A radioactive waste transportation package monitoring system for normal transport and accident emergency response conditions

International Nuclear Information System (INIS)

Brown, G.S.; Cashwell, J.W.; Apple, M.L.

1993-01-01

This paper addresses spent fuel and high level waste transportation history and prospects, discusses accident histories of radioactive material transport, discusses emergency responder needs and provides a general description of the Transportation Intelligent Monitoring System (TRANSIMS) design. The key objectives of the monitoring system are twofold: (1) to facilitate effective emergency response to accidents involving a radioactive waste transportation package, while minimizing risk to the public and emergency first-response personnel, and (2) to allow remote monitoring of transportation vehicle and payload conditions to enable research into radioactive material transportation for normal and accident conditions. (J.P.N.)

Safety evaluation for packaging transportation of equipment for tank 241-C-106 waste sluicing system

International Nuclear Information System (INIS)

Calmus, D.B.

1994-01-01

A Waste Sluicing System (WSS) is scheduled for installation in nd waste storage tank 241-C-106 (106-C). The WSS will transfer high rating sludge from single shell tank 106-C to double shell waste tank 241-AY-102 (102-AY). Prior to installation of the WSS, a heel pump and a transfer pump will be removed from tank 106-C and an agitator pump will be removed from tank 102-AY. Special flexible receivers will be used to contain the pumps during removal from the tanks. After equipment removal, the flexible receivers will be placed in separate containers (packagings). The packaging and contents (packages) will be transferred from the Tank Farms to the Central Waste Complex (CWC) for interim storage and then to T Plant for evaluation and processing for final disposition. Two sizes of packagings will be provided for transferring the equipment from the Tank Farms to the interim storage facility. The packagings will be designated as the WSSP-1 and WSSP-2 packagings throughout the remainder of this Safety Evaluation for Packaging (SEP). The WSSP-1 packagings will transport the heel and transfer pumps from 106-C and the WSSP-2 packaging will transport the agitator pump from 102-AY. The WSSP-1 and WSSP-2 packagings are similar except for the length

Aspiration requirements for the transportation of retrievably stored waste in the TRUPACT-2 package

International Nuclear Information System (INIS)

Djordjevic, S.; Drez, P.; Murthy, D.; Temus, C.

1990-01-01

The Transuranic Package Transporter-II (TRUPACT-II) is the shipping package to be used for the transportation of contact-handled transuranic (CH TRU) waste between the various US Department of Energy (DOE) sites, and to the Waste Isolation Pilot Plant (WIPP) located near Carlsbad, New Mexico. Waste (payload) containers to be transported in the TRUPACT-II package are required to be vented prior to being shipped. ''Venting'' refers to the installation of one or more carbon composite filters in the lid of the container, and the puncturing of a rigid liner (if present). This ensures that there is no buildup of pressure or potentially flammable gas concentrations in the container prior to transport. Payload containers in retrievable storage that have been stored in an unvented condition at the DOE sites, may have generated and accumulated potentially flammable concentrations of gases (primarily due to generation of hydrogen by radiolysis) during the unvented storage period. Such payload containers need to be aspirated for a sufficient period of time until safe pre-transport conditions (acceptably low hydrogen concentrations) are achieved. The period of time for which a payload container needs to be in a vented condition before qualifying for transport in a TRUPACT-II package is defined as the ''aspiration time.'' This paper presents the basis for evaluating the minimum aspiration time for a payload container that has been in unvented storage. Three different options available to the DOE sites for meeting the aspiration requirements are described in this paper. 4 refs., 2 figs

A radioactive waste transportation package monitoring system for normal transport and accident emergency response conditions

International Nuclear Information System (INIS)

Brown, G.S.; Cashwell, J.W.; Apple, M.L.

1991-01-01

Shipments of radioactive material (RAM) constitute but a small fraction of the total hazardous materials shipped in the United States each year. Public perception, however, of the potential consequences of a release from a transportation package containing RAM has resulted in significant regulation of transport operations, both to ensure the integrity of a package in accident conditions and to place operational constraints on the shipper. Much of this attention has focused on shipments of spent nuclear fuel and high level wastes which, although comprising a very small number of total shipments, constitute a majority of the total curies transported on an annual basis. This report discusses the shipment of these highly radioactive materials

Radiaoctive waste packaging for transport and final disposal

International Nuclear Information System (INIS)

Suarez, A.A.

1989-01-01

Prior and after the conditioning of radioactive wastes is the packaging design of uppermost importance since it will be the first barrier against water and human intrusion. The choice of the proper package according waste category as well criteria utilized for final disposal are shown. (author) [pt

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

final disposal of radioactive wastes, Brazilian Competent Authority specifies minimum performance requirements for operations that give wastes a suitable form for storage, transport and disposal. Taking into account the difficulties to demonstrate that the packagings were in compliance with such requirements, the packagings were refused. (author)

Transport categories for radioactive waste

International Nuclear Information System (INIS)

Goldfinch, E.P.

1993-01-01

The paper makes proposals for materials which are intrinsically safe without packaging other than for administrative convenience, and for wastes to be transported to the same levels of safety as Type A packages. It is proposed that waste forms to be transported to the same level of safety as Type B packages cannot be prescribed in advance without the need for Competent Authority approval for each specific form or combination of waste form and packaging. Finally it is proposed to revert to simple packaging requirements, equivalent to the earlier industrial and strong industrial packaging. The former have no quantitative performance requirements and the latter have requirements identical to Type A packages. (author)

Waste package/repository impact study: Final report

Energy Technology Data Exchange (ETDEWEB)

1985-09-01

The Waste Package/Repository Impact Study was conducted to evaluate the feasibility of using the current reference salt waste package in the salt repository conceptual design. All elements of the repository that may impact waste package parameters, i.e., (size, weight, heat load) were evaluated. The repository elements considered included waste hoist feasibility, transporter and emplacement machine feasibility, subsurface entry dimensions, feasibility of emplacement configuration, and temperature limits. The evaluations are discussed in detail with supplemental technical data included in Appendices to this report, as appropriate. Results and conclusions of the evaluations are discussed in light of the acceptability of the current reference waste package as the basis for salt conceptual design. Finally, recommendations are made relative to the salt project position on the application of the reference waste package as a basis for future design activities. 31 refs., 11 figs., 11 tabs.

Waste package/repository impact study: Final report

International Nuclear Information System (INIS)

1985-09-01

The Waste Package/Repository Impact Study was conducted to evaluate the feasibility of using the current reference salt waste package in the salt repository conceptual design. All elements of the repository that may impact waste package parameters, i.e., (size, weight, heat load) were evaluated. The repository elements considered included waste hoist feasibility, transporter and emplacement machine feasibility, subsurface entry dimensions, feasibility of emplacement configuration, and temperature limits. The evaluations are discussed in detail with supplemental technical data included in Appendices to this report, as appropriate. Results and conclusions of the evaluations are discussed in light of the acceptability of the current reference waste package as the basis for salt conceptual design. Finally, recommendations are made relative to the salt project position on the application of the reference waste package as a basis for future design activities. 31 refs., 11 figs., 11 tabs

Evaluation on the structural soundness of the transport package for low-level radioactive waste for subsurface disposal against aircraft impact by finite element method

International Nuclear Information System (INIS)

Itoh, Chihiro

2009-01-01

The structural analysis of aircraft crush on the transport package for low-level radioactive waste was performed using the impact force which was already used for the evaluation of the high-level waste transport package by LSDYNA code. The transport package was deformed, and stresses due to the crush exceeded elastic range. However, plastic strains yieled in the package were far than the elongation of the materials and the body of the package did not contact the disposal packages due to the deformation of the package. Therefore, it was confirmed that the package keeps its integrity against aircraft crush. (author)

Use of simple transport equations to estimate waste package performance requirements

International Nuclear Information System (INIS)

Wood, B.J.

1982-01-01

A method of developing waste package performance requirements for specific nuclides is described. The method is based on: Federal regulations concerning permissible concentrations in solution at the point of discharge to the accessible environment; a simple and conservative transport model; baseline and potential worst-case release scenarios. Use of the transport model enables calculation of maximum permissible release rates within a repository in basalt for each of the scenarios. The maximum permissible release rates correspond to performance requirements for the engineered barrier system. The repository was assumed to be constructed in a basalt layer. For the cases considered, including a well drilled into an aquifer 1750 m from the repository center, little significant advantage is obtained from a 1000-yr as opposed to a 100-yr waste package. A 1000-yr waste package is of importance only for nuclides with half-lives much less than 100 yr which travel to the accessible environment in much less than 1000 yr. Such short travel times are extremely unlikely for a mined repository. Among the actinides, the most stringent maximum permissible release rates are for 236 U and 234 U. A simple solubility calculation suggests, however, that these performance requirements can be readily met by the engineered barrier system. Under the reducing conditions likely to occur in a repository located in basalt, uranium would be sufficiently insoluble that no solution could contain more than about 0.01% of the maximum permissible concentration at saturation. The performance requirements derived from the one-dimensional modeling approach are conservative by at least one to two orders of magnitude. More quantitative three-dimensional modeling at specific sites should enable relaxation of the performance criteria derived in this study. 12 references, 8 figures, 8 tables

Konrad transport study: Safety analysis of the transportation of radioactive waste to the Konrad waste disposal site

International Nuclear Information System (INIS)

Lange, F.; Gruendler, D.; Schwarz, G.

1992-05-01

For the purpose of the study the anticipated waste transport volume and the waste properties were analysed in detail. This included information on the transport containers, waste product properties, activity inventories and local dose rates of the waste packages being transported. The envisaged practical implementation, i.e. the transport arrangements including shunting operations at the Braunschweig marshalling yard and the Beddingen interchange station, were also included. The two shipping scenarios 100% transportation by rail and 80% transportation by rail, 20% by road, which could be considered to bound the real conditions, were analysed. The relevant transport regulations contain the requirements to be met by the transport of shipping units carrying radioactive waste. In addition, the ''Konrad preliminary waste acceptance criteria'' contain activity limits for waste packages being disposed of in conjunction with further requirements relating to the properties of waste products and waste containers. (orig./DG)

Transport concept of new waste management system (inner packaging system)

International Nuclear Information System (INIS)

Hakozaki, K.; Wada, R.

2004-01-01

Kobe Steel, Ltd. (KSL) and Transnuclear Tokyo (TNT) have jointly developed a new waste management system concept (called ''Inner packaging system'') for high dose rate wastes generated from nuclear power plants under cooperation with Tokyo Electric Power Company (TEPCO). The inner packaging system is designed as a total management system dedicated to the wastes from nuclear plants in Japan, covering from the wastes conditioning in power plants up to the disposal in final repository. This paper presents the new waste management system concept

Waste Package and Material Testing for the Proposed Yucca Mountain High Level Waste Repository

International Nuclear Information System (INIS)

Doering, Thomas; Pasupathi, V.

2002-01-01

Over the repository lifetime, the waste package containment barriers will perform various functions that will change with time. During the operational period, the barriers will function as vessels for handling, emplacement, and waste retrieval (if necessary). During the years following repository closure, the containment barriers will be relied upon to provide substantially complete containment, through 10,000 years and beyond. Following the substantially complete containment phase, the barriers and the waste package internal structures help minimize release of radionuclides by aqueous- and gaseous-phase transport. These requirements have lead to a defense-in-depth design philosophy. A multi-barrier design will result in a lower breach rate distributed over a longer period of time, thereby ensuring the regulatory requirements are met. The design of the Engineered Barrier System (EBS) has evolved. The initial waste package design was a thin walled package, 3/8 inch of stainless steel 304, that had very limited capacity, (3 PWR and 4 BWR assemblies) and performance characteristics, 300 to 1,000 years. This design required over 35,000 waste packages compared to today's design of just over 10,000 waste packages. The waste package designs are now based on a defense-in-depth/multi-barrier philosophy and have a capacity similar to the standard storage and rail transported spent nuclear fuel casks. Concurrent with the development of the design of the waste packages, a comprehensive waste package materials testing program has been undertaken to support the selection of containment barrier materials and to develop predictive models for the long-term behavior of these materials under expected repository conditions. The testing program includes both long-term and short-term tests and the results from these tests combination with the data published in the open literature are being used to develop models for predicting performance of the waste packages

Shielding design of radioactive contaminated metal waste packaging

International Nuclear Information System (INIS)

Zou Wenhua; Dong Zhiqiang; Yao Zhenyu; Xu Shuhe; Wang Wen

2015-01-01

Focusing on the cylindrical source model to calculate γ dose field of waste packages with the relative formulae then derived. By comparing the calculated data of waste packages of type Ⅷ steel box with the monitoring data, it is found that the cylinder source model could accurately reflect the distributions of γ dose of the waste package. Based on the results of the cylindrical source model, a reasonable shielding technology applicable to waste package containers was designed to meet relevant requirements prescribed in standards about the transport and disposal of radioactive materials. The cylinder source model calculated dose distributions for single package in this paper is simple and easy to implement but slightly larger than the monitoring data providing a certain safety margin for the shielding design. It is suitable for radiological engineering practices. (authors)

Regulatory authority of the Rocky Mountain states for low-level radioactive waste packaging and transportation

International Nuclear Information System (INIS)

Whitman, M.; Tate, P.

1983-07-01

The newly-formed Rocky Mountain Low-Level Radioactive Waste Compact is an interstate agreement for the management of low-level radioactive waste (LLW). Eligible members of the compact are Arizona, Colorado, Nevada, New Mexico, Utah, and Wyoming. Each state must ratify the compact within its legislature for the compact to become effective in that state and to make that state a full-fledged member of the compact. By so adopting the compact, each state agrees to the terms and conditions specified therein. Among those terms and conditions are provisions requiring each member state to adopt and enforce procedures requiring low-level waste shipments originating within its borders and destined for a regional facility to conform to packaging and transportation requirements and regulations. These procedures are to include periodic inspections of packaging and shipping practices, periodic inspections of waste containers while in the custody of carriers and appropriate enforcement actions for violations. To carry out this responsibility, each state must have an adequate statutory and regulatory inspection and enforcement authority to ensure the safe transportation of low-level radioactive waste. Three states in the compact region, Arizona, Utah and Wyoming, have incorporated the Department of Transportation regulations in their entirety, and have no published rules and regulations of their own. The other states in the compact, Colorado, Nevada and New Mexico all have separate rules and regulations that incorporate the DOT regulations. A brief description of the regulatory requirements of each state is presented

Universal storage/transport/disposal packages

International Nuclear Information System (INIS)

Smith, M.L.

1992-01-01

In this paper a concept for a more robust Engineered Barrier System (EBS) that is part of an integrated waste management system is presented. This integrated system uses a thick walled metal package as the basic component of an integrated system for utility site storage, MRS storage, transportation, and disposal. Overpacks are used where necessary to supplement the basic package in each application. This integrated system combines the advantages of a robust EBS (improved margin and confidence in the repository) with a systems approach that can simplify the waste management system and reduce costs

Packages for radiactive waste disposal

International Nuclear Information System (INIS)

Oliveira, R. de.

1983-01-01

The development of multi-stage type package for sea disposal of compactable nuclear wastes, is presented. The basic requirements for the project followed the NEA and IAEA recommendations and observations of the solutions adopted by others countries. The packages of preliminary design was analysed, by computer, under several conditions arising out of its nature, as well as their conditions descent, dumping and durability in the deep of sea. The designed pressure equalization mechanic and the effect compacting on the package, by prototypes and specific tests, were studied. These prototypes were also submitted to the transport tests of the 'Regulament for the Safe Transport of Radioactive Materials'. Based on results of the testes and the re-evaluation of the preliminary design, final indications and specifications for excuting the package design, are presented. (M.C.K.) [pt

An analytical one-dimensional model for predicting waste package performance

International Nuclear Information System (INIS)

Relyea, J.F.; Wood, M.I.

1984-01-01

A method for allocating waste package performance requirements among waste package components with regard to radionuclide isolation has been developed. Modification or change in this approach can be expected as the understanding of radionuclide behavior in the waste package improves. Thus, the performance requirements derived in this document are preliminary and subject to change. However, this kind of analysis is a useful starting point. It has also proved useful for identifying a small group of radionuclides which should be emphasized in a laboratory experimental program designed to characterize the behavior of specific radionuclides in the waste package environment. A simple one-dimensional, two media transport model has been derived and used to calculate radionuclide transport from the waste form-packing material interface of the waste package into the host rock. Cumulative release over 10,000 years, maximum yearly releases and release rates at the packing material-host rock interface were evaluated on a radionuclide-by radionuclide basis. The major parameters controlling radionuclide release were found to be: radionuclide solubility, porosity of the rock, isotopic ratio of the radionuclide and surface area of the waste form-packing material interface. 15 refs., 2 figs., 16 tabs

Waste Package Lifting Calculation

International Nuclear Information System (INIS)

H. Marr

2000-01-01

The objective of this calculation is to evaluate the structural response of the waste package during the horizontal and vertical lifting operations in order to support the waste package lifting feature design. The scope of this calculation includes the evaluation of the 21 PWR UCF (pressurized water reactor uncanistered fuel) waste package, naval waste package, 5 DHLW/DOE SNF (defense high-level waste/Department of Energy spent nuclear fuel)--short waste package, and 44 BWR (boiling water reactor) UCF waste package. Procedure AP-3.12Q, Revision 0, ICN 0, calculations, is used to develop and document this calculation

Transuranic package transporter (TRUPACT) system design status and operational support equipment

International Nuclear Information System (INIS)

Johanson, N.W.; Meyer, R.J.; Romesberg, L.E.; Pope, R.B.

1983-01-01

A program was initiated in the late 1970's at Sandia National Laboratories to develop an efficient, safe, reliable, and cost-effective transportation packaging system for the carriage of contact-handled transuranic (CH-TRU) waste within the Department of Energy (DOE) complex. It is anticipated that eventually a family of TRUPACT (TRansUranic PACKage Transporter) systems having varied dimensions and weight/volume capacities will be needed by the DOE to transport different CH-TRU waste forms. Each TRUPACT system will be a Type B packaging. Large quantities of CH-TRU wastes having many different forms, isotopic contents, and contained in a variety of waste containers have been, are being, and will continue to be produced and stored for ultimate disposal. Packaging design is being closely coordinated with facility designs to ensure the rapid and economic integration of the TRUPACT system. The first packaging developed for transport by truck or rail (bimodal) is designated TRUPACT-I and will become operational in 1984. This paper provides an overview of progress on the TRUPACT-I design and details of equipment to be used for interfacing with users

Hazardous and mixed waste transportation program

International Nuclear Information System (INIS)

Hohnstreiter, G.F.; Glass, R.E.; McAllaster, M.E.; Nigrey, P.J.; Trennel, A.J.; Yoshimura, H.R.

1993-01-01

Sandia National Laboratories (SNL) has developed a program to address the packaging needs associated with the transport of hazardous and mixed waste during the United States' Department of Energy (DOE) remediation efforts. The program addresses the technology needs associated with the transport of materials which have components that are radioactive and chemically hazardous. The mixed waste transportation activities focus on on-site specific applications of technology to the transport of hazardous and mixed wastes. These activities were identified at a series of DOE-sponsored workshops. These activities will be composed of the following: (1) packaging concepts, (2) chemical compatibility studies, and (3) systems studies. This paper will address activities in each of these areas. (J.P.N.)

Hazardous and Mixed Waste Transportation Program

International Nuclear Information System (INIS)

Hohnstreiter, G.F.; Glass, R.E.; McAllaster, M.E.; Nigrey, P.J.; Trennel, A.J.; Yoshimura, H.R.

1991-01-01

Sandia National Laboratories (SNL) has developed a program to address the packaging needs associated with the transport of hazardous and mixed waste during the United States' Department of Energy (DOE) remediation efforts. The program addresses the technology needs associated with the transport of materials which have components that are radioactive and chemically hazardous. The mixed waste transportation activities focus on on-site specific applications of technology to the transport of hazardous and mixed wastes. These activities were identified at a series of DOE-sponsored workshops. These activities will be composed of the following: (1) packaging concepts, (2) chemical compatibility studies, and (3) systems studies. This paper will address activities in each of these areas

Shielding Calculations on Waste Packages – The Limits and Possibilities of different Calculation Methods by the example of homogeneous and inhomogeneous Waste Packages

OpenAIRE

Adams Mike; Smalian Silva

2017-01-01

For nuclear waste packages the expected dose rates and nuclide inventory are beforehand calculated. Depending on the package of the nuclear waste deterministic programs like MicroShield® provide a range of results for each type of packaging. Stochastic programs like “Monte-Carlo N-Particle Transport Code System” (MCNP®) on the other hand provide reliable results for complex geometries. However this type of program requires a fully trained operator and calculations are time consuming. The prob...

Status of ERDA TRU waste packaging study

International Nuclear Information System (INIS)

Doty, J.W. Jr.

1977-01-01

This paper discusses the status of Task 3 of the TRU Waste Cyclone Drum Incinerator and Treatment System program. This task covers acceptable TRU packaging for interim storage and terminal isolation. The kind of TRU wastes generated by contractors and its transport are discussed. Both drum and box systems are desirable

Thermal performance of a depleted uranium shielded storage, transportation, and disposal package

International Nuclear Information System (INIS)

Wix, S.D.; Yoshimura, H.R.

1994-01-01

The US Department of Energy (DOE) is responsible for management and disposal of large quantities of depleted uranium (DU) in the DOE complex. Viable economic options for the use and eventual disposal of the material are needed. One possible option is the use of DU as shielding material for vitrified Defense High-Level Waste (DHLW) storage, transportation, and disposal packages. Use of DU as a shielding material provides the potential benefit of disposing of significant quantities of DU during the DHLW storage and disposal process. Two DU package concepts have been developed by Sandia National Laboratories. The first concept is the Storage/Disposal plus Transportation (S/D+T) package. The S/D+T package consists of two major components: a storage/disposal (S/D) container and a transportation overpack. The second concept is the S/D/T package which is an integral storage, transportation, and disposal package. The package concept considered in this analysis is the S/D+T package with seven DHLW waste canisters

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

Assessment of management alternatives for LWR wastes. Volume 6. Cost determination of the LWR waste management routes (treatment/conditioning/packaging/transport operations)

International Nuclear Information System (INIS)

Thiels, G.M.; Kowa, S.

1993-01-01

This report deals with the cost determination of a number of schemes for the treatment, conditioning, packaging, interim storage and transport operations of LWR wastes drawn up on the basis of Belgian, French and German practices in this particular area. In addition to the general procedure elaborated for determining, actualizing and scaling of plant and transport costs associated with the various schemes, in-depth calculations of each intermediate management stage are included in this report. This study is part of an overall theoretical exercise aimed at evaluating a selection of management routes for LWR waste based on economical and radiological criteria

Transportation considerations related to waste forms and canisters for Defense TRU wastes

International Nuclear Information System (INIS)

Schneider, K.J.; Andrews, W.B.; Schreiber, A.M.; Rosenthal, L.J.; Odle, C.J.

1981-09-01

This report identifies and discusses the considerations imposed by transportation on waste forms and canisters for contact-handled, solid transuranic wastes from the US Department of Energy (DOE) activities. The report reviews (1) the existing raw waste forms and potential immobilized waste forms, (2) the existing and potential future DOE waste canisters and shipping containers, (3) regulations and regulatory trends for transporting commercial transuranic wastes on the ISA, (4) truck and rail carrier requirements and preferences for transporting the wastes, and (5) current and proposed Type B external packagings for transporting wastes

Thermal performance of a depleted uranium shielded storage, transportation, and disposal package

International Nuclear Information System (INIS)

Wix, S.D.; Yoshimura, H.R.

1994-01-01

The US Department of Energy (DOE) is responsible for management and disposal of large quantities of depleted uranium (DU) in the DOE complex. Viable economic options for the use and eventual disposal of the material are needed. One possible option is the use of DU as shielding material for vitrified Defense High-Level Waste (DHLW) storage, transportation, and disposal packages. Use of DU as a shielding material provides the potential benefit of disposing of significant quantities of DU during the DHLW storage and disposal process. Two DU package concepts have been developed by Sandia National Laboratories. The first concept is the Storage/Disposal plus Transportation (S/D+T) package. The S/D+T package consists of two major components: a storage/disposal (S/D) container and a transportation overpack. The second concept is the S/D/T package which is an integral storage, transportation, and disposal package. The package concept considered in this analysis is the S/D+T package with seven DHLW waste canisters. The S/D+T package provides shielding and containment for the DHLW waste canisters. The S/D container is intended to be used as an on-site storage and repository disposal container. In this analysis, the S/D container is constructed from a combination of stainless steel and DU. Other material combinations, such as mild steel and DU, are potential candidates. The transportation overpack is used to transport the S/D containers to a final geological repository and is not included in this analysis

Packaging, Transportation and Recycling of NPP Condenser Modules - 12262

Energy Technology Data Exchange (ETDEWEB)

Polley, G.M. [Perma-Fix Environmental Services, 575 Oak Ridge Turnpike, Oak Ridge, TN 37830 (United States)

2012-07-01

Perma-Fix was awarded contract from Energy Northwest for the packaging, transportation and disposition of the condenser modules, water boxes and miscellaneous metal, combustibles and water generated during the 2011 condenser replacement outage at the Columbia Generating Station. The work scope was to package the water boxes and condenser modules as they were removed from the facility and transfer them to the Perma-Fix Northwest facility for processing, recycle of metals and disposition. The condenser components were oversized and overweight (the condenser modules weighed ∼102,058 kg [225,000 lb]) which required special equipment for loading and transport. Additional debris waste was packaged in inter-modals and IP-1 boxes for transport. A waste management plan was developed to minimize the generation of virtually any waste requiring landfill disposal. The Perma-Fix Northwest facility was modified to accommodate the ∼15 m [50-ft] long condenser modules and equipment was designed and manufactured to complete the disassembly, decontamination and release survey. The condenser modules are currently undergoing processing for free release to a local metal recycler. Over three millions pounds of metal will be recycled and over 95% of the waste generated during this outage will not require land disposal. There were several elements of this project that needed to be addressed during the preparation for this outage and the subsequent packaging, transportation and processing. - Staffing the project to support 24/7 generation of large components and other wastes. - The design and manufacture of the soft-sided shipping containers for the condenser modules that measured ∼15 m X 4 m X 3 m [50 ft X 13 ft X 10 ft] and weighed ∼102,058 kg [225,000 lbs] - Developing a methodology for loading the modules into the shipping containers. - Obtaining a transport vehicle for the modules. - Designing and modifying the processing facility. - Movement of the modules at the processing

Integrated performance assessment model for waste package behavior and radionuclide release

International Nuclear Information System (INIS)

Kossik, R.; Miller, I.; Cunnane, M.

1992-01-01

Golder Associates Inc. (GAI) has developed a probabilistic total system performance assessment and strategy evaluation model (RIP) which can be applied in an iterative manner to evaluate repository site suitability and guide site characterization. This paper describes one component of the RIP software, the waste package behavior and radionuclide release model. The waste package component model considers waste package failure by various modes, matrix alteration/dissolution, and radionuclide mass transfer. Model parameters can be described as functions of local environmental conditions. The waste package component model is coupled to component models for far-field radionuclide transport and disruptive events. The model has recently been applied to the proposed repository at Yucca Mountain

Compatibility of packaging components with simulant mixed waste

International Nuclear Information System (INIS)

Nigrey, P.J.; Dickens, T.G.

1996-01-01

The purpose of hazardous and radioactive materials packaging is to enable these materials to be transported without posing a threat to the health or property of the general public. To achieve this aim, regulations in the US have been written establishing general design requirements for such packagings. While no regulations have been written specifically for mixed waste packaging, regulations for the constituents of mixed wastes, i.e., hazardous and radioactive substances, have been codified by the US Department of Transportation (US DOT, 49 CFR 173) and the US Nuclear Regulatory Commission (NRC, 10 CFR 71). Based on these national requirements, a Chemical Compatibility Testing Program was developed in the Transportation Systems Department at Sandia National Laboratories (SNL). The program provides a basis to assure any regulatory body that the issue of packaging material compatibility towards hazardous and radioactive materials has been addressed. In this paper, the authors present the results of the second phase of this testing program. The first phase screened five liner materials and six seal materials towards four simulant mixed wastes. This phase involved the comprehensive testing of five candidate liner materials to an aqueous Hanford Tank simulant mixed waste. The comprehensive testing protocol involved exposing the respective materials a matrix of four gamma radiation doses (∼ 1, 3, 6, and 40 kGy), three temperatures (18, 50, and 60 C), and four exposure times (7, 14, 28, and 180 days). Following their exposure to these combinations of conditions, the materials were evaluated by measuring five material properties. These properties were specific gravity, dimensional changes, hardness, stress cracking, and mechanical properties

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

Performance of the IAEA transport regulations in controlling doses and risks from a large-scale radioactive waste transport system

International Nuclear Information System (INIS)

Hutchinson, D.; Miles, R.; White, I.

2004-01-01

The role of United Kingdom Nirex Limited is to provide the UK with safe, environmentally sound and publicly acceptable options for the long-term management of radioactive materials generated by the UK's commercial, medical, research and defence activities. An important part of this role is to set standards and specifications for waste packaging. Waste producers in the UK are currently developing processes for packaging many different types of intermediatelevel waste (ILW), and also those forms of low-level waste that will require similar management to ILW. When packaging processes are at the proposal stage, the waste producers consult Nirex about the suitability of the resulting packages for all future aspects of waste management. The response that Nirex provides is based on detailed assessments of the proposed packages, their compliance with Nirex standards and specifications, and their predicted performance through the successive phases of waste management. One of those phases is transport through the public domain. This paper draws on experience gained from more than 200 separate transport safety assessments, which have cumulatively covered a wide range of waste types, waste packages and transport packages

Contribution to internal pressure and flammable gas concentration in RAM [radioactive material] transport packages

International Nuclear Information System (INIS)

Warrant, M.M.; Brown, N.

1989-01-01

Various facilities in the US generate wastes contaminated with transuranic (TRU) isotopes (such as plutonium and americium) that decay primarily by emission of alpha particles. The waste materials consist of a wide variety of commercially available plastics, paper, cloth, and rubber; concreted or sludge wastes containing water; and metals, glass, and other solid inorganic materials. TRU wastes that have surface dose rates of 200 mrem/hr or less are typically packaged in plastic bags placed inside metal drums or boxes that are vented through high efficiency particulate air (HEPA) filters. These wastes are to be transported from waste generation or storage sites to the Waste Isolation Pilot Plant (WIPP) in the TRUPACT-II, a Type B package. Radiolysis of organic wastes or packaging materials, or wastes containing water generates gas which may be flammable or simply contribute to the internal pressure of the radioactive material (RAM) transport package. This paper discusses the factors that affect the amount and composition of this gas, and summarizes maximum radiolytic G values (number of molecules produced per 100 eV absorbed energy) found in the technical literature for many common materials. These G values can be used to determine the combination of payload materials and decay heats that are safe for transport. G values are established for categories of materials, based on chemical functional groups. It is also shown using transient diffusion and quasi-equilibrium statistical mechanics methods that hydrogen, if generated, will not stratify at the top of the transport package void space. 9 refs., 1 tab

Interim storage of radioactive waste packages

International Nuclear Information System (INIS)

1998-01-01

This report covers all the principal aspects of production and interim storage of radioactive waste packages. The latest design solutions of waste storage facilities and the operational experiences of developed countries are described and evaluated in order to assist developing Member States in decision making and design and construction of their own storage facilities. This report is applicable to any category of radioactive waste package prepared for interim storage, including conditioned spent fuel, high level waste and sealed radiation sources. This report addresses the following issues: safety principles and requirements for storage of waste packages; treatment and conditioning methods for the main categories of radioactive waste; examples of existing interim storage facilities for LILW, spent fuel and high level waste; operational experience of Member States in waste storage operations including control of storage conditions, surveillance of waste packages and observation of the behaviour of waste packages during storage; retrieval of waste packages from storage facilities; technical and administrative measures that will ensure optimal performance of waste packages subject to various periods of interim storage

General Corrosion and Localized Corrosion of Waste Package Outer Barrier

Energy Technology Data Exchange (ETDEWEB)

K.G. Mon

2004-10-01

The waste package design for the License Application is a double-wall waste package underneath a protective drip shield (BSC 2004 [DIRS 168489]; BSC 2004 [DIRS 169480]). The purpose and scope of this model report is to document models for general and localized corrosion of the waste package outer barrier (WPOB) to be used in evaluating waste package performance. The WPOB is constructed of Alloy 22 (UNS N06022), a highly corrosion-resistant nickel-based alloy. The inner vessel of the waste package is constructed of Stainless Steel Type 316 (UNS S31600). Before it fails, the Alloy 22 WPOB protects the Stainless Steel Type 316 inner vessel from exposure to the external environment and any significant degradation. The Stainless Steel Type 316 inner vessel provides structural stability to the thinner Alloy 22 WPOB. Although the waste package inner vessel would also provide some performance for waste containment and potentially decrease the rate of radionuclide transport after WPOB breach before it fails, the potential performance of the inner vessel is far less than that of the more corrosion-resistant Alloy 22 WPOB. For this reason, the corrosion performance of the waste package inner vessel is conservatively ignored in this report and the total system performance assessment for the license application (TSPA-LA). Treatment of seismic and igneous events and their consequences on waste package outer barrier performance are not specifically discussed in this report, although the general and localized corrosion models developed in this report are suitable for use in these scenarios. The localized corrosion processes considered in this report are pitting corrosion and crevice corrosion. Stress corrosion cracking is discussed in ''Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material'' (BSC 2004 [DIRS 169985]).

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

Integrated performance assessment model for waste policy package behavior and radionuclide release

International Nuclear Information System (INIS)

Kossik, R.; Miller, I.; Cunnane, M.

1992-01-01

Golder Associates Inc. (GAI) has developed a probabilistic total system performance assessment and strategy evaluation model (RIP) which can be applied in an iterative manner to evaluate repository site suitability and guide site characterization. This paper describes one component of the RIP software, the waste package behavior and radionuclide release model. The waste package component model considers waste package failure by various modes, matrix alteration/dissolution, and radionuclide mass transfer. Model parameters can be described as functions of local environmental conditions. The waste package component model is coupled to component models for far-field radionuclide transport and disruptive events. The model has recently been applied to the proposed repository at Yucca Mountain

Demonstration tests for low level radioactive waste packaging safety

International Nuclear Information System (INIS)

Nagano, I.; Shimura, S.; Miki, T.; Tamamura, T.; Kunitomi, K.

1993-01-01

The transport packaging for low level radioactive waste (so-called the LLW packaging) has been developed to be utilized for transportation of LLW in 200 liter-drums from Japanese nuclear power stations to the LLW Disposal Center at Rokkashomura in Aomori Prefecture. Transportation is expected to start from December in 1992. We will explain the brief history of the development, technical features and specifications as well as two kinds of safety demonstration tests, namely one is '1.2 meter free drop test' and the other is 'ISO container standard test'. (J.P.N.)

Structural safety test and analysis of type IP-2 transport packages with bolted lid type and thick steel plate for radioactive waste drums in a NPP

International Nuclear Information System (INIS)

Kim, Dong Hak; Seo, Ki Seog; Lee, Sang Jin; Lee, Kyung Ho; Kim, Jeong Mook

2007-01-01

If a type IP-2 transport package were to be subjected to a free drop test and a penetration test under the normal conditions of transport, it should prevent a loss or dispersal of the radioactive contents and a more than 20% increase in the maximum radiation level at any external surface of the package. In this paper, we suggested the analytic method to evaluate the structural safety of a type IP-2 transport package using a thick steel plate for a structure part and a bolt for tying a bolt. Using an analysis a loss or disposal of the radioactive contents and a loss of shielding integrity were confirmed for two kinds of type IP-2 transport packages to transport radioactive waste drums from a waste facility to a temporary storage site in a nuclear power plant. Under the free drop condition the maximum average stress at the bolts and the maximum opening displacement of a lid were compared with the tensile stress of a bolt and the steps in a lid, which were made to avoid a streaming radiation in the shielding path, to evaluate a loss or dispersal of radioactive waste contents. Also a loss of shielding integrity was evaluated using the maximum decrease in a shielding thickness. To verify the impact dynamic analysis for free drop test condition and evaluate experimentally the safety of two kinds of type IP-2 transport packages, free drop tests were conducted with various drop directions

Research and Development Program for transportation packagings at Sandia National Laboratories

International Nuclear Information System (INIS)

Hohnstreiter, G.F.; Sorenson, K.B.

1995-01-01

This document contains information about the research and development programs dealing with waste transport at Sandia National Laboratories. This paper discusses topics such as: Why new packaging is needed; analytical methodologies and design codes;evaluation of packaging components; materials characterization; creative packaging concepts; packaging engineering and analysis; testing; and certification support

Testing of the permissible inventories in radioactive waste packages

International Nuclear Information System (INIS)

Stegmaier, W.

1988-01-01

The inventories of radionuclides in waste packages which are to be stored in repositories are determined in the Waste Acceptance Requirements of the repository and in the Act on Transport of Dangerous Goods. In this report limiting values of relevant radionuclides are given in such a way that it is possible to use them in a standardized manner. The limiting values apply to single radionuclides, for handling mixtures of nuclides it is necessary to use the sum formula. The minimized number of waste packages which must be produced from a given quantity of raw waste and an inventory of radionuclides keeping all parameters can be calculated with the help of the shown calculating sheet. (orig.) [de

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

Mass Transfer Model for a Breached Waste Package

International Nuclear Information System (INIS)

Hsu, C.; McClure, J.

2004-01-01

The degradation of waste packages, which are used for the disposal of spent nuclear fuel in the repository, can result in configurations that may increase the probability of criticality. A mass transfer model is developed for a breached waste package to account for the entrainment of insoluble particles. In combination with radionuclide decay, soluble advection, and colloidal transport, a complete mass balance of nuclides in the waste package becomes available. The entrainment equations are derived from dimensionless parameters such as drag coefficient and Reynolds number and based on the assumption that insoluble particles are subjected to buoyant force, gravitational force, and drag force only. Particle size distributions are utilized to calculate entrainment concentration along with geochemistry model abstraction to calculate soluble concentration, and colloid model abstraction to calculate colloid concentration and radionuclide sorption. Results are compared with base case geochemistry model, which only considers soluble advection loss

Transportation Packages to Support Savannah River Site Missions

International Nuclear Information System (INIS)

Opperman, E.

2001-01-01

The Savannah River Site's missions have expanded from primarily a defense mission to one that includes environmental cleanup and the stabilization, storage, and preparation for final disposition of nuclear materials. The development of packaging and the transportation of radioactive materials are playing an ever-increasing role in the successful completion of the site's missions. This paper describes the Savannah River Site and the three strategic mission areas of (1) nuclear materials stewardship, (2) environmental stewardship, and (3) nuclear weapons stockpile stewardship. The materials and components that need to be shipped, and associated packaging, will be described for each of the mission areas. The diverse range of materials requiring shipment include spent fuel, irradiated target assemblies, excess plutonium and uranium materials, high level waste canisters, transuranic wastes, mixed and low level wastes, and nuclear weapons stockpile materials and components. Since many of these materials have been in prolonged storage or resulted from disassembly of components, the composition, size and shape of the materials present packaging and certification challenges that need to be met. Over 30 different package designs are required to support the site's missions. Approximately 15 inbound shipping-legs transport materials into the Savannah River Site and the same number (15) of outgoing shipment-legs are carrying materials from the site for further processing or permanent disposal

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

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.

Waste package characterisation

Energy Technology Data Exchange (ETDEWEB)

Sannen, L.; Bruggeman, M.; Wannijn, J.P

1998-09-01

Radioactive wastes originating from the hot labs of the Belgian Nuclear Research Centre SCK-CEN contain a wide variety of radiotoxic substances. The accurate characterisation of the short- and long-term radiotoxic components is extremely difficult but required in view of geological disposal. This paper describes the methodology which was developed and adopted to characterise the high- and medium-level waste packages at the SCK-CEN hot laboratories. The proposed method is based on the estimation of the fuel inventory evacuated in a particular waste package; a calculation of the relative fission product contribution on the fuel fabrication and irradiation footing; a comparison of the calculated, as expected, dose rate and the real measured dose rate of the waste package. To cope with the daily practice an appropriate fuel inventory estimation route, a user friendly computer programme for fission product and corresponding dose rate calculation, and a simple dose rate measurement method have been developed and implemented.

Waste package characterisation

International Nuclear Information System (INIS)

Sannen, L.; Bruggeman, M.; Wannijn, J.P.

1998-09-01

Radioactive wastes originating from the hot labs of the Belgian Nuclear Research Centre SCK-CEN contain a wide variety of radiotoxic substances. The accurate characterisation of the short- and long-term radiotoxic components is extremely difficult but required in view of geological disposal. This paper describes the methodology which was developed and adopted to characterise the high- and medium-level waste packages at the SCK-CEN hot laboratories. The proposed method is based on the estimation of the fuel inventory evacuated in a particular waste package; a calculation of the relative fission product contribution on the fuel fabrication and irradiation footing; a comparison of the calculated, as expected, dose rate and the real measured dose rate of the waste package. To cope with the daily practice an appropriate fuel inventory estimation route, a user friendly computer programme for fission product and corresponding dose rate calculation, and a simple dose rate measurement method have been developed and implemented

Mixed waste chemical compatibility: A testing program for plastic packaging components

International Nuclear Information System (INIS)

Nigrey, P.J.

1995-01-01

The purpose of hazardous and radioactive materials packaging is to enable these materials to be transported without posing a threat to the health or property of the general public. To achieve this aim, regulations in the United States have been written establishing general design requirements for such packagings. While no regulations have been written specifically for mixed waste packaging, regulations for the constituents of mixed wastes, i.e., hazardous and radioactive substances, have been codified by the US Department of Transportation (DOT, 49 CFR 173) and the US Nuclear Regulatory Commission (NRC, 10 CFR 71). The design requirements for both hazardous [49 CFR 173.24 (e)(1)] and radioactive [49 CFR 173.412 (g)] materials packaging specify packaging compatibility, i.e., that the materials of the packaging at sign d any contents be chemically compatible with each other. Furthermore, Type A [49 CFR 173.412 (g)] and Type B (10 CFR 71.43) packaging design requirements stipulate that there be no significant chemical, galvanic, or other reaction between the materials and contents of the package. Based on these requirements, a Chemical Compatibility Testing Program was developed in the Transportation Systems Department at Sandia National Laboratories (SNL). The program attempts to assure any regulatory body that the issue of packaging material compatibility towards hazardous and radioactive materials has been addressed. This program has been described in considerable detail in an internal SNL document, the Chemical Compatibility Test Plan ampersand Procedure Report (Nigrey 1993)

Transportation system (TRUPACT) for contact-handled transuranic wastes

International Nuclear Information System (INIS)

Romesberg, L.E.; Pope, R.B.; Burgoyne, R.M.

1982-04-01

Contact-handled transuranic defense waste is being, and will continue to be, moved between a number of locations in the United States. The DOE is sponsoring development of safe, efficient, licensable, and cost-effective transportation systems to handle this waste. The systems being developed have been named TRUPACT which stands for TRansUranic PACkage Transporter. The system will be compatible with Type A packagings used by waste generators, interim storage facilities, and repositories. TRUPACT is required to be a Type B packaging since larger than Type A quantities of some radionuclides (particularly plutonium) may be involved in the collection of Type A packagings. TRUPACT must provide structural and thermal protection to the waste in hypothetical accident environments specified in DOT regulations 49CFR173 and NRC regulations 10CFR71. Preliminary design of the systems has been completed and final design for a truck system is underway. The status of the development program is reviewed in this paper and the reference design is described. Tests that have been conducted are discussed and long-term program objectives are reviewed

Development of a pneumatic stowing and chocking system for packages containing radioactive waste

International Nuclear Information System (INIS)

Baekelandt, L.; Libon, H.; Vandorpe, M.; Lafontaine, I.

1989-01-01

Since that goods are transported, their chocking and stowing is very often done by improvisation, successfully or disastrously. When the disaster appears in comics it is always a source of an enormous amusement, when it appears in road or maritime accidents it is most of the time a source of death or severe damages. Even if transport of radioactive materials could be considered as the exception where chains and tie-down systems are used abundantly, their strength relies always on the weakness of their components. Special attention has been paid to the transport of type A or type B packages, but obviously there was a lack of interest for the transport of low level radioactive waste, even knowing that the quantities of this waste are a hunderfold or a thousandfold of the first ones. On the subject of stowing and chocking systems for radioactive waste packages, TRANSNUBEL together with the CEA-France performed under the sponsorship of the Commission of the European Communities between 1980 and 1985 a study which clearly showed that during a road accident, in case of a front end impact, the stowing system must be able to absorb entirely the kinetic energy generated by the package deceleration, which is proportional to the package mass. The chocks must be able to absorb a deceleration energy generated by the package of about 30 g at a speed of about 50 km/h. This energy of course decreases at the same time as the speed. These conclusions served as basic principles for the development by TRANSNUBEL of a pneumatic stowing and chocking system for packagings containing radioactive waste

Radioactive wastes transport. A safety logic

International Nuclear Information System (INIS)

2005-01-01

The safety principle which applies to transport operations of radioactive wastes obeys to a very strict regulation. For the conditioning of wastes in package, the organisation of shipments and the qualification of carriers, the ANDRA, the French national agency of radioactive wastes, has implemented a rigorous policy based on the respect of a quality procedure and on the mastery of delivery fluxes. This brochure presents in a simple, illustrated and detailed manner the different steps of these transports. (J.S.)

Specific transport and storage solutions: Waste management facing current and future stakes of the nuclear fuel cycle

International Nuclear Information System (INIS)

Deniau, Helene; Gagner, Laurent; Gendreau, Francoise; Presta, Anne

2006-01-01

With major projects ongoing or being planned, and also with the daily management of radioactive waste from nuclear facilities, the role of transport and/or storage packaging has been often overlooked. Indeed, the packaging development process and transport solutions implemented are a key part of the waste management challenge: protection of people and environment. During over four decades, the AREVA Group has developed a complete and coherent system for the transport of waste produced by nuclear industries. The transport solutions integrate the factors to consider, as industrial transportation needs, various waste forms, associated hazards and current regulations. Thus, COGEMA LOGISTICS has designed, licensed and manufactured a large number of different transport, storage and dual purpose cask models for residues and all kinds of radioactive wastes. The present paper proposes to illustrate how a company acting both as a cask designer and a carrier is key to the waste management issue and how it can support the waste management policy of nuclear producers through their operational choices. We will focus on the COGEMA LOGISTICS technical solutions implemented to guarantee safe and secure transportation and storage solutions. We will describe different aspects of the cask design process, insisting on how it enables to fulfill both customer needs and regulation requirements. We will also mention the associated services developed by the AREVA Business Unit Logistics (COGEMA LOGISTICS, TRANSNUCLEAR, MAINCO, and LEMARECHAL CELESTIN) in order to manage transportation of liquid and solid waste towards interim or final storage sites. The paper has the following contents: About radioactive waste; - Radioactive waste classification; - High level activity waste and long-lived intermediate level waste; - Long-lived low level waste; - Short-lived low- and intermediate level waste; - Very low level waste; - The radioactive waste in nuclear fuel cycle; - Packaging design and

Waste Package Component Design Methodology Report

International Nuclear Information System (INIS)

D.C. Mecham

2004-01-01

This Executive Summary provides an overview of the methodology being used by the Yucca Mountain Project (YMP) to design waste packages and ancillary components. This summary information is intended for readers with general interest, but also provides technical readers a general framework surrounding a variety of technical details provided in the main body of the report. The purpose of this report is to document and ensure appropriate design methods are used in the design of waste packages and ancillary components (the drip shields and emplacement pallets). The methodology includes identification of necessary design inputs, justification of design assumptions, and use of appropriate analysis methods, and computational tools. This design work is subject to ''Quality Assurance Requirements and Description''. The document is primarily intended for internal use and technical guidance for a variety of design activities. It is recognized that a wide audience including project management, the U.S. Department of Energy (DOE), the U.S. Nuclear Regulatory Commission, and others are interested to various levels of detail in the design methods and therefore covers a wide range of topics at varying levels of detail. Due to the preliminary nature of the design, readers can expect to encounter varied levels of detail in the body of the report. It is expected that technical information used as input to design documents will be verified and taken from the latest versions of reference sources given herein. This revision of the methodology report has evolved with changes in the waste package, drip shield, and emplacement pallet designs over many years and may be further revised as the design is finalized. Different components and analyses are at different stages of development. Some parts of the report are detailed, while other less detailed parts are likely to undergo further refinement. The design methodology is intended to provide designs that satisfy the safety and operational

Waste Package Component Design Methodology Report

Energy Technology Data Exchange (ETDEWEB)

D.C. Mecham

2004-07-12

This Executive Summary provides an overview of the methodology being used by the Yucca Mountain Project (YMP) to design waste packages and ancillary components. This summary information is intended for readers with general interest, but also provides technical readers a general framework surrounding a variety of technical details provided in the main body of the report. The purpose of this report is to document and ensure appropriate design methods are used in the design of waste packages and ancillary components (the drip shields and emplacement pallets). The methodology includes identification of necessary design inputs, justification of design assumptions, and use of appropriate analysis methods, and computational tools. This design work is subject to ''Quality Assurance Requirements and Description''. The document is primarily intended for internal use and technical guidance for a variety of design activities. It is recognized that a wide audience including project management, the U.S. Department of Energy (DOE), the U.S. Nuclear Regulatory Commission, and others are interested to various levels of detail in the design methods and therefore covers a wide range of topics at varying levels of detail. Due to the preliminary nature of the design, readers can expect to encounter varied levels of detail in the body of the report. It is expected that technical information used as input to design documents will be verified and taken from the latest versions of reference sources given herein. This revision of the methodology report has evolved with changes in the waste package, drip shield, and emplacement pallet designs over many years and may be further revised as the design is finalized. Different components and analyses are at different stages of development. Some parts of the report are detailed, while other less detailed parts are likely to undergo further refinement. The design methodology is intended to provide designs that satisfy the safety

Safety evaluation for packaging (onsite) for concrete-shielded RHTRU waste drum for the 327 postirradiation testing laboratory

Energy Technology Data Exchange (ETDEWEB)

Adkins, H.E.

1996-10-29

This safety evaluation for packaging authorizes onsite transport of Type B quantities of radioactive material in the Concrete- Shielded Remote-Handled Transuranic Waste (RH TRU) Drum per WHC-CM-2-14, Hazardous Material Packaging and Shipping. The drum will be used for transport of 327 Building legacy waste from the 300 Area to the Transuranic Waste Storage and Assay Facility in the 200 West Area and on to a Solid Waste Storage Facility, also in the 200 Area.

The development of a type B(U) transport container design in cast and forged stainless steel for the transport of immobilised intermediate level waste

International Nuclear Information System (INIS)

Sievwright, B.; Dixon, P.; Tso, C.F.

2004-01-01

United Kingdom Nirex Limited (Nirex) is responsible for providing the United Kingdom with safe, environmentally sound and publicly acceptable options for the long-term management of radioactive materials. This includes intermediate level (ILW) and some low level (LLW) wastes. As part of its role Nirex has defined standards and specifications for the conditioning and packaging of these wastes, and carries out assessments of packaging proposals to ensure compatibility with the requirements for future phases of waste management. In order to facilitate this process and to provide a basis for the production of waste package specifications, Nirex has developed the Phased Disposal Concept, and produced a suite of underpinning safety and performance assessments. It has also undertaken work to assess the compatibility of its waste packaging specifications with other waste management options. The Phased Disposal Concept continues to be developed and updated to incorporate issues arising from dialogue with stakeholders, including members of the public; future changes arising from Government policy, legislation and regulations; information from waste producers, and the results from on-going research and development. One of the documents describing the Phased Disposal Concept is the Generic Transport System Design (GTSD). The GTSD outlines the range of waste packages to be transported and disposed of, and describes the design of the transport system needed to transport wastes from their sites of production or storage to a centralised phased disposal facility site. It also describes a range of re-usable transport containers which could be used to transport those waste packages, which require Type B standards for transport, through the public domain. This paper describes the development to date of such a design of reusable transport container, known as the SWTC-285, the Standard Waste Transport Container (SWTC) with 285 mm of shielding

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

Gas generation phenomena in radioactive waste transportation packaging

International Nuclear Information System (INIS)

Nigrey, P.J.

1998-01-01

The interaction of radiation from radioactive materials with the waste matrix can lead to the deterioration of the waste form resulting in the possible of gaseous species. Depending on the type and characteristics of the radiation source, the generation of hydrogen may predominate. Since the interaction of alpha particles with the waste form results in significant energy transfer, other gases such as carbon oxides, methane, nitrogen oxides, oxygen, water, and helium are possible. The type of gases produced from the waste forms is determined by the mechanisms involved in the waste degradation. For transuranic wastes, the identified degradation mechanisms are reported to be caused by radiolysis, thermal decomposition or dewatering, chemical corrosion, and bacterial action. While all these mechanisms may be responsible for the building of gases during the storage of wastes, radiolysis and thermal decomposition appear to be main contributors during waste transport operations. (authors)

Package materials, waste form

International Nuclear Information System (INIS)

Anon.

1980-01-01

The schedules for waste package development for the various host rocks were presented. The waste form subtask activities were reviewed, with the papers focusing on high-level waste, transuranic waste, and spent fuel. The following ten papers were presented: (1) Waste Package Development Approach; (2) Borosilicate Glass as a Matrix for Savannah River Plant Waste; (3) Development of Alternative High-Level Waste Forms; (4) Overview of the Transuranic Waste Management Program; (5) Assessment of the Impacts of Spent Fuel Disassembly - Alternatives on the Nuclear Waste Isolation System; (6) Reactions of Spent Fuel and Reprocessing Waste Forms with Water in the Presence of Basalt; (7) Spent Fuel Stabilizer Screening Studies; (8) Chemical Interactions of Shale Rock, Prototype Waste Forms, and Prototype Canister Metals in a Simulated Wet Repository Environment; (9) Impact of Fission Gas and Volatiles on Spent Fuel During Geologic Disposal; and (10) Spent Fuel Assembly Decay Heat Measurement and Analysis

Isotope production potential at Sandia National Laboratories: Product, waste, packaging, and transportation

International Nuclear Information System (INIS)

Trennel, A.J.

1995-01-01

The U.S. Congress directed the U.S. Department of Energy to establish a domestic source of molybdenum-99, an essential isotope used in nuclear medicine and radiopharmacology. An Environmental Impact Statement for production of 99 Mo at one of four candidate sites is being prepared. As one of the candidate sites, Sandia National Laboratories is developing the Isotope Production Project. Using federally approved processes and procedures now owned by the U.S. Department of Energy, and existing facilities that would be modified to meet the production requirements, the Sandia National Laboratories' Isotope Project would manufacture up to 30 percent of the U.S. market, with the capacity to meet 100 percent of the domestic need if necessary. This paper provides a brief overview of the facility, equipment, and processes required to produce isotopes. Packaging and transportation issues affecting both product and waste are addressed, and the storage and disposal of the four low-level radioactive waste types generated by the production program are considered. Recommendations for future development are provided

Review of DOE waste package program. Subtask 1.1. National waste package program, April-September 1982

International Nuclear Information System (INIS)

Soo, P.

1983-03-01

The current effort is part of an ongoing task to evaluate the national high-level waste package effort. It includes evaluations of reference waste form, container, and packing material components with respect to determining how they may contribute to the containment and controlled release of radionuclides after waste packages have been emplaced in salt and basalt repositories. Chemical and mechanical failure/degradation modes for the waste package have been reviewed and the licensing data requirements to demonstrate compliance with NRC performance objectives specified

Waste package reliability analysis

International Nuclear Information System (INIS)

Pescatore, C.; Sastre, C.

1983-01-01

Proof of future performance of a complex system such as a high-level nuclear waste package over a period of hundreds to thousands of years cannot be had in the ordinary sense of the word. The general method of probabilistic reliability analysis could provide an acceptable framework to identify, organize, and convey the information necessary to satisfy the criterion of reasonable assurance of waste package performance according to the regulatory requirements set forth in 10 CFR 60. General principles which may be used to evaluate the qualitative and quantitative reliability of a waste package design are indicated and illustrated with a sample calculation of a repository concept in basalt. 8 references, 1 table

Defense waste transportation: cost and logistics studies

International Nuclear Information System (INIS)

Andrews, W.B.; Cole, B.M.; Engel, R.L.; Oylear, J.M.

1982-08-01

Transportation of nuclear wastes from defense programs is expected to significantly increase in the 1980s and 1990s as permanent waste disposal facilities come into operation. This report uses models of the defense waste transportation system to quantify potential transportation requirements for treated and untreated contact-handled transuranic (CH-TRU) wastes and high-level defense wastes (HLDW). Alternative waste management strategies in repository siting, waste retrieval and treatment, treatment facility siting, waste packaging and transportation system configurations were examined to determine their effect on transportation cost and hardware requirements. All cost estimates used 1980 costs. No adjustments were made for future changes in these costs relative to inflation. All costs are reported in 1980 dollars. If a single repository is used for defense wastes, transportation costs for CH-TRU waste currently in surface storage and similar wastes expected to be generated by the year 2000 were estimated to be 109 million dollars. Recovery and transport of the larger buried volumes of CH-TRU waste will increase CH-TRU waste transportation costs by a factor of 70. Emphasis of truck transportation and siting of multiple repositories would reduce CH-TRU transportation costs. Transportation of HLDW to repositories for 25 years beginning in 1997 is estimated to cost $229 M in 1980 costs and dollars. HLDW transportation costs could either increase or decrease with the selection of a final canister configuration. HLDW transportation costs are reduced when multiple repositories exist and emphasis is placed on truck transport

Transportation of liquid mixed waste in the US: Is it really a problem?

International Nuclear Information System (INIS)

Chakraborti, S.; DeBiase, T.

1993-01-01

The transportation of liquid radioactive wastes has often been perceived to be a problem because of the potential consequences from hypothetical accident scenarios and the difficulties that may be encountered in the handling and containment of liquids. This paper focuses specifically to determine if the transportation of these wastes are severely restricted by the regulations. The paper also compares current practices for the transportation of liquid mixed waste in the US with that of France to provide an international perspective on the issue. The review of the regulations and current practices shows that the transportation of liquid mixed waste is by no means prohibited, and also that the majority of the regulations do not impose any additional restrictions because of the physical form of the waste. Rather, the selection of an authorized package primarily depends on the quantity of radioactivity and the specific radionuclides involved. Although the selection process for an authorized package for liquid mixed wastes is fairly straightforward, it seems that the difficulties in transporting liquid mixed waste can be attributed to the lack of readily available Type A packages designed for transporting liquids

Large packages for reactor decommissioning waste

International Nuclear Information System (INIS)

Price, M.S.T.

1991-01-01

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

Estimation of waste package performance requirements for a nuclear waste repository in basalt

International Nuclear Information System (INIS)

Wood, B.J.

1980-07-01

A method of developing waste package performance requirements for specific nuclides is described, and based on federal regulations concerning permissible concentrations in solution at the point of discharge to the accessible environment, a simple and conservative transport model, and baseline and potential worst-case release scenarios

Commercial Spent Nuclear Fuel Waste Package Misload Analysis

International Nuclear Information System (INIS)

J.K. Knudson

2003-01-01

The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M and O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis

Source, transport and dumping of radioactive waste

International Nuclear Information System (INIS)

1980-03-01

The results of an examination into the problems of radioactive waste are presented, in particular the sources, transport and dumping and the policy considerations in favour of specific methods. The theoretical background of radioactive waste is described, including the physical and chemical, ecological, medical and legal aspects. The practical aspects of radioactive waste in the Netherlands are considered, including the sources, the packaging and transport and dumping in the Atlantic Ocean. The politics and policies involved in this process are outlined. (C.F.)

Development of Specifications for Radioactive Waste Packages

International Nuclear Information System (INIS)

2006-10-01

The main objective of this publication is to provide guidelines for the development of waste package specifications that comply with waste acceptance requirements for storage and disposal of radioactive waste. It will assist waste generators and waste package producers in selecting the most significant parameters and in developing and implementing specifications for each individual type of waste and waste package. This publication also identifies and reviews the activities and technical provisions that are necessary to meet safety requirements; in particular, selection of the significant safety parameters and preparation of specifications for waste forms, waste containers and waste packages using proven approaches, methods and technologies. This report provides guidance using a systematic, stepwise approach, integrating the technical, organizational and administrative factors that need to be considered at each step of planning and implementing waste package design, fabrication, approval, quality assurance and control. The report reflects the considerable experience and knowledge that has been accumulated in the IAEA Member States and is consistent with the current international requirements, principles, standards and guidance for the safe management of radioactive waste

Development of Specifications for Radioactive Waste Packages

Energy Technology Data Exchange (ETDEWEB)

NONE

2006-10-15

The main objective of this publication is to provide guidelines for the development of waste package specifications that comply with waste acceptance requirements for storage and disposal of radioactive waste. It will assist waste generators and waste package producers in selecting the most significant parameters and in developing and implementing specifications for each individual type of waste and waste package. This publication also identifies and reviews the activities and technical provisions that are necessary to meet safety requirements; in particular, selection of the significant safety parameters and preparation of specifications for waste forms, waste containers and waste packages using proven approaches, methods and technologies. This report provides guidance using a systematic, stepwise approach, integrating the technical, organizational and administrative factors that need to be considered at each step of planning and implementing waste package design, fabrication, approval, quality assurance and control. The report reflects the considerable experience and knowledge that has been accumulated in the IAEA Member States and is consistent with the current international requirements, principles, standards and guidance for the safe management of radioactive waste.

The TN-GEMINI: experience on a versatile alpha waste transport container

International Nuclear Information System (INIS)

Roland, V.; Chanzy, Y.

2001-01-01

The present paper discusses experience gained in moving alpha wastes and its teachings regarding transport aspects of D and D. Alpha wastes are generated in fuel cycle facilities such as those involved in reprocessing, in manufacture of mixed oxide fuel, and by research laboratories. If a significant amount of wastes has to be transported, then a Type B packaging is required. Developed by Transnucleaire and COGEMA, the TN GEMINI container enables nuclear facilities operators to optimise their alpha waste transport management, and more generally contribute to their D and D projects. After describing succinctly the design of the TN GEMINI, the paper will explain how the packaging is being operated. Teachings from experience will be shared. (orig.)

Quality assurance requirements and methods for high level waste package acceptability

International Nuclear Information System (INIS)

1992-12-01

This document should serve as guidance for assigning the necessary items to control the conditioning process in such a way that waste packages are produced in compliance with the waste acceptance requirements. It is also provided to promote the exchange of information on quality assurance requirements and on the application of quality assurance methods associated with the production of high level waste packages, to ensure that these waste packages comply with the requirements for transportation, interim storage and waste disposal in deep geological formations. The document is intended to assist both the operators of conditioning facilities and repositories as well as national authorities and regulatory bodies, involved in the licensing of the conditioning of high level radioactive wastes or in the development of deep underground disposal systems. The document recommends the quality assurance requirements and methods which are necessary to generate data for these parameters identified in IAEA-TECDOC-560 on qualitative acceptance criteria, and indicates where and when the control methods can be applied, e.g. in the operation or commissioning of a process or in the development of a waste package design. Emphasis is on the control of the process and little reliance is placed on non-destructive or destructive testing. Qualitative criteria, relevant to disposal of high level waste, are repository dependent and are not addressed here. 37 refs, 3 figs, 2 tabs

IGNEOUS INTRUSION IMPACTS ON WASTE PACKAGES AND WASTE FORMS

International Nuclear Information System (INIS)

Bernot, P.

2004-01-01

The purpose of this model report is to assess the potential impacts of igneous intrusion on waste packages and waste forms in the emplacement drifts at the Yucca Mountain Repository. The models are based on conceptual models and includes an assessment of deleterious dynamic, thermal, hydrologic, and chemical impacts. The models described in this report constitute the waste package and waste form impacts submodel of the Total System Performance Assessment for the License Application (TSPA-LA) model assessing the impacts of a hypothetical igneous intrusion event on the repository total system performance. This submodel is carried out in accordance with Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of LA (BSC 2004 [DIRS:167796]) and Total System Performance Assessment-License Application Methods and Approaches (BSC 2003 [DIRS: 166296]). The technical work plan was prepared in accordance with AP-2.27Q, Planning for Science Activities. Any deviations from the technical work plan are documented in the following sections as they occur. The TSPA-LA approach to implementing the models for waste package and waste form response during igneous intrusion is based on identification of damage zones. Zone 1 includes all emplacement drifts intruded by the basalt dike, and Zone 2 includes all other emplacement drifts in the repository that are not in Zone 1. This model report will document the following model assessments: (1) Mechanical and thermal impacts of basalt magma intrusion on the invert, waste packages and waste forms of the intersected emplacement drifts of Zone 1. (2) Temperature and pressure trends of basaltic magma intrusion intersecting Zone 1 and their potential effects on waste packages and waste forms in Zone 2 emplacement drifts. (3) Deleterious volatile gases, exsolving from the intruded basalt magma and their potential effects on waste packages of Zone 2 emplacement drifts. (4) Post-intrusive physical

Development of the NIREX generic transport safety assessment to assist in the provision of waste packaging advice

International Nuclear Information System (INIS)

Hutchinson, D.L.; Marrison, A.R.; Sievwright, R.W.T.

2002-01-01

The current Nirex Mission is to provide the United Kingdom with safe, environmentally sound and publicly acceptable options for the long-term management of radioactive materials. As part of this role, Nirex has developed a phased deep geological disposal concept which is defined by six 'generic documents' that describe systems, processes and safety assessments that are not specific to any one location or geology. These generic documents give access to detailed information about the ideas and approaches that underpin the phased disposal concept, and have been published with an invitation to enter into dialogue with Nirex regarding these issues. The generic documents identify the requirements for an integrated transport system that would be necessary for the management of the intermediate-level (ILW) and low-level (LLW) wastes within Nirex's remit - the so-called reference case volume. This has involved Nirex in the development of transport hardware and associated safety reports and modelling and assessment tools for transport system logistics and system safety. Although the phased disposal concept is only one option for the long-term management of waste, the integrated transport system and associated modelling tools, is likely to be of equal relevance to other options. The safety assessment of the generic transport operation for the movement of ILW and LLW waste from waste producers' sites to a future radioactive waste disposal facility is described in one of the generic documents - the generic transport safety assessment (GTSA). The GTSA demonstrates that the transport operation is compliant with Nirex safety principles, and that the nuclear and non-nuclear risks to the public and workers from routine transport and from accidents are acceptable. This paper describes the types of risk that are calculated, and discusses the data requirements and calculation methodology. The verification and validation methodology is outlined, together with a discussion of the results

Management of waste from packaging of construction materials in building construction works

OpenAIRE

González Pericot, Natalia; Río Merino, Mercedes del

2011-01-01

Every material arriving at the construction site comes protected in some type of packaging, fundamentally cardboard, plastic or wood, and presently the great majority of these packagings finish in a container mixed with the rest of waste of the construction work. The increasing tendency to use prefabricated materials increases the volume of packaging necessary in product transport; in addition, the traditional materials also arrive more protected with packaging. A specific management for ...

Waste package materials selection process

International Nuclear Information System (INIS)

Roy, A.K.; Fish, R.L.; McCright, R.D.

1994-01-01

The office of Civilian Radioactive Waste Management (OCRWM) of the United States Department of Energy (USDOE) is evaluating a site at Yucca Mountain in Southern Nevada to determine its suitability as a mined geologic disposal system (MGDS) for the disposal of high-level nuclear waste (HLW). The B ampersand W Fuel Company (BWFC), as a part of the Management and Operating (M ampersand O) team in support of the Yucca Mountain Site Characterization Project (YMP), is responsible for designing and developing the waste package for this potential repository. As part of this effort, Lawrence Livermore National Laboratory (LLNL) is responsible for testing materials and developing models for the materials to be used in the waste package. This paper is aimed at presenting the selection process for materials needed in fabricating the different components of the waste package

Management of packaging waste in Poland--development agenda and accession to the EU.

Science.gov (United States)

Grodzińska-Jurczak, Małgorzata; Zakowska, Hanna; Read, Adam

2004-06-01

In recent years the issue of the municipal waste in Poland has become increasingly topical, with a considerable rise in the waste generation, much of which can be attributed to a boom in product packaging (mainly plastic). The annual production of plastics packaging has been constantly increasing over the last 20 to 30 years, and now exceeds 3.7 million tons. Due to a lack of processing technologies and poorly developed selective segregation system, packaging waste is still treated as a part of the municipal solid waste (MSW) stream, most of which is landfilled. As a result of Poland's access to the European Union, previous legal regulations governing municipal waste management have been harmonized with those binding on the member countries. One of the main changes, the most revolutionary one, is to make entrepreneurs liable for environmental risks resulting from the introduction of packaging to the market, and for its recycling. In practice, all entrepreneurs are to ensure recovery, and recycling, of used packaging from products introduced to the market at the required level. In recent year, the required recycling levels were fulfilled for all types of materials but mainly by large institutions using grouped and transport packaging waste for that matter. Household packaging gathered in the selective segregation system at the municipalities was practically left alone. This paper is an attempt to describe the system and assess the first year of functioning of the new, revamped system of packaging waste management in Poland. Recommendations are made relating to those features that need to be included in packaging waste management systems in order to maximize their sustainability and harmonization with the EU legal system.

Optimization of an impact limiter for radioactive waste packaging

International Nuclear Information System (INIS)

Mourao, Rogerio Pimenta; Mattar Neto, Miguel

1999-01-01

A certain class of packages for the transportation of radioactive wastes - type B packages in the transport jargon - is supposed to resist to a series of postulated tests, the most severe for the majority of the packages being the 9 m height drop test. To improve the performance of the packages under this test, impact limiters are added to them, normally as a removable overpack, with the primary goal of reducing the deceleration loads transmitted to the packages and their contents. The first impact limiter concept, developed during the '70s, used a shell-type impact limiter attached to both ends of the package. Later on, wood was tested as impact limiter filling, which improved the package's mechanical performance, but not its thermal resistance. The popularization of the polymeric materials and their growing use in engineer applications have led to the use of these materials in impact limiters, with the extra advantage of the polymers good thermal properties. This paper proposes a methodology for the optimization of an impact limiter for a package for the conditioning of spent sealed sources. Two simplified methods for the design of impact limiters are presented. Finally, a brief discussion is presented on the methodology usually employed in the design of accident-resisting packages. (author)

Waste disposal package

Science.gov (United States)

Smith, M.J.

1985-06-19

This is a claim for a waste disposal package including an inner or primary canister for containing hazardous and/or radioactive wastes. The primary canister is encapsulated by an outer or secondary barrier formed of a porous ceramic material to control ingress of water to the canister and the release rate of wastes upon breach on the canister. 4 figs.

A testing program to evaluate the effects of simulant mixed wastes on plastic transportation packaging components

International Nuclear Information System (INIS)

Nigrey, P.J.; Dickens, T.G.

1997-01-01

Based on regulatory requirements for Type A and B radioactive material packaging, a Testing Program was developed to evaluate the effects of mixed wastes on plastic materials which could be used as liners and seals in transportation containers. The plastics evaluated in this program were butadiene-acrylonitrile copolymer (Nitrile rubber), cross-linked polyethylene, epichlorohydrin, ethylene-propylene rubber (EPDM), fluorocarbons, high-density polyethylene (HDPE), butyl rubber, polypropylene, polytetrafluoroethylene, and styrene-butadiene rubber (SBR). These plastics were first screened in four simulant mixed wastes. The liner materials were screened using specific gravity measurements and seal materials by vapor transport rate (VTR) measurements. For the screening of liner materials, Kel-F, HDPE, and XLPE were found to offer the greatest resistance to the combination of radiation and chemicals. The tests also indicated that while all seal materials passed exposure to the aqueous simulant mixed waste, EPDM and SBR had the lowest VTRs. In the chlorinated hydrocarbon simulant mixed waste, only Viton passed the screening tests. In both the simulant scintillation fluid mixed waste and the ketone mixture waste, none of the seal materials met the screening criteria. Those materials which passed the screening tests were subjected to further comprehensive testing in each of the simulant wastes. The materials were exposed to four different radiation doses followed by exposure to a simulant mixed waste at three temperatures and four different exposure times (7, 14, 28, 180 days). Materials were tested by measuring specific gravity, dimensional, hardness, stress cracking, VTR, compression set, and tensile properties. The second phase of this Testing Program involving the comprehensive testing of plastic liner has been completed and for seal materials is currently in progress

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

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.

Development of waste packages for tuff

International Nuclear Information System (INIS)

Rothman, A.J.

1982-01-01

The objective of this program is to develop nuclear waste packages that meet the Nuclear Regulatory Commission's requirements for a licensed repository in tuff at the Nevada Test Site. Selected accomplishments for FY82 are: (1) Selection, collection of rock, and characterization of suitable outcrops (for lab experiments); (2) Rock-water interactions (Bullfrog Tuff); (3) Corrosion tests of ferrous metals; (4) Thermal modeling of waste package in host rock; (5) Preliminary fabrication tests of alternate backfills (crushed tuff); (6) Reviewed Westinghouse conceptual waste package designs for tuff and began modification for unsaturated zone; and (7) Waste Package Codes (BARIER and WAPPA) now running on our computer. Brief discussions are presented for rock-water interactions, corrosion tests of ferrous metals, and thermal and radionuclide migration modelling

Waste package performance in unsaturated rock

International Nuclear Information System (INIS)

Pigford, T.H.; Lee, W.W.-L.

1989-03-01

The unsaturated rock and near-atmospheric pressure of the potential nuclear waste repository at Yucca Mountain present new problems of predicting waste package performance. In this paper we present some illustrations of predictions of waste package performance and discuss important data needs. 11 refs., 9 figs., 1 tab

Chemical compatibility screening results of plastic packaging to mixed waste simulants

International Nuclear Information System (INIS)

Nigrey, P.J.; Dickens, T.G.

1995-01-01

We have developed a chemical compatibility program for evaluating transportation packaging components for transporting mixed waste forms. We have performed the first phase of this experimental program to determine the effects of simulant mixed wastes on packaging materials. This effort involved the screening of 10 plastic materials in four liquid mixed waste simulants. The testing protocol involved exposing the respective materials to ∼3 kGy of gamma radiation followed by 14 day exposures to the waste simulants of 60 C. The seal materials or rubbers were tested using VTR (vapor transport rate) measurements while the liner materials were tested using specific gravity as a metric. For these tests, a screening criteria of ∼1 g/m 2 /hr for VTR and a specific gravity change of 10% was used. It was concluded that while all seal materials passed exposure to the aqueous simulant mixed waste, EPDM and SBR had the lowest VTRs. In the chlorinated hydrocarbon simulant mixed waste, only VITON passed the screening tests. In both the simulant scintillation fluid mixed waste and the ketone mixture simulant mixed waste, none of the seal materials met the screening criteria. It is anticipated that those materials with the lowest VTRs will be evaluated in the comprehensive phase of the program. For specific gravity testing of liner materials the data showed that while all materials with the exception of polypropylene passed the screening criteria, Kel-F, HDPE, and XLPE were found to offer the greatest resistance to the combination of radiation and chemicals

Packaged low-level waste verification system

Energy Technology Data Exchange (ETDEWEB)

Tuite, K.; Winberg, M.R.; McIsaac, C.V. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

1995-12-31

The Department of Energy through the National Low-Level Waste Management Program and WMG Inc. have entered into a joint development effort to design, build, and demonstrate the Packaged Low-Level Waste Verification System. Currently, states and low-level radioactive waste disposal site operators have no method to independently verify the radionuclide content of packaged low-level waste that arrives at disposal sites for disposition. At this time, the disposal site relies on the low-level waste generator shipping manifests and accompanying records to ensure that low-level waste received meets the site`s waste acceptance criteria. The subject invention provides the equipment, software, and methods to enable the independent verification of low-level waste shipping records to ensure that the site`s waste acceptance criteria are being met. The objective of the prototype system is to demonstrate a mobile system capable of independently verifying the content of packaged low-level waste.

Risk management of onsite transportation of nuclear waste

International Nuclear Information System (INIS)

Field, J.G.; Wang, O.S.; Mercado, J.E.

1993-01-01

The United States Department of Energy (DOE) Hanford Site recently has undergone a significant change in mission. The focus of operations has shifted from plutonium production to environmental restoration. This transition has caused a substantial increase in quantities of nuclear waste and other hazardous materials packaged and transported onsite. In response to the escalating transportation activity, Westinghouse Hanford Company (Westinghouse Hanford), the Hanford Site operations and engineering contractor, is proposing an integrated risk assessment methodology and risk management strategy to enhance the safety of onsite packaging and transportation operations involving nuclear waste. The proposed methodology consists of three integral parts: risk assessment, risk acceptance criteria, and risk minimization. The purpose of the methodology is to ensure that the risk for each ongoing transportation activity is acceptable and to minimize the overall risk for current and future onsite operations. (authors). 2 figs., 6 refs

Risk management of onsite transportation of nuclear waste

International Nuclear Information System (INIS)

Field, J.G.; Wang, O.S.; Mercado, J.E.

1993-03-01

The United States Department of Energy (DOE) Hanford Site recently has undergone a significant change in mission. The focus of operations has shifted from plutonium production to environmental restoration. This transition has caused a substantial increase in quantities of nuclear waste and other hazardous materials packaged and transported onsite. In response to the escalating transportation activity, Westinghouse Hanford Company (Westinghouse Hanford), the Hanford Site operations and engineering contractor, is proposing an integrated risk assessment methodology and risk management strategy to enhance the safety of onsite packaging and transportation operations involving nuclear waste. The proposed methodology consists of three integral parts: risk assessment, risk acceptance criteria, and risk minimization. The purpose of the methodology is to ensure that the risk for each ongoing transportation activity is acceptable and to minimize the overall risk for current and future onsite operations

TRANSPORT LOCOMOTIVE AND WASTE PACKAGE TRANSPORTER ITS STANDARDS IDENTIFICATION STUDY

International Nuclear Information System (INIS)

Draper, K.D.

2005-01-01

To date, the project has established important to safety (ITS) performance requirements for structures, systems and components (SSCs) based on identification and categorization of event sequences that may result in a radiological release. These performance requirements are defined within the ''Nuclear Safety Design Basis for License Application'' (NSDB) (BSC 2005). Further, SSCs credited with performing safe functions are classified as ITS. In turn, performance confirmation for these SSCs is sought through the use of consensus code and standards. The purpose of this study is to identify applicable codes and standards for the waste package (WP) transporter and transport locomotive ITS SSCs. Further, this study will form the basis for selection and the extent of applicability of each code and standard. This study is based on the design development completed for License Application only. Accordingly, identification of ITS SSCs beyond those defined within the NSDB are based on designs that may be subject to further development during detail design. Furthermore, several design alternatives may still be under consideration to satisfy certain safety functions, and that final selection will not be determined until further design development has occurred. Therefore, for completeness, throughout this study alternative designs currently under consideration will be discussed. Further, the results of this study will be subject to evaluation as part of a follow-on gap analysis study. Based on the results of this study the gap analysis will evaluate each code and standard to ensure each ITS performance requirement is fully satisfied. When a performance requirement is not fully satisfied a ''gap'' is highlighted. Thereafter, the study will identify supplemental requirements to augment the code or standard to meet performance requirements. Further, the gap analysis will identify non-standard areas of the design that will be subject to a Development Plan. Non-standard components and

Transport, handling, and interim storage of intermediate-level transuranic waste at the INEL

International Nuclear Information System (INIS)

Metzger, J.C.; Snyder, A.M.

1977-09-01

The Idaho National Engineering Laboratory stores transuranic (TRU)-contaminated waste emitting significant amounts of beta-gamma radiation. This material is referred to as intermediate-level TRU waste. The Energy Research and Development Administration requires that this waste be stored retrievably during the interim before a Federal repository becomes operational. Waste form and packaging criteria for the eventual storage of this waste at a Federal repository, i.e., the Waste Isolation Pilot Plant (WIPP), have been tentatively established. The packaging and storage techniques now in use at the Idaho National Engineering Laboratory are compatible with these criteria and also meet the requirement that the waste containers remain in a readily-retrievable, contamination-free condition during the interim storage period. The Intermediate Level Transuranic Storage Facility (ILTSF) provides below-grade storage in steel pipe vaults for intermediate-level TRU waste prior to shipment to the WIPP. Designated waste generating facilities, operated for the Energy Research and Development Administration, use a variety of packaging and transportation methods to deliver this waste to the ILTSF. Transfer of the waste containers to the ILTSF storage vaults is accomplished using handling methods compatible with these waste packaging and transport methods

HMPT: Hazardous Waste Transportation Live 27928, Test 27929

Energy Technology Data Exchange (ETDEWEB)

Simpson, Lewis Edward [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-03-17

HMPT: Hazardous Waste Transportation (Live 27928, suggested one time and associated Test 27929, required initially and every 36 months) addresses the Department of Transportation (DOT) function-specific training requirements of the hazardous materials packagings and transportation (HMPT) Los Alamos National Laboratory (LANL) lab-wide training. This course addresses the requirements of the DOT that are unique to hazardous waste shipments. Appendix B provides the Title 40 Code of Federal Regulations (CFR) reference material needed for this course.

Current status and future considerations for a transportation system for spent fuel and radioactive waste

International Nuclear Information System (INIS)

Anderson, R.T.; Darr, D.G.; Godfrey, W.L.; Keely, R.B.; Lusk, E.C.; Peterson, R.W.; Ridihalgh, J.L.; Shallo, F.A.; Young, M.

1978-02-01

This report is part of the OWI Transportation/Logistics systems analysis of problems associated with shipping these wastes to waste terminal storage facilities. It covers governmental regulations and functional responsibilities, highway and rail transportation status and economic considerations, assessment of present industry capabilities and business-related considerations, important receiving facility considerations, necessary engineering and licensing-related aspects of packaging systems, and essential elements of reprocessing plant waste activities including packaging and transportation

Synthesis of knowledge on the long-term behaviour of concretes. Applications to cemented waste packages

International Nuclear Information System (INIS)

Richet, C.; Galle, C.; Le Bescop, P.; Peycelon, H.; Bejaoui, S.; Tovena, I.; Pointeau, I.; L'Hostis, V.; Levera, P.

2004-03-01

As stipulated in the former law of December 91 relating to 'concrete waste package', a progress report (phenomenological reference document) was first provided in 1999. The objective was to make an assessment of the knowledge acquired on the long-term behaviour of cement-based waste packages in the context of deep disposal and/or interim storage. The present document is an updated summary report. It takes into account a new knowledge assessment, considers coupled mechanisms and should contribute to the first performance studies (operational calculations). Handling and radio-nuclides (RN) confinement are the two major functional properties requested from the concrete used for the waste packages. In unsaturated environment (interim storage/disposal prior to closing), the main problem is the generation of cracks in the material. This aspect is a key parameter from the mechanical point of view (retrievability). It can have a major impact on the disposal phase (confinement). In saturated environment (disposal post-closing phase), the main concern is the chemical degradation of the waste package concrete submitted to underground waters leaching. In this context, the major thema are: the durability of the concretes under water (chemical degradation) and in unsaturated medium (corrosion of reinforcement), matter transport, RN retention, chemistry / transport / mechanical couplings. On the other hand, laboratory data on the behaviour of concretes are used to evaluate the RN source term of waste packages in function of time (concrete waste package OPerational Model, i.e. 'Concrete MOP'). The 'MOP' provides the physico-chemical description of the RN release in relationship with the waste package degradation itself. This description is based on simplified phenomenology for which only dimensioning mechanisms are taken into account. The use of Diffu-Ca code (basic module for the MOP) on the CASTEM numerical plate-form, already allows operational predictions. (authors)

From waste packages acceptance criteria to waste packages acceptance process at the Centre de l'Aube disposal facility

International Nuclear Information System (INIS)

Dutzer, M.

2003-01-01

The Centre de l'Aube disposal facility has now been operated for 10 years. At the end of 2001, about 124,000 m3 of low and intermediate level short lived waste packages, representing 180,000 packages, have been disposed, for a total capacity of 1,000,000 m3. The flow of waste packages is now between 12 and 15,000 m3 per year, that is one third of the flow that was taken into account for the design of the repository. It confirms the efforts by waste generators to minimise waste production. This flow represents 25 to 30,000 packages, 50% are conditioned into the compaction facility of the repository, so that 17,000 packages are disposed per year. 54 disposal vaults have been closed. In 1996-1999, the safety assessment of the repository have been reviewed, taking into account the experience of operation. This assessment was investigated by the regulatory body and, subsequently, a so-called 'definitive license' to operate was granted to ANDRA on September 2, 1999 with updated licensing requirements. Another review will be performed in 2004. To ensure a better consistency with the safety assessment of the facility, Andra issued new technical requirements for waste packages at the end of 2000. Discussions with waste generators also showed that the waste package acceptance process should be improved to provide a more precise definition of operational criteria to comply with in waste conditioning facilities. Consequently, a new approach has been implemented since 2000. (orig.)

Waste Generator Instructions: Key to Successful Implementation of the US DOE's 435.1 for Transuranic Waste Packaging Instructions (LA-UR-12-24155) - 13218

International Nuclear Information System (INIS)

French, David M.; Hayes, Timothy A.; Pope, Howard L.; Enriquez, Alejandro E.; Carson, Peter H.

2013-01-01

In times of continuing fiscal constraints, a management and operation tool that is straightforward to implement, works as advertised, and virtually ensures compliant waste packaging should be carefully considered and employed wherever practicable. In the near future, the Department of Energy (DOE) will issue the first major update to DOE Order 435.1, Radioactive Waste Management. This update will contain a requirement for sites that do not have a Waste Isolation Pilot Plant (WIPP) waste certification program to use two newly developed technical standards: Contact-Handled Defense Transuranic Waste Packaging Instructions and Remote-Handled Defense Transuranic Waste Packaging Instructions. The technical standards are being developed from the DOE O 435.1 Notice, Contact-Handled and Remote-Handled Transuranic Waste Packaging, approved August 2011. The packaging instructions will provide detailed information and instruction for packaging almost every conceivable type of transuranic (TRU) waste for disposal at WIPP. While providing specificity, the packaging instructions leave to each site's own discretion the actual mechanics of how those Instructions will be functionally implemented at the floor level. While the Technical Standards are designed to provide precise information for compliant packaging, the density of the information in the packaging instructions necessitates a type of Rosetta Stone that translates the requirements into concise, clear, easy to use and operationally practical recipes that are waste stream and facility specific for use by both first line management and hands-on operations personnel. The Waste Generator Instructions provide the operator with step-by-step instructions that will integrate the sites' various operational requirements (e.g., health and safety limits, radiological limits or dose limits) and result in a WIPP certifiable waste and package that can be transported to and emplaced at WIPP. These little known but widely productive Waste

Objectives for radioactive waste packaging

International Nuclear Information System (INIS)

Flowers, R.H.

1982-04-01

The report falls under the headings: introduction; the nature of radioactive wastes; how to manage radioactive wastes; packaging of radioactive wastes (supervised storage; disposal); waste form evaluation and test requirements (supervised storage; disposal); conclusions. (U.K.)

Evaluation of low and intermediate level radioactive solidified waste forms and packages

International Nuclear Information System (INIS)

1990-10-01

Evaluation of low and intermediate level radioactive waste forms and packages with respect to compliance with quality and safety requirements for transport, interim storage and disposal has become a very important part of the radioactive waste management strategy in many countries. The evaluation of waste forms and packages provides precise basic data for regulatory bodies to establish safety requirements, and implement quality control and quality assurance procedures for radioactive waste management programmes. The requirements depend very much upon the disposal option selected, treatment technology used, waste form characteristics, package quality and other factors. The regulatory requirements can also influence the methodology of waste form/package evaluation together with selection and analysis of data for quality control and safety assurance. A coordinated research programme started at the end of 1985 and brought together 12 participants from 11 countries. The results of the programme and each particular project were discussed at three Research Coordination Meetings held in Cairo, Egypt, in May, 1986; in Beijing, China, in April, 1998; and at Harwell Laboratory, United Kingdom, in November, 1989. This document summarises the salient features and results achieved during the four year investigation and a recommendation for future work in this area. Refs, figs and tabs

DEVELOPMENT OF THE TRU WASTE TRANSPORTATION FLEET--A SUCCESS STORY

International Nuclear Information System (INIS)

Devarakonda, Murthy; Morrison, Cindy; Brown, Mike

2003-01-01

Since March 1999, the Waste Isolation Pilot Plant (WIPP), located in southeastern New Mexico, has been operated by the U.S. Department of Energy (DOE), Carlsbad Field Office (CBFO), as a repository for the permanent disposal of defense-related transuranic (TRU) waste. More than 1,450 shipments of TRU waste for WIPP disposal have been completed, and the WIPP is currently receiving 12 to 16 shipments per week from five DOE sites around the nation. One of the largest fleets of Type B packagings supports the transportation of TRU waste to WIPP. This paper discusses the development of this fleet since the original Certificate of Compliance (C of C) for the Transuranic Package Transporter-II (TRUPACT-II) was issued by the U.S. Nuclear Regulatory Commission (NRC) in 1989. Evolving site programs, closure schedules of major sites, and the TRU waste inventory at the various DOE sites have directed the sizing and packaging mix of this fleet. This paper discusses the key issues that guided this fleet development, including the following: While the average weight of a 55-gallon drum packaging debris could be less than 300 pounds (lbs.), drums containing sludge waste or compacted waste could approach the maximum allowable weight of 1,000 lbs. A TRUPACT-II shipment may consist of three TRUPACT-II packages, each of which is limited to a total weight of 19,250 lbs. Payload assembly weights dictated by ''as-built'' TRUPACT-II weights limit each drum to an average weight of 312 lbs when three TRUPACT-IIs are shipped. To optimize the shipment of heavier drums, the HalfPACT packaging was designed as a shorter and lighter version of the TRUPACT-II to accommodate a heavier load. Additional packaging concepts are currently under development, including the ''TRUPACT-III'' packaging being designed to address ''oversized'' boxes that are currently not shippable in the TRUPACT-II or HalfPACT due to size constraints. Shipment optimization is applicable not only to the addition of new

Reference waste package environment report

International Nuclear Information System (INIS)

Glassley, W.E.

1986-01-01

One of three candidate repository sites for high-level radioactive waste packages is located at Yucca Mountain, Nevada, in rhyolitic tuff 700 to 1400 ft above the static water table. Calculations indicate that the package environment will experience a maximum temperature of ∼230 0 C at 9 years after emplacement. For the next 300 years the rock within 1 m of the waste packages will remain dehydrated. Preliminary results suggest that the waste package radiation field will have very little effect on the mechanical properties of the rock. Radiolysis products will have a negligible effect on the rock even after rehydration. Unfractured specimens of repository rock show no change in hydrologic characteristics during repeated dehydration-rehydration cycles. Fractured samples with initially high permeabilities show a striking permeability decrease during dehydration-rehydration cycling, which may be due to fracture healing via deposition of silica. Rock-water interaction studies demonstrate low and benign levels of anions and most cations. The development of sorptive secondary phases such as zeolites and clays suggests that anticipated rock-water interaction may produce beneficial changes in the package environment

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...

Transport of radioactive waste in Germany - a survey

International Nuclear Information System (INIS)

Alter, U.

1995-01-01

The transport of radioactive waste is centralised and coordinated by the German Railway Company (Deutsche Bahn AG, DB) in Germany. The conditioning of radioactive waste is now centralised and carried out by the Gesellschaft fuer Nucklear Service (GNS). The Germany Railway Company, DB, is totally and exclusively responsible for the transport, the GNS is totally and exclusively responsible for the conditioning of radioactive waste. The German Railway Company transports all radioactive waste from nuclear power plants, conditioning facilities and the existing intermediate storage facilities in Germany. In 1992 nearly 177 shipments of radioactive waste were carried out, in 1991 the total amount was 179 shipments. A brief description of the transport procedures, the use of different waste packages for radioactive waste with negligible heat generation and the transport routes within Germany will be given. For this purpose the inspection authorities in Germany have used a new documentation system, a special computer program for waste flow tracking and quality assurance and compliance assurance, developed by the electrical power companies in Germany. (Author)

Aging management assessment of type B transportation packages

International Nuclear Information System (INIS)

Sullivan, G.J.; Stahmer, U.; Freeman, E.L.

2004-01-01

The condition of a physical system such as a radioactive materials transportation package can change as it ages. The degree to which aging effects are identified, prevented or mitigated will depend on the types of inspections and maintenance performed on the critical components of the system. Routine inspections and maintenance may not address degradation mechanisms that are difficult to observe and can act over long periods of time. Aging management is a systematic effort to ensure that the system performs as designed over its entire service life and that degradation mechanisms do not prematurely end the service life. The Nuclear Waste Management Division (NWMD) of Ontario Power Generation (OPG) has developed an Aging Management Procedure and began performing aging management assessments on its Type B(U) packages. This paper discusses the Procedure and briefly describes the aging management assessment performed on the Roadrunner Transportation Package to demonstrate a practical application of the aging management process

Draft ASME code case on ductile cast iron for transport packaging

International Nuclear Information System (INIS)

Saegusa, T.; Arai, T.; Hirose, M.; Kobayashi, T.; Tezuka, Y.; Urabe, N.; Hueggenberg, R.

2004-01-01

The current Rules for Construction of ''Containment Systems for Storage and Transport Packagings of Spent Nuclear Fuel and High Level Radioactive Material and Waste'' of Division 3 in Section III of ASME Code (2001 Edition) does not include ductile cast iron in its list of materials permitted for use. The Rules specify required fracture toughness values of ferritic steel material for nominal wall thickness 5/8 to 12 inches (16 to 305 mm). New rule for ductile cast iron for transport packaging of which wall thickness is greater than 12 inches (305mm) is required

Regulatory aspects of the transport of high radiation level and alpha waste in France

International Nuclear Information System (INIS)

Devillers, C.; Grenier, M.; Lombard, J.; Mathieu, F.

1993-01-01

The introduction of the 10 mSv.h -1 at 3 m limit for LSA unshielded material makes it impossible to transport, as LSA material, the highest radiation level wastes from EdF PWR's operations. At present, the EdF's waste blocks can be transported as LSA III material by special arrangement. A new package design, equivalent to a Type B package, will be available for their transport before the end of the year 1995. It consists of a re-usable steel cylinder over-packing each block. Compliance of this package model with transport safety requirements will be demonstrated by taking into account the non-dispersability, as LSA III material, of the irradiating waste. A two-step approach has been accepted by the French Competent Authority for the transport of these wastes: (1) a specific ISO 20 container, thermally insulated, can be used by special arrangement for the transport of LSA combustible material having a total activity per conveyance higher than 100 A2. Furthermore, additional safety measures have to be implemented for these consignments. (2) After the end of 1995, a Type B package must be used for activity contents per conveyance higher than 100 A2. A specific 20' ISO container, complying with Type B requirements, is being developed for that purpose. (author)

Effects of sorption hysteresis on radionuclide releases from waste packages

International Nuclear Information System (INIS)

Barney, G.S.; Reed, D.T.

1985-01-01

A one-dimensional, numerical transport model was used to calculate radionuclide releases from waste packages emplaced in a nuclear waste repository in basalt. The model incorporates both sorption and desorption isotherm parameters measured previously for sorption of key radionuclides on the packing material component of the waste package. Sorption hysteresis as described by these isotherms lowered releases of some radionuclides by as much as two orders of magnitude. Radionuclides that have low molar inventories (relative to uranium), high solubility, and strongly sorbed, are most affected by sorption hysteresis. In these cases, almost the entire radionuclide inventory is sorbed on the packing material. The model can be used to help optimize the thickness of the packing material layer by comparing release rate versus packing material thickness curves with Nuclear Regulatory Commission (NRC) and Environmental Protection Agency (EPA) release limits

PACCOM: A nuclear waste packaging facility cost model: Draft technical report

International Nuclear Information System (INIS)

Dippold, D.G.; Tzemos, S.; Smith, D.J.

1985-05-01

PACCOM is a computerized, parametric model used to estimate the capital, operating, and decommissioning costs of a variety of nuclear waste packaging facility configurations. The model is based upon a modular waste packaging facility concept from which functional components of the overall facility have been identified and their design and costs related to various parameters such as waste type, waste throughput, and the number of operational shifts employed. The model may be used to either estimate the cost of a particular waste packaging facility configuration or to explore the cost tradeoff between plant capital and labor. That is, one may use the model to search for the particular facility sizes and associated cost which when coupled with a particular number of shifts, and thus staffing level, leads to the lowest overall total cost. The functional components which the model considers include hot cells and their supporting facilities, transportation, cask handling facilities, transuranic waste handling facilities, and administrative facilities such as warehouses, security buildings, maintenance buildings, etc. The cost of each of these functional components is related either directly or indirectly to the various independent design parameters. Staffing by shift is reported into direct and indirect support labor. These staffing levels are in turn related to the waste type, waste throughput, etc. 2 refs., 11 figs., 3 tabs

Scientific investigation plan for NNWSI WBS element 1.2.2.5.L: NNWSI waste package performance assessment: Revision 1

International Nuclear Information System (INIS)

Eggert, K.G.; O'Connell, W.J.; Lappa, D.A.

1986-01-01

Waste package performance assessment contains three broad categories of activities. These activities are: (1) development of a hydrothermal flow and transport model to test concepts to be used in establishing boundary conditions for performance calculations, and to interface EBS release calculations with total system performance calculations; (2) development of a waste package systems model to provide integrated deterministic assessments of performance and analyses of waste package designs; and (3) development of an uncertainty methodology for combination with the system model to perform probabilistic reliability and performance analysis waste package designs. The first category contains activities that aid in determining the scope of a separate, simplified set of hydrologic calculations needed to characterize the waste package environment for performance assessment calculations. The last two activity categories are directly concerned with waste package performance calculations. A rationale for each activity under these groups is presented. All of the activities of performance assessment are either code development or analyses of waste package problems

Engineered waste-package-system design specification

International Nuclear Information System (INIS)

1983-05-01

This report documents the waste package performance requirements and geologic and waste form data bases used in developing the conceptual designs for waste packages for salt, tuff, and basalt geologies. The data base reflects the latest geotechnical information on the geologic media of interest. The parameters or characteristics specified primarily cover spent fuel, defense high-level waste, and commercial high-level waste forms. The specification documents the direction taken during the conceptual design activity. A separate design specification will be developed prior to the start of the preliminary design activity

Igneous Intrusion Impacts on Waste Packages and Waste Forms

International Nuclear Information System (INIS)

P. Bernot

2004-01-01

The purpose of this model report is to assess the potential impacts of igneous intrusion on waste packages and waste forms in the emplacement drifts at the Yucca Mountain Repository. The model is based on conceptual models and includes an assessment of deleterious dynamic, thermal, hydrologic, and chemical impacts. This constitutes the waste package and waste form impacts submodel of the Total System Performance Assessment for the License Application (TSPA-LA) model assessing the impacts of a hypothetical igneous intrusion event on the repository total system performance. This submodel is carried out in accordance with Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of SR and LA (BSC 2003a) and Total System Performance Assessment-License Application Methods and Approaches (BSC 2002a). The technical work plan is governed by the procedures of AP-SIII.10Q, Models. Any deviations from the technical work plan are documented in the TSPA-LA approach to implementing the models for waste package and waste form response during igneous intrusion is based on identification of damage zones. Zone 1 includes all emplacement drifts intruded by the basalt dike, and Zone 2 includes all other emplacement drifts in the repository that are not in Zone 1. This model report will document the following model: (1) Impacts of magma intrusion on the components of engineered barrier system (e.g., drip shields and cladding) of emplacement drifts in Zone 1, and the fate of waste forms. (2) Impacts of conducting magma heat and diffusing magma gases on the drip shields, waste packages, and cladding in the Zone 2 emplacement drifts adjacent to the intruded drifts. (3) Impacts of intrusion on Zone 1 in-drift thermal and geochemical environments, including seepage hydrochemistry. The scope of this model only includes impacts to the components stated above, and does not include impacts to other engineered barrier system (EBS) components such as the invert and

Igneous Intrusion Impacts on Waste Packages and Waste Forms

Energy Technology Data Exchange (ETDEWEB)

P. Bernot

2004-08-16

The purpose of this model report is to assess the potential impacts of igneous intrusion on waste packages and waste forms in the emplacement drifts at the Yucca Mountain Repository. The model is based on conceptual models and includes an assessment of deleterious dynamic, thermal, hydrologic, and chemical impacts. This constitutes the waste package and waste form impacts submodel of the Total System Performance Assessment for the License Application (TSPA-LA) model assessing the impacts of a hypothetical igneous intrusion event on the repository total system performance. This submodel is carried out in accordance with Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of SR and LA (BSC 2003a) and Total System Performance Assessment-License Application Methods and Approaches (BSC 2002a). The technical work plan is governed by the procedures of AP-SIII.10Q, Models. Any deviations from the technical work plan are documented in the TSPA-LA approach to implementing the models for waste package and waste form response during igneous intrusion is based on identification of damage zones. Zone 1 includes all emplacement drifts intruded by the basalt dike, and Zone 2 includes all other emplacement drifts in the repository that are not in Zone 1. This model report will document the following model: (1) Impacts of magma intrusion on the components of engineered barrier system (e.g., drip shields and cladding) of emplacement drifts in Zone 1, and the fate of waste forms. (2) Impacts of conducting magma heat and diffusing magma gases on the drip shields, waste packages, and cladding in the Zone 2 emplacement drifts adjacent to the intruded drifts. (3) Impacts of intrusion on Zone 1 in-drift thermal and geochemical environments, including seepage hydrochemistry. The scope of this model only includes impacts to the components stated above, and does not include impacts to other engineered barrier system (EBS) components such as the invert and

Greater-than-Class C low-level radioactive waste shipping package/container identification and requirements study. National Low-Level Waste Management Program

Energy Technology Data Exchange (ETDEWEB)

Tyacke, M.

1993-08-01

This report identifies a variety of shipping packages (also referred to as casks) and waste containers currently available or being developed that could be used for greater-than-Class C (GTCC) low-level waste (LLW). Since GTCC LLW varies greatly in size, shape, and activity levels, the casks and waste containers that could be used range in size from small, to accommodate a single sealed radiation source, to very large-capacity casks/canisters used to transport or dry-store highly radioactive spent fuel. In some cases, the waste containers may serve directly as shipping packages, while in other cases, the containers would need to be placed in a transport cask. For the purpose of this report, it is assumed that the generator is responsible for transporting the waste to a Department of Energy (DOE) storage, treatment, or disposal facility. Unless DOE establishes specific acceptance criteria, the receiving facility would need the capability to accept any of the casks and waste containers identified in this report. In identifying potential casks and waste containers, no consideration was given to their adequacy relative to handling, storage, treatment, and disposal. Those considerations must be addressed separately as the capabilities of the receiving facility and the handling requirements and operations are better understood.

The effect of proposed changes to the IAEA transport regulations on decommissioning and other low level waste transportation

International Nuclear Information System (INIS)

Shetler, A.P.; Jayawardene, N.

1986-01-01

Ontario Hydro has studied the proposed changes to the IAEA Transport Regulations contained in the 1985 4th draft edition. The study shows that these proposed changes will have significant strategic and economic effects on the transportation of radioactive decommissioning and other low-level wastes. Under the 1985 4th draft edition, the definition of Low Specific Activity (LSA) material is revised and a new regulation is proposed which restricts the quantity of LSA material or Surface Contaminated Object (SCO) in a single package to that which would have an unshielded dose rate of 10 mSv/h (1 rem/h) at 3 m. The objective of this paper is to highlight the implications of the proposed regulatory changes. An example of the impact of these changes is presented by considering the transportation of typical CANDU decommissioning wastes which arise through piece-by-piece removal of a reactor assembly. The potential effect of the changes is that less decommissioning waste can be shipped in a single low-level waste package. This results in the requirement for so many small, low-level waste packages that Type B shipments are strategically and economically more attractive. However, use of Type B shipments would also result in higher dose uptake and waste management costs than under the 1973 Regulations

Contribution to internal pressure and flammable gas concentration in RAM transport packages

International Nuclear Information System (INIS)

Warrant, M.M.; Brown, N.

1989-01-01

Various facilities in the US operated by the US Department of Energy generate wastes contaminated with transuranic (TRU) isotopes (such as plutonium and americium) that decay primarily by emission of alpha particles. The alpha particles lose energy in their passage through matter and change the material chemically in the process called radiolysis. The waste materials consist of a wide variety of commercially available plastics, paper, cloth, and rubber; concreted or sludge wastes containing water; and metals, glass, and other solid inorganic materials. TRU wastes that have surface dose rates of 200 mrem/hr or less are typically packaged in plastic bags placed inside metal drums or boxes that are vented through high efficiency particulate air (HEPA) filters. These wastes are to be transported from waste generation or storage sites to the Waste Isolation Pilot Plant (WIPP) in the TRUPACT-II, a Type B package

Safety analysis report for packaging (onsite) for the Waste Encapsulation and Storage Facility ion exchange module

International Nuclear Information System (INIS)

Romano, T.

1997-01-01

The Waste Encapsulation and Storage Facility (WESF) is in need of providing an emergency ion exchange system to remove cesium or strontium from the pool cell in the event of a capsule failure. The emergency system is call the WESF Emergency Ion Exchange System and the packaging is called the WESF ion exchange module (WIXM). The packaging system will meet the onsite transportation requirements for a Type B, highway route controlled quantity package as well as disposal requirements for Category 3 waste

Waste Generator Instructions: Key to Successful Implementation of the US DOE's 435.1 for Transuranic Waste Packaging Instructions (LA-UR-12-24155) - 13218

Energy Technology Data Exchange (ETDEWEB)

French, David M. [LANL EES-12, Carlsbad, NM, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Hayes, Timothy A. [LANL EES-12, Carlsbad, NM, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Pope, Howard L. [Aspen Resources Ltd., Inc., P.O. Box 3038, Boulder, CO 80307 (United States); Enriquez, Alejandro E. [LANL NCO-4, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Carson, Peter H. [LANL NPI-7, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

2013-07-01

In times of continuing fiscal constraints, a management and operation tool that is straightforward to implement, works as advertised, and virtually ensures compliant waste packaging should be carefully considered and employed wherever practicable. In the near future, the Department of Energy (DOE) will issue the first major update to DOE Order 435.1, Radioactive Waste Management. This update will contain a requirement for sites that do not have a Waste Isolation Pilot Plant (WIPP) waste certification program to use two newly developed technical standards: Contact-Handled Defense Transuranic Waste Packaging Instructions and Remote-Handled Defense Transuranic Waste Packaging Instructions. The technical standards are being developed from the DOE O 435.1 Notice, Contact-Handled and Remote-Handled Transuranic Waste Packaging, approved August 2011. The packaging instructions will provide detailed information and instruction for packaging almost every conceivable type of transuranic (TRU) waste for disposal at WIPP. While providing specificity, the packaging instructions leave to each site's own discretion the actual mechanics of how those Instructions will be functionally implemented at the floor level. While the Technical Standards are designed to provide precise information for compliant packaging, the density of the information in the packaging instructions necessitates a type of Rosetta Stone that translates the requirements into concise, clear, easy to use and operationally practical recipes that are waste stream and facility specific for use by both first line management and hands-on operations personnel. The Waste Generator Instructions provide the operator with step-by-step instructions that will integrate the sites' various operational requirements (e.g., health and safety limits, radiological limits or dose limits) and result in a WIPP certifiable waste and package that can be transported to and emplaced at WIPP. These little known but widely

Nuclear Waste Package Mockups: A Study of In-situ Redox State

Science.gov (United States)

Helean, K.; Anderson, B.; Brady, P. V.

2006-05-01

The Yucca Mountain Repository (YMR), located in southern Nevada, is to be the first facility in the U.S. for the permanent disposal of high-level radioactive waste and spent nuclear fuels. Total system performance assessment(TSPA) has indicated that among the major radionuclides contributing to dose are Np, Tc, and I. These three radionuclides are mobile in most geochemical settings, and therefore sequestering them within the repository horizon is a priority for the Yucca Mountain Project (YMP). Corroding steel may offset radionuclide transport processes within the proposed waste packages at YMR by retaining radionuclides, creating locally reducing conditions, and reducing porosity. Ferrous iron has been shown to reduce UO22+ to UO2s, and some ferrous iron-bearing ion-exchange materials have been shown to adsorb radionuclides and heavy metals. Locally reducing conditions may lead to the reduction and subsequent immobilization of problematic dissolved species such as TcO4-, NpO2+, and UO22+ and can also inhibit corrosion of spent nuclear fuel. Water occluded during corrosion produces bulky corrosion products, and consequently less porosity is available for water and radionuclide transport. The focus of this study is on the nature of Yucca Mountain waste package corrosion products and their effects on local redox conditions, radionuclide transport, and porosity. In order to measure in-situ redox, six small-scale (1:40) waste package mockups were constructed using A516 and 316 stainless steel, the same materials as the proposed Yucca Mountain waste packages. The mockups are periodically injected with a simulated groundwater and the accumulated effluent and corrosion products are evaluated for their Fe(II)/Fe(III) content and mineralogy. Oxygen fugacities are then calculated and, thus, in-situ redox conditions are determined. Early results indicate that corrosion products are largely amorphous Fe-oxyhydroxides, goethite and magnetite. That information together with the

Nuclear-waste-package materials degradation modes and accelerated testing

International Nuclear Information System (INIS)

1981-09-01

This report reviews the materials degradation modes that may affect the long-term behavior of waste packages for the containment of nuclear waste. It recommends an approach to accelerated testing that can lead to the qualification of waste package materials in specific repository environments in times that are short relative to the time period over which the waste package is expected to provide containment. This report is not a testing plan but rather discusses the direction for research that might be considered in developing plans for accelerated testing of waste package materials and waste forms

Thermal analysis of NNWSI conceptual waste package designs

International Nuclear Information System (INIS)

Stein, W.; Hockman, J.N.; O'Neal, W.C.

1984-04-01

Lawrence Livermore National Laboratory is involved in the design and testing of high-level nuclear waste packages. Many of the aspects of waste package design and testing (e.g., corrosion and leaching) depend in part on the temperature history of the emplaced packages. This report discusses thermal modeling and analysis of various emplaced waste package conceptual designs including the models used, the assumptions and approximations made, and the results obtained. 16 references

EQ6 Calculations for Chemical Degradation of Navy Waste Packages

International Nuclear Information System (INIS)

S. LeStrange

1999-01-01

The Monitored Geologic Repository Waste Package Operations 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 Navy (Refs. 1 and 2). The Navy SNF has been considered for disposal at the potential Yucca Mountain site. For some waste packages, the containment may breach (Ref. 3), allowing the influx of water. Water in the waste package may moderate neutrons, increasing the likelihood of a criticality event within the waste package. The water may gradually leach the fissile components and neutron absorbers out of the waste package. In addition, the accumulation of silica (SiO 2 ) in the waste package over time may further affect the neutronics of the system. This study presents calculations of the long-term geochemical behavior of waste packages containing the Enhanced Design Alternative (EDA) II inner shell, Navy canister, and basket components. The calculations do not include the Navy SNF in the waste package. The specific study objectives were to determine the chemical composition of the water and the quantity of silicon (Si) and other solid corrosion products in the waste package during the first million years after the waste package is breached. The results of this calculation will be used to ensure that the type and amount of criticality control material used in the waste package design will prevent criticality

EQ6 Calculations for Chemical Degradation of Navy Waste Packages

Energy Technology Data Exchange (ETDEWEB)

S. LeStrange

1999-11-15

The Monitored Geologic Repository Waste Package Operations 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 Navy (Refs. 1 and 2). The Navy SNF has been considered for disposal at the potential Yucca Mountain site. For some waste packages, the containment may breach (Ref. 3), allowing the influx of water. Water in the waste package may moderate neutrons, increasing the likelihood of a criticality event within the waste package. The water may gradually leach the fissile components and neutron absorbers out of the waste package. In addition, the accumulation of silica (SiO{sub 2}) in the waste package over time may further affect the neutronics of the system. This study presents calculations of the long-term geochemical behavior of waste packages containing the Enhanced Design Alternative (EDA) II inner shell, Navy canister, and basket components. The calculations do not include the Navy SNF in the waste package. The specific study objectives were to determine the chemical composition of the water and the quantity of silicon (Si) and other solid corrosion products in the waste package during the first million years after the waste package is breached. The results of this calculation will be used to ensure that the type and amount of criticality control material used in the waste package design will prevent criticality.

Structural analysis of the TRansUranic PACkage Transporter (TRUPACT)

International Nuclear Information System (INIS)

Lamoreaux, G.H.; Sutherland, S.H.; Duffey, T.A.

1981-07-01

The TRansUranic PACkage Transporter (TRUPACT) is a Type B container under development at the Transportation Technology Center, Sandia National Laboratory, for use in the transportation of contact-handled transuranic waste. This report describes the numerical analyses of the container's response to end-on, side-on, and center of gravity over corner impacts on an unyielding surface following a 9 m free fall. The results of the analyses are compared to available experimental data. In general, the analytical predictions and experimental comparisons confirm the validity of the TRUPACT design concept

Low-level waste packaging--a managerial perspective

International Nuclear Information System (INIS)

Motl, G.P.; Hebbard, L.B. Jr.

1980-01-01

This paper emphasizes managerial responsibility for assuring that facility waste is properly packaged. Specifically, existing packaging regulations are summarized, several actual violations are reviewed and, lastly, some recommendations are made to assist managerial personnel in fulfilling their responsibility to ensure that low-level waste is packaged safely and properly before shipment to the disposal site

Broad survey of radioactive waste transports and trends

International Nuclear Information System (INIS)

Blum, P.T.

1986-01-01

This survey reviews Radioactive Waste (RW) transports in different countries, giving the origins, amounts, distances and modes of transport for the different RW categories currently moved. It appears that: present transport experience concerns mainly Low and Medium level Wastes (LMW) and Spent nuclear Fuel (SF); RW transports are implemented in compliance with IAEA recommendations which proved particularly helpful to gain public acceptance; the bulk of LMW is transported as Low Specific Activity (LSA) materials (i.e. with packaging requirements just equivalent to those needed for their disposal) and SF as well as liquid RW are transported in 'type B' packagings which withstand severe accident conditions; records indicate that transport worker irradiation is kept much below allowable limits and that accidents during RW transports did not cause significant release of activity to environment; and the cost of RW transports including insurances, is small compared to that of RW processing and disposal. Therefore, one may comment that RW movements are by no means slowed down by transport problems and just follow the demand which in some countries may be delayed by reasons linked to RW disposal strategy or its public acceptance. This survey presents also an assessment of RW transports until the end of this century

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

Greater-than-Class C low-level radioactive waste shipping package/container identification and requirements study

International Nuclear Information System (INIS)

Tyacke, M.

1993-08-01

This report identifies a variety of shipping packages (also referred to as casks) and waste containers currently available or being developed that could be used for greater-than-Class C (GTCC) low-level waste (LLW). Since GTCC LLW varies greatly in size, shape, and activity levels, the casks and waste containers that could be used range in size from small, to accommodate a single sealed radiation source, to very large-capacity casks/canisters used to transport or dry-store highly radioactive spent fuel. In some cases, the waste containers may serve directly as shipping packages, while in other cases, the containers would need to be placed in a transport cask. For the purpose of this report, it is assumed that the generator is responsible for transporting the waste to a Department of Energy (DOE) storage, treatment, or disposal facility. Unless DOE establishes specific acceptance criteria, the receiving facility would need the capability to accept any of the casks and waste containers identified in this report. In identifying potential casks and waste containers, no consideration was given to their adequacy relative to handling, storage, treatment, and disposal. Those considerations must be addressed separately as the capabilities of the receiving facility and the handling requirements and operations are better understood

Safety transport of radioactive waste in the nuclear power area

Directory of Open Access Journals (Sweden)

Tureková Ivana

2017-01-01

Full Text Available Radioactive wastes require strict rules for manipulation with them due to the hazards for the human health and environment, not excluding the hazards during their internal transport. The article deals with the transport of packing unit inside of the company and it proposes the possible alternatives so that meet the limit conditions and reduce the manipulation time with the radioactive material in the packing unit. The packing unite isolates fixated liquid waste from the environment while it also serves as protection. There are also important external radiation characteristics of package unit, which consist of measurable values of the scratch contamination surface and dose power on the surface of package unit. Thus, the paper is aimed to point out the necessity of the logistics during manipulation with the package unit in the process of internal transport so that the dose power of exposed employees would achieve the lowest possible level and meet the strict limits in a full extent.

Packaging and transportation of radioactive liquid at the U.S. Department of Energy Hanford Site

International Nuclear Information System (INIS)

Smith, R.J.

1995-02-01

Beginning in the 1940's, radioactive liquid waste has been generated at the US Department of Energy (DOE) Hanford Site as a result of defense material production. The liquid waste is currently stored in 177 underground storage tanks. As part of the tank remediation efforts, Type B quantity packagings for the transport of large volumes of radioactive liquids are required. There are very few Type B liquid packagings in existence because of the rarity of large-volume radioactive liquid payloads in the commercial nuclear industry. Development of aboveground transport systems for large volumes of radioactive liquids involves institutional, economic, and technical issues. Although liquid shipments have taken place under DOE-approved controlled conditions within the boundaries of the Hanford Site for many years, offsite shipment requires compliance with DOE, US Nuclear Regulatory Commission (NRC), and US Department of Transportation (DOT) directives and regulations. At the present time, no domestic DOE nor NRC-certified Type B packagings with the appropriate level of shielding are available for DOT-compliant transport of radioactive liquids in bulk volumes. This paper will provide technical details regarding current methods used to transport such liquids on and off the Hanford Site, and will provide a status of packaging development programs for future liquid shipments

TECHNICAL EVALUATION OF THE SAFE TRANSPORTATION OF WASTE CONTAINERS COATED WITH POLYUREA

Energy Technology Data Exchange (ETDEWEB)

VAIL, T.S.

2007-03-30

This technical report is to evaluate and establish that the transportation of waste containers (e.g. drums, wooden boxes, fiberglass-reinforced plywood (FRP) or metal boxes, tanks, casks, or other containers) that have an external application of polyurea coating between facilities on the Hanford Site can be achieved with a level of onsite safety equivalent to that achieved offsite. Utilizing the parameters, requirements, limitations, and controls described in the DOE/RL-2001-36, ''Hanford Sitewide Transportation Safety Document'' (TSD) and the Department of Energy Richland Operations (DOE-RL) approved package specific authorizations (e.g. Package Specific Safety Documents (PSSDs), One-Time Requests for Shipment (OTRSs), and Special Packaging Authorizations (SPAS)), this evaluation concludes that polyurea coatings on packages does not impose an undue hazard for normal and accident conditions. The transportation of all packages on the Hanford Site must comply with the transportation safety basis documents for that packaging system. Compliance with the requirements, limitations, or controls described in the safety basis for a package system will not be relaxed or modified because of the application of polyurea. The inspection criteria described in facility/projects procedures and work packages that ensure compliance with Container Management Programs and transportation safety basis documentation dictate the need to overpack a package without consideration for polyurea. This technical report reviews the transportation of waste packages coated with polyurea and does not credit the polyurea with enhancing the structural, thermal, containment, shielding, criticality, or gas generating posture of a package. Facilities/Projects Container Management Programs must determine if a container requires an overpack prior to the polyurea application recognizing that circumstances newly discovered surface contamination or loss of integrity may require a previously

TECHNICAL EVALUATION OF THE SAFE TRANSPORTATION OF WASTE CONTAINERS COATED WITH POLYUREA

International Nuclear Information System (INIS)

VAIL, T.S.

2007-01-01

This technical report is to evaluate and establish that the transportation of waste containers (e.g. drums, wooden boxes, fiberglass-reinforced plywood (FRP) or metal boxes, tanks, casks, or other containers) that have an external application of polyurea coating between facilities on the Hanford Site can be achieved with a level of onsite safety equivalent to that achieved offsite. Utilizing the parameters, requirements, limitations, and controls described in the DOE/RL-2001-36, ''Hanford Sitewide Transportation Safety Document'' (TSD) and the Department of Energy Richland Operations (DOE-RL) approved package specific authorizations (e.g. Package Specific Safety Documents (PSSDs), One-Time Requests for Shipment (OTRSs), and Special Packaging Authorizations (SPAS)), this evaluation concludes that polyurea coatings on packages does not impose an undue hazard for normal and accident conditions. The transportation of all packages on the Hanford Site must comply with the transportation safety basis documents for that packaging system. Compliance with the requirements, limitations, or controls described in the safety basis for a package system will not be relaxed or modified because of the application of polyurea. The inspection criteria described in facility/projects procedures and work packages that ensure compliance with Container Management Programs and transportation safety basis documentation dictate the need to overpack a package without consideration for polyurea. This technical report reviews the transportation of waste packages coated with polyurea and does not credit the polyurea with enhancing the structural, thermal, containment, shielding, criticality, or gas generating posture of a package. Facilities/Projects Container Management Programs must determine if a container requires an overpack prior to the polyurea application recognizing that circumstances newly discovered surface contamination or loss of integrity may require a previously un

Yucca Mountain Site Characterization Project Waste Package Plan

International Nuclear Information System (INIS)

Harrison-Giesler, D.J.; Jardine, L.J.

1991-02-01

The goal of the US Department of Energy's (DOE) Yucca Mountain Site Characterization Project (YMP) waste package program is to develop, confirm the effectiveness of, and document a design for a waste package and associated engineered barrier system (EBS) for spent nuclear fuel and solidified high-level nuclear waste (HLW) that meets the applicable regulatory requirements for a geologic repository. The Waste Package Plan describes the waste package program and establishes the technical approach against which overall progress can be measured. It provides guidance for execution and describes the essential elements of the program, including the objectives, technical plan, and management approach. The plan covers the time period up to the submission of a repository license application to the US Nuclear Regulatory Commission (NRC). 1 fig

WASTE PACKAGE CORROSION STUDIES USING SMALL MOCKUP EXPERIMENTS

International Nuclear Information System (INIS)

B.E. Anderson; K.B. Helean; C.R. Bryan; P.V. Brady; R.C. Ewing

2005-01-01

The corrosion of spent nuclear fuel and subsequent mobilization of radionuclides is of great concern in a geologic repository, particularly if conditions are oxidizing. Corroding A516 steel may offset these transport processes within the proposed waste packages at the Yucca Mountain Repository (YMR) by retaining radionuclides, creating locally reducing conditions, and reducing porosity. Ferrous iron, Fe 2+ , has been shown to reduce UO 2 2+ to UO 2(s) [1], and some ferrous iron-bearing ion-exchange materials adsorb radionuclides and heavy metals [2]. Of particular interest is magnetite, a potential corrosion product that has been shown to remove TcO 4 - from solution [3]. Furthermore, if Fe 2+ minerals, rather than fully oxidized minerals such as goethite, are produced during corrosion, then locally reducing conditions may be present. High electron availability leads to the reduction and subsequent immobilization of problematic dissolved species such as TcO 4 - , NpO 2 + , and UO 2 2+ and can also inhibit corrosion of spent nuclear fuel. Finally, because the molar volume of iron material increases during corrosion due to oxygen and water incorporation, pore space may be significantly reduced over long time periods. The more water is occluded, the bulkier the corrosion products, and the less porosity is available for water and radionuclide transport. The focus of this paper is on the nature of Yucca Mountain waste package steel corrosion products and their effects on local redox state, radionuclide transport, and porosity

Review of DOE waste package program. Subtask 1.1. National waste package program, April-September 1983. Volume 5

Energy Technology Data Exchange (ETDEWEB)

Soo, P. (ed.)

1984-08-01

The current effort is part of an ongoing task to review the national high-level waste package effort. It includes evaluations of reference waste form, container, and packing material components with respect to determining how they may contribute to the containment and controlled release of radionuclides after waste packages have been emplaced in salt, basalt, and tuff repositories. In the current Biannual Report a section on carbon steel container corrosion has been included to complement prior work on TiCode-12 and Type 304 stainless steel. The use of crushed tuff as a packing material is discussed and waste package component interaction test data are included. Licensing data requirements to estimate the degree of compliance with NRC performance objectives are specified. 41 figures, 24 tables.

Review of DOE waste package program. Subtask 1.1. National waste package program, April-September 1983. Volume 5

International Nuclear Information System (INIS)

Soo, P.

1984-08-01

The current effort is part of an ongoing task to review the national high-level waste package effort. It includes evaluations of reference waste form, container, and packing material components with respect to determining how they may contribute to the containment and controlled release of radionuclides after waste packages have been emplaced in salt, basalt, and tuff repositories. In the current Biannual Report a section on carbon steel container corrosion has been included to complement prior work on TiCode-12 and Type 304 stainless steel. The use of crushed tuff as a packing material is discussed and waste package component interaction test data are included. Licensing data requirements to estimate the degree of compliance with NRC performance objectives are specified. 41 figures, 24 tables

WIPP waste package testing on simulated DHLW: emplacement

International Nuclear Information System (INIS)

Molecke, M.A.

1984-01-01

Several series of simulated (nonradioactive) defense high-level waste (DHLW) package tests have been emplaced in the WIPP, a research and development facility authorized to demonstrate the safe disposal of defense-related wastes. The primary purpose of these 3-to-7 year duration tests is to evaluate the in situ materials performance of waste package barriers (canisters, overpacks, backfills, and nonradioactive DHLW glass waste form) for possible future application to a licensed waste repository in salt. This paper describes all test materials, instrumentation, and emplacement and testing techniques, and discusses progress of the various tests. These tests are intended to provide information on materials behavior (i.e., corrosion, metallurgical and geochemical alterations, waste form durability, surface interactions, etc.), as well as comparison between several waste package designs, fabrications details, and actual costs. These experiments involve 18 full-size simulated DHLW packages (approximately 3.0 m x 0.6 m diameter) emplaced in vertical boreholes in the salt drift floor. Six of the test packages contain internal electrical heaters (470 W/canister), and were emplace under approximately reference DHLW repository conditions. Twelve other simulated DHLW packages were emplaced under accelerated-aging or overtest conditions, including the artificial introduction of brine, and a thermal loading approximately three to four times higher than reference. Eight of these 12 test packages contain 1500 W/canister electrical heaters; the other four are filled with DHLW glass. 9 refs., 1 fig

Safety evaluation for packaging (onsite) concrete-lined waste packaging

Energy Technology Data Exchange (ETDEWEB)

Romano, T.

1997-09-25

The Pacific Northwest National Laboratory developed a package to ship Type A, non-transuranic, fissile excepted quantities of liquid or solid radioactive material and radioactive mixed waste to the Central Waste Complex for storage on the Hanford Site.

Slovak Nuclear Regulatory Body Position in the Transport of Radioactive Waste

International Nuclear Information System (INIS)

Homola, J.

2003-01-01

This paper describes safety requirements for transport of radioactive waste in Slovakia and the role of regulatory body in the transport licensing and assessment processes. Importance of radioactive waste shipments have been increased since 1999 by starting of NPP A-1 decommissioning and operation of near surface disposal facility. Also some information from history of shipment as well as future activities are given. Legal basis for radioactive waste transport is resulting from IAEA recommendations in this area. Different types of transport equipment were approved by regulatory body for both liquid and solid waste and transportation permits were issued to their shipment. Regulatory body attention during evaluation of transport safety is focused mainly on ability of individual packages to withstand different transport conditions and on safety analyses performed for transport equipment for liquid waste with high frequency of shipments. During past three years no event was occurred in connection with radioactive waste transport in Slovakia

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

DHLW Glass Waste Package Criticality Analysis (SCPB:N/A)

International Nuclear Information System (INIS)

Davis, J.W.

1996-01-01

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to determine the viability of the Defense High-Level Waste (DHLW) Glass waste package concept with respect to criticality regulatory requirements in compliance with the goals of the Waste Package Implementation Plan (Ref. 5.1) for conceptual design. These design calculations are performed in sufficient detail to provide a comprehensive comparison base with other design alternatives. The objective of this evaluation is to show to what extent the concept meets the regulatory requirements or indicate additional measures that are required for the intact waste package

A PC-based software package for modeling DOE mixed-waste management options

International Nuclear Information System (INIS)

Abashian, M.S.; Carney, C.; Schum, K.

1995-02-01

The U.S. Department of Energy (DOE) Headquarters and associated contractors have developed an IBM PC-based software package that estimates costs, schedules, and public and occupational health risks for a range of mixed-waste management options. A key application of the software package is the comparison of various waste-treatment options documented in the draft Site Treatment Plans prepared in accordance with the requirements of the Federal Facility Compliance Act of 1992. This automated Systems Analysis Methodology consists of a user interface for configuring complexwide or site-specific waste-management options; calculational algorithms for cost, schedule and risk; and user-selected graphical or tabular output of results. The mixed-waste management activities modeled in the automated Systems Analysis Methodology include waste storage, characterization, handling, transportation, treatment, and disposal. Analyses of treatment options identified in the draft Site Treatment Plans suggest potential cost and schedule savings from consolidation of proposed treatment facilities. This paper presents an overview of the automated Systems Analysis Methodology

Safety Evaluation of Radioactive Material Transport Package under Stacking Test Condition

International Nuclear Information System (INIS)

Lee, Ju Chan; Seo, Ki Seog; Yoo, Seong Yeon

2012-01-01

Radioactive waste transport package was developed to transport eight drums of low and intermediate level waste(LILW) in accordance with the IAEA and domestic related regulations. The package is classified with industrial package IP-2. IP-2 package is required to undergo a free drop test and a stacking test. After free drop and stacking tests, it should prevent the loss or dispersal of radioactive contents, and loss of shielding integrity which would result in more than 20 % increase in the radiation level at any external surface of the package. The objective of this study is to establish the safety test method and procedure for stacking test and to prove the structural integrities of the IP-2 package. Stacking test and analysis were performed with a compressive load equal to five times the weight of the package for a period of 24 hours using a full scale model. Strains and displacements were measured at the corner fitting of the package during the stacking test. The measured strains and displacements were compared with the analysis results, and there were good agreements. It is very difficult to measure the deflection at the container base, so the maximum deflection of the container base was calculated by the analysis method. The maximum displacement at the corner fitting and deflection at the container base were less than their allowable values. Dimensions of the test model, thickness of shielding material and bolt torque were measured before and after the stacking test. Throughout the stacking test, it was found that there were no loss or dispersal of radioactive contents and no loss of shielding integrity. Thus, the package was shown to comply with the requirements to maintain structural integrity under the stacking condition.

Waste Package Design Methodology Report

Energy Technology Data Exchange (ETDEWEB)

D.A. Brownson

2001-09-28

The objective of this report is to describe the analytical methods and processes used by the Waste Package Design Section to establish the integrity of the various waste package designs, the emplacement pallet, and the drip shield. The scope of this report shall be the methodology used in criticality, risk-informed, shielding, source term, structural, and thermal analyses. The basic features and appropriateness of the methods are illustrated, and the processes are defined whereby input values and assumptions flow through the application of those methods to obtain designs that ensure defense-in-depth as well as satisfy requirements on system performance. Such requirements include those imposed by federal regulation, from both the U.S. Department of Energy (DOE) and U.S. Nuclear Regulatory Commission (NRC), and those imposed by the Yucca Mountain Project to meet repository performance goals. The report is to be used, in part, to describe the waste package design methods and techniques to be used for producing input to the License Application Report.

Waste Package Design Methodology Report

International Nuclear Information System (INIS)

D.A. Brownson

2001-01-01

The objective of this report is to describe the analytical methods and processes used by the Waste Package Design Section to establish the integrity of the various waste package designs, the emplacement pallet, and the drip shield. The scope of this report shall be the methodology used in criticality, risk-informed, shielding, source term, structural, and thermal analyses. The basic features and appropriateness of the methods are illustrated, and the processes are defined whereby input values and assumptions flow through the application of those methods to obtain designs that ensure defense-in-depth as well as satisfy requirements on system performance. Such requirements include those imposed by federal regulation, from both the U.S. Department of Energy (DOE) and U.S. Nuclear Regulatory Commission (NRC), and those imposed by the Yucca Mountain Project to meet repository performance goals. The report is to be used, in part, to describe the waste package design methods and techniques to be used for producing input to the License Application Report

Greater-Than-Class C Low-Level Radioactive Waste Transportation Strategy report and institutional plan

Energy Technology Data Exchange (ETDEWEB)

Schmitt, R.C.; Tyacke, M.J.

1995-01-01

This document contains two parts. Part I, Greater-Than-Class-C Low-Level Radioactive Waste Transportation Strategy, addresses the requirements, responsibilities, and strategy to transport and receive these wastes. The strategy covers (a) transportation packaging, which includes shipping casks and waste containers; (b) transportation operations relating to the five facilities involved in transportation, i.e., waste originator, interim storage, dedicated storage, treatment, and disposal; (c) system safety and risk analysis; (d) routes; (e) emergency preparedness and response; and (o safeguards and security. A summary of strategic actions is provided at the conclusion of Part 1. Part II, Institutional Plan for Greater-Than-Class C Low-Level Radioactive Waste Packaging and Transportation, addresses the assumptions, requirements, and institutional plan elements and actions. As documented in the Strategy and Institutional Plan, the most challenging issues facing the GTCC LLW Program shipping campaign are institutional issues closely related to the strategy. How the Program addresses those issues and demonstrates to the states, local governments, and private citizens that the shipments can and will be made safely will strongly affect the success or failure of the campaign.

Greater-Than-Class C Low-Level Radioactive Waste Transportation Strategy report and institutional plan

International Nuclear Information System (INIS)

Schmitt, R.C.; Tyacke, M.J.

1995-01-01

This document contains two parts. Part I, Greater-Than-Class-C Low-Level Radioactive Waste Transportation Strategy, addresses the requirements, responsibilities, and strategy to transport and receive these wastes. The strategy covers (a) transportation packaging, which includes shipping casks and waste containers; (b) transportation operations relating to the five facilities involved in transportation, i.e., waste originator, interim storage, dedicated storage, treatment, and disposal; (c) system safety and risk analysis; (d) routes; (e) emergency preparedness and response; and (o safeguards and security. A summary of strategic actions is provided at the conclusion of Part 1. Part II, Institutional Plan for Greater-Than-Class C Low-Level Radioactive Waste Packaging and Transportation, addresses the assumptions, requirements, and institutional plan elements and actions. As documented in the Strategy and Institutional Plan, the most challenging issues facing the GTCC LLW Program shipping campaign are institutional issues closely related to the strategy. How the Program addresses those issues and demonstrates to the states, local governments, and private citizens that the shipments can and will be made safely will strongly affect the success or failure of the campaign

Shielding Calculations on Waste Packages - The Limits and Possibilities of different Calculation Methods by the example of homogeneous and inhomogeneous Waste Packages

Science.gov (United States)

Adams, Mike; Smalian, Silva

2017-09-01

For nuclear waste packages the expected dose rates and nuclide inventory are beforehand calculated. Depending on the package of the nuclear waste deterministic programs like MicroShield® provide a range of results for each type of packaging. Stochastic programs like "Monte-Carlo N-Particle Transport Code System" (MCNP®) on the other hand provide reliable results for complex geometries. However this type of program requires a fully trained operator and calculations are time consuming. The problem here is to choose an appropriate program for a specific geometry. Therefore we compared the results of deterministic programs like MicroShield® and stochastic programs like MCNP®. These comparisons enable us to make a statement about the applicability of the various programs for chosen types of containers. As a conclusion we found that for thin-walled geometries deterministic programs like MicroShield® are well suited to calculate the dose rate. For cylindrical containers with inner shielding however, deterministic programs hit their limits. Furthermore we investigate the effect of an inhomogeneous material and activity distribution on the results. The calculations are still ongoing. Results will be presented in the final abstract.

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

Packaging and transportation of radioactively contaminated lead

International Nuclear Information System (INIS)

Gleason, Eugene; Holden, Gerard

2007-01-01

Under the management of the Nuclear Decommissioning Authority (NDA) the government of the United Kingdom has launched an ambitious program to remediate the nation's nuclear waste legacy. Over a twenty-five year period NDA plans to decommission several first generation nuclear power plants and other radioactive facilities. The use innovative, safe 'fit for purpose' technologies will be a major part of this complex program. This paper will present a case study of a recently completed project undertaken in support of the nuclear decommissioning activities at the Sellafield site in the United Kingdom. The focus is on an innovative application of new packaging technology developed for the safe transportation of radioactively contaminated lead objects. Several companies collaborated on the project and contributed to its safe and successful conclusion. These companies include British Nuclear Group, Gravatom Engineering, W. F. Bowker Transport, Atlantic Container Lines, MHF Logistical Solutions and Energy Solutions. New containers and a new innovative inter-modal packaging system to transport the radioactive lead were developed and demonstrated during the project. The project also demonstrated the potential contribution of international nuclear recycling activities as a safe, economic and feasible technical option for nuclear decommissioning in the United Kingdom. (authors)

Generalized waste package containment model

International Nuclear Information System (INIS)

Liebetrau, A.M.; Apted, M.J.

1985-02-01

The US Department of Energy (DOE) is developing a performance assessment strategy to demonstrate compliance with standards and technical requirements of the Environmental Protection Agency (EPA) and the Nuclear Regulatory Commission (NRC) for the permanent disposal of high-level nuclear wastes in geologic repositories. One aspect of this strategy is the development of a unified performance model of the entire geologic repository system. Details of a generalized waste package containment (WPC) model and its relationship with other components of an overall repository model are presented in this paper. The WPC model provides stochastically determined estimates of the distributions of times-to-failure of the barriers of a waste package by various corrosion mechanisms and degradation processes. The model consists of a series of modules which employ various combinations of stochastic (probabilistic) and mechanistic process models, and which are individually designed to reflect the current state of knowledge. The WPC model is designed not only to take account of various site-specific conditions and processes, but also to deal with a wide range of site, repository, and waste package configurations. 11 refs., 3 figs., 2 tabs

Waste Form Release Data Package for the 2001 Immobilized Low-Activity Waste Performance Assessment

Energy Technology Data Exchange (ETDEWEB)

McGrail, B. Peter; Icenhower, Jonathan P.; Martin, Paul F.; Schaef, Herbert T.; O' Hara, Matthew J.; Rodriguez, Eugenio; Steele, Jackie L.

2001-02-01

This data package documents the experimentally derived input data on the representative waste glasses LAWABP1 and HLP-31 that will be used for simulations of the immobilized lowactivity waste disposal system with the Subsurface Transport Over Reactive Multiphases (STORM) code. The STORM code will be used to provide the near-field radionuclide release source term for a performance assessment to be issued in March of 2001. Documented in this data package are data related to 1) kinetic rate law parameters for glass dissolution, 2) alkali-H ion exchange rate, 3) chemical reaction network of secondary phases that form in accelerated weathering tests, and 4) thermodynamic equilibrium constants assigned to these secondary phases. The kinetic rate law and Na+-H+ ion exchange rate were determined from single-pass flow-through experiments. Pressurized unsaturated flow and vapor hydration experiments were used for accelerated weathering or aging of the glasses. The majority of the thermodynamic data were extracted from the thermodynamic database package shipped with the geochemical code EQ3/6. However, several secondary reaction products identified from laboratory tests with prototypical LAW glasses were not included in this database, nor are the thermodynamic data available in the open literature. One of these phases, herschelite, was determined to have a potentially significant impact on the release calculations and so a solubility product was estimated using a polymer structure model developed for zeolites. Although this data package is relatively complete, final selection of ILAW glass compositions has not been done by the waste treatment plant contractor. Consequently, revisions to this data package to address new ILAW glass formulations are to be regularly expected.

Second Generation Waste Package Design Study

International Nuclear Information System (INIS)

Armijo, J.S.; Misra, M.; Kar, Piyush

2007-01-01

The following describes the objectives of Project Activity 023 ''Second Generation Waste Package Design Study'' under DOE Cooperative Agreement DE-FC28-04RW12232. The objectives of this activity are: to review the current YMP baseline environment and establish corrosion test environments representative of the range of dry to intermittently wet conditions expected in the drifts as a function of time; to demonstrate the oxidation and corrosion resistance of A588 weathering steel and reference Alloy 22 samples in the representative dry to intermittently dry conditions; and to evaluate backfill and design features to improve the thermal performance analyses of the proposed second-generation waste packages using existing models developed at the University of Nevada, Reno(UNR). The work plan for this project activity consists of three major tasks: Task 1. Definition of expected worst-case environments (humidity, liquid composition and temperature) at waste package outer surfaces as a function of time, and comparison with environments defined in the YMP baseline; Task 2. Oxidation and corrosion tests of proposed second-generation outer container material; and Task 3. Second Generation waste package thermal analyses. Full funding was not provided for this project activity

10 CFR 60.143 - Monitoring and testing waste packages.

Science.gov (United States)

2010-01-01

....143 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN... repository operations area, the environment of the waste packages selected for the waste package monitoring program shall be representative of the environment in which the wastes are to be emplaced. (c) The waste...

Inspection and verification of waste packages for near surface disposal

International Nuclear Information System (INIS)

2000-01-01

Extensive experience has been gained with various disposal options for low and intermediate level waste at or near surface disposal facilities. Near surface disposal is based on proven and well demonstrated technologies. To ensure the safety of near surface disposal facilities when available technologies are applied, it is necessary to control and assure the quality of the repository system's performance, which includes waste packages, engineered features and natural barriers, as well as siting, design, construction, operation, closure and institutional controls. Recognizing the importance of repository performance, the IAEA is producing a set of technical publications on quality assurance and quality control (QA/QC) for waste disposal to provide Member States with technical guidance and current information. These publications cover issues on the application of QA/QC programmes to waste disposal, long term record management, and specific QA/QC aspects of waste packaging, repository design and R and D. Waste package QA/QC is especially important because the package is the primary barrier to radionuclide release from a disposal facility. Waste packaging also involves interface issues between the waste generator and the disposal facility operator. Waste should be packaged by generators to meet waste acceptance requirements set for a repository or disposal system. However, it is essential that the disposal facility operator ensure that waste packages conform with disposal facility acceptance requirements. Demonstration of conformance with disposal facility acceptance requirements can be achieved through the systematic inspection and verification of waste packages at both the waste generator's site and at the disposal facility, based on a waste package QA/QC programme established by the waste generator and approved by the disposal operator. However, strategies, approaches and the scope of inspection and verification will be somewhat different from country to country

Conceptual waste packaging options for deep borehole disposal

Energy Technology Data Exchange (ETDEWEB)

Su, Jiann -Cherng [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Hardin, Ernest L. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

2015-07-01

This report presents four concepts for packaging of radioactive waste for disposal in deep boreholes. Two of these are reference-size packages (11 inch outer diameter) and two are smaller (5 inch) for disposal of Cs/Sr capsules. All four have an assumed length of approximately 18.5 feet, which allows the internal length of the waste volume to be 16.4 feet. However, package length and volume can be scaled by changing the length of the middle, tubular section. The materials proposed for use are low-alloy steels, commonly used in the oil-and-gas industry. Threaded connections between packages, and internal threads used to seal the waste cavity, are common oilfield types. Two types of fill ports are proposed: flask-type and internal-flush. All four package design concepts would withstand hydrostatic pressure of 9,600 psi, with factor safety 2.0. The combined loading condition includes axial tension and compression from the weight of a string or stack of packages in the disposal borehole, either during lower and emplacement of a string, or after stacking of multiple packages emplaced singly. Combined loading also includes bending that may occur during emplacement, particularly for a string of packages threaded together. Flask-type packages would be fabricated and heat-treated, if necessary, before loading waste. The fill port would be narrower than the waste cavity inner diameter, so the flask type is suitable for directly loading bulk granular waste, or loading slim waste canisters (e.g., containing Cs/Sr capsules) that fit through the port. The fill port would be sealed with a tapered, threaded plug, with a welded cover plate (welded after loading). Threaded connections between packages and between packages and a drill string, would be standard drill pipe threads. The internal flush packaging concepts would use semi-flush oilfield tubing, which is internally flush but has a slight external upset at the joints. This type of tubing can be obtained with premium, low

NWTS waste package program plan. Volume II. Program logic networks

International Nuclear Information System (INIS)

1981-10-01

This document describes the work planned for developing the technology to design, test and produce packages used for the long-term isolation of nuclear waste in deep geologic repositories. Waste forms considered include spent fuel and high-level waste. The testing and selection effort for barrier materials for radionuclide containment is described. The NWTS waste package program is a design-driven effort; waste package conceptual designs are used as input for preliminary designs, which are upgraded to a final design as materials and testing data become available. Performance assessment models are developed and validated. Milestones and a detailed schedule are given for the waste package development effort. Program logic networks defining work flow, interfaces among the NWTS Projects, and interrelationships of specific activities are presented. Detailed work elements are provided for the Waste Package Program Plan subtasks - design and development, waste form, barrier materials, and performance evaluation - for salt and basalt, host rocks for which the state of waste package knowledge and the corresponding data base are advanced

TRU waste transport economics: an overview

International Nuclear Information System (INIS)

Edling, D.A.; Hopkins, D.R.; Walls, H.C.

1978-01-01

There are currently three predominant methods used to transport transuranium contaminated waste. These are: (1) ATMX Railcars--500 and 600 series, (2) Super Tigers, and (3) Poly Panthers. Both the ATMX-500 and 600 series railcars are massive doubly walled steel railcars which provide the equivalent protection of a Type B package. In ATMX-600 the rapid loading and unloading of the 9 x 9 x 50 feet cargo space is achieved by prepackaging the TRU waste into standard 20-foot steel cargo containers. The ATMX-500 railcars are divided into three inside bays, having dimensions of 16 (l) x 9.25 (w) x 6.25 (h) feet. A typical load consists of 128 55-gallon drums (however, space can accommodate 192 drums), 12 fiberglass boxes (4 x 4 x 7), or a combination of palletized drums and boxes. A Super Tiger is an overpack authorized for Type A, Type B, and large quantities of radioactive materials having outside dimensions of 8 x 8 x 20 feet. Maximum payload is approximately 28,700 lb with a gross weight of 45,000 lb. The primary factors influencing transport costs are examined including freight rates of transport mode, effective cargo (weight and volume) management, effective utilization of available space (package design), transport mileage, and rental fees or initial capital outlay. Miscellaneous factors are also examined

Shielding Calculations on Waste Packages – The Limits and Possibilities of different Calculation Methods by the example of homogeneous and inhomogeneous Waste Packages

Directory of Open Access Journals (Sweden)

Adams Mike

2017-01-01

Full Text Available For nuclear waste packages the expected dose rates and nuclide inventory are beforehand calculated. Depending on the package of the nuclear waste deterministic programs like MicroShield® provide a range of results for each type of packaging. Stochastic programs like “Monte-Carlo N-Particle Transport Code System” (MCNP® on the other hand provide reliable results for complex geometries. However this type of program requires a fully trained operator and calculations are time consuming. The problem here is to choose an appropriate program for a specific geometry. Therefore we compared the results of deterministic programs like MicroShield® and stochastic programs like MCNP®. These comparisons enable us to make a statement about the applicability of the various programs for chosen types of containers. As a conclusion we found that for thin-walled geometries deterministic programs like MicroShield® are well suited to calculate the dose rate. For cylindrical containers with inner shielding however, deterministic programs hit their limits. Furthermore we investigate the effect of an inhomogeneous material and activity distribution on the results. The calculations are still ongoing. Results will be presented in the final abstract.

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

Radioactive waste disposal package

Science.gov (United States)

Lampe, Robert F.

1986-11-04

A radioactive waste disposal package comprising a canister for containing vitrified radioactive waste material and a sealed outer shell encapsulating the canister. A solid block of filler material is supported in said shell and convertible into a liquid state for flow into the space between the canister and outer shell and subsequently hardened to form a solid, impervious layer occupying such space.

10 CFR 63.134 - Monitoring and testing waste packages.

Science.gov (United States)

2010-01-01

....134 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN A... geologic repository operations area, the environment of the waste packages selected for the waste package monitoring program must be representative of the environment in which the wastes are to be emplaced. (c) The...

Initial waste package interaction tests: status report

International Nuclear Information System (INIS)

Shade, J.W.; Bradley, D.J.

1980-12-01

This report describes the results of some initial investigations of the effects of rock media on the release of simulated fission products from a sngle waste form, PNL reference glass 76-68. All tests assemblies contained a minicanister prepared by pouring molten, U-doped 76-68 glass into a 2-cm-dia stanless steel tube closed at one end. The tubes were cut to 2.5 to 7.5 cm in length to expose a flat glass surface rimmed by the canister wall. A cylindrical, whole rock pellet, cut from one of the rock materials used, was placed on the glass surface then both the canister and rock pellet were packed in the same type of rock media ground to about 75 μm to complete the package. Rock materials used were a quartz monzonite basalt and bedded salt. These packages were run from 4 to 6 weeks in either 125 ml digestion bombs or 850 ml autoclaves capable of direct solution sampling, at either 250 or 150 0 C. Digestion bomb pressures were the vapor pressure of water, 600 psig at 250 0 C, and the autoclaves were pressurized at 2000 psig with an argon overpressure. In general, the solution chemistry of these initial package tests suggests that the rock media is the dominant controlling factor and that rock-water interaction may be similar to that observed in some geothermal areas. In no case was uranium observed in solution above 15 ppB. The observed leach rates of U glass not in contact with potential sinks (rock surfaces and alteration products) have been observed to be considerably higher. Thus the use of leach rates and U concentrations observed from binary leach experiments (waste-form water only) to ascertain long-term environmental consequences appear to be quite conservative compared to actual U release in the waste package experiments. Further evaluation, however, of fission product transport behavior and the role of alteration phases as fission product sinks is required

The approach of risk and safety evaluation in radioactive waste transport

International Nuclear Information System (INIS)

Vieru, G.

1996-01-01

Within Institute for Nuclear Research (INR) Pitesti, qualification tests were performed on packages, designed for transport and storage of low activity radioactive waste. Risk assessment activities aiming the evaluation of risk categories that many arise either during accident free transport or during accident conditions of waste transportation to the disposal center, in Romania, have been approached. The accident rates calculation, the distribution within accident scenarios and overall effective collective dose (man.Sv/year),for routine road transportation and the accidental Risk (man.Sv/Year) were determined

Low-level radioactive waste transportation safety history

International Nuclear Information System (INIS)

McClure, J.D.

1997-01-01

The Radioactive Materials Incident Report (RMIR) database was developed fin 1981 at the Transportation Technology Center of Sandia National Laboratories to support its research and development activities for the US department of Energy (DOE). This database contains information about radioactive material (RAM) transportation incidents that have occurred in the US since 1971. These data were drawn from the US Department of Transportation's (DOT) Hazardous Materials Incident Report system, from Nuclear Regulatory Commission (NRC) files, and from various agencies including state radiological control offices. Support for the RMIR data base is funded by the US DOE National Transportation Program (NTP). Transportation events in RMIR are classified in one of the following ways: as a transportation accident, as a handling accident, or as a reported incident. This presentation will provide definitions for these classifications and give examples of each. The primary objective of this presentation is to provide information on nuclear materials transportation accident/incident events involving low-level waste (LLW) that have occurred in the US for the period 1971 through 1996. Among the areas to be examined are: transportation accidents by mode, package response during accidents, and an examination of accidents where release of contents has occurred. Where information is available, accident and incident history and package response for LLW packages in transportation accidents will be described

Developments in the transport of radioactive waste. Proceedings of a seminar held in Vienna, 21-25 February 1994

International Nuclear Information System (INIS)

1995-06-01

This document is proceeding of the seminar on development in the transport of low and intermediate level radioactive wastes. The papers presented in this seminar covered the following areas: IAEA activities in radioactive waste transport; Member State experience and recommendations for international transport regulations; Member State experience with national transport regulations; waste transport and handling; waste generation volumes, characteristics, and disposal requirements; risk assessment; and transport and waste packages. Refs, figs and tabs

Developments in the transport of radioactive waste. Proceedings of a seminar held in Vienna, 21-25 February 1994

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-06-01

This document is proceeding of the seminar on development in the transport of low and intermediate level radioactive wastes. The papers presented in this seminar covered the following areas: IAEA activities in radioactive waste transport; Member State experience and recommendations for international transport regulations; Member State experience with national transport regulations; waste transport and handling; waste generation volumes, characteristics, and disposal requirements; risk assessment; and transport and waste packages. Refs, figs and tabs.

Methods for maintaining a record of waste packages during waste processing and storage

International Nuclear Information System (INIS)

2005-01-01

During processing, radioactive waste is converted into waste packages, and then sent for storage and ultimately for disposal. A principal condition for acceptance of a waste package is its full compliance with waste acceptance criteria for disposal or storage. These criteria define the radiological, mechanical, physical, chemical and biological properties of radioactive waste that can, in principle, be changed during waste processing. To declare compliance of a waste package with waste acceptance criteria, a system for generating and maintaining records should be established to record and track all relevant information, from raw waste characteristics, through changes related to waste processing, to final checking and verification of waste package parameters. In parallel, records on processing technology and the operational parameters of technological facilities should adhere to established and approved quality assurance systems. A records system for waste management should be in place, defining the data to be collected and stored at each step of waste processing and using a reliable selection process carried over into the individual steps of the waste processing flow stream. The waste management records system must at the same time ensure selection and maintenance of all the main information, not only providing evidence of compliance of waste package parameters with waste acceptance criteria but also serving as an information source in the case of any future operations involving the stored or disposed waste. Records generated during waste processing are a constituent part of the more complex system of waste management record keeping, covering the entire life cycle of radioactive waste from generation to disposal and even the post-closure period of a disposal facility. The IAEA is systematically working on the preparation of a set of publications to assist its Member States in the development and implementation of such a system. This report covers all the principal

Safety analysis of the transportation of radioactive waste to the Konrad final repository

International Nuclear Information System (INIS)

Sentuc, F.N.; Bruecher, W.

2010-01-01

A transport risk assessment study has been conducted for transport of radioactive waste with negligible heat-generation to the German final repository Konrad. This study is a revision of the former Konrad Transport Study performed by GRS in 1991 implementing updated waste data among other improved methods and assumptions for the purpose of a more realistic approach to risk assessment. The first part of the transport risk assessment study concerns the radiological consequences from normal (accident-free) transportation of radioactive material, i.e. the radiation exposure of transport personnel and the public. Based on the assessed detailed information on transport arrangements and on the average number and radiological characteristics of waste packages the maximum annual effective doses for the representative persons were estimated. The risk associated with transport incidents and accidents has been quantified for the area within a radius of 25 km around the repository site. The probabilistic method adopted in this study considers parameters as the frequency and severity of railway or road accidents, characteristics of radioactive waste and transport packagings and the frequency of atmospheric dispersion conditions. From a large set of parameter combinations the spectrum of potential radiological consequences and of the associated probability of occurrence was assessed. (orig.)

Safety evaluation report for packaging (onsite) concrete-lined waste packaging

International Nuclear Information System (INIS)

Romano, T.

1997-01-01

The Pacific Northwest National Laboratory developed a package to ship Type A, non-transuranic, fissile excepted quantities of liquid or solid radioactive material and radioactive mixed waste to the Central Waste Complex for storage on the Hanford Site

Update of Nuclear Waste Policy Act transportation activities

International Nuclear Information System (INIS)

Callaghan, E.F.

1987-01-01

As directed by the Nuclear Waste Policy Act of 1982 (NWPA), the Department of Energy (DOE) is developing a nationwide system for transporting spent nuclear fuel and high-level radioactive waste from commercial power plants to deep geologic repositories for disposal. Plans for the transportation system will consider the following factors: the President's 1985 decision to co-locate some defense high-level waste with commercial waste in a repository, the NWPA requirement that the private sector be used to the fullest extent possible in developing and operating the system, and the possible approval by Congress of the DOE's proposal for a Monitored Retrievable Storage (MRS) facility, submitted in March 1987. (The MRS, if approved, would provide for the consolidation, packaging, and perhaps the temporary storage of spent fuel from reactors.) The ''Transportation Business Plan'', published in January 1986, reflects these considerations. The transportation system, when operational, will consist of two elements: (1) the cask system, which includes the transportation casks, the vehicular conveyances, tie-downs, and associated equipment for handling the casks; and (2) the transportation support system which is comprised of facilities, equipment, and services to support waste transportation. Development of the transportation system incorporates the following work elements: operational planning, support systems development, cash system development, systems analysis, and institutional activities. This paper focusses on the technical aspects of the system

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

Influence of non-radioactive payload parameters on radioactive shipping packages

International Nuclear Information System (INIS)

Drez, P.E.; Murthy, D.V.S.; Temus, C.J.; Quinn, G.J.; Ozaki, C.

1989-01-01

The transport of radioactive waste materials in radioactive material (RAM) packages involves two components: the packaging used for transportation, and the waste which forms the payload. The payload is usually comprised of non-radioactive materials contaminated with radionuclides. The non-radionuclide payload characteristics can often be a controlling factor in determining the restrictions imposed on the certification of the package. This paper describes these package/payload interactions and the limiting parameters for the Transuranic Package Transporter-II (TRUPACT-II), designed for the transportation of Contact Handled Transuranic (CH-TRU) waste. The parameters discussed include the physical and chemical form of the payload, the configuration of the waste, and resulting gas generation and gas release phenomena. Brief descriptions of the TRUPACT-II package and its payload are presented initially

Symmetric Rock Fall on Waste Package

International Nuclear Information System (INIS)

Sreten Mastilovic

2001-01-01

The objective of this calculation is to determine the structural response of the Naval SNF (spent nuclear fuel) Waste Package (WP) and the emplacement pallet (EP) subjected to the rock fall DBE (design basis event) dynamic loads. The scope of this calculation is limited to reporting the calculation results in terms of stress intensities and residual stresses in the WP, and stress intensities and maximum permanent downward displacements of the EP-lifting surface. The information provided by the sketches (Attachment I) is that of the potential design of the type of WP and EP considered in this calculation, and all obtained results are valid for those designs only. This calculation is associated with the waste package design and is performed by the Waste Package Design Section in accordance with Reference 24. AP-3.124, ''Calculations'', is used to perform the calculation and develop the document

Impacts of transportation on a test and evaluation facility for nuclear waste disposal: a systems analysis

International Nuclear Information System (INIS)

Varadarajan, R.V.; Peterson, R.W.; Joy, D.S.; Gibson, S.M.

1983-01-01

An essential element of the Test and Evaluation Facility (TEF) is a waste packaging facility capable of producing a small number Test and Evaluation Facility of packages consisting of several different waste forms. The study envisions three scenarios for such a packaging facility: (1) modify an existing hot cell facility such as the Engine Maintenance Assembly and Disassembly (EMAD) facility at the Nevada Test Site so that it can serve as a packaging facility for the TEF. This scenario is referred to as the EMAD Option. (2) Build a new generic packaging facility (GPF) at the site of the TEF. In other words, colocate the GPF and the TEF. This scenario is referred to as the GPF Option, and (3) utilize the EMAD facility in conjunction with a colocated GPF (of minimal size and scope) at the TEF. This scenario is referred to as the Split Option. The results of the system study clearly bring out the fact that transportation has a significant impact on the selection and siting of the waste packaging facility. Preliminary conclusions, subject to the assumptions of the study, include the following: (1) regardless of the waste form, the GPF option is preferable to the other two in minimizing both transportation costs and logistical problems, (2) for any given scenario and choice of waste forms, there exists a candidate TEF location for which the transportation costs are at a minimum compared to the other locations, (3) in spite of the increased transportation costs and logistical complexity, the study shows that the overall system costs favor modification of an existing hot cell facility for the particular case considered

DESIGN ANALYSIS FOR THE NAVAL SNF WASTE PACKAGE

International Nuclear Information System (INIS)

T.L. Mitchell

2000-01-01

The purpose of this analysis is to demonstrate the design of the naval spent nuclear fuel (SNF) waste package (WP) using the Waste Package Department's (WPD) design methodologies and processes described in the ''Waste Package Design Methodology Report'' (CRWMS MandO [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000b). The calculations that support the design of the naval SNF WP will be discussed; however, only a sub-set of such analyses will be presented and shall be limited to those identified in the ''Waste Package Design Sensitivity Report'' (CRWMS MandO 2000c). The objective of this analysis is to describe the naval SNF WP design method and to show that the design of the naval SNF WP complies with the ''Naval Spent Nuclear Fuel Disposal Container System Description Document'' (CRWMS MandO 1999a) and Interface Control Document (ICD) criteria for Site Recommendation. Additional criteria for the design of the naval SNF WP have been outlined in Section 6.2 of the ''Waste Package Design Sensitivity Report'' (CRWMS MandO 2000c). The scope of this analysis is restricted to the design of the naval long WP containing one naval long SNF canister. This WP is representative of the WPs that will contain both naval short SNF and naval long SNF canisters. The following items are included in the scope of this analysis: (1) Providing a general description of the applicable design criteria; (2) Describing the design methodology to be used; (3) Presenting the design of the naval SNF waste package; and (4) Showing compliance with all applicable design criteria. The intended use of this analysis is to support Site Recommendation reports and assist in the development of WPD drawings. Activities described in this analysis were conducted in accordance with the technical product development plan (TPDP) ''Design Analysis for the Naval SNF Waste Package (CRWMS MandO 2000a)

Radcalc: A computer program to calculate the radiolytic production of hydrogen gas from radioactive wastes in packages

International Nuclear Information System (INIS)

Green, J.R.; Schwarz, R.A.; Hillesland, K.E.; Roetman, V.E.; Field, J.G.

1995-11-01

Radcalc for Windows' is a menu-driven Microsoft2 Windows-compatible computer code that calculates the radiolytic production of hydrogen gas in high- and low-level radioactive waste. In addition, the code also determines US Department of Transportation (DOT) transportation classifications, calculates the activities of parent and daughter isotopes for a specified period of time, calculates decay heat, and calculates pressure buildup from the production of hydrogen gas in a given package geometry. Radcalc for Windows was developed by Packaging Engineering, Transportation and Packaging, Westinghouse Hanford Company, Richland, Washington, for the US Department of Energy (DOE). It is available from Packaging Engineering and is issued with a user's manual and a technical manual. The code has been verified and validated

Review of DOE waste package program. Subtask 1.1 - National Waste Package Program, October 1983-March 1984. Volume 6

International Nuclear Information System (INIS)

Soo, P.

1985-03-01

The present effort is part of an ongoing task to review the national high-level waste package effort. It includes evaluation of reference waste form, container, and packing material components with respect to determining how they may contribute to the containment and controlled release of radionuclides after waste packages have been emplaced in salt, basalt, tuff, and granite repositories. In the current Biannual Report a review of progress in the new crystalline repository (granite) program is described. Other foreign data for this host rock have also been outlined where relevant. The use of crushed salt, and bentonite- and zeolite-containing packing materials is discussed. The effects of temperature and gamma irradiation are shown to be important with respect to defining the localized environmental conditions around a waste package and the long-term integrity of the packing

Study on retrievability of waste package in geological disposal

International Nuclear Information System (INIS)

Hasegawa, Hiroshi; Noda, Masaru

2002-02-01

Retrievability of waste packages in geological disposal of high-level radioactive waste has been investigated from a technical aspect in various foreign countries, reflecting a social concern while retrievability is not provided as a technical requirement. This study investigates the concept of reversibility and retrievability in foreign countries and a technical feasibility on retrievability of waste packages in the geological disposal concept shown in the H12 report. The conclusion obtained through this study is as follows: 1. Concept of reversibility and retrievability in foreign countries. Many organizations have reconsidered the retrievability as one option in the geological disposal to improve the reversibility of the stepwise decision-making process and provide the flexibility, even based upon the principle of the geological disposal that retrieval of waste from the repository is not intended. 2. Technical feasibility on the retrievability in disposal concept in the H12 report. It is confirmed to be able to remove the buffer and to retrieve the waste packages by currently available technologies even after the stages following emplacement of the buffer. It must be noted that a large effort and expense would be required for some activities such as the reconstruction of access route if the activities started after a stage of backfilling disposal tunnels. 3. Evaluation of feasibility on the retrievability and extraction of the issues. In the near future, it is necessary to study and confirm the practical workability and economical efficiency for the retrieving method of waste packages proposed in this study, the handling and processing method of removed buffer materials, and the retrieving method of waste packages in the case of degrading the integrity of waste packages or not emplacing the waste packages in the assumed attitude, etc. (author)

Waste package designs for disposal of high-level waste in salt formations

International Nuclear Information System (INIS)

Basham, S.J. Jr.; Carr, J.A.

1984-01-01

In the United States of America the selected method for disposal of radioactive waste is mined repositories located in suitable geohydrological settings. Currently four types of host rocks are under consideration: tuff, basalt, crystalline rock and salt. Development of waste package designs for incorporation in mined salt repositories is discussed. The three pertinent high-level waste forms are: spent fuel, as disassembled and close-packed fuel pins in a mild steel canister; commercial high-level waste (CHLW), as borosilicate glass in stainless-steel canisters; defence high-level waste (DHLW), as borosilicate glass in stainless-steel canisters. The canisters are production and handling items only. They have no planned long-term isolation function. Each waste form requires a different approach in package design. However, the general geometry and the materials of the three designs are identical. The selected waste package design is an overpack of low carbon steel with a welded closure. This container surrounds the waste forms. Studies to better define brine quantity and composition, radiation effects on the salt and brines, long-term corrosion behaviour of the low carbon steel, and the leaching behaviour of the spent fuel and borosilicate glass waste forms are continuing. (author)

Improved practices for packaging transuranic waste at Los Alamos National Laboratory (LA-UR-09-03293) - 16280

International Nuclear Information System (INIS)

Goyal, Kapil K.; Carson, Peter H.

2009-01-01

Transuranic (TRU) waste leaving the Plutonium Facility at Los Alamos National Laboratory (LANL) is packaged using LANL's waste acceptance criteria for onsite storage. Before shipment to the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico, each payload container is subject to rigorous characterization to ensure compliance with WIPP waste acceptance criteria and Department of Transportation regulations. Techniques used for waste characterization include nondestructive examination by WIPP-certified real-time radiography (RTR) and nondestructive assay (NDA) of containers, as well as headspace gas sampling to ensure that hydrogen and other flammable gases remain at safe levels during transport. These techniques are performed under a rigorous quality assurance program to confirm that results are accurate and reproducible. If containers are deemed problematic, corrective action is implemented before they are shipped to WIPP. A defensive approach was used for many years to minimize the number of problematic drums. However, based on review of data associated with headspace gas sampling, NDA and RTR results, and enhanced coordination with the entities responsible for waste certification, many changes have been implemented to facilitate packaging of TRU waste drums with higher isotopic loading at the Plutonium Facility at an unprecedented rate while ensuring compliance with waste acceptance criteria. This paper summarizes the details of technical changes and related administrative coordination activities, such as information sharing among the certification entities, generators, waste packagers, and shippers. It discusses the results of all such cumulative changes that have been implemented at the Plutonium Facility and gives readers a preview of what LANL has accomplished to expeditiously certify and dispose of newly generated TRU waste. (authors)

Status of standardization efforts for packaging and transportation of spent fuel and high-level waste

International Nuclear Information System (INIS)

Eggers, P.E.; Dawson, D.M.

1986-01-01

This paper provides a current review of the status of efforts to develop standards and guidelines related to the packaging and transportation of spent fuel and high-level waste. An overview of each of the organizations and agencies developing standards and guidelines is discussed and includes the efforts of the N14 Division of the American National Standards Institute (ANSI), NUPACK Committee of Section III of the American Society of Mechanical Engineers, Nuclear Regulatory Commission and Department of Energy. This comparative overview identifies the scope and areas of application of each standard and guideline. In addition, the current or proposed standards and guidelines are considered collectively with commentary on areas of apparent or potential complimentary fit, overlap and incompatability. Finally, the paper reviews initiatives now being taken within the N14 division of ANSI to identify where new standards development activities are required

Transportation package design using numerical optimization

International Nuclear Information System (INIS)

Harding, D.C.; Witkowski, W.R.

1991-01-01

The purpose of this overview is twofold: first, to outline the theory and basic elements of numerical optimization; and second, to show how numerical optimization can be applied to the transportation packaging industry and used to increase efficiency and safety of radioactive and hazardous material transportation packages. A more extensive review of numerical optimization and its applications to radioactive material transportation package design was performed previously by the authors (Witkowski and Harding 1992). A proof-of-concept Type B package design is also presented as a simplified example of potential improvements achievable using numerical optimization in the design process

Challenges in packaging waste management in the fast food industry

Energy Technology Data Exchange (ETDEWEB)

Aarnio, Teija [Digita Oy, P.O. Box 135, FI-00521 Helsinki (Finland); Haemaelaeinen, Anne [Department of Energy and Environmental Technology, Lappeenranta University of Technology, P.O. Box 20, FI-53851 Lappeenranta (Finland)

2008-02-15

The recovery of solid waste is required by waste legislation, and also by the public. In some industries, however, waste is mostly disposed of in landfills despite of its high recoverability. Practical experiences show that the fast food industry is one example of these industries. A majority of the solid waste generated in the fast food industry is packaging waste, which is highly recoverable. The main research problem of this study was to find out the means of promoting the recovery of packaging waste generated in the fast food industry. Additionally, the goal of this article was to widen academic understanding on packaging waste management in the fast food industry, as the subject has not gained large academic interest previously. The study showed that the theoretical recovery rate of packaging waste in the fast food industry is high, 93% of the total annual amount, while the actual recovery rate is only 29% of the total annual amount. The total recovery potential of packaging waste is 64% of the total annual amount. The achievable recovery potential, 33% of the total annual amount, could be recovered, but is not mainly because of non-working waste management practices. The theoretical recovery potential of 31% of the total annual amount of packaging waste cannot be recovered by the existing solid waste infrastructure because of the obscure status of commercial waste, the improper operation of producer organisations, and the municipal autonomy. The research indicated that it is possible to reach the achievable recovery potential in the existing solid waste infrastructure through new waste management practices, which are designed and operated according to waste producers' needs and demands. The theoretical recovery potential can be reached by increasing the consistency of the solid waste infrastructure through governmental action. (author)

Design aspects of plutonium air-transportable packages

International Nuclear Information System (INIS)

Allen, G.C.; Moya, J.L.; Pierce, J.D.; Attaway, S.W.

1989-01-01

Recent worldwide interest in transporting plutonium powders by air has created a need for expanding the packaging technology base as well as improving their understanding of how plutonium air transport (PAT) packagings perform during severe accident tests. Historically it has not been possible to establish design rules for individual package components because of the complex way parts interacted in forming a successful whole unit. Also, computer analyses were only considered valid for very limited portions of the design effort because of large deformations, localized tearing occurring in the package during accident testing, and extensive use of orthotropic materials. Consequently, iterative design and experimentation has historically been used to develop plutonium air-transportable packages. Full-scale prototypes have been tested since scaling of packages utilizing wood as an energy absorber and thermal insulator has not proven to be very successful. This is because the wood grain and dynamic performance of the wood during crush do not always scale. The high cost of full-scale testing of large packages has certainly hindered obtaining additional data and development new designs. The testing criteria for PAT packages, as described in the US Nuclear Regulatory Commission's Qualification Criteria to Certify a Package for Air Transport of Plutonium, NUREG-0360, 1978, are summarized. Computer modeling techniques have greatly improved over the last ten years, and there are some areas of opportunity for future applications to plutonium air-transportable package design problems. Having developed a better understanding of the performance of current packages, they have the opportunity to make major improvements in new packaging concepts. Each of these areas is explored in further depth to establish their impact on design practices for air-transportable packages

78 FR 75672 - New Jersey Regulations on Transportation of Regulated Medical Waste

Science.gov (United States)

2013-12-12

... placing it in a packaging as required by the HMR; 3. N.J.A.C. 7:26-3A.14 that the words ``Medical Waste... Environmental Protection (NJDEP) solid waste transporter registration number; and 3) either the words ``Medical... material does not include a waste concentrated stock culture of an infectious substance. Sharps containers...

Severities of transportation accidents involving large packages

Energy Technology Data Exchange (ETDEWEB)

Dennis, A.W.; Foley, J.T. Jr.; Hartman, W.F.; Larson, D.W.

1978-05-01

The study was undertaken to define in a quantitative nonjudgmental technical manner the abnormal environments to which a large package (total weight over 2 tons) would be subjected as the result of a transportation accident. Because of this package weight, air shipment was not considered as a normal transportation mode and was not included in the study. The abnormal transportation environments for shipment by motor carrier and train were determined and quantified. In all cases the package was assumed to be transported on an open flat-bed truck or an open flat-bed railcar. In an earlier study, SLA-74-0001, the small-package environments were investigated. A third transportation study, related to the abnormal environment involving waterways transportation, is now under way at Sandia Laboratories and should complete the description of abnormal transportation environments. Five abnormal environments were defined and investigated, i.e., fire, impact, crush, immersion, and puncture. The primary interest of the study was directed toward the type of large package used to transport radioactive materials; however, the findings are not limited to this type of package but can be applied to a much larger class of material shipping containers.

Severities of transportation accidents involving large packages

International Nuclear Information System (INIS)

Dennis, A.W.; Foley, J.T. Jr.; Hartman, W.F.; Larson, D.W.

1978-05-01

The study was undertaken to define in a quantitative nonjudgmental technical manner the abnormal environments to which a large package (total weight over 2 tons) would be subjected as the result of a transportation accident. Because of this package weight, air shipment was not considered as a normal transportation mode and was not included in the study. The abnormal transportation environments for shipment by motor carrier and train were determined and quantified. In all cases the package was assumed to be transported on an open flat-bed truck or an open flat-bed railcar. In an earlier study, SLA-74-0001, the small-package environments were investigated. A third transportation study, related to the abnormal environment involving waterways transportation, is now under way at Sandia Laboratories and should complete the description of abnormal transportation environments. Five abnormal environments were defined and investigated, i.e., fire, impact, crush, immersion, and puncture. The primary interest of the study was directed toward the type of large package used to transport radioactive materials; however, the findings are not limited to this type of package but can be applied to a much larger class of material shipping containers

Repository documentation rethought. A comprehensive approach from untreated waste to waste packages for final disposal

Energy Technology Data Exchange (ETDEWEB)

Anthofer, Anton Philipp; Schubert, Johannes [VPC GmbH, Dresden (Germany)

2017-11-15

The German Act on Reorganization of Responsibility for Nuclear Disposal (Entsorgungsuebergangsgesetz (EntsorgUebG)) adopted in June 2017 provides the energy utilities with the new option of transferring responsibility for their waste packages to the Federal Government. This is conditional on the waste packages being approved for delivery to the Konrad final repository. A comprehensive approach starts with the dismantling of nuclear facilities and extends from waste disposal and packaging planning to final repository documentation. Waste package quality control measures are planned and implemented as early as in the process qualification stage so that the production of waste packages that are suitable for final deposition can be ensured. Optimization of cask and loading configuration can save container and repository volume. Workflow planning also saves time, expenditure and exposure time for personnel at the facilities. VPC has evaluated this experience and developed it into a comprehensive approach.

Containers for packaging of solid and intermediate level radioactive wastes

International Nuclear Information System (INIS)

1993-01-01

Low and intermediate level radioactive wastes are generated at all stages in the nuclear fuel cycle and also from the medical, industrial and research applications of radiation. These wastes can potentially present risks to health and the environment if they are not managed adequately. Their effective management will require the wastes to be safely stored, transported and ultimately disposed of. The waste container, which may be defined as any vessel, drum or box, made from metals, concrete, polymers or composite materials, in which the waste form is placed for interim storage, for transport and/or for final disposal, is an integral part of the whole package for the management of low and intermediate level wastes. It has key roles to play in several stages of the waste management process, starting from the storage of raw wastes and ending with the disposal of conditioned wastes. This report provides an overview of the various roles that a container may play and the factors that are important in each of these roles. This report has two main objectives. The first is to review the main requirements for the design of waste containers. The second is to provide advice on the design, fabrication and handling of different types of containers used in the management of low and intermediate level radioactive solid wastes. Recommendations for design and testing are given, based on the extensive experience available worldwide in waste management. This report is not intended to have any regulatory status or objectives. 56 refs, 16 figs, 10 tabs

Sampling and transport of paraffin waste form from CWDS of nuclear power plant

International Nuclear Information System (INIS)

Lee, J. M.; Hwang, J. H.; Kim, C. R.; Park, J. W.

2000-01-01

Sampling and transport of paraffin waste form from concentrated waste drying system (CWDS) of domestic nuclear power plant were performed to collect the leaching characteristic data for the disposal of radioactive waste. Transport was performed according to the national regulations and the internal rules of the nuclear power plant. The sample of paraffin waste form was classified as L type package according to the regulation and radiation exposure of operator was measured in the range of 6 to 12 mrem that was less than the estimated amount

Overview of hydrothermal testing of waste-package barrier materials at the Basalt Waste Isolation Project

International Nuclear Information System (INIS)

1982-01-01

The current Waste Package Department (WPD) hydrothermal testing program for the Basalt Waste Isolation Project (BWIP) has followed a systematic approach for the testing of waste-barrier-basalt interactions based on sequential penetration of barriers by intruding groundwaters. Present test activities in the WPD program have focused on determining radionuclide solubility limits (or steady-state conditions) of simulated waste forms and the long-term stability of waste package barriers under site-specific hydrothermal conditions. The resulting data on solution compositions and solid alteration products have been used to evaluate waste form degradation under conditions specific to a nuclear waste repository located in basalt (NWRB). Isothermal, time-invariant compositional data on sampled solutions have been coupled with realistic hydrologic flow data for near-field and far-field modeling for the calculation of meaningful radionuclide release rates. Radionuclides that are not strongly sorbed or precipitated from solution and that, therefore, may require special attention to ensure their isolation within the waste package have been identified. Taken together, these hydrothermal test data have been used to establish design requirements for waste packages located in basalt

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

L. Transportation of fuel and wastes

International Nuclear Information System (INIS)

1976-01-01

The principles applied to the transport of nuclear fuels and wastes have been founded on the more general provisions governing the transport of radioactive materials. Safe shipment of radioactive materials has historically been sought by specifying required characteristics in the shipping packages and establishing minimum acceptable levels of package integrity. The reason for this is that in the course of transport by road, rail, sea, or air, consignments of radioactive material are in close proximity to members of the public, and in many cases they are loaded or unloaded by transport workers who have had no special training or experience in the handling of such substances. The procedures adopted to ensure transport safety have worked satisfactorily. Both in the USA and the UK, the industry and regulatory authorities have established outstanding safety records in shipping radioactive materials over a period of thirty years. It is claimed that there have been no injuries due to the radioactive nature of the shipments, nor has there been a release of nuclear materials serious enough to be a threat of death or injury. Admittedly, about 95% of the 800,000 shipments estimated in the USA each year involve small quantities for use in industry, medicine, agriculture and education. However the principals underlying the safe packaging of these and reactor fuels are the same, and there is little reason to doubt that a similar safety record can be maintained

A study on opening displacement of lid and decrease in shielding thickness of a type IP-2 transport package in drop events

International Nuclear Information System (INIS)

Kim, Dong Hak; Seo, Ki Seog; Kim, Jae Yong; Lee, Ju Chan; Yoon, Jeong Hyoun; Lee, Kyung Ho; Kim, Sung Hwan; Lee, Heung Young

2005-01-01

Radioactive waste generated from nuclear power plants shall be transported in accordance with designated regulations, which is to protect radiation workers and the public against potential radiation exposure caused by the transportations. Each transport package of radioactive waste is to be designed to have enough safety to fulfill with the regulations and technical standards in domestic and foreign regulations. In accordance with IAEA safety standard series TS-R-1 which is widely accepted by most of its member states, industrial package can be divided into IP-1, IP-2 and IP-3 along with other Type A and Type B packages, a conventional clarification. IP-2 package shall be designed to meet the designated requirements in addition to those for type IP-1 package. IP-2 package is subject to the free drop and stacking tests under normal conditions of transport as regulated in the regulation. In this paper, opening displacement of lid and body and decrease in shielding thickness of an IP-2 package are analytically evaluated, which is proposed for on-site transportation in domestic nuclear power plants. The results of the analysis is compared with design requirements of the package that loss or dispersal of the radioactive contents should be prevented and total loss of shielding effect from free drop shall be less than 20%

The basics in transportation of low-level radioactive waste

International Nuclear Information System (INIS)

Allred, W.E.

1998-06-01

This bulletin gives a basic understanding about issues and safety standards that are built into the transportation system for radioactive material and waste in the US. An excellent safety record has been established for the transport of commercial low-level radioactive waste, or for that matter, all radioactive materials. This excellent safety record is primarily because of people adhering to strict regulations governing the transportation of radioactive materials. This bulletin discusses the regulatory framework as well as the regulations that set the standards for packaging, hazard communications (communicating the potential hazard to workers and the public), training, inspections, routing, and emergency response. The excellent safety record is discussed in the last section of the bulletin

Non-Destructive Testing for Control of Radioactive Waste Package

Science.gov (United States)

Plumeri, S.; Carrel, F.

2015-10-01

Characterization and control of radioactive waste packages are important issues in the management of a radioactive waste repository. Therefore, Andra performs quality control inspection on radwaste package before disposal to ensure the compliance of the radwast characteristics with Andra waste disposal specifications and to check the consistency between Andra measurements results and producer declared properties. Objectives of this quality control are: assessment and improvement of producer radwaste packages quality mastery, guarantee of the radwaste disposal safety, maintain of the public confidence. To control radiological characteristics of radwaste package, non-destructive passive methods (gamma spectrometry and neutrons counting) are commonly used. These passive methods may not be sufficient, for instance to control the mass of fissile material contained inside radwaste package. This is particularly true for large concrete hull of heterogeneous radwaste containing several actinides mixed with fission products like 137Cs. Non-destructive active methods, like measurement of photofission delayed neutrons, allow to quantify the global mass of actinides and is a promising method to quantify mass of fissile material. Andra has performed different non-destructive measurements on concrete intermediate-level short lived nuclear waste (ILW-SL) package to control its nuclear material content. These tests have allowed Andra to have a first evaluation of the performance of photofission delayed neutron measurement and to identify development needed to have a reliable method, especially for fissile material mass control in intermediate-level long lived waste package.

Mechanical Assessment of the Waste Package Subject to Vibratory Motion

International Nuclear Information System (INIS)

M. Gross

2004-01-01

The purpose of this document is to provide an integrated overview of the calculation reports that define the response of the waste package and its internals to vibratory ground motion. The calculation reports for waste package response to vibratory ground motion are identified in Table 1-1. Three key calculation reports describe the potential for mechanical damage to the waste package, fuel assemblies, and cladding from a seismic event. Three supporting documents have also been published to investigate sensitivity of damage to various assumptions for the calculations. While these individual reports present information on a specific aspect of waste package and cladding response, they do not describe the interrelationship between the various calculations and the relationship of this information to the seismic scenario class for Total System Performance Assessment-License Application (TSPA-LA). This report is designed to fill this gap by providing an overview of the waste package structural response calculations

Mechanical Assessment of the Waste Package Subject to Vibratory Motion

Energy Technology Data Exchange (ETDEWEB)

M. Gross

2004-10-14

The purpose of this document is to provide an integrated overview of the calculation reports that define the response of the waste package and its internals to vibratory ground motion. The calculation reports for waste package response to vibratory ground motion are identified in Table 1-1. Three key calculation reports describe the potential for mechanical damage to the waste package, fuel assemblies, and cladding from a seismic event. Three supporting documents have also been published to investigate sensitivity of damage to various assumptions for the calculations. While these individual reports present information on a specific aspect of waste package and cladding response, they do not describe the interrelationship between the various calculations and the relationship of this information to the seismic scenario class for Total System Performance Assessment-License Application (TSPA-LA). This report is designed to fill this gap by providing an overview of the waste package structural response calculations.

Oxidation and waste-to-energy output of aluminium waste packaging during incineration: A laboratory study.

Science.gov (United States)

López, Félix A; Román, Carlos Pérez; García-Díaz, Irene; Alguacil, Francisco J

2015-09-01

This work reports the oxidation behaviour and waste-to-energy output of different semi-rigid and flexible aluminium packagings when incinerated at 850°C in an air atmosphere enriched with 6% oxygen, in the laboratory setting. The physical properties of the different packagings were determined, including their metallic aluminium contents. The ash contents of their combustion products were determined according to standard BS ISO 1171:2010. The net calorific value, the required energy, and the calorific gain associated with each packaging type were determined following standard BS EN 13431:2004. Packagings with an aluminium lamina thickness of >50μm did not fully oxidise. During incineration, the weight-for-weight waste-to-energy output of the packagings with thick aluminium lamina was lower than that of packagings with thin lamina. The calorific gain depended on the degree of oxidation of the metallic aluminium, but was greater than zero for all the packagings studied. Waste aluminium may therefore be said to act as an energy source in municipal solid waste incineration systems. Copyright © 2015 Elsevier Ltd. All rights reserved.

Computerized waste-accountability shipping and packaging system

International Nuclear Information System (INIS)

Jackson, J.A.; Baston, M. Jr.; DeVer, E.A.

1981-01-01

The Waste Accountability, Shipping and Packaging System (WASP) is a real-time computerized system designed and implemented by Mound Facility to meet the stringent packaging and reporting requirements of radioactive waste being shipped to burial sites. The system stores packaging data and inspection results for each unit and prepares all necessary documents at the time of shipment. Shipping data specific for each burial site are automatically prepared on magnetic tape for transmission to the computing center at that site. WASP has enabled Mound Facility to effectively meet the requirements of the burial sites, diminishing the possibility of being rejected from a site because of noncompliance

Nuclear waste management technical support in the development of nuclear waste form criteria for the NRC. Task 1. Waste package overview

International Nuclear Information System (INIS)

Dayal, R.; Lee, B.S.; Wilke, R.J.; Swyler, K.J.; Soo, P.; Ahn, T.M.; McIntyre, N.S.; Veakis, E.

1982-02-01

In this report the current state of waste package development for high level waste, transuranic waste, and spent fuel in the US and abroad has been assessed. Specifically, reviewed are recent and on-going research on various waste forms, container materials and backfills and tentatively identified those which are likely to perform most satisfactorily in the repository environment. Radiation effects on the waste package components have been reviewed and the magnitude of these effects has been identified. Areas requiring further research have been identified. The important variables affecting radionuclide release from the waste package have been described and an evaluation of regulatory criteria for high level waste and spent fuel is presented. Finally, for spent fuel, high level, and TRU waste, components which could be used to construct a waste package having potential to meet NRC performance requirements have been described and identified

Nuclear waste management technical support in the development of nuclear waste form criteria for the NRC. Task 1. Waste package overview

Energy Technology Data Exchange (ETDEWEB)

Dayal, R.; Lee, B.S.; Wilke, R.J.; Swyler, K.J.; Soo, P.; Ahn, T.M.; McIntyre, N.S.; Veakis, E.

1982-02-01

In this report the current state of waste package development for high level waste, transuranic waste, and spent fuel in the US and abroad has been assessed. Specifically, reviewed are recent and on-going research on various waste forms, container materials and backfills and tentatively identified those which are likely to perform most satisfactorily in the repository environment. Radiation effects on the waste package components have been reviewed and the magnitude of these effects has been identified. Areas requiring further research have been identified. The important variables affecting radionuclide release from the waste package have been described and an evaluation of regulatory criteria for high level waste and spent fuel is presented. Finally, for spent fuel, high level, and TRU waste, components which could be used to construct a waste package having potential to meet NRC performance requirements have been described and identified.

Methodologies for certification of transuranic waste packages

International Nuclear Information System (INIS)

Christensen, R.N.; Kok, K.D.

1980-10-01

The objective of this study was to postulate methodologies for certification that a waste package is acceptable for disposal in a licensed geologic repository. Within the context of this report, certification means the overall process which verifies that a waste package meets the criteria or specifications established for acceptance for disposal in a repository. The overall methodology for certification will include (1) certifying authorities, (2) tests and procedures, and (3) documentation and quality assurance programs. Each criterion will require a methodology that is specific to that criterion. In some cases, different waste forms will require a different methodology. The purpose of predicting certification methodologies is to provide additional information as to what changes, if any, are needed for the TRU waste in storage

Nuclear waste transportation package testing: A review of selected programs in the United States and abroad

International Nuclear Information System (INIS)

Snedeker, D.F.

1990-12-01

This report provides an overview of some recent nuclear waste transportation package development programs. This information is intended to aid the State of Nevada in its review of US Department of Energy (DOE) nuclear waste transportation programs. This report addresses cask testing programs in the United Kingdom and selected 1/4 and full scale testing in the US. Facilities that can provide cask testing services, both in the US and to a limited extent abroad, are identified. The costs for different type test programs are identified as a means to estimate costs for future test programs. Not addressed is the public impact such testing might have in providing an increased sense of safety or confidence. The British test program was apparently quite successful in demonstrating safety to the public at the time. There is no US test effort that is similar in scope for direct comparison. Also addressed are lessons learned from testing programs and areas that may merit possible future integrated examination. Areas that may require further examination are both technical and institutional. This report provides information which, when combined with other sources of information will enable the State of Nevada to assess the following areas: feasibility of full scale testing; costs of full scale tests; potential benefits of testing; limits that full scale testing impose; and disadvantages of emphasis on testing vs analytical solutions. This assessment will then allow the state to comment on DOE Office of Civilian Radioactive Waste Management (OCRWM) plans for the development and licensing of new shipping cask designs. These plans currently expect contractors to perform engineering testing for materials development, quarter scale model testing to validate analytical assessments and full scale prototype testing of operational features. DOE currently plans no full scale or extra-regulatory destructive testing to aid in cask licensing. 1 tab

NWTS waste package program plan. Volume I. Program strategy, description, and schedule

International Nuclear Information System (INIS)

1981-10-01

This document describes the work planned for developing the technology to design, test and produce packages used for the long-term isolation of nuclear waste in deep geologic repositories. Waste forms considered include spent fuel and high-level waste. The testing and selection effort for barrier materials for radionuclide containment is described. The NWTS waste package program is a design-driven effort; waste package conceptual designs are used as input for preliminary designs, which are upgraded to a final design as materials and testing data become available. Performance assessment models are developed and validated. Milestones and a detailed schedule are given for the waste package development effort. Program logic networks defining work flow, interfaces among the NWTS Projects, and interrelationships of specific activities are presented. Detailed work elements are provided for the Waste Package Program Plan subtasks - design and development, waste form, barrier materials, and performance evaluation - for salt and basalt, host rocks for which the state of waste package knowledge and the corresponding data base are advanced

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)

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

WAPDEG Analysis of Waste Package and Drip shield Degradation

International Nuclear Information System (INIS)

K. Mon

2004-01-01

As directed by ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]), an analysis of the degradation of the engineered barrier system (EBS) drip shields and waste packages at the Yucca Mountain repository is developed. The purpose of this activity is to provide the TSPA with inputs and methodologies used to evaluate waste package and drip shield degradation as a function of exposure time under exposure conditions anticipated in the repository. This analysis provides information useful to satisfy ''Yucca Mountain Review Plan, Final Report'' (NRC 2003 [DIRS 163274]) requirements. Several features, events, and processes (FEPs) are also discussed (Section 6.2, Table 15). The previous revision of this report was prepared as a model report in accordance with AP-SIII.10Q, Models. Due to changes in the role of this report since the site recommendation, it no longer contains model development. This revision is prepared as a scientific analysis in accordance with AP-SIII.9Q, ''Scientific Analyses'' and uses models previously validated in (1) ''Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material'' (BSC 2004 [DIRS 169985]); (2) ''General Corrosion and Localized Corrosion of Waste Package Outer Barrier'' (BSC 2004 [DIRS 169984]); and (3) ''General Corrosion and Localized Corrosion of Drip Shield'' (BSC 2004 [DIRS 169845]). The integrated waste package degradation (IWPD) analysis presented in this report treats several implementation-related issues, such as defining the number and size of patches per waste package that undergo stress corrosion cracking; recasting the weld flaw analysis in a form as implemented in the Closure Weld Defects (CWD) software; and, general corrosion rate manipulations (e.g., change of scale in Section 6.3.4). The weld flaw portion of this report takes input from an engineering calculation (BSC 2004

Integrity of radioactive waste packages at the Yucca mountain repository

International Nuclear Information System (INIS)

Sandquist, G.; Biaglow, A.; Huber, M.; Jagmin, C.

2004-01-01

Several of the important physical and chemical processes that impact the integrity of the radioactive waste packages planned for disposal at the proposed Repository at Yucca Mountain are examined. These processes are described by the aerodynamic, thermodynamic, and chemical interactions associated with the waste packages. The effects of chemical corrosion, mechanical erosion, temperature distributions throughout the repository environs, interactions of air, water, and solid particles, and radiological and biological influences are addressed. Materials will be exposed to at least 3 conditions threatening the integrity of the waste package: 1) accumulated dust and particles on the package surface and suspended in the air, 2) chemical reactions from deposits on the waste package infrastructure materials and tight contact areas, and crevices, and 3) environmental factors affecting chemical reactions such as moisture, pH, Eh, and radiolysis. All 3 of these conditions can combine and produce damaging impacts upon the thin protective layer on the alloy surface of the waste package. There are certain benefits from the low-temperature operating mode with ambient temperature below 85 Celsius degrees, but the materials could be subjected to a maximum temperature of 180 Celsius degrees which might introduce stress corrosion cracking and high temperature effects

In-Package Chemistry Abstraction

Energy Technology Data Exchange (ETDEWEB)

E. Thomas

2004-11-09

package has been breached but the drip shield remains intact, so all of the seepage flow is diverted from the waste package. The chemistry from the vapor influx case is used to determine the stability of colloids and the solubility of radionuclides available for transport by diffusion, and to determine the degradation rates for the waste forms. TSPA-LA uses the water influx case for the seismic scenario, where the waste package has been breached and the drip shield has been damaged such that seepage flow is actually directed into the waste package. The chemistry from the water influx case that is a function of the flow rate is used to determine the stability of colloids and the solubility of radionuclides available for transport by diffusion and advection, and to determine the degradation rates for the CSNF and HLW glass. TSPA-LA does not use this model for the igneous scenario. Outputs from the in-package chemistry model implemented inside TSPA-LA include pH, ionic strength, and total carbonate concentration. These inputs to TSPA-LA will be linked to the following principle factors: dissolution rates of the CSNF and HLWG, dissolved concentrations of radionuclides, and colloid generation.

In-Package Chemistry Abstraction

International Nuclear Information System (INIS)

E. Thomas

2004-01-01

breached but the drip shield remains intact, so all of the seepage flow is diverted from the waste package. The chemistry from the vapor influx case is used to determine the stability of colloids and the solubility of radionuclides available for transport by diffusion, and to determine the degradation rates for the waste forms. TSPA-LA uses the water influx case for the seismic scenario, where the waste package has been breached and the drip shield has been damaged such that seepage flow is actually directed into the waste package. The chemistry from the water influx case that is a function of the flow rate is used to determine the stability of colloids and the solubility of radionuclides available for transport by diffusion and advection, and to determine the degradation rates for the CSNF and HLW glass. TSPA-LA does not use this model for the igneous scenario. Outputs from the in-package chemistry model implemented inside TSPA-LA include pH, ionic strength, and total carbonate concentration. These inputs to TSPA-LA will be linked to the following principle factors: dissolution rates of the CSNF and HLWG, dissolved concentrations of radionuclides, and colloid generation

Isotopic analysis of radioactive waste packages (an inexpensive approach)

International Nuclear Information System (INIS)

Padula, D.A.; Richmond, J.S.

1983-01-01

A computer printout of the isotopic analysis for all radioactive waste packages containing resins, or other aqueous filter media is now required at the disposal sites at Barnwell, South Carolina, and Beatty, Nevada. Richland, Washington requires an isotopic analysis for all radioactive waste packages. The NRC (Nuclear Regulatory Commission), through 10 CFR 61, will require shippers of radioactive waste to classify and label for disposal all radioactive waste forms. These forms include resins, filters, sludges, and dry active waste (trash). The waste classification is to be based upon 10 CFR 61 (Section 1-7). The isotopes upon which waste classification is to be based are tabulated. 7 references, 8 tables

Comparison of transports expected under different waste management concepts: determination of basic data for application in risk analyses

International Nuclear Information System (INIS)

Alter, U.; Mielke, H.G.; Wehner, G.

1983-01-01

According to the Atomic Act, article 9a, paragraph 1, the licensees of nuclear power plants in the Federal Republic of Germany are obliged to provide for the management of radioactive wastes resulting from the operation of these plants. Concerning the provisions to be made for the management of such wastes, two concepts are discussed: nuclear reprocessing and final waste disposal center (Nukleares Entsorgungszentrum, NEZ); and the integrated spent fuel and waste management concept (Integriertes Entsorgungskonzept, IEK). Unlike the NEZ, the IEK-concept may have different sites for the following fuel cycle facilities: intermediate spent fuel storage, reprocessing, waste conditioning and final disposal, and uranium and plutonium fuel element fabrication facilities. The fundamental differences of the pertinent transports are presented. Transport scenarios expected under the two alternatives NEZ and IEK have been elaborated for the purpose of a data collection covering the following aspects: materials to be shipped, number of packages shipped, number of packages shipped per transport, transport by rail or by road, transport routes and distances, and duration of transports

Thermal modeling of nuclear waste package designs for disposal in tuff

International Nuclear Information System (INIS)

Hockman, J.N.; O'Neal, W.C.

1983-09-01

Lawrence Livermore National Laboratory is involved in the design and testing of high-level nuclear waste packages. Many of the aspects of waste package design and testing (e.g., corrosion and leaching) depend in part on the temperature history of the emplaced packages. This paper discusses thermal modeling and analysis of various emplaced waste package conceptual designs including the models used, the assumptions and approximations made, and the results obtained. 6 references, 6 figures, 3 tables

Thermal modeling of nuclear waste package designs for disposal in tuff

International Nuclear Information System (INIS)

Hockman, J.N.; O'Neal, W.C.

1984-02-01

Lawrence Livermore National Laboratory is involved in the design and testing of high-level nuclear waste packages. Many of the aspects of waste package design and testing (e.g., corrosion and leaching) depend in part on the temperature history of the emplaced packages. This paper discusses thermal modeling and analysis of various emplaced waste package conceptual designs including the models used, the assumptions and approximations made, and the results obtained. 6 references, 6 figures, 4 tables

A waste package strategy for regulatory compliance

International Nuclear Information System (INIS)

Stahl, D.; Cloninger, M.O.

1990-01-01

This paper summarizes the strategy given in the Site Characterization Plan for demonstrating compliance with the post closure performance objectives for the waste package and the Engineered Barrier System contained in the Code of Federal Regulations. The strategy consists of the development of a conservative waste package design that will meet the regulatory requirements with sufficient margin for uncertainty using a multi-barrier approach that takes advantage of the unsaturated nature of the Yucca Mountain site. 7 refs., 1 fig

An overview of the transportation of radioactive waste at Ontario Power Generation facilities

International Nuclear Information System (INIS)

Holmes, P.

2006-01-01

The Radioactive Material Transportation Department (RMT) ensures regulatory compliance in radioactive material shipping within Ontario Power Generation (OPG). OPG provides a radioactive shipping program, high quality carrier service, stringent packaging maintenance, and quality assurance oversight to the corporation's nuclear facilities and its customers. This paper will speak to the transport of radioactive waste in Ontario Power Generation. It will also mention non-waste shipments and the quality assurance programme used at Ontario Power Generation to ensure a high quality transportation system. (author)

Containment barrier metals for high-level waste packages in a Tuff repository

International Nuclear Information System (INIS)

Russell, E.W.; McCright, R.D.; O'Neal, W.C.

1983-01-01

The Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package project is part of the US Department of Energy's Civilian Radioactive Waste Management (CRWM) Program. The NNWSI project is working towards the development of multibarriered packages for the disposal of spent fuel and high-level waste in tuff in the unsaturated zone at Yucca Mountain at the Nevada Test Site (NTS). The final engineered barrier system design may be composed of a waste form, canister, overpack, borehole liner, packing, and the near field host rock, or some combination thereof. Lawrence Livermore National Laboratory's (LLNL) role is to design, model, and test the waste package subsystem for the tuff repository. At the present stage of development of the nuclear waste management program at LLNL, the detailed requirements for the waste package design are not yet firmly established. In spite of these uncertainties as to the detailed package requirements, we have begun the conceptual design stage. By conceptual design, we mean design based on our best assessment of present and future regulatory requirements. We anticipate that changes will occur as the detailed requirements for waste package design are finalized. 17 references, 4 figures, 10 tables

Containment barrier metals for high-level waste packages in a Tuff repository

Energy Technology Data Exchange (ETDEWEB)

Russell, E.W.; McCright, R.D.; O`Neal, W.C.

1983-10-12

The Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package project is part of the US Department of Energy`s Civilian Radioactive Waste Management (CRWM) Program. The NNWSI project is working towards the development of multibarriered packages for the disposal of spent fuel and high-level waste in tuff in the unsaturated zone at Yucca Mountain at the Nevada Test Site (NTS). The final engineered barrier system design may be composed of a waste form, canister, overpack, borehole liner, packing, and the near field host rock, or some combination thereof. Lawrence Livermore National Laboratory`s (LLNL) role is to design, model, and test the waste package subsystem for the tuff repository. At the present stage of development of the nuclear waste management program at LLNL, the detailed requirements for the waste package design are not yet firmly established. In spite of these uncertainties as to the detailed package requirements, we have begun the conceptual design stage. By conceptual design, we mean design based on our best assessment of present and future regulatory requirements. We anticipate that changes will occur as the detailed requirements for waste package design are finalized. 17 references, 4 figures, 10 tables.

Nuclear waste package design for the Vadose zone in tuff

International Nuclear Information System (INIS)

O'Neal, W.C.; Ballou, L.B.; Gregg, D.W.; Russell, E.W.

1984-02-01

This report presents an overview of the selection and analysis of conceptual waste package designs that will be used by the Nevada Nuclear Waste Storage Investigations (NNWSI) project for disposal of high-level nuclear waste (HLW) at the proposed Yucca Mountain, Nevada Site. The design requirements that the waste packages are required to meet are listed. Concept drawings for the reference designs and one alternative package design are shown. Four metal alloys; 304L SS, 321 SS, 316L SS and Incoloy 825 have been selected for candidate canister/overpack materials, and 1020 carbon steel has been selected as the reference metal for the borehole liners. A summary of the results of technical and economic analysis supporting the selection of the conceptual waste package designs is included. Post-closure containment and release rates are not discussed in this paper. 17 references, 2 figures, 2 tables

Role of waste packages in the safety of a high level waste repository in a deep geological formation

International Nuclear Information System (INIS)

Bretheau, F.; Lewi, J.

1990-06-01

The safety of a radioactive waste disposal facility lays on the three following barriers placed between the radioactive materials and the biosphere: the waste package; the engineered barriers; the geological barrier. The function assigned to each of these barriers in the performance assessment is an option taken by the organization responsible for waste disposal management (ANDRA in France), which must show that: expected performances of each barrier (confinement ability, life-time, etc.) are at least equal to those required to fulfill the assigned function; radiation protection requirements are met in all situations considered as credible, whether they be the normal situation or random event situations. The French waste management strategy is based upon two types of disposal depending on the nature and activity of waste packages: - surface disposal intended for low and medium level wastes having half-lives of about 30 years or less and alpha activity less than 3.7 MBq/kg (0.1 Ci/t), for individual packages and less than 0.37 MBq/kg (0.01 Ci/t) in the average. Deep geological disposal intended for TRU and high level wastes. The conditions of acceptance of packages in a surface disposal site are subject to the two fundamental safety rules no. I.2 and III.2.e. The present paper is only dealing with deep geological disposal. For deep geological repositories, three stages are involved: stage preceding definitive disposal (intermediate storage, transportation, handling, setting up in the disposal cavities); stage subsequent to definitive sealing of the disposal cavities but prior to the end of operation of the repository; stage subsequent to closure of the repository. The role of the geological barrier has been determined as the essential part of long term radioactivity confinement, by a working group, set up by the French safety authorities. Essential technical criteria relating to the choice of a site so defined by this group, are the following: very low permeability

Long-Term Waste Package Degradation Studies at the Yucca Mountain Potential High-Level Nuclear Waste Repository

International Nuclear Information System (INIS)

Mon, K. G.; Bullard, B. E.; Longsine, D. E.; Mehta, S.; Lee, J. H.; Monib, A. M.

2002-01-01

The Site Recommendation (SR) process for the potential repository for spent nuclear fuel (SNF) and high-level nuclear waste (HLW) at Yucca Mountain, Nevada is underway. Fulfillment of the requirements for substantially complete containment of the radioactive waste emplaced in the potential repository and subsequent slow release of radionuclides from the Engineered Barrier System (EBS) into the geosphere will rely on a robust waste container design, among other EBS components. Part of the SR process involves sensitivity studies aimed at elucidating which model parameters contribute most to the drip shield and waste package degradation characteristics. The model parameters identified included (a) general corrosion rate model parameters (temperature-dependence and uncertainty treatment), and (b) stress corrosion cracking (SCC) model parameters (uncertainty treatment of stress and stress intensity factor profiles in the Alloy 22 waste package outer barrier closure weld regions, the SCC initiation stress threshold, and the fraction of manufacturing flaws oriented favorably for through-wall penetration by SCC). These model parameters were reevaluated and new distributions were generated. Also, early waste package failures due to improper heat treatment were added to the waste package degradation model. The results of these investigations indicate that the waste package failure profiles are governed by the manufacturing flaw orientation model parameters and models used

Evolution of repository and waste package designs for Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste

International Nuclear Information System (INIS)

Rechard, Rob P.; Voegele, Michael D.

2014-01-01

This paper summarizes the evolution of the engineered barrier design for the proposed Yucca Mountain disposal system. Initially, the underground facility used a fairly standard panel and drift layout excavated mostly by drilling and blasting. By 1993, the layout of the underground facility was changed to accommodate construction by a tunnel boring machine. Placement of the repository in unsaturated zone permitted an extended period without backfilling; placement of the waste package in an open drift permitted use of much larger, and thus hotter packages. Hence in 1994, the underground facility design switched from floor emplacement of waste in small, single walled stainless steel or nickel alloy containers to in-drift emplacement of waste in large, double-walled containers. By 2000, the outer layer was a high nickel alloy for corrosion resistance and the inner layer was stainless steel for structural strength. Use of large packages facilitated receipt and disposal of high volumes of spent nuclear fuel. In addition, in-drift package placement saved excavation costs. Options considered for in-drift emplacement included different heat loads and use of backfill. To avoid dripping on the package during the thermal period and the possibility of localized corrosion, titanium drip shields were added for the disposal drifts by 2000. In addition, a handling canister, sealed at the reactor to eliminate further handling of bare fuel assemblies, was evaluated and eventually adopted in 2006. Finally, staged development of the underground layout was adopted to more readily adjust to changes in waste forms and Congressional funding. - Highlights: • Progression of events associated with repository design to accommodate tunnel boring machine and in-drift waste package emplacement are discussed. • Change in container design from small, single-layered stainless steel vessel to large, two-layered nickel alloy vessel is discussed. • The addition of drip shield to limit the

Release rates from waste packages in a salt repository

International Nuclear Information System (INIS)

Chambre, P.L.; Hwang, Y.; Lee, W.W.L.; Pigford, T.H.

1987-06-01

In this report we present estimates of radionuclide release rates from waste packages into salt. This conservative and bounding analysis shows that release rates from waste packages in salt are well below the US Nuclear Regulatory Commission's performance objectives for the engineered barrier system. 2 refs., 2 figs

Preclosure analysis of conceptual waste package designs for a nuclear waste repository in tuff

International Nuclear Information System (INIS)

O'Neal, W.C.; Gregg, D.W.; Hockman, J.N.; Russell, E.W.; Stein, W.

1984-01-01

This report discusses the selection and analysis of conceptual waste package developed by the Nevada Nuclear Waste Storage Investigations (NNWSI) project for possible disposal of high-level nuclear waste at a candidate site at Yucca Mountain, Nevada. The design requirements that the waste package must conform to are listed, as are several desirable design considerations. Illustrations of the reference and alternative designs are shown. Four austenitic stainless steels (316L SS, 321 SS, 304L SS and Incoloy 825 high nickel alloy) have been selected for candidate canister/overpack materials, and 1020 carbon steel has been selected as the reference metal for the borehole liners. A summary of the results of technical and ecnonmic analyses supporting the selection of the conceptual waste package designs is included. Postclosure containment and release rates are not analyzed in this report

REMOTE MATERIAL HANDLING IN THE YUCCA MOUNTAIN WASTE PACKAGE CLOSURE CELL AND SUPPORT AREA GLOVEBOX

International Nuclear Information System (INIS)

K.M. Croft; S.M. Allen; M.W. Borland

2005-01-01

The Yucca Mountain Waste Package Closure System (WPCS) cells provide for shielding of highly radioactive materials contained in unsealed waste packages. The purpose of the cells is to provide safe environments for package handling and sealing operations. Once sealed, the packages are placed in the Yucca Mountain Repository. Closure of a typical waste package involves a number of remote operations. Those involved typically include the placement of matched lids onto the waste package. The lids are then individually sealed to the waste package by welding. Currently, the waste package includes three lids. One lid is placed before movement of the waste package to the closure cell; the final two are placed inside the closure cell, where they are welded to the waste package. These and other important operations require considerable remote material handling within the cell environment. This paper discusses the remote material handling equipment, designs, functions, operations, and maintenance, relative to waste package closure

WASTE PACKAGE OPERATIONS FY99 CLOSURE METHODS REPORT

Energy Technology Data Exchange (ETDEWEB)

M. C. Knapp

1999-09-23

The waste package (WP) closure weld development task is part of a larger engineering development program to develop waste package designs. The purpose of the larger waste package engineering development program is to develop nuclear waste package fabrication and closure methods that the Nuclear Regulatory Commission will find acceptable and will license for disposal of spent nuclear fuel (SNF), non-fuel components, and vitrified high-level waste within a Monitored Geologic Repository (MGR). Within the WP closure development program are several major development tasks, which, in turn, are divided into subtasks. The major tasks include: WP fabrication development, WP closure weld development, nondestructive examination (NDE) development, and remote in-service inspection development. The purpose of this report is to present the objectives, technical information, and work scope relating to the WP closure weld development.and NDE tasks and subtasks and to report results of the closure weld and NDE development programs for fiscal year 1999 (FY-99). The objective of the FY-99 WP closure weld development task was to develop requirements for closure weld surface and volumetric NDE performance demonstrations, investigate alternative NDE inspection techniques, and develop specifications for welding, NDE, and handling system integration. In addition, objectives included fabricating several flat plate mock-ups that could be used for NDE development, stress relief peening, corrosion testing, and residual stress testing.

WASTE PACKAGE OPERATIONS FY-99 CLOSURE METHODS REPORT

International Nuclear Information System (INIS)

M. C. Knapp

1999-01-01

The waste package (WP) closure weld development task is part of a larger engineering development program to develop waste package designs. The purpose of the larger waste package engineering development program is to develop nuclear waste package fabrication and closure methods that the Nuclear Regulatory Commission will find acceptable and will license for disposal of spent nuclear fuel (SNF), non-fuel components, and vitrified high-level waste within a Monitored Geologic Repository (MGR). Within the WP closure development program are several major development tasks, which, in turn, are divided into subtasks. The major tasks include: WP fabrication development, WP closure weld development, nondestructive examination (NDE) development, and remote in-service inspection development. The purpose of this report is to present the objectives, technical information, and work scope relating to the WP closure weld development.and NDE tasks and subtasks and to report results of the closure weld and NDE development programs for fiscal year 1999 (FY-99). The objective of the FY-99 WP closure weld development task was to develop requirements for closure weld surface and volumetric NDE performance demonstrations, investigate alternative NDE inspection techniques, and develop specifications for welding, NDE, and handling system integration. In addition, objectives included fabricating several flat plate mock-ups that could be used for NDE development, stress relief peening, corrosion testing, and residual stress testing

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

International Nuclear Information System (INIS)

Lewis, Mark S.

2008-01-01

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

Hydrothermal waste package interactions with methane-containing basalt groundwater

International Nuclear Information System (INIS)

McGrail, B.P.

1984-01-01

Hydrothermal waste package interaction tests were conducted with a mixture of crushed glass, basalt, and steel in methane-containing synthetic basalt groundwater. In the absence of gamma radiolysis, methane was found to have little influence on the corrosion behavior of the waste package constituents. Under gamma radiolysis, methane was found to significantly lower the solution oxidation potential when compared to identical tests without methane. In addition, colloidal hydrocarbon polymers that have been produced under the irradiation conditions of these experiments were not formed. The presence of the waste package constituents apparently inhibited the formation of the polymers. However, the mechanism which prevented their formation was not determined

Test for radioactive material transport package safety

International Nuclear Information System (INIS)

Li Guoqiang; Zhao Bing; Zhang Jiangang; Wang Xuexin; Ma Anping

2012-01-01

Regulations on radioactive material transport in China were introduced. Test facilities and data acquiring instruments for radioactive material package in China Institute for Radiation Protection were also introduced in this paper, which were used in drop test and thermal test. Test facilities were constructed according to the requirements of IAEA's 'Regulations for the Safe Transport of Radioactive Material' (TS-R-l) and Chinese 'Regulations for the Safe Transport of Radioactive Material' (GB 11806-2004). Drop test facilities were used in free drop test, penetration test, mechanical test (free drop test Ⅰ, free drop test Ⅱ and free drop test Ⅲ) of type A and type B packages weighing less than thirteen tons. Thermal test of type B packages can be carried out in the thermal test facilities. Certification tests of type FCo70-YQ package, type 30A-HB-01 package, type SY-I package and type XAYT-I package according to regulations were done using these facilities. (authors)

Mixed waste chemical compatibility with packaging components

International Nuclear Information System (INIS)

Nigrey, P.J.; Conroy, M.; Blalock, L.B.

1994-01-01

In this paper, a chemical compatibility testing program for packaging of mixed wastes at will be described. We will discuss the choice of four y-radiation doses, four time durations, four temperatures and four waste solutions to simulate the hazardous waste components of mixed wastes for testing materials compatibility of polymers. The selected simulant wastes are (1) an aqueous alkaline mixture of sodium nitrate and sodium nitrite; (2) a chlorinated hydrocarbon mixture; (3) a simulant liquid scintillation fluid; and (4) a mixture of ketones. A selection of 10 polymers with anticipated high resistance to one or more of these types of environments are proposed for testing as potential liner or seal materials. These polymers are butadiene acrylonitrile copolymer, cross-linked polyethylene, epichlorhyarin, ethylene-propylene rubber, fluorocarbon, glass-filled tetrafluoroethylene, high-density poly-ethylene, isobutylene-isoprene copolymer, polypropylene, and styrene-butadiene rubber. We will describe the elements of the testing plan along with a metric for establishing time resistance of the packaging materials to radiation and chemicals

NWTS program criteria for mined geologic disposal of nuclear waste: functional requirements and performance criteria for waste packages for solidified high-level waste and spent fuel

International Nuclear Information System (INIS)

1982-07-01

The Department of Energy (DOE) has primary federal responsibility for the development and implementation of safe and environmentally acceptable nuclear waste disposal methods. Currently, the principal emphasis in the program is on emplacement of nuclear wastes in mined geologic repositories well beneath the earth's surface. A brief description of the mined geologic disposal system is provided. The National Waste Terminal Storage (NWTS) program was established under DOE's predecessor, the Energy Research and Development Administration, to provide facilities for the mined geologic disposal of radioactive wastes. The NWTS program includes both the development and the implementation of the technology necessary for designing, constructing, licensing, and operating repositories. The program does not include the management of processing radioactive wastes or of transporting the wastes to repositories. The NWTS-33 series, of which this document is a part, provides guidance for the NWTS program in the development and implementation of licensed mined geologic disposal systems for solidified high-level and transuranic (TRU) wastes. This document presents the functional requirements and performance criteria for waste packages for solidified high-level waste and spent fuel. A separate document to be developed, NWTS-33(4b), will present the requirements and criteria for waste packages for TRU wastes. The hierarchy and application of these requirements and criteria are discussed in Section 2.2

Determination of Radioisotope Content by Measurement of Waste Package Dose Rates - 13394

Energy Technology Data Exchange (ETDEWEB)

Souza, Daiane Cristini B.; Gimenes Tessaro, Ana Paula; Vicente, Roberto [Nuclear and Energy Research Institute Brazil, Radioactive Waste Management Department IPEN/GRR, Sao Paulo. SP. (Brazil)

2013-07-01

The objective of this communication is to report the observed correlation between the calculated air kerma rates produced by radioactive waste drums containing untreated ion-exchange resin and activated charcoal slurries with the measured radiation field of each package. Air kerma rates at different distances from the drum surface were calculated with the activity concentrations previously determined by gamma spectrometry of waste samples and the estimated mass, volume and geometry of solid and liquid phases of each waste package. The water content of each waste drum varies widely between different packages. Results will allow determining the total activity of wastes and are intended to complete the previous steps taken to characterize the radioisotope content of wastes packages. (authors)

Solvent extraction as additional purification method for postconsumer plastic packaging waste

NARCIS (Netherlands)

Thoden van Velzen, E.U.; Jansen, M.

2011-01-01

An existing solvent extraction process currently used to convert lightly polluted post-industrial packaging waste into high quality re-granulates was tested under laboratory conditions with highly polluted post-consumer packaging waste originating from municipal solid refuse waste. The objective was

Environmental effects of transporting radioactive materials in nuclear waste management systems

International Nuclear Information System (INIS)

Pope, R.B.; Yoshimura, H.R.; McClure, J.D.; Huerta, M.

1978-01-01

This paper discusses the environmental effects of radioactive materials transportation. The systems used or being designed for use in spent fuel and waste transportation are described. Accident rate and severity data are used to quantify risk. A test program in which subscale and full scale transportation systems were exposed to accident environments far in excess of those used in package design is used to relate package damage to accident severity levels. Analytical results and subscale and full scale test results are correlated to demonstrate that computational methods or scale modeling, or both, can be used to predict accident behavior of transportation systems. This work is used to show that the risks to the public from radioactive material transportation are low relative to other risks commonly accepted by the public

Production patterns of packaging waste categories generated at typical Mediterranean residential building worksites

Energy Technology Data Exchange (ETDEWEB)

González Pericot, N., E-mail: natalia.gpericot@upm.es [Escuela Técnica Superior de Edificación, Universidad Politécnica de Madrid, Calle Juan de Herrera n°6, 28040 Madrid (Spain); Villoria Sáez, P., E-mail: paola.villoria@upm.es [Escuela Técnica Superior de Edificación, Universidad Politécnica de Madrid, Calle Juan de Herrera n°6, 28040 Madrid (Spain); Del Río Merino, M., E-mail: mercedes.delrio@upm.es [Escuela Técnica Superior de Edificación, Universidad Politécnica de Madrid, Calle Juan de Herrera n°6, 28040 Madrid (Spain); Liébana Carrasco, O., E-mail: oscar.liebana@uem.es [Escuela de Arquitectura, Universidad Europea de Madrid, Calle Tajo s/n, 28670 Villaviciosa de Odón (Spain)

2014-11-15

Highlights: • On-site segregation level: 1.80%; training and motivation strategies were not effective. • 70% Cardboard waste: from switches and sockets during the building services stage. • 40% Plastic waste: generated during structures and partition works due to palletizing. • >50% Wood packaging waste, basically pallets, generated during the envelope works. - Abstract: The construction sector is responsible for around 28% of the total waste volume generated in Europe, which exceeds the amount of household waste. This has led to an increase of different research studies focusing on construction waste quantification. However, within the research studies made, packaging waste has been analyzed to a limited extent. This article focuses on the packaging waste stream generated in the construction sector. To this purpose current on-site waste packaging management has been assessed by monitoring ten Mediterranean residential building works. The findings of the experimental data collection revealed that the incentive measures implemented by the construction company to improve on-site waste sorting failed to achieve the intended purpose, showing low segregation ratios. Subsequently, through an analytical study the generation patterns for packaging waste are established, leading to the identification of the prevailing kinds of packaging and the products responsible for their generation. Results indicate that plastic waste generation maintains a constant trend throughout the whole construction process, while cardboard becomes predominant towards the end of the construction works with switches and sockets from the electricity stage. Understanding the production patterns of packaging waste will be beneficial for adapting waste management strategies to the identified patterns for the specific nature of packaging waste within the context of construction worksites.

Production patterns of packaging waste categories generated at typical Mediterranean residential building worksites

International Nuclear Information System (INIS)

González Pericot, N.; Villoria Sáez, P.; Del Río Merino, M.; Liébana Carrasco, O.

2014-01-01

Highlights: • On-site segregation level: 1.80%; training and motivation strategies were not effective. • 70% Cardboard waste: from switches and sockets during the building services stage. • 40% Plastic waste: generated during structures and partition works due to palletizing. • >50% Wood packaging waste, basically pallets, generated during the envelope works. - Abstract: The construction sector is responsible for around 28% of the total waste volume generated in Europe, which exceeds the amount of household waste. This has led to an increase of different research studies focusing on construction waste quantification. However, within the research studies made, packaging waste has been analyzed to a limited extent. This article focuses on the packaging waste stream generated in the construction sector. To this purpose current on-site waste packaging management has been assessed by monitoring ten Mediterranean residential building works. The findings of the experimental data collection revealed that the incentive measures implemented by the construction company to improve on-site waste sorting failed to achieve the intended purpose, showing low segregation ratios. Subsequently, through an analytical study the generation patterns for packaging waste are established, leading to the identification of the prevailing kinds of packaging and the products responsible for their generation. Results indicate that plastic waste generation maintains a constant trend throughout the whole construction process, while cardboard becomes predominant towards the end of the construction works with switches and sockets from the electricity stage. Understanding the production patterns of packaging waste will be beneficial for adapting waste management strategies to the identified patterns for the specific nature of packaging waste within the context of construction worksites

Peer Review of the Waste Package Material Performance Interim Report

International Nuclear Information System (INIS)

J. A. Beavers; T. M. Devine, Jr.; G. S. Frankel; R. H. Jones; R. G. Kelly; R. M. Latanision; J. H. Payer

2001-01-01

At the request of the U.S. Department of Energy, Bechtel SAIC Company, LLC, formed the Waste Package Materials Performance Peer Review Panel (the Panel) to review the technical basis for evaluating the long-term performance of waste package materials in a proposed repository at Yucca Mountain, Nevada. This is the interim report of the Panel; a final report will be issued in February 2002. In its work to date, the Panel has identified important issues regarding waste package materials performance. In the remainder of its work, the Panel will address approaches and plans to resolve these issues. In its review to date, the Panel has not found a technical basis to conclude that the waste package materials are unsuitable for long-term containment at the proposed Yucca Mountain Repository. Nevertheless, significant technical issues remain unsettled and, primarily because of the extremely long life required for the waste packages, there will always be some uncertainty in the assessment. A significant base of scientific and engineering knowledge for assessing materials performance does exist and, therefore, the likelihood is great that uncertainty about the long-term performance can be substantially reduced through further experiments and analysis

Recycling potential of post-consumer plastic packaging waste in Finland.

Science.gov (United States)

Dahlbo, Helena; Poliakova, Valeria; Mylläri, Ville; Sahimaa, Olli; Anderson, Reetta

2018-01-01

Recycling of plastics is urged by the need for closing material loops to maintain our natural resources when striving towards circular economy, but also by the concern raced by observations of plastic scrap in oceans and lakes. Packaging industry is the sector using the largest share of plastics, hence packaging dominates in the plastic waste flow. The aim of this paper was to sum up the recycling potential of post-consumer plastic packaging waste in Finland. This potential was evaluated based on the quantity, composition and mechanical quality of the plastic packaging waste generated by consumers and collected as a source-separated fraction, within the mixed municipal solid waste (MSW) or within energy waste. Based on the assessment 86,000-117,000 tons (18 kg/person/a) of post-consumer plastic packaging waste was generated in Finland in 2014. The majority, 84% of the waste was in the mixed MSW flow in 2014. Due to the launching of new sorting facilities and separate collections for post-consumer plastic packaging in 2016, almost 40% of the post-consumer plastic packaging could become available for recycling. However, a 50% recycling rate for post-consumer plastic packaging (other than PET bottles) would be needed to increase the overall MSW recycling rate from the current 41% by around two percentage points. The share of monotype plastics in the overall MSW plastics fraction was 80%, hence by volume the recycling potential of MSW plastics is high. Polypropylene (PP) and low density polyethylene (LDPE) were the most common plastic types present in mixed MSW, followed by polyethylene terephthalate (PET), polystyrene (PS) and high density polyethylene (HDPE). If all the Finnish plastic packaging waste collected through the three collection types would be available for recycling, then 19,000-25,000 tons of recycled PP and 6000-8000 tons of recycled HDPE would be available on the local market. However, this assessment includes uncertainties due to performing the

Nuclear waste package fabricated from concrete

International Nuclear Information System (INIS)

Pfeiffer, P.A.; Kennedy, J.M.

1987-03-01

After the United States enacted the Nuclear Waste Policy Act in 1983, the Department of Energy must design, site, build and operate permanent geologic repositories for high-level nuclear waste. The Department of Energy has recently selected three sites, one being the Hanford Site in the state of Washington. At this particular site, the repository will be located in basalt at a depth of approximately 3000 feet deep. The main concern of this site, is contamination of the groundwater by release of radionuclides from the waste package. The waste package basically has three components: the containment barrier (metal or concrete container, in this study concrete will be considered), the waste form, and other materials (such as packing material, emplacement hole liners, etc.). The containment barriers are the primary waste container structural materials and are intended to provide containment of the nuclear waste up to a thousand years after emplacement. After the containment barriers are breached by groundwater, the packing material (expanding sodium bentonite clay) is expected to provide the primary control of release of radionuclide into the immediate repository environment. The loading conditions on the concrete container (from emplacement to approximately 1000 years), will be twofold; (1) internal heat of the high-level waste which could be up to 400 0 C; (2) external hydrostatic pressure up to 1300 psi after the seepage of groundwater has occurred in the emplacement tunnel. A suggested container is a hollow plain concrete cylinder with both ends capped. 7 refs

Interim performance specifications for conceptual waste-package designs for geologic isolation in salt repositories

International Nuclear Information System (INIS)

1983-06-01

The interim performance specifications and data requirements presented apply to conceptual waste package designs for all waste forms which will be isolated in salt geologic repositories. The waste package performance specifications and data requirements respond to the waste package performance criteria. Subject areas treated include: containment and controlled release, operational period safety, criticality control, identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available

Waste Package Project quarterly report, July 1, 1995--September 30, 1995

International Nuclear Information System (INIS)

Ladkany, S.G.

1995-01-01

The following tasks are reported: overview and progress of nuclear waste package project and container design; nuclear waste container design considerations; structural investigation of multi purpose nuclear waste package canister; and design requirements of rock tunnel drift for long-term storage of high-level waste (faulted tunnel model study by photoelasticity/finite element analysis)

Safety Analysis Report for the KRI-ALM 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-ALM transport package to transport safely I-123, which is produced at Cyclotron in KIRAMS, was carried out. In the safety analyses results for the KRI-ALM 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-ALM transport package. Leak Test was performed after drop test, the measured leakage rate was lower than allowable leakage rate. It is revealed that the containment integrity of the KRI-ALM transport package is maintained. Therefore, it shows that the integrity of the KRI-ALM transport package is well maintained.

The packaging of intermediate and low level radioactive wastes

International Nuclear Information System (INIS)

Flowers, R.H.

1985-01-01

Solid radioactive wastes will generally require some kind of packaging to prepare them for a period of storage followed probably by a land burial. In this Paper the specification of the package is discussed in relation to the properties which will facilitate those two phases of the management of the waste. It is concluded that, by adopting the philosophy of redundant barriers for the disposal phase, a suitable package can be specified for any particular waste product even before the repository site has been selected. Low water flow and an appropriate depth to reduce the risk of accidental re-exposure are the technical site parameters for which particular values will have to be assured at that stage. (author)

The Role of Packaging in Solid Waste Management 1966 to 1976.

Science.gov (United States)

Darnay, Arsen; Franklin, William E.

The goals of waste processors and packagers obviously differ: the packaging industry seeks durable container material that will be unimpaired by external factors. Until recently, no systematic analysis of the relationship between packaging and solid waste disposal had been undertaken. This three-part document defines these interactions, and the…

Transportation package design using numerical optimization

International Nuclear Information System (INIS)

Harding, D.C.; Witkowski, W.R.

1992-01-01

The design of structures and engineering systems has always been an iterative process whose complexity was dependent upon the boundary conditions, constraints and available analytical tools. Transportation packaging design is no exception with structural, thermal and radiation shielding constraints based on regulatory hypothetical accident conditions. Transportation packaging design is often accomplished by a group of specialists, each designing a single component based on one or more simple criteria, pooling results with the group, evaluating the open-quotes pooledclose quotes design, and then reiterating the entire process until a satisfactory design is reached. The manual iterative methods used by the designer/analyst can be summarized in the following steps: design the part, analyze the part, interpret the analysis results, modify the part, and re-analyze the part. The inefficiency of this design practice and the frequently conservative result suggests the need for a more structured design methodology, which can simultaneously consider all of the design constraints. Numerical optimization is a structured design methodology whose maturity in development has allowed it to become a primary design tool in many industries. The purpose of this overview is twofold: first, to outline the theory and basic elements of numerical optimization; and second, to show how numerical optimization can be applied to the transportation packaging industry and used to increase efficiency and safety of radioactive and hazardous material transportation packages. A more extensive review of numerical optimization and its applications to radioactive material transportation package design was performed previously by the authors (Witkowski and Harding 1992). A proof-of-concept Type B package design is also presented as a simplified example of potential improvements achievable using numerical optimization in the design process

Waste package performance criteria for deepsea disposal of low-level radioactive wastes

International Nuclear Information System (INIS)

Colombo, P.; Fuhrmann, M.

1988-07-01

Sea disposal of low-level radioactive waste began in the United States in 1946, and was placed under the licensing authority of the Atomic Energy Commission (AEC). The practice stopped completely in 1970. Most of the waste disposed of at sea was packaged in second- hand or reconditioned 55-gallon drums filled with cement so that the average package density was sufficiently greater than that of sea water to ensure sinking. It was assumed that all the contents would eventually be released since the packages were not designed or required to remain intact for sustained periods of time after descent to the ocean bottom. Recently, there has been renewed interest in ocean disposal, both in this country and abroad, as a waste management alternative to land burial. The Marine Protection, Research and Sanctuaries Act of 1972 (PL 92-532) gives EPA the regulatory responsibility for ocean dumping of all materials, including radioactive waste. This act prohibits the ocean disposal of high-level radioactive waste and requires EPA to control the ocean disposal of all other radioactive waste through the issuance of permits. In implementing its permit authorities, EPA issued on initial set of regulations and criteria in 1973 to control the disposal of material into the ocean waters. It was in these regulations that EPA initially introduced the general requirement of isolation and containment of radioactive waste as the basic operating philosophy. 37 refs

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.

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

Radioactive waste packages stored at the Aube facility for low-intermediate activity wastes. A selective and controlled storage

International Nuclear Information System (INIS)

2005-01-01

The waste package is the first barrier designed to protect the man and the environment from the radioactivity contained in wastes. Its design is thus particularly stringent and controlled. This brochure describes the different types of packages for low to intermediate activity wastes like those received and stored at the Aube facility, and also the system implemented by the ANDRA (the French national agency of radioactive wastes) and by waste producers to safely control each step of the design and fabrication of these packages. (J.S.)

Life cycle assessment of a packaging waste recycling system in Portugal

International Nuclear Information System (INIS)

Ferreira, S.; Cabral, M.; Cruz, N.F. da; Simões, P.; Marques, R.C.

2014-01-01

Highlights: • We modeled a real packaging waste recycling system. • The analysis was performed using the life cycle assessment methodology. • The 2010 situation was compared with scenarios where the materials were not recycled. • The “Baseline” scenario seems to be more beneficial to the environment. - Abstract: Life Cycle Assessment (LCA) has been used to assess the environmental impacts associated with an activity or product life cycle. It has also been applied to assess the environmental performance related to waste management activities. This study analyses the packaging waste management system of a local public authority in Portugal. The operations of selective and refuse collection, sorting, recycling, landfilling and incineration of packaging waste were considered. The packaging waste management system in operation in 2010, which we called “Baseline” scenario, was compared with two hypothetical scenarios where all the packaging waste that was selectively collected in 2010 would undergo the refuse collection system and would be sent directly to incineration (called “Incineration” scenario) or to landfill (“Landfill” scenario). Overall, the results show that the “Baseline” scenario is more environmentally sound than the hypothetical scenarios

Life cycle assessment of a packaging waste recycling system in Portugal

Energy Technology Data Exchange (ETDEWEB)

Ferreira, S.; Cabral, M. [CEG-IST, ULisboa, Av. Rovisco Pais, 1049-001 Lisbon (Portugal); Cruz, N.F. da, E-mail: nunocruz@tecnico.ulisboa.pt [IST, ULisboa, Av. Rovisco Pais, 1049-001 Lisbon (Portugal); Simões, P. [IST, ULisboa, Av. Rovisco Pais, 1049-001 Lisbon (Portugal); Marques, R.C. [CESUR, IST, ULisboa, Av. Rovisco Pais, 1049-001 Lisbon (Portugal)

2014-09-15

Highlights: • We modeled a real packaging waste recycling system. • The analysis was performed using the life cycle assessment methodology. • The 2010 situation was compared with scenarios where the materials were not recycled. • The “Baseline” scenario seems to be more beneficial to the environment. - Abstract: Life Cycle Assessment (LCA) has been used to assess the environmental impacts associated with an activity or product life cycle. It has also been applied to assess the environmental performance related to waste management activities. This study analyses the packaging waste management system of a local public authority in Portugal. The operations of selective and refuse collection, sorting, recycling, landfilling and incineration of packaging waste were considered. The packaging waste management system in operation in 2010, which we called “Baseline” scenario, was compared with two hypothetical scenarios where all the packaging waste that was selectively collected in 2010 would undergo the refuse collection system and would be sent directly to incineration (called “Incineration” scenario) or to landfill (“Landfill” scenario). Overall, the results show that the “Baseline” scenario is more environmentally sound than the hypothetical scenarios.

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

Current issues in the transport of radioactive waste and spent fuel: work by the World Nuclear Transport Institute

Energy Technology Data Exchange (ETDEWEB)

Neau, H-J.; Bonnardel-Azzarelli, B. [World Nuclear Transport Inst., London (United Kingdom)

2014-07-01

Various kinds of radioactive waste are generated from nuclear power and fuel cycle facilities. These materials have to be treated, stored and eventually sent to a repository site. Transport of wastes between these various stages is crucial for the sustainable utilization of nuclear energy. The IAEA Regulations for the Safe Transport of Radioactive Material (SSR-6) have, for many decades, provided a safe and efficient framework for radioactive materials transport and continue to do so. However, some shippers have experienced that in the transport of certain specific radioactive wastes, difficulties can be encountered. For example, some materials produced in the decommissioning of nuclear facilities are unique in terms of composition or size and can be difficult to characterize as surface contaminated objects (SCO) or homogeneous. One way WNTI (World Nuclear Transport Institute) helps develop transport methodologies is through the use of Industry Working Groups, bringing together WNTI members with common interests, issues and experiences. The Back-End Transport Industry Working Group focuses on the following issues currently. - Characterization of Waste: techniques and methods to classify wastes - Large Objects: slightly contaminated large objects (ex. spent steam generators) transport - Dual Use Casks: transportable storage casks for spent nuclear fuels, including the very long term storage of spent fuel - Fissile Exceptions: new fissile exceptions provisions of revised TS-R-1 (SSR-6) The paper gives a broad overview of current issues for the packaging and transport of radioactive wastes and the associated work of the WNTI. (author)

Management and transport of radioactive wastes in Goiania, Brazil

International Nuclear Information System (INIS)

Xavier, A.M.; Mezrahi, A.

1988-01-01

The radiological accident occured in Goiania, which has led to the liberation to the environment of c.l.350 Ci of Cs-137, demanded the formulation of special procedures related to waste management and transport of radioactive materials to a provisory storage facility. The main objectives of the present article are to describe the work performed in the above mentioned fields and to point out the basic needs for Brazilian research and development in the areas of waste treatment and design and testing of packages. (author) [pt

Hydrothermal waste package interactions with methane-containing basalt groundwater

International Nuclear Information System (INIS)

McGrail, B.P.

1984-11-01

Hydrothermal waste package interaction tests with methane-containing synthetic basalt groundwater have shown that in the absence of gamma radiolysis, methane has little influence on the glass dissolution rate. Gamma radiolysis tests at fluxes of 5.5 x 10 5 and 4.4 x 10 4 R/hr showed that methane-saturated groundwater was more reducing than identical experiments where Ar was substituted for CH 4 . Dissolved methane, therefore, may be beneficial to the waste package in limiting the solubility of redox sensitive radionuclides such a 99 Tc. Hydrocarbon polymers known to form under the irradiation conditions of these tests were not produced. The presence of the waste package constituents apparently inhibited the formation of the polymers, however, the mechanism which prevented their formation was not determined

CH Packaging Operations for High Wattage Waste at LANL

International Nuclear Information System (INIS)

Washington TRU Solutions LLC

2002-01-01

This procedure provides instructions for assembling the following contact-handled (CH) packaging payloads: - Drum payload assembly - Standard Waste Box (SWB) assembly - Ten-Drum Overpack (TDOP) In addition, this procedure provides operating instructions for the TRUPACT-II CH waste packaging. This document also provides instructions for performing ICV and OCV preshipment leakage rate tests on the following packaging seals, using a nondestructive helium (He) leak test: - ICV upper main O-ring seal - ICV outer vent port plug O-ring seal - OCV upper main O-ring seal - OCV vent port plug O-ring seal

CH Packaging Operations for High Wattage Waste at LANL

International Nuclear Information System (INIS)

Washington TRU Solutions LLC

2002-01-01

This procedure provides instructions for assembling the following contact-handled (CH) packaging payloads: - Drum payload assembly - Standard Waste Box (SWB) assembly - Ten-Drum Overpack (TDOP) In addition, this procedure also provides operating instructions for the TRUPACT-II CH waste packaging. This document also provides instructions for performing ICV and OCV preshipment leakage rate tests on the following packaging seals, using a nondestructive helium (He) leak test: - ICV upper main O-ring seal - ICV outer vent port plug O-ring seal - OCV upper main O-ring seal - OCV vent port plug O-ring seal

CH Packaging Operations for High Wattage Waste at LANL

International Nuclear Information System (INIS)

Washington TRU Solutions LLC

2003-01-01

This procedure provides instructions for assembling the following contact-handled (CH) packaging payloads: - Drum payload assembly - Standard Waste Box (SWB) assembly - Ten-Drum Overpack (TDOP) In addition, this procedure also provides operating instructions for the TRUPACT-II CH waste packaging. This document also provides instructions for performing ICV and OCV preshipment leakage rate tests on the following packaging seals, using a nondestructive helium (He) leak test: - ICV upper main O-ring seal - ICV outer vent port plug O-ring seal - OCV upper main O-ring seal - OCV vent port plug O-ring seal

Transport experience of new ''TNF-XI'' powder package

International Nuclear Information System (INIS)

Nomura, I.; Fujiwara, T.; Naigeon, P.

2004-01-01

Since the Tokai criticality accident in 1999, there has been no specialized manufacturer conducting uranium re-conversion in Japan. For this reason, Nuclear Fuel Industries, Ltd. (NFI) imports from overseas almost all the uranium oxide powder used for manufacturing pellets for nuclear fuel assemblies. To date, an NT-IX package has been used for transporting the uranium oxide powder. However, due to the adoption of IAEA TS-R-1 into Japanese domestic regulations, we have begun to use a new TNF-XI powder package because the NT-IX package can suffer major deformation under the drop test III condition. The TNF-XI package was jointly developed by COGEMA LOGISTICS of France and NFI from 2000, and started to be used for actual transportation in 2003. This package has improved transport efficiency, handling operability and safety performance in comparison to its predecessor. This paper describes the characteristics of the new TNF-XI package and its actual transportation records and performance

Tritium waste package

Science.gov (United States)

Rossmassler, Rich; Ciebiera, Lloyd; Tulipano, Francis J.; Vinson, Sylvester; Walters, R. Thomas

1995-01-01

A containment and waste package system for processing and shipping tritium xide waste received from a process gas includes an outer drum and an inner drum containing a disposable molecular sieve bed (DMSB) seated within outer drum. The DMSB includes an inlet diffuser assembly, an outlet diffuser assembly, and a hydrogen catalytic recombiner. The DMSB absorbs tritium oxide from the process gas and converts it to a solid form so that the tritium is contained during shipment to a disposal site. The DMSB is filled with type 4A molecular sieve pellets capable of adsorbing up to 1000 curies of tritium. The recombiner contains a sufficient amount of catalyst to cause any hydrogen add oxygen present in the process gas to recombine to form water vapor, which is then adsorbed onto the DMSB.

Assessment of Quality Assurance Measures for Radioactive Material Transport Packages not Requiring Competent Authority Design Approval - 13282

International Nuclear Information System (INIS)

Komann, Steffen; Groeke, Carsten; Droste, Bernhard

2013-01-01

The majority of transports of radioactive materials are carried out in packages which don't need a package design approval by a competent authority. Low-active radioactive materials are transported in such packages e.g. in the medical and pharmaceutical industry and in the nuclear industry as well. Decommissioning of NPP's leads to a strong demand for packages to transport low and middle active radioactive waste. According to IAEA regulations the 'non-competent authority approved package types' are the Excepted Packages and the Industrial Packages of Type IP-1, IP-2 and IP-3 and packages of Type A. For these types of packages an assessment by the competent authority is required for the quality assurance measures for the design, manufacture, testing, documentation, use, maintenance and inspection (IAEA SSR 6, Chap. 306). In general a compliance audit of the manufacturer of the packaging is required during this assessment procedure. Their regulatory level in the IAEA regulations is not comparable with the 'regulatory density' for packages requiring competent authority package design approval. Practices in different countries lead to different approaches within the assessment of the quality assurance measures in the management system as well as in the quality assurance program of a special package design. To use the package or packaging in a safe manner and in compliance with the regulations a management system for each phase of the life of the package or packaging is necessary. The relevant IAEA-SSR6 chap. 801 requires documentary verification by the consignor concerning package compliance with the requirements. (authors)

Assessment of Quality Assurance Measures for Radioactive Material Transport Packages not Requiring Competent Authority Design Approval - 13282

Energy Technology Data Exchange (ETDEWEB)

Komann, Steffen; Groeke, Carsten; Droste, Bernhard [BAM Federal Institute for Materials Research and Testing, Unter den Eichen 44-46, 12203 Berlin (Germany)

2013-07-01

The majority of transports of radioactive materials are carried out in packages which don't need a package design approval by a competent authority. Low-active radioactive materials are transported in such packages e.g. in the medical and pharmaceutical industry and in the nuclear industry as well. Decommissioning of NPP's leads to a strong demand for packages to transport low and middle active radioactive waste. According to IAEA regulations the 'non-competent authority approved package types' are the Excepted Packages and the Industrial Packages of Type IP-1, IP-2 and IP-3 and packages of Type A. For these types of packages an assessment by the competent authority is required for the quality assurance measures for the design, manufacture, testing, documentation, use, maintenance and inspection (IAEA SSR 6, Chap. 306). In general a compliance audit of the manufacturer of the packaging is required during this assessment procedure. Their regulatory level in the IAEA regulations is not comparable with the 'regulatory density' for packages requiring competent authority package design approval. Practices in different countries lead to different approaches within the assessment of the quality assurance measures in the management system as well as in the quality assurance program of a special package design. To use the package or packaging in a safe manner and in compliance with the regulations a management system for each phase of the life of the package or packaging is necessary. The relevant IAEA-SSR6 chap. 801 requires documentary verification by the consignor concerning package compliance with the requirements. (authors)

Improved permeation barriers for tritiated waste packaging

International Nuclear Information System (INIS)

Vassallo, G.; Van Den Bergh, R.; Forcey, K.S.; Perujo, A.

1994-01-01

High-density polyethylene (HDPE) is extensively used as flexible bagging or packaging for soft tritiated waste in the tritium community because of its low permeability to the more radiotoxic form of tritium, i.e., tritiated water (HTO). However, HDPE does not represent a perfect barrier to HTO nor does it effectively hinder the permeation of elemental tritium, i.e, HT. This latter drawback is particularly important considering that the elemental form may readily convert to HTO outside of the waste package. The possible use of a multilayer film as packing material for the conditioning of tritiated waste is assessed, and its capability to hinder the permeation of elemental tritium is measured and compared with that of bare HDPE. The material investigated is readily available from the food industry. 5 refs., 1 tab

Impact limiters for radioactive materials transport packagings: a methodology for assessment

International Nuclear Information System (INIS)

Mourao, Rogerio Pimenta

2002-01-01

This work aims at establishing a methodology for design assessment of a cellular material-filled impact limiter to be used as part of a radioactive material transport packaging. This methodology comprises the selection of the cellular material, its structural characterization by mechanical tests, the development of a case study in the nuclear field, preliminary determination of the best cellular material density for the case study, performance of the case and its numerical simulation using the finite element method. Among the several materials used as shock absorbers in packagings, the polyurethane foam was chosen, particularly the foam obtained from the castor oil plant (Ricinus communis), a non-polluting and renewable source. The case study carried out was the 9 m drop test of a package prototype containing radioactive wastes incorporated in a cement matrix, considered one of the most severe tests prescribed by the Brazilian and international transport standards. Prototypes with foam density pre-determined as ideal as well as prototypes using lighter and heavier foams were tested for comparison. The results obtained validate the methodology in that expectations regarding the ideal foam density were confirmed by the drop tests and the numerical simulation. (author)

Criticality Potential of Waste Packages Containing DOE SNF Affected by Igneous Intrusion

International Nuclear Information System (INIS)

D.S. Kimball; C.E. Sanders

2006-01-01

The Department of Energy (DOE) is currently preparing an application to submit to the U.S. Nuclear Regulatory Commission for a construction authorization for a monitored geologic repository. The repository will contain spent nuclear fuel (SNF) and defense high-level waste (DHLW) in waste packages placed in underground tunnels, or drifts. The primary objective of this paper is to perform a criticality analysis for waste packages containing DOE SNF affected by a disruptive igneous intrusion event in the emplacement drifts. The waste packages feature one DOE SNF canister placed in the center and surrounded by five High-Level Waste (HLW) glass canisters. The effective neutron multiplication factor (k eff ) is determined for potential configurations of the waste package during and after an intrusive igneous event. Due to the complexity of the potential scenarios following an igneous intrusion, finding conservative and bounding configurations with respect to criticality requires some additional considerations. In particular, the geometry of a slumped and damaged waste package must be examined, drift conditions must be modeled over a range of parameters, and the chemical degradation of DOE SNF and waste package materials must be considered for the expected high temperatures. The secondary intent of this calculation is to present a method for selecting conservative and bounding configurations for a wide range of end conditions

Methodologies for assessing long-term performance of high-level radioactive waste packages

International Nuclear Information System (INIS)

Stephens, K.; Boesch, L.; Crane, B.; Johnson, R.; Moler, R.; Smith, S.; Zaremba, L.

1986-01-01

Several methods the Nuclear Regulatory Commission (NRC) can use to independently assess Department of Energy (DOE) waste package performance were identified by The Aerospace Corporation. The report includes an overview of the necessary attributes of performance assessment, followed by discussions of DOE methods, probabilistic methods capable of predicting waste package lifetime and radionuclide releases, process modeling of waste package barriers, sufficiency of the necessary input data, and the applicability of probability density functions. It is recommended that the initial NRC performance assessment (for the basalt conceptual waste package design) should apply modular simulation, using available process models and data, to demonstrate this assessment method

Number of Waste Package Hit by Igneous Intrusion

International Nuclear Information System (INIS)

M. Wallace

2004-01-01

The purpose of this scientific analysis report is to document calculations of the number of waste packages that could be damaged in a potential future igneous event through a repository at Yucca Mountain. The analyses include disruption from an intrusive igneous event and from an extrusive volcanic event. This analysis supports the evaluation of the potential consequences of future igneous activity as part of the total system performance assessment for the license application (TSPA-LA) for the Yucca Mountain Project (YMP). Igneous activity is a disruptive event that is included in the TSPA-LA analyses. Two igneous activity scenarios are considered: (1) The igneous intrusion groundwater release scenario (also called the igneous intrusion scenario) considers the in situ damage to waste packages or failure of waste packages that occurs if they are engulfed or otherwise affected by magma as a result of an igneous intrusion. (2) The volcanic eruption scenario depicts the direct release of radioactive waste due to an intrusion that intersects the repository followed by a volcanic eruption at the surface. An igneous intrusion is defined as the ascent of a basaltic dike or dike system (i.e., a set or swarm of multiple dikes comprising a single intrusive event) to repository level, where it intersects drifts. Magma that does reach the surface from igneous activity is an eruption (or extrusive activity) (Jackson 1997 [DIRS 109119], pp. 224, 333). The objective of this analysis is to develop a probabilistic measure of the number of waste packages that could be affected by each of the two scenarios

ERG review of waste package corrosion mechanisms

International Nuclear Information System (INIS)

Geisert, R.E.

1988-01-01

The Engineering Review Group (ERG) was established by the Office of Nuclear Waste Isolation (ONWI) to help evaluate engineering-related issues in the US Department of Energy's nuclear waste repository program. The ERG reviewed the waste package corrosion mechanisms. This report documents the ERG's comments and recommendations on these subjects and the ONWI response to the specific points raised by the ERG. 1 ref

Fracture mechanics based design for radioactive material transport packagings -- Historical review

International Nuclear Information System (INIS)

Smith, J.A.; Salzbrenner, D.; Sorenson, K.; McConnell, P.

1998-04-01

The use of a fracture mechanics based design for the radioactive material transport (RAM) packagings has been the subject of extensive research for more than a decade. Sandia National Laboratories (SNL) has played an important role in the research and development of the application of this technology. Ductile iron has been internationally accepted as an exemplary material for the demonstration of a fracture mechanics based method of RAM packaging design and therefore is the subject of a large portion of the research discussed in this report. SNL's extensive research and development program, funded primarily by the U. S. Department of Energy's Office of Transportation, Energy Management and Analytical Services (EM-76) and in an auxiliary capacity, the office of Civilian Radioactive Waste Management, is summarized in this document along with a summary of the research conducted at other institutions throughout the world. In addition to the research and development work, code and standards development and regulatory positions are also discussed

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

Application of systems engineering to determine performance requirements for repository waste packages

International Nuclear Information System (INIS)

Aitken, E.A.; Stimmell, G.L.

1987-01-01

The waste package for a nuclear waste repository in salt must contribute substantially to the performance objectives defined by the Salt Repository Project (SRP) general requirements document governing disposal of high-level waste. The waste package is one of the engineered barriers providing containment. In establishing the performance requirements for a project focused on design and fabrication of the waste package, the systems engineering methodology has been used to translate the hierarchy requirements for the repository system to specific performance requirements for design and fabrication of the waste package, a subsystem of the repository. This activity is ongoing and requires a methodology that provides traceability and is capable of iteration as baseline requirements are refined or changed. The purpose of this summary is to describe the methodology being used and the way it can be applied to similar activities in the nuclear industry

A comprehensive waste collection cost model applied to post-consumer plastic packaging waste

NARCIS (Netherlands)

Groot, J.J.; Bing, X.; Bos-Brouwers, H.E.J.; Bloemhof, J.M.

2014-01-01

Post-consumer plastic packaging waste (PPW) can be collected for recycling via source separation or post-separation. In source separation, households separate plastics from other waste before collection, whereas in post-separation waste is separated at a treatment centre after collection. There are

Packagings in the silicon era

International Nuclear Information System (INIS)

Beone, G.; Mione, A.; Orsini, A.; Forasassi, G.

1993-01-01

ENEA is studying, with the collaboration of the DCMN of the Pisa University, a new packaging to collect wastes in various facilities while proceeding to find a final disposal. Following a survey on the wastes that could be transported in the future, it was agreed to design a packaging able to contain an industrial drum, with a maximum capacity of 220 litres and a total weight less than 4000 N, previously filled with solid wastes in bulk or in a solid binding material. The packaging, to be approved as a Type B in agreement with the IAEA Regulations, will be useful to transport not only radioactive wastes but any kind of dangerous goods and also those not in agreement with the UNO Regulations. The 1/2 scale model of the packaging is formed by two concentric vessels of mild steel obtained by welding commercial shells to cylindrical walls and joined through a flange. The new packaging under development presents features that seem to be proper for its envisaged waste collection main use such as construction simplicity, relatively low cost, time and use endurance, low maintenance requirements. The design analysis and testing program ongoing at present allowed for the preliminary definition of the packaging geometry and confirmed the necessity of further investigations in some key areas as the determination of actual behaviour of the silicon foam, used as energy absorbing/thermal insulating material, in the specific conditions of interest. (J.P.N.)

EBS Radionuclide Transport Abstraction

International Nuclear Information System (INIS)

J.D. Schreiber

2005-01-01

The purpose of this report is to develop and analyze the engineered barrier system (EBS) radionuclide transport abstraction model, consistent with Level I and Level II model validation, as identified in ''Technical Work Plan for: Near-Field Environment and Transport: Engineered Barrier System: Radionuclide Transport Abstraction Model Report Integration'' (BSC 2005 [DIRS 173617]). The EBS radionuclide transport abstraction (or EBS RT Abstraction) is the conceptual model used in the total system performance assessment for the license application (TSPA-LA) to determine the rate of radionuclide releases from the EBS to the unsaturated zone (UZ). The EBS RT Abstraction conceptual model consists of two main components: a flow model and a transport model. Both models are developed mathematically from first principles in order to show explicitly what assumptions, simplifications, and approximations are incorporated into the models used in the TSPA-LA. The flow model defines the pathways for water flow in the EBS and specifies how the flow rate is computed in each pathway. Input to this model includes the seepage flux into a drift. The seepage flux is potentially split by the drip shield, with some (or all) of the flux being diverted by the drip shield and some passing through breaches in the drip shield that might result from corrosion or seismic damage. The flux through drip shield breaches is potentially split by the waste package, with some (or all) of the flux being diverted by the waste package and some passing through waste package breaches that might result from corrosion or seismic damage. Neither the drip shield nor the waste package survives an igneous intrusion, so the flux splitting submodel is not used in the igneous scenario class. The flow model is validated in an independent model validation technical review. The drip shield and waste package flux splitting algorithms are developed and validated using experimental data. The transport model considers

EBS Radionuclide Transport Abstraction

Energy Technology Data Exchange (ETDEWEB)

J.D. Schreiber

2005-08-25

The purpose of this report is to develop and analyze the engineered barrier system (EBS) radionuclide transport abstraction model, consistent with Level I and Level II model validation, as identified in ''Technical Work Plan for: Near-Field Environment and Transport: Engineered Barrier System: Radionuclide Transport Abstraction Model Report Integration'' (BSC 2005 [DIRS 173617]). The EBS radionuclide transport abstraction (or EBS RT Abstraction) is the conceptual model used in the total system performance assessment for the license application (TSPA-LA) to determine the rate of radionuclide releases from the EBS to the unsaturated zone (UZ). The EBS RT Abstraction conceptual model consists of two main components: a flow model and a transport model. Both models are developed mathematically from first principles in order to show explicitly what assumptions, simplifications, and approximations are incorporated into the models used in the TSPA-LA. The flow model defines the pathways for water flow in the EBS and specifies how the flow rate is computed in each pathway. Input to this model includes the seepage flux into a drift. The seepage flux is potentially split by the drip shield, with some (or all) of the flux being diverted by the drip shield and some passing through breaches in the drip shield that might result from corrosion or seismic damage. The flux through drip shield breaches is potentially split by the waste package, with some (or all) of the flux being diverted by the waste package and some passing through waste package breaches that might result from corrosion or seismic damage. Neither the drip shield nor the waste package survives an igneous intrusion, so the flux splitting submodel is not used in the igneous scenario class. The flow model is validated in an independent model validation technical review. The drip shield and waste package flux splitting algorithms are developed and validated using experimental data. The transport

EBS Radionuclide Transport Abstraction

Energy Technology Data Exchange (ETDEWEB)

J. Prouty

2006-07-14

The purpose of this report is to develop and analyze the engineered barrier system (EBS) radionuclide transport abstraction model, consistent with Level I and Level II model validation, as identified in Technical Work Plan for: Near-Field Environment and Transport: Engineered Barrier System: Radionuclide Transport Abstraction Model Report Integration (BSC 2005 [DIRS 173617]). The EBS radionuclide transport abstraction (or EBS RT Abstraction) is the conceptual model used in the total system performance assessment (TSPA) to determine the rate of radionuclide releases from the EBS to the unsaturated zone (UZ). The EBS RT Abstraction conceptual model consists of two main components: a flow model and a transport model. Both models are developed mathematically from first principles in order to show explicitly what assumptions, simplifications, and approximations are incorporated into the models used in the TSPA. The flow model defines the pathways for water flow in the EBS and specifies how the flow rate is computed in each pathway. Input to this model includes the seepage flux into a drift. The seepage flux is potentially split by the drip shield, with some (or all) of the flux being diverted by the drip shield and some passing through breaches in the drip shield that might result from corrosion or seismic damage. The flux through drip shield breaches is potentially split by the waste package, with some (or all) of the flux being diverted by the waste package and some passing through waste package breaches that might result from corrosion or seismic damage. Neither the drip shield nor the waste package survives an igneous intrusion, so the flux splitting submodel is not used in the igneous scenario class. The flow model is validated in an independent model validation technical review. The drip shield and waste package flux splitting algorithms are developed and validated using experimental data. The transport model considers advective transport and diffusive transport

EBS Radionuclide Transport Abstraction

International Nuclear Information System (INIS)

J. Prouty

2006-01-01

The purpose of this report is to develop and analyze the engineered barrier system (EBS) radionuclide transport abstraction model, consistent with Level I and Level II model validation, as identified in Technical Work Plan for: Near-Field Environment and Transport: Engineered Barrier System: Radionuclide Transport Abstraction Model Report Integration (BSC 2005 [DIRS 173617]). The EBS radionuclide transport abstraction (or EBS RT Abstraction) is the conceptual model used in the total system performance assessment (TSPA) to determine the rate of radionuclide releases from the EBS to the unsaturated zone (UZ). The EBS RT Abstraction conceptual model consists of two main components: a flow model and a transport model. Both models are developed mathematically from first principles in order to show explicitly what assumptions, simplifications, and approximations are incorporated into the models used in the TSPA. The flow model defines the pathways for water flow in the EBS and specifies how the flow rate is computed in each pathway. Input to this model includes the seepage flux into a drift. The seepage flux is potentially split by the drip shield, with some (or all) of the flux being diverted by the drip shield and some passing through breaches in the drip shield that might result from corrosion or seismic damage. The flux through drip shield breaches is potentially split by the waste package, with some (or all) of the flux being diverted by the waste package and some passing through waste package breaches that might result from corrosion or seismic damage. Neither the drip shield nor the waste package survives an igneous intrusion, so the flux splitting submodel is not used in the igneous scenario class. The flow model is validated in an independent model validation technical review. The drip shield and waste package flux splitting algorithms are developed and validated using experimental data. The transport model considers advective transport and diffusive transport

Transportation and packaging resource guide

Energy Technology Data Exchange (ETDEWEB)

Arendt, J.W.; Gove, R.M.; Welch, M.J.

1994-12-01

The purpose of this resource guide is to provide a convenient reference document of information that may be useful to the U.S. Department of Energy (DOE) and DOE contractor personnel involved in packaging and transportation activities. An attempt has been made to present the terminology of DOE community usage as it currently exists. DOE`s mission is changing with emphasis on environmental cleanup. The terminology or nomenclature that has resulted from this expanded mission is included for the packaging and transportation user for reference purposes. Older terms still in use during the transition have been maintained. The Packaging and Transportation Resource Guide consists of four sections: Sect. 1, Introduction; Sect. 2, Abbreviations and Acronyms; Sect. 3, Definitions; and Sect. 4, References for packaging and transportation of hazardous materials and related activities, and Appendices A and B. Information has been collected from DOE Orders and DOE documents; U.S Department of Transportation (DOT), U.S. Environmental Protection Agency (EPA), and U.S. Nuclear Regulatory Commission (NRC) regulations; and International Atomic Energy Agency (IAEA) standards and other international documents. The definitions included in this guide may not always be a regulatory definition but are the more common DOE usage. In addition, the definitions vary among regulatory agencies. It is, therefore, suggested that if a definition is to be used in a regulatory or a legal compliance issue, the definition should be verified with the appropriate regulation. To assist in locating definitions in the regulations, a listing of all definition sections in the regulations are included in Appendix B. In many instances, the appropriate regulatory reference is indicated in the right-hand margin.

Transportation and packaging resource guide

International Nuclear Information System (INIS)

Arendt, J.W.; Gove, R.M.; Welch, M.J.

1994-12-01

The purpose of this resource guide is to provide a convenient reference document of information that may be useful to the U.S. Department of Energy (DOE) and DOE contractor personnel involved in packaging and transportation activities. An attempt has been made to present the terminology of DOE community usage as it currently exists. DOE's mission is changing with emphasis on environmental cleanup. The terminology or nomenclature that has resulted from this expanded mission is included for the packaging and transportation user for reference purposes. Older terms still in use during the transition have been maintained. The Packaging and Transportation Resource Guide consists of four sections: Sect. 1, Introduction; Sect. 2, Abbreviations and Acronyms; Sect. 3, Definitions; and Sect. 4, References for packaging and transportation of hazardous materials and related activities, and Appendices A and B. Information has been collected from DOE Orders and DOE documents; U.S Department of Transportation (DOT), U.S. Environmental Protection Agency (EPA), and U.S. Nuclear Regulatory Commission (NRC) regulations; and International Atomic Energy Agency (IAEA) standards and other international documents. The definitions included in this guide may not always be a regulatory definition but are the more common DOE usage. In addition, the definitions vary among regulatory agencies. It is, therefore, suggested that if a definition is to be used in a regulatory or a legal compliance issue, the definition should be verified with the appropriate regulation. To assist in locating definitions in the regulations, a listing of all definition sections in the regulations are included in Appendix B. In many instances, the appropriate regulatory reference is indicated in the right-hand margin

Assessment of plastic packaging waste : material origin, methods, properties

NARCIS (Netherlands)

Luijsterburg, B.J.; Goossens, J.G.P.

2014-01-01

The global plastics production has increased annually and a substantial part is used for packaging (in Europe 39%). Most plastic packages are discarded after a relatively short service life and the resulting plastic packaging waste is subsequently landfilled, incinerated or recycled. Laws of several

Safety analysis report for packaging onsite long-length contaminated equipment transport system

International Nuclear Information System (INIS)

McCormick, W.A.

1997-01-01

This safety analysis report for packaging describes the components of the long-length contaminated equipment (LLCE) transport system (TS) and provides the analyses, evaluations, and associated operational controls necessary for the safe use of the LLCE TS on the Hanford Site. The LLCE TS will provide a standardized, comprehensive approach for the disposal of approximately 98% of LLCE scheduled to be removed from the 200 Area waste tanks

Safety analysis report for packaging, onsite, long-length contaminated equipment transport system

Energy Technology Data Exchange (ETDEWEB)

McCormick, W.A.

1997-05-09

This safety analysis report for packaging describes the components of the long-length contaminated equipment (LLCE) transport system (TS) and provides the analyses, evaluations, and associated operational controls necessary for the safe use of the LLCE TS on the Hanford Site. The LLCE TS will provide a standardized, comprehensive approach for the disposal of approximately 98% of LLCE scheduled to be removed from the 200 Area waste tanks.

Full-scale testing of waste package inspection system

International Nuclear Information System (INIS)

Yagi, T.; Kuribayashi, H.; Moriya, Y.; Fujisawa, H.; Takebayashi, N.

1989-01-01

In land disposal of low-level radioactive waste (LLW) in Japan, it is legally required that the waste packages to be disposed of be inspected for conformance to applicable technical regulations prior to shipment from each existing power station. JGC has constructed a fully automatic waste package inspection system for the purpose of obtaining the required design data and proving the performance of the system. This system consists of three inspection units (for visual inspection, surface contamination/dose rate measurement and radioactivity/weight measurement), a labelling unit, a centralized control unit and a drum handling unit. The outstanding features of the system are as follows: The equipment and components are modularized and designed to be of the most compact size and the quality control functions are performed by an advanced centralized control system. The authors discuss how, as a result of the full-scale testing, it has been confirmed that this system satisfies all the performance requirements for the inspection of disposal packages

Packaging of hazardous materials and their transport in national and international road, rail, sea and air transport. Summaries of papers. Gefahrgutverpackung und deren Befoerderung im nationalen und internationalen Strassen-, Schienen-, See- und Luftverkehr. Kurzfassungen der Referate

Energy Technology Data Exchange (ETDEWEB)

1986-01-01

The packaging and transport of hazardous goods demands a high degree of knowledge and responsibility from those involved. The symposium aims to refresh and bring up to date this knowledge with contributions about the legal fundamentals for packaging for the transport of hazardous goods; classification of materials; examination, licensing and identification of packaging; responsibilities, hability, irregularities, insurance; compatibility of filling materials; hazardous goods as additional packaging; re-use, re-conditioning, recycling, waste management, etc. (orig./HSCH).

Waste package materials testing for a salt repository: 1983 status summary report

International Nuclear Information System (INIS)

Moak, D.P.

1986-09-01

The United States plans to safely dispose of nuclear waste in deep, stable geologic formations. As part of these plans, the US Department of Energy is sponsoring research on the designing and testing of waste packages and waste package materials. This fiscal year 1983 status report summarizes recent results of waste package materials testing in a salt environment. The results from these tests will be used by waste package designers and performance assessment experts. Release characteristics data are available on two waste forms (spent fuel and waste-containing glass) that were exposed to leaching tests at various radiation levels, temperatures, pH, glass surface area to solution volume ratios, and brine solutions simulating expected salt repository conditions. Candidate materials tested for corrosion resistance and other properties include iron alloys; TI-CODE 12, the most promising titanium alloy for containment; and nickel alloys. In component interaction testing, synergistic effects have not ruled out any candidate material. 21 refs., 37 figs., 15 tabs

Salt Repository Project: Waste Package Program (WPP) modeling activiteis: FY 1984 annual report

International Nuclear Information System (INIS)

Kuhn, W.L.; Simonson, S.A.; Pulsipher, B.A.

1987-03-01

The Pacific Northwest Laboratory (PNL) is supporting the US Department of Energy's (DOE) Salt Repository Project (SRP) through its Waste Package Program (WPP). During FY 1984, the WPP continued its program of waste package component development and interactions testing and application of the resulting data base to develop predictive models describing waste package degradation and radionuclide release. Within the WPP, the Modeling Task (Task 04 during FY 1984) was conducted to interpret the tests in such a way that scientifically defensible models can be developed for use in qualification of the waste package

Analysis and model testing of a Super Tiger Type B waste transport system in accident environments

International Nuclear Information System (INIS)

May, R.A.; Yoshimura, H.R.; Romesberg, L.E.; Joseph, B.J.

1980-01-01

Sandia National Laboratories is investigating the response of a Type B packaging containing drums of contact-handled transuranic waste (CH-TRU) as a part of a program to evaluate the adequacy of experimental and analytical methods for assessing the safety of waste transport systems in accident environments. A US NRC certified Type B package known as the Super Tiger was selected for the study. This overpack consists of inner and outer steel shells separated by rigid polyurethane foam and can be used for either highway or rail transportation. Tests using scale models of the vehicular system are being conducted in conjunction with computer analyses

Radiological safety assessment of transporting radioactive waste to the Gyeongju disposal facility in Korea

Energy Technology Data Exchange (ETDEWEB)

Jeong, Jong Tae; Baik, Min Hoon; Kang, Mun Ja; Ahn, Hong Joo; Hwang, Doo Seong; Hong, Dae Seok; Jeong, Yong Hwan; Kim, Kyung Su [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-12-15

A radiological safety assessment study was performed for the transportation of low level radioactive wastes which are temporarily stored in Korea Atomic Energy Research Institute (KAERI), Daejeon, Korea. We considered two kinds of wastes: (1) operation wastes generated from the routine operation of facilities; and (2) decommissioning wastes generated from the decommissioning of a research reactor in KAERI. The important part of the radiological safety assessment is related to the exposure dose assessment for the incident-free (normal) transportation of wastes, i.e., the radiation exposure of transport personnel, radiation workers for loading and unloading of radioactive waste drums, and the general public. The effective doses were estimated based on the detailed information on the transportation plan and on the radiological characteristics of waste packages. We also estimated radiological risks and the effective doses for the general public resulting from accidents such as an impact and a fire caused by the impact during the transportation. According to the results, the effective doses for transport personnel, radiation workers, and the general public are far below the regulatory limits. Therefore, we can secure safety from the viewpoint of radiological safety for all situations during the transportation of radioactive wastes which have been stored temporarily in KAERI.

Radiological Safety Assessment of Transporting Radioactive Wastes to the Gyeongju Disposal Facility in Korea

Directory of Open Access Journals (Sweden)

Jongtae Jeong

2016-12-01

Full Text Available A radiological safety assessment study was performed for the transportation of low level radioactive wastes which are temporarily stored in Korea Atomic Energy Research Institute (KAERI, Daejeon, Korea. We considered two kinds of wastes: (1 operation wastes generated from the routine operation of facilities; and (2 decommissioning wastes generated from the decommissioning of a research reactor in KAERI. The important part of the radiological safety assessment is related to the exposure dose assessment for the incident-free (normal transportation of wastes, i.e., the radiation exposure of transport personnel, radiation workers for loading and unloading of radioactive waste drums, and the general public. The effective doses were estimated based on the detailed information on the transportation plan and on the radiological characteristics of waste packages. We also estimated radiological risks and the effective doses for the general public resulting from accidents such as an impact and a fire caused by the impact during the transportation. According to the results, the effective doses for transport personnel, radiation workers, and the general public are far below the regulatory limits. Therefore, we can secure safety from the viewpoint of radiological safety for all situations during the transportation of radioactive wastes which have been stored temporarily in KAERI.

Preliminary assessment of the controlled release of radionuclides from waste packages containing borosilicate waste glass

International Nuclear Information System (INIS)

Strachan, D.M.; McGrail, B.P.; Apted, M.J.; Engle, D.W.; Eslinger, P.W.

1990-06-01

The purpose of this report is to provide a preliminary assessment of the release-rate for an engineered barriers subsystem (EBS) containing waste packages of defense high-level waste borosilicate glass at geochemical and hydrological conditions similar to the those at Yucca Mountain. The relationship between the proposed Waste Acceptance Preliminary Specifications (WAPS) test of glass- dissolution rate and compliance with the NRC's release-rate criterion is also evaluated. Calculations are reported for three hierarchical levels: EBS analysis, waste-package analysis, and waste-glass analysis. The following conclusions identify those factors that most acutely affect the magnitude of, or uncertainty in, release-rate performance

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

Development of waste packages for TRU-disposal. 5. Development of cylindrical metal package for TRU wastes

International Nuclear Information System (INIS)

Mine, Tatsuya; Mizubayashi, Hiroshi; Asano, Hidekazu; Owada, Hitoshi; Otsuki, Akiyoshi

2005-01-01

Development of the TRU waste package for hulls and endpieces compression canisters, which include long-lived and low sorption nuclides like C-14 is essential and will contribute a lot to a reasonable enhancement of safety and economy of the TRU-disposal system. The cylindrical metal package made of carbon steel for canisters to enhance the efficiency of the TRU-disposal system and to economically improve their stacking conditions was developed. The package is a welded cylindrical construction with a deep drawn upper cover and a disc plate for a bottom cover. Since the welding is mainly made only for an upper cover and a bottom disc plate, this package has a better containment performance for radioactive nuclide and can reduce the cost for construction and manufacturing including its welding control. Furthermore, this package can be laid down in pile for stacking in the circular cross section disposal tunnel for the sedimentary rock, which can drastically minimize the space for disposal tunnel as mentioned previously in TRU report. This paper reports the results of the study for application of newly developed metal package into the previous TRU-disposal system and for the stacking equipment for the package. (author)

Sensitivity of the engineered barrier system (EBS) release rate to alternative conceptual models of advective release from waste packages under dripping fractures

International Nuclear Information System (INIS)

Lee, J.H.; Atkins, J.E.; McNeish, J.A.; Vallikat, V.

1996-01-01

Simulations were conducted to analyze the sensitivity of the engineered barrier system (EBS) release rate to alternative conceptual models of the advective release from waste packages under dripping fractures. The first conceptual model assumed that dripping water directly contacts the waste form inside the 'failed' waste package, and radionuclides are released from the EBS by advection. The second conceptual model assumed that dripping water is diverted around the 'failed' waste package (because of the presence of corrosion products plugging the perforations) and dripping water is prevented from directly contacting the waste form. In the second model, radionuclides were assumed to transport through the perforations by diffusion, and, once outside the waste package, to be released from the EBS by advection. The second model was to incorporate more realism into the EBS release calculations. For the case with the second EBS release model, most radionuclides had significantly lower peak EBS release rates (from at least one to several orders of magnitude) than with the first EBS release model. The impacts of the alternative EBS release models were greater for the radionuclides with a low solubility (or solubility-limited radionuclides) than for the radionuclides with a high solubility (or waste form dissolution-limited radionuclides). The analyses indicated that the EBS release model representing advection through a 'failed' waste package (the first EBS release model) may be too conservative in predicting the EBS performance. One major implication from this sensitivity study was that a 'failed' waste package container with multiple perforations may still be able to perform effectively as an important barrier to radionuclide release. (author)

A history of solid waste packaging at the Hanford Site

International Nuclear Information System (INIS)

Duncan, D.R.; Weyns-Rollosson, D.I.; Pottmeyer, J.A.; Stratton, T.J.

1995-02-01

Since the initiation of the defense materials product mission, a total of more than 600,000 m 3 of radioactive solid waste has been stored or disposed at the US Department of Energy's (DOE) Hanford Site, located in southeastern Washington State. As the DOE complex prepares for its increasing role in environmental restoration and waste remediation, the characterization of buried and retrievably stored waste will become increasingly important. Key to this characterization is an understanding of the standards and specifications to which waste was packaged; the regulations that mandated these standards and specifications; the practices used for handling and packaging different waste types; and the changes in these practices with time

A mechanistic model for leaching from low-level radioactive waste packages

International Nuclear Information System (INIS)

Kempf, C.R.

1988-01-01

The development of a waste leaching model to predict radionuclide releases from porous wastes in corrodible outer containers in unsaturated conditions and/or conditions of intermittent water flow is summarized in this paper. Three major processes have been conceptualized as necessarily participating in waste leaching: infiltration of water to the waste package; interaction of this water with the waste; and exit of radionuclide-laden water from the waste package. Through the exit point, the main features of the whole leaching process ware held in common. The departure occurs in two main ways: 1) the method of entrance of the radionuclides to leachant (i.e. part of the waste-water interaction phase outlined earlier); and 2) the mode of exit from waste form/waste package (i.e., the exit of radionuclide-laden water phase). The first branching point, which occurs in relation to 1), leads to either readily soluble species directly entering leachant on contact, or to other processes - mainly expected to be diffusion, dissolution or ion exchange, or some combination thereof

Phosphates as packaging materials for separated nuclear wastes

International Nuclear Information System (INIS)

Audubert, F.

2006-10-01

The author gives an overview of fifteen years of research activities performed within the context of the so-called Bataille bill which recommended in 1991 new investigations on the management of nuclear wastes. She presents studies aimed at the elaboration of phosphates with an apatite structure, and outlines the determination of compositions adapted to iodine, caesium and tri- or tetravalent actinide incorporation. She reports the synthesis of phosphates with a monazite structure for caesium and actinide confinement. Finally, she reports studies dealing with the waste packaging issue (elaboration of packaging matrices, properties)

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

Quality control concept for radioactive waste packages

International Nuclear Information System (INIS)

Warnecke, E.; Martens, B.R.; Odoj, R.

1990-01-01

In the Federal Republic of Germany a contract with the BfS for the performance of quality control measures is necessary. It is principally possible to apply two alternative methods: random checks on waste packages or qualification of conditioning processes with subsequent inspections. Priority is given to the control by the process qualification. Both methods have successfully been developed in the Federal Republic of Germany and can be applied. In the course of the qualification of conditioning processes it must be demonstrated by inactive and/or active runs that waste packages are produced which fulfil the waste acceptance requirements. The qualification results in the fixation of a handbook for the operation of the respective conditioning process including the process instrumentation and the operational margins. The qualified process will be inspected to assure the compliance of the actual operation with the conditions fixed in the handbook. (orig./DG)

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

Defense Waste Processing Facility Process Simulation Package Life Cycle

International Nuclear Information System (INIS)

Reuter, K.

1991-01-01

The Defense Waste Processing Facility (DWPF) will be used to immobilize high level liquid radioactive waste into safe, stable, and manageable solid form. The complexity and classification of the facility requires that a performance based operator training to satisfy Department of Energy orders and guidelines. A major portion of the training program will be the application and utilization of Process Simulation Packages to assist in training the Control Room Operators on the fluctionality of the process and the application of the Distribution Control System (DCS) in operating and managing the DWPF process. The packages are being developed by the DWPF Computer and Information Systems Simulation Group. This paper will describe the DWPF Process Simulation Package Life Cycle. The areas of package scope, development, validation, and configuration management will be reviewed and discussed in detail

Packaging waste recycling in Europe: is the industry paying for it?

Science.gov (United States)

da Cruz, Nuno Ferreira; Ferreira, Sandra; Cabral, Marta; Simões, Pedro; Marques, Rui Cunha

2014-02-01

This paper describes and examines the schemes established in five EU countries for the recycling of packaging waste. The changes in packaging waste management were mainly implemented since the Directive 94/62/EC on packaging and packaging waste entered into force. The analysis of the five systems allowed the authors to identify very different approaches to cope with the same problem: meet the recovery and recycling targets imposed by EU law. Packaging waste is a responsibility of the industry. However, local governments are generally in charge of waste management, particularly in countries with Green Dot schemes or similar extended producer responsibility systems. This leads to the need of establishing a system of financial transfers between the industry and the local governments (particularly regarding the extra costs involved with selective collection and sorting). Using the same methodological approach, the authors also compare the costs and benefits of recycling from the perspective of local public authorities for France, Portugal and Romania. Since the purpose of the current paper is to take note of who is paying for the incremental costs of recycling and whether the industry (i.e. the consumer) is paying for the net financial costs of packaging waste management, environmental impacts are not included in the analysis. The work carried out in this paper highlights some aspects that are prone to be improved and raises several questions that will require further research. In the three countries analyzed more closely in this paper the industry is not paying the net financial cost of packaging waste management. In fact, if the savings attained by diverting packaging waste from other treatment (e.g. landfilling) and the public subsidies to the investment on the "recycling system" are not considered, it seems that the industry should increase the financial support to local authorities (by 125% in France, 50% in Portugal and 170% in Romania). However, in France and

Progress in waste package and engineered barrier system performance assessment and design

International Nuclear Information System (INIS)

Van Luik, A.; Stahl, D.; Harrison, D.

1993-01-01

As part of the U.S. Department of Energy's evaluation of site suitability for a potential high-level radioactive waste repository, long-term interactions between the engineered barrier system and the site must be determined. This requires a waste-package/engineered-system design, a description of the environment around the emplacement zone, and models that simulate operative processes describing these engineered/natural systems interactions. Candidate designs are being evaluated, including a more robust, multi-barrier waste package, and a drift emplacement mode. Tools for evaluating designs, and emplacement mode are the currently available waste-package/engineered-system performance assessment codes development for the project. For assessments that support site suitability, environmental impact, or licensing decisions, more capable codes are needed. Code capability requirements are being written, and existing codes are to be evaluated against those requirements. Recommendations are being made to focus waste-packaging/engineered-system code-development

Evaluation of alternatives for a second-generation transportation system for Department of Energy transuranic waste

International Nuclear Information System (INIS)

1984-01-01

Department of Energy (DOE) waste storage sites will ship their contact-handled (CH) and remote-handled (RH) transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) beginning FY 1989. The CH-TRU waste will be shipped in the Transuranic Package Transported (TRUPACT-I), a new packaging being developed by Sandia National Laboratories, Albuquerque/Transportation Technology Center. Some of the DOE TRU waste, however, might be unsuitable for shipment in TRUPACT-I, and is designated special-shipped (SS) TRU waste. The purposes of this study were to: (1) identify the quantity and characteristics of SS-TRU waste stored and generated at DOE facilities; (2) identify alternatives for managing the SS-TRU waste; and (3) make overall recommendations for managing the SS-TRU waste. Data on quantity and characteristics were gathered through coordinating visits to the sites and extracting information from each site's records. Representatives of DOE organizations and contractors set objectives for managing the SS-TRU waste. Alternative shipping systems were then identified for CH SS-TRU waste and RH SS-TRU waste. Evaluations of these alternatives considered how well they would satisfy each objective, and associated potential problems. The study recommends delaying the decision on how best to transport the CH SS-TRU waste to WIPP until the amount of SS-TRU processed waste in heavy drums is known. These conditions and choices are presented: a relatively small number of processed, heavy drums could be shipped most economically via TRUPACT-I, mixed with lighter drums of unprocessed waste. If a large number of heavy drums is to be shipped, a shorter and narrower version of TRUPACT-I would be preferred alternative. The Defense High-Level Waste cask is the recommended alternative system for shipping RH SS-TRU waste. 12 references, 15 figures, 22 tables