Structural evaluation of the Shippingport Reactor Pressure Vessel and Neutron Shield Tank package for impact and puncture loads

International Nuclear Information System (INIS)

Fischer, L.E.; Chou, C.K.; Lo, T.; Schwartz, M.W.

1988-06-01

A structural evaluation of Shippingport Reactor Pressure Vessel and Neutron Shield Tank package for impact and puncture loads under the normal and hypothetical accident conditions of 10 CFR 71 was performed. Component performance criteria for the Shippingport package and the corresponding structural acceptance criteria for these components were developed based on a review of the package geometry, the planned transport environment, and the external radiation standards and dispersal limits of 10 CFR 71. The evaluation was performed using structural analysis methods. A demonstration combining simplified model tests and nonlinear finite element analyses was made to substantiate the structural analysis methods used to evaluate the Shippingport package. The package was analyzed and the results indicate that the package meets external radiation standards and release limits of 10 CFR 71. 13 refs., 50 figs., 19 tabs

Development of a new control software package for Pakistan Research Reactor-2

International Nuclear Information System (INIS)

Qazi, M.K.

1993-05-01

The development of a new control software package for Pakistan Research Reactor-2 is presented. The software operates in different modes which comprises of surveillance, pre-operational self tests, operator, supervisor and robotic control. The control logic critically damp the system minimizing power overshoots. The software, handles multiple abnormal conditions, provides an elaborate access control and maintains startup/shutdown record. The report describes the functional details and covers the operational aspects of the new control software. (author)

User's guide to input for WRAP: a water reactor analysis package

International Nuclear Information System (INIS)

Gregory, M.V.

1977-06-01

The document describes the input records required to execute the Water Reactor Analysis Package (WRAP) for the analysis of thermal-hydraulic transients in primarily light water reactors. The card input required by RELAP4 has been significantly modified to broaden the code's input processing capabilities: (1) All input is in the form of templated, named records. (2) All components (volumes, junctions, etc.) are named rather than numbered, and system relationships are formed by defining associations between the names. (3) A hierarchical part structure is used which allows collections of components to be described as discrete parts (these parts may then be catalogued for use in a wide range of cases). A sample problem, the small break analysis of the Westinghouse Trojan Plant, is discussed and detailed, step-by-step instructions in setting up an input data base are presented. A master list of all input templates for WRAP is compiled

IN-PACKAGE CHEMISTRY ABSTRACTION

Energy Technology Data Exchange (ETDEWEB)

E. Thomas

2005-07-14

This report was developed in accordance with the requirements in ''Technical Work Plan for Postclosure Waste Form Modeling'' (BSC 2005 [DIRS 173246]). The purpose of the in-package chemistry model is to predict the bulk chemistry inside of a breached waste package and to provide simplified expressions of that chemistry as a function of time after breach to Total Systems Performance Assessment for the License Application (TSPA-LA). The scope of this report is to describe the development and validation of the in-package chemistry model. The in-package model is a combination of two models, a batch reactor model, which uses the EQ3/6 geochemistry-modeling tool, and a surface complexation model, which is applied to the results of the batch reactor model. The batch reactor model considers chemical interactions of water with the waste package materials, and the waste form for commercial spent nuclear fuel (CSNF) waste packages and codisposed (CDSP) waste packages containing high-level waste glass (HLWG) and DOE spent fuel. The surface complexation model includes the impact of fluid-surface interactions (i.e., surface complexation) on the resulting fluid composition. The model examines two types of water influx: (1) the condensation of water vapor diffusing into the waste package, and (2) seepage water entering the waste package as a liquid from the drift. (1) Vapor-Influx Case: The condensation of vapor onto the waste package internals is simulated as pure H{sub 2}O and enters at a rate determined by the water vapor pressure for representative temperature and relative humidity conditions. (2) Liquid-Influx Case: The water entering a waste package from the drift is simulated as typical groundwater and enters at a rate determined by the amount of seepage available to flow through openings in a breached waste package.

IN-PACKAGE CHEMISTRY ABSTRACTION

International Nuclear Information System (INIS)

E. Thomas

2005-01-01

This report was developed in accordance with the requirements in ''Technical Work Plan for Postclosure Waste Form Modeling'' (BSC 2005 [DIRS 173246]). The purpose of the in-package chemistry model is to predict the bulk chemistry inside of a breached waste package and to provide simplified expressions of that chemistry as a function of time after breach to Total Systems Performance Assessment for the License Application (TSPA-LA). The scope of this report is to describe the development and validation of the in-package chemistry model. The in-package model is a combination of two models, a batch reactor model, which uses the EQ3/6 geochemistry-modeling tool, and a surface complexation model, which is applied to the results of the batch reactor model. The batch reactor model considers chemical interactions of water with the waste package materials, and the waste form for commercial spent nuclear fuel (CSNF) waste packages and codisposed (CDSP) waste packages containing high-level waste glass (HLWG) and DOE spent fuel. The surface complexation model includes the impact of fluid-surface interactions (i.e., surface complexation) on the resulting fluid composition. The model examines two types of water influx: (1) the condensation of water vapor diffusing into the waste package, and (2) seepage water entering the waste package as a liquid from the drift. (1) Vapor-Influx Case: The condensation of vapor onto the waste package internals is simulated as pure H 2 O and enters at a rate determined by the water vapor pressure for representative temperature and relative humidity conditions. (2) Liquid-Influx Case: The water entering a waste package from the drift is simulated as typical groundwater and enters at a rate determined by the amount of seepage available to flow through openings in a breached waste package

The Innovations, Technology and Waste Management Approaches to Safely Package and Transport the World's First Radioactive Fusion Research Reactor for Burial

International Nuclear Information System (INIS)

Keith Rule; Erik Perry; Jim Chrzanowski; Mike Viola; Ron Strykowsky

2003-01-01

Original estimates stated that the amount of radioactive waste that will be generated during the dismantling of the Tokamak Fusion Test Reactor will approach two million kilograms with an associated volume of 2,500 cubic meters. The materials were activated by 14 MeV neutrons and were highly contaminated with tritium, which present unique challenges to maintain integrity during packaging and transportation. In addition, the majority of this material is stainless steel and copper structural metal that were specifically designed and manufactured for this one-of-a-kind fusion research reactor. This provided further complexity in planning and managing the waste. We will discuss the engineering concepts, innovative practices, and technologies that were utilized to size reduce, stabilize, and package the many unique and complex components of this reactor. This waste was packaged and shipped in many different configurations and methods according to the transportation regulations and disposal facility requirements. For this particular project, we were able to utilize two separate disposal facilities for burial. This paper will conclude with a complete summary of the actual results of the waste management costs, volumes, and best practices that were developed from this groundbreaking and successful project

CONSUL code package application for LMFR core calculations

Energy Technology Data Exchange (ETDEWEB)

Chibinyaev, A.V.; Teplov, P.S.; Frolova, M.V. [RNC ' Kurchatovskiy institute' , Kurchatov sq.1, Moscow (Russian Federation)

2008-07-01

CONSUL code package designed for the calculation of reactor core characteristics has been developed at the beginning of 90's. The calculation of nuclear reactor core characteristics is carried out on the basis of correlated neutron, isotope and temperature distributions. The code package has been generally used for LWR core characteristics calculations. At present CONSUL code package was adapted to calculate liquid metal fast reactors (LMFR). The comparisons with IAEA computational test 'Evaluation of benchmark calculations on a fast power reactor core with near zero sodium void effect' and BN-1800 testing calculations are presented in the paper. The IAEA benchmark core is based on the innovative core concept with sodium plenum above the core BN-800. BN-1800 core is the next development step which is foreseen for the Russian fast reactor concept. The comparison of the operational parameters has shown good agreement and confirms the possibility of CONSUL code package application for LMFR core calculation. (authors)

Naval Waste Package Design Report

International Nuclear Information System (INIS)

M.M. Lewis

2004-01-01

A design methodology for the waste packages and ancillary components, viz., the emplacement pallets and drip shields, has been developed to provide designs that satisfy the safety and operational requirements of the Yucca Mountain Project. This methodology is described in the ''Waste Package Design Methodology Report'' Mecham 2004 [DIRS 166168]. To demonstrate the practicability of this design methodology, four waste package design configurations have been selected to illustrate the application of the methodology. These four design configurations are the 21-pressurized water reactor (PWR) Absorber Plate waste package, the 44-boiling water reactor (BWR) waste package, the 5-defense high-level waste (DHLW)/United States (U.S.) Department of Energy (DOE) spent nuclear fuel (SNF) Co-disposal Short waste package, and the Naval Canistered SNF Long waste package. Also included in this demonstration is the emplacement pallet and continuous drip shield. The purpose of this report is to document how that design methodology has been applied to the waste package design configurations intended to accommodate naval canistered SNF. This demonstrates that the design methodology can be applied successfully to this waste package design configuration and support the License Application for construction of the repository

Design/licensing of on-site package for core component

International Nuclear Information System (INIS)

Ogasawara, K.; Chohzuka, T.; Shimura, T.; Kikuchi, T.; Fujiwara, R.; Karigome, S.; Takani, M.

1993-01-01

For storage of used core components which are produced from reactors, Tohoku EPCO decided to construct a site bunker at Onagawa site. It was also decided to develop and fabricate one packaging to transport core components from the reactor buildings to the site bunker. The packaging will be used within the power station; therefore, it shall comply with 'The Law for the Business of Electric Power' and relevant Notification. The main requirements of the packaging are as follows: 1) The number of contents, such as channel boxes and control rods, shall be as large as possible. 2) The weight and the outer dimensions of the packaging shall be within the limitation of the reactor building and the site bunker. 3) Materials shall be selected from those which have been already applied for existing packagings and utilized without any problems. 4) It shall be considered during design of trunnions that handling equipment, such as lifting beam, can be used for not only this packaging but also for existing spent fuel packagings. The design of the packaging is completed and has been licensed. The packaging is scheduled to be utilized from November, 1993. (J.P.N.)

Package Flow Model and its fuzzy implementation for simulating nuclear reactor system dynamics

International Nuclear Information System (INIS)

Matsuoka, Hiroshi; Ishiguro, Misako.

1996-01-01

A simple intuitive simulation model, which we call 'Package Flow Model', has been developed to evaluate physical processes in nuclear reactor system from a macroscopic point of view. In the previous paper, we showed the physical process of each energy generation and transfer stage in a PWR could be modeled by PFM, and its dynamics could be approximately simulated by fuzzy implementation. In this paper, a PFMs network approach for a total PWR system simulation is proposed and some transients of nuclear ship 'MUTSU' reactor system are evaluated. The simulated results are consistent with those from Nuclear Ship Engineering Simulation System developed by JAERI. Furthermore, a visual representation method is proposed to intuitively capture the profile of fuel safety transient. Using the PFMs network, we can handily calculate the transient phenomena of the system even by a notebook-type personal computer. In addition, we can easily interpret the results of calculation surveying a small number of parameters. (author)

Waste Package Component Design Methodology Report

International Nuclear Information System (INIS)

D.C. Mecham

2004-01-01

requirements of the YMP. Four waste package configurations have been selected to illustrate the application of the methodology during the licensing process. These four configurations are the 21-pressurized water reactor absorber plate waste package (21-PWRAP), the 44-boiling water reactor waste package (44-BWR), the 5 defense high-level radioactive waste (HLW) DOE spent nuclear fuel (SNF) codisposal short waste package (5-DHLWDOE SNF Short), and the naval canistered SNF long waste package (Naval SNF Long). Design work for the other six waste packages will be completed at a later date using the same design methodology. These include the 24-boiling water reactor waste package (24-BWR), the 21-pressurized water reactor control rod waste package (21-PWRCR), the 12-pressurized water reactor waste package (12-PWR), the 5 defense HLW DOE SNF codisposal long waste package (5-DHLWDOE SNF Long), the 2 defense HLW DOE SNF codisposal waste package (2-MC012-DHLW), and the naval canistered SNF short waste package (Naval SNF Short). This report is only part of the complete design description. Other reports related to the design include the design reports, the waste package system description documents, manufacturing specifications, and numerous documents for the many detailed calculations. The relationships between this report and other design documents are shown in Figure 1

Waste Package Component Design Methodology Report

Energy Technology Data Exchange (ETDEWEB)

D.C. Mecham

2004-07-12

and operational requirements of the YMP. Four waste package configurations have been selected to illustrate the application of the methodology during the licensing process. These four configurations are the 21-pressurized water reactor absorber plate waste package (21-PWRAP), the 44-boiling water reactor waste package (44-BWR), the 5 defense high-level radioactive waste (HLW) DOE spent nuclear fuel (SNF) codisposal short waste package (5-DHLWDOE SNF Short), and the naval canistered SNF long waste package (Naval SNF Long). Design work for the other six waste packages will be completed at a later date using the same design methodology. These include the 24-boiling water reactor waste package (24-BWR), the 21-pressurized water reactor control rod waste package (21-PWRCR), the 12-pressurized water reactor waste package (12-PWR), the 5 defense HLW DOE SNF codisposal long waste package (5-DHLWDOE SNF Long), the 2 defense HLW DOE SNF codisposal waste package (2-MC012-DHLW), and the naval canistered SNF short waste package (Naval SNF Short). This report is only part of the complete design description. Other reports related to the design include the design reports, the waste package system description documents, manufacturing specifications, and numerous documents for the many detailed calculations. The relationships between this report and other design documents are shown in Figure 1.

Packaging and transport case of test fuel assembly irradiated in the Creys-Malville reactor

International Nuclear Information System (INIS)

Geffroy, J.; Vivien, J.; Pouard, M.; Dujardin, G.N.; Veron, B.; Michoux, H.

1986-06-01

Some irradiated fuel assemblies from the fast neutron Creys Malville reactor will be sent to hot laboratories to follow fuel behavior. These test assemblies will be examined after a limited cooling time and transport is realized at high residual power (about 10kW) and cladding temperature should not rise over 500deg C. The fuel assemblies are not dismantled and transported into sodium. The assembly is placed into a case containing sodium plugged and put into a packaging. Dimensioning, thermal behavior, radiation protection and containment are examined [fr

RELAP/SCDAPSIM Reactor System Simulator Development and Training for University and Reactor Applications

International Nuclear Information System (INIS)

Hohorst, J.K.; Allison, C.M.

2010-01-01

The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology development program called SDTP (SCDAP Development and Training Program). SDTP involves more than 60 organizations in 28 countries. One of the important applications of the code is for simulator training of university faculty and students, reactor analysts, and reactor operations and technical support staff. Examples of RELAP/SCDAPSIM-based system thermal hydraulic and severe accident simulator packages include the SAFSIM simulator developed by NECSA for the SAFARI research reactor in South Africa, university-developed simulators at the University of Mexico and Shanghai Jiao Tong University in China, and commercial VISA and RELSIM packages used for analyst and reactor operations staff training. This paper will briefly describe the different packages/facilities. (authors)

RELAP/SCDAPSIM Reactor System Simulator Development and Training for University and Reactor Applications

Energy Technology Data Exchange (ETDEWEB)

Hohorst, J.K.; Allison, C.M. [Innovative Systems Software, 1242 South Woodruff Avenue, Idaho Falls, Idaho 83404 (United States)

2010-07-01

The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology development program called SDTP (SCDAP Development and Training Program). SDTP involves more than 60 organizations in 28 countries. One of the important applications of the code is for simulator training of university faculty and students, reactor analysts, and reactor operations and technical support staff. Examples of RELAP/SCDAPSIM-based system thermal hydraulic and severe accident simulator packages include the SAFSIM simulator developed by NECSA for the SAFARI research reactor in South Africa, university-developed simulators at the University of Mexico and Shanghai Jiao Tong University in China, and commercial VISA and RELSIM packages used for analyst and reactor operations staff training. This paper will briefly describe the different packages/facilities. (authors)

The UK core performance code package

International Nuclear Information System (INIS)

Hutt, P.K.; Gaines, N.; McEllin, M.; White, R.J.; Halsall, M.J.

1991-01-01

Over the last few years work has been co-ordinated by Nuclear Electric, originally part of the Central Electricity Generating Board, with contributions from the United Kingdom Atomic Energy Authority and British Nuclear Fuels Limited, to produce a generic, easy-to-use and integrated package of core performance codes able to perform a comprehensive range of calculations for fuel cycle design, safety analysis and on-line operational support for Light Water Reactor and Advanced Gas Cooled Reactor plant. The package consists of modern rationalized generic codes for lattice physics (WIMS), whole reactor calculations (PANTHER), thermal hydraulics (VIPRE) and fuel performance (ENIGMA). These codes, written in FORTRAN77, are highly portable and new developments have followed modern quality assurance standards. These codes can all be run ''stand-alone'' but they are also being integrated within a new UNIX-based interactive system called the Reactor Physics Workbench (RPW). The RPW provides an interactive user interface and a sophisticated data management system. It offers quality assurance features to the user and has facilities for defining complex calculational sequences. The Paper reviews the current capabilities of these components, their integration within the package and outlines future developments underway. Finally, the Paper describes the development of an on-line version of this package which is now being commissioned on UK AGR stations. (author)

Pyrolysis of aseptic packages (tetrapak) in a laboratory screw type reactor and secondary thermal/catalytic tar decomposition

Energy Technology Data Exchange (ETDEWEB)

Haydary, J., E-mail: juma.haydary@stuba.sk [Institute of Chemical and Environmental Engineering, Faculty of Chemical and Food Technology, Slovak University of Technology, Radlinského 9, 812 37 Bratislava (Slovakia); Susa, D.; Dudáš, J. [Institute of Chemical and Environmental Engineering, Faculty of Chemical and Food Technology, Slovak University of Technology, Radlinského 9, 812 37 Bratislava (Slovakia)

2013-05-15

Highlights: ► Pyrolysis of aseptic packages was carried out in a laboratory flow reactor. ► Distribution of tetrapak into the product yields was obtained. ► Composition of the pyrolysis products was estimated. ► Secondary thermal and catalytic decomposition of tars was studied. ► Two types of catalysts (dolomite and red clay marked AFRC) were used. - Abstract: Pyrolysis of aseptic packages (tetrapak cartons) in a laboratory apparatus using a flow screw type reactor and a secondary catalytic reactor for tar cracking was studied. The pyrolysis experiments were realized at temperatures ranging from 650 °C to 850 °C aimed at maximizing of the amount of the gas product and reducing its tar content. Distribution of tetrapak into the product yields at different conditions was obtained. The presence of H{sub 2}, CO, CH{sub 4}, CO{sub 2} and light hydrocarbons, HCx, in the gas product was observed. The Aluminum foil was easily separated from the solid product. The rest part of char was characterized by proximate and elemental analysis and calorimetric measurements. The total organic carbon in the tar product was estimated by elemental analysis of tars. Two types of catalysts (dolomite and red clay marked AFRC) were used for catalytic thermal tar decomposition. Three series of experiments (without catalyst in a secondary cracking reactor, with dolomite and with AFRC) at temperatures of 650, 700, 750, 800 and 850 °C were carried out. Both types of catalysts have significantly affected the content of tars and other components in pyrolytic gases. The effect of catalyst on the tetrapack distribution into the product yield on the composition of gas and on the total organic carbon in the tar product is presented in this work.

In-Package Chemistry Abstraction

Energy Technology Data Exchange (ETDEWEB)

E. Thomas

2004-11-09

This report was developed in accordance with the requirements in ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The purpose of the in-package chemistry model is to predict the bulk chemistry inside of a breached waste package and to provide simplified expressions of that chemistry as function of time after breach to Total Systems Performance Assessment for the License Application (TSPA-LA). The scope of this report is to describe the development and validation of the in-package chemistry model. The in-package model is a combination of two models, a batch reactor model that uses the EQ3/6 geochemistry-modeling tool, and a surface complexation model that is applied to the results of the batch reactor model. The batch reactor model considers chemical interactions of water with the waste package materials and the waste form for commercial spent nuclear fuel (CSNF) waste packages and codisposed waste packages that contain both high-level waste glass (HLWG) and DOE spent fuel. The surface complexation model includes the impact of fluid-surface interactions (i.e., surface complexation) on the resulting fluid composition. The model examines two types of water influx: (1) the condensation of water vapor that diffuses into the waste package, and (2) seepage water that enters the waste package from the drift as a liquid. (1) Vapor Influx Case: The condensation of vapor onto the waste package internals is simulated as pure H2O and enters at a rate determined by the water vapor pressure for representative temperature and relative humidity conditions. (2) Water Influx Case: The water entering a waste package from the drift is simulated as typical groundwater and enters at a rate determined by the amount of seepage available to flow through openings in a breached waste package. TSPA-LA uses the vapor influx case for the nominal scenario for simulations where the waste

In-Package Chemistry Abstraction

International Nuclear Information System (INIS)

E. Thomas

2004-01-01

This report was developed in accordance with the requirements in ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The purpose of the in-package chemistry model is to predict the bulk chemistry inside of a breached waste package and to provide simplified expressions of that chemistry as function of time after breach to Total Systems Performance Assessment for the License Application (TSPA-LA). The scope of this report is to describe the development and validation of the in-package chemistry model. The in-package model is a combination of two models, a batch reactor model that uses the EQ3/6 geochemistry-modeling tool, and a surface complexation model that is applied to the results of the batch reactor model. The batch reactor model considers chemical interactions of water with the waste package materials and the waste form for commercial spent nuclear fuel (CSNF) waste packages and codisposed waste packages that contain both high-level waste glass (HLWG) and DOE spent fuel. The surface complexation model includes the impact of fluid-surface interactions (i.e., surface complexation) on the resulting fluid composition. The model examines two types of water influx: (1) the condensation of water vapor that diffuses into the waste package, and (2) seepage water that enters the waste package from the drift as a liquid. (1) Vapor Influx Case: The condensation of vapor onto the waste package internals is simulated as pure H2O and enters at a rate determined by the water vapor pressure for representative temperature and relative humidity conditions. (2) Water Influx Case: The water entering a waste package from the drift is simulated as typical groundwater and enters at a rate determined by the amount of seepage available to flow through openings in a breached waste package. TSPA-LA uses the vapor influx case for the nominal scenario for simulations where the waste package has been

Alternatives for packaging and transport of greater-than-class C low-level waste

International Nuclear Information System (INIS)

Smith, R.I.

1990-06-01

Viable methods for packaging greater-than-class C (GTCC) low-level wastes and for transporting those wastes from the waste generator sites or from an eastern interim storage site to the Yucca Mountain repository site have been identified and evaluated. Estimated costs for packaging and transporting the population of GTCC wastes expected to be accumulated through the year 2040 have been developed for three waste volume scenarios, for two preferred packaging methods for activated metals from reactor operations and from reactor decommissioning, and for two packaging density assumptions for the activated metals from reactor decommissioning. 7 refs. 7 tabs

Dry reloading and packaging of spent fuel at TRIGA MARK I reactor of Medical University Hanover (MHH), Germany

International Nuclear Information System (INIS)

Haferkamp, D.

2008-01-01

Between 1994 and 1998 the equipment for dry reloading of a research reactor was developed by Noell, which was funded by the German Federal Government and State of Saxonia. The task of this development programme was the design and delivery of an equipment able to load the spent fuel into the shipping casks in a dry mode for research reactors, where wet loading inside the storage pool is impossible. ALARA and infrastructure conditions had to be taken into consideration. Most of the research reactors of TRIGA MARK I type or WWR-SM have operating modes for handling of spent fuel inside the pond or for transfer of spent fuel from pond to dry/wet storage pools. On the other hand, most of them cannot handle heavy weighted shipping casks inside the reactor building because of the crane capacity, or inside water pool because of dimensions and weight of shipping casks. A typical licensed normal operating procedure for spent fuel in research reactors (TRIGA MARK I) is shown. Dry unloading procedure is described. Additionally to the normal operating procedures at the MHH research reactor the following steps were necessary: - dry packaging of spent fuel elements into the loading units (six packs) in order to minimise the transfer and loading steps between the pool and shipping cask; - transfer of spent fuel loading units from dry storage pool to the shipping cask (outside the reactor building) in a shielded transfer cask; - dry reloading of loading units, into the shipping casks outside the reactor building. The Dry Reloading Equipment implies the following 5 items: 1. loading units (six packs), which includes: - capacity up to six spent fuel elements; - criticality safe placement of spent fuel elements; - handling of several spent fuel elements in an aluminium loading unit. 2. Special Transfer Cask, which includes: - shielded housing with locks; - gripper inside housing; - hoist outside housing; - computer aided operation mode for loading and unloading. 3. Transfer Vehicle

Packaging radioactive wastes for geologic disposal

International Nuclear Information System (INIS)

Benton, H.A.

1996-01-01

The M ampersand O contractor for the DOE Office of Civilian Radioactive Waste Management is developing designs of waste packages that will contain the spent nuclear fuel assemblies from commercial and Navy reactor plants and various civilian and government research reactor plants, as well as high-level wastes vitrified in glass. The safe and cost effective disposal of the large and growing stockpile of nuclear waste is of national concern and has generated political and technical debate. This paper addresses the technical aspects of disposing of these wastes in large and robust waste packages. The paper discusses the evolution of waste package design and describes the current concepts. In addition, the engineering and regulatory issues that have governed the development are summarized and the expected performance in meeting the requirements are discussed

Thermal Analysis of the SAFKEG Package for Long Term Storage

International Nuclear Information System (INIS)

NARENDRA, GUPTA

2005-01-01

Interim plutonium storage for up to 10 years in the K-reactor building is currently being planned at Savannah River Site (SRS). SAFKEG package could be used to store Pu metal and oxide (PuO2) in the K-reactor complex with other packagings like 9975. The SAFKEG is designed for carrying Type-B materials across the DOE complex and meets the 10CFR71 requirements. Thermal analyses were performed to ensure that the temperatures of the SAFKEG components will not exceed their temperature limits under the K-reactor storage conditions. Thermal analyses of the SAFKEG packaging with three content configurations using BNFL 3013 outer container (Rocky Flats, SRS bagless transfer cans, and BNFL inner containers) were performed for storage of PuO2 and plutonium metal

Validation and verification of the MTRPC thermohydraulic package

International Nuclear Information System (INIS)

Doval, Alicia

1998-01-01

The MTR P C v2.6 is a computational package developed for research reactor design and calculation. It covers three of the main aspects of a research reactor: neutronic, shielding and thermohydraulic. In this work only the thermohydraulic package will be covered, dealing with verification and validation aspects. The package consists of the following steady state programs: CAUDVAP 2.60 for the hydraulic calculus, estimates the velocity distribution through different parallel channels connected to a common inlet and outlet common plenum. TERMIC 1H v3.0, used for the thermal design of research reactors, provides information about heat flux for a given maximum wall temperature, onset of nucleate boiling, redistribution phenomena and departure from nucleate boiling. CONVEC V3.0 allows natural convection calculations, giving information on heat fluxes for onset of nucleate boiling, pulsed and burn-out phenomena as well as total coolant flow. Results have been validated against experimental values and verified against theoretical and computational programmes results, showing a good agreement. (author)

Spent fuel packaging and its safety analysis

International Nuclear Information System (INIS)

Takada, Kimitaka; Nakaoki, Kozo; Tamamura, Tadao; Matsuda, Fumio; Fukudome, Kazuyuki

1983-01-01

An all stainless steel B(U) type packaging is proposed to transport spent fuels discharged from research reactors and other radioactive materials. The package is used dry and provided with surface fins to absorb drop shock and to dissipate decay heat. Safety was analyzed for structural, thermal, containment shielding and criticality factors, and the integrity of the package was confirmed with the MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, and KENO computer codes. (author)

Reactor calculations for improving utilization of TRIGA reactor

International Nuclear Information System (INIS)

Ravnik, M.

1986-01-01

A brief review of our work on reactor calculations of 250 kW TRIGA with mixed core (standard + FLIP fuel) will be presented. The following aspects will be treated: - development of computer programs; - optimization of in-core fuel management with respect to fuel costs and irradiation channels utilization. TRIGAP programme package will be presented as an example of computer programs. It is based on 2-group 1-D diffusion approximation and besides calculations offers possibilities for operational data logging and fuel inventory book-keeping as well. It is developed primarily for the research reactor operators as a tool for analysing reactor operation and fuel management. For this reason it is arranged for a small (PC) computer. Second part will be devoted to reactor physics properties of the mixed cores. Results of depletion calculations will be presented together with measured data to confirm some general guidelines for optimal mixed core fuel management. As the results are obtained using TRIGAP program package results can be also considered as an illustration and qualification for its application. (author)

Nuclear criticality safety assessment of ORR, NBS, and HFBR fuel element shipping package

International Nuclear Information System (INIS)

Thomas, J.T.

1979-01-01

A fuel element shipping package employing a borated-phenolic foam as a thermal insulating material is designed to transport as many as seven fuel elements for use in the Oak Ridge Research Reactor, the Brookhaven Fast Beam Reactor, or the National Bureau of Standards Reactor. This report presents the criticality safety evaluation and demonstrates that the requirements for a Fissile Class I package are satisfied by the design

Packaging development needs to support environmental restoration

International Nuclear Information System (INIS)

Hummer, J.H.; Kuklinski, J.L.

1993-01-01

The U.S. Department of Energy is bringing its facilities into compliance with present environmental protection regulations. At the Hanford Site, this includes cleanup of its vast nuclear and chemical wastes. Cleanup will involve extensive collecting, consolidating, and processing of radioactive and other hazardous wastes. The Hanford Site was established by the Federal government in 1943 to produce plutonium. Natural uranium was fabricated into fuel slugs, inserted into nuclear reactors, and converted into plutonium. The irradiated slugs were then sent through plutonium extraction facilities. Process waste was discharged to the ground, stored on-site, or shipped off-site for disposal. Activities grew to include nine production reactors, five coal-fired power plants, nuclear fuel fabrication, other support facilities including underground waste storage tanks, and numerous chemical and waste processing plants. Cleanup activities will require extensive transport of radioactive and other hazardous materials. Packaging developments and research are required in the following areas to enhance environmental cleanup; (1) Packaging for Large Contaminated and Activated Components. (2) Bulk Packaging for Contaminated Solids. (3) Bulk Packaging for Contaminated Liquids. (4) Environmental Samples. (J.P.N.)

COOLOD, Steady-State Thermal Hydraulics of Research Reactors

International Nuclear Information System (INIS)

Kaminaga, Masanori

1997-01-01

1 - Description of program or function: The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is a revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode. A 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is a subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. 2 - Method of solution: The 'Heat Transfer Package' is a subprogram for calculating heat transfer coefficients, ONB temperature, heat flux at onset of flow instability and DNB heat flux. The 'Heat transfer package' was especially developed for research reactors which are operated under low pressure and low temperature conditions using plate-type fuel, just like the JRR-3M. Heat transfer correlations adopted in the 'Heat Transfer Package' were obtained or estimated based on the heat transfer experiments in which thermal-hydraulic features of the upgraded JRR-3 core were properly reflected. The 'Heat Transfer Package' is applicable to upward and downward flow

RASPLAV package

International Nuclear Information System (INIS)

1990-01-01

The RASPLAV package for investigation of post-accident mass transport and heat transfer processes is presented. The package performs three dimensional thermal conduction calculations in space nonuniform and temperature dependent conductivities and variable heat sources, taking into account phase transformations. The processes of free-moving bulk material, mixing of melting fuel due to advection and dissolution, and also evaporation/adsorption are modelled. Two-dimensional hydrodynamics with self-consistent heat transfer are also performed. The paper briefly traces the ways the solution procedures are carried out in the program package and outlines the major results of the simulation of reactor vessel melting after a core meltdown. The theoretical analysis and the calculations in this case were carried out in order to define the possibility of localization of the zone reminders. The interactions between the reminders and the concrete are simulated and evaluation of the interaction parameters is carried out. 4 refs. (R.Ts)

PWR internals segmentation and packaging experience in the U.S

International Nuclear Information System (INIS)

Kreitman, P.J.

2008-01-01

Seven commercial nuclear power plants of the Pressurized Water Reactor (PWR) design have been permanently shut down in the US to date. Six of these plants have been decommissioned using dismantling methods. The remaining plant, Indian Point 1, located in Buchanan, NY, has been placed in Safe Storage. Of the six dismantled plants, five underwent extensive segmentation, separation, and packaging of their internal components to allow removal and disposal of the reactor vessel assembly. Only the Trojan Plant in Portland, Oregon, was able to ship the reactor vessel assembly with its internals intact to the final disposal site near Richland, Washington. An important part of the planning for the upcoming decommissioning of similar plants in Europe such as Chooz A in France and Zorita in Spain will be to study the lessons learned from the US efforts and apply the best practices to future projects. This paper will chronicle the evolution of the reactor internals segmentation and packaging process to date, including the planning, methodology, equipment, waste management, and packaging strategy. (author)

IAEA activities in nuclear reactors simulation for educational purposes

International Nuclear Information System (INIS)

Lyon, R.B.

1998-01-01

The International Atomic Energy Agency (IAEA) has established a programme in nuclear reactor simulation computer programs to assist its Member States in education and training. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and training material, sponsors training courses and workshops, and distributes documentation and computer programs. Currently, the IAEA has two simulation programs: the Classroom-based Advanced Reactor Demonstrators package, and the Advanced Reactor Simulator. Both packages simulate the behaviour of BWR, PWR and HWR reactor types. For each package, the modeling approach and assumptions are broadly described, together with a general description of the operation of the computer programs. (author)

TU Electric reactor physics model verification: Power reactor benchmark

International Nuclear Information System (INIS)

Willingham, C.E.; Killgore, M.R.

1988-01-01

Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

ROBATEL RS 24 packaging

International Nuclear Information System (INIS)

Robatel, M.; Bochard, C.

1986-01-01

The ROBATEL RS 24 packaging for light-water reactor fuel assemblies is discussed. The cask uses a strength frame of carbon steel sheets, lead gamma ray shielding, neutron shielding of aluminous concrete, a thermal protection system for the gamma shielding, and a natural convection thermal transfer system

Case study for one-piece removal method of reactor vessel of nuclear ship 'Mutsu'

International Nuclear Information System (INIS)

Nagane, Satoru; Kitahara, Katsumi; Yoshikawa, Seiji; Miyasaka, Yasuhiko; Fukumura, Nobuo; Nisizawa, Ichiou

2010-01-01

A reactor installed at the center part of the nuclear ship 'Mutsu' has been stored safely and exhibited in a reactor room building since 1996. The reactor vessel and its internals are key components because of main radioactive wastes for the reasonable decommissioning plan in the future. This report describes the one-piece removal method as the one package of the reactor vessel with its internals intact with a shipping container or additional shields. The reactor vessel package (Max.100ton) will be classified acceptable for burial at the low level radioactive waste (LLW), which will be buried at a LLW pit facility under waste disposal regulations. And also, the package will be classified as an IP-2-equivalent package according to the requirement for Shipments and Packagings. (author)

Criticality benchmark guide for light-water-reactor fuel in transportation and storage packages

International Nuclear Information System (INIS)

Lichtenwalter, J.J.; Bowman, S.M.; DeHart, M.D.; Hopper, C.M.

1997-03-01

This report is designed as a guide for performing criticality benchmark calculations for light-water-reactor (LWR) fuel applications. The guide provides documentation of 180 criticality experiments with geometries, materials, and neutron interaction characteristics representative of transportation packages containing LWR fuel or uranium oxide pellets or powder. These experiments should benefit the U.S. Nuclear Regulatory Commission (NRC) staff and licensees in validation of computational methods used in LWR fuel storage and transportation concerns. The experiments are classified by key parameters such as enrichment, water/fuel volume, hydrogen-to-fissile ratio (H/X), and lattice pitch. Groups of experiments with common features such as separator plates, shielding walls, and soluble boron are also identified. In addition, a sample validation using these experiments and a statistical analysis of the results are provided. Recommendations for selecting suitable experiments and determination of calculational bias and uncertainty are presented as part of this benchmark guide

Benchmarking of FA2D/PARCS Code Package

International Nuclear Information System (INIS)

Grgic, D.; Jecmenica, R.; Pevec, D.

2006-01-01

FA2D/PARCS code package is used at Faculty of Electrical Engineering and Computing (FER), University of Zagreb, for static and dynamic reactor core analyses. It consists of two codes: FA2D and PARCS. FA2D is a multigroup two dimensional transport theory code for burn-up calculations based on collision probability method, developed at FER. It generates homogenised cross sections both of single pins and entire fuel assemblies. PARCS is an advanced nodal code developed at Purdue University for US NRC and it is based on neutron diffusion theory for three dimensional whole core static and dynamic calculations. It is modified at FER to enable internal 3D depletion calculation and usage of neutron cross section data in a format produced by FA2D and interface codes. The FA2D/PARCS code system has been validated on NPP Krsko operational data (Cycles 1 and 21). As we intend to use this code package for development of IRIS reactor loading patterns the first logical step was to validate the FA2D/PARCS code package on a set of IRIS benchmarks, starting from simple unit fuel cell, via fuel assembly, to full core benchmark. The IRIS 17x17 fuel with erbium burnable absorber was used in last full core benchmark. The results of modelling the IRIS full core benchmark using FA2D/PARCS code package have been compared with reference data showing the adequacy of FA2D/PARCS code package model for IRIS reactor core design.(author)

IAEA activities in nuclear reactor simulation for educational purposes

International Nuclear Information System (INIS)

Lyon, R.B.

2001-01-01

The International Atomic Energy Agency (IAEA) has established a programme in nuclear reactor simulation computer programs to assist its Member States in education and training. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and training material, sponsors training courses and workshops, and distributes documentation and computer programs. Two simulation programs are presented at this workshop: the Classroom-based Advanced Reactor Demonstrators package, and the Advanced Reactor Simulator. Both packages simulate the behaviour of BWR, PWR and HWR reactor types. For each package, the modeling approach and assumptions are broadly described, together with a general description of the operation of the computer programs. (author)

Feedback from Westinghouse experience on segmentation of reactor vessel internals - 59013

International Nuclear Information System (INIS)

Kreitman, Paul J.; Boucau, Joseph; Segerud, Per; Fallstroem, Stefan

2012-01-01

With more than 25 years of experience in the development of reactor vessel internals segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. Building on tooling concepts and cutting methodologies developed decades ago for the successful removal of nuclear fuel from the damaged Three Mile Island Unit 2 reactor (TMI-2), Westinghouse has continuously improved its approach to internals segmentation and packaging by incorporating lessons learned and best practices into each successive project. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive water-jet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Westinghouse has applied its technology to all types of reactors covering Pressurized Water Reactors (PWR's), Boiling Water Reactors (BWR's), Gas Cooled Reactors (GCR's) and sodium reactors. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since space is almost always a limiting factor it is therefore important to plan and optimize the available room in the segmentation areas. The choice of the optimum cutting technology is important for a successful project implementation and depends on some specific constraints like disposal costs, project schedule, available areas or safety. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. Westinghouse has also developed a variety of special handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a

Safety Evaluation for Packaging for the N Reactor/single pass reactor fuel characterization shipments

International Nuclear Information System (INIS)

Stevens, P.F.

1994-01-01

The purpose of this Safety Evaluation for Packaging (SEP) is to authorize the ChemNuclear CNS 1-13G packaging to ship samples of irradiated fuel elements from the 100 K East and 100 K West basins to the Postirradiation Testing Laboratory (PTL) in support of the spent nuclear fuel characterization effort. It also authorizes the return of the fuel element samples to the 100 K East facility using the same packaging. The CNS 1-13G cask has been-chosen to transport the fuel because it has a Certificate of Compliance (CoC) issued by the US Nuclear Regulatory Commission (NRC) for transporting irradiated oxide and metal fuel in commerce. It is capable of being loaded and offloaded underwater and may be shipped with water in the payload compartment

Safety analysis report: packages. DOT specification 7A - Type A container Mark 15 sludge shipping package

International Nuclear Information System (INIS)

Zeh, C.W.

1985-03-01

Sludge or filter cake containing 1.1 wt % 235 U enriched uranium from Mark 15 fabrication will be packaged in 55-gallon containers and shipped from SRP to NLO, Fernald, Ohio for recovery of product. About 7 Metric Tons (MT) of filter cake will be produced from fabricating Mark 15 slugs each reactor charge. Packaged shipments of this material, consigned as exclusive use, will be shipped to NLO in Department of Transportation (DOT) Specification 7A - Type A packages under a DOE Certificate of Compliance for Fissile Class III shipments. ''Type A packaging'' is designed to retain containment and shielding integrity under normal conditions of transport. This report documents compliance of the package design, construction methods, material and test performance with DOT Specification 7A. This DOT 7A Type A package contains a carbon steel outer container which is a 0.060-in.-thick 55-gal, galvanized drum equipped with a gasketed closure. The outer container encloses a 0.090-in.-thick, open head, polyethylene liner with lid

Computational geometry for reactor applications

International Nuclear Information System (INIS)

Brown, F.B.; Bischoff, F.G.

1988-01-01

Monte Carlo codes for simulating particle transport involve three basic computational sections: a geometry package for locating particles and computing distances to regional boundaries, a physics package for analyzing interactions between particles and problem materials, and an editing package for determining event statistics and overall results. This paper describes the computational geometry methods in RACER, a vectorized Monte Carlo code used for reactor physics analysis, so that comparisons may be made with techniques used in other codes. The principal applications for RACER are eigenvalue calculations and power distributions associated with reactor core physics analysis. Successive batches of neutrons are run until convergence and acceptable confidence intervals are obtained, with typical problems involving >10 6 histories. As such, the development of computational geometry methods has emphasized two basic needs: a flexible but compact geometric representation that permits accurate modeling of reactor core details and efficient geometric computation to permit very large numbers of histories to be run. The current geometric capabilities meet these needs effectively, supporting a variety of very large and demanding applications

Packaging LLW and ILW

International Nuclear Information System (INIS)

Flowers, R.H.; Owen, R.G.

1991-01-01

Low level waste (LLW) accounts for 70-80% by volume of all radioactive wastes produced by the nuclear industry. It has low specific activity, negligible actinide content and requires little, if any, shielding to protect workers. Volume reduction for LLW of high volume but low density may be achieved by incineration and compaction as appropriate, before packaging for disposal by near surface burial. Intermediate level waste (ILW) is treated and packed to convert it into a stable form to minimize any release of activity and make handling easier. The matrix chosen for immobilization, usually cement, polymers or bitumen, depends on the nature of the waste and the acceptance criteria of the disposal facility. The special case of LLW and ILW which will arise from reactor decommissioning is discussed. Packaging methods adopted by individual countries are reviewed. The range of costs involved for packaging ILW is indicated. There is no international consensus on the performance required from packaged waste to ensure its suitability both for interim storage and final disposal. (UK)

In-Package Chemistry Abstraction

Energy Technology Data Exchange (ETDEWEB)

P.S. Domski

2003-07-21

The work associated with the development of this model report was performed in accordance with the requirements established in ''Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of SR and LA'' (BSC 2002a). The in-package chemistry model and in-package chemistry model abstraction are developed to predict the bulk chemistry inside of a failed waste package and to provide simplified expressions of that chemistry. The purpose of this work is to provide the abstraction model to the Performance Assessment Project and the Waste Form Department for development of geochemical models of the waste package interior. The scope of this model report is to describe the development and validation of the in-package chemistry model and in-package chemistry model abstraction. The in-package chemistry model will consider chemical interactions of water with the waste package materials and the waste form for commercial spent nuclear fuel (CSNF) and codisposed high-level waste glass (HLWG) and N Reactor spent fuel (CDNR). The in-package chemistry model includes two sub-models, the first a water vapor condensation (WVC) model, where water enters a waste package as vapor and forms a film on the waste package components with subsequent film reactions with the waste package materials and waste form--this is a no-flow model, the reacted fluids do not exit the waste package via advection. The second sub-model of the in-package chemistry model is the seepage dripping model (SDM), where water, water that may have seeped into the repository from the surrounding rock, enters a failed waste package and reacts with the waste package components and waste form, and then exits the waste package with no accumulation of reacted water in the waste package. Both of the submodels of the in-package chemistry model are film models in contrast to past in-package chemistry models where all of the waste package pore space was filled with water. The

In-Package Chemistry Abstraction

International Nuclear Information System (INIS)

P.S. Domski

2003-01-01

The work associated with the development of this model report was performed in accordance with the requirements established in ''Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of SR and LA'' (BSC 2002a). The in-package chemistry model and in-package chemistry model abstraction are developed to predict the bulk chemistry inside of a failed waste package and to provide simplified expressions of that chemistry. The purpose of this work is to provide the abstraction model to the Performance Assessment Project and the Waste Form Department for development of geochemical models of the waste package interior. The scope of this model report is to describe the development and validation of the in-package chemistry model and in-package chemistry model abstraction. The in-package chemistry model will consider chemical interactions of water with the waste package materials and the waste form for commercial spent nuclear fuel (CSNF) and codisposed high-level waste glass (HLWG) and N Reactor spent fuel (CDNR). The in-package chemistry model includes two sub-models, the first a water vapor condensation (WVC) model, where water enters a waste package as vapor and forms a film on the waste package components with subsequent film reactions with the waste package materials and waste form--this is a no-flow model, the reacted fluids do not exit the waste package via advection. The second sub-model of the in-package chemistry model is the seepage dripping model (SDM), where water, water that may have seeped into the repository from the surrounding rock, enters a failed waste package and reacts with the waste package components and waste form, and then exits the waste package with no accumulation of reacted water in the waste package. Both of the submodels of the in-package chemistry model are film models in contrast to past in-package chemistry models where all of the waste package pore space was filled with water. The current in-package

Reactor physics computer code development for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs

International Nuclear Information System (INIS)

Rastogi, B.P.

1989-01-01

This report discusses various reactor physics codes developed for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs. These code packages have been utilized for nuclear design of 500 MWe and new 235 MWe PHWRs. (author)

Validation and verification of the MTR{sub P}C thermohydraulic package

Energy Technology Data Exchange (ETDEWEB)

Doval, Alicia [INVAP S.E., Bariloche, Rio Negro (Argentina). Nuclear Engineering Dept.]. E-mail: doval@invap.com.ar

1998-07-01

The MTR{sub P}C v2.6 is a computational package developed for research reactor design and calculation. It covers three of the main aspects of a research reactor: neutronic, shielding and thermohydraulic. In this work only the thermohydraulic package will be covered, dealing with verification and validation aspects. The package consists of the following steady state programs: CAUDVAP 2.60 for the hydraulic calculus, estimates the velocity distribution through different parallel channels connected to a common inlet and outlet common plenum. TERMIC 1H v3.0, used for the thermal design of research reactors, provides information about heat flux for a given maximum wall temperature, onset of nucleate boiling, redistribution phenomena and departure from nucleate boiling. CONVEC V3.0 allows natural convection calculations, giving information on heat fluxes for onset of nucleate boiling, pulsed and burn-out phenomena as well as total coolant flow. Results have been validated against experimental values and verified against theoretical and computational programmes results, showing a good agreement. (author)

Packaging Design Criteria for the MCO Cask

International Nuclear Information System (INIS)

FLANAGAN, B.D.

2000-01-01

Approximately 2,100 metric tons of unprocessed, irradiated, nuclear fuel elements are presently stored in the K Basins (including approximately 700 additional elements from the Plutonium-Uranium Extraction Plant, N Reactor, and 327 Laboratory). To permit cleanup of the K Basins and fuel conditioning, the fuel will be transported from the 100 K Area to a Canister Storage Building (CSB) in the 200 East Area. The purpose of this packaging design criteria is to provide criteria for the design, fabrication, and use of a packaging system to transport the large quantities of irradiated nuclear fuel elements positioned within Multi-canister Overpacks. Concurrent with the K Basin cleanup, 72 Shippingport Pressurized Water Reactor Core 2 fuel assemblies will be transported from T Plant to the CSB to provide space at T Plant for K Basin sludge canisters

44 years of testing radioactive materials packages at ORNL

Energy Technology Data Exchange (ETDEWEB)

Shappert, L.B.; Ludwig, S.B. [Oak Ridge National Lab., Oak Ridge, TN (United States)

2004-07-01

This paper briefly reviews the package testing at the Oak Ridge National Laboratory (ORNL) since 1960 and then examines the trends in the testing activities that occurred during the same period. Radioactive material shipments have been made from ORNL since the 1940s. The first fully operating reactor built at the ORNL site was patterned after the graphite pile constructed by Enrico Fermi under Stagg Field in Chicago. After serving as a test bed for future reactors, it became useful as a producer of radioactive isotopes. The Isotopes Division was established at ORNL to furnish radioactive materials used in the medical community. Often these shipments have been transported by aircraft worldwide due to the short half-lives of many of the materials. This paper touches briefly on the lighter and smaller radioisotope packages that were being shipped from ORNL in large numbers and then deals with the testing of packages designed to handle large radioactive sources, such as spent fuel, and other fissile materials.

44 years of testing radioactive materials packages at ORNL

International Nuclear Information System (INIS)

Shappert, L.B.; Ludwig, S.B.

2004-01-01

This paper briefly reviews the package testing at the Oak Ridge National Laboratory (ORNL) since 1960 and then examines the trends in the testing activities that occurred during the same period. Radioactive material shipments have been made from ORNL since the 1940s. The first fully operating reactor built at the ORNL site was patterned after the graphite pile constructed by Enrico Fermi under Stagg Field in Chicago. After serving as a test bed for future reactors, it became useful as a producer of radioactive isotopes. The Isotopes Division was established at ORNL to furnish radioactive materials used in the medical community. Often these shipments have been transported by aircraft worldwide due to the short half-lives of many of the materials. This paper touches briefly on the lighter and smaller radioisotope packages that were being shipped from ORNL in large numbers and then deals with the testing of packages designed to handle large radioactive sources, such as spent fuel, and other fissile materials

Safety Analysis Report for the KRI-ASM Transport Package

Energy Technology Data Exchange (ETDEWEB)

Bang, K. S.; Lee, J. C.; Kim, D. H.; Park, H. Y.; Kim, J. B.; Kim, H. J.; Seo, K. S

2005-11-15

Safety evaluation for the KRI-ASM transport package to transport safely I-131, which is produced at HANARO research reactor in KAERI, was carried out. In the safety analyses results for the KRI-ASM transport package, all the maximum stresses as well as the maximum temperature of the surface are lower than their allowable limits. The safety tests were performed by using the test model of the KRI-ASM transport package. Leak Test was performed after drop test and penetration test, the measured leakage rate was lower than allowable leakage rate. It is revealed that the containment integrity of the KRI-ASM transport package is maintained. Therefore, it shows that the integrity of the KRI-ASM transport package is well maintained.

Development of packagings for 'MONJU' blanket fuel assemblies

International Nuclear Information System (INIS)

Shibata, Kan; Ouchi, Yuichiro; Matsuzaki, Masaaki; Okuda, Yoshihisa

1995-01-01

Blanket assemblies for prototype Fast Breeder Reactor 'MONJU' are made at commercial fuel fabrication plants capable of handling deplete Uranium in Japan. For the purpose of transport the assemblies are inserted into a packaging that is set horizontally at the fabrication plants because of compatibility with equipment installed at the plants. On the other hand, the assemblies must be taken out from the packaging set vertically at 'MONJU' due to compatibility. For this reason development of a new packaging, which makes it possible to take assemblies in and out both horizontally and vertically, is needed to carry out transport of assemblies for reload. The development and fabrication of the packagings, taking about two years, were completed in March 1995. The packagings were used in transport of assemblies in June 1995 for the first change. This report introduces the outline of the packaging and confirmation tests done in the process of development. (author)

Certifying the decommissioned Shippingport reactor vessel for transport

International Nuclear Information System (INIS)

Towell, R.H.

1990-01-01

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

Conceptual Assessment of a Fresh Fuel Transport Package for KJRR

Energy Technology Data Exchange (ETDEWEB)

Lee, Ju-Chan; Choi, W. S.; Bang, K. S.; Yu, S. H.; Park, J. S.; Yang, Y. Y. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The IAEA and domestic regulations stipulate that the fissile material transport package be subjected to the cumulative effects of a 9 m drop, 1 m puncture, 800 ℃ thermal and water leakage tests. A fissile material transport package should be maintained the subcriticality during the normal and accident conditions for contingency of leakage of water into or out of package, rearrangement of the contents, reduction of spaces and temperature changes. KAERI has been developing a fresh fuel transport package for Kijang research reactor (KJRR). This paper describes a conceptual design and preliminary safety analysis of the transport package for KJRR. The transport package was designed for shipment of a fresh fuel and a FM (Fission Molybdenum) target. Low-enriched uranium (LEU) of U-Mo fuel with U-235 enrichment of 19.75 w/o is used as a research reactor fuel. And LEU of UAlx-Al with U-235 enrichment of 19.75 w/o is used as a FM target material. The transport package was designed for shipment of a fresh fuel and a FM target. Safety analyses were conducted on all areas, including criticality, structural, and thermal fields. In the criticality analysis, effective neutron multiplication factors were below the criticality safety limit. In the structural analysis, the maximum stress satisfied the stress requirement stipulated in the ASME code. After 9 m free drop and 1 m puncture test, there was no significant deformation of fuel basket to cause a criticality. In the thermal analysis, the maximum temperatures at each part were lower than the allowable values.

Development of a fresh MOX fuel transport package for disposition of weapons plutonium

International Nuclear Information System (INIS)

Ludwig, S.B.; Pope, R.B.; Shappert, L.B.; Michelhaugh, R.D.; Chae, S.M.

1998-01-01

The US Department of Energy announced its Record of Decision on January 14, 1997, to embark on a dual-track approach for disposition of surplus weapons-usable plutonium using immobilization in glass or ceramics and burning plutonium as mixed-oxide (MOX) fuel in reactors. In support of the MOX fuel alternative, Oak Ridge National Laboratory initiated development of conceptual designs for a new package for transporting fresh (unirradiated) MOX fuel assemblies between the MOX fabrication facility and existing commercial light-water reactors in the US. This paper summarizes progress made in development of new MOX transport package conceptual designs. The development effort has included documentation of programmatic and technical requirements for the new package and development and analysis of conceptual designs that satisfy these requirements

Design of a nuclear reactor cooperative controller

International Nuclear Information System (INIS)

Alang-Rashid, N.K.; Heger, A.S.

1991-01-01

This paper describes the development of a fuzzy logic controller software package and explores the feasibility of its use in nuclear reactor operation. The controller complements reactor operator actions, and the operators can override the controller decisions. Techniques of providing learning capability to the controller are also being investigated to improve the reasoning and control skill of the controller. The fuzzy logic controller is implemented in C language and its overall structure is shown. The heart of the systems consists of a fuzzifier, a rule interpreter, and a defuzzifier. The controller is designed as a stand-alone package that can be interfaced to a simulated model of a nuclear reactor. Since no model is an accurate representation of the actual process being modeled, some tuning must be performed to use the controller in an actual reactor. This is accomplished using the learning feature of the controller

Introduction of SCIENCE code package

International Nuclear Information System (INIS)

Lu Haoliang; Li Jinggang; Zhu Ya'nan; Bai Ning

2012-01-01

The SCIENCE code package is a set of neutronics tools based on 2D assembly calculations and 3D core calculations. It is made up of APOLLO2F, SMART and SQUALE and used to perform the nuclear design and loading pattern analysis for the reactors on operation or under construction of China Guangdong Nuclear Power Group. The purpose of paper is to briefly present the physical and numerical models used in each computation codes of the SCIENCE code pack age, including the description of the general structure of the code package, the coupling relationship of APOLLO2-F transport lattice code and SMART core nodal code, and the SQUALE code used for processing the core maps. (authors)

Extreme temperature packaging: challenges and opportunities

Science.gov (United States)

Johnson, R. Wayne

2016-05-01

Consumer electronics account for the majority of electronics manufactured today. Given the temperature limits of humans, consumer electronics are typically rated for operation from -40°C to +85°C. Military applications extend the range to -65°C to +125°C while underhood automotive electronics may see +150°C. With the proliferation of the Internet of Things (IoT), the goal of instrumenting (sensing, computation, transmission) to improve safety and performance in high temperature environments such as geothermal wells, nuclear reactors, combustion chambers, industrial processes, etc. requires sensors, electronics and packaging compatible with these environments. Advances in wide bandgap semiconductors (SiC and GaN) allow the fabrication of high temperature compatible sensors and electronics. Integration and packaging of these devices is required for implementation into actual applications. The basic elements of packaging are die attach, electrical interconnection and the package or housing. Consumer electronics typically use conductive adhesives or low melting point solders for die attach, wire bonds or low melting solder for electrical interconnection and epoxy for the package. These materials melt or decompose in high temperature environments. This paper examines materials and processes for high temperature packaging including liquid transient phase and sintered nanoparticle die attach, high melting point wires for wire bonding and metal and ceramic packages. The limitations of currently available solutions will also be discussed.

IBM-PC-based reactor neutronics analysis package

International Nuclear Information System (INIS)

Nigg, D.W.; Wessol, D.E.; Grimesey, R.A.; Parsons, D.K.; Wheeler, F.J.; Yoon, W.Y.; Lake, J.A.

1985-01-01

Technical advances over the past few years have led to a situation where a wide range of complex scientific computations can now be done on properly configured microcomputers such as the IBM-PC (personal computer). For a number of reasons, including security, economy, and user convenience, the development of a comprehensive system of reactor neutronics codes suitable for operation on the IBM-PC has been undertaken at the Idaho National Engineering Laboratory (INEL). It is anticipated that a PC-based code system could also have wide applicability in the nuclear engineering education community since conversion of software generated by national laboratories and others to college and university mainframe hardware has historically been a time-consuming process that has sometimes met with only limited success. This paper discusses the philosophy behind the INEL reactor neutronics PC code system and describes those parts of the system that are currently complete, those that are now under development, and those that are still in the planning stage

HORIZONTAL LIFTING OF 5 DHLW/DOE LONG, 12-PWR LONG AND 24-BWR WASTE PACKAGES

International Nuclear Information System (INIS)

V. de la Brosse

2001-01-01

The objective of this calculation was to determine the structural response of a 12-Pressurized Water Reactor (PWR) Long, a 24-Boiling Water Reactor (BWR) and a 5-Defense High Level Waste/Department of Energy (DHLW/DOE)--Long spent nuclear fuel waste packages lifted in a horizontal position. The scope of this calculation was limited to reporting the calculation results in terms of maximum stress intensities in the trunnion collar sleeves. In addition, the maximum stress intensities in the inner and outer shells of the waste packages were presented for illustrative purposes. The information provided by the sketches (Attachments I, II and III) is that of the potential design of the types of waste packages considered in this calculation, and all obtained results are valid for these designs only. This calculation is associated with the waste package design and was performed by the Waste Package Design Section in accordance with the ''Technical work plan for: Waste Package Design Description for LA'' (Ref. 7). AP-3.12Q, Calculations (Ref. 13), was used to perform the calculation and develop the document

Improving plant availability by predicting reactor trips

International Nuclear Information System (INIS)

Frank, M.V.; Epstein, S.A.

1986-01-01

Management Ahnalysis Company (MAC) has developed and applied two complementary software packages called RiTSE and RAMSES. Together they provide an mini-computer workstation for maintenance and operations personnel to dramatically reduce inadvertent reactor trips. They are intended to be used by those responsible at the plant for authorizing work during operation (such as a clearance coordinator or shift foreman in U.S. plants). They discover and represent all components, processes, and their interactions that could case a trip. They predict if future activities at the plant would cause a reactor trip, provide a reactor trip warning system and aid in post-trip cause analysis. RAMSES is a general reliability engineering software package that uses concepts of artificial intelligence to provide unique capabilities on personal and mini-computers

Tritium immobilization and packaging using metal hydrides

International Nuclear Information System (INIS)

Holtslander, W.J.; Yaraskavitch, J.M.

1981-04-01

Tritium recovered from CANDU heavy water reactors will have to be packaged and stored in a safe manner. Tritium will be recovered in the elemental form, T 2 . Metal tritides are effective compounds in which to immobilize the tritium as a stable non-reactive solid with a high tritium capacity. The technology necessary to prepare hydrides of suitable metals, such as titanium and zirconium, have been developed and the properties of the prepared materials evaluated. Conceptual designs of packages for containing metal tritides suitable for transportation and long-term storage have been made and initial testing started. (author)

Thermal Response of the 21-PWR Waste Package to a Fire Accident

International Nuclear Information System (INIS)

F.P. Faucher; H. Marr; M.J. Anderson

2000-01-01

The objective of this calculation is to evaluate the thermal response of the 21-PWR WP (pressurized water reactor waste package) to the regulatory fire event. The scope of this calculation is limited to the two-dimensional waste package temperature calculations to support the waste package design. The information provided by the sketches attached to this calculation (Attachment IV) is that of the potential design of the type of waste package considered in this calculation. The procedure AP-3.12Q.Calculations (Reference 1), and the Development Plan (Reference 24) are used to develop this calculation

Dismantling id the reactor pressure vessel insulation and dissecting of the MZFR reactor pressure vessel

International Nuclear Information System (INIS)

Loeb, Andreas; Stanke, Dieter; Thoma, Markus; Eisenmann, Beata; Prechtl, Erwin; Dehnke, Burckhard

2008-01-01

The MZFR reactor was decommissioned in 1984. The authors describe the dismantling of the reactor pressure vessel insulation that consists of asbestos containing mineral fiber wool. The appropriate remote handling and cutting tools had to be adapted with respect to the restrained space in the containment. The dismantling of the reactor pressure vessel has been completed, the dissected parts have been packaged into 200 containers for the final repository Konrad. During the total project time no reportable events and no damage to persons occurred.

Cleanup Verification Package for the 118-F-1 Burial Ground

Energy Technology Data Exchange (ETDEWEB)

E. J. Farris and H. M. Sulloway

2008-01-10

This cleanup verification package documents completion of remedial action for the 118-F-1 Burial Ground on the Hanford Site. This burial ground is a combination of two locations formerly called Minor Construction Burial Ground No. 2 and Solid Waste Burial Ground No. 2. This waste site received radioactive equipment and other miscellaneous waste from 105-F Reactor operations, including dummy elements and irradiated process tubing; gun barrel tips, steel sleeves, and metal chips removed from the reactor; filter boxes containing reactor graphite chips; and miscellaneous construction solid waste.

Development of a core follow calculational system for research reactors

International Nuclear Information System (INIS)

Muller, E.Z.; Ball, G.; Joubert, W.R.; Schutte, H.C.; Stoker, C.C.; Reitsma, F.

1994-01-01

Over the last few years a comprehensive Pressurized Water Reactor and Materials Testing Reactor core analysis code system based on modern reactor physics methods has been under development by the Atomic Energy Corporation of South Africa. This system, known as OSCAR-3, will incorporate a customized graphical user interface and data management system to ensure user-friendliness and good quality control. The system has now reached the stage of development where it can be used for practical MTR core analyses. This paper describes the current capabilities of the components of the OSCAR-3 package, their integration within the package, and outlines future developments. 10 refs., 1 tab., 1 fig

Environmental impact assessment of a package type IFAS reactor during construction and operational phases: a life cycle approach.

Science.gov (United States)

Singh, Nitin Kumar; Singh, Rana Pratap; Kazmi, Absar Ahmad

2017-05-01

In the present study, a life cycle assessment (LCA) approach was used to analyse the environmental impacts associated with the construction and operational phases of an integrated fixed-film activated sludge (IFAS) reactor treating municipal wastewater. This study was conducted within the boundaries of a research project that aimed to investigate the implementation related challenges of a package type IFAS reactor from an environmental perspective. Along with the LCA results of the construction phase, a comparison of the LCA results of seven operational phases is also presented in this study. The results showed that among all the inputs, the use of stainless steel in the construction phase caused the highest impact on environment, followed by electricity consumption in raw materials production. The impact of the construction phase on toxicity impact indicators was found to be significant compared to all operational phases. Among the seven operational phases of this study, the dissolved oxygen phase III, having a concentration of ∼4.5 mg/L, showed the highest impact on abiotic depletion, acidification, global warming, ozone layer depletion, human toxicity, fresh water eco-toxicity, marine aquatic eco-toxicity, terrestrial eco-toxicity, and photochemical oxidation. However, better effluent quality in this phase reduced the eutrophication load on environment.

The decommissioning of a small nuclear reactor

International Nuclear Information System (INIS)

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

1990-02-01

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

Space reactor system and subsystem investigations: assessment of technology issues for the reactor and shield subsystem. SP-100 Program

International Nuclear Information System (INIS)

Atkins, D.F.; Lillie, A.F.

1983-01-01

As part of Rockwell's effort on the SP-100 Program, preliminary assessment has been completed of current nuclear technology as it relates to candidate reactor/shield subsystems for the SP-100 Program. The scope of the assessment was confined to the nuclear package (to the reactor and shield subsystems). The nine generic reactor subsystems presented in Rockwell's Subsystem Technology Assessment Report, ESG-DOE-13398, were addressed for the assessment

Cost targets for at-reactor spent fuel rod consolidation

International Nuclear Information System (INIS)

Macnabb, W.V.

1985-01-01

The high-level nuclear waste management system in the US currently envisions the disposal of spent fuel rods that have been removed from their assemblies and reconfigured into closely packed arrays. The process of fuel rod removal and packaging, referred to as rod consolidation, can occur either at reactors or at an integrated packaging facility, monitored retrievable storage (MRS). Rod consolidation at reactors results in cost savings down stream of reactors by reducing needs for additional storage, reducing the number of shipments, and reducing (eliminating, in the extreme) the amount of fuel handling and consolidation at the MRS. These savings accrue to the nuclear waste fund. Although private industry is expected to pay for at-reactor activities, including rod consolidation, it is of interest to estimate cost savings to the waste system if all fuel were consolidated at reactors. If there are savings, the US Department of Energy (DOE) may find it advantageous to pay for at-reactor rod consolidation from the nuclear waste fund. This paper assesses and compares the costs of rod consolidation at reactors and at the MRS in order to determine at what levels the former could be cost competitive with the latter

Nuclear Reactor RA Safety Report, Vol. 12, Accidents during reactor operation

International Nuclear Information System (INIS)

1986-11-01

This volume includes description and analysis of typical accidents occurred during operation of RA reactor in chronological order, as follows: contamination of primary coolant circuit; leakage of heavy water from the primary coolant loop; contamination of vertical experimental channel; air contamination in the reactor building and loss of circulation of the primary coolant; failures of the vacuum pump and spent fuel packaging device; rupture of the spent fuel element cladding; dethronement's of capsule for irradiation of fuel element; rupture of the vertical experimental channel and contamination of the surroundings; swelling of a fuel element; appearance of deposits on the surface of the fuel elements cladding. The last chapter describes similar accidents occurred on nuclear reactors in the world [sr

Waste package performance allocation system study report

International Nuclear Information System (INIS)

Memory, R.D.

1994-01-01

The Waste Package Performance Allocation system study was performed in order to provide a technical basis for the selection of the waste package period of substantially complete containment and its resultant contribution to the overall total system performance. This study began with a reference case based on the current Mined Geologic Disposal System (MGDS) baseline design and added a number of alternative designs. The waste package designs were selected from the designs being considered in detail during Advanced Conceptual Design (ACD). The waste packages considered were multi-barrier packages with a 0.95 cm Alloy 825 inner barrier and a 10, 20, or 45 cm thick carbon steel outer barrier. The waste package capacities varied from 6 to 12 to 21 Pressurized Water Reactor (PWR) fuel assemblies. The vertical borehole and in-drift emplacement modes were also considered, as were thermal loadings of 25, 57, and 114 kW/acre. The repository cost analysis indicated that the 21 PWR in-drift emplacement mode option with the 10 cm and 20 cm outer barrier thicknesses are the least expensive and that the 12 PWR in-drift case has approximately the same cost as the 6 PWR vertical borehole. It was also found that the cost increase from the 10 cm outer barrier waste package to the 20 cm waste package was less per centimeter than the increase from the 20 cm outer barrier waste package to the 45 cm outer barrier waste package. However, the repository cost was nearly linear with the outer barrier thickness for the 21 PWR in-drift case. Finally, corrosion rate estimates are provided and the relationship of repository cost versus waste package lifetime is discussed as is cumulative radionuclide release from the waste package and to the accessible environment for time periods of 10,000 years and 100,000 years

EQ6 Calculations for Chemical Degradation Of N Reactor (U-Metal) Spent Nuclear Fuel Waste Packages

International Nuclear Information System (INIS)

P. Bernot

2001-01-01

The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management and Operating Contractor (CRWMS M and O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the N Reactor, a graphite moderated reactor at the Department of Energy's (DOE) Hanford Site (ref. 1). The N Reactor core was fueled with slightly enriched (0.947 wt% and 0.947 to 1.25 wt% 235 U in Mark IV and Mark IA fuels, respectively) U-metal clad in Zircaloy-2 (Ref. 1, Sec. 3). Both types of N Reactor SNF have been considered for disposal at the proposed Yucca Mountain site. For some WPs, the outer shell and inner shell may breach (Ref. 3) allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing two multi-canister overpacks (MCO) with either six baskets of Mark IA or five baskets of Mark IV intact N Reactor SNF rods (Ref. 1, Sec. 4) and two high-level waste (HLW) glass pour canisters (GPCs) arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which fissile uranium will remain in the WP after corrosion/dissolution of the initial WP configuration (2) The extent to which fissile uranium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this calculation, the chemical compositions (and subsequent criticality evaluations) of the simulations, is limited to

EQ6 Calculations for Chemical Degradation Of N Reactor (U-Metal) Spent Nuclear Fuel Waste Packages

Energy Technology Data Exchange (ETDEWEB)

P. Bernot

2001-02-27

The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management & Operating Contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the N Reactor, a graphite moderated reactor at the Department of Energy's (DOE) Hanford Site (ref. 1). The N Reactor core was fueled with slightly enriched (0.947 wt% and 0.947 to 1.25 wt% {sup 235}U in Mark IV and Mark IA fuels, respectively) U-metal clad in Zircaloy-2 (Ref. 1, Sec. 3). Both types of N Reactor SNF have been considered for disposal at the proposed Yucca Mountain site. For some WPs, the outer shell and inner shell may breach (Ref. 3) allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing two multi-canister overpacks (MCO) with either six baskets of Mark IA or five baskets of Mark IV intact N Reactor SNF rods (Ref. 1, Sec. 4) and two high-level waste (HLW) glass pour canisters (GPCs) arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which fissile uranium will remain in the WP after corrosion/dissolution of the initial WP configuration (2) The extent to which fissile uranium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this calculation, the chemical compositions (and subsequent criticality evaluations) of the simulations, is limited

In-core fuel management programs for nuclear power reactors

International Nuclear Information System (INIS)

1984-10-01

In response to the interest shown by Member States, the IAEA organized a co-ordinated research programme to develop and make available in the open domain a set of programs to perform in-core fuel management calculations. This report summarizes the work performed in the context of the CRP. As a result of this programme, complete in-core fuel management packages for three types of reactors, namely PWR's, BWR's and PHWR are now available from the NEA Data Bank. For some reactor types, these program packages are available with three levels of sophistication ranging from simple methods for educational purposes to more comprehensive methods that can be used for reactor design and operation. In addition some operating data have been compiled to allow code validation. (author)

Innovative Approaches to Large Component Packaging

International Nuclear Information System (INIS)

Freitag, A.; Hooper, M.; Posivak, E.; Sullivan, J.

2006-01-01

Radioactive waste disposal often times requires creative approaches in packaging design, especially for large components. Innovative design techniques are required to meet the needs for handling, transporting, and disposing of these large packages. Large components (i.e., Reactor Pressure Vessel (RPV) heads and even RPVs themselves) require special packaging for shielding and contamination control, as well as for transport and disposal. WMG Inc designed and used standard packaging for RPV heads without control rod drive mechanisms (CRDMs) attached for five RPV heads and has also more recently met an even bigger challenge and developed the innovative Intact Vessel Head Transport System (IVHTS) for RPV heads with CRDMs intact. This packaging system has been given a manufacturer's exemption by the United States Department of Transportation (USDOT) for packaging RPV heads. The IVHTS packaging has now been successfully used at two commercial nuclear power plants. Another example of innovative packaging is the large component packaging that WMG designed, fabricated, and utilized at the West Valley Demonstration Project (WVDP). In 2002, West Valley's high-level waste vitrification process was shut down in preparation for D and D of the West Valley Vitrification Facility. Three of the major components of concern within the Vitrification Facility were the Melter, the Concentrate Feed Makeup Tank (CFMT), and the Melter Feed Holdup Tank (MFHT). The removal, packaging, and disposition of these three components presented significant radiological and handling challenges for the project. WMG designed, fabricated, and installed special packaging for the transport and disposal of each of these three components, which eliminated an otherwise time intensive and costly segmentation process that WVDP was considering. Finally, WMG has also designed and fabricated special packaging for both the Connecticut Yankee (CY) and San Onofre Nuclear Generating Station (SONGS) RPVs. This paper

Revised conceptual designs for the FMDP MOX fresh fuel transport package

International Nuclear Information System (INIS)

Ludwig, S.B.; Michelhaugh, R.D.; Shappert, L.B.; Chae, S.M.; Tang, J.S.

1998-03-01

The revised conceptual designs described in this document provide a foundation for the development and certification of final transport package designs that will be needed to support the disposition of surplus weapons-grade plutonium as mixed-oxide (MOX) fuel in commercial light-water reactors in the US. This document is intended to describe the revised package design concepts and summarize the results of preliminary analyses and assessments of two new concepts for fresh MOX fuel transport packages that have been developed by Oak Ridge National Laboratory during the past year in support of the Department of Energy/Office of Fissile Materials Disposition

Conceptual designs for waste packages for horizontal or vertical emplacement in a repository in salt for reference in the site characterization plan

International Nuclear Information System (INIS)

1987-06-01

This report includes the options of horizontal and vertical emplacement, the addition of a phased repository, an additional waste form (intact spent fuel), revised geotechnical data appropriate for the Deaf Smith County site, new corrosion data for the container, and new repository design data. The waste package consists of waste form and canister within a thick-walled, low-carbon steel container surrounded by packing. The container is a hollow cylinder with a flat head welded to each end. The design concepts for the waste container or vertical and horizontal emplacement are identical. This report discusses the results of analyses of aspects of the reference waste package concept needing changes because of new data and information believed applicable to the Deaf Smith County site. Included are waste package conceptual designs or (1) the reference defense high-level waste form from the Savannah River Plant; (2) intact spent fuel with our pressurized-water-reactor or nine boiling-water-reactor assemblies per package for emplacement during Phase 1 of repository operation; and (3) spent fuel which has been disassembled and consolidated into a segmented cylindrical canister with rods from either 12 pressurized-water-reactor or 30 boiling-water-reactor assemblies per package for emplacement during Phase 2. 30 refs., 61 figs., 30 tabs

Development of heat transfer package for core thermal-hydraulic design and analysis of upgraded JRR-3

International Nuclear Information System (INIS)

Sudo, Yukio; Ikawa, Hiromasa; Kaminaga, Masanori

1985-01-01

A heat transfer package was developed for the core thermal-hydraulic design and analysis of the Japan Research Reactor-3 (JRR-3) which is to be remodeled to a 20 MWt pool-type, light water-cooled reactor with 20 % low enriched uranium (LEU) plate-type fuel. This paper presents the constitution of the developed heat transfer package and the applicability of the heat transfer correlations adopted in it, based on the heat transfer experiments in which thermal-hydraulic features of the new JRR-3 core were properly reflected. (author)

TU electric reactor model verification

International Nuclear Information System (INIS)

Willingham, C.E.; Killgore, M.R.

1989-01-01

Power reactor benchmark calculations using the code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles include gadolinia as a burnable absorber, natural uranium axial blankets, and increased water-to-fuel ratio. The calculated results for both low-power physics tests (boron end points, control rod worths, and isothermal temperature coefficients) and full-power operation (power distributions and boron letdown) are compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important physics parameters for power reactors

Structural and Shielding Safety of a Transport Package for Radioisotope Sealed Source Assembly

Energy Technology Data Exchange (ETDEWEB)

Seo, Kiseog; Cho, Ilje; Kim, Donghak [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2006-07-01

As some kinds of radioisotope (RI) sealed source are produced by HANARO research reactor, a demand of RI transport package is increasing gradually. Foreign countries, which produce the various RIs, have the intrinsic model of the RI transport package. It is necessary to develop a RI and its transport package simultaneously. It is difficult to design a shielding part for this transport package because the passage for this source assembly should be provided from the center of shielding part to the outside of the package. In order to endure the accident conditions such as a 9 m drop and puncture, this transport package consists of the guide tubes, a gamma shield and a shock absorber. This paper describe that a shielding and structural safety of RI sealed source transport package are evaluated under the accident conditions.

Structural and Shielding Safety of a Transport Package for Radioisotope Sealed Source Assembly

International Nuclear Information System (INIS)

Seo, Kiseog; Cho, Ilje; Kim, Donghak

2006-01-01

As some kinds of radioisotope (RI) sealed source are produced by HANARO research reactor, a demand of RI transport package is increasing gradually. Foreign countries, which produce the various RIs, have the intrinsic model of the RI transport package. It is necessary to develop a RI and its transport package simultaneously. It is difficult to design a shielding part for this transport package because the passage for this source assembly should be provided from the center of shielding part to the outside of the package. In order to endure the accident conditions such as a 9 m drop and puncture, this transport package consists of the guide tubes, a gamma shield and a shock absorber. This paper describe that a shielding and structural safety of RI sealed source transport package are evaluated under the accident conditions

Swing-Down of 21-PWR Waste Package

International Nuclear Information System (INIS)

A.K. Scheider

2001-01-01

The objective of this calculation is to determine the structural response of the waste package (WP) swinging down from a horizontally suspended height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 13). AP-3.12Q, ''Calculations'' (Ref. 18) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 21-PWR WP design considered in this calculation and provides the potential dimensions and materials for the 21-PWR WP design

The development of the Nuclear Electric core performance and fault transient analysis code package in support of Sizewell B

International Nuclear Information System (INIS)

Hall, P.; Hutt, P.

1994-01-01

This paper describes Nuclear Electric's (NE) development of an integrated code package in support of all its reactors including Sizewell B, designed for the provision of fuel management design, core performance studies, operational support and fault transient analysis. The package uses the NE general purpose three-dimensional transient reactor physics code PANTHER with cross-sections derived in the PWR case from the LWRWIMS LWR lattice neutronics code. The package also includes ENIGMA a generic fuel performance code and for PWR application VIPRE-01 a subchannel thermal hydraulics code, RELAP5 the system thermal hydraulics transient code and SCORPIO an on-line surveillance system. The paper describes the capabilities and validation of the elements of this package for PWR, how they are coupled within the package and the way in which they are being applied for Sizewell B to on-line surveillance and fault transient analysis. (Author)

Logistics of the research reactor fuel cycle: AREVA solutions

International Nuclear Information System (INIS)

Ohayon, David; Halle, Laurent; Naigeon, Philippe; Falgoux, Jean-Louis; Franck Obadia, Franck; Auziere, Philippe

2005-01-01

The AREVA Group Companies offer comprehensive solutions for the entire fuel cycle of Research Reactors comply with IAEA standards. CERCA and Cogema Logistics have developed a full partnership in the front end cycle. In the field of uranium CERCA and Cogema Logistics have the long term experience of the shipment from Russia, USA to the CERCA plant.. Since 1960, CERCA has manufactured over 300,000 fuel plates and 15,000 fuel elements of more than 70 designs. These fuel elements have been delivered to 40 research reactors in 20 countries. For the Back-End stage, Cogema and Cogema Logistics propose customised solutions and services for international shipments. Cogema Logistics has developed a new generation of packaging to meet the various needs and requirements of the Laboratories and Research Reactors all over the world, and complex regulatory framework. Comprehensive assistance dedicated, services, technical studies, packaging and transport systems are provided by AREVA for every step of research reactor fuel cycle. (author)

10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

Science.gov (United States)

2010-01-01

... nonpower reactors. 73.60 Section 73.60 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION... requirements for physical protection at nonpower reactors. Each nonpower reactor licensee who, pursuant to the... perform their duties. (6) Prior to entry into a material access area, packages shall be searched for...

Cleanup Verification Package for the 118-F-7, 100-F Miscellaneous Hardware Storage Vault

International Nuclear Information System (INIS)

Appel, M.J.

2006-01-01

This cleanup verification package documents completion of remedial action for the 118-F-7, 100-F Miscellaneous Hardware Storage Vault. The site consisted of an inactive solid waste storage vault used for temporary storage of slightly contaminated reactor parts that could be recovered and reused for the 100-F Area reactor operations

Cleanup Verification Package for the 118-F-7, 100-F Miscellaneous Hardware Storage Vault

Energy Technology Data Exchange (ETDEWEB)

M. J. Appel

2006-11-02

This cleanup verification package documents completion of remedial action for the 118-F-7, 100-F Miscellaneous Hardware Storage Vault. The site consisted of an inactive solid waste storage vault used for temporary storage of slightly contaminated reactor parts that could be recovered and reused for the 100-F Area reactor operations.

21-PWR Waste Package Side and End Impacts

International Nuclear Information System (INIS)

V. Delabrosse

2003-01-01

The objective of this calculation is to determine the structural response of a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities and initial angles between the waste package and the unyielding surface is studied. The scope of this calculation is limited to estimating the area of the outer shell (OS) where the residual stress exceeds a given limit (hereafter ''damaged area''). The stress limit is defined as a fraction of the yield strength of the OS material, Alloy 22 (SB-575 N06022), at the appropriate temperature. The design of the 21-PWR waste package used in this calculation is that defined in Reference 8. However, a value of 4 mm was used for the gap between the inner shell and the OS, and the thickness of the OS was reduced by 2 mm. The sketch in Attachment I provides additional information not included in Reference 8. All obtained results are valid for this design only. This calculation is associated with the waste package design and was performed by the Specialty Analyses and Waste Package Design Section. The waste package (i.e. uncanistered spent nuclear fuel disposal container) is classified as Quality Level 1

21-PWR Waste Package Side and End Impacts

International Nuclear Information System (INIS)

T. Schmitt

2005-01-01

The objective of this calculation is to determine the structural response of a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities and initial angles between the waste package and the unyielding surface is studied. The scope of this calculation is limited to estimating the area of the outer shell (OS) where the residual stress exceeds a given limit (hereafter ''damaged area''). The stress limit is defined as a fraction of the yield strength of the OS material, Alloy 22 (SB-575 N06022), at the appropriate temperature. The design of the 21-PWR waste package used in this calculation is that defined in Reference 8. However, a value of 4 mm was used for the gap between the inner shell and the OS, and the thickness of the OS was reduced by 2 mm. The sketch in Attachment I provides additional information not included in Reference 8. All obtained results are valid for this design only. This calculation is associated with the waste package design and was performed by the Specialty Analyses and Waste Package Design Section. The waste package (i.e. uncanistered spent nuclear fuel disposal container) is classified as Quality Level 1

Safety Analysis Report for Packaging: The unirradiated fuel shipping container USA/9853/AF

International Nuclear Information System (INIS)

1991-01-01

The HFBR Unirradiated Fuel Shipping Container was designed and fabricated at the Oak Ridge National Laboratory in 1978 for the transport of fuel for the High Flux Beam Reactor (HFBR) for Brookhaven National Laboratory. The package has been evaluated analytically, as well as the comparison to tests on similar packages, to demonstrate compliance with the applicable regulations governing packages in which radioactive and fissile materials are transported. The contents of this Safety Analysis Report for Packaging (SARP) are based on Regulatory Guide 7.9 (proposed Revision 2 - May 1986), 10 CFR Part 71, DOE Order 1540.2, DOE Order 5480.3, and 49 CFR Part 173

Anticipated development in radioactive materials packaging and transport systems

International Nuclear Information System (INIS)

Williams, L.D.; Rhoads, R.E.; Hall, R.J.

1976-07-01

Closing the light water reactor fuel cycle and the use of mixed oxide fuels will produce materials such as solidified high level waste, cladding hulls and plutonium from Pu recycle fuel that have not been transported extensively in the past. Changes in allowable gaseous emissions from fuel cycle facilities may require the collection and transportation of radioactive noble gases and tritium. Although all of these materials could be transported in existing radioactive material packaging, economic considerations will make it desirable to develop new packaging specifically designed for each material. Conceptual package designs for these materials are reviewed. Special Nuclear Material transportation safeguards are expected to have a significant impact on future fuel cycle transportation. This subject is reviewed briefly. Other factors that could affect fuel cycle transportation are also discussed. Development of new packaging for radioactive materials is not believed to require the development of new technologies. New package designs will be primarily an adaptation of existing technology to fit the changing needs of a growing nuclear power industry. 23 references

New developments in transportation for research reactors

Energy Technology Data Exchange (ETDEWEB)

Mondanel, J.L. [Transnucleaire, F-75008 Paris (France)

1998-07-01

For more than 30 years, Transnucleaire has been performing safely a large number of national and international transports of radioactive material. Transnucleaire has also designed and supplied numerous packagings for all types of nuclear fuel cycle radioactive materials: for front-end and back-end products and for power and research reactors. Since the last meeting held in Bruges, Transnucleaire has been continuously involved in transportation activities for fresh and irradiated materials for research reactors. We are pleased to take the opportunity in this meeting to share with reactor operators, official bodies and other partners, the on-going developments in transportation and associated services. Special attention will be paid to the starting of transports of MTR spent fuel elements to the La Hague reprocessing plant where COGEMA offers reprocessing services on a long-term basis to reactors operators. Detailed information is provided on regulatory issues, which may affect transport activities: evolution of the regulations, real experiences of recent transportation and development of new packaging designs. Options and solutions will be proposed by Transnucleaire to improve the situation for continuation of national and international transports at an acceptable price whilst maintaining an ultimate level of safety (author)

Cleanup Verification Package for the 118-C-1, 105-C Solid Waste Burial Ground

Energy Technology Data Exchange (ETDEWEB)

M. J. Appel and J. M. Capron

2007-07-25

This cleanup verification package documents completion of remedial action for the 118-C-1, 105-C Solid Waste Burial Ground. This waste site was the primary burial ground for general wastes from the operation of the 105-C Reactor and received process tubes, aluminum fuel spacers, control rods, reactor hardware, spent nuclear fuel and soft wastes.

Cleanup Verification Package for the 118-C-1, 105-C Solid Waste Burial Ground

International Nuclear Information System (INIS)

Appel, M.J.; Capron, J.M.

2007-01-01

This cleanup verification package documents completion of remedial action for the 118-C-1, 105-C Solid Waste Burial Ground. This waste site was the primary burial ground for general wastes from the operation of the 105-C Reactor and received process tubes, aluminum fuel spacers, control rods, reactor hardware, spent nuclear fuel and soft wastes

Development of the processing software package for RPV neutron fluence determination methodology

International Nuclear Information System (INIS)

Belousov, S.; Kirilova, K.; Ilieva, K.

2001-01-01

According to the INRNE methodology the neutron transport calculation is carried out by two steps. At the first step reactor core eigenvalue calculation is performed. This calculation is used for determination of the fixed source for the next step calculation of neutron transport from the reactor core to the RPV. Both calculation steps are performed by state of the art and tested codes. The interface software package DOSRC developed at INRNE is used as a link between these two calculations. The package transforms reactor core calculation results to neutron source input data in format appropriate for the neutron transport codes (DORT, TORT and ASYNT) based on the discrete ordinates method. These codes are applied for calculation of the RPV neutron flux and its responses - induced activity, radiation damage, neutron fluence etc. Fore more precise estimation of the neutron fluence, the INRNE methodology has been supplemented by the next improvements: - implementation of more advanced codes (PYTHIA/DERAB) for neutron-physics parameter calculations; - more detailed neutron source presentation; - verification of neutron fluence by statistically treated experimental data. (author)

FFTF and CRBRP reactor vessels

International Nuclear Information System (INIS)

Morgan, R.E.

1977-01-01

The Fast Flux Test Facility (FFTF) reactor vessel and the Clinch River Breeder Reactor Plant (CRBRP) reactor vessel each serve to enclose a fast spectrum reactor core, contain the sodium coolant, and provide support and positioning for the closure head and internal structure. Each vessel is located in its reactor cavity and is protected by a guard vessel which would ensure continued decay heat removal capability should a major system leak develop. Although the two plants have significantly different thermal power ratings, 400 megawatts for FFTF and 975 megawatts for CRBRP, the two reactor vessels are comparable in size, the CRBRP vessel being approximately 28% longer than the FFTF vessel. The FFTF vessel diameter was controlled by the space required for the three individual In-Vessel Handling Machines and Instrument Trees. Utilization of the triple rotating plug scheme for CRBRP refueling enables packaging of the larger CRBRP core in a vessel the same diameter as the FFTF vessel

Benchmarking of the computer code and the thirty foot side drop analysis for the Shippingport (RPV/NST package)

International Nuclear Information System (INIS)

Bumpus, S.E.; Gerhard, M.A.; Hovingh, J.; Trummer, D.J.; Witte, M.C.

1989-01-01

This paper presents the benchmarking of a finite element computer code and the subsequent results from the code simulating the 30 foot side drop impact of the RPV/NST transport package from the decommissioned Shippingport Nuclear Power Station. The activated reactor pressure vessel (RPV), thermal shield, and other reactor external components were encased in concrete contained by the neutron shield tank (NST) and a lifting skirt. The Shippingport RPV/NST package, a Type B Category II package, weighs approximately 900 tons and has 17.5 ft diameter and 40.7 ft. length. For transport of the activated components from Shippingport to the burial site, the Safety Analysis Report for Packaging (SARP) demonstrated that the package can withstand the hypothetical accidents of DOE Order 5480.3 including 10 CFR 71. Mathematical simulations of these accidents can substitute for actual tests if the simulated results satisfy the acceptance criteria. Any such mathematical simulation, including the modeling of the materials, must be benchmarked to experiments that duplicate the loading conditions of the tests. Additional confidence in the simulations is justified if the test specimens are configured similar to the package

Reactor neutron dosimetry

International Nuclear Information System (INIS)

Najzer, M.; Pauko, M.; Glumac, B.; Acquah, I.N.; Moskon, F.

1977-01-01

An analysis of requirements and possibilities for experimental neutron spectrum determination during the reactor pressure vessel surveil lance programme is given. Fast neutron spectrum and neutron dose rate were measured in the Fast neutron irradiation facility of our TRIGA reactor. It was shown that the facility can be used for calibration of neutron dosimeters and for irradiation of samples sensitive to neutron radiation. The investigation of the unfolding algorithm ITER was continued. Based on this investigations are two specialized unfolding program packages ITERAD and ITERGS written this year. They are able to unfold data from activation detectors and NaI(T1) gamma spectrometer respectively

An approach to neutronics analysis of candu reactors

International Nuclear Information System (INIS)

Gul, S.; Arshad, M.

1982-12-01

An attempt is made to tackle the problem of neutronics analysis of CANDU reactors. Until now CANDU reactors have been analysed by the methods developed at AECL and CGE using mainly receipe methods. Relying on multigroup transport codes GAM-GATHER in combination with diffusion code CITATION a package of codes is established to use it for survey as well as production purposes. (authors)

Development of an automated core model for nuclear reactors

International Nuclear Information System (INIS)

Mosteller, R.D.

1998-01-01

This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The objective of this project was to develop an automated package of computer codes that can model the steady-state behavior of nuclear-reactor cores of various designs. As an added benefit, data produced for steady-state analysis also can be used as input to the TRAC transient-analysis code for subsequent safety analysis of the reactor at any point in its operating lifetime. The basic capability to perform steady-state reactor-core analysis already existed in the combination of the HELIOS lattice-physics code and the NESTLE advanced nodal code. In this project, the automated package was completed by (1) obtaining cross-section libraries for HELIOS, (2) validating HELIOS by comparing its predictions to results from critical experiments and from the MCNP Monte Carlo code, (3) validating NESTLE by comparing its predictions to results from numerical benchmarks and to measured data from operating reactors, and (4) developing a linkage code to transform HELIOS output into NESTLE input

Cleanup Verification Package for the 118-B-1, 105-B Solid Waste Burial Ground

International Nuclear Information System (INIS)

Capron, J.M.

2008-01-01

This cleanup verification package documents completion of remedial action, sampling activities, and compliance criteria for the 118-B-1, 105-B Solid Waste Burial Ground. This waste site was the primary burial ground for general wastes from the operation of the 105-B Reactor and P-10 Tritium Separation Project and also received waste from the 105-N Reactor. The burial ground received reactor hardware, process piping and tubing, fuel spacers, glassware, electrical components, tritium process wastes, soft wastes and other miscellaneous debris

A computer code for Tokamak reactor concepts evaluation

International Nuclear Information System (INIS)

Rosatelli, F.; Raia, G.

1985-01-01

A computer package has been developed which could preliminarily investigate the engineering configuration of a tokamak reactor concept. The code is essentially intended to synthesize, starting from a set of geometrical and plasma physics parameters and the required performances and objectives, three fundamental components of a tokamak reactor core: blanket+shield, TF magnet, PF magnet. An iterative evaluation of the size, power supply and cooling system requirements of these components allows the judgment and the preliminary design optimization on the considered reactor concept. The versatility of the code allows its application both to next generation tokamak devices and power reactor concepts

Cleanup Verification Package for the 118-F-3, Minor Construction Burial Ground

International Nuclear Information System (INIS)

Appel, M.J.

2007-01-01

This cleanup verification package documents completion of remedial action for the 118-F-3, Minor Construction Burial Ground waste site. This site was an open field covered with cobbles, with no vegetation growing on the surface. The site received irradiated reactor parts that were removed during conversion of the 105-F Reactor from the Liquid 3X to the Ball 3X Project safety systems and received mostly vertical safety rod thimbles and step plugs

Nuclear Reactor RA Safety Report, Vol. 8, Auxiliary system

International Nuclear Information System (INIS)

1986-11-01

This volume describes RA reactor auxiliary systems, as follows: special ventilation system, special drainage system, hot cells, systems for internal transport. Ventilation system is considered as part of the reactor safety and protection system. Its role is eliminate possible radioactive particles dispersion in the environment. Special drainage system includes pipes and reservoirs with the safety role, meaning absorption or storage of possible radioactive waste water from the reactor building. Hot cells existing in the RA reactor building are designed for production of sealed radioactive sources, including packaging and transport [sr

Spent fuel receipt and lag storage facility for the spent fuel handling and packaging program

International Nuclear Information System (INIS)

Black, J.E.; King, F.D.

1979-01-01

Savannah River Laboratory (SRL) is participating in the Spent Fuel Handling and Packaging Program for retrievable, near-surface storage of spent light water reactor (LWR) fuel. One of SRL's responsibilities is to provide a technical description of the wet fuel receipt and lag storage part of the Spent Fuel Handling and Packaging (SFHP) facility. This document is the required technical description

The Siemens-Argonaut reactor as a driver zone for a high-temperature reactor cell. Der Siemens-Argonaut-Reaktor als Treiberzone fuer eine Hochtemperaturreaktorzelle

Energy Technology Data Exchange (ETDEWEB)

Mueller, H; Schuerrer, F; Ninaus, W; Oswald, K; Rabitsch, H; Kreiner, H [Technische Univ., Graz (Austria). Inst. fuer Theoretische Physik und Reaktorphysik; Neef, R D [Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Reaktorentwicklung

1984-12-15

To enable a validation of neutron physics calculation methods for pebble bed reactors the inner reflector of an Argonaut research reactor was substituted by a full of about 1200 fuel elements of the AVR-Juelich type. The report describes the measuring instruments and the reactor physical layout of the arrangement by the code packages GAMTEREX, ZUT-D.G.L. and MUPO. Comparison of calculated reaction rates with measurements show good agreement. Application of the codes to high-temperature reactors in abnormal states is envisaged. (Author, translated by G.Q.)

Cleanup Verification Package for the 116-K-2 Effluent Trench

International Nuclear Information System (INIS)

Capron, J.M.

2006-01-01

This cleanup verification package documents completion of remedial action for the 116-K-2 effluent trench, also referred to as the 116-K-2 mile-long trench and the 116-K-2 site. During its period of operation, the 116-K-2 site was used to dispose of cooling water effluent from the 105-KE and 105-KW Reactors by percolation into the soil. This site also received mixed liquid wastes from the 105-KW and 105-KE fuel storage basins, reactor floor drains, and miscellaneous decontamination activities

The Reactor Analysis Support Package (RASP): Volume 10, Guidelines for developing a reload licensing capability:Final report

International Nuclear Information System (INIS)

Diamond, D.J.

1988-08-01

The EPRI Reactor Analysis Support Package (RASP) consists of computer codes and documentation for calculating core performance and plant transients. This report was written to help a utility to use these tools properly for reload design and safety evaluations. The emphasis is on the steps that the utility can take to develop a methodology that is approved by the Nuclear Regulatory Commission for reload licensing submittals. The report treats both the planning and the implementation of this type of calculational capability. With regard to planning, there is discussion defining objectives, resource requirements, organization and scheduling. In order to help the engineering staff implement the plan there is discussion of the development of a methodology for event analysis, qualification of the methods, and the writing of design control procedures and topical reports. The experience of utilities, and especially of GPU Nuclear Corporation (GPUN), in developing a reload licensing capability is cited throughout the report and extracts from GPUN design control procedures are included in the appendices. 16 refs., 23 figs., 9 tabs

Characterizing and packaging BN-350 spent fuel for long-term dry storage

International Nuclear Information System (INIS)

Lambert, J. D. B.; Bolshinsky, I.; Haues, S.L.; Allen, K.J.; Howden, E.A.; Hill, R.N.; Planchon, H.P.; Staples, P.; Karaulov, V.N.; Blynskij, A.P.; Yakovlev, I.K.; Maev, V.; Dumchev, I. A.

2000-01-01

The Republic of Kazakhstan is being assisted by the U.S. Department of Energy in preparing spent fuel from the BN-350 fast reactor for long term dry storage. Argonne National Laboratory was assigned responsibility for the physical and nuclear characterization of the spent fuel, for the design and safety analysis of 6-pac and 4-pac canisters used to contain spent fuel assemblies for storage, and for the design, testing and installation of a closure station at the reactor in which the canisters of fuel are dried, filled with inert gas and welded shut. This paper briefly describes the specialized components and equipment used, the process followed, and experience gained in packaging the spent fuel. Olsen et al and Schaefer separately discuss overall safety and criticality considerations of the packaging process in parallel papers to this conference

Project Management Plan 105-KE Basin sludge retrieval and packaging

International Nuclear Information System (INIS)

McWethy, L.M.

1994-01-01

The KE Basin contains over 1,100 metric tons of spent nuclear fuel (SNF). The bulk of this inventory consists of over 50,000 zircaloy clad, uranium metal N-Reactor fuel element assemblies, along with less than half a metric ton of single-pass reactor fuel elements, stored in over 3,600 open top canister assemblies. In addition, sludge containing fissile and fission product material from damaged/degraded fuel has accumulated in the basin. The sludge, particularly the fines, impacts basin operations by clouding the water and making activities requiring a clear view impossible to complete until after sludge settles. Packaging would get the sludge out of the operator's way and allow it to be moved within the basin in a more manageable state. The primary project objective is to develop, procure, and quality the equipment needed to remove all sludge from the KE Basin with minimal dose commitment, minimal cost, and on schedule. The project will provide: (1) the development, testing, and installation of equipment for sludge retrieval and packaging; (2) understanding of and experience with actual sludge through near-term sludge packaging feature tests in the KE Basin; (3) sludge removal and handling equipment required to support debris removal, fuel handling, and other activities involving sludge within the KE Basin; and (4) enlist industry expertise in all phases of the project. This Project Management Plant establishes the organizational responsibilities, control systems, and procedures for the execution of project activities for KE Basin sludge retrieval packaging, to meet programmatic requirements within authorized funding and approved schedules

Effect of Core Configurations on Burn-Up Calculations For MTR Type Reactors

International Nuclear Information System (INIS)

Hussein, H.M.; Sakr, A.M.; Amin, E.H.

2011-01-01

Three-dimensional burn-up calculations of MTR-type research reactor were performed using different patterns of control rods , to examine their effect on power density and neutron flux distributions throughout the entire core and on the local burn-up distribution. Calculations were performed using the computer codes' package M TR P C system , using the cell calculation transport code WIMS-D4 and the core calculation diffusion code CITVAP. A depletion study was done and the effects on the reactor fuel were studied, then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Keywords: Neutronic Calculations, Burn-Up, MTR-Type Research Reactors, MTR P C Package, Empirical Formula For Fuel Burn-Up.

HORECA. Hoger onderwijs reactor elementary core analysis system. User's manual

International Nuclear Information System (INIS)

Battum, E. van; Serov, I.V.

1993-07-01

HORECA is developed at IRI Delft for quick analysis of power distribution, burnup and safety for the HOR. It can be used for the manual search of a better loading of the reactor. HORECA is based on the Penn State Fuel Management Package and uses the MCRAC code included in this package as a calculation engine. (orig./HP)

Packaging configurations and handling requirements for nuclear materials

International Nuclear Information System (INIS)

Jefferson, R.M.

1981-01-01

The basic safety concepts for radioactive material are that the package is the primary protection for the public, that the protection afforded by the package should be proportional to the hazard and that the package must be proved by performance. These principles are contained in Department of Energy (DOE), Nuclear Regulatory Commission (NRC) and Department of Transportation (DOT) regulations which classify hazards of various radioactive materials and link packaging requirements to the physical form and quantities being shipped. Packaging requirements are reflected in performance standards to guarantee that shipments of low hazard quantities will survive the rigors of normal transportation and that shipments of high hazard quantities will survive extreme severity transportation accidents. Administrative controls provide for segregation of radioactive material from people and other sensitive or hazardous material. They also provide the necessary information function to control the total amounts in a conveyance and to assure that appropriate emergency response activities be started in case of accidents or other emergencies. Radioactive materials shipped in conjunction with the nuclear reactor programs include, ores, concentrates, gaseous diffusion feedstocks, enriched and depleted uranium, fresh fuel, spent fuel, high level wastes, low level wastes and transuranic wastes. Each material is packaged and shipped in accordance with regulations and all hazard classes, quantity limits and packaging types are called into use. From the minimal requirements needed to ship the low hazard uranium ores or concentrates to the very stringent requirements in packaging and moving high level wastes or spent fuel, the regulatory system provides a means for carrying out transportation of radioactive material which assures low and controlled risk to the public

Evaluating and planning the radioactive waste options for dismantling the Tokamak Fusion Test Reactor

Energy Technology Data Exchange (ETDEWEB)

Rule, K.; Scott, J.; Larson, S. [Princeton Plasma Physics Lab., NJ (United States)] [and others

1995-12-31

The Tokamak Fusion Test Reactor (TFTR) is a one-of-a kind tritium fusion research reactor, and is planned to be decommissioned within the next several years. This is the largest fusion reactor in the world and as a result of deuterium-tritum reactions is tritium contaminated and activated from 14 Mev neutrons. This presents many unusual challenges when dismantling, packaging and disposing its components and ancillary systems. Special containers are being designed to accommodate the vacuum vessel, neutral beams, and tritium delivery and processing systems. A team of experienced professionals performed a detailed field study to evaluate the requirements and appropriate methods for packaging the radioactive materials. This team focused on several current and innovative methods for waste minimization that provides the oppurtunmost cost effective manner to package and dispose of the waste. This study also produces a functional time-phased schedule which conjoins the waste volume, weight, costs and container requirements with the detailed project activity schedule for the entire project scope. This study and project will be the first demonstration of the decommissioning of a tritium fusion test reactor. The radioactive waste disposal aspects of this project are instrumental in demonstrating the viability of a fusion power reactor with regard to its environmental impact and ultimate success.

Benchmark shielding calculations for the NEACRP [Nuclear Energy Agency-Committee on Reactor Physics] Working Group on shielding assessment of transportation packages

International Nuclear Information System (INIS)

Broadhead, B.L.; Brady, M.C.; Parks, C.V.

1990-11-01

In 1985, the Nuclear Energy Agency-Committee on Reactor Physics (NEACRP) established a working group on shielding assessment of transportation packages. Following the initial distribution of a set of six problems, discussions were held at the Organization for Economic Cooperation and Development (OECD) Headquarters in Paris, France, in June/July 1986, May 1988, and February/March 1990. The US contribution to the working group is documented in this report. The results from this effort permit the evaluation of a number of approximations and effects that must be considered in a typical shielding analysis of a transportation cask. Among the effects reported here are the performance of multiple cross-section sets, the comparison of several source generation codes, and multidimensional versus one-dimensional (1-D) analyses. 18 refs., 16 figs., 33 tabs

NURESIM lecture on reactor physics (visual aids)

International Nuclear Information System (INIS)

Nguyen Tien Nguyen

1998-01-01

The purpose of the NURESIM software (NUclear REactor SIMulation) is to be used as a computer guide in quick view of the texts and pictures in the fields of nuclear reactor physics. This software is designed so that it can be used by users of different knowledge levels. Students could find here elementary concepts, researchers - important calculation codes as GRACE, PEACO, THERMOS, HEX120. The NURESIM software is composed of four parts: units, pictures, simulations and calculations. In the terminology of IAEA-TECDOC-314 (1984) the first three parts may be classified as a level 2 of sophistication IFM code package: ''Code package useful as a first introduction for nuclear engineers''. The last one (calculations) is classified as a level higher. Details about each part are explained in Paragraph 2. A users guide is in Paragraph 3. (author)

Running the source term code package in Elebra MX-850

International Nuclear Information System (INIS)

Guimaraes, A.C.F.; Goes, A.G.A.

1988-01-01

The source term package (STCP) is one of the main tools applied in calculations of behavior of fission products from nuclear power plants. It is a set of computer codes to assist the calculations of the radioactive materials leaving from the metallic containment of power reactors to the environment during a severe reactor accident. The original version of STCP runs in SDC computer systems, but as it has been written in FORTRAN 77, is possible run it in others systems such as IBM, Burroughs, Elebra, etc. The Elebra MX-8500 version of STCP contains 5 codes:March 3, Trapmelt, Tcca, Vanessa and Nava. The example presented in this report has taken into consideration a small LOCA accident into a PWR type reactor. (M.I.)

Evaluation of nuclear reactor based activation analysis techniques

International Nuclear Information System (INIS)

Obrusnik, I.; Kucera, J.

1977-09-01

A survey is presented of the basic types of activation analysis applied in environmental control. Reactor neutron activation analysis is described (including the reactor as a neutron source, sample activation in the reactor, methodology of neutron activation analysis, sample transport into the reactor and sample packaging after irradiation, instrumental activation analysis with radiochemical separation, data measurement and evaluation, sampling and sample preparation). Sources of environmental contamination with trace elements, sampling and sample analysis by neutron activation are described. The analysis is described of soils, waters and biological materials. Methods are shown of evaluating neutron activation analysis results and of their interpretation for purposes of environmental control. (J.B.)

MVPACK: a package for the computer-aided design of multivariable control systems

International Nuclear Information System (INIS)

Mensah, S.

1984-01-01

The design and analysis of high performing controllers for large complex plants require a collection of interactive, powerful computer software. MVPACK, an open-ended package for the computer aided design of control systems has been developed in the Reactor Control Branch of the Chalk River Nuclear Laboratories. The package is fully interactive, and includes a comprehensive state-of-the-art mathematical library to support development of complex multivariable control algorithms. Coded in RATFOR, MVPACK operates with a flexible data structure which makes efficient use of minicomputer resources and provides a standard framework for program generation. The existence of a help mechanism enhances the simplicity of package utilization. This report provides the technical description of the package. It reviews the specifications used in the design and implementation of the package. The database structure, the supporting libraries and the design and analysis modules of MVPACK are described. The report includes several application examples to illustrate the capability of the package. Experience with MVPACK shows that the package provides a synergistic environment for control and regulation systems design, and that it is a unique tool in training of control system engineers

Waste package reference conceptual designs for a repository in salt

International Nuclear Information System (INIS)

1986-02-01

This report provides the reference conceptual waste package designs for the Office of Nuclear Waste Isolation to baseline these designs, thereby establishing the configuration and interface controls necessary, within the Civilian Radioactive Waste Management Program, formerly the National Waste Terminal Storage Program, to proceed in an orderly manner with preliminary design. Included are designs for the current reference defense high-level waste form from the Savannah River Plant, an optimized commercial high-level waste form, and spent fuel which has been disassembled and compacted into a circular bundle containing either 12 pressurized-water reactor or 30 boiling-water reactor assemblies. For compacted spent fuel, it appears economically attractive to standardize the waste package diameter for all fuel types. The reference waste packages consist of the containerized waste form, a low carbon steel overpack, and, after emplacement, a cover of salt. The overpack is a hollow cylinder with a flat head welded to each end. Its design thickness is the sum of the structural thickness required to resist the 15.4-MPa lithostatic pressure plus the corrosion allowance necessary to assure the required structural thickness will exist through the 1000-year containment period. Based on available data and completed analyses, the reference concepts described in this report satisfy all requirements of the US Department of Energy and the US Nuclear Regulatory Commission with reasonable assurance. In addition, sufficient design maturity exists to form a basis for preliminary design; these concepts can be brought under configuration control to serve as reference package designs. Development programs are identified that will be required to support these designs during the licensing process. 19 refs., 37 figs., 31 tabs

Computerized nuclear material database management system for power reactors

International Nuclear Information System (INIS)

Cheng Binghao; Zhu Rongbao; Liu Daming; Cao Bin; Liu Ling; Tan Yajun; Jiang Jincai

1994-01-01

The software packages for nuclear material database management for power reactors are described. The database structure, data flow and model for management of the database are analysed. Also mentioned are the main functions and characterizations of the software packages, which are successfully installed and used at both the Daya Bay Nuclear Power Plant and the Qinshan Nuclear Power Plant for the purposed of handling nuclear material database automatically

JANUS reactor d and d project

International Nuclear Information System (INIS)

Fellhauer, C. R.

1998-01-01

Argonne National Laboratory (ANL-E) has recently completed the decontamination and decommissioning (D and D) of the JANUS Reactor Facility located in Building 202. The 200 KW reactor operated from August 1963 to March 1992. The facility was used to study the effects of both high and low doses of fission neutrons in animals. There were two exposure rooms on opposite sides of the reactor and the reactor was therefore named after the two-faced Roman god. The High Dose Room was capable of specimen exposure at a dose rate of 3,600 rads per hour. During calendar year 1996 a detailed characterization of the facility was performed by ANL-E Health Physics personnel. ANL-E Analytical Services performed the required sample analysis. An Auditable Safety Analysis and an Environmental Assessment were completed. D and D plans, procedures and procurement documents were prepared and approved. A D and D subcontractor was selected and a firm, fixed price contract awarded for the field work and final survey effort. The D and D subcontractor was mobilized to ANL-E in January 1997. Electrical isolation of all reactor equipment and control panels was accomplished and the equipment removed. A total of 207,230 pounds (94,082 Kg) of lead shielding was removed, surveyed and sampled, and free-released for recycle. All primary and secondary piping was removed, size reduced and packaged for disposal or recycled as appropriate. The reactor vessel was removed, sized reduced and packaged as radioactive waste in April. The activated graphite block reflector was removed next, followed by the bioshield concrete and steel. All of this material was packaged as low level waste. Total low level radioactive waste generation was 4002.1 cubic feet (113.3 cubic meters). Mixed waste generation was 538 cubic feet (15.2 cubic meters). The Final Release Survey was completed in September. The project field work was completed in 38 weeks without any lost-time accidents, personnel contaminations or unplanned

Packaging design criteria for the MCO cask

International Nuclear Information System (INIS)

Edwards, W.S.

1996-01-01

Approximately 2,100 metric tons of unprocessed, irradiated nuclear fuel elements are presently stored in the K Basins (including possibly 700 additional elements from PUREX, N Reactor, and 327 Laboratory). The basin water, particularly in the K East Basin, contains significant quantities of dissolved nuclear isotopes and radioactive fuel corrosion particles. To permit cleanup of the K Basins and fuel conditioning, the fuel will be transported from the 100 K Area to a Canister Storage Building (CSB) in the 200 East area. In order to initiate K Basin cleanup on schedule, the two-year fuel-shipping campaign must begin by December 1997. The purpose of this packaging design criteria is to provide criteria for the design, fabrication, and use of a packaging system to transport the large quantities of irradiated nuclear fuel elements positioned within Multiple Canister Overpacks

Accident analysis for new reactor concepts and VVER type reactor design with advanced fuel. STC with Russia. Final report

International Nuclear Information System (INIS)

Grundmann, U.; Kliem, S.; Mittag, S.; Rohde, U.; Seidel, A.

2000-10-01

In the frame of a project on scientific-technical cooperation funded by BMBF/BMWi, the 3D reactor dynamics code DYN3D developed at Forschungszentrum Rossendorf (FZR), has been transferred to the Institute of Physics and Power Engineering (IPPE) Obninsk in Russia and integrated into the software package of IPPE. DYN3D has been coupled to a thermohydraulic system code used in IPPE making available 3D neutron kinetics within this software package. A new macroscopic cross section library has been created using a modified version of the WIMS/D4 code. This library includes data for modernized fuel design containing burnable absorbers in different concentrations, which is tested in VVER-1000 type reactors. The cross section library has been connected to DYN3D. Calculations were performed to check the library in comparison with other data libraries and codes. The code DYN3D and the coupled 3D neutron kinetics/thermal hydraulics code system were used to perform analyses of Anticipated Transients Without Scram (ATWS) for the reactor design ABV-67, an integral reactor concept with small power developed under participation of IPPE. The fluid dynamics code DINCOR developed at IPPE was transferred to FZR. It was used in validation calculations on test problems for the short-term core melt behaviour (CORVIS experiments). (orig.) [de

Qualification of the coupled RELAP5/PANTHER/COBRA code package for licensing applications

International Nuclear Information System (INIS)

Schneidesch, C.R.; Zhang Jinzhao

2004-01-01

A coupled thermal hydraulics-neutronics code package has been developed at Tractebel Engineering (TE), in which the best-estimate thermal-hydraulic system code, RELAP5/mod2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via the dynamic data exchange interface, TALINK. The Departure from Nucleate Boiling Ratio (DNBR) is calculated by the sub-channel thermal-hydraulic analysis code COBRA-3C. The package provides the capability to accurately simulate the key physical phenomena in nuclear power plant accidents with strong asymmetric behaviours and system-core interactions. This paper presents the TE coupled code package and focuses on the methodology followed for qualifying it for licensing applications. The qualification of the coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been qualified and will be used at Tractebel Engineering (TE) for analyzing asymmetric PWR accidents with strong core-system interactions

Extension of ship accident analysis to multiple-package shipments

International Nuclear Information System (INIS)

Mills, G.S.; Neuhauser, K.S.

1997-11-01

Severe ship accidents and the probability of radioactive material release from spent reactor fuel casks were investigated previously. Other forms of RAM, e.g., plutonium oxide powder, may be shipped in large numbers of packagings rather than in one to a few casks. These smaller, more numerous packagings are typically placed in ISO containers for ease of handling, and several ISO containers may be placed in one of several holds of a cargo ship. In such cases, the size of a radioactive release resulting from a severe collision with another ship is determined not by the likelihood of compromising a single, robust package but by the probability that a certain fraction of 10's or 100's of individual packagings is compromised. The previous analysis involved a statistical estimation of the frequency of accidents which would result in damage to a cask located in one of seven cargo holds in a collision with another ship. The results were obtained in the form of probabilities (frequencies) of accidents of increasing severity and of release fractions for each level of severity. This paper describes an extension of the same general method in which the multiple packages are assumed to be compacted by an intruding ship's bow until there is no free space in the hold. At such a point, the remaining energy of the colliding ship is assumed to be dissipated by progressively crushing the RAM packagings and the probability of a particular fraction of package failures is estimated by adaptation of the statistical method used previously. The parameters of a common, well characterized packaging, the 6M with 2R inner containment vessel, were employed as an illustrative example of this analysis method. However, the method is readily applicable to other packagings for which crush strengths have been measured or can be estimated with satisfactory confidence

Extension of ship accident analysis to multiple-package shipments

International Nuclear Information System (INIS)

Mills, G.S.; Neuhauser, K.S.

1998-01-01

Severe ship accidents and the probability of radioactive material release from spent reactor fuel casks were investigated previously (Spring, 1995). Other forms of RAM, e.g., plutonium oxide powder, may be shipped in large numbers of packagings rather than in one to a few casks. These smaller, more numerous packagings are typically placed in ISO containers for ease of handling, and several ISO containers may be placed in one of several holds of a cargo ship. In such cases, the size of a radioactive release resulting from a severe collision with another ship is determined not by the likelihood of compromising a single, robust package but by the probability that a certain fraction of 10's or 100's of individual packagings is compromised. The previous analysis (Spring, 1995) involved a statistical estimation of the frequency of accidents which would result in damage to a cask located in one of seven cargo holds in a collision with another ship. The results were obtained in the form of probabilities (frequencies) of accidents of increasing severity and of release fractions for each level of severity. This paper describes an extension of the same general method in which the multiple packages are assumed to be compacted by an intruding ship's bow until there is no free space in the hold. At such a point, the remaining energy of the colliding ship is assumed to be dissipated by progressively crushing the RAM packagings and the probability of a particular fraction of package failures is estimated by adaptation of the statistical method used previously. The parameters of a common, well-characterized packaging, the 6M with 2R inner containment vessel, were employed as an illustrative example of this analysis method. However, the method is readily applicable to other packagings for which crush strengths have been measured or can be estimated with satisfactory confidence. (authors)

Second generation waste package design and storage concept for the Yucca Mountain Repository

International Nuclear Information System (INIS)

Armijo, Joseph Sam; Kar, Piyush; Misra, Manoranjan

2006-01-01

The reference waste package design and operating mode to be used in the Yucca Mountain Repository is reviewed. An alternate (second generation) operating concept and waste package design is proposed to reduce the risk of localized corrosion of waste packages and to reduce repository costs. The second generation waste package design and storage concept is proposed for implementation after the initial licensing and operation of the reference repository design. Implementation of the second generation concept at Yucca Mountain would follow regulatory processes analogous to those used successfully to extend the design life and uprate the power of commercial light water nuclear reactors in the United States. The second generation concept utilizes the benefits of hot dry storage to minimize the potential for localized corrosion of the waste package by liquid electrolytes. The second generation concept permits major reductions in repository costs by increasing the number of fuel assemblies stored in each waste package, by eliminating the need for titanium drip shields and by fabricating the outer container from corrosion resistant low alloy carbon steel

Example of a quality control operation performed on a nuclear reactor waste package

International Nuclear Information System (INIS)

Nouguier, H.; Nomine, J.S.; Vaunois, P.

1983-01-01

The need has emerged to secure the possibility of the spot sampling and inspection of a package, in the production line or on its arrival at the disposal center. This external inspection (or superinspection) has a threefold objective: to check the conformity of the package and of all its components with the mechanical, chemical and radiochemical specifications; to ensure, by thorough inspection, that the product is similar to the product which was characterized and qualified, and which secured ANDRA's approval of the process; and, through the detection of any variations and discrepancies, to advise the process promoter and the research laboratories, so as to improve the process, the product and the specifications. This paper describes an operation of this type conducted in 1982 by ANDRA, with the help of the CEA's laboratories and technical units, on low and medium-level waste packages produced by an industrial installation in an EDF nuclear power plant. The following are described: dismantling and sampling methods implemented, means employed, and the specific characteristics tested and the first results obtained

On steady-state concentrations of ammonia and molecular hydrogen in the primary circuit of the WWER-1000 reactors

International Nuclear Information System (INIS)

Arkhipov, O.P.; Bugaenko, V.L.; Kamakchi, S.A.

1997-01-01

It is shown that the MORAVA-N2 software package describes well the coolant state in the primary circuit of an actual reactor facility with the WWER-1000 during on-load operation. It permits using the package for analysis of process perturbation effect on the coolant composition. Specific feature of ammonia radiation chemistry in the primary circuit of a reactor facility with the WWER-1000, assuring the rates hydrogen concentration in the coolant with ammonia concentration variation in the coolant within wide limits, when reactor operates on power, can be mentioned by way of example, the fact being ascertained in this study

TARMS, an on-line boiling water reactor operation management system

International Nuclear Information System (INIS)

Iwamoto, T.; Sakurai, S.; Uematsu, H.; Tsuiki, M.; Makino, K.

1984-01-01

The TARMS (Toshiba Advanced Reactor Management System) software package was developed as an effective on-line, on-site tool for boiling water reactor core operation management. It was designed to support a complete function set to meet the requirement to the current on-line process computers. The functions can be divided into two categories. One is monitoring of the present core power distribution as well as related limiting parameters. The other is aiding site engineers or reactor operators in making the future reactor operating plan. TARMS performs these functions with a three-dimensional BWR core physics simulator LOGOS 2, which is based on modified one-group, coarse-mesh nodal diffusion theory. A method was developed to obtain highly accurate nodal powers by coupling LOGOS 2 calculations with the readings of an in-core neutron flux monitor. A sort of automated machine-learning method also was developed to minimize the errors caused by insufficiency of the physics model adopted in LOGOS 2. In addition to these fundamental calculational methods, a number of core operation planning aid packages were developed and installed in TARMS, which were designed to make the operator's inputs simple and easy. (orig.) [de

Aqueous Corrosion Rates for Waste Package Materials

Energy Technology Data Exchange (ETDEWEB)

S. Arthur

2004-10-08

The purpose of this analysis, as directed by ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]), is to compile applicable corrosion data from the literature (journal articles, engineering documents, materials handbooks, or standards, and national laboratory reports), evaluate the quality of these data, and use these to perform statistical analyses and distributions for aqueous corrosion rates of waste package materials. The purpose of this report is not to describe the performance of engineered barriers for the TSPA-LA. Instead, the analysis provides simple statistics on aqueous corrosion rates of steels and alloys. These rates are limited by various aqueous parameters such as temperature (up to 100 C), water type (i.e., fresh versus saline), and pH. Corrosion data of materials at pH extremes (below 4 and above 9) are not included in this analysis, as materials commonly display different corrosion behaviors under these conditions. The exception is highly corrosion-resistant materials (Inconel Alloys) for which rate data from corrosion tests at a pH of approximately 3 were included. The waste package materials investigated are those from the long and short 5-DHLW waste packages, 2-MCO/2-DHLW waste package, and the 21-PWR commercial waste package. This analysis also contains rate data for some of the materials present inside the fuel canisters for the following fuel types: U-Mo (Fermi U-10%Mo), MOX (FFTF), Thorium Carbide and Th/U Carbide (Fort Saint Vrain [FSVR]), Th/U Oxide (Shippingport LWBR), U-metal (N Reactor), Intact U-Oxide (Shippingport PWR, Commercial), aluminum-based, and U-Zr-H (TRIGA). Analysis of corrosion rates for Alloy 22, spent nuclear fuel, defense high level waste (DHLW) glass, and Titanium Grade 7 can be found in other analysis or model reports.

Aqueous Corrosion Rates for Waste Package Materials

International Nuclear Information System (INIS)

Arthur, S.

2004-01-01

The purpose of this analysis, as directed by ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]), is to compile applicable corrosion data from the literature (journal articles, engineering documents, materials handbooks, or standards, and national laboratory reports), evaluate the quality of these data, and use these to perform statistical analyses and distributions for aqueous corrosion rates of waste package materials. The purpose of this report is not to describe the performance of engineered barriers for the TSPA-LA. Instead, the analysis provides simple statistics on aqueous corrosion rates of steels and alloys. These rates are limited by various aqueous parameters such as temperature (up to 100 C), water type (i.e., fresh versus saline), and pH. Corrosion data of materials at pH extremes (below 4 and above 9) are not included in this analysis, as materials commonly display different corrosion behaviors under these conditions. The exception is highly corrosion-resistant materials (Inconel Alloys) for which rate data from corrosion tests at a pH of approximately 3 were included. The waste package materials investigated are those from the long and short 5-DHLW waste packages, 2-MCO/2-DHLW waste package, and the 21-PWR commercial waste package. This analysis also contains rate data for some of the materials present inside the fuel canisters for the following fuel types: U-Mo (Fermi U-10%Mo), MOX (FFTF), Thorium Carbide and Th/U Carbide (Fort Saint Vrain [FSVR]), Th/U Oxide (Shippingport LWBR), U-metal (N Reactor), Intact U-Oxide (Shippingport PWR, Commercial), aluminum-based, and U-Zr-H (TRIGA). Analysis of corrosion rates for Alloy 22, spent nuclear fuel, defense high level waste (DHLW) glass, and Titanium Grade 7 can be found in other analysis or model reports

The incorporation of boron in fissile transport packages for the transport and interim storage of irradiated light water reactor fuels

International Nuclear Information System (INIS)

Hunter, I.J.

1998-01-01

Boron is widely used in the nuclear industry for capturing neutrons and it is particularly useful in the criticality control of packages for the transport and interim storage of irradiated light water reactor fuels. Such fuels are typically located in an internal frame of stainless steel or aluminium, referred to as a basket, which locates the fuel assemblies in channels. Transnucleaire has designed and supplied more than 100 baskets of varying types during the past 30 years. Boron has been incorporated in many forms. Early designs of baskets used boron in specific zones to contribute to the control of criticality. Later developments in new materials dispersed boron throughout the basket and gave designers more options for the basic forms which make up the channels. New basket concepts have been developed by Transnucleaire to meet the changing market needs for transport and interim storage and boron continues to play an important role as an efficient thermal neutron absorber. (author)

Moderation control in low enriched 235U uranium hexafluoride packaging operations and transportation

International Nuclear Information System (INIS)

Dyer, R.H.; Kovac, F.M.; Pryor, W.A.

1993-01-01

Moderation control is the basic parameter for ensuring nuclear criticality safety during the packaging and transport of low 235 U enriched uranium hexafluoride before its conversion to nuclear power reactor fuel. Moderation control has permitted the shipment of bulk quantities in large cylinders instead of in many smaller cylinders and, therefore, has resulted in economies without compromising safety. Overall safety and uranium accountability have been enhanced through the use of the moderation control. This paper discusses moderation control and the operating procedures to ensure that moderation control is maintained during packaging operations and transportation

COOLOD-N2: a computer code, for the analyses of steady-state thermal-hydraulics in research reactors

International Nuclear Information System (INIS)

Kaminaga, Masanori

1994-03-01

The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode as well as COOLOD-N code. In the COOLOD-N2 code, a 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. (author)

Research reactor decommissioning experience - concrete removal and disposal -

International Nuclear Information System (INIS)

Manning, Mark R.; Gardner, Frederick W.

1990-01-01

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

Code package {open_quotes}SVECHA{close_quotes}: Modeling of core degradation phenomena at severe accidents

Energy Technology Data Exchange (ETDEWEB)

Veshchunov, M.S.; Kisselev, A.E.; Palagin, A.V. [Nuclear Safety Institute, Moscow (Russian Federation)] [and others

1995-09-01

The code package SVECHA for the modeling of in-vessel core degradation (CD) phenomena in severe accidents is being developed in the Nuclear Safety Institute, Russian Academy of Science (NSI RAS). The code package presents a detailed mechanistic description of the phenomenology of severe accidents in a reactor core. The modules of the package were developed and validated on separate effect test data. These modules were then successfully implemented in the ICARE2 code and validated against a wide range of integral tests. Validation results have shown good agreement with separate effect tests data and with the integral tests CORA-W1/W2, CORA-13, PHEBUS-B9+.

The use of an object oriented technique for fault diagnosis in nuclear reactors

International Nuclear Information System (INIS)

Jalel, N.A.; Nicholson, H.

1990-11-01

In recent years there has been an increased growth of interest in object oriented programming which is a new approach to software construction having wide application. The possibility of using object oriented programming to build a software package that will assist the nuclear reactor operator in diagnosing any faults or alarms in the Loss Of Fluid Test (LOFT), a small scale pressurised water reactor, reactor is investigated in this work. (author)

AGC-2 Graphite Preirradiation Data Package

Energy Technology Data Exchange (ETDEWEB)

David Swank; Joseph Lord; David Rohrbaugh; William Windes

2012-10-01

The NGNP Graphite R&D program is currently establishing the safe operating envelope of graphite core components for a Very High Temperature Reactor (VHTR) design. The program is generating quantitative data necessary for predicting the behavior and operating performance of the new nuclear graphite grades. To determine the in-service behavior of the graphite for pebble bed and prismatic designs, the Advanced Graphite Creep (AGC) experiment is underway. This experiment is examining the properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences and compressive loads. Each experiment consists of over 400 graphite specimens that are characterized prior to irradiation and following irradiation. Six experiments are planned with the first, AGC-1, currently being irradiated in the Advanced Test Reactor (ATR) and pre-irradiation characterization of the second, AGC-2, completed. This data package establishes the readiness of 512 specimens for assembly into the AGC-2 capsule.

Transportation and packaging issues involving the disposition of surplus plutonium as MOX fuel in commercial LWRs

International Nuclear Information System (INIS)

Ludwig, S.B.; Welch, D.E.; Best, R.E.; Schmid, S.P.

1997-08-01

This report provides a view of anticipated transportation, packaging, and facility handling operations that are expected to occur at mixed-oxide (MOX) fuel fabrication and commercial reactor facilities. This information is intended for use by prospective contractors to the U.S. Department of Energy (DOE) who plan to submit proposals to DOE to manufacture and irradiate MOX fuel assemblies in domestic commercial light-water reactors. The report provides data to prospective consortia regarding packaging and pickup of MOX nuclear fuel assemblies at a MOX fuel manufacturing plant and transport and delivery of the MOX assemblies to nuclear power plants. The report also identifies areas where data are incomplete either because of the status of development or lack of sufficient information and specificity regarding the nuclear power plant(s) where deliveries will take place

EC-FP7 ARCAS: technical and economical comparison of Fast Reactors and Accelerator Driven Systems for transmutation of Minor Actinides

International Nuclear Information System (INIS)

Van den Eynde, G.; Romanello, V.; Heek, A. van; Martin-Fuertes, F.; Zimmerman, C.; Lewin, B.

2015-01-01

The ARCAS project aims to compare, on a technological and economical basis, Accelerator Driven Systems and Fast Reactors as Minor Actinide burners. It is split in five work packages: the reference scenario definition, the fast reactor system definition, the accelerator driven system definition, the fuel reprocessing and fabrication facilities definition and the economical comparison. This paper summarizes the status of the project and its five work packages. (author)

Salt Repository Project Waste Package Program Plan: Draft

International Nuclear Information System (INIS)

Carr, J.A.; Cunnane, J.C.

1986-01-01

Under the direction of the Office of Civilian Radioactive Waste Management (OCRWM) created within the DOE by direction of the Nuclear Waste Policy Act of 1982 (NWPA), the mission of the Salt Repository Project (SRP) is to provide for the development of a candidate salt repository for disposal of high-level radioactive waste (HLW) and spent reactor fuel in a manner that fully protects the health and safety of the public and the quality of the environment. In consideration of the program needs and requirements discussed above, the SRP has decided to develop and issue this SRP Waste Package Program Plan. This document is intended to outline how the SRP plans to develop the waste package design and to show, with reasonable assurance, that the developed design will satisfy applicable requirements/performance objectives. 44 refs., 16 figs., 16 tabs

Using package MESSAGE for optimization studies of nuclear power structures

International Nuclear Information System (INIS)

Andrianov, A.A.; Fedorova, E.V.; Korobejnikov, V.V.; Poplavskaya, E.V.; Rachkova, E.N.

2010-01-01

The results of optimization research of Russia nuclear power strategies, obtained for different assumptions concerning availability of natural uranium resources were presented. The ability of energy planning package MESSAGE (Model for Energy Supply Strategy Alternatives and their General Environmental Impacts) application for elaborating breeding parameters requirements of fast sodium reactors and assessing the required scale of nuclear fuel cycle enterprises development was demonstrated [ru

Cleanup Verification Package for the 118-F-2 Burial Ground

International Nuclear Information System (INIS)

Capron, J.M.; Anselm, K.A.

2008-01-01

This cleanup verification package documents completion of remedial action, sampling activities, and compliance with cleanup criteria for the 118-F-2 Burial Ground. This burial ground, formerly called Solid Waste Burial Ground No. 1, was the original solid waste disposal site for the 100-F Area. Eight trenches contained miscellaneous solid waste from the 105-F Reactor and one trench contained solid waste from the biology facilities

Validation of the DRAGON/DONJON code package for MNR using the IAEA 10 MW benchmark problem

International Nuclear Information System (INIS)

Day, S.E.; Garland, W.J.

2000-01-01

The first step in developing a framework for reactor physics analysis is to establish the appropriate and proven reactor physics codes. The chosen code package is tested, by executing a benchmark problem and comparing the results to the accepted standards. The IAEA 10 MW Benchmark problem is suitable for static reactor physics calculations on plate-fueled research reactor systems and has been used previously to validate codes for the McMaster Nuclear (MNR). The flexible and advanced geometry capabilities of the DRAGON transport theory code make it a desirable tool, and the accompanying DONJON diffusion theory code also has useful features applicable to safety analysis work at MNR. This paper describes the methodology used to benchmark the DRAGON/DONJON code package against this problem and the results herein extend the domain of validation of this code package. The results are directly applicable to MNR and are relevant to a reduced-enrichment fuel program. The DRAGON transport code models, used in this study, are based on the 1-D infinite slab approximation whereas the DONJON diffusion code models are defined in 3-D Cartesian geometry. The cores under consideration are composed of HEU (93% enrichment), MEU (45% enrichment) and LEU (20% enrichment) fuel and are examined in a fresh state, as well as at beginning-of-life (BOL) and end-of-life (EOL) exposures. The required flux plots and flux-ratio plots are included, as are transport theory code k∞and diffusion theory code k eff results. In addition to this, selected isotope atom densities are charted as a function of fuel burnup. Results from this analysis are compared to and are in good agreement with previously published results. (author)

Intact and Degraded Component Criticality Calculations of N Reactor Spent Nuclear Fuel

International Nuclear Information System (INIS)

L. Angers

2001-01-01

The objective of this calculation is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) N Reactor Spent Nuclear Fuel codisposed in a 2-Defense High-Level Waste (2-DHLW)/2-Multi-Canister Overpack (MCO) Waste Package (WP) and emplaced in a monitored geologic repository (MGR) (see Attachment I). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (k eff ) for both intact and degraded mode internal configurations of the codisposal waste package. This calculation will support the analysis that will be performed to demonstrate the technical viability for disposing of U-metal (N Reactor) spent nuclear fuel in the potential MGR

Research on V and V strategy of reactor physics code of COSINE

International Nuclear Information System (INIS)

Liu Zhanquan; Chen Yixue; Yang Chao; Dang Halei

2013-01-01

Verification and validation (V and V) is very important for the software quality assurance. Reasonable and efficient V and V strategy can achieve twice the result with half the effort. Core and system integrated engine for design and analysis (COSINE) software package contains three reactor physics codes, the lattice code (LATC), the core simulator (CORE) and the kinetics code (KIND), which is called the reactor physics subsystem. The V and V strategy for the physics subsystem was researched based on the foundation of scientific software's V and V method. The module based verification method and the function based validation method were proposed, composing the physical subsystem V and V strategy of COSINE software package. (authors)

World-wide French experience in research reactor fuel cycle transportation

International Nuclear Information System (INIS)

Raisonnier, D.

1997-01-01

Since 1963 Transnucleaire has safely performed a large number of national and international transports of radioactive material. Transnucleaire has also designed and supplied suitable packagings for all types of nuclear fuel cycle radioactive material from front-end and back-end products and for power or for research reactors. Transportation of the nuclear fuel material for power reactors is made on a regular and industrial basis. The transportation of material for the research reactor fuel cycle is quite different due to the small quantities involved, the categorisation of material and the numerous places of delivery world-wide. Adapted solutions exist, which require a reactive organisation dealing with all the transportation issues for LEU and HEU products as metal, oxide, fresh fuel elements, spent fuel elements including supply of necessary transport packaging and equipment. This presentation will: - explain the choices made by Transnucleaire and its associates to provide and optimise the corresponding services, - demonstrate the capability to achieve, through reliable partnership, transport operations involving new routes, specific equipment and new political constraints while respecting sophisticated safety and security regulations. (author)

SEJV2 software package for radiation monitoring system of WWER 440 NPP

International Nuclear Information System (INIS)

Kapisovsky, V.; Jancik, O.; Kubik, I.; Bena, J.

1993-01-01

The main part of the radiation monitoring system at a WWER-440 (213 reactor type) nuclear power plant is the centralized 400-channel monitoring system 'SEJVAL' servicing twin reactor units. The SEJV2 software package is described developed to run on a PC with an IFS2 interface to the SEJVAL radiation monitoring system. It provides enhanced data presentation, record keeping and report generation, thus improving the efficiency of the health physics shift. The system was for the first time implemented at the Jaslovske Bohunice V-2 nuclear power plant with encouraging results. (Z.S.) 3 refs

Safety Analysis Report - Packages, 9965, 9968, 9972-9975 Packages

International Nuclear Information System (INIS)

Blanton, P.

2000-01-01

This Safety Analysis Report for Packaging (SARP) documents the analysis and testing performed on four type B Packages: the 9972, 9973, 9974, and 9975 packages. Because all four packages have similar designs with very similar performance characteristics, all of them are presented in a single SARP. The performance evaluation presented in this SARP documents the compliance of the 9975 package with the regulatory safety requirements. Evaluations of the 9972, 9973, and 9974 packages support that of the 9975. To avoid confusion arising from the inclusion of four packages in a single document, the text segregates the data for each package in such a way that the reader interested in only one package can progress from Chapter 1 through Chapter 9. The directory at the beginning of each chapter identifies each section that should be read for a given package. Sections marked ''all'' are generic to all packages

Experience in dismantling and packaging of pressure vessel and core internals

International Nuclear Information System (INIS)

Pillokat, Peter; Bruhn, Jan Hendrik

2011-01-01

Nuclear Company AREVA is proud to look back on versatile experience in successfully dismantling nuclear components. After performing several minor dismantling projects and studies for nuclear power plants, AREVA completed the order for dismantling of all remaining Reactor Pressure Vessel internals at German Boiling Water Reactor Wuergassen NPP in October '08. During the onsite activities about 121 tons of steel were successfully cut and packed under water into 200l- drums, as the dismantling was performed partly in situ and partly in an underwater working tank. AREVA deployed a variety of different cutting techniques such as band sawing, milling, nibbling, compass sawing and water jet cutting throughout this project. After successfully finishing this task, AREVA dismantled the cylindrical part of the Wuergassen Pressure Vessel. During this project approximately 320 tons of steel were cut and packaged for final disposal, as dismantling was mainly performed by on air use of water jet cutting with vacuum suction of abrasive and kerfs material. The main clue during this assignment was the logistic challenge to handle and convey cut pieces from the pressure vessel to the packing area. For this, an elevator was installed to transport cut segments into the turbine hall, where a special housing was built for final storage conditioning. At the beginning of 2007, another complex dismantling project of great importance was acquired by AREVA. The contract included dismantling and conditioning for final storage of the complete RPV Internals of the German Pressurized Water Reactor Stade NPP. Very similar cutting techniques turned out to be the proper policy to cope this task. On-site activities took place in up to 5 separate working areas including areas for post segmentation and packaging to perform optimized parallel activities. All together about 85 tons of Core Internals were successfully dismantled at Stade NPP until September '09. To accomplish the best possible on

Full scale tests on remote handled FFTF fuel assembly waste handling and packaging

International Nuclear Information System (INIS)

Allen, C.R.; Cash, R.J.; Dawson, S.A.; Strode, J.N.

1986-01-01

Handling and packaging of remote handled, high activity solid waste fuel assembly hardware components from spent FFTF reactor fuel assemblies have been evaluated using full scale components. The demonstration was performed using FFTF fuel assembly components and simulated components which were handled remotely using electromechanical manipulators, shielding walls, master slave manipulators, specially designed grapples, and remote TV viewing. The testing and evaluation included handling, packaging for current and conceptual shipping containers, and the effects of volume reduction on packing efficiency and shielding requirements. Effects of waste segregation into transuranic (TRU) and non-transuranic fractions also are discussed

Using SAFRAN Software to Assess Radiological Hazards from Dismantling of Tammuz-2 Reactor Core at Al-tuwaitha Nuclear Site

Science.gov (United States)

Abed Gatea, Mezher; Ahmed, Anwar A.; jundee kadhum, Saad; Ali, Hasan Mohammed; Hussein Muheisn, Abbas

2018-05-01

The Safety Assessment Framework (SAFRAN) software has implemented here for radiological safety analysis; to verify that the dose acceptance criteria and safety goals are met with a high degree of confidence for dismantling of Tammuz-2 reactor core at Al-tuwaitha nuclear site. The activities characterizing, dismantling and packaging were practiced to manage the generated radioactive waste. Dose to the worker was considered an endpoint-scenario while dose to the public has neglected due to that Tammuz-2 facility is located in a restricted zone and 30m berm surrounded Al-tuwaitha site. Safety assessment for dismantling worker endpoint-scenario based on maximum external dose at component position level in the reactor pool and internal dose via airborne activity while, for characterizing and packaging worker endpoints scenarios have been done via external dose only because no evidence for airborne radioactivity hazards outside the reactor pool. The in-situ measurements approved that reactor core components are radiologically activated by Co-60 radioisotope. SAFRAN results showed that the maximum received dose for workers are (1.85, 0.64 and 1.3mSv/y) for activities dismantling, characterizing and packaging of reactor core components respectively. Hence, the radiological hazards remain below the low level hazard and within the acceptable annual dose for workers in radiation field

Multimodal shipments under program on Russian-origin research reactor SFA return to Russian Federation

International Nuclear Information System (INIS)

Dorofeev, A.N.; Ivashchenko, A.A.; Kanashov, B.A.; Komarov, S.V.; Komarov, S.N.; Barinkov, O.P.

2010-01-01

The paper describes experience in preparation and organization of research reactor nuclear material import under the Program on Russian-Origin Highly Enriched Uranium Return to the Russian Federation. It also summarizes evolution of transport equipment, conveyances and routes and describes types of packages, their adaptation and certification, safety issues, peculiarities and prospective use of the packagings and conveyances. (author)

Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

Energy Technology Data Exchange (ETDEWEB)

Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

2013-07-01

Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle

Programmatic and technical requirements for the FMDP fresh MOX fuel transport package

International Nuclear Information System (INIS)

Ludwig, S.B.; Michelhaugh, R.D.; Pope, R.B.

1997-12-01

This document is intended to guide the designers of the package to all pertinent regulatory and other design requirements to help ensure the safe and efficient transport of the weapons-grade (WG) fresh MOX fuel under the Fissile Materials Disposition Program. To accomplish the disposition mission using MOX fuel, the unirradiated MOX fuel must be transported from the MOX fabrication facility to one or more commercial reactors. Because the unirradiated fuel contains large quantities of plutonium and is not sufficient radioactive to create a self-protecting barrier to deter the material from theft, DOE intends to use its fleet of safe secure trailers (SSTs) to provide the necessary safeguards and security for the material in transit. In addition to these requirements, transport of radioactive materials must comply with regulations of the Department of Transportation and the Nuclear Regulatory Commission (NRC). In particular, NRC requires that the packages must meet strict performance requirements. The requirements for shipment of MOX fuel (i.e., radioactive fissile materials) specify that the package design is certified by NRC to ensure the materials contained in the packages are not released and remain subcritical after undergoing a series of hypothetical accident condition tests. Packages that pass these tests are certified by NRC as a Type B fissile (BF) package. This document specifies the programmatic and technical design requirements a package must satisfy to transport the fresh MOX fuel assemblies

Optimization of reactor pressure vessel internals segmentation in Korea

Energy Technology Data Exchange (ETDEWEB)

Lee, Byung-Sik [Dankook Univ., Chungnam (Korea, Republic of). Dept. of Nuclear Engineering

2017-11-15

One of the most challenging tasks during plant decommissioning is the removal of highly radioactive internal components from the reactor pressure vessel (RPV). For RPV internals dismantling, it is essential that all activities are thoroughly planned and discussed in the early stage of the decommissioning project. One of the key activities in the detailed planning is to prepare the segmentation and packaging plan that describes the sequential steps required to segment, separate, and package each individual component of RPV, based on an activation analysis and component characterization study.

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

International Nuclear Information System (INIS)

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

1984-01-01

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

French experience in research reactor fuel transportation

International Nuclear Information System (INIS)

Raisonnier, Daniele

1996-01-01

Since 1963 Transnucleaire has safely performed a large number of national and international transports of radioactive material. Transnucleaire has also designed and supplied suitable packaging for all types of nuclear fuel cycle radioactive material from front-end and back-end products and for power or for research reactors. Transportation of spent fuel from power reactors are made on a regular and industrial basis, but this is not yet the case for the transport of spent fuel coming from research reactors. Each shipment is a permanent challenge and requires a reactive organization dealing with all the transportation issues. This presentation will explain the choices made by Transnucleaire and its associates to provide and optimize the corresponding services while remaining in full compliance with the applicable regulations and customer requirements. (author)

Use of source term code package in the ELEBRA MX-850 system

International Nuclear Information System (INIS)

Guimaraes, A.C.F.; Goes, A.G.A.

1988-12-01

The implantation of source term code package in the ELEBRA-MX850 system is presented. The source term is formed when radioactive materials generated in nuclear fuel leakage toward containment and the external environment to reactor containment. The implantated version in the ELEBRA system are composed of five codes: MARCH 3, TRAPMELT 3, THCCA, VANESA and NAVA. The original example case was used. The example consists of a small loca accident in a PWR type reactor. A sensitivity study for the TRAPMELT 3 code was carried out, modifying the 'TIME STEP' to estimate the processing time of CPU for executing the original example case. (M.C.K.) [pt

Packaging Solutions : Delivering customer value through Logistical Packaging: A Case Study at Stora Enso Packaging

OpenAIRE

Shan, Kun; Julius, Joezer

2015-01-01

AbstractBackground;Despite of the significant role of packaging within logistics and supply chain management, packaging is infrequently studied as focal point in supply chain. Most of the previous logistics research studies tend to explain the integration between packaging and logistics through logistical packaging. In very rare cases, the studies mentioned about customer value. Therefore the major disadvantage of these studies is that, they didn’t consider logistical packaging and customer v...

Cleanup Verification Package for the 600-259 Waste Site

Energy Technology Data Exchange (ETDEWEB)

J. M. Capron

2006-02-09

This cleanup verification package documents completion of remedial action for the 600-259 waste site. The site was the former site of the Special Waste Form Lysimeter, consisting of commercial reactor isotope waste forms in contact with soils within engineered caissons, and was used by Pacific Northwest National Laboratory to collect data regarding leaching behavior for target analytes. A Grout Waste Test Facility also operated at the site, designed to test leaching rates of grout-solidified low-level radioactive waste.

Cleanup Verification Package for the 600-259 Waste Site

International Nuclear Information System (INIS)

Capron, J.M.

2006-01-01

This cleanup verification package documents completion of remedial action for the 600-259 waste site. The site was the former site of the Special Waste Form Lysimeter, consisting of commercial reactor isotope waste forms in contact with soils within engineered caissons, and was used by Pacific Northwest National Laboratory to collect data regarding leaching behavior for target analytes. A Grout Waste Test Facility also operated at the site, designed to test leaching rates of grout-solidified low-level radioactive waste

Packaging fluency

DEFF Research Database (Denmark)

Mocanu, Ana; Chrysochou, Polymeros; Bogomolova, Svetlana

2011-01-01

Research on packaging stresses the need for packaging design to read easily, presuming fast and accurate processing of product-related information. In this paper we define this property of packaging as “packaging fluency”. Based on the existing marketing and cognitive psychology literature on pac...

A reactor vessel and its internals disassembled and packaged

International Nuclear Information System (INIS)

Eisenmann, B.; Prechtl, E.; Suessdorf, W.

2008-01-01

2007 was a successful year for the Disassembly Unit of the Karlsruhe Research Center: dismantling the highly activated Karlsruhe Multipurpose Research Reactor (MZFR) was completed successfully and without any incident. A vote of thanks is expressed at this point to all staff members, participating industries and institutions for their extraordinary commitment and their outstanding innovation and work. Preserving and advancing existing knowledge is one of the important pillars securing a sustainable future for generations to come in Germany. Securing and advancing know-how in nuclear technology was defined as a major duty last in late January 2008 by Dr. Peter Fritz, Chairman of the Kerntechnische Gesellschaft e.V. (KTG) and member of the Executive Board of the Karlsruhe Research Center, for instance, in cooperation between the KERNTECHNIK (i.e. nuclear technology) Southwestern Research and Teaching Association and the Karlsruhe Institute of Technology (KIT). The borders of Germany do not constitute a natural radiological barrier, despite the efforts by political groups in this country to convey this impression. This report therefore is to document that the demolition of a nuclear reactor with a high radioactivity inventory can be managed safely. In the light of the experience accumulated with the MZFR, this is feasible only if demolition is carried out immediately instead of the 'problem' of disassembly and conditioning being shifted to future generations. The article is also meant to be a piece of advice by showing the unplannable difficulties which came up, and the technical solutions implemented by the competence team successfully and efficiently. (orig.)

Proposal of guidelines for selecting optimum options in packagings and transportation systems of spent fuel

International Nuclear Information System (INIS)

Saegusa, T.; Abe, H.; Fukuda, S.

1983-01-01

Type and size of spent fuel shipping packagings and packaging transport ships in spent fuel transport system would have been determined separately in response to technical requirements etc. of reactor sites and reprocessing plants. However, since more and more spent fuel will be generated from world's nuclear power plants and will be transported much frequently to reprocessing plants or storage facilities, the current spent fuel transport system will have to be necessarily reexamined from the operational and economical aspects or an optimum transport system may have to be newly determined in the near future. In the literature, a variety of options are found, particularly of spent fuel packagings. This paper listed and classified options of spent fuel packagings and packaging transport ships in the transportation systems of spent fuel on the basis of literature surveys. These options were discussed from viewpoints of designers and users and compared in terms of transport efficiency. Finally, one way to determine an optimum transport system of spent fuel was indicated considering the total transport system in the light of safety, operational efficiency and economy

Automation of reactor neutron activation analysis

International Nuclear Information System (INIS)

Pavlov, S.S.; Dmitriev, A.Yu.; Frontasyeva, M.V.

2013-01-01

The present status of the development of a software package designed for automation of NAA at the IBR-2 reactor of FLNP, JINR, Dubna, is reported. Following decisions adopted at the CRP Meeting in Delft, August 27-31, 2012, the missing tool - a sample changer - will be installed for NAA in compliance with the peculiar features of the radioanalytical laboratory REGATA at the IBR-2 reactor. The details of the design are presented. The software for operation with the sample changer consists of two parts. The first part is a user interface and the second one is a program to control the sample changer. The second part will be developed after installing the tool.

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

International Nuclear Information System (INIS)

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

1995-09-01

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

Microelectronic packaging

CERN Document Server

Datta, M; Schultze, J Walter

2004-01-01

Microelectronic Packaging analyzes the massive impact of electrochemical technologies on various levels of microelectronic packaging. Traditionally, interconnections within a chip were considered outside the realm of packaging technologies, but this book emphasizes the importance of chip wiring as a key aspect of microelectronic packaging, and focuses on electrochemical processing as an enabler of advanced chip metallization.Divided into five parts, the book begins by outlining the basics of electrochemical processing, defining the microelectronic packaging hierarchy, and emphasizing the impac

The Packaging Handbook -- A guide to package design

International Nuclear Information System (INIS)

Shappert, L.B.

1995-01-01

The Packaging Handbook is a compilation of 14 technical chapters and five appendices that address the life cycle of a packaging which is intended to transport radioactive material by any transport mode in normal commerce. Although many topics are discussed in depth, this document focuses on the design aspects of a packaging. The Handbook, which is being prepared under the direction of the US Department of Energy, is intended to provide a wealth of technical guidance that will give designers a better understanding of the regulatory approval process, preferences of regulators in specific aspects of packaging design, and the types of analyses that should be seriously considered when developing the packaging design. Even though the Handbook is concerned with all packagings, most of the emphasis is placed on large packagings that are capable of transporting large radioactive sources that are also fissile (e.g., spent fuel). These are the types of packagings that must address the widest range of technical topics in order to meet domestic and international regulations. Most of the chapters in the Handbook have been drafted and submitted to the Oak Ridge National Laboratory for editing; the majority of these have been edited. This report summarizes the contents

Reprocessing technology for present water reactor fuels

International Nuclear Information System (INIS)

McMurray, P.R.

1977-01-01

The basic Purex solvent extraction technology developed and applied in the U.S. in the 1950's provides a well-demonstrated and efficient process for recovering uranium and plutonium for fuel recycle and separating the wastes for further treatment and packaging. The technologies for confinement of radioactive effluents have been developed but have had limited utilization in the processing of commercial light water reactor fuels. Technologies for solidification and packaging of radioactive wastes have not yet been demonstrated but significant experience has been gained in laboratory and engineering scale experiments with simulated commercial reprocessing wastes and intermediate level wastes. Commercial scale experience with combined operations of all the required processes and equipment are needed to demonstrate reliable reprocessing centers

Evaluating the loss of a LWR spent fuel or plutonium shipping package into the sea

International Nuclear Information System (INIS)

Heaberlin, S.W.; Baker, D.A.

1976-06-01

As the nations of the world turn to nuclear power for an energy source, commerce in nuclear fuel cycle materials will increase. Some of this commerce will be transported by sea. Such shipments give rise to the possibility of loss of these materials into the sea. This paper discusses the postulated accidental loss of two materials, light water reactor (LWR) spent fuel and plutonium, at sea. The losses considered are that of a single shipping package which is either undamaged or damaged by fire prior to the loss. The containment failure of the package in the sea,

Benchmark calculation for GT-MHR using HELIOS/MASTER code package and MCNP

International Nuclear Information System (INIS)

Lee, Kyung Hoon; Kim, Kang Seog; Noh, Jae Man; Song, Jae Seung; Zee, Sung Quun

2005-01-01

The latest research associated with the very high temperature gas-cooled reactor (VHTR) is focused on the verification of a system performance and safety under operating conditions for the VHTRs. As a part of those, an international gas-cooled reactor program initiated by IAEA is going on. The key objectives of this program are the validation of analytical computer codes and the evaluation of benchmark models for the projected and actual VHTRs. New reactor physics analysis procedure for the prismatic VHTR is under development by adopting the conventional two-step procedure. In this procedure, a few group constants are generated through the transport lattice calculations using the HELIOS code, and the core physics analysis is performed by the 3-dimensional nodal diffusion code MASTER. We evaluated the performance of the HELIOS/MASTER code package through the benchmark calculations related to the GT-MHR (Gas Turbine-Modular Helium Reactor) to dispose weapon plutonium. In parallel, MCNP is employed as a reference code to verify the results of the HELIOS/MASTER procedure

A personal computer based console monitor for a TRIGA reactor

International Nuclear Information System (INIS)

Rieke, Phillip E.; Hood, William E.; Razvi, Junaid

1990-01-01

Numerous improvements have been made to the Mark F facility to provide a minimum reactor down time, giving a high reactor availability. A program was undertaken to enhance the monitoring capabilities of the instrumentation and control system on this reactor. To that end, a personal computer based console monitoring system has been developed, installed in the control room and is operational to provide real-time monitoring and display of a variety of reactor operating parameters. This system is based on commercially available hardware and an applications software package developed internally at the GA facility. It has (a) assisted the operator in controlling reactor parameters to maintain the high degree of power stability required during extended runs with thermionic devices in-core, and (b) provided data trending and archiving capabilities on all monitored channels to allow a post-mortem analysis to be performed on any of the monitored parameters

Overview of the MCU Monte Carlo software package

International Nuclear Information System (INIS)

Kalugin, M.A.; Oleynik, D.S.; Shkarovsky, D.A.

2013-01-01

MCU (Monte Carlo Universal) is a project on development and practical use of a universal computer code for simulation of particle transport (neutrons, photons, electrons, positrons) in three-dimensional systems by means of the Monte Carlo method. This paper provides the information on the current state of the project. The developed libraries of constants are briefly described, and the potentialities of the MCU-5 package modules and the executable codes compiled from them are characterized. Examples of important problems of reactor physics solved with the code are presented. It is shown that the MCU constructor tool is able to assemble a full-scale 3D model from templates describing single components using simple and intuitive graphic user interface. The templates are prepared by a skilled user and stored in constructor's templates library. Ordinary user works with the graphic user interface and does not deal with MCU input data directly. At the present moment there are template libraries for several types of reactors

Chooz A, First Pressurized Water Reactor to be Dismantled in France - 13445

Energy Technology Data Exchange (ETDEWEB)

Boucau, Joseph [Westinghouse Electric Company, 43 rue de l' Industrie, Nivelles (Belgium); Mirabella, C. [Westinghouse Electric France, Orsay (France); Nilsson, Lennart [Westinghouse Electric Sweden, Vaesteraas (Sweden); Kreitman, Paul J. [Westinghouse Electric Company, Lake Bluff, IL 60048 (United States); Obert, Estelle [EDF - DPI - CIDEN, Lyon (France)

2013-07-01

Nine commercial nuclear power plants have been permanently shut down in France to date, of which the Chooz A plant underwent an extensive decommissioning and dismantling program. Chooz Nuclear Power Station is located in the municipality of Chooz, Ardennes region, in the northeast part of France. Chooz B1 and B2 are 1,500 megawatt electric (MWe) pressurized water reactors (PWRs) currently in operation. Chooz A, a 305 MWe PWR implanted in two caves within a hill, began operations in 1967 and closed in 1991, and will now become the first PWR in France to be fully dismantled. EDF CIDEN (Engineering Center for Dismantling and Environment) has awarded Westinghouse a contract for the dismantling of its Chooz A reactor vessel (RV). The project began in January 2010. Westinghouse is leading the project in a consortium with Nuvia France. The project scope includes overall project management, conditioning of the reactor vessel (RV) head, RV and RV internals segmentation, reactor nozzle cutting for lifting the RV out of the pit and seal it afterwards, dismantling of the RV thermal insulation, ALARA (As Low As Reasonably Achievable) forecast to ensure acceptable doses for the personnel, complementary vacuum cleaner to catch the chips during the segmentation work, needs and facilities, waste characterization and packaging, civil work modifications, licensing documentation. The RV and RV internals will be segmented based on the mechanical cutting technology that Westinghouse applied successfully for more than 13 years. The segmentation activities cover the cutting and packaging plan, tooling design and qualification, personnel training and site implementation. Since Chooz A is located inside two caves, the project will involve waste transportation from the reactor cave through long galleries to the waste buffer area. The project will end after the entire dismantling work is completed, and the waste storage is outside the caves and ready to be shipped either to the ANDRA (French

Interim waste storage for the Integral Fast Reactor

International Nuclear Information System (INIS)

Benedict, R.W.; Phipps, R.D.; Condiff, D.W.

1991-01-01

The Integral Fast Reactor (IFR), which Argonne National Laboratory is developing, is an innovative liquid metal breeder reactor that uses metallic fuel and has a close coupled fuel recovery process. A pyrochemical process is used to separate the fission products from the actinide elements. These actinides are used to make new fuel for the reactor. As part of the overall IFR development program, Argonne has refurbished an existing Fuel Cycle Facility at ANL-West and is installing new equipment to demonstrate the remote reprocessing and fabrication of fuel for the Experimental Breeder Reactor II (EBR-II). During this demonstration the wastes that are produced will be treated and packaged to produce waste forms that would be typical of future commercial operations. These future waste forms would, assuming Argonne development goals are fulfilled, be essentially free of long half-life transuranic isotopes. Promising early results indicate that actinide extraction processes can be developed to strip these isotopes from waste stream and return them to the IFR type reactors for fissioning. 1 fig

In-core fuel management code package validation for BWRs

International Nuclear Information System (INIS)

1995-12-01

The main goal of the present CRP (Coordinated Research Programme) was to develop benchmarks which are appropriate to check and improve the fuel management computer code packages and their procedures. Therefore, benchmark specifications were established which included a set of realistic data for running in-core fuel management codes. Secondly, the results of measurements and/or operating data were also provided to verify and compare with these parameters as calculated by the in-core fuel management codes or code packages. For the BWR it was established that the Mexican Laguna Verde 1 BWR would serve as the model for providing data on the benchmark specifications. It was decided to provide results for the first 2 cycles of Unit 1 of the Laguna Verde reactor. The analyses of the above benchmarks are performed in two stages. In the first stage, the lattice parameters are generated as a function of burnup at different voids and with and without control rod. These lattice parameters form the input for 3-dimensional diffusion theory codes for over-all reactor analysis. The lattice calculations were performed using different methods, such as, Monte Carlo, 2-D integral transport theory methods. Supercell Model and transport-diffusion model with proper correction for burnable absorber. Thus the variety of results should provide adequate information for any institute or organization to develop competence to analyze In-core fuel management codes. 15 refs, figs and tabs

Modal properties of the flexural vibrating package of rods linked by spacer grids

Directory of Open Access Journals (Sweden)

Zeman V.

2011-06-01

Full Text Available The paper deals with the modelling and modal analysis of the large package of identical parallel rods linked by transverse springs (spacer grids placed on several level spacings. The rod discretization by finite element method is based on Rayleigh beam theory. For the cyclic and central symmetric package of rods (such as fuel rods in nuclear fuel assembly the system decomposition on the identical revolved rod segments was applied. A modal synthesis method with condensation is used for modelling of the whole system. The presented method is the first step for modelling the nuclear fuel assembly vibration caused by excitation determined by the support plate motion of the reactor core.

Thirty years of transport package development for spent fuels

International Nuclear Information System (INIS)

Cory, A.R.

2005-01-01

probe technical weaknesses in both package designs and safety cases. Many packages currently operated were designed around the reactor and fuel characteristics prevailing in the 1970s. Over the lifetime of the packages the responsible engineering teams have faced and met the successive challenges to develop the capability of the Package to face the changing requirements of the industry and the Transport Regulations. Both Package and Safety Case have been revised such that they have continued to service the needs of the industry around thirty years after the initial designs were produced. (authors)

Reactor core design optimization of the 200 MWt Pb-Bi cooled fast reactor for hydrogen production

International Nuclear Information System (INIS)

Bahrum, Epung Saepul; Su'ud, Zaki; Waris, Abdul; Fitriyani, Dian; Wahjoedi, Bambang Ari

2008-01-01

In this study reactor core geometrical optimization of 200 MWt Pb-Bi cooled long life fast reactor for hydrogen production has been conducted. The reactor life time is 20 years and the fuel type is UN-PuN. Geometrical core configurations considered in this study are balance, pancake and tall cylindrical cores. For the hydrogen production unit we adopt steam membrane reforming hydrogen gas production. The optimum operating temperature for the catalytic reaction is 540degC. Fast reactor design optimization calculation was run by using FI-ITB-CHI software package. The design criteria were restricted by the multiplication factor that should be less than 1.002, the average outlet coolant temperature 550degC and the maximum coolant outlet temperature less than 700degC. By taking into account of the hydrogen production as well as corrosion resulting from Pb-Bi, the balance cylindrical geometrical core design with diameter and height of the active core of 157 cm each, the inlet coolant temperature of 350degC and the coolant flow rate of 7000 kg/s were preferred as the best design parameters. (author)

Reactor physics verification of the MCNP6 unstructured mesh capability

International Nuclear Information System (INIS)

Burke, T. P.; Kiedrowski, B. C.; Martz, R. L.; Martin, W. R.

2013-01-01

The Monte Carlo software package MCNP6 has the ability to transport particles on unstructured meshes generated from the Computed-Aided Engineering software Abaqus. Verification is performed using benchmarks with features relevant to reactor physics - Big Ten and the C5G7 computational benchmark. Various meshing strategies are tested and results are compared to reference solutions. Computational performance results are also given. The conclusions show MCNP6 is capable of producing accurate calculations for reactor physics geometries and the computational requirements for small lattice benchmarks are reasonable on modern computing platforms. (authors)

Reactor physics verification of the MCNP6 unstructured mesh capability

Energy Technology Data Exchange (ETDEWEB)

Burke, T. P. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Boulevard, Ann Arbor, MI 48109 (United States); Kiedrowski, B. C.; Martz, R. L. [X-Computational Physics Division, Monte Carlo Codes Group, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Martin, W. R. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Boulevard, Ann Arbor, MI 48109 (United States)

2013-07-01

The Monte Carlo software package MCNP6 has the ability to transport particles on unstructured meshes generated from the Computed-Aided Engineering software Abaqus. Verification is performed using benchmarks with features relevant to reactor physics - Big Ten and the C5G7 computational benchmark. Various meshing strategies are tested and results are compared to reference solutions. Computational performance results are also given. The conclusions show MCNP6 is capable of producing accurate calculations for reactor physics geometries and the computational requirements for small lattice benchmarks are reasonable on modern computing platforms. (authors)

New package systems for EDF recycling policy

International Nuclear Information System (INIS)

Leboucher, I.; Michels, L.; Hugon, F.C.; Penoty, C.; Froment, A.

2004-01-01

Since the beginning of the nineties, EDF (Electricite France) has been carrying out a policy towards increasing burn-up. As a consequence and to keep energetic performances of the MOX (Mixed OXide uranium and plutonium) fresh fuel assemblies at parity with UO2 fuel, the concentration in plutonium increased up to 7.1%, soon up to 8.65% and even more. To take into account the increased dose rate and the increased thermal power of these MOX fresh fuels in the transportation system, COGEMA LOGISTICS designed, manufactured and licensed the MX8 package to replace the previous packaging. Simultaneously new high-secured transport system was developed. Due to the capacity of the reactor pool, the spent fuel assemblies must be evacuated with a short cooling time and high residual power. Because of the neutron source of these future MOX spent fuel assemblies, the TN 12/2 packaging, which has been in operation since 1980's to transport UO2 or MOX spent fuel to COGEMA La Hague, from EDF, other European and Japanese NPP lacks flexibility. COGEMA LOGISTICS is designing a new cask (TN trademark 112) to transport the future EDF MOX spent fuel assemblies. The TN trademark 112 has to comply with the 1996 edition of the IAEA transport regulations and with the transport and handling equipment presently used for the transport of spent fuel in France. It can be loaded with 12 assemblies, either MOX or UO2. This paper gives a description of two packages belonging to the new generation: - one for fresh MOX fuel assemblies transportation, used on EDF's power plants for MOX reloads, - one for spent MOX fuel assemblies transportation, presently under development with a first transport planned for mid-2007

New package systems for EDF recycling policy

Energy Technology Data Exchange (ETDEWEB)

Leboucher, I.; Michels, L.; Hugon, F.C. [COGEMA Logistics (AREVA Group) (France); Penoty, C. [EDF - Electricite de France (France); Froment, A. [COGEMA (AREVA Group) (France)

2004-07-01

Since the beginning of the nineties, EDF (Electricite France) has been carrying out a policy towards increasing burn-up. As a consequence and to keep energetic performances of the MOX (Mixed OXide uranium and plutonium) fresh fuel assemblies at parity with UO2 fuel, the concentration in plutonium increased up to 7.1%, soon up to 8.65% and even more. To take into account the increased dose rate and the increased thermal power of these MOX fresh fuels in the transportation system, COGEMA LOGISTICS designed, manufactured and licensed the MX8 package to replace the previous packaging. Simultaneously new high-secured transport system was developed. Due to the capacity of the reactor pool, the spent fuel assemblies must be evacuated with a short cooling time and high residual power. Because of the neutron source of these future MOX spent fuel assemblies, the TNa12/2 packaging, which has been in operation since 1980's to transport UO2 or MOX spent fuel to COGEMA La Hague, from EDF, other European and Japanese NPP lacks flexibility. COGEMA LOGISTICS is designing a new cask (TN trademark 112) to transport the future EDF MOX spent fuel assemblies. The TN trademark 112 has to comply with the 1996 edition of the IAEA transport regulations and with the transport and handling equipment presently used for the transport of spent fuel in France. It can be loaded with 12 assemblies, either MOX or UO2. This paper gives a description of two packages belonging to the new generation: - one for fresh MOX fuel assemblies transportation, used on EDF's power plants for MOX reloads, - one for spent MOX fuel assemblies transportation, presently under development with a first transport planned for mid-2007.

The arison data acquisition and elaboration software package running on Hewlett-Packard minicomputer

International Nuclear Information System (INIS)

Diamantidis, Z.

1987-01-01

In this article, a data acquisition and elaboration system is described consisting in a PCM data acquisition and a spectrum analyser system and their data elaboration package for reactor safety or other general purposes. Measurements on temperature, noise fluctuations of temperature or other noise analysis dynamics, pressure, etc. Time series and their conversion in engineering units and their statistical and frequency analysis is provided

Proposition of innovative and safe design of grid plate for Tehran research reactor

International Nuclear Information System (INIS)

Jalali, H.R.; Fadaei, A.H.

2017-01-01

Highlights: • An innovative and safe design for grid plate in research reactors proposed. • New grid plate acts as an independent shutdown system. • Neutronic and transient calculation was done using MTR-PC package. • Calculations show that the performance and safety of new design are acceptable. - Abstract: The purpose of this paper is to propose an innovative and safe design of grid plate for Tehran research reactor (TRR) without any reduction in its performance in comparison with the current operation. The new grid plate consisted of two joined cubic with empty walls which are place of fuels and heavy water, respectively. The proposed design is such that the reactor core is divided into two distinct parts using the heavy water. The heavy water is inserted in the walls of the new grid plate. The new design of grid plate by keeping the characteristics of the previous version creates the possibility of shutting the reactor down in critical condition. In this paper, at initial step, a simulation of acceptable benchmark for Tehran research reactor is performed which could be considered reliable and comparable with SAR (Safety Analysis Report) data. In the next step, two different designs are proposed for grid plate and then are applied to reactor core using simulation tools. For the proposed design: core excess reactivity, shutdown margin, control rod worth, neutron flux and kinetic parameters are calculated. Furthermore, the transient analysis was performed for the new design to check the status of reactor safety. Obtained results show that all neutronic parameters for the first operating core and the new design are comparable, and there is no reduction in the efficiency of reference core. Moreover, in the current design, a diverse and independent shutdown system for TRR was included. Nuclear reactor analysis codes including MTR-PC package were employed to carry out these calculations.

MEMS packaging

CERN Document Server

Hsu , Tai-Ran

2004-01-01

MEMS Packaging discusses the prevalent practices and enabling techniques in assembly, packaging and testing of microelectromechanical systems (MEMS). The entire spectrum of assembly, packaging and testing of MEMS and microsystems, from essential enabling technologies to applications in key industries of life sciences, telecommunications and aerospace engineering is covered. Other topics included are bonding and sealing of microcomponents, process flow of MEMS and microsystems packaging, automated microassembly, and testing and design for testing.The Institution of Engineering and Technology is

Packaging design criteria for the Hanford Ecorok Packaging

International Nuclear Information System (INIS)

Mercado, M.S.

1996-01-01

The Hanford Ecorok Packaging (HEP) will be used to ship contaminated water purification filters from K Basins to the Central Waste Complex. This packaging design criteria documents the design of the HEP, its intended use, and the transportation safety criteria it is required to meet. This information will serve as a basis for the safety analysis report for packaging

CORRELATIONS BETWEEN HOMOLOGUE CONCENTRATIONS OF PCDD/FS AND TOXIC EQUIVALENCY VALUES IN LABORATORY-, PACKAGE BOILER-, AND FIELD-SCALE INCINERATORS

Science.gov (United States)

The toxic equivalency (TEQ) values of polychlorinated dibenzo-p-dioxins and polychlorinated dibenzofurans (PCDD/Fs) are predicted with a model based on the homologue concentrations measured from a laboratory-scale reactor (124 data points), a package boiler (61 data points), and ...

Packaging microservices

DEFF Research Database (Denmark)

Montesi, Fabrizio; Thrane, Dan Sebastian

2017-01-01

We describe a first proposal for a new packaging system for microservices based on the Jolie programming language, called the Jolie Package Manager (JPM). Its main features revolve around service interfaces, which make the functionalities that a service provides and depends on explicit. For the f......We describe a first proposal for a new packaging system for microservices based on the Jolie programming language, called the Jolie Package Manager (JPM). Its main features revolve around service interfaces, which make the functionalities that a service provides and depends on explicit...

TARMS, an on-line boiling water reactor operation management system. [3 D core simulator LOGOS 2

Energy Technology Data Exchange (ETDEWEB)

Iwamoto, T.; Sakurai, S.; Uematsu, H.; Tsuiki, M.; Makino, K.

1984-12-01

The TARMS (Toshiba Advanced Reactor Management System) software package was developed as an effective on-line, on-site tool for boiling water reactor core operation management. It was designed to support a complete function set to meet the requirement to the current on-line process computers. The functions can be divided into two categories. One is monitoring of the present core power distribution as well as related limiting parameters. The other is aiding site engineers or reactor operators in making the future reactor operating plan. TARMS performs these functions with a three-dimensional BWR core physics simulator LOGOS 2, which is based on modified one-group, coarse-mesh nodal diffusion theory. A method was developed to obtain highly accurate nodal powers by coupling LOGOS 2 calculations with the readings of an in-core neutron flux monitor. A sort of automated machine-learning method also was developed to minimize the errors caused by insufficiency of the physics model adopted in LOGOS 2. In addition to these fundamental calculational methods, a number of core operation planning aid packages were developed and installed in TARMS, which were designed to make the operator's inputs simple and easy.

Risk Management for Sodium Fast Reactors.

Energy Technology Data Exchange (ETDEWEB)

Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Groth, Katrina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cardoni, Jeffrey N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-01-01

Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.

An updated status of Department of Energy safety reviews of packages for transporting radioactive material

International Nuclear Information System (INIS)

Kapoor, A.

1995-01-01

The Department of Energy conducts conformance reviews and issues Certificates of Compliance for Type B packaging for radioactive materials. Several offices within DOE perform these reviews which are required by the Department of Transportation to be to the regulations promulgated by the Nuclear Regulatory Commission or their safety equivalent. This paper focuses on one of these offices, the Office of Facility Safety Analysis, EH-32, which is responsible for reviewing and certifying packages other than those used for weapons and weapons component, for Naval Reactors, and for Civilian Radioactive Waste Management. This paper gives the background and organizational history of EH-32, discusses the version of regulations to which the packaging is reviewed, updates the status of these reviews, describes the effectiveness of the reviews, updates the training courses sponsored by EH-32, and mentions the new Quality Assurance Evaluations being started by EH-32

Vertical Drop Of 21-PWR Waste Package On Unyielding Surface

International Nuclear Information System (INIS)

S. Mastilovic; A. Scheider; S.M. Bennett

2001-01-01

The objective of this calculation is to determine the structural response of a 21-PWR (pressurized-water reactor) Waste Package (WP) subjected to the 2-m vertical drop on an unyielding surface at three different temperatures. The scope of this calculation is limited to reporting the calculation results in terms of stress intensities in two different WP components. The information provided by the sketches (Attachment I) is that of the potential design of the type of WP considered in this calculation, and all obtained results are valid for that design only

PRIDE Surveillance Projects Data Packaging Project, Information Package Specification Version 1.0

Energy Technology Data Exchange (ETDEWEB)

Kelleher, D.M.; Shipp, R. L.; Mason, J. D.

2009-09-28

This document contains a specification for a standard XML document format called an information package that can be used to store information and the context required to understand and use that information in information management systems and other types of information archives. An information package consists of packaged information, a set of information metadata that describes the packaged information, and an XML signature that protects the packaged information. The information package described in this specification was designed to be used to store Department of Energy (DOE) and National Nuclear Security Administration (NNSA) information and includes the metadata required for that information: a unique package identifier, information marking that conforms to DOE and NNSA requirements, and access control metadata. Information package metadata can also include information search terms, package history, and notes. Packaged information can be text content, binary content, and the contents of files and other containers. A single information package can contain multiple types of information. All content not in a text form compatible with XML must be in a text encoding such as base64. Package information is protected by a digital XML signature that can be used to determine whether the information has changed since it was signed and to identify the source of the information. This specification has been tested but has not been used to create production information packages. The authors expect that gaps and unclear requirements in this specification will be identified as this specification is used to create information packages and as information stored in information packages is used. The authors expect to issue revised versions of this specification as needed to address these issues.

PR-EDB: Power Reactor Embrittlement Database Version 3

International Nuclear Information System (INIS)

Wang, Jy-An John; Subramani, Ranjit

2008-01-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. 'User-friendly' utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

PR-EDB: Power Reactor Embrittlement Database - Version 3

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [ORNL; Subramani, Ranjit [ORNL

2008-03-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

Actinide behavior in the integral fast reactor

International Nuclear Information System (INIS)

Courtney, J.C.

1993-05-01

Goal of this project is to determine the consumption of Np-237, Pu-240, Am-241, and Am-243 in the Integral Fast Reactor (IFR) fuel cycle. These four actinides set the long term waste management criteria for spent nuclear fuel; if it can be demonstrated that they can be efficiently consumed in the IFR, then requirements for nuclear waste repositories can be much less demanding. Irradiations in the Experimental Breeder Reactor II (EBR-II) at Argonne National Laboratory's site near Idaho Falls, Idaho, will be conducted to determine fission and transmutation rates for the four nuclides. The experimental effort involves target package design, fabrication, quality assurance, and irradiation. Post irradiation analyses are required to determine the fission rates and neutron spectra in the EBR-II core

Replacement Energy Cost Analysis Package (RECAP): User's guide. Revision 1

International Nuclear Information System (INIS)

VanKuiken, J.C.; Willing, D.L.

1994-07-01

A microcomputer program called the Replacement Energy Cost Analysis Package (RECAP) has been developed to assist the US Nuclear Regulatory Commission (NRC) in determining the replacement energy costs associated with short-term shutdowns or deratings of one or more nuclear reactors. The calculations are based on the seasonal, unit-specific cost estimates for 1993--1996 previously published in NRC Report NUREG/CR--4012, Vol. 3 (1992), for all 112 US reactors. Because the RECAP program is menu-driven, the user can define specific case studies in terms of such parameters as the units to be included, the length and timing of the shutdown or derating period, the unit capacity factors, and the reference year for reporting cost results. In addition to simultaneous shutdown cases, more complicated situations, such as overlapping shutdown periods or shutdowns that occur in different years, can be examined through the use of a present-worth calculation option

3D simulation of CANDU reactor regulating system

International Nuclear Information System (INIS)

Venescu, B.; Zevedei, D.; Jurian, M.

2013-01-01

Present paper shows the evaluation of the performance of the 3-D modal synthesis based reactor kinetic model in a closed-loop environment in a MATLAB/SIMULINK based Reactor Regulating System (RRS) simulation platform. A notable advantage of the 3-D model is the level of details that it can reveal as compared to the coupled point kinetic model. Using the developed RRS simulation platform, the reactor internal behaviours can be revealed during load-following tests. The test results are also benchmarked against measurements from an existing (CANDU) power plant. It can be concluded that the 3-D reactor model produces more realistic view of the core neutron flux distribution, which is closer to the real plant measurements than that from a coupled point kinetic model. It is also shown that, through a vectorization process, the computational load of the 3-D model is comparable with that of the 14-zone coupled point kinetic model. Furthermore, the developed Graphical User Interface (GUI) software package for RRS implementation represents a user friendly and independent application environment for education training and industrial utilizations. (authors)

Simulation of the behaviour of small and medium nuclear reactors on PCs

International Nuclear Information System (INIS)

Lyon, R.B.

1999-01-01

The International Atomic Energy Agency (IAEA) has established a programme in nuclear reactor simulation computer programs to assist its Member States in education and training. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and training material, sponsors training courses and workshops, and distributes documentation and computer programs. One of the simulation programs distributed by the IAEA is the the Advanced Reactor Simulator which simulates the behaviour of BWR, PWR and HWR reactor types. For this package, the modeling approach and assumptions are broadly described, together with a general description of the operation of the computer program. (author)

Advancements in meat packaging.

Science.gov (United States)

McMillin, Kenneth W

2017-10-01

Packaging of meat provides the same or similar benefits for raw chilled and processed meats as other types of food packaging. Although air-permeable packaging is most prevalent for raw chilled red meat, vacuum and modified atmosphere packaging offer longer shelf life. The major advancements in meat packaging have been in the widely used plastic polymers while biobased materials and their integration into composite packaging are receiving much attention for functionality and sustainability. At this time, active and intelligent packaging are not widely used for antioxidant, antimicrobial, and other functions to stabilize and enhance meat properties although many options are being developed and investigated. The advances being made in nanotechnology will be incorporated into food packaging and presumably into meat packaging when appropriate and useful. Intelligent packaging using sensors for transmission of desired information and prompting of subsequent changes in packaging materials, environments or the products to maintain safety and quality are still in developmental stages. Copyright © 2017 Elsevier Ltd. All rights reserved.

Westinghouse experience in using mechanical cutting for reactor vessel internals segmentation

International Nuclear Information System (INIS)

Boucau, Joseph; Fallstroem, Stefan; Segerud, Per; Kreitman, Paul J.

2010-01-01

Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques. Mechanical cutting has been used by Westinghouse since 1999 for both PWRs and BWRs and its process has been continuously improved over the years. Detailed planning is essential to a successful project, and typically a 'Segmentation and Packaging Plan' is prepared to document the effort. The usual method is to start at the end of the process, by evaluating the waste disposal requirements imposed by the waste disposal agency, what type and size of containers are available for the different disposal options, and working backwards to select the best cutting tools and finally the cut geometry required. These plans are made utilizing advanced 3-D CAD software to model the process. Another area where the modelling has proven invaluable is in determining the logistics of component placement and movement in the reactor cavity, which is typically very congested when all the internals are out of the reactor vessel in various stages of segmentation. The main objective of the segmentation and packaging plan is to determine the strategy for separating the highly activated components from the less activated material, so that they can be disposed of in the most cost effective manner. Usually, highly activated components cannot be shipped off-site, so they must be packaged such that they can be dry stored with the spent fuel in an Independent Spent Fuel Storage Installation (ISFSI). Less activated components can be shipped to an off-site disposal site depending on space availability. Several of the

Color in packaging design : Case: ZheJiang JinSheng packaging Co,Ltd

OpenAIRE

Hu, Cuicui

2010-01-01

Color occupies an important position in packaging design, with the improvement of living standard, the higher requirement of color design in packaging. The aim of this thesis was to discuss key issues concerning aesthetics of packaging design. Topics will include an overview of the packaging design, the influence factor of packaging design, and introduce the aesthetics from packaging aspect. This thesis will also identify common problems of the production process, and list the phases of ho...

Packaging for Sustainability

CERN Document Server

Lewis, Helen; Fitzpatrick, Leanne

2012-01-01

The packaging industry is under pressure from regulators, customers and other stakeholders to improve packaging’s sustainability by reducing its environmental and societal impacts. This is a considerable challenge because of the complex interactions between products and their packaging, and the many roles that packaging plays in the supply chain. Packaging for Sustainability is a concise and readable handbook for practitioners who are trying to implement sustainability strategies for packaging. Industry case studies are used throughout the book to illustrate possible applications and scenarios. Packaging for Sustainability draws on the expertise of researchers and industry practitioners to provide information on business benefits, environmental issues and priorities, environmental evaluation tools, design for environment, marketing strategies, and challenges for the future.

CDIAC catalog of numeric data packages and computer model packages

International Nuclear Information System (INIS)

Boden, T.A.; Stoss, F.W.

1993-05-01

The Carbon Dioxide Information Analysis Center acquires, quality-assures, and distributes to the scientific community numeric data packages (NDPs) and computer model packages (CMPs) dealing with topics related to atmospheric trace-gas concentrations and global climate change. These packages include data on historic and present atmospheric CO 2 and CH 4 concentrations, historic and present oceanic CO 2 concentrations, historic weather and climate around the world, sea-level rise, storm occurrences, volcanic dust in the atmosphere, sources of atmospheric CO 2 , plants' response to elevated CO 2 levels, sunspot occurrences, and many other indicators of, contributors to, or components of climate change. This catalog describes the packages presently offered by CDIAC, reviews the processes used by CDIAC to assure the quality of the data contained in these packages, notes the media on which each package is available, describes the documentation that accompanies each package, and provides ordering information. Numeric data are available in the printed NDPs and CMPs, in CD-ROM format, and from an anonymous FTP area via Internet. All CDIAC information products are available at no cost

Burn-Up Calculation of the Fuel Element in RSG-GAS Reactor using Program Package BATAN-FUEL

International Nuclear Information System (INIS)

Mochamad Imron; Ariyawan Sunardi

2012-01-01

Calculation of burn lip distribution of 2.96 gr U/cc Silicide fuel element at the 78 th reactor cycle using computer code program of BATAN-FUEL has been done. This calculation uses inputs such as generated power, operation time and a core assumption model of 5/1. Using this calculation model burn up for the entire fuel elements at the reactor core are able to be calculated. From the calculation it is obtained that the minimum burn up of 6.82% is RI-50 at the position of A-9, while the maximum burn up of 57.57% is RI 467 at the position of 8-7. Based on the safety criteria as specified in the Safety Analysis Report (SAR) RSG-GAS reactor, the maximum fuel burn up allowed is 59.59%. It then can be concluded that pattern that elements placement at the reactor core are properly and optimally done. (author)

Safety analysis report for packaging (onsite) contaminated well cars

International Nuclear Information System (INIS)

Mercado, J.E.

1998-01-01

In support of past operations, railcars were used to ship irradiated fuel from the 100 Area fuel storage basins to the Plutonium Uranium Extraction (PUREX) Facility. There are two configurations for the packaging systems that transported the fuel: the Three-Well Cask Car, which is outfitted with three casks, and the taller, single well, New Production Reactor (NPR) Cask Car. In this document, these cask cars are referred to collectively as well cars. The purpose of this document is to evaluate and authorize the onsite transportation of well cars that contain significant levels of contamination. No irradiated fuel will be transported in the well cars. Neutron detection data confirmed that the well cars do not contain fuel. The intention is to move 14 retired well cars from their current locations in the 100 Area to a suitable storage location in the 200 Area. Each well car contains Type B quantities of radioactivity; so that the hazard of the transport operation is relatively low. This safety analysis report for packaging (SARP) provides the analyses and evaluations necessary to demonstrate that the contaminated well cars meet the transportation safety requirements of HNF-PRO-154, Responsibilities and Procedures for Hazardous Material Shipments for an onsite packaging. The scope of this document addresses the preparation and transportation of the contaminated well cars

Radiation protection at the RA Reactor in 1985, Part -4, decontamination and treatment of solid radioactive materials for the needs of RA reactor

International Nuclear Information System (INIS)

Plecas, I.; Vukovic, Z.; Blagojevic, R.; Kostadinovic, A.

1985-01-01

This report describes the activity of the decontamination and treatment team for the needs of the RA reactor, its equipment, working conditions, methods for decontamination, means of decontamination, type and quantity of decontaminated surfaces, number of decontaminated objects, quantity of collected radioactive solid wastes, their packaging, transport to the storage place and topography od radiation field in the storage during 1985 [sr

Properties of light water reactor spent fuel cladding. Interim report

International Nuclear Information System (INIS)

Farwick, D.G.; Moen, R.A.

1979-08-01

The Commercial Waste and Spent Fuel Packaging Program will provide containment packages for the safe storage or disposal of spent Light Water Reactor (LWR) fuel. Maintaining containment of radionuclides during transportation, handling, processing and storage is essential, so the best understanding of the properties of the materials to be stored is necessary. This report provides data collection, assessment and recommendations for spent LWR fuel cladding materials properties. Major emphasis is placed on mechanical properties of the zircaloys and austenitic stainless steels. Limited information on elastic constants, physical properties, and anticipated corrosion behavior is also provided. Work is in progress to revise these evaluations as the program proceeds

A Restructuring of the CAV and FDI Package for the MIDAS Computer Code

International Nuclear Information System (INIS)

Park, S. H.; Kim, D. H.; Cho, S. W.

2006-01-01

As one of the processes for a localized severe accident analysis code, KAERI is developing a severe accident code MIDAS. The MIDAS code is being developed based on MELCOR. The existing data saving method of MELCOR uses pointer variables for a fix-sized storage management, and it deteriorates the readability, maintainability and portability of the code. As a most important process for a localized severe accident analysis code, it is needed convenient method for data handling. So, it has been used the new features in FORTRAN90 such as a dynamic allocation for the restructuring. The restructuring of the data saving and transferring method of the existing code makes it easy to understand the code. Before an entire restructuring of the code, a restructuring template for a simple package was developed and tested. The target for the restructuring in this paper was the CAV and FDI packages. The CAV(cavity) package is responsible for modeling the attack on the basement concrete by hot core materials. The FDI(Fuel Dispersal Interactions) package is responsible for modeling both low and high pressure molten fuel ejection from the RPV into the reactor cavity, control volumes and surfaces. The verification was done through comparing the results before and after the restructuring

Waste package performance assessment

International Nuclear Information System (INIS)

Lester, D.H.

1981-01-01

This paper describes work undertaken to assess the life-expectancy and post-failure nuclide release behavior of high-level and waste packages in a geologic repository. The work involved integrating models of individual phenomena (such as heat transfer, corrosion, package deformation, and nuclide transport) and using existing data to make estimates of post-emplacement behavior of waste packages. A package performance assessment code was developed to predict time to package failure in a flooded repository and subsequent transport of nuclides out of the leaking package. The model has been used to evaluate preliminary package designs. The results indicate, that within the limitation of model assumptions and data base, packages lasting a few hundreds of years could be developed. Very long lived packages may be possible but more comprehensive data are needed to confirm this

Edible packaging materials.

Science.gov (United States)

Janjarasskul, Theeranun; Krochta, John M

2010-01-01

Research groups and the food and pharmaceutical industries recognize edible packaging as a useful alternative or addition to conventional packaging to reduce waste and to create novel applications for improving product stability, quality, safety, variety, and convenience for consumers. Recent studies have explored the ability of biopolymer-based food packaging materials to carry and control-release active compounds. As diverse edible packaging materials derived from various by-products or waste from food industry are being developed, the dry thermoplastic process is advancing rapidly as a feasible commercial edible packaging manufacturing process. The employment of nanocomposite concepts to edible packaging materials promises to improve barrier and mechanical properties and facilitate effective incorporation of bioactive ingredients and other designed functions. In addition to the need for a more fundamental understanding to enable design to desired specifications, edible packaging has to overcome challenges such as regulatory requirements, consumer acceptance, and scaling-up research concepts to commercial applications.

Testing of a transport cask for research reactor spent fuel

International Nuclear Information System (INIS)

Mourao, Rogerio P.; Silva, Luiz Leite da; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

2011-01-01

Since the beginning of the last decade three Latin American countries which operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the reactors operated in the region. As a step in this direction, a packaging for the transport of irradiated fuel from research reactors was designed by a tri-national team and a half-scale model for MTR fuel constructed in Argentina and tested in Brazil. Two test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. Although the specimen has not successfully performed the tests, its overall performance was considered very satisfactory, and improvements are being introduced to the design. A third test sequence is planned for 2011. (author)

THE PHASE REACTOR INDUCTANCE SELECTION TECHNIQUE FOR POWER ACTIVE FILTER

Directory of Open Access Journals (Sweden)

D. V. Tugay

2016-12-01

Full Text Available Purpose. The goal is to develop technique of the phase inductance power reactors selection for parallel active filter based on the account both low-frequency and high-frequency components of the electromagnetic processes in a power circuit. Methodology. We have applied concepts of the electrical circuits theory, vector analysis, mathematical simulation in Matlab package. Results. We have developed a new technique of the phase reactors inductance selection for parallel power active filter. It allows us to obtain the smallest possible value of THD network current. Originality. We have increased accuracy of methods of the phase reactor inductance selection for power active filter. Practical value. The proposed technique can be used in the design and manufacture of the active power filter for real objects of energy supply.

Criticality evaluation of BWR MOX fuel transport packages using average Pu content

International Nuclear Information System (INIS)

Mattera, C.; Martinotti, B.

2004-01-01

Currently in France, criticality studies in transport configurations for Boiling Water Reactor Mixed Oxide fuel assemblies are based on conservative hypothesis assuming that all rods (Mixed Oxide (Uranium and Plutonium), Uranium Oxide, Uranium and Gadolinium Oxide rods) are Mixed Oxide rods with the same Plutonium-content, corresponding to the maximum value. In that way, the real heterogeneous mapping of the assembly is masked and covered by a homogeneous Plutonium-content assembly, enriched at the maximum value. As this calculation hypothesis is extremely conservative, COGEMA LOGISTICS has studied a new calculation method based on the average Plutonium-content in the criticality studies. The use of the average Plutonium-content instead of the real Plutonium-content profiles provides a highest reactivity value that makes it globally conservative. This method can be applied for all Boiling Water Reactor Mixed Oxide complete fuel assemblies of type 8 x 8, 9 x 9 and 10 x 10 which Plutonium-content in mass weight does not exceed 15%; it provides advantages which are discussed in our approach. With this new method, for the same package reactivity, the Pu-content allowed in the package design approval can be higher. The COGEMA LOGISTICS' new method allows, at the design stage, to optimise the basket, materials or geometry for higher payload, keeping the same reactivity

Active Packaging Coatings

Directory of Open Access Journals (Sweden)

Luis J. Bastarrachea

2015-11-01

Full Text Available Active food packaging involves the packaging of foods with materials that provide an enhanced functionality, such as antimicrobial, antioxidant or biocatalytic functions. This can be achieved through the incorporation of active compounds into the matrix of the commonly used packaging materials, or by the application of coatings with the corresponding functionality through surface modification. The latter option offers the advantage of preserving the packaging materials’ bulk properties nearly intact. Herein, different coating technologies like embedding for controlled release, immobilization, layer-by-layer deposition, and photografting are explained and their potential application for active food packaging is explored and discussed.

Study on the Neutrino Oscillation with a Next Generation Medium-Baseline Reactor Experiment

International Nuclear Information System (INIS)

Joo, Kyung Kwang; Shin, Chang Dong

2014-01-01

For over fifty years, reactor experiments have played an important role in neutrino physics, in both discoveries and precision measurements. One of the methods to verify the existence of neutrino is the observation of neutrino oscillation phenomena. Electron antineutrinos emitted from a reactor provide the measurement of the small mixing angle θ 13 , providing rich programs of neutrino properties, detector development, nuclear monitoring, and application. Using reactor neutrinos, future reactor neutrino experiments, more precise measurements of θ 12 ,Δm 12 2 , and mass hierarchy will be explored. The precise measurement of θ 13 would be crucial for measuring the CP violation parameters at accelerators. Therefore, reactor neutrino physics will assist in the complete understanding of the fundamental nature and implications of neutrino masses and mixing. In this paper, we investigated several characteristics of RENO-50, which is a future medium-baseline reactor neutrino oscillation experiment, by using the GloBES simulation package

Safety Analysis Report for packaging (onsite) steel waste package

International Nuclear Information System (INIS)

BOEHNKE, W.M.

2000-01-01

The steel waste package is used primarily for the shipment of remote-handled radioactive waste from the 324 Building to the 200 Area for interim storage. The steel waste package is authorized for shipment of transuranic isotopes. The maximum allowable radioactive material that is authorized is 500,000 Ci. This exceeds the highway route controlled quantity (3,000 A 2 s) and is a type B packaging

Pumps for German pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Dernedde, R.

1984-01-01

The article describes the development of a selection of pumps which are used in the primary coolant system and the high-pressure safety injection system and feed water system during the past 2 decades. The modifications were caused by the step-wise increasing power output of the plants from 300 MW up to 1300 MW. Additional important influences were given be the increased requirements for quality assurance and final-documentation. The good operating results of the delivered pumps proved that the reliability is independent of the volume of the software-package. The outlook expects that consolidation will be followed by additional steps for the order processing of components for the convoy pumps. KW: main coolant pump; primary system; boiler feed pump; reactor pump; secondary system; barrel insert pump; pressure water reactor; convoy pump; state of the art.

Overview of the RPV-2 and INTERN-1 packages: From primary damage to microplasticity

International Nuclear Information System (INIS)

Adjanor, G.; Bugat, S.; Domain, C.; Barbu, A.

2010-01-01

In the framework of the European project PERFECT, four multiscale simulation packages dedicated to the prediction of evolution of material properties were developed. Among them, the RPV-2 and INTERN-1 are two simulation sequences of similar structure dealing with radiation damage in the reactor pressure vessel and the reactor internal structures, respectively. Both start at the atomic scale, where the neutron spectrum of the specified reactor is used to determine the energy distribution of the primary knocked-on atoms (PKA). A database of molecular dynamics results is then used to integrate the instantaneous production of defect clusters resulting from the displacement cascades initiated by each PKA. Depending on the type of calculation chosen to model long-term diffusion and reactions of defect clusters, precipitates and mixed-clusters, this primary damage enters either in rate equations or in Object Kinetic Monte Carlo simulations. The later correspond to a more accurate (but also more computationally demanding) physical model for diffusion as positions of objects on a lattice are explicitly treated. Finally, the increase of critical resolved shear stress is estimated from these cluster distributions either using an analytical model, taking into account the self and mutual dipole interactions of dislocations pinned on randomly dispersed unshearable obstacles, or by simulating the glide of a single dislocation line in its main slip system. Dislocation dynamics simulations were already used to validate some of the assumptions of the latter models, and will be fully integrated in the next versions of the packages.

PRIDE Surveillance Projects Data Packaging Project Information Package Specification Version 1.1

Energy Technology Data Exchange (ETDEWEB)

Kelleher, D. M.; Shipp, R. L.; Mason, J. D.

2010-08-31

Information Package Specification version 1.1 describes an XML document format called an information package that can be used to store information in information management systems and other information archives. An information package consists of package information, the context required to understand and use that information, package metadata that describes the information, and XML signatures that protect the information. The information package described in this specification was designed to store Department of Energy (DOE) and National Nuclear Security Administration (NNSA) information and includes the metadata required for that information: a unique package identifier, information marking that conforms to DOE and NNSA requirements, and access control metadata. It is an implementation of the Open Archival Information System (OAIS) Reference Model archival information package tailored to meet NNSA information storage requirements and designed to be used in the computing environments at the Y-12 National Security Complex and at other NNSA sites.

The integral fast reactor fuels reprocessing laboratory at Argonne National Laboratory, Illinois

International Nuclear Information System (INIS)

Wolson, R.D.; Tomczuk, Z.; Fischer, D.F.; Slawecki, M.A.; Miller, W.E.

1986-09-01

The processing of Integral Fast Reactor (IFR) metal fuel utilizes pyrochemical fuel reprocessing steps. These steps include separation of the fission products from uranium and plutonium by electrorefining in a fused salt, subsequent concentration of uranium and plutonium for reuse, removal, concentration, and packaging of the waste material. Approximately two years ago a facility became operational at Argonne National Laboratory-Illinois to establish the chemical feasibility of proposed reprocessing and consolidation processes. Sensitivity of the pyroprocessing melts to air oxidation necessitated operation in atmosphere-controlled enclosures. The Integral Fast Reactor Fuels Reprocessing Laboratory is described

Specification of the Advanced Burner Test Reactor Multi-Physics Coupling Demonstration Problem

Energy Technology Data Exchange (ETDEWEB)

Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Grudzinski, J. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. Q. [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-12-21

This document specifies the multi-physics nuclear reactor demonstration problem using the SHARP software package developed by NEAMS. The SHARP toolset simulates the key coupled physics phenomena inside a nuclear reactor. The PROTEUS neutronics code models the neutron transport within the system, the Nek5000 computational fluid dynamics code models the fluid flow and heat transfer, and the DIABLO structural mechanics code models structural and mechanical deformation. The three codes are coupled to the MOAB mesh framework which allows feedback from neutronics, fluid mechanics, and mechanical deformation in a compatible format.

HPLOT: the graphics interface package for the HBOOK histogramming package

International Nuclear Information System (INIS)

Watkins, H.

1978-01-01

The subroutine package HPLOT described in this report, enables the CERN histogramming package HBOOK to produce high-quality pictures by means of high-resolution devices such as plotters. HPLOT can be implemented on any scientific computing system with a Fortran IV compiler and can be interfaced with any graphics package; spectral routines in addition to the basic ones enable users to embellish their histograms. Examples are also given of the use of HPLOT as a graphics package for plotting simple pictures without histograms. (Auth.)

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-06-15

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

Qualification testing facility for packages to be used for transport and storage of radioactive materials

International Nuclear Information System (INIS)

Vieru, Gheorghe

2009-01-01

The radioactive materials (RAM) packaging have to comply to all modes and transport condition, routine or in accident conditions possibly to occur during transportation operations. It is well known that the safety in the transport of RAM is dependent on packaging appropriate for the contents being shipped rather than on operational and/or administrative actions required for the package. The quality of these packages - type A, B or C has to be proved by performing qualification tests in accordance with the ROMANIAN nuclear regulation conditions provided by CNCAN Order no. 357/22.12.2005- 'Norms for a Safe Transport of Radioactive Material', the IAEA Vienna Recommendation stipulated in the Safety standard TS-R-1- Regulation for the Safe Transport of Radioactive Material, 2005 Edition, and other applicable international recommendations. The paper will describe the components of the designed testing facilities, and the qualification testing to be performed for all type A, B and C packages subjected to the testing. In addition, a part of the qualification tests for a package (designed and manufactured in INR Pitesti) used for transport and storage of spent fuel LEU elements of a TRIGA nuclear reactor will be described and analyzed. Quality assurance and quality controls measures taken in order to meet technical specification provided by the design are also presented and commented. The paper concludes that the new Romanian Testing Facilities for RAM packages will comply with the national safe standards as well as with the IAEA applicable recommendation provided by the TS-R-1 safety standard. (author)

Package

Directory of Open Access Journals (Sweden)

Arsić Zoran

2013-01-01

Full Text Available It is duty of the seller to pack the goods in a manner which assures their safe arrival and enables their handling in transit and at the place of destination. The problem of packing is relevant in two main respects. First of all the buyer is in certain circumstances entitled to refuse acceptance of the goods if they are not properly packed. Second, the package is relevant to calculation of price and freight based on weight. In the case of export trade, the package should conform to the legislation in the country of destination. The impact of package on environment is regulated by environment protection regulation of Republic if Serbia.

Packaging-radiation disinfestation relationships

International Nuclear Information System (INIS)

Highland, H.A.

1985-01-01

Foods that are susceptible to insect infestation can be irradiated to destroy the infestation; however, the food must be kept essentially insect-free until consumed, or it must be disinfested again, perhaps repeatedly. Insect-resistant packages can be used to prevent reinfestation, but there are certain requirements that must be fulfilled before a package can be made insect resistant. These include the use of insect-light construction and packaging materials that resist boring insects. The relative insect resistance of various packages and packaging materials is discussed, as are behavior traits such as egressive boring that enables insects to escape from packages and the ability of insects to climb on various packaging materials. Some successful and unsuccessful attempts to make various types of packages insect resistant are discussed, as are factors that must be considered in the selection or development of insect-resistant packages for radiation disinfested foods. The latter factors include biological and physical environments, length of storage periods, stresses on packages during shipment, types of storage facilities, governmental regulations, health requirements, and others

Packaging Printing Today

Directory of Open Access Journals (Sweden)

Stanislav Bolanča

2015-12-01

Full Text Available Printing packaging covers today about 50% of all the printing products. Among the printing products there are printing on labels, printing on flexible packaging, printing on folding boxes, printing on the boxes of corrugated board, printing on glass packaging, synthetic and metal ones. The mentioned packaging are printed in flexo printing technique, offset printing technique, intaglio halftone process, silk – screen printing, ink ball printing, digital printing and hybrid printing process. The possibilities of particular printing techniques for optimal production of the determined packaging were studied in the paper. The problem was viewed from the technological and economical aspect. The possible printing quality and the time necessary for the printing realization were taken as key parameters. An important segment of the production and the way of life is alocation value and it had also found its place in this paper. The events in the field of packaging printing in the whole world were analyzed. The trends of technique developments and the printing technology for packaging printing in near future were also discussed.

Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

International Nuclear Information System (INIS)

Mourao, Rogerio P.; Leite da Silva, Luiz; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

2013-01-01

Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

Thermal simulations and tests in the development of a helmet transport spent fuel elements Research Reactor

International Nuclear Information System (INIS)

Saliba, R.; Quintana, F.; Márquez Turiello, R.; Furnari, J.C.; Pimenta Mourão, R.

2013-01-01

A packaging for the transport of irradiated fuel from research reactors was designed by a group of researchers to improve the capability in the management of spent fuel elements from the reactors operated in the region. Two half-scale models for MTR fuel were constructed and tested so far and a third one for both MTR and TRIGA fuels will be constructed and tested next. Four test campaigns have been carried out, covering both normal and hypothetical accident conditions of transportation. The thermal test is part of the requirements for the qualification of transportation packages for nuclear reactors spent fuel elements. In this paper both the numerical modelling and experimental thermal tests performed are presented and discussed. The cask is briefly described as well as the finite element model developed and the main adopted hypotheses for the thermal phenomena. The results of both numerical runs and experimental tests are discussed as a tool to validate the thermal modelling. The impact limiters, attached to the cask for protection, were not modelled. (author) [es

Packaged die heater

Science.gov (United States)

Spielberger, Richard; Ohme, Bruce Walker; Jensen, Ronald J.

2011-06-21

A heater for heating packaged die for burn-in and heat testing is described. The heater may be a ceramic-type heater with a metal filament. The heater may be incorporated into the integrated circuit package as an additional ceramic layer of the package, or may be an external heater placed in contact with the package to heat the die. Many different types of integrated circuit packages may be accommodated. The method provides increased energy efficiency for heating the die while reducing temperature stresses on testing equipment. The method allows the use of multiple heaters to heat die to different temperatures. Faulty die may be heated to weaken die attach material to facilitate removal of the die. The heater filament or a separate temperature thermistor located in the package may be used to accurately measure die temperature.

Safety Analysis Report for packaging (onsite) steel waste package

Energy Technology Data Exchange (ETDEWEB)

BOEHNKE, W.M.

2000-07-13

The steel waste package is used primarily for the shipment of remote-handled radioactive waste from the 324 Building to the 200 Area for interim storage. The steel waste package is authorized for shipment of transuranic isotopes. The maximum allowable radioactive material that is authorized is 500,000 Ci. This exceeds the highway route controlled quantity (3,000 A{sub 2}s) and is a type B packaging.

Internal Corrosion Analysis of Model 9975 Packaging Containing Pu or PuO2 During Shipping and Storage

International Nuclear Information System (INIS)

Vormelker, P.

1999-01-01

The Materials Consultation Group of SRTC has completed an internal corrosion analysis of the Model 9975 packaging assembly containing either Pu or PuO2 for storage in K Reactor under ambient conditions for a period of 12 years. The 12-year storage period includes two years for shipping and up to ten years for storage

Decontamination and decommissioning the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

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

1994-01-01

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

Packaging Printing Today

OpenAIRE

Stanislav Bolanča; Igor Majnarić; Kristijan Golubović

2015-01-01

Printing packaging covers today about 50% of all the printing products. Among the printing products there are printing on labels, printing on flexible packaging, printing on folding boxes, printing on the boxes of corrugated board, printing on glass packaging, synthetic and metal ones. The mentioned packaging are printed in flexo printing technique, offset printing technique, intaglio halftone process, silk – screen printing, ink ball printing, digital printing and hybrid printing process. T...

k-eff of the Bn-350 reactor fuel by transportation

International Nuclear Information System (INIS)

Lado, A. V.; Romanenko, O. G.; Tazhibaeva, I. L.

2001-01-01

There is packaging of nuclear fuel on the BN-350 fast breeder reactor, Actau, now. The analysis of criticality while this procedure was done in the Safety Analysis Report . Keeping in mind the planning displacement of the fuel to a site of long-term storage, the criticality assessment of the fuel packed into transportation cask carried out in this paper

EMERIS: an advanced information system for a materials testing reactor

International Nuclear Information System (INIS)

Adorjan, F.; Buerger, L.; Lux, I.; Mesko, L.; Szabo, K.; Vegh, J.; Ivanov, V.V.; Mozhaev, A.A.; Yakovlev, V.V.

1990-06-01

The basic features of the Materials Testing Reactor of IAE, Moscow (MR) Information System (EMERIS) are outlined. The purpose of the system is to support reactor and experimental test loop operators by a flexible, fully computerized and user-friendly tool for the aquisition, analysis, archivation and presentation of data obtained during operation of the experimental facility. High availability of EMERIS services is ensured by redundant hardware and software components, and by automatic configuration procedure. A novel software feature of the system is the automatic Disturbance Analysis package, which is aimed to discover primary causes of irregularities occurred in the technology. (author) 2 refs.; 2 figs

Updated neutron spectrum characterization of SNL baseline reactor environments

International Nuclear Information System (INIS)

Griffin, P.J.; Kelly, J.G.; Vehar, D.W.

1994-04-01

This document provides SAND-II and MANIPULATE output listings from calculations used to derive the new spectrum-averaged integral parameters that were reported in volume 1. When used in conjunction with volume 1, this document provides an audit trail for the neutron radiation field characterization and supports current quality assurance initiatives. This document provides detailed information on the neutron spectrum characteristics of the primary Sandia National Laboratories' (SNL) reactor environments. The information in this volume is not intended for the casual user of the SNL reactor facilities. This detailed characterization of the neutron and gamma environments at the Sandia Pulsed Reactor (SPR) and the Annular Core Research Reactor (ACRR) is provided to aid the users who wish to convert the information given in the Radiation Metrology Laboratory (RML) dosimetry reports into other (non-silicon) measures of neutron damage. The spectra provided in these appendices can be used as a source term for Monte Carlo radiation transport calculations to study the impact of experimenter's test package on the neutron environment

Advance Liquid Metal Reactor Discrete Dynamic Event Tree/Bayesian Network Analysis and Incident Management Guidelines (Risk Management for Sodium Fast Reactors)

Energy Technology Data Exchange (ETDEWEB)

Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Groth, Katrina M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cardoni, Jeffrey N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-04-01

Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self-correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the system's design to manage the accident. Inherently and passively safe designs are laudable, but nonetheless extreme boundary conditions can interfere with the design attributes which facilitate inherent safety, thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayesian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The authors would like to acknowledge the U.S. Department of Energy's Office of Nuclear Energy for funding this research through Work Package SR-14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at Argonne National Laboratory, Oak Ridge National Laboratory, and Idaho National Laboratory for their continue d contributions to the advanced reactor PRA mission area.

Summary Describing Integration of ERM Methodology into Supervisory Control Framework with Software Package Documentation; Advanced Reactor Technology Milestone: M4AT-16PN2301052

Energy Technology Data Exchange (ETDEWEB)

Ramuhalli, Pradeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hirt, Evelyn H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Dib, Gerges [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Veeramany, Arun [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bonebrake, Christopher A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Roy, Surajit [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2016-09-20

This project involved the development of enhanced risk monitors (ERMs) for active components in Advanced Reactor (AdvRx) designs by integrating real-time information about equipment condition with risk monitors. Health monitoring techniques in combination with predictive estimates of component failure based on condition and risk monitors can serve to indicate the risk posed by continued operation in the presence of detected degradation. This combination of predictive health monitoring based on equipment condition assessment and risk monitors can also enable optimization of maintenance scheduling with respect to the economics of plant operation. This report summarizes PNNL’s multi-year project on the development and evaluation of an ERM concept for active components while highlighting FY2016 accomplishments. Specifically, this report provides a status summary of the integration and demonstration of the prototypic ERM framework with the plant supervisory control algorithms being developed at Oak Ridge National Laboratory (ORNL), and describes additional case studies conducted to assess sensitivity of the technology to different quantities. Supporting documentation on the software package to be provided to ONRL is incorporated in this report.

Robotics take heat out of reactor. [Windscale AGR decommissioning

Energy Technology Data Exchange (ETDEWEB)

Rufford, N

1986-12-04

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

Thermal Management of Power Semiconductor Packages - Matching Cooling Technologies with Packaging Technologies (Presentation)

Energy Technology Data Exchange (ETDEWEB)

Bennion, K.; Moreno, G.

2010-04-27

Heat removal for power semiconductor devices is critical for robust operation. Because there are different packaging options, different thermal management technologies, and a range of applications, there is a need for a methodology to match cooling technologies and package configurations to target applications. To meet this need, a methodology was developed to compare the sensitivity of cooling technologies on the overall package thermal performance over a range of power semiconductor packaging configurations. The results provide insight into the trade-offs associated with cooling technologies and package configurations. The approach provides a method for comparing new developments in power semiconductor packages and identifying potential thermal control technologies for the package. The results can help users select the appropriate combination of packaging configuration and cooling technology for the desired application.

Development and experimental qualification of the new safety-criticality CRISTAL package

International Nuclear Information System (INIS)

Mattera, Ch.

1998-11-01

This thesis is concerned with Criticality-Safety studies related to the French Nuclear Fuel Cycle. We first describe the steps in the nuclear fuel cycle and the specific characteristics of these studies compared with those performed in Reactor Physics. In order to respond to the future requirements of the French Nuclear Program, we have developed a new package CRISTAL based on a recent cross sections library (CEA 93) and the newest accurate codes (APOLLO 2, MORET 4, TRIPOLI 4). The CRISTAL system includes two calculations routes: a design route which will be used by French Industry (COGEMA/SGN) and a reference route. To transfer this package to the French industry, we have elaborated calculation schemes for fissile solutions, dissolver media, transport casks and storage pools. Afterwards, these schemes have been used for the CRISTAL experimental validation. We have also contributed to the CRISTAL experimental database by reevaluating a French storage pool experiment: the CRISTO II experiment. This revaluation has been submitted to the OECD working group in order that this experiment can be used by international criticality safety engineers to validate calculations methods. This work represents a large contribution to the recommendation of accurate calculation schemes and to the experimental validation of the CRISTAL package. These studies came up to the French Industry expectations. (author)

Development and experimental testing of the new safety-criticality Cristal package

International Nuclear Information System (INIS)

Mattera, Ch.

1998-01-01

This thesis is concerned with Criticality-Safety studies related to the French Nuclear Fuel Cycle. We first describe the steps in the nuclear fuel cycle and the specific characteristics of these studies compared with those performed in Reactor Physics. In order to respond to the future requirements of the French Nuclear Program, we have developed a new package CRISTAL based on a recent cross sections library (CEA93) and the newest accurate codes (APOLLO2, MORET4, TRIPOLI4). The cristal system includes two calculations routes: a design route which will be used by French Industry (COGEMA/SGN) and a reference route.) To transfer this package to the French industry, we have elaborated calculation schemes for fissile solutions, dissolver media, transport casks and storage pools. Afterwards, these schemes have been used for the CRISTAL experimental validation. We have also contributed to the CRISTAL experimental database by reevaluating a French storage pool experiment: the CRISTO II experiment. This revaluation has been submitted to the OCDE working group in order that this experiment can be used by international criticality safety engineers to validate calculations methods. This work represents a large contribution to the recommendation of accurate calculation schemes and to the experimental validation of the CRISTAL package. These studies came up to the French Industry expectations. (author)

Waste package emplacement borehole option study

International Nuclear Information System (INIS)

Streeter, W.S.

1992-03-01

This study evaluates the cost and thermal effects of various waste package emplacement configurations that differ in emplacement orientation, number of containers per borehole, and standoff distance at the potential Yucca Mountain nuclear waste repository. In this study, eight additional alternatives to the vertical and horizontal orientation options presented in the Site Characterization Plan Conceptual Design Report are considered. Typical panel layout configurations based on thermal analysis of the waste and cost estimates for design and construction, operations, and closure and decommissioning were made for each emplacement option. For the thermal analysis average waste 10 years out of reactor and the SIM code were used to determine whether the various configurations temperatures would exceed the design criteria for temperature. This study does not make a recommendation for emplacement configuration, but does provide information for comparison of alternatives

Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

Energy Technology Data Exchange (ETDEWEB)

Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.; Holdaway, K. K.; Housley, G. K.; Rabin, B. H.

2016-10-01

The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, and other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.

Waste transport and storage: Packaging refused due to failure in fulfilling QC/QA requirements

International Nuclear Information System (INIS)

Bruno, N.C.; Brandao, R.O.; Cavalcante, V.L.

2001-01-01

The Brazilian Nuclear Programme comprises several nuclear and radioactive facilities including Angra I Nuclear Power Plant, in operation since 1981, and Angra II, scheduled to start its operation by the end of 1999. Among the other ones there are uranium mining and milling facilities, four research reactors and one industrial facility of monazite sands processing. The already existing waste generation and near future ones claim to a solution regarding waste disposal. Although site selection criteria for waste repository in Brazil has already been defined, political and psychosocial aspects have strong impact. Trauma generated by Goiania's radiological accident has led to difficulties when decisions about this matter have to be taken. As a consequence, the waste generated by Angra I is still in a provisional facility at the plant's site. Wastes from the medical sources are stored in research institutes while waste generated from monazite sands is kept in a dam system. In order to overpack non-qualified packages containing waste of Angra I NPP, 70 lost concrete shielding packagings had to be provided. Based on successfully designed and tested prototype, packagings and respective lids specifications were written, approved and released for serial production. As part of packaging certification process, Brazilian Competent Authority performed a regulatory inspection and audit. Various findings, such as weaknesses in quality control and quality assurance records, unacceptable test results as well as failure in modify the concrete composition during a testified packaging manufacturing, led Competent Authority to refuse the packagings as containers until complementary tests could be performed. Further tests and evaluations led the Competent Authority to conclude that the manufacturer failed to both comply with requirements established in packaging specification and fulfill quality control/quality assurance requirements. As responsible by federal law for the reception and

Certification testing of the MOX Fresh Fuel Package (MFFP)

International Nuclear Information System (INIS)

Nichols, J.C. III; Yapuncich, F.L.

2004-01-01

Packaging Technology, Inc. (PacTec) is designing the MFFP as part of the Duke, COGEMA, Stone and Webster (DCS) consortium. DCS is tasked with providing the Department of Energy (DOE) with domestic MOX fuel fabrication and reactor irradiation services for the purpose of disposing of surplus weapons usable plutonium. This paper will discuss the development of the MFFP certification test program. The MFFP was subjected to a total of eleven free and puncture drops of the course of the certification testing. Because of the plutonium content, the design must be a Type BF, which among other things requires a containment boundary with a tested leakage rate of 1 x 10 -7 cm 3 /s air at 1 atm absolute and 25 C, or less. Both economics (desire for maximized payload) and operational (conveyance mode restricts size and weight) constraints lead to a highly optimized design. The optimized package design led to a significant test program which needed to address the containment boundary stability, puncture resistance of the package and lid end impact limiter, structural performance of the light weight lid structure, and stability of the internal structures. The test program efficiently balanced the test objectives while minimizing the number of costly hardware items used during this destructive testing. This balance achieved by strategic replacement of mock and prototypic payloads, impact limiters, and by careful test order considerations. The paper will conclude with a selected summary of the testing and an assessment of the test programs thoroughness

Packaging for Food Service

Science.gov (United States)

Stilwell, E. J.

1985-01-01

Most of the key areas of concern in packaging the three principle food forms for the space station were covered. It can be generally concluded that there are no significant voids in packaging materials availability or in current packaging technology. However, it must also be concluded that the process by which packaging decisions are made for the space station feeding program will be very synergistic. Packaging selection will depend heavily on the preparation mechanics, the preferred presentation and the achievable disposal systems. It will be important that packaging be considered as an integral part of each decision as these systems are developed.

Certification of packagings: compliance with DOT specification 7A packaging requirements

International Nuclear Information System (INIS)

Edling, D.A.

1976-01-01

A study was conducted to determine which of the packagings currently listed in CFR 49 Section 173.395 a.1-5, meet the Specification 7A requirements (CFR 49 Section 173.350). According to DOT HM-111 the present listing of various authorized DOT specifications in Section 173.394 and Section 173.395 (Type A containers) of ICC Tariff No. 27 would be deleted with complete reliance being placed on the use of DOT 7A, Type A general packaging specification. Each user of a Specification 7A package would be required to document and maintain on file for one year a written record of his determination of compliance with the DOT Specification 7A performance requirements. All the specification packagings listed in CFR 49 Section 173.395a.1-5 were tested and shown to meet the Specification 7A criteria; however, in many cases qualifications were placed on their use. Forty-nine specification packagings were tested and shown to meet the DOT Specification 7A performance requirements and since there were several styles of some specific packagings, this amounts to greater than 80 packagings. The extensive testing generally indicated a high degree of containment integrity in the packagings tested and the documentation discussed is a valuable tool for shippers of Type A quantities of radioactive materials

Burnable absorber for the PIK reactor

International Nuclear Information System (INIS)

Gostev, V.V.; Smolskii, S.L.; Tchmshkyan, D.V.; Zakharov, A.S.; Zvezdkin, V.S.; Konoplev, K.A.

1998-01-01

In the reactor PIK design a burnable absorber is not used and the cycle duration is limited by the rods weight. Designed cycle time is two weeks and seams to be not enough for the 100 MW power research reactor equipped by many neutron beams and experimental facilities. Relatively frequent reloading reduces the reactor time on full power and in this way increases the maintenance expenses. In the reactor core fuel elements well mastered by practice are used and its modification was not approved. We try to find the possibilities of installation in the core separate burnable elements to avoid poison of the fuel. It is possible to replace a part of the fuel elements by absorbers, since the fuel elements are relatively small (diameter 5.15mm, uranium 235 content 7.14g) and there are more then 3800 elements in the core. Nevertheless, replacing decreases the fuel burnup and its consumption. In the PIK fuel assembles a little part of the volume is occupied by the dumb elements to create a complete package of the assembles shroud, that is necessary in the hydraulic reasons. In the presented report the assessment of such a replacement is done. As a burnable material Gadolinium was selected. The measurements or the beginning of cycle were performed on the critical facility PIK. The burning calculation was confirmed by measurements on the 18MW reactor WWR-M. The results give the opportunity to twice the cycle duration. The proposed modification of the fuel assembles does not lead to alteration in the other reactor systems, but it touch the burned fuel reprocessing technology. (author)

New flux detectors for CANDU 6 reactors

International Nuclear Information System (INIS)

Cuttler, J.M.; Medak, N.

1992-06-01

CANDU reactors utilize large numbers of in-core self-powered detectors for control and protection. In the original design, the detectors (coaxial cables) were wound on carrier tubes and immersed in the heavy water moderator. Failures occurred due to corrosion and other factors, and replacement was very costly because the assemblies were not designed with maintenance in mind. A new design was conceived based on straight detectors, of larger diameter, in a sealed package of individual 'well' tubes. This protected the detectors from hostile environments and enabled individual failed sensors to be replaced by inserting spares in vacant neighbouring tubes. The new design was made retrofittable to older CANDU reactors. Provision was made for on-line scanning of the core with a miniature fission chamber. The modified detectors were tested in a lengthy development program and found to exhibit superior performance to that of the original detectors. Most of the CANDU reactors have now adopted the new design. In the case of the Gentilly-2 and Point Lepreau reactors, advantage was taken of the opportunity to redesign the detector layout (using better codes and the increased flexibility in positioning detectors) to achieve better coverage of abnormal events, leading to higher trip setpoints and wider operating margins

Anhydrous Ammonia Training Module. Trainer's Package. Participant's Package.

Science.gov (United States)

Beaudin, Bart; And Others

This document contains a trainer's and a participant's package for teaching employees on site safe handling procedures for working with anhydrous ammonia, especially on farms. The trainer's package includes the following: a description of the module; a competency; objectives; suggested instructional aids; a training outline (or lesson plan) for…

Gamma and Neutron Radiolysis in the 21-PWR Waste Package

Energy Technology Data Exchange (ETDEWEB)

J.S. Tang

2001-05-03

The objective of this calculation is to compute gamma and neutron dose rates in order to determine the maximum radiolytic production of nitric acid and other chemical species inside the 21-PWR (pressurized-water reactor) waste package (WP). The scope of this calculation is limited to the time period between 5,000 and 100,000 years after emplacement. The information provided by the sketches attached to this calculation is that of the potential design for the type of WP considered in this calculation. The results of this calculation will be used to evaluate nitric acid corrosion of fuel cladding from radiolysis in the 21-PWR WP. This calculation was performed in accordance with the Technical Work Plan for: Waste Package Design Description for LA (Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M&O) 2000a). AP-3.124, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the total system performance assessment (TSPA) of which the spent fuel cladding integrity is to be evaluated.

Gamma and Neutron Radiolysis in the 21-PWR Waste Package

International Nuclear Information System (INIS)

J.S. Tang

2001-01-01

The objective of this calculation is to compute gamma and neutron dose rates in order to determine the maximum radiolytic production of nitric acid and other chemical species inside the 21-PWR (pressurized-water reactor) waste package (WP). The scope of this calculation is limited to the time period between 5,000 and 100,000 years after emplacement. The information provided by the sketches attached to this calculation is that of the potential design for the type of WP considered in this calculation. The results of this calculation will be used to evaluate nitric acid corrosion of fuel cladding from radiolysis in the 21-PWR WP. This calculation was performed in accordance with the Technical Work Plan for: Waste Package Design Description for LA (Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M and O) 2000a). AP-3.124, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the total system performance assessment (TSPA) of which the spent fuel cladding integrity is to be evaluated

Verification of a BWR code package by gamma scan measurements

International Nuclear Information System (INIS)

Nakajima, Tsuyoshi; Iwamoto, Tatsuya; Kumanomido, Hironori

1996-01-01

High-burnup 8 x 8 fuel with a large central water rod (called step 2 fuel) has been recently introduced to the latest Japanese boiling water reactor (BWR) plants. Lanthanum-140 gamma intensity is almost directly related to nodal powers. By gamma scan measurement, the axial distribution of 140 La in the exposed fuel was measured at the end of cycle (EOC) 1 and was compared with the calculation by a BWR code package TGBLA/LOGOS. The multienrichment fuel-type core (MEC) design was adopted for the initial cycle core of the plants. The MEC design contains three different enrichment types of fuels to simulate the equilibrium cycles, achieve much higher discharge exposure, and save fuel cycle cost, and the low-enrichment fuels are loaded in periphery and in control cells. Such MEC design could be a challenge to the BWR design methods because of the large spectrum mismatch among the fuel assemblies of the different enrichments. The aforementioned comparison has shown that the accuracy of the TGBLA/LOGOS code package is satisfactory

ASAMPSA2 best-practices guidelines for L2 PSA development and applications. Volume 3 - Extension to Gen IV reactors

International Nuclear Information System (INIS)

Bassi, C.; Bonneville, H.; Brinkman, H.; Burgazzi, L.; Polidoro, F.; Vincon, L.; Jouve, S.

2010-01-01

The main objective assigned to the Work Package 4 (WP4) of the 'ASAMPSA2' project (EC 7. FPRD) consist in the verification of the potential compliance of L2PSA guidelines based on PWR/BWR reactors (which are specific tasks of WP2 and WP3) with Generation IV representative concepts. Therefore, in order to exhibit potential discrepancies between LWRs and new reactor types, the following work was based on the up-to-date designs of: - The European Fast Reactor (EFR) which will be considered as prototypical of a pool-type Sodium-cooled Fast Reactor (SFR); - The ELSY design for the Lead-cooled Fast Reactor (LFR) technology; - The ANTARES project which could be representative of a Very-High Temperature Reactor (VHTR); - The CEA 2400 MWth Gas-cooled Fast Reactor (GFR). (authors)

General meeting. Technical reunion: the numerical and experimental simulation applied to the Reactor Physics

International Nuclear Information System (INIS)

2001-10-01

The SFEN (French Society on Nuclear Energy), organized the 18 october 2001 at Paris, a technical day on the numerical and experimental simulation, applied to the reactor Physics. Nine aspects were discussed, giving a state of the art in the domain:the french nuclear park; the future technology; the controlled thermonuclear fusion; the new organizations and their implications on the research and development programs; Framatome-ANP markets and industrial code packages; reactor core simulation at high temperature; software architecture; SALOME; DESCARTES. (A.L.B.)

Green Packaging Development. : A way to efficient, effective and more environmental friendly packaging solutions.

OpenAIRE

Mian Muhammad, Masoud

2011-01-01

Growing pressure on the packaging design to enhance the environmental and logistics performance of a packaging system stresses the packaging designers to search new design strategies that not only fulfill logistics requirements in the supply chain, but also reduce the CO 2emissions during the packaging life cycle. This thesis focuses on the packaging design process and suggests some improvements by considering its logistics performance and CO 2emissions. A Green packaging development model wa...

Hazardous materials package performance regulations

International Nuclear Information System (INIS)

Russell, N.A.; Glass, R.E.; McClure, J.D.; Finley, N.C.

1992-01-01

The hazardous materials (hazmat) packaging development and certification process is currently defined by two different regulatory philosophies, one based on specification packagings and the other based on performance standards. With specification packagings, a packaging is constructed according to an agreed set of design specifications. In contrast, performance standards do not specify the packaging design; they specify performance standards that a packaging design must be able to pass before it can be certified for transport. The packaging can be designed according to individual needs as long as it meets these performance standards. Performance standards have been used nationally and internationally for about 40 years to certify radioactive materials (RAM) packagings. It is reasonable to state that for RAM transport, performance specifications have maintained transport safety. A committee of United Nation's experts recommended the performance standard philosophy as the preferred regulation method for hazmat packaging. Performance standards for hazmat packagings smaller than 118 gallons have been adopted in 49CFR178. Packagings for materials that are classified as toxic-by-inhalation must comply with the performance standards by October 1, 1993, and packagings for all other classes of hazardous materials covered must comply by October 1, 1996. For packages containing bulk (in excess of 188 gallons) quantities of materials that are extremely toxic by inhalation, there currently are no performance requirements. This paper discusses a Hazmat Packaging Performance Evaluation (HPPE) project to look at the subset of bulk packagings that are larger than 2000 gallons. The objectives of this project are the evaluate current hazmat specification packagings and develop supporting documentation for determining performance requirements for packagings in excess of 2000 gallons that transport hazardous materials that have been classified as extremely toxic by inhalation (METBI)

Lithium niobate packaging challenges

International Nuclear Information System (INIS)

Murphy, E.J.; Holmes, R.J.; Jander, R.B.; Schelling, A.W.

1988-01-01

The use of lithium niobate integrated optic devices outside of the research laboratory is predicated on the development of a sound packaging method. The authors present a discussion of the many issues that face the development of a viable, robust packaging technology. The authors emphasize the interaction of lithium niobate's physical properties with available packaging materials and technologies. The broad range of properties (i.e. electro-optic, piezo-electric, pyro-electric, photorefractive...) that make lithium niobate an interesting material in many device applications also make it a packaging challenge. The package design, materials and packaging technologies must isolate the device from the environment so that lithium niobate's properties do not adversely affect the device performance

User friendly packaging

DEFF Research Database (Denmark)

Geert Jensen, Birgitte

2010-01-01

Most consumers have experienced occasional problems with opening packaging. Tomato sauce from the tinned mackerel splattered all over the kitchen counter, the unrelenting pickle jar lid, and the package of sliced ham that cannot be opened without a knife or a pair of scissors. The research project...... “User-friendly Packaging” aims to create a platform for developing more user-friendly packaging. One intended outcome of the project is a guideline that industry can use in development efforts. The project also points the way for more extended collaboration between companies and design researchers. How...... can design research help industry in packaging innovation?...

Methods and codes for neutronic calculations of the MARIA research reactor

International Nuclear Information System (INIS)

Andrzejewski, K.; Kulikowska, T.; Bretscher, M.M.; Hanan, N.A.; Matos, J.E.

1998-01-01

The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6x8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminium. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with RERTR program. At IAE the package of programs was developed to help its operator in optimization of fuel utilization. (author)

Application study of EPICS-based redundant method for reactor control system

International Nuclear Information System (INIS)

Zhang Ning; Han Lifeng; Chen Yongzhong; Guo Bing; Yin Congcong

2013-01-01

In the reactor control system prototype development of TMSR (Thorium Molten Salt Reactor system, CAS) project, EPICS (Experimental Physics and Industrial Control System) is adopted as Instrument and Control software platform. For the aim of IOC (Input/Output Controller) redundancy and data synchronization of the system, the EPICS-based RMT (Redundancy Monitor Task ) software package and its data-synchronization component CCE (Continuous Control Executive) were introduced. By the development of related IOC driver, redundant switch-over control of server IOC was implemented. The method of redundancy implementation using RMT in server and redundancy performance test for power control system are discussed in this paper. (authors)

Spent Fuel Handling and Packaging Program: a survey of hot cell facilities

International Nuclear Information System (INIS)

Menon, M.N.

1978-07-01

Hot cell facilities in the United States were surveyed to determine their capabilities for conducting integral fuel assembly and individual fuel rod examinations that are required in support of the Spent Fuel Handling and Packaging Program. The ability to receive, handle, disassemble and reconstitute full-length light water reactor spent fuel assemblies, and the ability to conduct nondestructive and destructive examinations on full-length fuel rods were of particular interest. Three DOE-supported facilities and three commercial facilities were included in the survey. This report provides a summary of the findings

Physics-magnetics trade studies for tandem mirror reactors

International Nuclear Information System (INIS)

Campbell, R.B.; Perkins, L.J.; Blackfield, D.T.

1985-01-01

We describe and present results obtained from the optimization package of the Tandem Mirror Reactor Systems Code. We have found it to be very useful in searching through multidimensional parameter space, and have applied it here to study the effect of choke coil field strength and net electric power on cost of electricity (COE) and mass utilization factor (MUF) for MINIMARS type reactors. We have found that a broad optimum occurs at B/sub choke/ = 26 T for both COE and MUF. The COE economy of scale approaches saturation at quite low powers, around 600 MW(e). The saturation is mainly due to longer construction times for large plants, and the associated time related costs. The MUF economy of scale does not saturate, at least for powers up to 2400 MW(e)

Effect of core configuration on the burnup calculations of MTR research reactors

International Nuclear Information System (INIS)

Hussein, H.M.; Amin, E.H.; Sakr, A.M.

2014-01-01

Highlights: • 3D burn-up calculations of MTR-type research reactor were performed. Examination of the effect of control rod pattern on power density and neutron flux distributions is presented. • The calculations are performed using the MTR P C package and the programs (WIMS and CITVAP). • An empirical formula was generated for every fuel element type, to correlate irradiation to burn-up. - Abstract: In the present paper, three-dimensional burn-up calculations were performed using different patterns of control rods, in order to examine their effect on power density and neutron flux distributions through out the entire core and hence on the local burn-up distribution. These different cores burn-up calculations are carried out for an operating cycle equivalent to 15 Full Power Days (FPDs), with a power rating of 22 MW. Calculations were performed using an example of a typical research reactor of MTR-type using the internationally known computer codes’ package “MTR P C system”, using the cell calculation transport code WIMS-D4 with 12 energy groups and the core calculation diffusion code CITVAP with 5 energy groups. A depletion study was done and the effects on the research reactor fuel (U-235) were performed. The burn-up percentage (B.U.%) curves for every fuel element type were drawn versus irradiation (MWD/TE). Then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Charts of power density and neutron flux distribution for each core were plotted at different sections of each fuel element of the reactor core. Then a complete discussion and analysis of these curves are performed with comparison between the different core configurations, illustrating the effect of insertion or extraction of either of the four control rods directly on the neutron flux and consequently on the power distribution and burn-up. A detailed study of fuel burn-up gives detailed insight on the different B.U.% calculations

Packaging Review Guide for Reviewing Safety Analysis Reports for Packagings

Energy Technology Data Exchange (ETDEWEB)

DiSabatino, A; Biswas, D; DeMicco, M; Fisher, L E; Hafner, R; Haslam, J; Mok, G; Patel, C; Russell, E

2007-04-12

This Packaging Review Guide (PRG) provides guidance for Department of Energy (DOE) review and approval of packagings to transport fissile and Type B quantities of radioactive material. It fulfills, in part, the requirements of DOE Order 460.1B for the Headquarters Certifying Official to establish standards and to provide guidance for the preparation of Safety Analysis Reports for Packagings (SARPs). This PRG is intended for use by the Headquarters Certifying Official and his or her review staff, DOE Secretarial offices, operations/field offices, and applicants for DOE packaging approval. This PRG is generally organized at the section level in a format similar to that recommended in Regulatory Guide 7.9 (RG 7.9). One notable exception is the addition of Section 9 (Quality Assurance), which is not included as a separate chapter in RG 7.9. Within each section, this PRG addresses the technical and regulatory bases for the review, the manner in which the review is accomplished, and findings that are generally applicable for a package that meets the approval standards. This Packaging Review Guide (PRG) provides guidance for DOE review and approval of packagings to transport fissile and Type B quantities of radioactive material. It fulfills, in part, the requirements of DOE O 460.1B for the Headquarters Certifying Official to establish standards and to provide guidance for the preparation of Safety Analysis Reports for Packagings (SARPs). This PRG is intended for use by the Headquarters Certifying Official and his review staff, DOE Secretarial offices, operations/field offices, and applicants for DOE packaging approval. The primary objectives of this PRG are to: (1) Summarize the regulatory requirements for package approval; (2) Describe the technical review procedures by which DOE determines that these requirements have been satisfied; (3) Establish and maintain the quality and uniformity of reviews; (4) Define the base from which to evaluate proposed changes in scope

DNA Packaging by λ-Like Bacteriophages: Mutations Broadening the Packaging Specificity of Terminase, the λ-Packaging Enzyme

OpenAIRE

Feiss, Michael; Reynolds, Erin; Schrock, Morgan; Sippy, Jean

2010-01-01

The DNA-packaging specificities of phages λ and 21 depend on the specific DNA interactions of the small terminase subunits, which have support helix-turn-recognition helix-wing DNA-binding motifs. λ-Terminase with the recognition helix of 21 preferentially packages 21 DNA. This chimeric terminase's ability to package λDNA is reduced ∼20-fold. Phage λ with the chimeric terminase is unable to form plaques, but pseudorevertants are readily obtained. Some pseudorevertants have trans-acting suppre...

Solution of the reactor point kinetics equations by MATLAB computing

Directory of Open Access Journals (Sweden)

Singh Sudhansu S.

2015-01-01

Full Text Available The numerical solution of the point kinetics equations in the presence of Newtonian temperature feedback has been a challenging issue for analyzing the reactor transients. Reactor point kinetics equations are a system of stiff ordinary differential equations which need special numerical treatments. Although a plethora of numerical intricacies have been introduced to solve the point kinetics equations over the years, some of the simple and straightforward methods still work very efficiently with extraordinary accuracy. As an example, it has been shown recently that the fundamental backward Euler finite difference algorithm with its simplicity has proven to be one of the most effective legacy methods. Complementing the back-ward Euler finite difference scheme, the present work demonstrates the application of ordinary differential equation suite available in the MATLAB software package to solve the stiff reactor point kinetics equations with Newtonian temperature feedback effects very effectively by analyzing various classic benchmark cases. Fair accuracy of the results implies the efficient application of MATLAB ordinary differential equation suite for solving the reactor point kinetics equations as an alternate method for future applications.

The paradox of packaging optimization – a characterization of packaging source reduction in the Netherlands

NARCIS (Netherlands)

van Sluisveld, M.A.E.; Worrell, E.

2013-01-01

The European Council Directive 94/62/EC for Packaging and Packaging Waste requires that Member States implement packaging waste prevention measures. However, consumption and subsequently packaging waste figures are still growing annually. It suggests that policies to accomplish packaging waste

A 3D transport-based core analysis code for research reactors with unstructured geometry

International Nuclear Information System (INIS)

Zhang, Tengfei; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi; Li, Yunzhao

2013-01-01

Highlights: • A core analysis code package based on 3D neutron transport calculation in complex geometry is developed. • The fine considerations on flux mapping, control rod effects and isotope depletion are modeled. • The code is proved to be with high accuracy and capable of handling flexible operational cases for research reactors. - Abstract: As an effort to enhance the accuracy in simulating the operations of research reactors, a 3D transport core analysis code system named REFT was developed. HELIOS is employed due to the flexibility of describing complex geometry. A 3D triangular nodal S N method transport solver, DNTR, endows the package the capability of modeling cores with unstructured geometry assemblies. A series of dedicated methods were introduced to meet the requirements of research reactor simulations. Afterwards, to make it more user friendly, a graphical user interface was also developed for REFT. In order to validate the developed code system, the calculated results were compared with the experimental results. Both the numerical and experimental results are in close agreement with each other, with the relative errors of k eff being less than 0.5%. Results for depletion calculations were also verified by comparing them with the experimental data and acceptable consistency was observed in results

Uncertainty analysis of the 35% reactor inlet header break in a CANDU 6 reactor using RELAP/SCDAPSIM/MOD4.0 with integrated uncertainty analysis option

International Nuclear Information System (INIS)

Dupleac, D.; Perez, M.; Reventos, F.; Allison, C.

2011-01-01

The RELAP/SCDAPSIM/MOD4.0 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology Software Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD4.0, which is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards, uses the publicly available RELAP5 and SCDAP models in combination with (a) advanced programming and numerical techniques, (b) advanced SDTP-member-developed models for LWR, HWR, and research reactor analysis, and (c) a variety of other member-developed computational packages. One such computational package is an integrated uncertainty analysis (IUA) package being developed jointly by the Technical University of Catalonia (UPC) and Innovative Systems Software (ISS). RELAP/SCDAPSIM/MOD4.0(IUA) follows the input-propagation approach using probability distribution functions to define the uncertainty of the input parameters. The main steps for this type of methodologies, often referred as to statistical approaches or Wilks’ methods, are the ones that follow: 1. Selection of the plant; 2. Selection of the scenario; 3. Selection of the safety criteria; 4. Identification and ranking of the relevant phenomena based on the safety criteria; 5. Selection of the appropriate code parameters to represent those phenomena; 6. Association of uncertainty by means of Probability Distribution Functions (PDFs) for each selected parameter; 7. Random sampling of the selected parameters according to its PDF and performing multiple computer runs to obtain uncertainty bands with a certain percentile and confidence level; 8. Processing the results of the multiple computer runs to estimate the uncertainty bands for the computed quantities associated with the selected safety criteria. RELAP/SCDAPSIM/MOD4.0(IUA) calculates the number of required code runs given the desired percentile and confidence level, performs the sampling process for the

Uncertainty analysis of the 35% reactor inlet header break in a CANDU 6 reactor using RELAP/SCDAPSIM/MOD4.0 with integrated uncertainty analysis option

Energy Technology Data Exchange (ETDEWEB)

Dupleac, D., E-mail: danieldu@cne.pub.ro [Politehnica Univ. of Bucharest (Romania); Perez, M.; Reventos, F., E-mail: marina.perez@upc.edu, E-mail: francesc.reventos@upc.edu [Technical Univ. of Catalonia (Spain); Allison, C., E-mail: iss@cableone.net [Innovative Systems Software (United States)

2011-07-01

The RELAP/SCDAPSIM/MOD4.0 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology Software Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD4.0, which is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards, uses the publicly available RELAP5 and SCDAP models in combination with (a) advanced programming and numerical techniques, (b) advanced SDTP-member-developed models for LWR, HWR, and research reactor analysis, and (c) a variety of other member-developed computational packages. One such computational package is an integrated uncertainty analysis (IUA) package being developed jointly by the Technical University of Catalonia (UPC) and Innovative Systems Software (ISS). RELAP/SCDAPSIM/MOD4.0(IUA) follows the input-propagation approach using probability distribution functions to define the uncertainty of the input parameters. The main steps for this type of methodologies, often referred as to statistical approaches or Wilks’ methods, are the ones that follow: 1. Selection of the plant; 2. Selection of the scenario; 3. Selection of the safety criteria; 4. Identification and ranking of the relevant phenomena based on the safety criteria; 5. Selection of the appropriate code parameters to represent those phenomena; 6. Association of uncertainty by means of Probability Distribution Functions (PDFs) for each selected parameter; 7. Random sampling of the selected parameters according to its PDF and performing multiple computer runs to obtain uncertainty bands with a certain percentile and confidence level; 8. Processing the results of the multiple computer runs to estimate the uncertainty bands for the computed quantities associated with the selected safety criteria. RELAP/SCDAPSIM/MOD4.0(IUA) calculates the number of required code runs given the desired percentile and confidence level, performs the sampling process for the

Immobilization of Cesium Traps from the BN-350 Fast Reactor (Aktau, Kazakhstan)

Energy Technology Data Exchange (ETDEWEB)

J. A. Michelbacher; C. Knight; O. G. Romanenko; I. L. Tazhibaeva; I. L. Yakovlev; A. V. Rovneyko; V. I. Maev; D. Wells; A. Herrick

2011-03-01

During BN-350 reactor operations and also during the initial stages of decommissioning, cesium traps were used to decontaminate the reactor’s primary sodium coolant. Two different types of carbon-based trap were used – the MAVR series, low ash granulated graphite adsorber (LAG) contained in a carrier designed to be inserted into the reactor core during shutdown; and a series of ex-reactor trap accumulators(TAs) which used reticulated vitreous carbon (RVC) to reduce Cs-137 levels in the sodium after final reactor shutdown. In total four MAVRs and seven TAs were used at BN-350 to remove an estimated cumulative 755 TBq of cesium. The traps, which also contain residual sodium, need to be immobilized in an appropriate way to allow them to be consigned as waste packages for long term storage and, ultimately, disposal. The present paper reports on the current status of the implementation phase, with particular reference to the work done to date on the trap accumulators, which have the most similarity with the cesium traps used at other reactors.

Review of PSI studies on reactor physics and thermal fluid dynamics of pebble bed reactors

International Nuclear Information System (INIS)

Prasser, Horst-Michael

2014-01-01

Switzerland is member of the Generation IV International Forum (GIF). The related work takes entirely place at PSI in the working groups of Gas-Cooled Fast Reactors and Very High Temperature Reactors. In the past, PSI has performed experimental and theoretical studies on criticality issues of pebble beds at the PROTEUS reactor, as well as a preliminary risk assessment of a prototypal HTR as an input for a comparison of energy supply options. PROTEUS was a critical assembly with an annular driver zone. The central region was filled by arrangements of fuel spheres. The reactivity effect of a water ingress was investigated by simulating the water by polyethylene rods of different diameter inserted into the gaps of a regular package. For sub-criticality measurements in pebble beds, a built-in pulsed neutron source was used. The experimental results were used to validate diffusion and higher order neutron transport models. Concerning thermal hydraulics of gas flows, the vast experience of PSI is focused on hydrogen transport, accumulation, and dispersion in containments of light water reactors. The phenomena are comparable in many aspects to the fluid dynamic issues relevant to HTR. Experiments on hydrogen flows are performed for numerous scenarios in the large-scale containment test facility PANDA. Hydrogen is substituted by helium as a model fluid. An important generic aspect is turbulent mixing in the presence of strong stratification, which is relevant for HTR as well. In a parallel project, generic small-scale mixing experiments with a high density ratio of 1:7 are carried out in a horizontal rectangular channel, where helium and nitrogen flows are brought into contact downstream of the rear edge of a splitter plate. Due to the high density ratio, turbulent mixing is affected by strong non-Boussinesq effects. The measurements taken by Particle Imaging Velocimetry (PIV) and Laser Induced Fluorescence techniques are compared to RANS and LES simulations. Similar large

Prevention policies addressing packaging and packaging waste: Some emerging trends.

Science.gov (United States)

Tencati, Antonio; Pogutz, Stefano; Moda, Beatrice; Brambilla, Matteo; Cacia, Claudia

2016-10-01

Packaging waste is a major issue in several countries. Representing in industrialized countries around 30-35% of municipal solid waste yearly generated, this waste stream has steadily grown over the years even if, especially in Europe, specific recycling and recovery targets have been fixed. Therefore, an increasing attention starts to be devoted to prevention measures and interventions. Filling a gap in the current literature, this explorative paper is a first attempt to map the increasingly important phenomenon of prevention policies in the packaging sector. Through a theoretical sampling, 11 countries/states (7 in and 4 outside Europe) have been selected and analyzed by gathering and studying primary and secondary data. Results show evidence of three specific trends in packaging waste prevention policies: fostering the adoption of measures directed at improving packaging design and production through an extensive use of the life cycle assessment; raising the awareness of final consumers by increasing the accountability of firms; promoting collaborative efforts along the packaging supply chains. Copyright © 2016 Elsevier Ltd. All rights reserved.

Food packages for Space Shuttle

Science.gov (United States)

Fohey, M. F.; Sauer, R. L.; Westover, J. B.; Rockafeller, E. F.

1978-01-01

The paper reviews food packaging techniques used in space flight missions and describes the system developed for the Space Shuttle. Attention is directed to bite-size food cubes used in Gemini, Gemini rehydratable food packages, Apollo spoon-bowl rehydratable packages, thermostabilized flex pouch for Apollo, tear-top commercial food cans used in Skylab, polyethylene beverage containers, Skylab rehydratable food package, Space Shuttle food package configuration, duck-bill septum rehydration device, and a drinking/dispensing nozzle for Space Shuttle liquids. Constraints and testing of packaging is considered, a comparison of food package materials is presented, and typical Shuttle foods and beverages are listed.

Radioactive material packaging performance testing

International Nuclear Information System (INIS)

Romano, T.; Cruse, J.M.

1991-02-01

To provide uniform packaging of hazardous materials on an international level, the United Nations has developed packaging recommendations that have been implemented worldwide. The United Nations packaging recommendations are performance oriented, allowing for a wide variety of package materials and systems. As a result of this international standard, efforts in the United States are being directed toward use of performance-oriented packaging and elimination of specification (designed) packaging. This presentation will focus on trends, design evaluation, and performance testing of radioactive material packaging. The impacts of US Department of Transportation Dockets HM-181 and HM-169A on specification and low-specific activity radioactive material packaging requirements are briefly discussed. The US Department of Energy's program for evaluating radioactive material packings per US Department of Transportation Specification 7A Type A requirements, is used as the basis for discussing low-activity packaging performance test requirements. High-activity package testing requirements are presented with examples of testing performed at the Hanford Site that is operated by Westinghouse Hanford Company for the US Department of Energy. 5 refs., 2 tabs

The design and installation of a core discharge monitor for CANDU-type reactors

International Nuclear Information System (INIS)

Halbig, J.K.; Monticone, A.C.; Ksiezak, L.; Smiltnieks, V.

1990-01-01

A new type of surveillance systems that monitors neutron and gamma radiation in a reactor containment is being installed at the Ontario Hydro Darlington Nuclear Generating Station A, Unit 2. Unlike video or film surveillance that monitors mechanical motion, this system measures fuel-specific radiation emanating from irradiated fuel as it is pushed from the core of CANDU-type reactors. Proof-of-principle measurements have been carried out at Bruce Nuclear Generating Station A, Unit 3. The system uses (γ,n) threshold detectors and ionization detectors. A microprocessor-based electronics package, GRAND-II (Gamma Ray and Neutron Detector electronics package), provides detector bias, preamplifier power, and signal processing. Firmware in the GRAND-2 controls the surveillance activities, including data acquisition and a level of detector authentication, and it handles authenticated communication with a central data logging computer. Data from the GRAND-II are transferred to an MS-DOS-compatible computer and stored. These data are collected and reviewed for fuel-specific radiation signatures from the primary detector and proper ratios of signals from secondary detectors. 5 figs

Experimental and theoretical analyses of package-on-package structure under three-point bending loading

International Nuclear Information System (INIS)

Jia Su; Wang Xi-Shu; Ren Huai-Hui

2012-01-01

High density packaging is developing toward miniaturization and integration, which causes many difficulties in designing, manufacturing, and reliability testing. Package-on-Package (PoP) is a promising three-dimensional high-density packaging method that integrates a chip scale package (CSP) in the top package and a fine-pitch ball grid array (FBGA) in the bottom package. In this paper, in-situ scanning electron microscopy (SEM) observation is carried out to detect the deformation and damage of the PoP structure under three-point bending loading. The results indicate that the cracks occur in the die of the top package, then cause the crack deflection and bridging in the die attaching layer. Furthermore, the mechanical principles are used to analyse the cracking process of the PoP structure based on the multi-layer laminating hypothesis and the theoretical analysis results are found to be in good agreement with the experimental results. (condensed matter: structural, mechanical, and thermal properties)

WASTE PACKAGE TRANSPORTER DESIGN

International Nuclear Information System (INIS)

Weddle, D.C.; Novotny, R.; Cron, J.

1998-01-01

The purpose of this Design Analysis is to develop preliminary design of the waste package transporter used for waste package (WP) transport and related functions in the subsurface repository. This analysis refines the conceptual design that was started in Phase I of the Viability Assessment. This analysis supports the development of a reliable emplacement concept and a retrieval concept for license application design. The scope of this analysis includes the following activities: (1) Assess features of the transporter design and evaluate alternative design solutions for mechanical components. (2) Develop mechanical equipment details for the transporter. (3) Prepare a preliminary structural evaluation for the transporter. (4) Identify and recommend the equipment design for waste package transport and related functions. (5) Investigate transport equipment interface tolerances. This analysis supports the development of the waste package transporter for the transport, emplacement, and retrieval of packaged radioactive waste forms in the subsurface repository. Once the waste containers are closed and accepted, the packaged radioactive waste forms are termed waste packages (WP). This terminology was finalized as this analysis neared completion; therefore, the term disposal container is used in several references (i.e., the System Description Document (SDD)) (Ref. 5.6). In this analysis and the applicable reference documents, the term ''disposal container'' is synonymous with ''waste package''

Optimization of refueling loading pattern of uranium zirconium hydride research reactor

International Nuclear Information System (INIS)

Chen Wei; Xie Zhongsheng; Chen Da

1999-01-01

The orthogonal design method is used in the optimization of in-core fuel management. A code package of in-core fuel management in hexagonal geometry HEX-ORTH is developed. The loading pattern after the end of 3 cycle of Xi'an Pulsed Reactor is optimized using the HEX-ORTH. The optimistic loading pattern of the core are obtained as the objective function is Max(k eff BOC )

Models development for the fuel design of a reactor GT-MHR

International Nuclear Information System (INIS)

Telesforo R, D.; Francois L, J. L.

2009-10-01

The very high temperature reactor (VHTR) it as arisen as an option for the following reactors generation, due to their characteristics as they are inherent security, modularity and relative low cost. One of the VHTR variants, for its fuel based on prismatic blocks, is the modular reactor cooled by gas GT-MHR that uses a fuel particle of multiple layers called TRISO. These particles are small containers with fuel in their interior and they function as small pressure vessels that retain the fission products. They are absorbed inconstant ing in a cylindrical graphite matrix to form the fuel named Compact; the quantity of fuel inventory is proportional to the packaging fraction on the total volume of the Compact. The reactor consists of a matrix of 12 X 12 graphite hexagonal assemblies contained in a cylinder of 3.5 radio meters and 10.0 meters high. The nucleus has ten axial regions with 36 X 3 fuel assemblies distributed in three rings. For the neutronic modeling of the fuel and the nucleus it was employee the Monte Carlo method, using the code MCNPX (Monte Carlo N-Particle version X) that is a transport code of general purposes that uses this method with a great versatility in the representation of arbitrary three-dimensional configurations and materials configuration. The heterogeneous model of the reactor GT-MHR core was obtained, adjusting the relative parameters at core prototype GT-MHR presented by General Atomics. To prove the model it was employee the fuel formed by TRISO particles, with a nucleus of 150 μm and packaging fraction of 37.55%, with a Uranium-235 mixture, as fissile nuclide, and Thorium-232. To create a simplified model of the nucleus, or homogeneous model, without modeling any particle, it was employee the reactivity-equivalent physical transformation method that captures the effects of the heterogeneity double of the fuel region in two homogeneous equivalent cells, being obtained very good results. (Author)

Reactor physics special problem in 11. ENFIR

International Nuclear Information System (INIS)

Leszczynski, Francisco

1997-01-01

In this report, the computation method and the results of the work performed of the special topic on reactor physics proposed for the 11. ENFIR is presented. MCNP 4.2 has been adopted as the only code to perform the calculations. The full core of the IPEN-MB-1 critical unit has been modelled without important approximations. The specifications given by the Organizer Commission of the Special Topic were followed. The nuclear libraries adopted were those included on the MCNPDAT package, mainly from ENDF/B-V, except indium data, not included in this package. For indium, data obtained from LANL, based on ENDF/B-VI were used. The results are: critical position of the control banks assuming simultaneous movement: percent of extraction: (49±1)% ; excess of reactivity of the core: ρ =( 3590 ±50)pcm ; total reactivity of the one control rod bank: ρ= (4000±50) pcm. The reactivity curve of the control rods is included also. (author)

Reactor alarm system development and application issues

Energy Technology Data Exchange (ETDEWEB)

Drexler, J E; Oicese, G O [INVAP S.E. (Argentina)

1997-09-01

The new hardware and software technologies, and the need in research reactors for assistance systems in operation and maintenance, have given an appropriate background to develop a computer based system named ``Reactor Alarm System`` (RAS). RAS is a software package, user oriented, with emphasis on production, experiments and maintenance goals. It is designed to run on distributed systems conformed with microcomputers under QNX operating system. RAS main features are: (a) Alarm Panel Display; (b) Alarm Page; (c) Alarm Masking and Inhibition; (d) Alarms Color and Attributes; (e) Condition Classification; and (f) Arrangement Presentation. RAS design allows it to be installed as a part of a computer based Supervision and Control System in new installations or retrofit existing reactor instrumentation systems. The analysis of human factors during development stage and successive user feedback from different applications, brought out several RAS improvements: (a) Multiple-copy alarm summaries; (b) Improved alarm handling; (c) Extended dictionary; and (d) Enhanced hardware availability. It has proved successful in providing new capabilities for operators, and also has shown the continuous increase of user-demands, reflecting the expectations placed today on computer-based systems. (author). 6 figs, 1 tabs.

Reactor alarm system development and application issues

International Nuclear Information System (INIS)

Drexler, J.E.; Oicese, G.O.

1997-01-01

The new hardware and software technologies, and the need in research reactors for assistance systems in operation and maintenance, have given an appropriate background to develop a computer based system named ''Reactor Alarm System'' (RAS). RAS is a software package, user oriented, with emphasis on production, experiments and maintenance goals. It is designed to run on distributed systems conformed with microcomputers under QNX operating system. RAS main features are: a) Alarm Panel Display; b) Alarm Page; c) Alarm Masking and Inhibition; d) Alarms Color and Attributes; e) Condition Classification; and f) Arrangement Presentation. RAS design allows it to be installed as a part of a computer based Supervision and Control System in new installations or retrofit existing reactor instrumentation systems. The analysis of human factors during development stage and successive user feedback from different applications, brought out several RAS improvements: a) Multiple-copy alarm summaries; b) Improved alarm handling; c) Extended dictionary; and d) Enhanced hardware availability. It has proved successful in providing new capabilities for operators, and also has shown the continuous increase of user-demands, reflecting the expectations placed today on computer-based systems. (author). 6 figs, 1 tabs

A study on the radiometric method for evaluating element migration from plastic packagings to its contents

International Nuclear Information System (INIS)

Soares, Eufemia Paez

2008-01-01

Over the past few years, problems related to food contamination by substances or elements that can be a risk to human health have became a concern, not only to government authorities, but to the general population as well. Within this context, plastic packaging can constitute a source of food contamination since plastic manufacturing processes involve the use of catalysts and different types of additives that may contain toxic elements. When food comes into contact with this packaging, components of the package may migrate to the food. In order to control the material used as food packaging, the National Health Surveillance Agency (ANVISA) in Brazil, has established boundary values of migrant substances and procedures to determine migration from plastic packagings to food. In this study the radiometric method was evaluated for element migration determination from plastic packaging to food simulating or to the food itself. This radiometric method consisted in irradiating plastic packaging samples with a thermal neutron flux from the IEA-R1 nuclear research reactor in order to produce radionuclides of elements present in the packagings. The irradiated plastic was then exposed to food simulant or food for element migration. Gamma ray spectrometry was used to measure radioactivity in the simulant or food in order to quantify the migration. The food simulating types and experimental conditions were established according to the ANVISA regulations. Element migration was studied for plastic packaging used for soft drinks, drinking water, milk, dairy products, juices and fatty foods. In the instrumental neutron activation analysis of these packagings the presence of As, Cd, Cr, Co and Sb II was verified. Results obtained from the migration experiments by the radiometric method indicated that Cd, Co, Cr and Sb present in these plastics migrated to the simulant or to the food. In some packagings, the migration of only some of these elements was observed. In these cases the

Components of Adenovirus Genome Packaging

Science.gov (United States)

Ahi, Yadvinder S.; Mittal, Suresh K.

2016-01-01

Adenoviruses (AdVs) are icosahedral viruses with double-stranded DNA (dsDNA) genomes. Genome packaging in AdV is thought to be similar to that seen in dsDNA containing icosahedral bacteriophages and herpesviruses. Specific recognition of the AdV genome is mediated by a packaging domain located close to the left end of the viral genome and is mediated by the viral packaging machinery. Our understanding of the role of various components of the viral packaging machinery in AdV genome packaging has greatly advanced in recent years. Characterization of empty capsids assembled in the absence of one or more components involved in packaging, identification of the unique vertex, and demonstration of the role of IVa2, the putative packaging ATPase, in genome packaging have provided compelling evidence that AdVs follow a sequential assembly pathway. This review provides a detailed discussion on the functions of the various viral and cellular factors involved in AdV genome packaging. We conclude by briefly discussing the roles of the empty capsids, assembly intermediates, scaffolding proteins, portal vertex and DNA encapsidating enzymes in AdV assembly and packaging. PMID:27721809

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Science.gov (United States)

2010-01-01

... fuel and nuclear waste. 71.97 Section 71.97 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... advance notification of transportation of nuclear waste was published in the Federal Register on June 30...

Evolution of Framatome pressurized water reactor systems

International Nuclear Information System (INIS)

Leroy, C.; Bitsch, D.; Millot, J.P.

1985-10-01

FRAMATOME's PWR experience covers a total of 63 units, 36 of which are operating by end of 1984. More than 10 units were operated in load follow mode. Progress features, resulting from the feedback of construction and operating experience, and from the returns of a vast research and development program, were incorporated in their design through subsequent series of standard units. The last four loop standard, the N4 model, integrates in a rational way all those progress features, together with a significant design effort. The core design is based on the new Advanced Fuel Assemblies. The reactor control implements the ''Reactor Maximum Flexibility Package'' (R-MAX) which provides a high level of automatic reactor control. The steam generator incorporates an axial-mixed flow economizer design. The triangular-pitch tube bundle, together with modular steam/water separators and a rearrangement of the dryers resulted in a compact design. The reactor coolant pump benefits of higher performances over that of previous models due to an optimal hydraulic design, and of mechanical features which increase margins and facilitate the maintenance work. Following the N4 project, design work on advanced concepts is pursued by FRAMATOME. A main way of research is focused on the optimal use of fissile materials. These concepts are based on tight pitch fuel arrays, associated with a mechanical spectral shift device

Development of a remote monitoring system, through monitoring of key safety parameters for a nuclear research reactor

International Nuclear Information System (INIS)

Urcia, Agustin; Arrieta, Rolando; Baltuano, Oscar; Chan, Renzo; Tincopa, Jean Pierre; Urquizo, Rafael; Rosas, Bernick

2014-01-01

This paper presents the detailed development, installation and commissioning of water level sensors and exposure rate range in the 11 meters level (mouth of tank) of the RP-10 nuclear reactor used to continuously monitor these values and use them as security for the periods of no presence of operating personnel (overlooking situation) with the reactor in shutdown state. The levels of these parameters are packaged and transmitted to a controller in the control room of reactor for display and activation of alarm levels. Additionally, the design of these warning signs is shown in conjunction with the fire alarm in the building of reactor and auxiliary laboratories to be transmitted to the physical security facility, located at a distance of 500 meters. (authors).

Genome packaging in viruses

OpenAIRE

Sun, Siyang; Rao, Venigalla B.; Rossmann, Michael G.

2010-01-01

Genome packaging is a fundamental process in a viral life cycle. Many viruses assemble preformed capsids into which the genomic material is subsequently packaged. These viruses use a packaging motor protein that is driven by the hydrolysis of ATP to condense the nucleic acids into a confined space. How these motor proteins package viral genomes had been poorly understood until recently, when a few X-ray crystal structures and cryo-electron microscopy structures became available. Here we discu...

Development of a remote monitoring system, through monitoring of key safety parameters for a nuclear research reactor; Desarrollo de un sistema de vigilancia remota, por medio del monitoreo de parametros claves de seguridad, para un reactor nuclear de investigacion

Energy Technology Data Exchange (ETDEWEB)

Urcia, Agustin; Arrieta, Rolando [Direccion de Produccion, Instituto Peruano de Energia Nuclear, Lima (Peru); Baltuano, Oscar; Chan, Renzo [Direccion de Investigacion y Desarrollo, Instituto Peruano de Energia Nuclear, Lima (Peru); Tincopa, Jean Pierre [Facultad de Ingenieria Electrica y Electronica, Universidad Nacional del Callao, Callao (Peru); Urquizo, Rafael [Facultad de Ingenieria Electronica, Universidad Tecnologica del Peru, Lima (Peru); Rosas, Bernick [Facultad de Ingenieria Electronica, Universidad Nacional de Ingenieria, Lima (Peru)

2014-07-01

This paper presents the detailed development, installation and commissioning of water level sensors and exposure rate range in the 11 meters level (mouth of tank) of the RP-10 nuclear reactor used to continuously monitor these values and use them as security for the periods of no presence of operating personnel (overlooking situation) with the reactor in shutdown state. The levels of these parameters are packaged and transmitted to a controller in the control room of reactor for display and activation of alarm levels. Additionally, the design of these warning signs is shown in conjunction with the fire alarm in the building of reactor and auxiliary laboratories to be transmitted to the physical security facility, located at a distance of 500 meters. (authors).

Proposed rule package on fracture toughness and thermal annealing requirements and guidance for light water reactor vessels

International Nuclear Information System (INIS)

Allen Hiser, J.R.

1993-01-01

In the framework of updating and clarification of the fracture toughness and thermal annealing requirements and guidance for light water reactor pressure vessels, proposed revisions concerning the pressurized thermal shock rule, fracture toughness requirements and reactor vessel material surveillance program requirements, are described. A new rule concerning thermal annealing requirements and a draft regulatory guide on 'Format and Content of Application for Approval for Thermal Annealing of RPV' are also proposed

Proposed rule package on fracture toughness and thermal annealing requirements and guidance for light water reactor vessels

Energy Technology Data Exchange (ETDEWEB)

Allen Hiser, J R [UKAEA Harwell Lab. (United Kingdom). Engineering Div.

1994-12-31

In the framework of updating and clarification of the fracture toughness and thermal annealing requirements and guidance for light water reactor pressure vessels, proposed revisions concerning the pressurized thermal shock rule, fracture toughness requirements and reactor vessel material surveillance program requirements, are described. A new rule concerning thermal annealing requirements and a draft regulatory guide on `Format and Content of Application for Approval for Thermal Annealing of RPV` are also proposed.

Food packaging history and innovations.

Science.gov (United States)

Risch, Sara J

2009-09-23

Food packaging has evolved from simply a container to hold food to something today that can play an active role in food quality. Many packages are still simply containers, but they have properties that have been developed to protect the food. These include barriers to oxygen, moisture, and flavors. Active packaging, or that which plays an active role in food quality, includes some microwave packaging as well as packaging that has absorbers built in to remove oxygen from the atmosphere surrounding the product or to provide antimicrobials to the surface of the food. Packaging has allowed access to many foods year-round that otherwise could not be preserved. It is interesting to note that some packages have actually allowed the creation of new categories in the supermarket. Examples include microwave popcorn and fresh-cut produce, which owe their existence to the unique packaging that has been developed.

Trends in Food Packaging.

Science.gov (United States)

Ott, Dana B.

1988-01-01

This article discusses developments in food packaging, processing, and preservation techniques in terms of packaging materials, technologies, consumer benefits, and current and potential food product applications. Covers implications due to consumer life-style changes, cost-effectiveness of packaging materials, and the ecological impact of…

WASTE PACKAGE TRANSPORTER DESIGN

Energy Technology Data Exchange (ETDEWEB)

D.C. Weddle; R. Novotny; J. Cron

1998-09-23

The purpose of this Design Analysis is to develop preliminary design of the waste package transporter used for waste package (WP) transport and related functions in the subsurface repository. This analysis refines the conceptual design that was started in Phase I of the Viability Assessment. This analysis supports the development of a reliable emplacement concept and a retrieval concept for license application design. The scope of this analysis includes the following activities: (1) Assess features of the transporter design and evaluate alternative design solutions for mechanical components. (2) Develop mechanical equipment details for the transporter. (3) Prepare a preliminary structural evaluation for the transporter. (4) Identify and recommend the equipment design for waste package transport and related functions. (5) Investigate transport equipment interface tolerances. This analysis supports the development of the waste package transporter for the transport, emplacement, and retrieval of packaged radioactive waste forms in the subsurface repository. Once the waste containers are closed and accepted, the packaged radioactive waste forms are termed waste packages (WP). This terminology was finalized as this analysis neared completion; therefore, the term disposal container is used in several references (i.e., the System Description Document (SDD)) (Ref. 5.6). In this analysis and the applicable reference documents, the term ''disposal container'' is synonymous with ''waste package''.

Calculation of mixed HEU-LEU cores for the HOR research reactor with the scale code system

International Nuclear Information System (INIS)

Leege, P.F.A. de; Gibcus, H.P.M.; Hoogenboom, J.E.; Vries, J.W. de

1997-01-01

The HOR reactor of Interfaculty Reactor Institute (IRI), Delft, The Netherlands, will be converted to use low enriched fuel (LEU) assemblies. As there are still many usable high enriched (HEU) fuel assemblies present, there will be a considerable reactor operation time with mixed cores with both HEU and LEU fuel assemblies. At IRI a comprehensive reactor physics code system and evaluated nuclear data is implemented for detailed core calculations. One of the backbones of the IRI code system is the well-known SCALE code system package. Full core calculations are performed with the diffusion theory code BOLD VENTURE, the nodal code SILWER, and the Monte Carlo code KENO Va. Results are displayed of a strategy from a HEU core to a mixed HEU-LEU core and eventually a LEU core. (author)

Development of a fresh plutonium fuel container for a prototype fast breeder reactor

International Nuclear Information System (INIS)

Ohtake, T.; Takahashi, S.; Mishima, T.; Kurakami, J.; Yamamoto, Y.; Ohuchi, Y.

1989-01-01

Japan gives a good deal of encouragement to development of a fast breeder reactor (which is considered as the most likely candidate for nuclear power generation) to secure long-term energy source. And, following an experimental fast breeder reactor Joyo, a prototype fast breeder reactor Monju is now under vigorous construction. Related to development of the prototype fast breeder reactor, it is necessary and important to develop transport container which is used for transporting fresh fuel assemblies from Plutonium Fuel Production Facility to the Monju power plant. Therefore, the container is now being developed by Power Reactor and Nuclear Fuel Development Corporation (PNC). Currently, shipment and vibration tests, handling performance tests, shielding performance tests and prototype container tests are executed with prototype containers fabricated according to a final design, in order to experimentally confirm soundness of transport container and its contents, and propriety of design technique. This paper describes the summary of general specifications and structures of this container and the summary of preliminary safety analysis of package

The Innovative Approaches to Packaging – Comparison Analysis of Intelligent and Active Packaging Perceptions in Slovakia

Directory of Open Access Journals (Sweden)

Loucanova Erika

2017-06-01

Full Text Available Packaging has always served a practical function - to hold goods together and protect it when moving toward the customer through distribution channel. Today packaging is also a container for promoting the product and making it easier and safer to use. The sheer importance of packaging functions is still growing and consequently the interest of the company is to access to the packaging more innovative and creative. The paper deals with the innovative approaches to packaging resulting in the creation of packaging with interactive active features in the form of active and intelligent packaging. Using comparative analysis, we monitored the perception of the active packaging functions in comparison to intelligent packaging function among different age categories. We identified the age categories which are most interested in these functions.

Applications of Active Packaging in Breads

Directory of Open Access Journals (Sweden)

Ali Göncü

2017-10-01

Full Text Available Changes on consumer preferences lead to innovations and improvements in new packaging technologies. With these new developments passive packaging technologies aiming to protect food nowadays have left their place to active and intelligent packaging technologies that have other various functions beside protection of food. Active packaging is defined as an innovative packaging type and its usage increases the shelf life of food significantly. Applications of active packaging have begun to be used for packaging of breads. In this study active packaging applications in breads have been reviewed.

Food Packaging Materials

Science.gov (United States)

1978-01-01

The photos show a few of the food products packaged in Alure, a metallized plastic material developed and manufactured by St. Regis Paper Company's Flexible Packaging Division, Dallas, Texas. The material incorporates a metallized film originally developed for space applications. Among the suppliers of the film to St. Regis is King-Seeley Thermos Company, Winchester, Ma'ssachusetts. Initially used by NASA as a signal-bouncing reflective coating for the Echo 1 communications satellite, the film was developed by a company later absorbed by King-Seeley. The metallized film was also used as insulating material for components of a number of other spacecraft. St. Regis developed Alure to meet a multiple packaging material need: good eye appeal, product protection for long periods and the ability to be used successfully on a wide variety of food packaging equipment. When the cost of aluminum foil skyrocketed, packagers sought substitute metallized materials but experiments with a number of them uncovered problems; some were too expensive, some did not adequately protect the product, some were difficult for the machinery to handle. Alure offers a solution. St. Regis created Alure by sandwiching the metallized film between layers of plastics. The resulting laminated metallized material has the superior eye appeal of foil but is less expensive and more easily machined. Alure effectively blocks out light, moisture and oxygen and therefore gives the packaged food long shelf life. A major packaging firm conducted its own tests of the material and confirmed the advantages of machinability and shelf life, adding that it runs faster on machines than materials used in the past and it decreases product waste; the net effect is increased productivity.

Code package to analyse behavior of the WWER fuel rods in normal operation: TOPRA's code

International Nuclear Information System (INIS)

Scheglov, A.; Proselkov, V.

2001-01-01

This paper briefly describes the code package intended for analysis of WWER fuel rod characteristics. The package includes two computer codes: TOPRA-1 and TOPRA-2 for full-scale fuel rod analyses; MRZ and MKK codes for analyzing the separate sections of fuel rods in r-z and r-j geometry. The TOPRA's codes are developed on the base of PIN-mod2 version and verified against experimental results obtained in MR, MIR and Halden research reactors (in the framework of SOFIT, FGR-2 and FUMEX experimental programs). Comparative analysis of calculation results and results from post-reactor examination of the WWER-440 and WWER-1000 fuel rod are also made as additional verification of these codes. To avoid the enlarging of uncertainties in fuel behavior prediction as a result of simplifying of the fuel geometry, MKK and MRZ codes are developed on the basis of the finite element method with use of the three nodal finite elements. Results obtained in the course of the code verification indicate the possibility for application of the method and TOPRA's code for simplified engineering calculations of WWER fuel rods thermal-physical parameters. An analysis of maximum relative errors for predicting of the fuel rod characteristics in the range of the accepted parameter values is also presented in the paper

Plasma physics plotting package

International Nuclear Information System (INIS)

Hyman, D.H.

1981-02-01

We describe a package of plotting routines that do up to six two- or three-dimensional plots on a frame with minimal loss of resolution. The package now runs on a PDP-10 with PLOT-10 TCS primitives and on a Control Data Corporation-7600 and a Cray-1 with TV80LIB primitives on the National Magnetic Fusion Energy Computer Center network. The package is portable to other graphics systems because only the primitive plot calls are used from the underlying system's graphics package

Active food packaging technologies.

Science.gov (United States)

Ozdemir, Murat; Floros, John D

2004-01-01

Active packaging technologies offer new opportunities for the food industry, in the preservation of foods. Important active packaging systems currently known to date, including oxygen scavengers, carbon dioxide emitters/absorbers, moisture absorbers, ethylene absorbers, ethanol emitters, flavor releasing/absorbing systems, time-temperature indicators, and antimicrobial containing films, are reviewed. The principle of operation of each active system is briefly explained. Recent technological advances in active packaging are discussed, and food related applications are presented. The effects of active packaging systems on food quality and safety are cited.

Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core

International Nuclear Information System (INIS)

Loetsch, T.; Khalimonchuk, V.; Kuchin, A.

2009-01-01

In the framework of a project supported by the German BMU the code DYN3D should be further validated and verified. During the work a lack of a benchmark on core burnup calculations for VVER-1000 reactors was noticed. Such a benchmark is useful for validating and verifying the whole package of codes and data libraries for reactor physics calculations including fuel assembly modelling, fuel assembly data preparation, few group data parametrisation and reactor core modelling. The benchmark proposed specifies the core loading patterns of burnup cycles for a VVER-1000 reactor core as well as a set of operational data such as load follow, boron concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions. The reactor core characteristics chosen for comparison and the first results obtained during the work with the reactor physics code DYN3D are presented. This work presents the continuation of efforts of the projects mentioned to estimate the accuracy of calculated characteristics of VVER-1000 reactor cores. In addition, the codes used for reactor physics calculations of safety related reactor core characteristics should be validated and verified for the cases in which they are to be used. This is significant for safety related evaluations and assessments carried out in the framework of licensing and supervision procedures in the field of reactor physics. (authors)

Technical considerations for evaluating substantially complete containment of high-level waste within the waste package

Energy Technology Data Exchange (ETDEWEB)

Manaktala, H.K. (Southwest Research Inst., San Antonio, TX (USA). Center for Nuclear Waste Regulatory Analyses); Interrante, C.G. (Nuclear Regulatory Commission, Washington, DC (USA). Div. of High-Level Waste Management)

1990-12-01

This report deals with technical information that is considered essential for demonstrating the ability of the high-level radioactive waste package to provide substantially complete containment'' of its contents (vitrified waste form or spent light-water reactor fuel) for a period of 300 to 1000 years in a geological repository environment. The discussion is centered around technical considerations of the repository environment, materials and fabrication processes for the waste package components, various degradation modes of the materials of construction of the waste packages, and inspection and monitoring of the waste package during the preclosure and retrievability period, which could begin up to 50 years after initiation of waste emplacement. The emphasis in this report is on metallic materials. However, brief references have been made to other materials such as ceramics, graphite, bonded ceramic-metal systems, and other types of composites. The content of this report was presented to an external peer review panel of nine members at a workshop held at the Center for Nuclear Waste Regulatory Analyses (CNWRA), Southwest Research Institute, San Antonio, Texas, April 2--4, 1990. The recommendations of the peer review panel have been incorporated in this report. There are two companion reports; the second report in the series provides state-of-the-art techniques for uncertainty evaluations. 97 refs., 1 fig.

Technical considerations for evaluating substantially complete containment of high-level waste within the waste package

International Nuclear Information System (INIS)

Manaktala, H.K.; Interrante, C.G.

1990-12-01

This report deals with technical information that is considered essential for demonstrating the ability of the high-level radioactive waste package to provide ''substantially complete containment'' of its contents (vitrified waste form or spent light-water reactor fuel) for a period of 300 to 1000 years in a geological repository environment. The discussion is centered around technical considerations of the repository environment, materials and fabrication processes for the waste package components, various degradation modes of the materials of construction of the waste packages, and inspection and monitoring of the waste package during the preclosure and retrievability period, which could begin up to 50 years after initiation of waste emplacement. The emphasis in this report is on metallic materials. However, brief references have been made to other materials such as ceramics, graphite, bonded ceramic-metal systems, and other types of composites. The content of this report was presented to an external peer review panel of nine members at a workshop held at the Center for Nuclear Waste Regulatory Analyses (CNWRA), Southwest Research Institute, San Antonio, Texas, April 2--4, 1990. The recommendations of the peer review panel have been incorporated in this report. There are two companion reports; the second report in the series provides state-of-the-art techniques for uncertainty evaluations. 97 refs., 1 fig

49 CFR 173.24 - General requirements for packagings and packages.

Science.gov (United States)

2010-10-01

... identifiable (without the use of instruments) release of hazardous materials to the environment; (2) The effectiveness of the package will not be substantially reduced; for example, impact resistance, strength... significant chemical or galvanic reaction between the materials and contents of the package. (3) Plastic...

Packaging systems for animal origin food

Directory of Open Access Journals (Sweden)

2011-03-01

Full Text Available The main task of food packaging is to protect the product during storage and transport against the action of biological, chemical and mechanical factors. The paper presents packaging systems for food of animal origin. Vacuum and modified atmosphere packagings were characterised together with novel types of packagings, referred to as intelligent packaging and active packaging. The aim of this paper was to present all advantages and disadvantages of packaging used for meat products. Such list enables to choose the optimal type of packaging for given assortment of food and specific conditions of the transport and storing.

Wastes from selected activities in two light-water reactor fuel cycles

International Nuclear Information System (INIS)

Palmer, C.R.; Hill, O.F.

1980-07-01

This report presents projected volumes and radioactivities of wastes from the production of electrical energy using light-water reactors (LWR). The projections are based upon data developed for a recent environmental impact statement in which the transuranic wastes (i.e., those wastes containing certain long-lived alpha emitters at concentrations of at least 370 becquerels, or 10 nCi, per gram of waste) from fuel cycle activities were characterized. In addition, since the WG.7 assumed that all fuel cycle wastes except mill tailings are placed in a mined geologic repository, the nontransuranic wastes from several activities are included in the projections reported. The LWR fuel cycles considered are the LWR, once-through fuel cycle (Strategy 1), in which spent fuel is packaged in metal canisters and then isolated in geologic formations; and the LWR U/Pu recycle fuel cycle (Strategy 2), wherein spent fuel is reprocessed for recovery and recycle of uranium and plutonium in LWRs. The wastes projected for the two LWR fuel cycles are summarized. The reactor operations and decommissioning were found to dominate the rate of waste generation in each cycle. These activities account for at least 85% of the fuel cycle waste volume (not including head-end wastes) when normalized to per unit electrical energy generated. At 10 years out of reactor, however, spent fuel elements in Strategy 1 represent 98% of the fuel cycle activity but only 4% of the volume. Similarly, the packaged high-level waste, fuel hulls and hardware in Strategy 2 concentrate greater than 95% of the activity in 2% of the waste volume

ITER [International Thermonuclear Experimental Reactor] shield and blanket work package report

International Nuclear Information System (INIS)

1988-06-01

This report summarizes nuclear-related work in support of the US effort for the International Thermonuclear Experimental Reactor (ITER) Study. The purpose of this work was to prepare for the first international ITER workshop devoted to defining a basic ITER concept that will serve as a basis for an indepth conceptual design activity over the next 2-1/2 years. Primary tasks carried out during the past year included: design improvements of the inboard shield developed for the TIBER concept, scoping studies of a variety of tritium breeding blanket options, development of necessary design guidelines and evaluation criteria for the blanket options, further safety considerations related to nuclear components and issues regarding structural materials for an ITER device. 44 refs., 31 figs., 29 tabs

Merganser Download Package

Data.gov (United States)

U.S. Environmental Protection Agency — This data download package contains an Esri 10.0 MXD, file geodatabase and copy of this FGDC metadata record. The data in this package are used in support of the...

Modular assembly of chimeric phi29 packaging RNAs that support DNA packaging.

Science.gov (United States)

Fang, Yun; Shu, Dan; Xiao, Feng; Guo, Peixuan; Qin, Peter Z

2008-08-08

The bacteriophage phi29 DNA packaging motor is a protein/RNA complex that can produce strong force to condense the linear-double-stranded DNA genome into a pre-formed protein capsid. The RNA component, called the packaging RNA (pRNA), utilizes magnesium-dependent inter-molecular base-pairing interactions to form ring-shaped complexes. The pRNA is a class of non-coding RNA, interacting with phi29 motor proteins to enable DNA packaging. Here, we report a two-piece chimeric pRNA construct that is fully competent in interacting with partner pRNA to form ring-shaped complexes, in packaging DNA via the motor, and in assembling infectious phi29 virions in vitro. This is the first example of a fully functional pRNA assembled using two non-covalently interacting fragments. The results support the notion of modular pRNA architecture in the phi29 packaging motor.

The art of packaging: An investigation into the role of color in packaging, marketing, and branding

Directory of Open Access Journals (Sweden)

Behzad Mohebbi

2014-12-01

Full Text Available The purpose of this study is to contribute to the existing research in the field of packaging and marketing and shed more light on the psychology of colors and their effect on packaging and marketing. Nowadays, packaging is proved to be one of the significant factors in the success of promoting product sale. However, there is a perceived gap with respect to the different aspects of packaging, in particular the graphics, design, and color of packaging. The current study provides a comprehensive overview of packaging. It elaborates on different aspects of packaging and summarizes the findings of the most recent research conducted to date probing into packaging from different perspectives. It also discusses the role of color, i.e., the psychology of colors, and graphics in packaging and product sale. It is argued that graphics and color play key roles in promoting product sale and designers and marketers should attach a great deal of importance to color in packaging. The implications for producers, marketers, practitioners, and researchers are discussed in detail and suggestions for future research are provided.

MINUIT package parallelization and applications using the RooFit package

International Nuclear Information System (INIS)

Lazzaro, Alfio; Moneta, Lorenzo

2010-01-01

The fitting procedures are based on numerical minimization of functions. The MINUIT package is the most common package used for such procedures in High Energy Physics community. The main algorithm in this package, MIGRAD, searches the minimum of a function using the gradient information. For each minimization iteration, MIGRAD requires the calculation of the derivative for each free parameter of the function to be minimized. Minimization is required for data analysis problems based on the maximum likelihood technique. The calculation of complex likelihood functions, with several free parameters, many independent variables and large data samples, can be very CPU-time consuming. Then, the minimization process requires the calculation of the likelihood functions several times for each minimization iteration. In this paper we will show how MIGRAD algorithm and the likelihood function calculation can be easily parallelized using Message Passing Interface techniques. We will present the speed-up improvements obtained in typical physics applications such as complex maximum likelihood fits using the RooFit package.

Waste package performance analysis

International Nuclear Information System (INIS)

Lester, D.H.; Stula, R.T.; Kirstein, B.E.

1982-01-01

A performance assessment model for multiple barrier packages containing unreprocessed spent fuel has been applied to several package designs. The resulting preliminary assessments were intended for use in making decisions about package development programs. A computer model called BARIER estimates the package life and subsequent rate of release of selected nuclides. The model accounts for temperature, pressure (and resulting stresses), bulk and localized corrosion, and nuclide retardation by the backfill after water intrusion into the waste form. The assessment model assumes a post-closure, flooded, geologic repository. Calculations indicated that, within the bounds of model assumptions, packages could last for several hundred years. Intact backfills of appropriate design may be capable of nuclide release delay times on the order of 10 7 yr for uranium, plutonium, and americium. 8 references, 6 figures, 9 tables

Comparative Packaging Study

Science.gov (United States)

Perchonok, Michele H.; Oziomek, Thomas V.

2009-01-01

Future long duration manned space flights beyond low earth orbit will require the food system to remain safe, acceptable and nutritious. Development of high barrier food packaging will enable this requirement by preventing the ingress and egress of gases and moisture. New high barrier food packaging materials have been identified through a trade study. Practical application of this packaging material within a shelf life test will allow for better determination of whether this material will allow the food system to meet given requirements after the package has undergone processing. The reason to conduct shelf life testing, using a variety of packaging materials, stems from the need to preserve food used for mission durations of several years. Chemical reactions that take place during longer durations may decrease food quality to a point where crew physical or psychological well-being is compromised. This can result in a reduction or loss of mission success. The rate of chemical reactions, including oxidative rancidity and staling, can be controlled by limiting the reactants, reducing the amount of energy available to drive the reaction, and minimizing the amount of water available. Water not only acts as a media for microbial growth, but also as a reactant and means by which two reactants may come into contact with each other. The objective of this study is to evaluate three packaging materials for potential use in long duration space exploration missions.

Reactor decommissioning strategy: a new start for BNFL

International Nuclear Information System (INIS)

Woollam, P.; Nurden, P.

2001-01-01

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

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

International Nuclear Information System (INIS)

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

1997-05-01

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

Intelligent food packaging - research and development

Directory of Open Access Journals (Sweden)

Renata Dobrucka

2015-03-01

Full Text Available Packaging also fosters effective marketing of the food through distribution and sale channels. It is of the utmost importance to optimize the protection of the food, a great quality and appearance - better than typical packaged foods. In recent years, intelligent packaging became very popular. Intelligent packaging is becoming more and more widely used for food products. Application of this type of solution contributes to improvement of the quality consumer life undoubtedly. Intelligent packaging refers to a package that can sense environmental changes, and in turn, informs the users about the changes. These packaging systems contain devices that are capable of sensing and providing information about the functions and properties of the packaged foods. Also, this paper will review intelligent packaging technologies and describe different types of indicators (time-temperature indicators, freshness indicators.

Creative Thinking Package

Science.gov (United States)

Jones, Clive

1972-01-01

A look at the latest package from a British managment training organization, which explains and demonstrates creative thinking techniques, including brainstorming. The package, designed for groups of twelve or more, consists of tapes, visuals, and associated exercises. (Editor/JB)

Perspectives on the Elements of Packaging Design : A Qualitative Study on the Communication of Packaging

OpenAIRE

Alervall, Viktoria; Saied, Juan Sdiq

2013-01-01

Background: In today’s markets almost all products we buy come packaged. We use packaging to protect, contain and identify products. Furthermore if this is executed in a skillful way consumers often choose products based on packaging. The work of a designer and marketer is therefore extremely valuable when it comes to the design of a package. Problem: How are packages used to communicate marketing information? Purpose: The focus of this thesis is to identify differences and similarities of a ...

21 CFR 355.20 - Packaging conditions.

Science.gov (United States)

2010-04-01

... conditions. (a) Package size limitation. Due to the toxicity associated with fluoride active ingredients, the... (toothpastes and tooth powders) packages shall not contain more than 276 milligrams (mg) total fluorine per... packages shall not contain more than 120 mg total fluorine per package. (3) Exception. Package size...

19 CFR 191.13 - Packaging materials.

Science.gov (United States)

2010-04-01

... 19 Customs Duties 2 2010-04-01 2010-04-01 false Packaging materials. 191.13 Section 191.13 Customs... (CONTINUED) DRAWBACK General Provisions § 191.13 Packaging materials. (a) Imported packaging material... packaging material when used to package or repackage merchandise or articles exported or destroyed pursuant...

Warpage of QFN Package in Post Mold Cure Process of integrated circuit packaging

Science.gov (United States)

Sriwithoon, Nattha; Ugsornrat, Kessararat; Srisuwitthanon, Warayoot; Thonglor, Panakamon

2017-09-01

This research studied about warpage of QFN package in post mold cure process of integrated circuit (IC) packages using pre-plated (PPF) leadframe. For IC package, epoxy molding compound (EMC) are molded by cross linking of compound stiffness but incomplete crosslinked network and leading the fully cured thermoset by post mold cure (PMC) process. The cure temperature of PMC can change microstructure of EMC in term of stress inside the package and effect to warpage of the package due to coefficient of thermal expansion (CTE) between EMC and leadframe. In experiment, cure temperatures were varied to check the effect of internal stress due to different cure temperature after completed post mold cure for TDFN 2×3 8L. The cure temperature were varied with 180 °C, 170 °C, 160 °C, and 150°C with cure time 4 and 6 hours, respectively. For analysis, the TDFN 2×3 8L packages were analyzed the warpage by thickness gauge and scanning acoustic microscope (SAM) after take the test samples out from the oven cure. The results confirmed that effect of different CTE between EMC and leadframe due to different cure temperature resulting to warpage of the TDFN 2×3 8L packages.

CYPROS - Cybernetic Program Packages

Directory of Open Access Journals (Sweden)

Arne Tyssø

1980-10-01

Full Text Available CYPROS is an interactive program system consisting of a number of special purpose packages for simulation, identification, parameter estimation and control system design. The programming language is standard FORTRAN IV and the system is implemented on a medium size computer system (Nord-10. The system is interactive and program control is obtained by the use of numeric terminals. Output is rapidly examined by extensive use of video colour graphics. The subroutines included in the packages are designed and documented according to standardization rules given by the SCL (Scandinavian Control Library organization. This simplifies the exchange of subroutines throughout the SCL system. Also, this makes the packages attractive for implementation by industrial users. In the simulation package, different integration methods are available and it can be easily used for off-line, as well as real time, simulation problems. The identification package consists of programs for single-input/single-output and multivariablc problems. Both transfer function models and state space models can be handled. Optimal test signals can be designed. The control package consists of programs based on multivariable time domain and frequency domain methods for analysis and design. In addition, there is a package for matrix and time series manipulation. CYPROS has been applied successfully to industrial problems of various kinds, and parts of the system have already been implemented on different computers in industry. This paper will, in some detail, describe the use and the contents of the packages and some examples of application will be discussed.

Characterization of integrated circuit packaging materials

CERN Document Server

Moore, Thomas

1993-01-01

Chapters in this volume address important characteristics of IC packages. Analytical techniques appropriate for IC package characterization are demonstrated through examples of the measurement of critical performance parameters and the analysis of key technological problems of IC packages. Issues are discussed which affect a variety of package types, including plastic surface-mount packages, hermetic packages, and advanced designs such as flip-chip, chip-on-board and multi-chip models.

Reactor scale modeling of multi-walled carbon nanotube growth

International Nuclear Information System (INIS)

Lombardo, Jeffrey J.; Chiu, Wilson K.S.

2011-01-01

As the mechanisms of carbon nanotube (CNT) growth becomes known, it becomes important to understand how to implement this knowledge into reactor scale models to optimize CNT growth. In past work, we have reported fundamental mechanisms and competing deposition regimes that dictate single wall carbon nanotube growth. In this study, we will further explore the growth of carbon nanotubes with multiple walls. A tube flow chemical vapor deposition reactor is simulated using the commercial software package COMSOL, and considered the growth of single- and multi-walled carbon nanotubes. It was found that the limiting reaction processes for multi-walled carbon nanotubes change at different temperatures than the single walled carbon nanotubes and it was shown that the reactions directly governing CNT growth are a limiting process over certain parameters. This work shows that the optimum conditions for CNT growth are dependent on temperature, chemical concentration, and the number of nanotube walls. Optimal reactor conditions have been identified as defined by (1) a critical inlet methane concentration that results in hydrogen abstraction limited versus hydrocarbon adsorption limited reaction kinetic regime, and (2) activation energy of reaction for a given reactor temperature and inlet methane concentration. Successful optimization of a CNT growth processes requires taking all of those variables into account.

Capability to model reactor regulating system in RFSP

Energy Technology Data Exchange (ETDEWEB)

Chow, H C; Rouben, B; Younis, M H; Jenkins, D A [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Baudouin, A [Hydro-Quebec, Montreal, PQ (Canada); Thompson, P D [New Brunswick Electric Power Commission, Point Lepreau, NB (Canada). Point Lepreau Generating Station

1996-12-31

The Reactor Regulating System package extracted from SMOKIN-G2 was linked within RFSP to the spatial kinetics calculation. The objective is to use this new capability in safety analysis to model the actions of RRS in hypothetical events such as in-core LOCA or moderator drain scenarios. This paper describes the RRS modelling in RFSP and its coupling to the neutronics calculations, verification of the RRS control routine functions, sample applications and comparisons to SMOKIN-G2 results for the same transient simulations. (author). 7 refs., 6 figs.

The ZOOM minimization package

International Nuclear Information System (INIS)

Fischler, Mark S.; Sachs, D.

2004-01-01

A new object-oriented Minimization package is available for distribution in the same manner as CLHEP. This package, designed for use in HEP applications, has all the capabilities of Minuit, but is a re-write from scratch, adhering to modern C++ design principles. A primary goal of this package is extensibility in several directions, so that its capabilities can be kept fresh with as little maintenance effort as possible. This package is distinguished by the priority that was assigned to C++ design issues, and the focus on producing an extensible system that will resist becoming obsolete

Two-dimensional calculation by finite element method of velocity field and temperature field development in fast reactor fuel assembly. II

International Nuclear Information System (INIS)

Schmid, J.

1985-11-01

A package of updated computer codes for velocity and temperature field calculations for a fast reactor fuel subassembly (or its part) by the finite element method is described. Isoparametric triangular elements of the second degree are used. (author)

Packaging Concerns/Techniques for Large Devices

Science.gov (United States)

Sampson, Michael J.

2009-01-01

This slide presentation reviews packaging challenges and options for electronic parts. The presentation includes information about non-hermetic packages, space challenges for packaging and complex package variations.

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-07-01

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

Divertor modelling for conceptual studies of tokamak fusion reactor FDS-III

International Nuclear Information System (INIS)

Chen Yiping; Liu Songlin

2010-01-01

Divertor modelling for the conceptual studies of tokamak fusion reactor FDS-III was carried out by using the edge plasma code package B2.5-Eirene (SOLPS5.0). The modelling was performed by taking real MHD equilibrium and divertor geometry of the reactor into account. The profiles of plasma temperature, density and heat fluxes in the computational region and at the target plates have been obtained. The modelling results show that, with the fusion power P fu =2.6 GW and the edge density N edge =6.0x10 19 l/m 3 , the peak values of electron and ion heat fluxes at the outer target plate of divertor are respectively 93.92 MW/m 2 and 58.50 MW/m 2 . According to the modelling results it is suggested that some methods for reducing the heat fluxes at the target plates should be used in order to get acceptable level of power flux at the target plates for the divertor design of the reactor.

Evaluation on radioactive waste disposal amount of Kori Unit 1 reactor vessel considering cutting and packaging methods

International Nuclear Information System (INIS)

Choi, Yu Jong; Lee, Seong Cheol; Kim, Chang Lak

2016-01-01

Decommissioning of nuclear power plants has become a big issue in South Korea as some of the nuclear power plants in operation including Kori unit 1 and Wolsung unit 1 are getting old. Recently, Wolsung unit 1 received permission to continue operation while Kori unit 1 will shut down permanently in June 2017. With the consideration of segmentation method and disposal containers, this paper evaluated final disposal amount of radioactive waste generated from decommissioning of the reactor pressure vessel in Kori unit 1 which will be decommissioned as the first in South Korea. The evaluation results indicated that the final disposal amount from the top and bottom heads of the reactor pressure vessel with hemisphere shape decreased as they were cut in smaller more effectively than the cylindrical part of the reactor pressure vessel. It was also investigated that 200 L and 320 L radioactive waste disposal containers used in Kyung-Ju disposal facility had low payload efficiency because of loading weight limitation

LOFA and RIA analysis of the Indonesian Multipurpose research reactor RSG-GAS 1)

International Nuclear Information System (INIS)

Endiah Puji Hastuti; Hudi Hastowo; Iman Kuntoro

1999-01-01

Investigation on accident of the Indonesian Multipurpose research reactor RSG-GAS has been performed by computer simulation technique. Two groups of transients were considered, namely transient due to loss of primary cooling system (LOFA) and power excursion due to reactivity insertion (RIA). In such a transient condition, the Common Mode Failure (CMF) is considered and it will induce a situation so called unprotected transient or Anticipated Transient Without Scram (ATWS). RELAP5, PARET-ANL and EUREKA-2RR computer packages have been applied for these analyses. Simulations result done using these computer packages showed that in the occurrence of LOFA and RIA, failure on fuel elements is limited to the region with the highest power factor. (author)

Numerical benchmark for the deep-burn modular helium-cooled reactor (DB-MHR)

International Nuclear Information System (INIS)

Taiwo, T. A.; Kim, T. K.; Buiron, L.; Varaine, F.

2006-01-01

Numerical benchmark problems for the deep-burn concept based on the prismatic modular helium-cooled reactor design (a Very High Temperature Reactor (VHTR)) are specified for joint analysis by U.S. national laboratories and industry and the French CEA. The results obtained with deterministic and Monte Carlo codes have been inter-compared and used to confirm the underlying feature of the DB-MHR concept (high transuranics consumption). The results are also used to evaluate the impact of differences in code methodologies and nuclear data files on the code predictions for DB-MHR core physics parameters. The code packages of the participating organizations (ANL and CEA) are found to give very similar results. (authors)

Development of a transport cask for spent fuel elements of research reactors

International Nuclear Information System (INIS)

Quintana, F.; Saliba, R.O.; Furnari, J.C.; Mourao, R.P; Leite da Silva, L.; Novara, O.; Alexandre Miranda, C.; Mattar Neto, M.

2012-01-01

This article presents an overview of the development of a research reactor spent fuel transport cask. Through a project funded by the IAEA, Argentina, Brazil and Chile have collaborated to enhance regional capacity in the management of spent fuel elements from research reactors operated in the region. A packaging for the transport of research reactors spent fuel was developed. It was designed by a team of researchers from the countries mentioned and a 1:2 scale model for MTR type fuel was constructed in Argentina and subsequently tested in CDTN facilities in Belo Horizonte, Brazil. There were three test sequences to test the cask for normal transport and hypothetical accident conditions. It has successfully passed the tests and the overall performance was considered satisfactory. As part of the licensing process, a test sequence with the presence of regulatory authorities is scheduled for December, 2012 (author)

Reactor refurbishment in an outage environment

International Nuclear Information System (INIS)

Gowthorpe, P.; Hoare, R.

2012-01-01

Reactor life extension has typically been performed during specific refurbishment outages. These outages are long and costly due to the sheer complexity of the scope, not to mention the ever present discovery work. A scope of this size requires a huge labour force to execute, which poses significant challenges. The work is difficult to staff with qualified people able to execute the work smoothly and managing the required labour pool problematic. Cost and time overruns are inevitable in that environment. Reducing the cost and schedule is critical to the long term viability of reactor refurbishment projects. With planning, the total cost of the refurbishment can be reduced by managing the inspection and repairs during normal outages. Identifying what activities need to be done each outage for the life of the reactor and bringing the latest technology can make this viable. Tightly planned outages with a small well trained labour force will go a long way to reducing costs. The suite of services and tooling available to the utilities to manage their reactor integrity has improved significantly in recent years and continues to evolve. New feeder inspection technologies can provide improved inspection results for the complex feeder geometry. These improvements lead to more accurate wear rates and better predictions of component life. Feeders that need replacement based on improved inspection techniques can be replaced systematically during regular outages rather than specific refurbishment outages. Targeting areas rather than entire feeders reduces time, dose and cost. In cases where feeder replacement isn't feasible or where unpredicted wear is found, a feeder weld overlay process can be used. To manage the reactor work, new data systems are under development that allow for effective tracking of each activity performed and outcomes in a single package. (author)

Packaging and transport of radioisotopes

International Nuclear Information System (INIS)

Taylor, C.B.G.

1976-01-01

The importance of radioisotope traffic is emphasized. More than a million packages are being transported each year, mostly for medical uses. The involvement of public transport services and the incidental dose to the public (which is very small) are appreciably greater than for movements connected with the nuclear fuel cycle. Modern isotope packages are described, and an outline given of the problems of a large radioisotope manufacturer who has to package many different types of product. Difficulties caused by recent uncoordinated restrictions on the use of passenger aircraft are mentioned. Some specific problems relating to radioisotope packaging are discussed. These include the crush resistance of Type A packages, the closure of steel drums, the design of secure closures for large containers, the Type A packaging of liquids, leak tightness criteria of Type B packages, and the use of 'unit load' overpacks to consign a group of individually approved packages together as a single shipment. Reference is made to recent studies of the impact of radioisotope shipments on the environment. Cost/benefit analysis is important in this field - an important public debate is only just beginning. (author)

Safety Analysis Report - Packages, 9965, 9968, 9972-9975 Packages

International Nuclear Information System (INIS)

Van Alstine, M.N.

1999-01-01

This Safety Analysis Report for Packaging (SARP) documents the performance of the 9965 B, 9968 B, 9972 B(U), 9973 B(U), 9974 B(U), and 9975 B(U) packages in satisfying the regulatory safety requirements of the Code of Federal Regulations (CFR) 711 and the International Atomic Energy Agency (IAEA) Safety Series No. 6, Regulations for the Safe Transport of Radioactive Material, 1985 edition2. Results of the analysis and testing performed on the 9965 B, 9968 B, 9972 B(U), 9973 B(U), 9974 B(U), and 9975 B(U) packages are presented in this SARP, which was prepared in accordance with U.S. Department of energy (DOE) Order 5480.33 and in the format specified in the Nuclear Regulatory Commission (NRC) Regulatory Guides 7.94 and 7.10.5

Safety analysis report - packages 9965, 9968, 9972-9975 packages

International Nuclear Information System (INIS)

Van Alstine, M.N.

1997-10-01

This Safety Analysis Report for Packaging (SARP) documents the performance of the 9965 B( ), 9968 B( ), 9972 B(U), 9973 B(U), 9974 B(U), and 9975 B(U) packages in satisfying the regulatory safety requirements of the Code of Federal Regulations (CFR) 10 CFR 71 and the International Atomic Energy Agency (IAEA) Safety Series No. 6, Regulations for the Safe Transport of Radioactive Material, 1985 edition. Results of the analysis and testing performed on the 9965 B(), 9968 B(), 9972 B(U), 9973 B(U), and 9975 B(U) packages are presented in this SARP, which was prepared in accordance with U.S. Department of Energy (DOE) Order 5480.3 and in the format specified in the Nuclear Regulatory Commission (NRC) Regulatory Guides 7.9 and 7.10

Healthy package, healthy product? Effects of packaging design as a function of purchase setting

NARCIS (Netherlands)

van Rompay, Thomas Johannes Lucas; Deterink, Florien; Fenko, Anna

2016-01-01

Inspired by research testifying to the influence of visual packaging appearance and meaning portrayal on food evaluation, here it is argued that effects of packaging design vary depending on purchase context. Realistic packaging variants for a fictitious yoghurt brand varying in health connotation

Integrated Approach to Industrial Packaging Design

Science.gov (United States)

Vorobeva, O.

2017-11-01

The article reviews studies in the field of industrial packaging design. The major factors which influence technological, ergonomic, economic and ecological features of packaging are established. The main modern trends in packaging design are defined, the principles of marketing communications and their influence on consumers’ consciousness are indicated, and the function of packaging as a transmitter of brand values is specified. Peculiarities of packaging technology and printing techniques in modern printing industry are considered. The role of designers in the stage-by-stage development of the construction, form and graphic design concept of packaging is defined. The examples of authentic packaging are given and the mention of the tetrahedron packaging history is made. At the end of the article, conclusions on the key research aspects are made.

Directory of certificates of compliance for radioactive materials packages. Summary report of NRC approved packages

International Nuclear Information System (INIS)

1979-10-01

This directory contains a Summary Report of NRC Approved Packages for radioactive material packages effective September 14, 1979. Purpose of this directory is to make available a convenient source of information on packagings which have been approved by the US Nuclear Regulatory Commission. To assist in identifying packaging, an index by Model Number and corresponding Certificate of Compliance number is included at the back of each volume of the directory. The Summary Report includes a listing of all users of each package design prior to the publication date of the directory

Directory of certificates of compliance for radioactive materials packages. Summary report of NRC approved packages

Energy Technology Data Exchange (ETDEWEB)

None

1979-10-01

This directory contains a Summary Report of NRC Approved Packages for radioactive material packages effective September 14, 1979. Purpose of this directory is to make available a convenient source of information on packagings which have been approved by the US Nuclear Regulatory Commission. To assist in identifying packaging, an index by Model Number and corresponding Certificate of Compliance number is included at the back of each volume of the directory. The Summary Report includes a listing of all users of each package design prior to the publication date of the directory.

Directory of Certificates of Compliance for Radioactive Materials Packages: Report of NRC approved packages

International Nuclear Information System (INIS)

1988-12-01

This directory contains a Report of the US Nuclear Regulatory Commission's Approved Packages (Volume 1), all Certificates of Compliance (Volume 2), and a Report of NRC Approved Quality Assurance Programs (Volume 3) for Radioactive Material Packages effective October 1, 1988. The purpose of this directory is to make available a convenient source of information on packagings which have been approved by the US Nuclear Regulatory Commission. To assist in identifying packaging, an index by Model Number and corresponding Certificate of Compliance Number is included at the front of Volume 2 of the directory. A listing by packaging types is included in the back of Volume 2. An alphabetical listing by Company name is included in the back of Volume 3 for approved QA programs. The Reports include a listing of all users of each package design and approved QA programs prior to the publication date of the directory

Directory of certificates of compliance for radioactive materials packages, Report of NRC approved packages

International Nuclear Information System (INIS)

1990-10-01

This directory contains a Report of the US Nuclear Regulatory Commission's Approved Packages (Volume 1), Certificates of Compliance (Volume 2), and a Report of NRC Approved Quality Assurance Programs (Volume 3) for Radioactive Materials Packages effective October 1, 1990. The purpose of this directory is to make available a convenient source of information on packagings which have been approved by the US Nuclear Regulatory Commission. To assist in identifying packaging, an index by Model Number and corresponding Certificate of Compliance Number is included at the front of Volume 2 of the directory. A listing by packaging types is included in the back of Volume 2. An alphabetical listing by Company name is included in the back of Volume 3 for approved QA programs. The Reports include a listing of all users of each package design and approved QA programs prior to the publication date of the directory

Safety analysis report for packages: packaging of fissile and other radioactive materials. Final report

International Nuclear Information System (INIS)

Chalfant, G.G.

1984-01-01

The 9965, 9966, 9967, and 9968 packages are designed for surface shipment of fissile and other radioactive materials where a high degree of containment (either single or double) is required. Provisions are made to add shielding material to the packaging as required. The package was physically tested to demonstrate that it meets the criteria specified in USDOE Order No. 5480.1, chapter III, dated 5/1/81, which invokes Title 10, Code of Federal Regulations, Part 71 (10 CFR 71), Packing and Transportation of Radioactive Material, and Title 49, Code of Federal Regulations, Part 100-179, Transportation. By restricting the maximum normal operating pressure of the packages to less than 7 kg/cm 2 (gauge) (99 to 54 psig), the packages will comply with Type B(U) regulations of the International Atomic Energy Agency (IAEA) in its Regulations for the Safe Transport of Radioactive Materials, Safety Series No. 6, 1973 Revised Edition, and may be used for export and import shipments. These packages have been assessed for transport of up to 14.5 kilograms of uranium, excluding uranium-233, or 4.4 kilograms of plutonium metal, oxides, or scrap having a maximum radioactive decay energy of 30 watts. Specific maximum package contents are given. This quantity and the configuration of uranium or plutonium metal cannot be made critical by any combination of hydrogeneous reflection and moderation regardless of the condition of the package. For a uranium-233 shipment, a separate criticality evaluation for the specific package is required

Ensuring socially responsible packaging design

DEFF Research Database (Denmark)

Geert Jensen, Birgitte

Most consumers have experienced occasional problems with opening packaging. Tomato sauce from the tinned mackerel splattered all over the kitchen counter, the unrelenting pickle jar lid, and the package of sliced ham that cannot be opened without a knife or a pair of scissors. The research project...... "User‐friendly Packaging" aims to create a platform for developing more user‐friendly packaging. One intended outcome of the project is a guideline that industry can use in development efforts. The project also points the way for more extended collaboration between companies and design researchers. How...... can design research help industry in packaging innovation?...

Packaging based on polymeric materials

Directory of Open Access Journals (Sweden)

Jovanović Slobodan M.

2005-01-01

Full Text Available In the past two years the consumption of common in the developed countries world wide (high tonnage polymers for packaging has approached a value of 50 wt.%. In the same period more than 50% of the packaging units on the world market were made of polymeric materials despite the fact that polymeric materials present 17 wt.% of all packaging materials. The basic properties of polymeric materials and their environmental and economical advantages, providing them such a position among packaging materials, are presented in this article. Recycling methods, as well as the development trends of polymeric packaging materials are also presented.

Advanced flip chip packaging

CERN Document Server

Lai, Yi-Shao; Wong, CP

2013-01-01

Advanced Flip Chip Packaging presents past, present and future advances and trends in areas such as substrate technology, material development, and assembly processes. Flip chip packaging is now in widespread use in computing, communications, consumer and automotive electronics, and the demand for flip chip technology is continuing to grow in order to meet the need for products that offer better performance, are smaller, and are environmentally sustainable. This book also: Offers broad-ranging chapters with a focus on IC-package-system integration Provides viewpoints from leading industry executives and experts Details state-of-the-art achievements in process technologies and scientific research Presents a clear development history and touches on trends in the industry while also discussing up-to-date technology information Advanced Flip Chip Packaging is an ideal book for engineers, researchers, and graduate students interested in the field of flip chip packaging.

Challenges in the Packaging of MEMS

Energy Technology Data Exchange (ETDEWEB)

Malshe, A.P.; Singh, S.B.; Eaton, W.P.; O' Neal, C.; Brown, W.D.; Miller, W.M.

1999-03-26

The packaging of Micro-Electro-Mechanical Systems (MEMS) is a field of great importance to anyone using or manufacturing sensors, consumer products, or military applications. Currently much work has been done in the design and fabrication of MEMS devices but insufficient research and few publications have been completed on the packaging of these devices. This is despite the fact that packaging is a very large percentage of the total cost of MEMS devices. The main difference between IC packaging and MEMS packaging is that MEMS packaging is almost always application specific and greatly affected by its environment and packaging techniques such as die handling, die attach processes, and lid sealing. Many of these aspects are directly related to the materials used in the packaging processes. MEMS devices that are functional in wafer form can be rendered inoperable after packaging. MEMS dies must be handled only from the chip sides so features on the top surface are not damaged. This eliminates most current die pick-and-place fixtures. Die attach materials are key to MEMS packaging. Using hard die attach solders can create high stresses in the MEMS devices, which can affect their operation greatly. Low-stress epoxies can be high-outgassing, which can also affect device performance. Also, a low modulus die attach can allow the die to move during ultrasonic wirebonding resulting to low wirebond strength. Another source of residual stress is the lid sealing process. Most MEMS based sensors and devices require a hermetically sealed package. This can be done by parallel seam welding the package lid, but at the cost of further induced stress on the die. Another issue of MEMS packaging is the media compatibility of the packaged device. MEMS unlike ICS often interface with their environment, which could be high pressure or corrosive. The main conclusion we can draw about MEMS packaging is that the package affects the performance and reliability of the MEMS devices. There is a

CH Packaging Program Guidance

International Nuclear Information System (INIS)

2006-01-01

The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' They further state: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant| (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations(CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations

CH Packaging Program Guidance

International Nuclear Information System (INIS)

2007-01-01

The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' They further state: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations

User friendly packaging

DEFF Research Database (Denmark)

Geert Jensen, Birgitte

2010-01-01

Most consumers have experienced occasional problems with opening packaging. Tomato sauce from the tinned mackerel splattered all over the kitchen counter, the unrelenting pickle jar lid, and the package of sliced ham that cannot be opened without a knife or a pair of scissors. The research project...

Packaging of control system software

International Nuclear Information System (INIS)

Zagar, K.; Kobal, M.; Saje, N.; Zagar, A.; Sabjan, R.; Di Maio, F.; Stepanov, D.

2012-01-01

Control system software consists of several parts - the core of the control system, drivers for integration of devices, configuration for user interfaces, alarm system, etc. Once the software is developed and configured, it must be installed to computers where it runs. Usually, it is installed on an operating system whose services it needs, and also in some cases dynamically links with the libraries it provides. Operating system can be quite complex itself - for example, a typical Linux distribution consists of several thousand packages. To manage this complexity, we have decided to rely on Red Hat Package Management system (RPM) to package control system software, and also ensure it is properly installed (i.e., that dependencies are also installed, and that scripts are run after installation if any additional actions need to be performed). As dozens of RPM packages need to be prepared, we are reducing the amount of effort and improving consistency between packages through a Maven-based infrastructure that assists in packaging (e.g., automated generation of RPM SPEC files, including automated identification of dependencies). So far, we have used it to package EPICS, Control System Studio (CSS) and several device drivers. We perform extensive testing on Red Hat Enterprise Linux 5.5, but we have also verified that packaging works on CentOS and Scientific Linux. In this article, we describe in greater detail the systematic system of packaging we are using, and its particular application for the ITER CODAC Core System. (authors)

SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor Critical Configurations

International Nuclear Information System (INIS)

Gauld, I.C.

2000-01-01

Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the Waste Package Design team at the Yucca Mountain Project in the US, who performed the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The CRCs involved relatively low-cycle burnups, and therefore contained a relatively high gadolinium poison content in the reactor assemblies. This report summarizes the data and methods used in analyzing the critical configurations and assesses the sensitivity of the results to some of the modeling approximations used to represent the gadolinium poison distribution within the assemblies. The KENO V.a calculations, performed using the SCALE 44GROUPNDF5 ENDF/B-V cross-section library, yield predicted k eff values within about 1% Δk/k relative to reactor measurements for the five CRCs using general 8-pin and 9-pin heterogeneous gadolinium poison pin assembly models

Refurbishment programme of the reactor and progress of work

International Nuclear Information System (INIS)

Astruc, J.M.

1992-01-01

During 20 years of operation, since its start-up the ILL there have been some problems, like ruptured heavy water collector, in the upper part of the reflector tank, replacement of all the beam tubes due to the evolution of the mechanical characteristics of the aluminium alloy under irradiation. Some days after regular shutdown for maintenance, an inspection of the internal elements of the reactor discovered cracks on the grids which ensure the regular flow of cooling water. The investigations showed that the cracks are due to a design fault, aggravated by the effects of mechanical fatigue on highly irradiated material. It was not possible to repair the cracked grid, and it had to be replaced. This involved the dismantling of the internals parts of the reactor tank. The reactor refurbishment programme was set up. It provides for the replacement of the reactor block, the coupling sleeves, the anti turbulence grids and the diffuser, and of the ancillary elements. The main items to be replaced are: the reactor block consisting of the reactor vessel and its cover, known as the 'upper structure'; the heavy water collectors; connecting sleeves between the reactor block and the flanges of the various beam tubes. These three items constitute the primary circuit in the swimming pool. It is also planned to replace some internal parts of the reactor tank, such as the beam-tubes, the grid and diffuser and the chimney. Some parts of the present reactor, which are not at the end of their life, would be reused, for instance the two cold sources, the safety rods, and some other pieces. The parts replaced would be cut up and packaged in accordance with current standards and disposed of. All items are in principle to be replaced by identical equipment. This concerns in particular performance, mechanical characteristics and the choice of materials. The replacement of the reactor block necessitates a complete dismantling of the equipment in the reactor block, and of the structures in

Sensory impacts of food-packaging interactions.

Science.gov (United States)

Duncan, Susan E; Webster, Janet B

2009-01-01

Sensory changes in food products result from intentional or unintentional interactions with packaging materials and from failure of materials to protect product integrity or quality. Resolving sensory issues related to plastic food packaging involves knowledge provided by sensory scientists, materials scientists, packaging manufacturers, food processors, and consumers. Effective communication among scientists and engineers from different disciplines and industries can help scientists understand package-product interactions. Very limited published literature describes sensory perceptions associated with food-package interactions. This article discusses sensory impacts, with emphasis on oxidation reactions, associated with the interaction of food and materials, including taints, scalping, changes in food quality as a function of packaging, and examples of material innovations for smart packaging that can improve sensory quality of foods and beverages. Sensory evaluation is an important tool for improved package selection and development of new materials.

Plutonium stabilization and packaging system

International Nuclear Information System (INIS)

1996-01-01

This document describes the functional design of the Plutonium Stabilization and Packaging System (Pu SPS). The objective of this system is to stabilize and package plutonium metals and oxides of greater than 50% wt, as well as other selected isotopes, in accordance with the requirements of the DOE standard for safe storage of these materials for 50 years. This system will support completion of stabilization and packaging campaigns of the inventory at a number of affected sites before the year 2002. The package will be standard for all sites and will provide a minimum of two uncontaminated, organics free confinement barriers for the packaged material

RH Packaging Operations Manual

International Nuclear Information System (INIS)

Washington TRU Solutions LLC

2003-01-01

This procedure provides operating instructions for the RH-TRU 72-B Road Cask, Waste Shipping Package. In this document, ''Packaging'' refers to the assembly of components necessary to ensure compliance with the packaging requirements (not loaded with a payload). ''Package'' refers to a Type B packaging that, with its radioactive contents, is designed to retain the integrity of its containment and shielding when subject to the normal conditions of transport and hypothetical accident test conditions set forth in 10 CFR Part 71. Loading of the RH 72-B cask can be done two ways, on the RH cask trailer in the vertical position or by removing the cask from the trailer and loading it in a facility designed for remote-handling (RH). Before loading the 72-B cask, loading procedures and changes to the loading procedures for the 72-B cask must be sent to CBFO at sitedocuments at wipp.ws for approval

MARS software package status

International Nuclear Information System (INIS)

Azhgirej, I.L.; Talanov, V.V.

2000-01-01

The MARS software package is intended for simulating the nuclear-electromagnetic cascades and the secondary neutrons and muons transport in the heterogeneous medium of arbitrary complexity in the magnetic fields presence. The inclusive approach to describing the particle production in the nuclear and electromagnetic interactions and by the unstable particles decay is realized in the package. The MARS software package was actively applied for solving various radiation physical problems [ru

CH Packaging Program Guidance

International Nuclear Information System (INIS)

Washington TRU Solutions LLC

2003-01-01

The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: ''each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.'' They further state: ''each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.'' Chapter 9.0 of the SARP charges the WIPP management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with 10 CFR 71.11. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. This document provides the instructions to be followed to operate, maintain, and test the TRUPACT-II and HalfPACT packaging. The intent of these instructions is to standardize operations. All users will follow these instructions or equivalent instructions that assure operations are safe and meet the requirements of the SARPs

CH Packaging Program Guidance

International Nuclear Information System (INIS)

Washington TRU Solutions LLC

2002-01-01

The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT Shipping Package, and directly related components. This document complies with the minimum requirements as specified in TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event there is a conflict between this document and the SARP or C of C, the SARP and/or C of C shall govern. C of Cs state: ''each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.'' They further state: ''each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.'' Chapter 9.0 of the SAR P charges the WIPP Management and Operation (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with 10 CFR 71.11. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. This document details the instructions to be followed to operate, maintain, and test the TRUPACT-II and HalfPACT packaging. The intent of these instructions is to standardize these operations. All users will follow these instructions or equivalent instructions that assure operations are safe and meet the requirements of the SARPs

German Approach for the Transport of Spent Fuel Packages after Interim Storage

International Nuclear Information System (INIS)

Wille, Frank; Wolff, Dietmar; Droste, Bernhard; Voelzke, Holger

2014-01-01

In Germany the concept of dry interim storage of spent nuclear fuel in dual purpose metal casks is implemented, currently for periods of up to 40 years. The casks being used have an approved package design in accordance with the international transport regulations. The license for dry storage is granted on the German Atomic Energy Act with respect to the recently (in 2012) revised 'Guidelines for dry cask storage of spent nuclear fuel and heat-generating waste' by the German Waste management Commission (ESK) which are very similar to the former RSK (reactor safety commission) guidelines. For transport on public routes between or after long term interim storage periods, it has to be ensured that the transport and storage casks fulfil the specifications of the transport approval or other sufficient properties which satisfy the proofs for the compliance of the safety objectives at that time. In recent years the validation period of transport approval certificates for manufactured, loaded and stored packages were discussed among authorities and applicants. A case dependent system of 3, 5 and 10 years was established. There are consequences for the safety cases in the Package Design Safety Report including evaluation of long term behavior of components and specific operating procedures of the package. Present research and knowledge concerning the long term behavior of transport and storage cask components have to be consulted as well as experiences from interim cask storage operations. Challenges in the safety assessment are e.g. the behavior of aged metal and elastomeric seals under IAEA test conditions to ensure that the results of drop tests can be transferred to the compliance of the safety objectives at the time of transport after the interim storage period (aged package). Assessment methods for the material compatibility, the behavior of fuel assemblies and the aging behavior of shielding parts are issues as well. This paper describes the state

Consumer response to packaging design

NARCIS (Netherlands)

Steenis, Nigel D.; Herpen, van Erica; Lans, van der Ivo A.; Ligthart, Tom N.; Trijp, van Hans C.M.

2017-01-01

Building on theories of cue utilization, this paper investigates whether and how packaging sustainability influences consumer perceptions, inferences and attitudes towards packaged products. A framework is tested in an empirical study among 249 students using soup products varying in packaging

Natural biopolimers in organic food packaging

Science.gov (United States)

Wieczynska, Justyna; Cavoski, Ivana; Chami, Ziad Al; Mondelli, Donato; Di Donato, Paola; Di Terlizzi, Biagio

2014-05-01

Concerns on environmental and waste problems caused by use of non-biodegradable and non-renewable based plastic packaging have caused an increase interest in developing biodegradable packaging using renewable natural biopolymers. Recently, different types of biopolymers like starch, cellulose, chitosan, casein, whey protein, collagen, egg white, soybean protein, corn zein, gelatin and wheat gluten have attracted considerable attention as potential food packaging materials. Recyclable or biodegradable packaging material in organic processing standards is preferable where possible but specific principles of packaging are not precisely defined and standards have to be assessed. There is evidence that consumers of organic products have specific expectations not only with respect to quality characteristics of processed food but also in social and environmental aspects of food production. Growing consumer sophistication is leading to a proliferation in food eco-label like carbon footprint. Biopolymers based packaging for organic products can help to create a green industry. Moreover, biopolymers can be appropriate materials for the development of an active surfaces designed to deliver incorporated natural antimicrobials into environment surrounding packaged food. Active packaging is an innovative mode of packaging in which the product and the environment interact to prolong shelf life or enhance safety or sensory properties, while maintaining the quality of the product. The work will discuss the various techniques that have been used for development of an active antimicrobial biodegradable packaging materials focusing on a recent findings in research studies. With the current focus on exploring a new generation of biopolymer-based food packaging materials with possible applications in organic food packaging. Keywords: organic food, active packaging, biopolymers , green technology

Physical and chemical characteristics of cashew nut flour stored and packaged with different packages

Directory of Open Access Journals (Sweden)

Bruna Carolina GADANI

Full Text Available Abstract The aim of this study was to evaluate the physical and chemical changes in cashew nut flour under different packaging and storage conditions. The flour samples were characterized according to their chemical composition, packaged in polystyrene trays associated with covering chloride polyvinyl (PVC, plastic pot of polyethylene terephthalate (PET, plastic packaging polyethylene (POL and without coating polystyrene trays (CON. All packages were stored for 225 days, evaluating every 45 days the moisture content of the flour as well as extracting their oil by the cold method for characterization on the following indexes: acidity, saponification, iodine, peroxide and refraction. There was slight and gradual increase in the moisture content from the 45th to the 225th day, especially for products packaged in PVC and without coating polystyrene trays (CON. The oil saponification indexes showed gradual decrease during storage, especially in flour without packaging (CON. There was a gradual increase in the iodine index until the 135th day, followed by decay. The acidity and peroxide indexes increased along with the storage time. However packages with lower oxygen and moisture permeability, such as PET and POL, minimized such changes, being therefore the most suitable for long-term storage.

Cobalt-60 production in CANDU power reactors

International Nuclear Information System (INIS)

Malkoske, G.R.; Norton, J.L.; Slack, J.

2002-01-01

MDS Nordion has been supplying cobalt-60 sources to industry for industrial and medical purposes since 1946. These cobalt-60 sources are used in many market and product segments, but are primarily used to sterilize single-use medical products including; surgical kits, gloves, gowns, drapes, and cotton swabs. Other applications include sanitization of cosmetics, microbial reduction of pharmaceutical raw materials, and food irradiation. The technology for producing the cobalt-60 isotope was developed by MDS Nordion and Atomic Energy of Canada Limited (AECL) almost 55 years ago using research reactors at the AECL Chalk River Laboratories in Ontario, Canada. The first cobalt-60 source produced for medical applications was manufactured by MDS Nordion and used in cancer therapy. The benefits of cobalt-60 as applied to medical product manufacturing, were quickly realized and the demand for this radioisotope quickly grew. The same technology for producing cobalt-60 in research reactors was then designed and packaged such that it could be conveniently transferred to a utility/power reactor. In the early 1970's, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production for industrial irradiation applications was initiated in the four Pickering A CANDU reactors. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology for producing cobalt-60 in additional CANDU reactors. CANDU is unique among the power reactors of the world, being heavy water moderated and fuelled with natural uranium. They are also designed and supplied with stainless steel adjusters, the primary function of which is to shape the neutron flux to optimize reactor power and fuel bum-up, and to provide excess reactivity needed to overcome xenon-135 poisoning following a reduction of power. The reactor is designed to develop full power output with all of the adjuster

CH Packaging Program Guidance

International Nuclear Information System (INIS)

2008-01-01

The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the pplication.' They further state: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations

CH Packaging Program Guidance

International Nuclear Information System (INIS)

2009-01-01

The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' They further state: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations

RH Packaging Program Guidance

International Nuclear Information System (INIS)

2008-01-01

The purpose of this program guidance document is to provide the technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package (also known as the 'RH-TRU 72-B cask') and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the C of C shall govern. The C of C states: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' It further states: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M and O) Contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8, 'Deliberate Misconduct.' Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, 'Packaging and Transportation of Radioactive Material,' certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21, 'Reporting of Defects and Noncompliance,' regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous

RH Packaging Program Guidance

International Nuclear Information System (INIS)

2006-01-01

The purpose of this program guidance document is to provide the technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the C of C shall govern. The C of C states: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' It further states: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M and O) Contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with 10 Code of Federal Regulations (CFR) 1.8, 'Deliberate Misconduct.' Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, 'Packaging and Transportation of Radioactive Material,' certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21, 'Reporting of Defects and Noncompliance,' regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these

Safety analysis report: packages. LP-12 tritium package (packaging of fissile and other radioactive materials). Final report

International Nuclear Information System (INIS)

Gates, A.A.; McCarthy, P.G.; Edl, J.W.

1975-05-01

Elemental tritium is shipped at low pressure in a stainless steel container (LP-12) within an aluminum vessel and surrounded by 3.9 in.-thick Celotex insulation in a steel drum. Information is presented on the packaging design, evaluation of the structural, thermal, containment, shielding, and criticality characteristics of the package, procedures for loading, unloading, transporting, and testing the LP-12, and quality assurance requirements. (U.S.)

40 CFR 157.27 - Unit packaging.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 23 2010-07-01 2010-07-01 false Unit packaging. 157.27 Section 157.27 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) PESTICIDE PROGRAMS PACKAGING REQUIREMENTS FOR PESTICIDES AND DEVICES Child-Resistant Packaging § 157.27 Unit packaging. Pesticide products...

CH Packaging Program Guidance

International Nuclear Information System (INIS)

2005-01-01

The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: ''each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.'' They further state: ''each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.'' Chapter 9.0 of the SARP charges the Waste Isolation Pilot Plant (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.

DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE, GENERATION IV REACTOR SYSTEMS

Energy Technology Data Exchange (ETDEWEB)

Mynatt Fred R.; Townsend, L.W.; Williamson, Martin; Williams, Wesley; Miller, Laurence W.; Khan, M. Khurram; McConn, Joe; Kadak, Andrew C.; Berte, Marc V.; Sawhney, Rapinder; Fife, Jacob; Sedler, Todd L.; Conway, Larry E.; Felde, Dave K.

2003-11-12

The purpose of this research project is to develop compact (100 to 400 MWe) Generation IV nuclear power plant design and layout concepts that maximize the benefits of factory-based fabrication and optimal packaging, transportation and siting. The reactor concepts selected were compact designs under development in the 2000 to 2001 period. This interdisciplinary project was comprised of three university-led nuclear engineering teams identified by reactor coolant type (water, gas, and liquid metal) and a fourth Industrial Engineering team. The reactors included a Modular Pebble Bed helium-cooled concept being developed at MIT, the IRIS water-cooled concept being developed by a team led by Westinghouse Electric Company, and a Lead-Bismuth-cooled concept developed by UT. In addition to the design and layout concepts this report includes a section on heat exchanger manufacturing simulations and a section on construction and cost impacts of proposed modular designs.

DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE, GENERATION IV REACTOR SYSTEMS

International Nuclear Information System (INIS)

Mynatt, Fred R.; Townsend, L.W.; Williamson, Martin; Williams, Wesley; Miller, Laurence W.; Khan, M. Khurram; McConn, Joe; Kadak, Andrew C.; Berte, Marc V.; Sawhney, Rapinder; Fife, Jacob; Sedler, Todd L.; Conway, Larry E.; Felde, Dave K.

2003-01-01

The purpose of this research project is to develop compact (100 to 400 MWe) Generation IV nuclear power plant design and layout concepts that maximize the benefits of factory-based fabrication and optimal packaging, transportation and siting. The reactor concepts selected were compact designs under development in the 2000 to 2001 period. This interdisciplinary project was comprised of three university-led nuclear engineering teams identified by reactor coolant type (water, gas, and liquid metal) and a fourth Industrial Engineering team. The reactors included a Modular Pebble Bed helium-cooled concept being developed at MIT, the IRIS water-cooled concept being developed by a team led by Westinghouse Electric Company, and a Lead-Bismuth-cooled concept developed by UT. In addition to the design and layout concepts this report includes a section on heat exchanger manufacturing simulations and a section on construction and cost impacts of proposed modular designs

40 CFR 262.30 - Packaging.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 25 2010-07-01 2010-07-01 false Packaging. 262.30 Section 262.30... APPLICABLE TO GENERATORS OF HAZARDOUS WASTE Pre-Transport Requirements § 262.30 Packaging. Before... the waste in accordance with the applicable Department of Transportation regulations on packaging...

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

International Nuclear Information System (INIS)

Benest, T.G.

2008-01-01

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

Project monitoring package (PMP) : A package for project activity monitoring

International Nuclear Information System (INIS)

Vyas, K.N.; Kannan, A.; Susandhi, R.; Basu, S.

1987-01-01

A package for preparing PERT/CPM network diagrams has been written for PDP-11/34. The program uses PLOT-10 library calls for device interfacing. The package is essentially non-interactive in nature, and reads input data in the form of activity description and duration. It calculates the critical path time and performs time scaling of the events. The report gives a brief outline of the logic used, a sample plot and tabular output for reference. An additional facility for performing project activity monitoring has also been implemented. Activity monitoring generally requires various reports such as feed back reports from various group co-ordinators, information report for project co-ordinator and brief periodical reports for management. A package 'DATATRIEVE' (DTR) on PDP-11/34 system is utilized for generating the above mentioned reports. As DTR can also use normal sequential files, an interfacing program has been written which reformats the files accepted by PERT program acceptable to DTR. Various types of reports as generated by DTR are included. However this part of the package is not transportable and can be implemented only on systems having DTR. 6 figures. (author)

Perfume Packaging, Seduction and Gender

Directory of Open Access Journals (Sweden)

Magdalena Petersson McIntyre

2013-06-01

Full Text Available This article examines gender and cultural sense-making in relation to perfumes and their packaging. Gendered meanings of seduction, choice, consumption and taste are brought to the fore with the use of go-along interviews with consumers in per-fume stores. Meeting luxury packages in this feminized environment made the interviewed women speak of bottles as objects to fall in love with and they de-scribed packages as the active part in an act of seduction where they were expect-ing packages to persuade them into consumption. The interviewed men on the other hand portrayed themselves as active choice-makers and stressed that they were always in control and not seduced by packaging. However, while their ways of explaining their relationship with packaging on the surface seems to confirm cultural generalizations in relation to gender and seduction, the article argues that letting oneself be seduced is no less active than seducing. Based on a combination of actor network theories and theories of gender performativity the article points to the agency of packaging for constructions of gender and understands the inter-viewees as equally animated by the flows of passion which guide their actions.

Naval Waste Package Design Sensitivity

International Nuclear Information System (INIS)

T. Schmitt

2006-01-01

The purpose of this calculation is to determine the sensitivity of the structural response of the Naval waste packages to varying inner cavity dimensions when subjected to a comer drop and tip-over from elevated surface. This calculation will also determine the sensitivity of the structural response of the Naval waste packages to the upper bound of the naval canister masses. The scope of this document is limited to reporting the calculation results in terms of through-wall stress intensities in the outer corrosion barrier. This calculation is intended for use in support of the preliminary design activities for the license application design of the Naval waste package. It examines the effects of small changes between the naval canister and the inner vessel, and in these dimensions, the Naval Long waste package and Naval Short waste package are similar. Therefore, only the Naval Long waste package is used in this calculation and is based on the proposed potential designs presented by the drawings and sketches in References 2.1.10 to 2.1.17 and 2.1.20. All conclusions are valid for both the Naval Long and Naval Short waste packages

Radioactive material packaging performance testing

International Nuclear Information System (INIS)

Romano, T.

1992-06-01

In an effort to provide uniform packaging of hazardous material on an international level, recommendations for the transport of dangerous goods have been developed by the United Nations. These recommendations are performance oriented and contrast with a large number of packaging specifications in the US Department of Transportation's hazard materials regulations. This dual system presents problems when international shipments enter the US Department of Transportation's system. Faced with the question of continuing a dual system or aligning with the international system, the Research and Special Programs Administration of the US Department of Transportation responded with Docket HM-181. This began the transition toward the international transportation system. Following close behind is Docket HM-169A, which addressed low specific activity radioactive material packaging. This paper will discuss the differences between performance-oriented and specification packaging, the transition toward performance-oriented packaging by the US Department of Transportation, and performance-oriented testing of radioactive material packaging by Westinghouse Hanford Company. Dockets HM-181 and HM-169A will be discussed along with Type A (low activity) and Type B (high activity) radioactive material packaging evaluations

DTS : un logiciel d'aide à l'élaboration de modèles d'écoulement dans les réacteurs Dts: a Software Package for Flow Simulation in Reactors

Directory of Open Access Journals (Sweden)

Leclerc J. P.

2006-11-01

Full Text Available Cet article a pour but la présentation d'un logiciel d'aide à la construction de modèles d'écoulement à partir de la méthode des distributions de temps de séjour (DTS. La théorie des distributions de temps de séjour ainsi que des méthodes de traçage couramment utilisées sont rappelées. Une description du logiciel est ensuite effectuée. Le modèle est basé sur l'association de modules élémentaires (réacteur parfaitement agité, réacteur piston. . . , la détermination de la fonction de transfert étant effectuée en résolvant les bilans de matière dans l'espace de Laplace. Le logiciel peut donner la réponse à un signal d'entrée quelconque et plusieurs paramètres du modèle peuvent être optimisés. Différents couples de traceurs et de détecteurs sont donnés et les précautions à prendre lors des traçages sont décrites. Finalement, différentes études effectuées à l'aide du logiciel sont présentées afin de montrer les utilisations possibles. Improving the performance of an existing reactor or studying a new design requires modeling the flow of the different phases. Computational fluid dynamics can give an accurate a priori description of the velocity and concentration fields. However, this approach is too complicated in some cases (complicated geometry, random flow media, etc. , and this results in an abundance of information. A simpler approach relies on the theory of Residence Time Distribution (RTD, which is a theoretical interpretation framework for tracer experiments. This paper describes a software package that simulates hydrodynamic models derived from RTD experiments. The concept of RTD, introduced by Danckwerts in 1953, is briefly reviewed. Then tracer experiments, commonly used tracers and detectors are dealt with. Finally, the care required in tracer experiments is described in detail. Visual inspection of the tracer response makes it is possible to guess the main characteristics of the flow pattern

How Packaging Fleet Renewal Fits French CEA Programs

International Nuclear Information System (INIS)

Dumesnil, J.; Malvache, P.; Hugon, F.C.; Sollacaro, M.

2006-01-01

CEA's (French Atomic Energy Agency) packaging fleet is dedicated to transportation of test irradiated fuels, of research reactors fuels, of navy propulsion fuels, and of waste coming from and to nuclear plants or facilities. This fleet encompasses more than 30 types of casks ranging from 5 to 30 tons, with either recent designs or other dating back to the seventies. A study has been launched in order to perform a global analysis of the life expectancy of the existing CEA and COGEMA Logistics cask fleets with respect to a 2015 target, in order to anticipate its renewal, while limiting the number of type of cask. Key elements like periodical evolutions of design and transport regulations, lessons learnt of existing casks (design, approval and extensions, operational feedback, maintenance and dismantling) are taken into account in order to ensure compliance and availability of the fleet. Moreover, from design to cask delivery, including regulatory tests, safety analysis report/ CoC, and manufacturing, 3 to 5 years is needed. Therefore cask development should be taken into account earlier of invest and research's programs. The paper will address the current life expectancy study of CEA and COGEMA Logistics packaging fleet, based on lessons learnt and regulation evolution and on general R and D plans by user facilities. It will show how a comprehensive optimized fleet is made available to CEA and other customers. Such a fleet combines optimized investment and uses, thus entailing synergies for well-mastered costs of transports. (authors)

Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

Directory of Open Access Journals (Sweden)

Po Hu

2014-01-01

Full Text Available This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal.

Latest feedback from a major reactor vessel dismantling project

International Nuclear Information System (INIS)

Boucau, J.; Segerud, P.; Sanchez, M.; Garcia, R.

2015-01-01

Westinghouse performed two large segmentation projects in 2010-2013 and then 2013-2015 at the Jose Cabrera nuclear power plant in Spain. The power plant is located in Almonacid de Zorita, 43 miles east of Madrid, Spain and was in operation between 1968 and 2006. This paper will describe the sequential steps required to prepare, segment, separate, and package the individual component segments using under water mechanical techniques. The paper will also include experiences and lessons learned that Westinghouse has collected from the activities performed during the reactor vessel and vessel internals segmentation projects. (authors)

7 CFR 58.640 - Packaging.

Science.gov (United States)

2010-01-01

... 7 Agriculture 3 2010-01-01 2010-01-01 false Packaging. 58.640 Section 58.640 Agriculture Regulations of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (Standards... Procedures § 58.640 Packaging. The packaging of the semifrozen product shall be done by means which will in...

Emergency response packaging: A conceptual outline

International Nuclear Information System (INIS)

Luna, R.E.; McClure, J.D.; Bennett, P.C.; Wheeler, T.A.

1992-01-01

The Packaging and Transportation Needs in the 1990's (PATN) component of the Transportation Assessment and Integration (TRAIN) program (DOE Nov. 1991) was designed to survey United States Department of Energy programs, both ongoing and planned, to identify needs for packaging and transportation services over the next decade. PATN also identified transportation elements that should be developed by the DOE Office of Environmental Restoration and Waste Management (DOE EM) Transportation Management Program (TMP). As a result of the predominant involvement of the TMP in radioactive material shipment issues and DOE EM's involvement with waste management issues, the primary focus of PATN was on waste packaging issues. Pending DOE regulations will formalize federal guidelines and regulations for transportation of hazardous and radioactive materials within the boundaries of DOE reservations and facilities and reflect a growing awareness of concern regarding safety environmental responsibility activities on DOE reservations. Future practices involving the transportation of radioactive material within DOE reservations will closely parallel those used for commercial and governmental transportation across the United States. This has added to the perceived need for emergency recovery packaging and emergency response features on primary packaging, for both on-site shipments and shipments between DOE facilities (off-site). Historically, emergency response and recovery functions of packaging have not been adequately considered in packaging design and construction concepts. This paper develops the rationale for emergency response packaging, including both overpack concepts for repackaging compromised packaging and primary packaging redesign to facilitate the recovery of packages via mobile remote handling equipment. The rationale will examine concepts for determination of likely use patterns to identify types of shipments where recovery packaging may have the most favorable payoff

Sustainable packaging. Packaging for a circular economy; Duurzaam verpakken. Verpakken voor de circulaire economie

Energy Technology Data Exchange (ETDEWEB)

Haffmans, S. [Partners for Innovation, Amsterdam (Netherlands); Standhardt, G. [Nederlands Verpaskkingscentrum NVC, Gouda (Netherlands); Hamer, A. [Agentschap NL, Utrecht (Netherlands)

2013-10-15

What is Sustainable Packaging? And what is the most sustainable packaging for a product? The publication is intended for anyone who wants to take into account the environment in the design of a product and packaging. It offers concrete suggestions and inspiring examples to bring sustainable packaging into practice [Dutch] Wat is Duurzaam Verpakken? En wat is de duurzaamste verpakking voor mijn product? De publicatie is bestemd voor iedereen die rekening wil houden met het milieu bij het ontwerp van een product-verpakkingscombinatie. Ze biedt concrete aanknopingspunten en inspirerende voorbeelden om hier praktisch mee aan de slag te gaan.

Big rock point restoration project BWR major component removal, packaging and shipping - planning and experience

International Nuclear Information System (INIS)

Milner, T.; Dam, S.; Papp, M.; Slade, J.; Slimp, B.; Nurden, P.

2001-01-01

The Big Rock Point boiling water reactor (BWR) at Charlevoix, MI was permanently shut down on August 29th 1997. In 1999 BNFL Inc.'s Reactor Decommissioning Group (RDG) was awarded a contract by Consumers Energy (CECo) for the Big Rock Point (BRP) Major Component Removal (MCR) project. BNFL Inc. RDG has teamed with MOTA, Sargent and Lundy and MDM Services to plan and execute MCR in support of the facility restoration project. The facility restoration project will be completed by 2005. Key to the success of the project has been the integration of best available demonstrated technology into a robust and responsive project management approach, which places emphasis on safety and quality assurance in achieving project milestones linked to time and cost. To support decommissioning of the BRP MCR activities, a reactor vessel (RV) shipping container is required. Discussed in this paper is the design and fabrication of a 10 CFR Part 71 Type B container necessary to ship the BRP RV. The container to be used for transportation of the RV to the burial site was designed as an Exclusive Use Type B package for shipment and burial at the Barnwell, South Carolina (SC) disposal facility. (author)

AN ADA NAMELIST PACKAGE

Science.gov (United States)

Klumpp, A. R.

1994-01-01

The Ada Namelist Package, developed for the Ada programming language, enables a calling program to read and write FORTRAN-style namelist files. A namelist file consists of any number of assignment statements in any order. Features of the Ada Namelist Package are: the handling of any combination of user-defined types; the ability to read vectors, matrices, and slices of vectors and matrices; the handling of mismatches between variables in the namelist file and those in the programmed list of namelist variables; and the ability to avoid searching the entire input file for each variable. The principle user benefits of this software are the following: the ability to write namelist-readable files, the ability to detect most file errors in the initialization phase, a package organization that reduces the number of instantiated units to a few packages rather than to many subprograms, a reduced number of restrictions, and an increased execution speed. The Ada Namelist reads data from an input file into variables declared within a user program. It then writes data from the user program to an output file, printer, or display. The input file contains a sequence of assignment statements in arbitrary order. The output is in namelist-readable form. There is a one-to-one correspondence between namelist I/O statements executed in the user program and variables read or written. Nevertheless, in the input file, mismatches are allowed between assignment statements in the file and the namelist read procedure statements in the user program. The Ada Namelist Package itself is non-generic. However, it has a group of nested generic packages following the nongeneric opening portion. The opening portion declares a variety of useraccessible constants, variables and subprograms. The subprograms are procedures for initializing namelists for reading, reading and writing strings. The subprograms are also functions for analyzing the content of the current dataset and diagnosing errors. Two nested

Application of RELAP/SCDAPSIM with integrated uncertainty options to research reactor systems thermal hydraulic analysis

International Nuclear Information System (INIS)

Allison, C.M.; Hohorst, J.K.; Perez, M.; Reventos, F.

2010-01-01

The RELAP/SCDAPSIM/MOD4.0 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of the international SCDAP Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD4.0, which is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards, uses publicly available RELAP5 and SCDAP models in combination with advanced programming and numerical techniques and other SDTP-member modeling/user options. One such member developed option is an integrated uncertainty analysis package being developed jointly by the Technical University of Catalonia (UPC) and Innovative Systems Software (ISS). This paper briefly summarizes the features of RELAP/SCDAPSIM/MOD4.0 and the integrated uncertainty analysis package, and then presents an example of how the integrated uncertainty package can be setup and used for a simple pipe flow problem. (author)

Man machine interaction for operator information systems : a general purpose display package on PC/AT

International Nuclear Information System (INIS)

Chandra, A.K.; Dubey, B.P.; Deshpande, S.V.; Vaidya, U.W.; Khandekar, A.B.

1991-01-01

Several operator information systems for nuclear plants have been developed at Reactor Control Division of BARC and these have involved extensive operator interaction to extract the maximum information from the systems. Each of these systems used a different scheme for operator interaction. A composite package has now been developed on PC/AT with EGA/VGA for use with any system to obviate the necessity to develop new software for each project. This permits information to be displayed in various formats viz. trend and history curves, tabular data, bar graphs and core matrix (both for 235 and 500 MWe cores). It also allows data to be printed and plotted using multi colour plotter. This package thus integrates all the features of the earlier systems. It also integrates the operator interaction scheme. It uses window based pull down menus to select parameters to be fed into a particular display format. Within any display format the operator has significant flexibility to modify the selected parameters using context dependent soft keys. The package also allows data to be retrieved in machine readable form. This report describes the various user friendly functions implemented and also the design of the system software. (author). 1 tab., 10 fig., 3 refs

On geometric simulating in nuclear reactor calculations by the Monte-Carlo method

International Nuclear Information System (INIS)

Ostashenko, S.V.

1988-01-01

Analysis of existing geometric modules makes it possible to reveal their disadvantages and to formulate requirements list, which should be satisfied by any usefull geometry system. Short description of GDL language used for complex reactor systems simulating is given. GDL language applies hierarchical representation scheme to assemblies, which aids to reduce significantly amount of input data. The language is part of GDL geometry system designed for MCU package and implemented on ES computers

Optimization of advanced gas-cooled reactor fuel performance by a stochastic method

International Nuclear Information System (INIS)

Parks, G.T.

1987-01-01

A brief description is presented of a model representing the in-core behaviour of a single advanced gas-cooled reactor fuel channel, developed specifically for optimization studies. The performances of the only suitable Numerical Algorithms Group (NAG) library package and a Metropolis algorithm routine on this problem are discussed and contrasted. It is concluded that, for the problem in question, the stochastic Metropolis algorithm has distinct advantages over the deterministic NAG routine. (author)

21 CFR 820.130 - Device packaging.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Device packaging. 820.130 Section 820.130 Food and... QUALITY SYSTEM REGULATION Labeling and Packaging Control § 820.130 Device packaging. Each manufacturer shall ensure that device packaging and shipping containers are designed and constructed to protect the...

76 FR 30551 - Specifications for Packagings

Science.gov (United States)

2011-05-26

... design qualification test and each periodic retest on a packaging, a test report must be prepared. The... where the design qualification tests are conducted, for as long as the packaging is produced and for at... report; (5) Manufacturer of the packaging; (6) Description of the packaging design type (e.g. dimensions...

Package Formats for Preserved Digital Material

DEFF Research Database (Denmark)

Zierau, Eld

2012-01-01

This paper presents an investigation of the best suitable package formats for long term digital preservation. The choice of a package format for preservation is crucial for future access, thus a thorough analysis of choice is important. The investigation presented here covers setting up requireme......This paper presents an investigation of the best suitable package formats for long term digital preservation. The choice of a package format for preservation is crucial for future access, thus a thorough analysis of choice is important. The investigation presented here covers setting up...... requirements for package formats used for long term preserved digital material, and using these requirements as the basis for analysing a range of package formats. The result of the concrete investigation is that the WARC format is the package format best suited for the listed requirements. Fulfilling...

Directory of Certificates of Compliance for Radioactive-Materials Packages. Summary report of NRC approved packages

International Nuclear Information System (INIS)

1983-01-01

This directory contains a Summary Report of the US Nuclear Regulatory Commission's Approved Packages (Volume I), all Certificates of Compliance (Volume 2), and Summary Report of NRC Approved Quality Assurance Programs (Volume 3) for Radioactive Material Packages effective December 31, 1982. The purpose of this directory is to make available a convenient source of information on packagings which have been approved by the US Nuclear Regulatory Commission. To assist in identifying packaging, an index by Model Number and corresponding Certificate of Compliance Number is included at the back of Volumes 1 and 2 of the directory. A listing by packaging types is included in the back of Volume 2. An alphabetical listing by company name is included in the back of Volume 3 for approved QA programs. The Summary Reports include a listing of all users of each package design and approved QA programs prior to the publication date of the directory

9 CFR 354.72 - Packaging.

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Packaging. 354.72 Section 354.72... CERTIFICATION VOLUNTARY INSPECTION OF RABBITS AND EDIBLE PRODUCTS THEREOF Supervision of Marking and Packaging § 354.72 Packaging. No container which bears or may bear any official identification or any abbreviation...

Antimicrobial packaging with natural compunds - a review

Directory of Open Access Journals (Sweden)

Renata Dobrucka

2016-12-01

Full Text Available Background:  Packaging problems are an integral part of logistics and the implementation of packaging significantly affects the effectiveness of logistics processes, as a factor which increases the safety and the quality of products being transported. Active packaging is an area of technology needed to meet the requirements of the contemporary consumer. Active packaging creates additional opportunities in systems for packing goods, as well as offering a solution in which the packaging, the product and surroundings interact. Furthermore, active packaging allows packaging to interact with food and the environment and play a dynamic role in food preservation. The main role of antimicrobial packaging is to inhibit the growth of microorganisms that reduce the quality of the packaged product. Methods: The application of natural antimicrobial agents appears to be safe for food products. Also, these compounds have potential applications as a natural preservative in the food packaging industry. This study presents some antibacterial agents, namely chitosan, nisin and pectins. Results and conclusion: Natural substances used in active packaging can eliminate the danger of chemical substances migrating to food.

Overview of the DOE packaging certification process

Energy Technology Data Exchange (ETDEWEB)

Liu, Y.Y.; Carlson, R.D. [Argonne National Lab., IL (United States); Carlson, R.W. [Lawrence Livermore National Lab., CA (United States); Kapoor, A. [USDOE, Washington, DC (United States)

1995-12-31

This paper gives an overview of the DOE packaging certification process, which is implemented by the Office of Facility Safety Analysis, under the Assistance Secretary for Environment, Safety and Health, for packagings that are not used for weapons and weapons components, nor for naval nuclear propulsion. The overview will emphasize Type B packagings and the Safety Analysis Report for Packaging (SARP) review that parallels the NRC packaging review. Other important elements in the DOE packaging certification program, such as training, methods development, data bases, and technical assistance, are also emphasized, because they have contributed significantly to the improvement of the certification process since DOE consolidated its packaging certification function in 1985. The paper finishes with a discussion of the roles and functions of the DOE Packaging Safety Review Steering Committee, which is chartered to address issues and concerns of interest to the DOE packaging and transportation safety community. Two articles related to DOE packaging certification were published earlier on the SARP review procedures and the DOE Packaging Review Guide. These articles may be consulted for additional information.

Education and training at the Rensselaer Polytechnic Institute reactor critical facility

International Nuclear Information System (INIS)

Harris, D.R.

1989-01-01

The Rensselaer Polytechnic Institute (RPI) Reactor Critical Facility (RCF) has provided hands-on education and training for RPI and other students for almost a quarter of a century. The RCF was built in the 1950s by the American Locomotive Company (ALCO) as a critical facility in which to carry out experiments in support of the Army Package power Reactor (APPR) program. A number of APPRs were built and operated. In the middle 1960s, ALCO went out of business and provided the facility to RPI. Since that time, RPI has operated the RCF primarily in a teaching mode in the nuclear engineering department, although limited amounts of reactor research, activation analysis, and reactivity assays have been carried out as well. Recently, a U.S. Department of Energy (DOE) upgrade program supported refueling the RCF with 4.81 wt% enriched UO 2 high-density pellets clad in stainless steel rods. The use of these SPERT (F1) fuel rods in the RCF provided a cost-effective approach to conversion from high-enrichment bombgrade fuel to low-enrichment fuel. More important, however, is the fact that the new fuel is of current interest for light water power reactors with extended lifetime fuel. Thus, not only are critical reactor experiments being carried out on the fuel but, more importantly, the quality of the education and training has been enhanced

Transportation of failed or damaged foreign research reactor spent nuclear fuel

International Nuclear Information System (INIS)

Messick, C.E.; Mustin, T.P.; Massey, C.D.

1998-01-01

Since resuming the Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance Program in 1996, the Program has had to deal with difficult issues associated with the transportation of failed or damaged spent fuel. In several instances, problems with failed or damaged fuel have prevented the acceptance of the fuel at considerable cost to both the Department of Energy (DOE) and research reactor operators. In response to the problems faced by the Acceptance Program, DOE has undertaken significant steps to better define the spent fuel acceptance criteria. DOE has worked closely with the U.S. Nuclear Regulatory Commission to address failed or damaged research reactor spent fuel and to identify cask certificate issues which must be resolved by cask owners and foreign regulatory authorities. The specific issues associated with the transport of Materials Testing Reactor (MTR)-type FRR SNF will be discussed. The information presented will include U.S. Nuclear Regulatory Commission regulatory issues, cask certificate issues, technical constraints, and lessons learned. Specific information will also be provided on the latest efforts to revise DOE's Appendix B, Transport Package (Cask) Acceptance Criteria. The information presented in this paper will be important to foreign research reactor operators, shippers, and cask vendors, so that appropriate amendments to the Certificate of Compliance for spent fuel casks can be submitted in a timely manner to facilitate the safe and scheduled transport of FRR SNF

The development and application of advanced analytical methods to commercial ICF reactor chambers. Final report

Energy Technology Data Exchange (ETDEWEB)

Cousseau, P.; Engelstad, R.; Henderson, D.L. [and others

1997-10-01

Progress is summarized in this report for each of the following tasks: (1) multi-dimensional radiation hydrodynamics computer code development; (2) 2D radiation-hydrodynamic code development; (3) ALARA: analytic and Laplacian adaptive radioactivity analysis -- a complete package for analysis of induced activation; (4) structural dynamics modeling of ICF reactor chambers; and (5) analysis of self-consistent target chamber clearing.

The development and application of advanced analytical methods to commercial ICF reactor chambers. Final report

International Nuclear Information System (INIS)

Cousseau, P.; Engelstad, R.; Henderson, D.L.

1997-10-01

Progress is summarized in this report for each of the following tasks: (1) multi-dimensional radiation hydrodynamics computer code development; (2) 2D radiation-hydrodynamic code development; (3) ALARA: analytic and Laplacian adaptive radioactivity analysis -- a complete package for analysis of induced activation; (4) structural dynamics modeling of ICF reactor chambers; and (5) analysis of self-consistent target chamber clearing

Safety analysis report: packages. LP-50 tritium package (packaging of fissile and other radioactive materials). Final report

International Nuclear Information System (INIS)

Gates, A.A.; McCarthy, P.G.; Edl, J.W.

1975-04-01

Elemental tritium is shipped at low pressure in a stainless steel container (LP-50) sealed within an aluminum vessel and surrounded by a minimum of 4-in. thick Celotex insulation in a steel drum. The structural, thermal, containment, shielding, and criticality safety aspects of this package are evaluated. Procedures for loading and unloading, empty cask transport, acceptance testing and maintenance, and quality assurance requirements for the LP-50 package are described in detail. (U.S.)

Safety analysis report; packages LP-50 tritium package. (Packaging of fissile and other radioactive materials). Final report

International Nuclear Information System (INIS)

Gates, A.A.; McCarthy, P.G.; Edl, J.W.; Chalfant, G.G.

1975-05-01

Elemental tritium is shipped at low pressure in a stainless steel container (LP-50) surrounded by an aluminum vessel and Celotex insulation at least 4 in. thick in a steel drum. The total weight of the package is 260 lbs maximum. The various components that constitute the package are described and are shown in 7 figures. The safety analysis includes: structural evaluations; thermal evaluations; containment; operating procedures; acceptance tests and maintenance program; and design review

Large transport packages for decommissioning waste

International Nuclear Information System (INIS)

Price, M.S.T.

1988-08-01

This document reports progress on a study of large transport packages for decommissioning waste and is the semi-annual report for the period 1 January - 30 June 1988. The main tasks performed during the period related to the assembly of package design criteria ie those aspects of manufacture, handling, storage, transport and disposal which impose constraints on design. This work was synthesised into a design specification for packages which formed the conclusion of that task and was the entry into the final task - the development of package design concepts. The design specifications, which concentrated on the Industrial Package category of the IAEA Transport Regulations, has been interpreted for the two main concepts (a) a self-shielded package disposed of in its entirety and (b) a package with returnable shielding. Preliminary information has been prepared on the cost of providing the package as well as transport to a repository and disposal. There is considerable uncertainty about the cost of disposal and variations of over a factor of 10 are possible. Under these circumstances there is merit in choosing a design concept which is relatively insensitive to disposal cost variations. The initial results indicate that on these grounds the package with returnable shielding is preferred. (author)

Foundational development of an advanced nuclear reactor integrated safety code

International Nuclear Information System (INIS)

Clarno, Kevin; Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth; Hooper, Russell Warren; Humphries, Larry LaRon

2010-01-01

This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

Foundational development of an advanced nuclear reactor integrated safety code.

Energy Technology Data Exchange (ETDEWEB)

Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

2010-02-01

This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

RF and microwave microelectronics packaging II

CERN Document Server

Sturdivant, Rick

2017-01-01

Reviews RF, microwave, and microelectronics assembly process, quality control, and failure analysis Bridges the gap between low cost commercial and hi-res RF/Microwave packaging technologies Engages in an in-depth discussion of challenges in packaging and assembly of advanced high-power amplifiers This book presents the latest developments in packaging for high-frequency electronics. It is a companion volume to “RF and Microwave Microelectronics Packaging” (2010) and covers the latest developments in thermal management, electrical/RF/thermal-mechanical designs and simulations, packaging and processing methods, and other RF and microwave packaging topics. Chapters provide detailed coverage of phased arrays, T/R modules, 3D transitions, high thermal conductivity materials, carbon nanotubes and graphene advanced materials, and chip size packaging for RF MEMS. It appeals to practicing engineers in the electronic packaging and high-frequency electronics domain, and to academic researchers interested in underst...

49 CFR 173.3 - Packaging and exceptions.

Science.gov (United States)

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Packaging and exceptions. 173.3 Section 173.3... SHIPMENTS AND PACKAGINGS General § 173.3 Packaging and exceptions. (a) The packaging of hazardous materials.... standard packaging must be open to inspection by a representative of the Department. (b) The regulations...

Sensitivity analysis of reactor safety parameters with automated adjoint function generation

International Nuclear Information System (INIS)

Kallfelz, J.M.; Horwedel, J.E.; Worley, B.A.

1992-01-01

A project at the Paul Scherrer Institute (PSI) involves the development of simulation models for the transient analysis of the reactors in Switzerland (STARS). This project, funded in part by the Swiss Federal Nuclear Safety Inspectorate, also involves the calculation and evaluation of certain transients for Swiss light water reactors (LWRs). For best-estimate analyses, a key element in quantifying reactor safety margins is uncertainty evaluation to determine the uncertainty in calculated integral values (responses) caused by modeling, calculational methodology, and input data (parameters). The work reported in this paper is a joint PSI/Oak Ridge National Laboratory (ORNL) application to a core transient analysis code of an ORNL software system for automated sensitivity analysis. The Gradient-Enhanced Software System (GRESS) is a software package that can in principle enhance any code so that it can calculate the sensitivity (derivative) to input parameters of any integral value (response) calculated in the original code. The studies reported are the first application of the GRESS capability to core neutronics and safety codes

Materials for advanced packaging

CERN Document Server

Wong, CP

2017-01-01

This second edition continues to be the most comprehensive review on the developments in advanced electronic packaging technologies, with a focus on materials and processing. Recognized experts in the field contribute to 22 updated and new chapters that provide comprehensive coverage on various 3D package architectures, novel bonding and joining techniques, wire bonding, wafer thinning techniques, organic substrates, and novel approaches to make electrical interconnects between integrated circuit and substrates. Various chapters also address advances in several key packaging materials, including: Lead-free solders Flip chip underfills Epoxy molding compounds Conductive adhesives Die attach adhesives/films Thermal interface materials (TIMS) Materials for fabricating embedded passives including capacitors, inductors, and resistors Materials and processing aspects on wafer-level chip scale package (CSP) and MicroElectroMechanical system (MEMS) Contributors also review new and emerging technologies such as Light ...

Improvement and validation of the wall heat transfer package of RELAP5/MOD3.3

International Nuclear Information System (INIS)

Wu, Pan; Xiong, Xiaofei; Shan, Jianqiang; Gou, Junli; Zhang, Bin; Zhang, Bo

2016-01-01

Highlights: • A new heat transfer package has been developed. • It has been incorporated into RELAP5/MOD3.3 to verify its advantages. • The results of modified code were compared with available experimental data. • The results showed that higher prediction accuracy was achieved. - Abstract: The process of energy transfer from heat structure to control volume is determined by the wall-to-fluid heat transfer package, which is crucial for nuclear reactor safety analysis codes. The current logic for selection of heat transfer modes of RELAP5/MOD3.3 code is too complex and may result in incorrect heat transfer mode judgment. Also, the narrow application scope of film boiling heat transfer correlations may result in large errors in film boiling region which is of paramount importance for the predicted peak clad temperatures during hypothetical LB-LOCAs in PWRs. In this study, a new heat transfer package has been developed and incorporated into the RELAP5/MOD3.3 code. Differing from the original package, the modified one consists of twelve heat transfer modes and proposes a new logic for selection of heat transfer modes. For each mode, the models in the existing safety analysis codes and the leading models in literature have been reviewed in order to determine the best model which can easily be applicable to the RELAP5/MOD3.3 code. Particularly (1) a new package of heat transfer correlations are produced; (2) a new logic for selection of film boiling and transition boiling heat transfer modes is proposed which use minimum film boiling temperature and critical heat flux temperature as distinguished points. The modified code has been validated by comparing the analysis results with available experimental data from tube post dryout experiments and loss-of-fluid test (LOFT) facility. The calculation results showed that the improved package could better predict the experimental phenomena with higher prediction accuracy.

Improvement and validation of the wall heat transfer package of RELAP5/MOD3.3

Energy Technology Data Exchange (ETDEWEB)

Wu, Pan; Xiong, Xiaofei; Shan, Jianqiang, E-mail: jqshan@mail.xjtu.edu.cn; Gou, Junli; Zhang, Bin; Zhang, Bo

2016-12-15

Highlights: • A new heat transfer package has been developed. • It has been incorporated into RELAP5/MOD3.3 to verify its advantages. • The results of modified code were compared with available experimental data. • The results showed that higher prediction accuracy was achieved. - Abstract: The process of energy transfer from heat structure to control volume is determined by the wall-to-fluid heat transfer package, which is crucial for nuclear reactor safety analysis codes. The current logic for selection of heat transfer modes of RELAP5/MOD3.3 code is too complex and may result in incorrect heat transfer mode judgment. Also, the narrow application scope of film boiling heat transfer correlations may result in large errors in film boiling region which is of paramount importance for the predicted peak clad temperatures during hypothetical LB-LOCAs in PWRs. In this study, a new heat transfer package has been developed and incorporated into the RELAP5/MOD3.3 code. Differing from the original package, the modified one consists of twelve heat transfer modes and proposes a new logic for selection of heat transfer modes. For each mode, the models in the existing safety analysis codes and the leading models in literature have been reviewed in order to determine the best model which can easily be applicable to the RELAP5/MOD3.3 code. Particularly (1) a new package of heat transfer correlations are produced; (2) a new logic for selection of film boiling and transition boiling heat transfer modes is proposed which use minimum film boiling temperature and critical heat flux temperature as distinguished points. The modified code has been validated by comparing the analysis results with available experimental data from tube post dryout experiments and loss-of-fluid test (LOFT) facility. The calculation results showed that the improved package could better predict the experimental phenomena with higher prediction accuracy.

The decommissioning of a research reactor in the USA: A case study from planning to site release

International Nuclear Information System (INIS)

Boing, L.E.

1997-01-01

Argonne National Laboratory (ANL) has completed the D ampersand D of the Experimental Boiling Water Reactor (EBWR). The project consisted of the decontamination and/or packaging as radioactive waste the reactor vessel and internals, contaminated piping systems, miscellaneous tanks, pumps, and associated equipment. The dismantling process involved the removal and size reduction of equipment and associated plumbing, ductwork, drain lines, etc. Size reduction of reactor vessel internals was performed in the fuel pool. All radioactive and mixed waste was packaged and manifested. A thorough survey of the facility was performed after the removal of contaminated and activated material. Non-radioactive waste was disposed of in the ANL landfill or recycled as appropriate. The EBWR D ampersand D project was divided into four phases. Phases I and II were completed by ANL personnel, while Phases III and IV were done by a contractor under ANL management. The final survey was performed by the contractor, while the verification survey was performed by ANL. The project lasted 118 months. Phase I was initiated in April 1986 and the final report was issued February 1996. The duration of the project was driven by the availability of funding for decommissioning. Total exposure to project personnel was 208.7 person-mSv (20.87 person-rem), with no personnel exceeding the EBWR project dose limit of 15 mSv (1.5 rem)