this study are to find the influencing factors of systems and sub-systems of OEPRM, to find the interdependency among the influencing factors, and to develop a conceptual qualitative OEPRM for densely populated NPP country in case of SA using system dynamics (SD). NPP accidents are classified as nuclear accidents and incidents depending on the severity. Severe accident (SA) is certain low probability accident that are beyond design basis accident which may arise due to multiple failures of safety systems leading to significant core degradation and jeopardize the integrity of many or all of the barriers to the release of radioactive material. The weakness to the off-site emergency response in the time of Fukushima accident was observed. So, it is crucial to develop an off-site emergency preparedness and responses model (OEPRM) for radiological emergency in densely populated country from the Fukushima emergency response lesson. In this study, an OEPRM is developed for densely populated NPP country to mitigate radiological effects in case of SA using SD approach. Besides, this study focuses the weakness of emergency response in Fukushima accident and proposed solution approach. The development of OEPRM in case of SA of NPP is very complex because of the involvement of various organization and it requires highly specialized agencies and technical experts. Moreover, if the country is agriculture based, it will make completely sophisticated

  13. Hot Laboratories and Remote Handling

    International Nuclear Information System (INIS)

    Bart, G.; Blanc, J.Y.; Duwe, R.

    2003-01-01

    The European Working Group on ' Hot Laboratories and Remote Handling' is firmly established as the major contact forum for the nuclear R and D facilities at the European scale. The yearly plenary meetings intend to: - Exchange experience on analytical methods, their implementation in hot cells, the methodologies used and their application in nuclear research; - Share experience on common infrastructure exploitation matters such as remote handling techniques, safety features, QA-certification, waste handling; - Promote normalization and co-operation, e.g., by looking at mutual complementarities; - Prospect present and future demands from the nuclear industry and to draw strategic conclusions regarding further needs. The 41. plenary meeting was held in CEA Saclay from September 22 to 24, 2003 in the premises and with the technical support of the INSTN (National Institute for Nuclear Science and Technology). The Nuclear Energy Division of CEA sponsored it. The Saclay meeting was divided in three topical oral sessions covering: - Post irradiation examination: new analysis methods and methodologies, small specimen technology, programmes and results; - Hot laboratory infrastructure: decommissioning, refurbishment, waste, safety, nuclear transports; - Prospective research on materials for future applications: innovative fuels (Generation IV, HTR, transmutation, ADS), spallation source materials, and candidate materials for fusion reactor. A poster session was opened to transport companies and laboratory suppliers. The meeting addressed in three sessions the following items: Session 1 - Post Irradiation Examinations. Out of 12 papers (including 1 poster) 7 dealt with surface and solid state micro analysis, another one with an equally complex wet chemical instrumental analytical technique, while the other four papers (including the poster) presented new concepts for digital x-ray image analysis; Session 2 - Hot laboratory infrastructure (including waste theme) which was

  14. Safety analysis, 200 Area, Savannah River Plant: Separations area operations. Receiving Basin for Offsite Fuel (Supplement 3)

    Energy Technology Data Exchange (ETDEWEB)

    Allen, P M

    1983-09-01

    Analysis of the Savannah River Plant RBOF and RRF included an evaluation of the reliability of process equipment and controls, administrative controls, and engineered safety features. The evaluation also identified potential scenarios and radiological consequences. Risks were calculated in terms of 50-year population dose commitment per year (man-rem/year) to the onsite and offsite population within an 80 Km radius of RBOF and RRF, and to an individual at the plant boundary. The total 50-year onsite and offsite population radiological risks of operating the RBOF and RRF were estimated to be 1.0 man-rem/year. These risks are significantly less than the population dose of 54,000 man/rem/yr for natural background radiation in a 50-mile radius. The 50-year maximum offsite individual risk from operating the facility was estimated to be 2.1 {times} 10{sup 5} rem/yr. These risks are significantly lower than 93 mrem/yr an individual is expected to receive from natural background radiation in this area. The analysis shows. that the RBOF and RRF can be operated without undue risk to onsite personnel or to the general public.

  15. A New Format for Handling Nuclear Data

    CERN Document Server

    Bak, S I; Tenreiro, C; Kadi, Y; Hong, S W; Manchanda, V; Gheata, M; Chai, J S; Carminati, F; Park, T S; Brun, R

    2011-01-01

    The ASCII ENDF format for nuclear data has been used for four decades. It is practical for human inspection and portability, but; it is not very effective for manipulating and displaying the data or for using them in Monte-Carlo applications. In this paper we present a prototype of a nuclear data manipulation package (TNudy) based on the ROOT system (http://root.cern.ch). The ROOT object-oriented C++ framework is the de-facto standard in high energy and nuclear physics since ten years. Starting from the ENDF format, the data. is stored in machine-portable binary format. Root files also offer a powerful direct access capability to their different sections and compressibility upon writing, minimising the disk occupancy. ROOT offers a complete library of visualisation and mathematical routines and the Virtual Monte-Carlo system, which allows running different transport Monte-Carlo (Geant 4, Geant 3) with common scoring and geometry modellers, which comes as part of ROOT. ROOT contains isotope decay data and the ...

  16. Application service provider (ASP) financial models for off-site PACS archiving

    Science.gov (United States)

    Ratib, Osman M.; Liu, Brent J.; McCoy, J. Michael; Enzmann, Dieter R.

    2003-05-01

    For the replacement of its legacy Picture Archiving and Communication Systems (approx. annual workload of 300,000 procedures), UCLA Medical Center has evaluated and adopted an off-site data-warehousing solution based on an ASP financial with a one-time single payment per study archived. Different financial models for long-term data archive services were compared to the traditional capital/operational costs of on-site digital archives. Total cost of ownership (TCO), including direct and indirect expenses and savings, were compared for each model. Financial parameters were considered: logistic/operational advantages and disadvantages of ASP models versus traditional archiving systems. Our initial analysis demonstrated that the traditional linear ASP business model for data storage was unsuitable for large institutions. The overall cost markedly exceeds the TCO of an in-house archive infrastructure (when support and maintenance costs are included.) We demonstrated, however, that non-linear ASP pricing models can be cost-effective alternatives for large-scale data storage, particularly if they are based on a scalable off-site data-warehousing service and the prices are adapted to the specific size of a given institution. The added value of ASP is that it does not require iterative data migrations from legacy media to new storage media at regular intervals.

  17. Disposal and handling of nuclear steam generator chemical cleaning wastes

    International Nuclear Information System (INIS)

    Larrick, A.P.; Schneidmiller, D.

    1978-01-01

    A large number of pressurized water nuclear reactor electrical generating plants have experienced a corrosion-related problem with their steam generators known as denting. Denting is a mechanical deformation of the steam generator tubes that occurs at the tube support plates. Corrosion of the tube support plates occurs within the annuli through which the tubes pass and the resulting corrosion oxides, which are larger in volume than the original metal, compress and deform the tubes. In some cases, the induced stresses have been severe enough to cause tube and/or support cracking. The problem was so severe at the Turkey Point and Surrey plants that the tubing is being replaced. For less severe cases, chemical cleaning of the oxides, and other materials which deposit in the annuli from the water, is being considered. A Department of Energy-sponsored program was conducted by Consolidated Edison Co. of New York which identified several suitable cleaning solvents and led to in-plant chemical cleaning pilot demonstrations in the Indian Point Unit 1 steam generators. Current programs to improve the technology are being conducted by the Electric Power Research Institute, and the three PWR NSSS vendors with the assistance of numerous consultants, vendors, and laboratories. These programs are expected to result in more effective, less corrosive solvents. However, after a chemical cleaning is conducted, a large problem still remains- that of disposing of the spent wastes. The paper summarizes some of the methods currently available for handling and disposal of the wastes

  18. 340 Waste Handling Facility interim safety basis

    International Nuclear Information System (INIS)

    Bendixsen, R.B.

    1995-01-01

    This document establishes the interim safety basis (ISB) for the 340 Waste Handling Facility (340 Facility). An ISB is a documented safety basis that provides a justification for the continued operation of the facility until an upgraded final safety analysis report is prepared that complies with US Department of Energy (DOE) Order 5480.23, Nuclear Safety Analysis Reports. The ISB for the 340 Facility documents the current design and operation of the facility. The 340 Facility ISB (ISB-003) is based on a facility walkdown and review of the design and operation of the facility, as described in the existing safety documentation. The safety documents reviewed, to develop ISB-003, include the following: OSD-SW-153-0001, Operating Specification Document for the 340 Waste Handling Facility (WHC 1990); OSR-SW-152-00003, Operating Limits for the 340 Waste Handling Facility (WHC 1989); SD-RE-SAP-013, Safety Analysis Report for Packaging, Railroad Liquid Waste Tank Cars (Mercado 1993); SD-WM-TM-001, Safety Assessment Document for the 340 Waste Handling Facility (Berneski 1994a); SD-WM-SEL-016, 340 Facility Safety Equipment List (Berneski 1992); and 340 Complex Fire Hazard Analysis, Draft (Hughes Assoc. Inc. 1994)

  19. Handling of tritium-bearing wastes

    International Nuclear Information System (INIS)

    1981-01-01

    The generation of nuclear power and reprocessing of nuclear fuel results in the production of tritium and the possible need to control the release of tritium-contaminated effluents. In assessing the need for controls, it is necessary to know the production rates of tritium at different nuclear facilities, the technologies available for separating tritium from different gaseous and liquid streams, and the methods that are satisfactory for storage and disposal of tritiated wastes. The intention in applying such control technologies and methods is to avoid undesirable effects on the environment, and to reduce the radiation burden on operational personnel and the general population. This technical report is a result of the IAEA Technical Committee Meeting on Handling of Tritium-bearing Effluents and Wastes, which was held in Vienna, 4 - 8 December 1978. It summarizes the main topics discussed at the meeting and appends the more detailed reports on particular aspects that were prepared for the meeting by individual participants

  20. The World Nuclear Industry Status Report 2017

    International Nuclear Information System (INIS)

    Schneider, Mycle; Froggatt, Antony; Hazemann, Julie; Katsuta, Tadahiro; Ramana, M.V.; Rodriguez, Juan C.; Ruedinger, Andreas; Stienne, Agnes

    2017-09-01

    The World Nuclear Industry Status Report 2017 (WNISR2017) provides a comprehensive overview of nuclear power plant data, including information on operation, production and construction. The WNISR assesses the status of new-build programs in current nuclear countries as well as in potential newcomer countries. The WNISR2017 edition includes a new assessment from an equity analyst view of the financial crisis of the nuclear sector and some of its biggest industrial players. The Fukushima Status Report provides not only an update on onsite and offsite issues six years after the beginning of the catastrophe, but also the latest official and new independent cost evaluations of the disaster. Focus chapters provide in-depth analysis of France, Japan, South Korea, the United Kingdom and the United States. The Nuclear Power vs. Renewable Energy chapter provides global comparative data on investment, capacity, and generation from nuclear, wind and solar energy. Finally, Annex 1 presents a country-by-country overview of all other countries operating nuclear power plants

  1. Ontario Hydro Pickering Generating Station fuel handling system performance

    International Nuclear Information System (INIS)

    Underhill, H.J.

    1986-01-01

    The report briefly describes the Pickering Nuclear Generating Station (PNGS) on-power fuel handling system and refuelling cycle. Lifetime performance parameters of the fuelling system are presented, including station incapability charged to the fuel handling system, cost of operating and maintenance, dose expenditure, events causing system unavailability, maintenance and refuelling strategy. It is concluded that the 'CANDU' on-power fuelling system, by consistently contributing less than 1% to the PNGS incapability, has been credited with a 6 to 20% increase in reactor capacity factor, compared to off-power fuelling schemes. (author)

  2. Conceptual design report for a remotely operated cask handling system. Revision 1

    International Nuclear Information System (INIS)

    Yount, J.A.; Berger, J.D.

    1984-09-01

    Recent advances in remote handling utilizing commercial robotics are conceptually applied to lowering operator cumulative radiation exposure and increasing throughput during cask handling operations in nuclear shipping and receiving facilities. Revision 1 incorporates functional criteria for facility equipment, equipment technical outline specifications, and interface control drawings to assist Architect Engineers in the application of remote handling to waste shipping and receiving facilities. The document has also been updated to show some of the equipment used in proof-of-principle testing during fiscal year 1984. 10 references, 50 figures, 1 table

  3. Nuclear power manpower and training requirements

    International Nuclear Information System (INIS)

    Whan, G.A.

    1984-01-01

    A broad spectrum of technical personnel is required to conduct a national nuclear power program, predominantly electrical, mechanical, and nuclear engineers and health physicists. The need for nuclear education and training, even in the early planning states, is the topic of this paper. Experience gained in the United States can provide useful information to Asia-Pacific countries developing nuclear power programs. Including both on-site and off-site personnel, U.S. plants average about 570 workers for BWRs and 700 for PWRs. The need for an additional 57,000 technical employees over the next decade is projected. The technical backgrounds of the manpower required to operate and support a nuclear power plant are distinctly different from those used by non-nuclear utilities. Manpower cannot be transferred from fossil fuel plants without extensive training. Meeting the demand for nuclear education and training must be a friendly partnership among universities, government, and industry. The long-term supply of nuclear-educated personnel requires strong, government-supported universities. Most specific training, however, must be provided by industry. (author)

  4. Final environmental impact statement for the Nevada Test Site and off-site locations in the State of Nevada. Transportation study, Volume 1, Appendix I

    International Nuclear Information System (INIS)

    1996-08-01

    This report has been prepared to address local transportation issues concerning current and potential operations at the Nevada Test Site (NTS), to document the results of the NTS transportation risk analysis, and to provide information and supporting documentation for the Environmental Impact Statement (EIS) for the NTS and Off-Site Locations in the State of Nevada. Four alternatives are evaluated in the NTS EIS: Alternative 1, Continue Current Operations, (No Action); Alternative 2, Discontinue Operations; Alternative 3, Expanded Use; and Alternative 4, Alternate Use of Withdrawn Lands. The transportation risk analysis estimated the health risk from highway transportation of DOE-generated low-level waste, mixed waste, and defense-related nuclear materials for each of the four alternatives

  5. Environmental Assessment for the off-site commercial cleaning of lead and asbestos contaminated laundry from the Savannah River Site

    International Nuclear Information System (INIS)

    1995-12-01

    This Environmental Assessment (EA) has been prepared by the Department of Energy (DOE) to assess the potential environmental impacts of off-site commercial cleaning of lead and asbestos contaminated laundry generated at the Savannah River Site (SRS), located near Aiken, South Carolina. The proposed action constitutes an addition to the already-implemented action of sending controlled and routine SRS laundry to an off-site commercial facility for cleaning. This already-implemented action was evaluated in a previous EA (i.e., DOE/EA-0990; DOE, 1994) prepared under the National Environmental Policy Act of 1969 (NEPA)

  6. Emergency power systems at nuclear power plants

    International Nuclear Information System (INIS)

    1982-01-01

    This Guide applies to nuclear power plants for which the total power supply comprises normal power supply (which is electric) and emergency power supply (which may be electric or a combination of electric and non-electric). In its present form the Guide provides general guidance for all types of emergency power systems (EPS) - electric and non-electric, and specific guidance (see Appendix A) on the design principles and the features of the emergency electric power system (EEPS). Future editions will include a second appendix giving specific guidance on non-electric power systems. Section 3 of this Safety Guide covers information on considerations that should be taken into account relative to the electric grid, the transmission lines, the on-site electrical supply system, and other alternative power sources, in order to provide high overall reliability of the power supply to the EPS. Since the nuclear power plant operator does not usually control off-site facilities, the discussion of methods of improving off-site reliability does not include requirements for facilities not under the operator's control. Sections 4 to 11 of this Guide provide information, recommendations and requirements that would apply to any emergency power system, be it electric or non-electric

  7. Development of CSA N1600-14: general requirements for nuclear emergency management programs

    Energy Technology Data Exchange (ETDEWEB)

    Sellar, C. [Canadian Standards Association Group, Mississauga, ON (Canada); Coles, J. [Ontario Power Generation, Darlington, ON (Canada)

    2014-07-01

    CSA Group has published a new standard on General requirements for nuclear emergency management programs (CSA N1600-14). The standard establishes criteria for the emergency management programs of on- and off-site organizations to address nuclear emergencies at Canadian nuclear power plants (NPPs). It provides the requirements to develop, implement, evaluate, maintain, and continuously improve a nuclear emergency management program for prevention and mitigation, preparedness, response, and recovery from a nuclear emergency at a NPP. This paper discusses the development of the standard, and provides the key drivers, structure, scope, and outline of the standard, while highlighting key features, impacts, and benefits. (author)

  8. Examples of remote handling of irradiated fuel assemblies in Germany

    International Nuclear Information System (INIS)

    Peehs, M.; Knecht, K.

    1999-01-01

    Examples for the remote handling of irradiated fuel in Germany are presented in the following areas: - fuel assembling pool service activities; - early encapsulation of spent fuel in the pool of a nuclear power plant (NPP) at the end of the wet storage period. All development in remote fuel assembly handling envisages minimization of the radioactive dose applied to the operating staff. In the service area a further key objective for applying advanced methods is to perform the work faster and at a higher quality standard. The early encapsulation is a new technology to provide the final packaging of spent fuel already in the pool of a NPP to ensure reliable handling for all further back end processes. (author)

  9. Off-site emergency planning in Czech Republic

    International Nuclear Information System (INIS)

    Prouza, Z.; Drabova, D.

    1996-01-01

    In the Czech Republic, the NPP Dukovany - PWR 440/213-type (4 blocks) is currently in operation (from 1985) and NPP Temelin - PWR 1000 (2 blocks) is under construction. Radiation accident on the NPP is defined as an unexpected or unintentional event in a facility with a potential of off-site consequences. The principles of emergency planning in Czech Republic now are based on the philosophy and principles described in the ICRP Publication 40 and the IAEA Safety Series No. 55, 72, and includes already the post Chernobyl experiences. Nevertheless, Czech Republic legislation experiences an extensive reconstruction. The Atomic Act, which will be based from point of view the structure, philosophy and principles on new International Basic Safety Standards, already being elaborated. That acts and related laws should solve our legislative problems on field of emergency planning and preparedness

  10. INES: The International Nuclear Event Scale user's manual

    International Nuclear Information System (INIS)

    1992-09-01

    The revised and extended addition of the International Nuclear Event Scale is presented. The manual is comprised of four parts. Part 1 contains a summary of the basis of the scale and of the procedure to be used for rating events. Part 2 contains the detailed guidance required to rate events in terms of off-site and on-site impact. These two parts are applicable to all nuclear facilities. Parts 3 and 4 contain the detailed guidance required to rate events in terms of defence in depth for reactors and other facilities, respectively. 5 figs, 2 tabs

  11. Nuclear power plant status diagnostics using artificial neural networks

    International Nuclear Information System (INIS)

    Bartlett, E.B.; Uhrig, R.E.

    1991-01-01

    In this work, the nuclear power plant operating status recognition issue is investigated using artificial neural networks (ANNs). The objective is to train an ANN to classify nuclear power plant accident conditions and to assess the potential of future work in the area of plant diagnostics with ANNS. To this end, an ANN was trained to recognize normal operating conditions as well as potentially unsafe conditions based on nuclear power plant training simulator generated accident scenarios. These scenarios include; hot and cold leg loss of coolant, control rod ejection, loss of offsite power, main steam line break, main feedwater line break and steam generator tube leak accidents. Findings show that ANNs can be used to diagnose and classify nuclear power plant conditions with good results

  12. Potential exposure in nuclear safety. INSAG-9. A report by the International Nuclear Safety Advisory Group

    International Nuclear Information System (INIS)

    1995-01-01

    The report defines potential exposure in terms of probability of its occurrence and possible consequences. Individual risk is expressed as the probability of the exposure and the conditional probability of death resulting from the exposure. Societal risk is more than the sum of individual risks. This report explores the relationship between individual risk and societal risk and the relevant criteria. For accidents causing serious damage to a nuclear power plant or having off-site consequences, individual risk is not sufficiently limiting because of the many aspects of societal impact. The approach to dealing with potential exposures in nuclear safety results in risks that are consistent with or more stringent than the ICRP's recommendations in its recent publications. 10 refs

  13. Radiological effluents released from nuclear rocket and ramjet engine tests at the Nevada Test Site 1959 through 1969: Fact Book

    Energy Technology Data Exchange (ETDEWEB)

    Friesen, H.N.

    1995-06-01

    Nuclear rocket and ramjet engine tests were conducted on the Nevada Test Site (NTS) in Area 25 and Area 26, about 80 miles northwest of Las Vegas, Nevada, from July 1959 through September 1969. This document presents a brief history of the nuclear rocket engine tests, information on the off-site radiological monitoring, and descriptions of the tests.

  14. Analysis of emergency response to fukushima nuclear accident in Japan and suggestions for China's nuclear emergency management

    International Nuclear Information System (INIS)

    Li Wei; Ding Qihua; Wu Haosong

    2014-01-01

    On March 11, 2011, the Fukushima Dai-ichi Nuclear Power Station of the Tokyo Electric Power Company ('TEPCO') was hit and damaged by a magnitude 9 earthquake and accompanying tsunami. The accident is determined to be of the highest rating on the International Nuclear Event Scale. The Government of Japan and TEPCO have taken emergency response actions on-site and off-site at the accident. It became clear through the investigation that the accident had been initiated on the occasion of a natural disaster of an earthquake and tsunami, but there have been various complex problems behind this very serious and large scale accident. For an example, the then-available accident preventive measures and disaster preparedness of TEPCO were insufficient against tsunami and severe accidents; inadequate TEPCO emergency responses to the accident at the site were also identified. The accident rang the alarm for the nuclear safety of nuclear power plants. It also taught us a great of lessons in nuclear emergency management. (authors)

  15. Handling and storage of conditioned high-level wastes

    International Nuclear Information System (INIS)

    1983-01-01

    This report deals with certain aspects of the management of one of the most important wastes, i.e. the handling and storage of conditioned (immobilized and packaged) high-level waste from the reprocessing of spent nuclear fuel and, although much of the material presented here is based on information concerning high-level waste from reprocessing LWR fuel, the principles, as well as many of the details involved, are applicable to all fuel types. The report provides illustrative background material on the arising and characteristics of high-level wastes and, qualitatively, their requirements for conditioning. The report introduces the principles important in conditioned high-level waste storage and describes the types of equipment and facilities, used or studied, for handling and storage of such waste. Finally, it discusses the safety and economic aspects that are considered in the design and operation of handling and storage facilities

  16. The handling of radiation accidents

    International Nuclear Information System (INIS)

    Macdonald, H.F.; Orchard, H.C.; Walker, C.W.

    1977-04-01

    Some of the more interesting and important contributions to a recent International Symposium on the Handling of Radiation Accidents are discussed and personal comments on many of the papers presented are included. The principal conclusion of the Symposium was that although the nuclear industry has an excellent safety record, there is no room for complacency. Continuing attention to emergency planning and exercising are essential in order to maintain this position. A full list of the papers presented at the Symposium is included as an Appendix. (author)

  17. Evolution of the Darlington NGS fuel handling computer systems

    International Nuclear Information System (INIS)

    Leung, V.; Crouse, B.

    1996-01-01

    The ability to improve the capabilities and reliability of digital control systems in nuclear power stations to meet changing plant and personnel requirements is a formidable challenge. Many of these systems have high quality assurance standards that must be met to ensure adequate nuclear safety. Also many of these systems contain obsolete hardware along with software that is not easily transported to newer technology computer equipment. Combining modern technology upgrades into a system of obsolete hardware components is not an easy task. Lastly, as users become more accustomed to using modern technology computer systems in other areas of the station (e.g. information systems), their expectations of the capabilities of the plant systems increase. This paper will present three areas of the Darlington NGS fuel handling computer system that have been or are in the process of being upgraded to current technology components within the framework of an existing fuel handling control system. (author). 3 figs

  18. Evolution of the Darlington NGS fuel handling computer systems

    Energy Technology Data Exchange (ETDEWEB)

    Leung, V; Crouse, B [Ontario Hydro, Bowmanville (Canada). Darlington Nuclear Generating Station

    1997-12-31

    The ability to improve the capabilities and reliability of digital control systems in nuclear power stations to meet changing plant and personnel requirements is a formidable challenge. Many of these systems have high quality assurance standards that must be met to ensure adequate nuclear safety. Also many of these systems contain obsolete hardware along with software that is not easily transported to newer technology computer equipment. Combining modern technology upgrades into a system of obsolete hardware components is not an easy task. Lastly, as users become more accustomed to using modern technology computer systems in other areas of the station (e.g. information systems), their expectations of the capabilities of the plant systems increase. This paper will present three areas of the Darlington NGS fuel handling computer system that have been or are in the process of being upgraded to current technology components within the framework of an existing fuel handling control system. (author). 3 figs.

  19. Importance of Promocat in nuclear emergency management in ANAV; Impacto del Promocat en la gestion de emergencias en ANAV

    Energy Technology Data Exchange (ETDEWEB)

    Gil Cortiella, R.; Torres Gurdel, C.

    2016-08-01

    The nuclear power plant emergency management tool PROMOCAT, has been developed for supporting and facilitating the nuclear emergency management in the Spanish NPPs of Asco and Vandellos II. PROMOCAT is a computerized tool that comprises all the activities made by the Technical Support Centre (TSC) and Offsite facilities. In order to ease the management and decision-making during a nuclear emergency. In addition, the drill mode helps to improve and strengthen emergency personnel training. (Author)

  20. Estimation of initiating event distribution at nuclear power plants by Bayesian procedure

    International Nuclear Information System (INIS)

    Chen Guangming

    1995-01-01

    Initiating events at nuclear power plants such as human errors or components failures may lead to a nuclear accident. The study of the frequency of these events or the distribution of the failure rate is necessary in probabilistic risk assessment for nuclear power plants. This paper presents Bayesian modelling methods for the analysis of the distribution of the failure rate. The method can also be utilized in other related fields especially where the data is sparse. An application of the Bayesian modelling in the analysis of distribution of the time to recover Loss of Off-Site Power ( LOSP) is discussed in the paper

  1. The NUREG-1150 probabilistic risk assessment for the Sequoyah nuclear plant

    International Nuclear Information System (INIS)

    Gregory, J.J.; Breeding, R.J.; Higgins, S.J.; Shiver, A.W.; Helton, J.C.; Murfin, W.B.

    1992-01-01

    This paper summarizes the findings of the probabilistic risk assessment (PRA) for Unit 1 of the Sequoyah Nuclear Plant performed in support of NUREG-1150. The emphasis is on the 'back-end' analyses, the accident progression, source term, and consequence analyses, and the risk results obtained when the results of these analyses are combined with the accident frequency analysis. The results of this PRA indicate that the offsite risk from internal initiating events at Sequoyah are quite low with respect to the safety goals. The containment appears likely to withstand the loads that might be placed upon it if the reactor vessel fails. A good portion of the risk, in this analysis, comes from initiating events which bypass the containment. These events are estimated to have a relatively low frequency of occurrence, but their consequences are quite large. Other events that contribute to offsite risk involve early containment failures that occur during degradation of the core or near the time of vessel breach. Considerable uncertainty is associated with the risk estimates produced in this analysis. Offsite risk from external initiating events was not included in this analysis. (orig.)

  2. Implementations of the hydrological dispersion models of RODOS for assessment of off-site nuclear impacts caused by non energy production sources

    International Nuclear Information System (INIS)

    Zheleznyak, M.; Maderich, V.; Onishi, Y.

    2003-01-01

    Full text: The main aim of the Hydrological Dispersion Module (HDM) of the RODOS - the real-time on-line decision support system for nuclear emergency management in Europe, is the prediction of the dispersion of radionuclides in aquatic systems following atmospheric fallout. HDM provides also the supplementary possibility to simulate the transport of radionuclide in aquatic systems following a direct release of radioactivity into the surface waters. The hydrological model chain of RODOS was developed to simulate processes, such as runoff of radionuclides from watersheds following deposition from the atmosphere (RETRACE model), transport of radionuclides in river systems (1-D model RIVTOX), the radionuclide behaviour in lakes and reservoirs (radioecological box model LAKECO, 2-D lateral-longitudinal model COASTOX), river floodplains (COASTOX), estuaries and seas ( 3-D model THREETOX). The output from the HDM is transferred into the aquatic food chain and dose model FDMA (part of RODOS) to assess the main exposure pathways. The RODOS system was developed to support the offsite nuclear emergency management in a case of the accidents at the acting nuclear power plants. However the models, integrated into the RODOS are more generic and therefore ones can be applied for management of nuclear emergencies caused also by other sources. Three case studies are considered in this presentation for analyses of such possibilities. Case study 1: the simulations of hypothetical nuclear releases from the nuclear submarine 'Kursk' sunken on August the 12 th 2000 in the Barents Sea were carried out to show the ability of cope with RODOS-HDM model THREETOX the emergency situations an the board of ships with nuclear engines. 'Kursk' contains two pressurized-water nuclear reactors OK-650b of 190 MW. The submarine was salvaged in 2002. Two scenarios of 137 Cs releases - the permanent leakage of 1 TBq/year of and an instantaneous leakage of 4500 TBq have been simulated for the period of

  3. Nuclear power. Volume 2: nuclear power project management

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    The following topics are discussed: review of nuclear power plants; licensing procedures; safety analysis; project professional services; quality assurance and project organization; construction, scheduling and operation; construction, scheduling and operation; nuclear fuel handling and fuel management; and plant cost management. 116 references, 115 figures, 33 tables

  4. Examination of some assumed severe reactor accidents at the Olkiluoto nuclear power plant

    International Nuclear Information System (INIS)

    Pekkarinen, E.; Rossi, J.

    1989-02-01

    Knowledge and analysis methods of severe accidents at nuclear power plants and of subsequent response of primary system and containment have been developed in last few years to the extent that realistic source tems of the specified accident sequences can be calculated for the Finnish nuclear power plants. The objective of this investigation was to calculate the source terms of off-site consequences brought about by some selected severe accident sequences initiated by the total loss of on-site and off-site AC power at the Olkiluoto nuclear power plant. The results describing the estimated off-site health risks are expressed as conditional assuming that the accident has taken place, because the probabilities of the occurence of the accident sequences considered have not been analysed in this study. The range and probabilities of occurence of health detriments are considered by calculating consequences in different weeather conditions and taking into account the annual frequency of each weather condition and statistical population distribution. The calculational results indicate that the reactor building provides and additional holdup and deposition of radioactive substance (except coble gases) released from the containment. Furthermore, the release fractions of the core inventory to the environment of volatile fission products such as iodine, cesium and tellurium remain under 0.03. No early health effects are predicted for the surrounding population in case the assumed short-tem countermeasures are performed effectively. Acute health effects are extremely improbable even without any active countermeasure. By reducing the long-term exposure from contaminated agricultural products, the collective dose from natural long-term background radiation, for instance in the sector of 30 degrees towards the southern Finland up to the distance of 300 kilometers, would be expected to increase with 2-20 percent depending on the release considered

  5. How nuclear liability practices have been implemented in US. US nuclear claims experience

    International Nuclear Information System (INIS)

    Bardes, C.R.

    2000-01-01

    Three Mile Island has been only major nuclear incident in US involving a power plant that resulted in payments to public. In addition to Three Mile Island, there have been only 3 lawsuits by members of the public against nuclear power plant operators; these alleged bodily injury and property damage resulting from normal operations. Of 202 claims handled by ANI, 161 involved individual nuclear facilities workers. Costs of the worker claims (through 1998) was US $1.5 million for indemnity (losses) and US$35.9 million for legal defense costs. By far, 1979 TMI accident produced largest number of third-party claims. ANI's emergency claims handling procedure for large nuclear accident tested and proved itself at Three Mile Island

  6. Determination of action zone in the nuclear / radiology handling process

    International Nuclear Information System (INIS)

    Ade Awalludin

    2013-01-01

    Assessment has been conducted on determination of action zone in nuclear or radiological emergency. The assessment is taken into account radiological risk level in nuclear or radiological emergency management process outside nuclear installation. Managing of nuclear emergency is same as that one of other emergency by adding the principles of radiation protection. This study aims to provide guidance in making of safety and security perimeter outside the nuclear installation for first responders during nuclear/radiological emergency based on dose rate, contamination level or distance from the scene. Separation of working zone is important for first responder safety that works in radiological environment in the event of nuclear or radiation emergency without violating their standard operating procedure. Value limit of safety and security perimeter has been made according to the conditions in Indonesia and considering the applicability in practical. (author)

  7. Evolution of a test article handling system for the SP-100 GES test

    International Nuclear Information System (INIS)

    Shen, E.J.; Schweiger, L.J.; Miller, W.C.; Gluck, R.; Davies, S.M.

    1987-01-01

    A simulated space environment test of a flight prototypic SP-100 reactor, control system, and flight shield will be conducted at the Hanford Engineering Development Laboratory (HEDL). The flight prototypic components and the supporting primary heat removal system are collectively known as the nuclear assembly test article (TA). The unique configuration and materials of fabrication for the Test Article require a specialized handling facility to support installation, maintenance, and final disposal operation. The test site operator, working in conjunction with the test article supplier, developed and evaluated several handling concepts resulting in the selection of a reference test article handling system. The development of the reference concept for the handling system is presented

  8. Manpower requirements in the nuclear power industry, 1982-1991

    International Nuclear Information System (INIS)

    Johnson, R.C.

    1982-09-01

    The objective of this study is to project occupational employment needs, created by growth and employee turnover, for the nuclear power industry over the next decade. Employment data for 1981 were collected in a survey conducted by the Institute of Nuclear Power Operations of its 60 member utilities. The data were analyzed statistically to identify factors that account for variations in power plant staffing and the number of off-site nuclear support personnel employed by a utility. Total employment in the nuclear power industry is predicted to increase from 54,400 in 1981 to 73,600 in 1991. Nuclear generating capacity will increase from 58 to 124 gigawatts, based on the midline forecast of the Energy Information Administration. The projections assume that current regulations will remain in effect and no new plans for additional generating facilities will be initiated

  9. Aerial radiological survey of the US Department of Energy's Mound Facility, Miamisburg, Ohio

    International Nuclear Information System (INIS)

    1978-03-01

    An aerial radiological survey to measure terrestrial gamma radiation was carried out by helicopter over an area centered on Mound Facility, a 180 acre area adjacent to the southern edge of the city of Miamisburg, Ohio. This survey was part of an effort to document background radiation levels around nuclear processing and handling facilities owned or contracted by the United States Department of Energy (DOE). Survey activities were conducted and performed by EG and G for the DOE. Wright-Patterson Air Force Base served as the survey base of operations. During the survey, gamma ray data were collected over a 12.3 km 2 area by flying an east-west grid of lines spaced 61 m apart, flying slowly over several selected areas, and hovering over several spots of interest. The processed data indicated that the on-site radioactivity was primarily due to radionuclides currently being handled or processed at the Facility, and that lesser activity could be attributed to previously handled or processed nuclear materials. Off-site data showed the radioactivity to be that only due to naturally occurring radionuclides

  10. Potential applications of advanced remote handling and maintenance technology to future waste handling facilities

    International Nuclear Information System (INIS)

    Kring, C.T.; Herndon, J.N.; Meacham, S.A.

    1987-01-01

    The Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) has been advancing the technology in remote handling and remote maintenance of in-cell systems planned for future US nuclear fuel reprocessing plants. Much of the experience and technology developed over the past decade in this endeavor are directly applicable to the in-cell systems being considered for the facilities of the Federal Waste Management System (FWMS). The ORNL developments are based on the application of teleoperated force-reflecting servomanipulators controlled by an operator completely removed from the hazardous environment. These developments address the nonrepetitive nature of remote maintenance in the unstructured environments encountered in a waste handling facility. Employing technological advancements in dexterous manipulators, as well as basic design guidelines that have been developed for remotely maintained equipment and processes, can increase operation and maintenance system capabilities, thereby allowing the attainment of two Federal Waste Management System major objectives: decreasing plant personnel radiation exposure and increasing plant availability by decreasing the mean-time-to-repair in-cell maintenance and process equipment

  11. Potential applications of advanced remote handling and maintenance technology to future waste handling facilities

    International Nuclear Information System (INIS)

    Kring, C.T.; Herndon, J.N.; Meacham, S.A.

    1987-01-01

    The Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) has been advancing the technology in remote handling and remote maintenance of in-cell systems planned for future U.S. nuclear fuel reprocessing plants. Much of the experience and technology developed over the past decade in this endeavor are directly applicable to the in-cell systems being considered for the facilities of the Federal Waste Management System (FWMS). The ORNL developments are based on the application of teleoperated force-reflecting servomanipulators controlled by an operator completely removed from the hazardous environment. These developments address the nonrepetitive nature of remote maintenance in the unstructured environments encountered in a waste handling facility. Employing technological advancements in dexterous manipulators, as well as basic design guidelines that have been developed for remotely maintained equipment and processes, can increase operation and maintenance system capabilities, thereby allowing the attainment of two Federal Waste Management System major objectives: decreasing plant personnel radiation exposure and increasing plant availability by decreasing the mean-time-to-repair in-cell maintenance and process equipment

  12. Preliminary screening analysis of the off-site environment downstream of the US Department of Energy Oak Ridge Reservation

    International Nuclear Information System (INIS)

    Blaylock, B.G.; Hoffman, F.O.; Frank, M.L.

    1990-01-01

    Operations and waste disposal activities at the Y-12 Plant, the Oak Ridge National Laboratory (ORNL), and the Oak Ridge Gaseous Diffusion Plant (ORGDP), located on the US Department of Energy (DOE) Oak Ridge Reservation (ORR) in eastern Tennessee, have introduced airborne, liquid, and solid wastes into the surrounding environment. Some of these wastes may affect off-site areas by entering local streams that ultimately drain into the Clinch River. Previously reported concentrations of radionuclides, metals, and organic compounds in water, sediment, and biota of the Clinch River and Watts Bar Reservoir suggest the presence of contaminants of possible concern to the protection of human health and the environment. A preliminary screening was conducted of contaminants in the off-site surface water environments downstream of the DOE ORR. This screening analysis represents part of a scoping phase of the Clinch River Resource Conservation and Recovery Facilities Investigation (CRRFI). The purpose of this preliminary screening analysis is to use existing data on off-site contaminant concentrations to identify and prioritize potential contaminants of concern for further evaluation and investigation. The primary objective of this screening analysis is to ensure that CRRFI sampling and analysis efforts focus on those contaminants that may possibly contribute to human health or environmental risk. 8 refs., 3 figs., 6 tabs

  13. Fission reactor recycling pump handling device

    International Nuclear Information System (INIS)

    Togasawa, Hiroshi; Komita, Hideo; Susuki, Shoji; Endo, Takio; Yamamoto, Tetsuzo; Takahashi, Hideaki; Saito, Noboru.

    1991-01-01

    This invention provides a device for handling a recycling pump in a nuclear reactor upon periodical inspections in a BWR type power plant. That is, in a handling device comprising a support for supporting components of a recycling pump, and a lifter for vertically moving the support below a motor case disposed passing through a reactor pressure vessel, a weight is disposed below the support. Then, the center of gravity of the components, the support and the entire weight is substantially aligned with the position for the support. With such a constitution, the components can be moved vertically to the motor case extremely safely, to remarkably suppress vibrations. Further, the operation safety can remarkably be improved by preventing turning down upon occurrence of earthquakes. Further, since vibration-proof jigs as in a prior art can be saved, operation efficiency can be improved. (I.S.)

  14. Fission reactor recycling pump handling device

    Energy Technology Data Exchange (ETDEWEB)

    Togasawa, Hiroshi; Komita, Hideo; Susuki, Shoji; Endo, Takio; Yamamoto, Tetsuzo; Takahashi, Hideaki; Saito, Noboru

    1991-06-24

    This invention provides a device for handling a recycling pump in a nuclear reactor upon periodical inspections in a BWR type power plant. That is, in a handling device comprising a support for supporting components of a recycling pump, and a lifter for vertically moving the support below a motor case disposed passing through a reactor pressure vessel, a weight is disposed below the support. Then, the center of gravity of the components, the support and the entire weight is substantially aligned with the position for the support. With such a constitution, the components can be moved vertically to the motor case extremely safely, to remarkably suppress vibrations. Further, the operation safety can remarkably be improved by preventing turning down upon occurrence of earthquakes. Further, since vibration-proof jigs as in a prior art can be saved, operation efficiency can be improved. (I.S.).

  15. Reevaluation of environmental monitoring program for radiological emergency at Almirante Alvaro Alberto Nuclear Power Plant, Angra dos Reis, Rio de Janeiro - Brazil

    International Nuclear Information System (INIS)

    Ferreira, Lilia M.J. Belem; Ramos Junior, Anthenor C.; Gomes, Carlos A.; Carvalho, Zenildo L.; Gouveia, Vandir; Estrada, Julio; Ney, Cezar

    1996-01-01

    In order to respond to a major radiological emergency at the Almirante Alvaro Alberto Nuclear Power Plant, located in Angra dos Reis, Rio de Janeiro, the Emergency Response Team of the Institute for Radiation Protection and Dosimetry (IRD/CNEN) established a program of environmental monitoring. The purpose of this program is to define a monitoring trend to assess the off-site radiological conditions and give support to decision making for implementing protective measures in case of a radiological accident, taking into account atmospheric diffusion, population conglomerates and their habits, water and land use, contemplating the entire Emergency Planning Zone of 15 km radius. This program has been reevaluated recently, aiming to optimize it and keep it up to date to assure adequacy of environmental surveillance data in support to a prompt response in case of an emergency situation in the nuclear power plant. It has been organized in the form of a handbook to facilitate handling by field teams. Future revisions will be necessary to incorporate additional pertinent information and keep the handbook up to date, since Angra dos Reis is a summer resort region, subject to constant changes. This paper discusses the structure of the environmental monitoring program and describes the content and preparation of this handbook. (author)

  16. Characterisation of focal liver lesions with unenhanced and contrast enhanced low MI real time ultrasound: On-site unblinded versus off-site blinded reading

    International Nuclear Information System (INIS)

    Hohmann, Joachim; Skrok, Jan; Basilico, Raffaella; Jennett, Manfred; Müller, Anja; Wolf, Karl-Jürgen; Albrecht, Thomas

    2012-01-01

    Objective: To compare on-site and blinded off-site reading of baseline ultrasound (US) and contrast enhanced ultrasound (CEUS) for classification and characterisation of focal liver lesions. Materials and methods: 99 patients (57 women and 42 men, age range 18–89 years, mean age: 59 years) with 53 malignant and 46 benign liver lesions were studied with unenhanced US followed by contrast enhanced US after injection of 2.4 ml SonoVue ® (Bracco, Milano, Italy). Image interpretation was performed on-site with clinical information available by consensus of two readers and off-site by two independent blinded readers at two different centers. Comparison of pre and post contrast scans and of the different readers was performed. Reference examinations were histology, intraoperative US, MRI or CT. Results: Sensitivity for malignancy improved from 81/89/66% (on-site/off-site reader 1/2) before to 100/96/96% post contrast administration (p < 0.05, except for reader 1). Specificity improved from 48/48/54% on baseline US to 89/80/76% on CEUS (p < 0.05). Accuracy for specific lesion diagnosis was 62/59/50% pre and 90/77/72% post contrast (p < 0.05). Classification and characterisation post contrast were mildly inferior for off-site reading. Agreement between on-site and off-site readers of unenhanced scans was fair (κ = 0.29–0.39) while it was good for CEUS (κ = 0.63–0.79). Conclusions: CEUS improves classification and characterisation of focal liver lesions and interobserver agreement compared to conventional US. Classification and characterisation post contrast were mildly but statistically significantly better for on-site than for off-site reading.

  17. Nuclear emergency preparedness in Canada

    International Nuclear Information System (INIS)

    1993-03-01

    The preparedness of utilities and government agencies at various levels for dealing with nuclear emergencies occurring at nuclear reactors in Canada is reviewed and assessed. The review is centered on power reactors, but selected research reactors are included also. Emergency planning in the U.S.A., Germany and France, and international recommendations on emergency planning are reviewed to provide background and a basis for comparison. The findings are that Canadians are generally well protected by existing nuclear emergency plans at the electric utility and provincial levels but there are improvements that can be made, mainly at the federal level and in federal-provincial coordination. Ten issues of importance are identified: commitment to nuclear emergency planning by the federal government; division of federal and provincial roles and responsibilities; auditing of nuclear emergency preparedness of all levels of government and of electric utilities; the availability of technical guidance appropriate to Canada; protective action levels for public health and safety; communication with the public; planning and response for the later phases of a nuclear emergency; off-site exercises and training; coordination of international assistance; and emergency planning for research reactors. (L.L.) 79 refs., 2 tabs

  18. Analysis of loss of offsite power transient using RELAP5/MOD1/NSC

    International Nuclear Information System (INIS)

    Kim, Hho Jung; Chung, Bub Dong; Lee, Young Jin; Kim, Jin Soo

    1986-01-01

    System thermal-hydraulic parameters and simulated, using the best-estimate system code(RELAP5/MOD1/NSC), based upon the sequence of events for the KNU1( Korea Nuclear Unit 1) loss of offsite power transient at 77.5% power which occurred on June 9,1981. The results are compared with the actual plant transient data and show good agreements. After the flow coastdown following the trips of both reactor coolant pumps, the establishment of natural circulation by the temperature difference between the hot and the cold legs is confirmed. The calculated reactor coolant flowrate closely approximate the plant data indicating the validity of relevant thermal-hydraulic models in the RELAP5/MOD1/NSC. Results also show that the sufficient heat removal capability is secured by the appropriate supply of the auxiliary feedwater without the operation of S/G PORVs. In addition, a scenario accident at full power, based upon the same sequence of events described above, is also analysed and the results confirmed that the safety of KNU1 is secured by the appropriate operation of the S/G PORVs coupled with the supply of auxiliary feedwater which ensures sufficient heat removal capability. The characteristics of the non-safety related components such as the turbine stop valve closing time, S/G PORV setting etc. are recognized to be important in the transient analyses on a bestestimate basis. (Author)

  19. 10 years and 20,000 sources: the offsite source recovery project

    Energy Technology Data Exchange (ETDEWEB)

    Whitworth, Julia R [Los Alamos National Laboratory; Abeyta, Cristy L [Los Alamos National Laboratory; Pearson, Michael W [Los Alamos National Laboratory

    2009-01-01

    The Global Threat Reduction Initiative's (GTRI) Offsite Source Recovery Project (OSRP) has been recovering excess and unwanted sealed sources for ten years. In January 2009, GTRI announced that the project had recovered 20,000 sealed radioactive sources. This project grew out of early efforts at Los Alamos National Laboratory (LANL) to recover and disposition excess Plutonium-239 (Pu-239) sealed sources that were distributed in the 1960s and 1970s under the Atoms for Peace Program. Sealed source recovery was initially considered a waste management activity, as evidenced by its initial organization under the Department of Energy's (DOE's) Environmental Management (EM) program. After the terrorist attacks of 2001, however, the interagency community began to recognize the threat posed by excess and unwanted radiological material, particularly those that could not be disposed at the end of their useful life. After being transferred to the National Nuclear Security Administration (NNSA) to be part of GTRI, OSRP's mission was expanded to include not only material that would be classified as Greater-than-Class-C (GTCC) when it became waste, but also any other materials that might be a 'national security consideration.' This paper discusses OSRP's history, recovery operations, expansion to accept high-activity beta-gamma-emitting sealed sources and devices and foreign-possessed sources, and more recent efforts such as cooperative projects with the Council on Radiation Control Program Directors (CRCPD) and involvement in GTRI's Search and Secure project. Current challenges and future work will also be discussed.

  20. Minutes of the workshop on off-site release criteria for contaminated materials

    International Nuclear Information System (INIS)

    Singh, S.P.N.

    1989-11-01

    A one and one-half-day workshop was held May 2-3, 1989, at the Pollard Auditorium in Oak Ridge, Tennessee, with the objective of formulating a strategy for developing reasonable and uniform criteria for releasing radioactively contaminated materials from the US Department of Energy (DOE) sites. This report contains the minutes of the workshop. At the conclusion of the workshop, a plan was formulated to facilitate the development of the above-mentioned off-site release criteria

  1. Development of remote handling system based on 3-D shape recognition technique

    International Nuclear Information System (INIS)

    Tomizuka, Chiaki; Takeuchi, Yutaka

    2006-01-01

    In a nuclear facility, the maintenance and repair activities must be done remotely in a radioactive environment. Fuji Electric Systems Co., Ltd. has developed a remote handling system based on 3-D recognition technique. The system recognizes the pose and position of the target to manipulate, and visualizes the scene with the target in 3-D, enabling an operator to handle it easily. This paper introduces the concept and the key features of this system. (author)

  2. Guidance on offsite emergency radiation measurement systems. Phase 2: The milk pathway

    International Nuclear Information System (INIS)

    1984-04-01

    This document provides guidance to State and local governments and to Federal agencies on offsite emergency measurement of radionuclides after an accident involving a light-water nuclear power plant; in particular, this document provides guidance on determining the dose commitment from the milk pathway. Protective action levels proposed by the Food and Drug Administration (FDA) for milk are used as the basis for monitoring requirements. Measurement of radionuclides in milk should be made at the earliest practical point in the production chain: dairy farms, receiving and transfer stations, processing plants or marketing facilities. Early monitoring will provide data to keep significantly contaminated milk out of distribution and will provide the basis for the most timely emergency response action. Radioiodine plus four other radionuclides, cesium-134, cesium-137, strontium-89, and strontium-90, contribute significantly to dose via the milk pathway; of the most severe potential accident, the short-term dose via the milk pathway from radioiodine is significantly greater than that of cesium or strontium. There is no emergency field monitoring instrumentation available for accurately monitoring cesium and strontium, particularly in the presence of radioiodine. Radioiodine can be a potential contamination problem in liquid milk, whereas radiocesium and radiostrontium can be a contamination problem in processed milk products. Monitoring for the long half-life nuclides such as cesium and strontium requires sophisticated equipment or chemistry procedures which are only available in a laboratory. 2 references, 21 figures, 21 tables

  3. 10 years and 20,000 sources: the GTRI offsite source recovery project

    International Nuclear Information System (INIS)

    Whitworth, Julia; Streeper, Charles; Cuthbertson, Abigail

    2009-01-01

    Full text: The Global Threat Reduction Initiative's (GTRI) Offsite Source Recovery Project (OSRP) has been recovering excess and unwanted radioactive sealed sources for ten years. In January 2009, GTRI announced that the project had recovered 20,000 sealed radioactive sources. This project grew out of early efforts at Los Alamos National Laboratory (LANL) to recover and disposition excess Plutonium-239 ( 239 Pu) sealed sources that were distributed in the 1960s and 1970s under the Atoms for Peace Program. Decades later, these sources began to exceed their design life or fall out of regular use. Sealed source recovery was initially considered a waste management activity, but after the terrorist attacks of 2001, the interagency community began to recognize the threat posed by excess and unwanted radiological materials, particularly those that could not be disposed at the end of their useful life. After being transferred to the U.S. National Nuclear Security Administration (NNSA) to be part of GTRI, OSRP's mission was expanded to include not only material that would be classified as Greater-than-Class-C (GTCC) when it became waste, but also any other materials that might constitute a 'national security consideration'. This paper discusses OSRP's history, recovery operations, expansion to accept high-activity beta-gamma-emitting sealed sources and devices and foreign-possessed sources, and more recent efforts such as involvement in GTRI's Search and Secure project. Current challenges and future work will also be discussed

  4. Status of the RODOS system for off-site emergency management after nuclear and radiological accidents and its enhancement under the EURANOS project

    International Nuclear Information System (INIS)

    Raskob, W.

    2007-01-01

    Full text: Under the auspices of its Euratom Research Framework Programmes, the European Commission (EC) has supported the development of the RODOS (Real-time On-line Decision Support) system for off-site emergency management after nuclear accidents for more than a decade. Significant additional funds have been provided by many national RTD programmes, research institutes and industrial collaborators. In particular, the German Ministry of Environment, Nature Conservation and Reactor Safety (BMU) financially contributed to the project with emphasis on early emergency response. As a result of these collaborative actions, a comprehensive decision support system (RODOS) has been developed which can be applied generally within and across Europe. The current version of the system has been, or is being, installed in national emergency centres in several European countries for (pre-operational) use (Germany, Finland, Spain, Portugal, Austria, the Netherlands, Poland, Hungary, Slovakia, Ukraine, Slovenia, and the Czech Republic). Installation is foreseen or under consideration in Switzerland, Greece, Romania, Bulgaria, and Russia within the next few years. RODOS provides coherent support at all stages of an accident (i.e., before, during and after a release), including the long term management and restoration of contaminated areas. The system is able to support decisions about the introduction of a wide range of potentially useful countermeasures (e.g., sheltering and evacuation of people, distribution of iodine tablets, food restrictions, agricultural countermeasures, relocation, decontamination, restoration, etc.) mitigating the consequences of an accident with respect to health, the environment, and the economy. It can be applied to accidental releases into the atmosphere and into various aquatic environments. Appropriate interfaces exist with local and national radiological monitoring data, meteorological measurements and forecasts, and for adaptation to local, regional

  5. Automated nuclear materials accounting

    International Nuclear Information System (INIS)

    Pacak, P.; Moravec, J.

    1982-01-01

    An automated state system of accounting for nuclear materials data was established in Czechoslovakia in 1979. A file was compiled of 12 programs in the PL/1 language. The file is divided into four groups according to logical associations, namely programs for data input and checking, programs for handling the basic data file, programs for report outputs in the form of worksheets and magnetic tape records, and programs for book inventory listing, document inventory handling and materials balance listing. A similar automated system of nuclear fuel inventory for a light water reactor was introduced for internal purposes in the Institute of Nuclear Research (UJV). (H.S.)

  6. Analysis of results from a loss-of-offsite-power-initiated ATWS experiment in the LOFT Facility

    International Nuclear Information System (INIS)

    Varacalle, D.J.; Giri, A.M.; Koizumi, Y.; Koske, J.E.

    1983-01-01

    An anticipated transient without scram (ATWS), initiated by loss-of-offsite power, was experimentally simulated in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). Primary system pressure was controlled using a scaled safety relief valve (SRV) representative of those in a commercial PWR, while reactor power was reduced by moderator reactivity feedback in a natural circulation mode. The experiment showed that reactor power decreases more rapidly when the primary pumps are tripped in a loss-of-offsite-power ATWS than in a loss-of-feedwater induced ATWS when the primary pumps are left on. During the experiment, the SRV had sufficient relief capacity to control primary system pressure. Natural circulation was effective in removing core heat at high temperature, pressure, and core power. The system transient response predicted using the RELAPS/MOD1 computer code showed good agreement with the experimental data

  7. Handling of spent nuclear fuel and final storage of nitrified high level reprocessing waste

    International Nuclear Information System (INIS)

    The following stages of handling and transport of the fuel on its way to final storage are dealt with in the report. 1) The spent nuclear fuel is stored at the power station or in the central fuel storage facility awaiting reprocessing. 2) The fuel is reprocessed, i.e. uranium, plutonium and waste are separated from each other. Reprocessing does not take place in Sweden. The highlevel waste is vitrified and can be sent back to Sweden in the 1990s. 3) Vitrified waste is stored for about 30 years awaiting deposition in the final repository. 4) The waste is encapsulated in highly durable materials to prevent groundwater from coming into contact with the waste glass while the radioactivity of the waste is still high. 5) The canisters are emplaced in a final repository which is built at a depth of 500 m in rock of low permeability. 6) All tunnels and shafts are filled with a mixture of clay and sand of low permeability. A detailed analysis of possible harmful effects resulting from normal acitivties and from conceivable accidents is presented in a special section. (author)

  8. Nuclear criticality safety guide

    International Nuclear Information System (INIS)

    Ro, Seong Ki; Shin, Hee Seong; Park, Seong Won; Shin, Young Joon.

    1997-06-01

    Nuclear criticality safety guide was described for handling, transportation and storage of nuclear fissile materials in this report. The major part of the report was excerpted frp, TID-7016(revision 2) and nuclear criticality safety written by Knief. (author). 16 tabs., 44 figs., 5 refs

  9. Simulation of beta radiator handling procedures in nuclear medicine by means of a movable hand phantom.

    Science.gov (United States)

    Blunck, Ch; Becker, F; Urban, M

    2011-03-01

    In nuclear medicine therapies, people working with beta radiators such as (90)Y may be exposed to non-negligible partial body doses. For radiation protection, it is important to know the characteristics of the radiation field and possible dose exposures at relevant positions in the working area. Besides extensive measurements, simulations can provide these data. For this purpose, a movable hand phantom for Monte Carlo simulations was developed. Specific beta radiator handling scenarios can be modelled interactively with forward kinematics or automatically with an inverse kinematics procedure. As a first investigation, the dose distribution on a medical doctor's hand injecting a (90)Y solution was measured and simulated with the phantom. Modelling was done with the interactive method based on five consecutive frames from a video recorded during the injection. Owing to the use of only one camera, not each detail of the radiation scenario is visible in the video. In spite of systematic uncertainties, the measured and simulated dose values are in good agreement.

  10. Effective nuclear and radiation emergency planning

    International Nuclear Information System (INIS)

    Grlicarev, I.

    2000-01-01

    The paper describes how to develop a balanced emergency plan, which realistically reflect the interfaces with various emergency organizations. The use of resources should be optimized with focusing on the most likely accidents. The pitfalls of writing an emergency plan without ''big picture'' in mind should be avoided. It is absolutely essential to have a clear definition of responsibilities and to have proper understanding of the tasks in between all counterparts in the emergency preparedness. Special attention should be paid to off-site part of the nuclear emergency preparedness, because the people involved in it usually receive less training than the on-site personnel and they are not specialized for nuclear emergencies but deal with all sorts of emergencies. (author)

  11. Role of non-destructive examinations in leak testing of glove boxes for industrial scale plutonium handling at nuclear fuel fabrication facility along with case study

    International Nuclear Information System (INIS)

    Aher, Sachin

    2015-01-01

    Non Destructive Examinations has the prominent role at Nuclear Fuel Fabrication Facilities. Specifically NDE has contributed at utmost stratum in Leak Testing of Glove Boxes and qualifying them as a Class-I confinement for safe Plutonium handling at industrial scale. Advanced Fuel Fabrication Facility, BARC, Tarapur is engaged in fabrication of Plutonium based MOX (PuO 2 , DDUO 2 ) fuel with different enrichments for first core of PFBR reactor. Alpha- Leak Tight Glove Boxes along with HEPA Filters and dynamic ventilation form the promising engineering system for safe and reliable handling of plutonium bearing materials considering the radiotoxicity and risk associated with handling of plutonium. Leak Testing of Glove Boxes which involves the leak detection, leak rectification and leak quantifications is major challenging task. To accomplish this challenge, various Non Destructive Testing methods have assisted in promising way to achieve the stringent leak rate criterion for commissioning of Glove Box facilities for plutonium handling. This paper highlights the Role of various NDE techniques like Soap Solution Test, Argon Sniffer Test, Pressure Drop/Rise Test etc. in Glove Box Leak Testing along with procedure and methodology for effective rectification of leakage points. A Flow Chart consisting of Glove Box leak testing procedure starting from preliminary stage up to qualification stage along with a case study and observations are discussed in this paper. (author)

  12. Potential Indoor Worker Exposure From Handling Area Leakage: Example Event Sequence Frequency Analysis

    International Nuclear Information System (INIS)

    Benke, Roland R.; Adams, George R.

    2008-01-01

    The U.S. Department of Energy (DOE) is currently considering design options for the facilities that will handle spent nuclear fuel and high-level radioactive waste at the potential nuclear waste repository at Yucca Mountain, Nevada. The license application must demonstrate compliance with the performance objectives of 10 CFR Part 63, which include occupational dose limits from 10 CFR Part 20. If DOE submits a license application under 10 CFR Part 63, the U.S. Nuclear Regulatory Commission (NRC) will conduct a risk-informed, performance-based review of the DOE license application and its preclosure safety analysis, in which in-depth technical evaluations are focused on technical areas that are significant to preclosure safety and risk. As part of pre-licensing activities, the Center for Nuclear Waste Regulatory Analyses (CNWRA) developed the Preclosure Safety Analysis Tool software to aid in the regulatory review of a DOE license application and support any independent confirmatory assessments that may be needed. Recent DOE information indicates a primarily canister-based handling approach that includes the wet transfer of individual assemblies where Heating, Ventilation, and Air Conditioning (HVAC) systems may be relied on to provide confinement and limit the spread of any airborne radioactive material from handling operations. Workers may be involved in manual and remote operations in handling transportation casks, canisters, waste packages, or bare spent nuclear fuel assemblies inside facility buildings. As part of routine operations within these facilities, radioactive material may potentially become airborne if canisters are opened or bare fuel assemblies are handled. Leakage of contaminated air from the handling area into adjacent occupied areas, therefore, represents a potential radiological exposure pathway for indoor workers. The objective of this paper is to demonstrate modeling capabilities that can be used by the regulator to estimate frequencies of

  13. Reliability of emergency ac power systems at nuclear power plants

    International Nuclear Information System (INIS)

    Battle, R.E.; Campbell, D.J.

    1983-07-01

    Reliability of emergency onsite ac power systems at nuclear power plants has been questioned within the Nuclear Regulatory Commission (NRC) because of the number of diesel generator failures reported by nuclear plant licensees and the reactor core damage that could result from diesel failure during an emergency. This report contains the results of a reliability analysis of the onsite ac power system, and it uses the results of a separate analysis of offsite power systems to calculate the expected frequency of station blackout. Included is a design and operating experience review. Eighteen plants representative of typical onsite ac power systems and ten generic designs were selected to be modeled by fault trees. Operating experience data were collected from the NRC files and from nuclear plant licensee responses to a questionnaire sent out for this project

  14. EEVADO-ontology based construction of a knowledge based system for handling radiation emergency

    International Nuclear Information System (INIS)

    Datta, D.; Hegde, A.G.

    2001-01-01

    The basic aim of handling a radiation emergency is to protect the site personnel and public in the off-site areas. Hence it is required to assess the accident to develop the protective action recommendations for the site personnel, members of the public and emergency workers. An explanation facility is required to analyze the emergency situation and make needful decision. Algorithmic based software cannot provide this support as it does not have any such facility. A knowledge based system (KBS) obviously would act as a special purpose intelligent agent on behalf or at the behest of the emergency organization. Knowledge of the KBS has been elicited on the basis of ontology, a set of definitions of classes, objects, attributes, relations and constraints. Ontology provides a vocabulary for the expression of domain knowledge. The present paper discusses the development and utility of a knowledge based system - EEVADO (Emergency Evaluation and Dosimetry). The present software is verified and validated by ontology and by conducting Desktop Exercise in training program on Emergency Preparedness. (author)

  15. Remote technologies for handling spent fuel

    International Nuclear Information System (INIS)

    Ramakumar, M.S.

    1999-01-01

    The nuclear programme in India involves building and operating power and research reactors, production and use of isotopes, fabrication of reactor fuel, reprocessing of irradiated fuel, recovery of plutonium and uranium-233, fabrication of fuel containing plutonium-239, uranium-233, post-irradiation examination of fuel and hardware and handling solid and liquid radioactive wastes. Fuel that could be termed 'spent' in thermal reactors is a source for second generation fuel (plutonium and uranium-233). Therefore, it is only logical to extend remote techniques beyond handling fuel from thermal reactors to fuel from fast reactors, post-irradiation examination etc. Fabrication of fuel containing plutonium and uranium-233 poses challenges in view of restriction on human exposure to radiation. Hence, automation will serve as a step towards remotisation. Automated systems, both rigid and flexible (using robots) need to be developed and implemented. Accounting of fissile material handled by robots in local area networks with appropriate access codes will be possible. While dealing with all these activities, it is essential to pay attention to maintenance and repair of the facilities. Remote techniques are essential here. There are a number of commonalities in these requirements and so development of modularized subsystems, and integration of different configurations should receive attention. On a long-term basis, activities like decontamination, decommissioning of facilities and handling of waste generated have to be addressed. While robotized remote systems have to be designed for existing facilities, future designs of facilities should take into account total operation with robotic remote systems. (author)

  16. A unique manual method for emergency offsite dose calculations

    International Nuclear Information System (INIS)

    Wildner, T.E.; Carson, B.H.; Shank, K.E.

    1987-01-01

    This paper describes a manual method developed for performance of emergency offsite dose calculations for PP and L's Susquehanna Steam Electric Station. The method is based on a three-part carbonless form. The front page guides the user through selection of the appropriate accident case and inclusion of meteorological and effluent data data. By circling the applicable accident descriptors, the user circles the dose factors on pages 2 and 3 which are then simply multiplied to yield the whole body and thyroid dose rates at the plant boundary, two, five, and ten miles. The process used to generate the worksheet is discussed, including the method used to incorporate the observed terrain effects on airflow patterns caused by the Susquehanna River Valley topography

  17. Advancement on safety management system of nuclear power for safety and non-anxiety of society

    International Nuclear Information System (INIS)

    Yoshikawa, Hidekazu

    2004-01-01

    Advancement on safety management system is investigated to improve safety and non-anxiety of society for nuclear power, from the standpoint of human machine system research. First, the recent progress of R and D works of human machine interface technologies since 1980 s are reviewed and then the necessity of introducing a new approach to promote technical risk communication activity to foster safety culture in nuclear industries. Finally, a new concept of Offsite Operation and Maintenance Support Center (OMSC) is proposed as the core facility to assemble human resources and their expertise in all organizations of nuclear power, for enhancing safety and non-anxiety of society for nuclear power. (author)

  18. Design basis of off-site emergency response plans for fuel cycle installations

    International Nuclear Information System (INIS)

    Rzepka, J.P.; Dubiau, Ph.; Jouve, A.C.; Charles, T.; Mercier, J.P.

    1995-01-01

    In France, the term 'off-site emergency response plan' refers to all the arrangements which should be made by the government authorities to protect the population in the event of an accident affecting the installations of the site considered. The outline of the method of defining typical accidents, evaluation of 'source-terms' and health consequences is presented. Two applications to installations from the front-end and from the back-end of the fuel cycle are discussed. (K.A.). 1 tab

  19. Developments in emergency planning within Scottish nuclear

    International Nuclear Information System (INIS)

    Simpson, A.

    2000-01-01

    Scottish Nuclear has recently completed a major program of improvements to its nuclear emergency facilities. The improvements include the construction of a purpose built Off-Site Emergency Centre for each of its two power stations and the development of a computer based information management system to facilitate the rapid distribution of information on an emergency to local, regional and national agencies. A computer code has also been developed to allow the rapid assessment of the effects of any accidental release on the local population. The improvements to the emergency facilities have been coupled with changes in local and national arrangements for dealing with a civil nuclear emergency. The use of airborne surveying techniques for rapidly determining levels of deposited activity following an accident is also being examined and preliminary airborne surveys have been carried out. (author)

  20. Analysis on approach of safeguards implementation at research reactor handling item count and bulk material

    International Nuclear Information System (INIS)

    Kim, Hyun Jo; Lee, Sung Ho; Lee, Byung Doo; Jung, Juang

    2016-01-01

    KiJang research reactor (KJRR) will be constructed to produce the radioisotope such as Mo-99 etc., provide the neutron transmutation doping (NTD) service of silicon, and develop the core technologies of research reactor. In this paper, the features of the process and nuclear material flow are reviewed and the material balance area (MBA) and key measurement point (KMP) are established based on the nuclear material flow. Also, this paper reviews the approach on safeguards and nuclear material accountancy at the facility level for Safeguards-by-Design at research reactor handling item count and bulk material. In this paper, MBA and KMPs are established through the analysis on facility features and major process at KJRR handling item count and bulk material. Also, this paper reviews the IAEA safeguards implementation and nuclear material accountancy at KJRR. It is necessary to discuss the safeguards approach on the fresh FM target assemblies and remaining uranium in the intermediate level liquid wastes

  1. Analysis on approach of safeguards implementation at research reactor handling item count and bulk material

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun Jo; Lee, Sung Ho; Lee, Byung Doo; Jung, Juang [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    KiJang research reactor (KJRR) will be constructed to produce the radioisotope such as Mo-99 etc., provide the neutron transmutation doping (NTD) service of silicon, and develop the core technologies of research reactor. In this paper, the features of the process and nuclear material flow are reviewed and the material balance area (MBA) and key measurement point (KMP) are established based on the nuclear material flow. Also, this paper reviews the approach on safeguards and nuclear material accountancy at the facility level for Safeguards-by-Design at research reactor handling item count and bulk material. In this paper, MBA and KMPs are established through the analysis on facility features and major process at KJRR handling item count and bulk material. Also, this paper reviews the IAEA safeguards implementation and nuclear material accountancy at KJRR. It is necessary to discuss the safeguards approach on the fresh FM target assemblies and remaining uranium in the intermediate level liquid wastes.

  2. Socioeconomic consequences of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Tawil, J.J.; Callaway, J.W.; Coles, B.L.; Cronin, F.J.; Currie, J.W.; Imhoff, K.L.; Lewis, P.M.; Nesse, R.J.; Strenge, D.L.

    1984-06-01

    This report identifies and characterizes the off-site socioeconomic consequences that would likely result from a severe radiological accident at a nuclear power plant. The types of impacts that are addressed include economic impacts, health impacts, social/psychological impacts and institutional impacts. These impacts are identified for each of several phases of a reactor accident - from the warning phase through the post-resettlement phase. The relative importance of the impact during each accident phase and the degree to which the impact can be predicted are indicated. The report also examines the methods that are currently used for assessing nuclear reactor accidents, including development of accident scenarios and the estimating of socioeconomic accident consequences with various models. Finally, a critical evaluation is made regarding the use of impact analyses in estimating the contribution of socioeconomic consequences to nuclear accident reactor accident risk. 116 references, 7 figures, 15 tables

  3. Survey on national practices and lessons learnt from off-site nuclear emergency exercises

    International Nuclear Information System (INIS)

    Viktorsson, C.

    1993-01-01

    Nuclear emergency exercises are considered to make an important contribution to the efficiency of emergency preparedness. Generally, the details of the emergency exercises are specified for each country and often for each site, reflecting the particular features that exist in relation to general emergency arrangements. The Chernobyl accident brought a new dimension into the arena of emergency arrangements - the international dimension. New conventions and revised international guidance have been issued and have been or are being included in national emergency plans. The OECD Nuclear Energy Agency decided in 1990 to promote international co-operation in the field of emergency exercises and has adopted a programme of work in this field. One component of this programme, which concerns a survey on national practices and lessons learnt from the planning and conduct of emergency exercises, is dealt with in this paper

  4. Recovery operations in the event of a nuclear accident or radiological emergency

    International Nuclear Information System (INIS)

    1990-01-01

    Much progress has been made over the last decade in the field of emergency planning and preparedness, including the development of guidance, criteria, training programmes, regulations and comprehensive plans in the support of nuclear facilities. To provide a forum for international review and discussion of actual experiences gained and lessons learned from the different aspects of recovery techniques and operations in response to serious accidents at nuclear facilities and accidents associated with radioactive materials, the IAEA organized the International Symposium on Recovery Operations in the Event of a Nuclear Accident or Radiological Emergency. The symposium was held from 6 to 10 November 1989 in Vienna, Austria, and was attended by over 250 experts from 35 Member State and 7 international organizations. Although the prime focus was on on-site and off-site recovery from nuclear reactor accidents and on recovery from radiological accidents unrelated to nuclear power plants, development of emergency planning and preparedness resources was covered as well. From the experiences reported, lessons learned were identified. While further work remains to be done to improve concepts, plans, materials, communications and mechanisms to assemble quickly all the special resources needed in the event of an accident, there was general agreement that worldwide preparations to handle any possible future radiological emergencies had vastly improved. A special feature of the symposium programme was the inclusion of a full session on an accident involving a chemical explosion in a high level waste tank a a plutonium extraction plant in the Southern Urals in the USSR in 1957. Information was presented on the radioactive release, its dissemination and deposition, the resultant radiation situation, dose estimates, health effects follow-up, and the rehabilitation of contaminated land. This volume contains the full text of the 49 papers presented at the symposium together with a

  5. Handling, treatment, conditioning and storage of biological radioactive wastes. Technical manual for the management of low and intermediate level wastes generated at small nuclear research centres and by radioisotope users in medicine, research and industry

    International Nuclear Information System (INIS)

    1994-12-01

    Biological materials that contain radioactive isotopes have many important applications. During the production and use of these materials, waste will inevitably arise which must be managed with particular care due to their potential biological as well as radiological hazards. This report deals with wastes that arise outside the nuclear fuel cycle and is directed primarily to countries without nuclear power programmes. It is intended to provide guidance to Member States in the handling, treatment and conditioning of biological radioactive materials. The objective of radioactive waste management is to handle, pretreat, treat, condition, store, transport and dispose of radioactive waste in a manner that protects human health and the environment without imposing undue burdens on future generations. 31 refs, 15 figs, 3 tabs

  6. Handling, treatment, conditioning and storage of biological radioactive wastes. Technical manual for the management of low and intermediate level wastes generated at small nuclear research centres and by radioisotope users in medicine, research and industry

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-01

    Biological materials that contain radioactive isotopes have many important applications. During the production and use of these materials, waste will inevitably arise which must be managed with particular care due to their potential biological as well as radiological hazards. This report deals with wastes that arise outside the nuclear fuel cycle and is directed primarily to countries without nuclear power programmes. It is intended to provide guidance to Member States in the handling, treatment and conditioning of biological radioactive materials. The objective of radioactive waste management is to handle, pretreat, treat, condition, store, transport and dispose of radioactive waste in a manner that protects human health and the environment without imposing undue burdens on future generations. 31 refs, 15 figs, 3 tabs.

  7. Fuel handling machine and auxiliary systems for a fuel handling cell

    International Nuclear Information System (INIS)

    Suikki, M.

    2013-10-01

    This working report is an update for as well as a supplement to an earlier fuel handling machine design (Kukkola and Roennqvist 2006). A focus in the earlier design proposal was primarily on the selection of a mechanical structure and operating principle for the fuel handling machine. This report introduces not only a fuel handling machine design but also auxiliary fuel handling cell equipment and its operation. An objective of the design work was to verify the operating principles of and space allocations for fuel handling cell equipment. The fuel handling machine is a remote controlled apparatus capable of handling intensely radiating fuel assemblies in the fuel handling cell of an encapsulation plant. The fuel handling cell is air tight space radiation-shielded with massive concrete walls. The fuel handling machine is based on a bridge crane capable of traveling in the handling cell along wall tracks. The bridge crane has its carriage provided with a carousel type turntable having mounted thereon both fixed and telescopic masts. The fixed mast has a gripper movable on linear guides for the transfer of fuel assemblies. The telescopic mast has a manipulator arm capable of maneuvering equipment present in the fuel handling cell, as well as conducting necessary maintenance and cleaning operations or rectifying possible fault conditions. The auxiliary fuel handling cell systems consist of several subsystems. The subsystems include a service manipulator, a tool carrier for manipulators, a material hatch, assisting winches, a vacuum cleaner, as well as a hose reel. With the exception of the vacuum cleaner, the devices included in the fuel handling cell's auxiliary system are only used when the actual encapsulation process is not ongoing. The malfunctions of mechanisms or actuators responsible for the motion actions of a fuel handling machine preclude in a worst case scenario the bringing of the fuel handling cell and related systems to a condition appropriate for

  8. Fuel handling machine and auxiliary systems for a fuel handling cell

    Energy Technology Data Exchange (ETDEWEB)

    Suikki, M. [Optimik Oy, Turku (Finland)

    2013-10-15

    This working report is an update for as well as a supplement to an earlier fuel handling machine design (Kukkola and Roennqvist 2006). A focus in the earlier design proposal was primarily on the selection of a mechanical structure and operating principle for the fuel handling machine. This report introduces not only a fuel handling machine design but also auxiliary fuel handling cell equipment and its operation. An objective of the design work was to verify the operating principles of and space allocations for fuel handling cell equipment. The fuel handling machine is a remote controlled apparatus capable of handling intensely radiating fuel assemblies in the fuel handling cell of an encapsulation plant. The fuel handling cell is air tight space radiation-shielded with massive concrete walls. The fuel handling machine is based on a bridge crane capable of traveling in the handling cell along wall tracks. The bridge crane has its carriage provided with a carousel type turntable having mounted thereon both fixed and telescopic masts. The fixed mast has a gripper movable on linear guides for the transfer of fuel assemblies. The telescopic mast has a manipulator arm capable of maneuvering equipment present in the fuel handling cell, as well as conducting necessary maintenance and cleaning operations or rectifying possible fault conditions. The auxiliary fuel handling cell systems consist of several subsystems. The subsystems include a service manipulator, a tool carrier for manipulators, a material hatch, assisting winches, a vacuum cleaner, as well as a hose reel. With the exception of the vacuum cleaner, the devices included in the fuel handling cell's auxiliary system are only used when the actual encapsulation process is not ongoing. The malfunctions of mechanisms or actuators responsible for the motion actions of a fuel handling machine preclude in a worst case scenario the bringing of the fuel handling cell and related systems to a condition appropriate for

  9. Current US strategy and technologies for spent fuel handling

    International Nuclear Information System (INIS)

    Bennett, P.C.; Stringer, J.B.

    1999-01-01

    The United States Department of Energy has recently completed a topical safety analysis report outlining the design and operation of a Centralized Interim Storage Facility for spent commercial nuclear fuel. During the course of the design, dose assessments indicated the need for remote operation of many of the cask handling operations. Use of robotic equipment was identified as a desirable handling solution that is capable of automating many of the operations to maintain throughput, and sufficiently flexible to handle five or more different storage cask designs in varying numbers on a given day. This paper discusses the facility and the dose assessment leading to this choice, and reviews factors to be considered when choosing robotics or automation. Further, a new computer simulation tool to quantify dose to humans working in radiological environments, the Radiological Environment Modeling System (REMS), is introduced. REMS has been developed to produce a more accurate estimate of dose to radiation workers in new activities with radiological hazards. (author)

  10. The Y-12 Plant No Rad-Added Program for off-site shipment of nonradioactive hazardous waste

    International Nuclear Information System (INIS)

    Cooper, K.H.; Mattie, B.K.; Williams, J.L.; Jacobs, D.G.; Roberts, K.A.

    1994-01-01

    On May 17, 1991, the US Department of Energy (DOE) issued a directive for DOE operations to cease off-site shipments of non-radioactive hazardous waste pending further clarification and approvals. A DOE Performance Objective for Certification of Non-Radioactive Hazardous Waste was issued in November 1991. In response to these directives, the Waste Management Division of Oak Ridge Y-12 Plant, with assistance from Roy F. Weston, Inc., has developed a No Rad-Added Program to provide small programmatic guidance and a set of procedures, approved by DOE, which will permit hazardous waste to be shipped from the Y-12 Plant to commercial treatment, storage, or disposal facilities after ensuring and certifying that hazardous waste has no radioactivity added as a result of DOE operations. There are serious legal and financial consequences of shipping waste containing radioactivity to an off-site facility not licensed to receive radioactive materials. Therefore, this program is designed with well-defined responsibilities and stringent documentation requirements

  11. Radiation-tolerant cable management systems for remote handling applications in the nuclear industry

    International Nuclear Information System (INIS)

    Cullen, S.; Thom, M.

    1993-01-01

    Experience has shown that one of the most vulnerable areas within remote handling equipment is the umbilical cable and termination system. Repairs of a damaged system can be very long due to poorly designed termination techniques. Over the past five years W.L. Gore has gained considerable experience in the design and manufacture of cable systems, utilising unique radiation tolerant materials and manufacturing processes. The cable systems manufactured at the W.L. Gore, Dunfermline, Scotland facility have proven to give excellent performance in the most demanding of remote handling applications. (author)

  12. Handling device for nuclear fuel assemblies and assembly appropriate for such a device

    International Nuclear Information System (INIS)

    Cransac, J.P.; Jaquelin, R.; Renaux, C.

    1985-01-01

    The handling device comprises a guide tube of which axis is vertical, in which a grab moves, hanging from a chain, under the action of a back-geared motor. The grab being stopped in its rotation in the guide tube, an assembly can be gripped with a bayonet system while controlling the rotation of the grab - guide tube system a back-geared motor. The device can be hanged from the small or large rotating plug of a fast neutron reactor. It can be used in a handling flask [fr

  13. Design of simulation-based medical education and advantages and disadvantages of in situ simulation versus off-site simulation

    DEFF Research Database (Denmark)

    Sørensen, Jette Led; Østergaard, Doris; LeBlanc, Vicki

    2017-01-01

    that choice of setting for simulations does not seem to influence individual and team learning. Department-based local simulation, such as simulation in-house and especially in situ simulation, leads to gains in organisational learning. The overall objectives of simulation-based education and factors......BACKGROUND: Simulation-based medical education (SBME) has traditionally been conducted as off-site simulation in simulation centres. Some hospital departments also provide off-site simulation using in-house training room(s) set up for simulation away from the clinical setting, and these activities...... simulations. DISCUSSION: Non-randomised studies argue that in situ simulation is more effective for educational purposes than other types of simulation settings. Conversely, the few comparison studies that exist, either randomised or retrospective, show that choice of setting does not seem to influence...

  14. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

    Yoon, Ji Sup; Park, B. S.; Park, Y. S.; Oh, S. C.; Kim, S. H.; Cho, M. W.; Hong, D. H.

    1997-12-01

    Since the nation's policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs

  15. Disposal of slightly contaminated radioactive wastes from nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Minns, J.L. [Nuclear Regulatory Commission, Washington, DC (United States)

    1995-02-01

    With regard to the disposal of solid wastes, nuclear power plants basically have two options, disposal in a Part 61 licensed low-level waste site, or receive approval pursuant to 20.2002 for disposal in a manner not otherwise authorized by the NRC. Since 1981, the staff has reviewed and approved 30 requests for disposal of slightly contaminated radioactive materials pursuant to Section 20.2002 (formerly 20.302) for nuclear power plants located in non-Agreement States. NRC Agreement States have been delegated the authority for reviewing and approving such disposals (whether onsite or offsite) for nuclear power plants within their borders. This paper describes the characteristics of the waste disposed of, the review process, and the staff`s guidelines.

  16. Nuclear energy in question

    International Nuclear Information System (INIS)

    Simon, D.N.; Carvalho, J.F. de; Goldemberg, J.; Menezes, L.C.; Rosa, L.P.; Oliveira, R.G. de.

    1981-01-01

    The basic requirements demanded for the physical protection of nuclear operational units, is established. These units can be, production, utilization, processing, reprocessing, handling, transport or storage of materials of interesting to Brazilian Nuclear Program. (E.G.) [pt

  17. Meteorological considerations in emergency response capability at nuclear power plant

    International Nuclear Information System (INIS)

    Fairobent, J.E.

    1985-01-01

    Meteorological considerations in emergency response at nuclear power plants are discussed through examination of current regulations and guidance documents, including discussion of the rationale for current regulatory requirements related to meteorological information for emergency response. Areas discussed include: major meteorological features important to emergency response; onsite meteorological measurements programs, including redundant and backup measurements; access to offsite sources of meteorological information; consideration of real-time and forecast conditions and atmospheric dispersion modeling

  18. Fuel handling at Cernavoda 1 N.P.S. - commissioning and training philosophy

    Energy Technology Data Exchange (ETDEWEB)

    Standen, G W [AECL-Ansaldo Consortium, Cernavoda (Romania); Tiron, C; Marinescu, S [Regia Nationala de Electricitate (RENEL), Cernavoda (Romania); [Filiala Centrala Nuclearo Electrica (FCNE), Cernavoda (Romania)

    1997-12-31

    Efficient operation of a Candu nuclear power plant depends greatly on the reliable and safe operation of the fuel handling system. Successful commissioning of the system is obviously a key aspect of the reliability of the system and this coupled with a rigorous training programme for the fuel handling staff will ensure the system`s safe operation. This paper describes the philosophy used at Cernavoda 1 N.P.S. for the commissioning of the fuel handling systems and for the training of staff for operation and maintenance of these systems. The paper also reviews the commissioning programme, describing the milestones achieved and discussing some of the more interesting technical aspects which includes some unique Romanian input. In conclusion the paper looks at the organization of the mature fuel handling department from the operations, maintenance and technical support points of view and the long term plans for the future. (author). 1 fig.

  19. Fuel handling at Cernavoda 1 N.P.S. - commissioning and training philosophy

    International Nuclear Information System (INIS)

    Standen, G.W.; Tiron, C.; Marinescu, S.

    1996-01-01

    Efficient operation of a Candu nuclear power plant depends greatly on the reliable and safe operation of the fuel handling system. Successful commissioning of the system is obviously a key aspect of the reliability of the system and this coupled with a rigorous training programme for the fuel handling staff will ensure the system's safe operation. This paper describes the philosophy used at Cernavoda 1 N.P.S. for the commissioning of the fuel handling systems and for the training of staff for operation and maintenance of these systems. The paper also reviews the commissioning programme, describing the milestones achieved and discussing some of the more interesting technical aspects which includes some unique Romanian input. In conclusion the paper looks at the organization of the mature fuel handling department from the operations, maintenance and technical support points of view and the long term plans for the future. (author). 1 fig

  20. Constitutional problems in the handling of plutonium

    International Nuclear Information System (INIS)

    Witt, S. de.

    1989-01-01

    Reprocessing and final storage involve two different systems of nuclear energy utilization: with or without the use of plutonium. There is a choice available between these two systems. The paper discusss the constitutional implications of this choice. The permission of the use of plutonium as nuclear fuel by the Atomic Energy Law is irreconcilable with the Basic Law, i.e. the Constitution. If the corresponding provisions of the Atomic Energy Law are repealed, then only the plutonium-related branch will be revoked and not the legal permission of nuclear energy as a whole. The fact is not ignored that the Atomic Energy law does not permit the construction and operation of a plant or the handling of plutonium if this were to violate a basic right. However, the plutonium-related branch of nculear energy utilization inevitably results in such basic right violations; hence the Atomic Energy law is unconstitutional in this respect. (orig./HSCH) [de