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Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
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1

Radioisotope Thermoelectric Generator Transportation System licensed hardware second certification test series and package shock mount system test  

SciTech Connect (OSTI)

This paper presents a summary of two separate drop test activities that were performed in support of the Radioisotope Thermoelectric Generator (RTG) Transportation System (RTGTS). The first portion of this paper presents the second series of drop testing required to demonstrate that the RTG package design meets the requirements of {ital Title} 10, {ital Code} {ital of} {ital Federal} {ital Regulations}, {open_quote}{open_quote}Part 71{close_quote}{close_quote} (10 CFR 71). Results of the first test series, performed in July 1994, demonstrated that some design changes were necessary. The package design was modified to improve test performance and the design changes were incorporated into the Safety Analysis Report for Packaging (SARP). The second full-size certification test article (CTA-2) incorporated the modified design and was tested at the U.S. Department of Energy{close_quote}s (DOE) Hanford Site near Richland, Washington. With the successful completion of the test series, and pending DOE Office of Facility Safety Analysis approval of the SARP, a certificate of compliance will be issued for the RTG package allowing its use. The second portion of this paper presents the design and testing of the RTG Package Mount System. The RTG package mount was designed to protect the RTG from excessive vibration during transport, provide shock protection during on/off loading, and provide a mechanism for moving the RTG package with a forklift. Military Standard (MIL-STD) 810E, {ital Transit} {ital Drop} {ital Procedure} (DOE 1989), was used to verify that the shock limiting system limited accelerations in excess of 15 G{close_quote}s at frequencies below 150 Hz. Results of the package mount drop tests indicate that an impact force of 15 G{close_quote}s was not exceeded in any test from a free drop height of 457 mm (18 in.). {copyright} {ital 1996 American Institute of Physics.}

Ferrell, P.C.; Moody, D.A. [Westinghouse Hanford Company, P.O. Box 1970, Richland, Washington 99352 (United States)

1996-03-01T23:59:59.000Z

2

Packaging and Transportation Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

Establishes safety requirements for the proper packaging and transportation of offsite shipments and onsite transfers of hazardous materials andor modal transport. Cancels DOE 1540.2 and DOE 5480.3

1995-09-27T23:59:59.000Z

3

Packaging and Transportation Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

Establishes safety requirements for the proper packaging and transportation of Department of Energy (DOE) offsite shipments and onsite transfers of hazardous materials and for modal transport. Canceled by DOE 460.1A

1995-09-27T23:59:59.000Z

4

Packaging and Transportation Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

Establishes safety requirements for the proper packaging and transportation of Department of Energy (DOE) offsite shipments and onsite transfers of hazardous materials and for modal transport. Cancels DOE O 460.1.

1996-10-02T23:59:59.000Z

5

Romanian experience on packaging testing  

SciTech Connect (OSTI)

With more than twenty years ago, the Institute for Nuclear Research Pitesti (INR), through its Reliability and Testing Laboratory, was licensed by the Romanian Nuclear Regulatory Body- CNCAN and to carry out qualification tests [1] for packages intended to be used for the transport and storage of radioactive materials. Radioactive materials, generated by Romanian nuclear facilities [2] are packaged in accordance with national [3] and the IAEA's Regulations [1,6] for a safe transport to the disposal center. Subjecting these packages to the normal and simulating test conditions accomplish the evaluation and certification in order to prove the package technical performances. The paper describes the qualification tests for type A and B packages used for transport and storage of radioactive materials, during a period of 20 years of experience. Testing is used to substantiate assumption in analytical models and to demonstrate package structural response. The Romanian test facilities [1,3,6] are used to simulate the required qualification tests and have been developed at INR Pitesti, the main supplier of type A packages used for transport and storage of low radioactive wastes in Romania. The testing programme will continue to be a strong option to support future package development, to perform a broad range of verification and certification tests on radioactive material packages or component sections, such as packages used for transport of radioactive sources to be used for industrial or medical purposes [2,8]. The paper describes and contain illustrations showing some of the various tests packages which have been performed during certain periods and how they relate to normal conditions and minor mishaps during transport. Quality assurance and quality controls measures taken in order to meet technical specification provided by the design there are also presented and commented. (authors)

Vieru, G. [IAEA Technical Expert, Head, Reliability and Testing Lab., Institute for Nuclear Research (Romania)

2007-07-01T23:59:59.000Z

6

Packaging and Transportation Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The order establishes safety requirements for the proper packaging and transportation of DOE, including NNSA, offsite shipments and onsite transfers of radioactive and other hazardous materials and for modal transportation. Cancels DOE O 460.1B, 5-14-10

2010-05-14T23:59:59.000Z

7

Packaging and Transportation Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

To establish safety requirements for the proper packaging and transportation of Department of Energy (DOE)/National Nuclear Security Administration (NNSA) offsite shipments and onsite transfers of hazardous materials and for modal transport. Cancels DOE O 460.1A. Canceled by DOE O 460.1C.

2003-04-04T23:59:59.000Z

8

Underground Test Area Subproject Phase I Data Analysis Task. Volume VII - Tritium Transport Model Documentation Package  

SciTech Connect (OSTI)

Volume VII of the documentation for the Phase I Data Analysis Task performed in support of the current Regional Flow Model, Transport Model, and Risk Assessment for the Nevada Test Site Underground Test Area Subproject contains the tritium transport model documentation. Because of the size and complexity of the model area, a considerable quantity of data was collected and analyzed in support of the modeling efforts. The data analysis task was consequently broken into eight subtasks, and descriptions of each subtask's activities are contained in one of the eight volumes that comprise the Phase I Data Analysis Documentation.

None

1996-12-01T23:59:59.000Z

9

Departmental Materials Transportation and Packaging Management  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

Establishes requirements and responsibilities for management of Department of Energy (DOE), including National Nuclear Security Administration, materials transportation and packaging and ensures the safe, secure, efficient packaging and transportation of materials, both hazardous and non-hazardous.

2010-11-18T23:59:59.000Z

10

Transportation and packaging resource guide  

SciTech Connect (OSTI)

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.

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

1994-12-01T23:59:59.000Z

11

Hazardous Materials Packaging and Transportation Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The Order establishes safety requirements for the proper packaging and transportation of Department of offsite shipments and onsite transfers of radioactive and other hazardous materials, and for modal transportation.

2015-04-20T23:59:59.000Z

12

TYPE A FISSILE PACKAGING FOR AIR TRANSPORT PROJECT OVERVIEW  

SciTech Connect (OSTI)

This paper presents the project status of the Model 9980, a new Type A fissile packaging for use in air transport. The Savannah River National Laboratory (SRNL) developed this new packaging to be a light weight (<150-lb), drum-style package and prepared a Safety Analysis for Packaging (SARP) for submission to the DOE/EM. The package design incorporates unique features and engineered materials specifically designed to minimize packaging weight and to be in compliance with 10CFR71 requirements. Prototypes were fabricated and tested to evaluate the design when subjected to Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC). An overview of the design details, results of the regulatory testing, and lessons learned from the prototype fabrication for the 9980 will be presented.

Eberl, K.; Blanton, P.

2013-10-11T23:59:59.000Z

13

Departmental Materials Transportation and Packaging Management  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The Order establishes requirements and responsibilities for management of Department of Energy (DOE), including National Nuclear Security Administration (NNSA), materials transportation and packaging to ensure the safe, secure, efficient packaging and transportation of materials, both hazardous and nonhazardous. Cancels DOE O 460.2 and DOE O 460.2 Chg 1

2004-12-22T23:59:59.000Z

14

Departmental Materials Transportation and Packaging Management  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

Establishes Department of Energy (DOE) policies and requirements to supplement applicable laws, rules, regulations, and other DOE Orders for materials transportation and packaging operations. Cancels: DOE 1540.1A, DOE 1540.2, and DOE 1540.3A.

1995-10-26T23:59:59.000Z

15

Departmental Materials Transportation and Packaging Management  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

Establishes Department of Energy (DOE) policies and requirements to supplement applicable laws, rules, regulations, and other DOE Orders for materials transportation and packaging operations. Cancels DOE 1540.1A, DOE 1540.2, DOE 1540.3A.

1995-09-27T23:59:59.000Z

16

Radioactive material package seal tests  

SciTech Connect (OSTI)

General design or test performance requirements for radioactive materials (RAM) packages are specified in Title 10 of the US Code of Federal Regulations Part 71 (US Nuclear Regulatory Commission, 1983). The requirements for Type B packages provide a broad range of environments under which the system must contain the RAM without posing a threat to health or property. Seals that provide the containment system interface between the packaging body and the closure must function in both high- and low-temperature environments under dynamic and static conditions. A seal technology program, jointly funded by the US Department of Energy Office of Environmental Restoration and Waste Management (EM) and the Office of Civilian Radioactive Waste Management (OCRWM), was initiated at Sandia National Laboratories. Experiments were performed in this program to characterize the behavior of several static seal materials at low temperatures. Helium leak tests on face seals were used to compare the materials. Materials tested include butyl, neoprene, ethylene propylene, fluorosilicone, silicone, Eypel, Kalrez, Teflon, fluorocarbon, and Teflon/silicone composites. Because most elastomer O-ring applications are for hydraulic systems, manufacturer low-temperature ratings are based on methods that simulate this use. The seal materials tested in this program with a fixture similar to a RAM cask closure, with the exception of silicone S613-60, are not leak tight (1.0 {times} 10{sup {minus}7} std cm{sup 3}/s) at manufacturer low-temperature ratings. 8 refs., 3 figs., 1 tab.

Madsen, M.M.; Humphreys, D.L.; Edwards, K.R.

1990-01-01T23:59:59.000Z

17

Y-12 defense programs: Nuclear Packaging Systems testing capabilities  

SciTech Connect (OSTI)

The Nuclear Packaging Systems (NPS) Department can manage/accomplish any packaging task. The NPS organization is responsible for managing the design, testing, certification, procurement, operation, refurbishment, maintenance, and disposal of packaging used to transport radioactive materials, other hazardous materials, and general cargoes on public roads and within the Oak Ridge Y-12 Plant. Additionally, the NPS Department has developed a Quality Assurance plan for all packaging, design and procurement of nonweapon shipping containers for radioactive materials, and design and procurement of performance-oriented packaging for hazardous materials. Further, the NPS Department is responsible for preparation and submittal of Safety Analysis Reports for Packaging (SARP). The NPS Department coordinates shipping container procurement and safety certification activities that have lead-times of up to two years. A Packaging Testing Capabilities Table at the Oak Ridge complex is included as a table.

NONE

1995-06-01T23:59:59.000Z

18

Packaging and transportation manual. Chapter on the packaging and transportation of hazardous and radioactive waste  

SciTech Connect (OSTI)

The purpose of this chapter is to outline the requirements that Los Alamos National Laboratory employees and contractors must follow when they package and ship hazardous and radioactive waste. This chapter is applied to on-site, intra-Laboratory, and off-site transportation of hazardous and radioactive waste. The chapter contains sections on definitions, responsibilities, written procedures, authorized packaging, quality assurance, documentation for waste shipments, loading and tiedown of waste shipments, on-site routing, packaging and transportation assessment and oversight program, nonconformance reporting, training of personnel, emergency response information, and incident and occurrence reporting. Appendices provide additional detail, references, and guidance on packaging for hazardous and radioactive waste, and guidance for the on-site transport of these wastes.

NONE

1998-03-01T23:59:59.000Z

19

Testing of the CANDU Spent Fuel Storage Basket Package  

SciTech Connect (OSTI)

The paper described the results of testing for a CANDU Spent Fuel Storage Basket Package Prototype intended to be used for transport and storage of the CANDU spent fuel bundles within NPP CANDU Cernavoda, Romania. The results obtained proved that the objectives of those tests were achieved

Vieru, G.

2002-02-28T23:59:59.000Z

20

Regulatory and extra-regulatory testing to demonstrate radioactive material packaging safety  

SciTech Connect (OSTI)

Packages for the transportation of radioactive material must meet performance criteria to assure safety and environmental protection. The stringency of the performance criteria is based on the degree of hazard of the material being transported. Type B packages are used for transporting large quantities of radioisotopes (in terms of A{sub 2} quantities). These packages have the most stringent performance criteria. Material with less than an A{sub 2} quantity are transported in Type A packages. These packages have less stringent performance criteria. Transportation of LSA and SCO materials must be in {open_quotes}strong-tight{close_quotes} packages. The performance requirements for the latter packages are even less stringent. All of these package types provide a high level of safety for the material being transported. In this paper, regulatory tests that are used to demonstrate this safety will be described. The responses of various packages to these tests will be shown. In addition, the response of packages to extra-regulatory tests will be discussed. The results of these tests will be used to demonstrate the high level of safety provided to workers, the public, and the environment by packages used for the transportation of radioactive material.

Ammerman, D.J.

1997-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

Radioisotope thermoelectric generator licensed hardware package and certification tests  

SciTech Connect (OSTI)

This paper presents the Licensed Hardware package and the Certification Test portions of the Radioisitope Themoelectric Generator Transportation System. This package has been designed to meet those portions of the {ital Code} {ital of} {ital Federal} {ital Regulations} (10 CFR 71) relating to ``Type B`` shipments of radioactive materials. The licensed hardware is now in the U. S. Department of Energy licensing process that certifies the packaging`s integrity under accident conditions. The detailed information for the anticipated license is presented in the safety analysis report for packaging, which is now in process and undergoing necessary reviews. As part of the licensing process, a full-size Certification Test Article unit, which has modifications slightly different than the Licensed Hardware or production shipping units, is used for testing. Dimensional checks of the Certification Test Article were made at the manufacturing facility. Leak testing and drop testing were done at the 300 Area of the U.S. Department of Energy`s Hanford Site near Richland, Washington. The hardware includes independent double containments to prevent the environmental spread of {sup 238}Pu, impact limiting devices to protect portions of the package from impacts, and thermal insulation to protect the seal areas from excess heat during accident conditions. The package also features electronic feed-throughs to monitor the Radioisotope Thermoelectric Generator`s temperature inside the containment during the shipment cycle. This package is designed to safely dissipate the typical 4,500 thermal watts produced in the largest Radioisotope Thermoelectric Generators. The package also contains provisions to ensure leak tightness when radioactive materials, such as a Radioisotope Thermoelectric Generator for the Cassini Mission, planned for 1997 by the National Aeronautics and Space Administration, are being prepared for shipment. (Abstract Truncated)

Goldmann, L.H.; Averette, H.S. [Westinghouse Hanford Company, P.O. Box 1970, M/S R3-86 or N1-32, Richland, Washington 99352 (United States)

1995-01-20T23:59:59.000Z

22

Base Technology for Radioactive Material Transportation Packaging Systems  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

To establish Department of Energy (DOE) policies and responsibilities for coordinating and planning base technology for radioactive material transportation packaging systems.

1992-07-08T23:59:59.000Z

23

Testing of Packaging Materials for Improved PV Module Reliability  

SciTech Connect (OSTI)

A number of candidate alternative encapsulant and soft backsheet materials have been evaluated in terms of their suitability for photovoltaic (PV) module packaging applications. Relevant properties, including interfacial adhesion and moisture transport, have been measured as a function of damp-heat (85 C/85% relative humidity) exposure. Based on these tests, promising new encapsulants with improved properties have been identified. Backsheets prepared by industry and at NREL have been found to provide varying levels of moisture ingress protection. To achieve significantly improved products, further development of these candidates is ongoing. The relative effectiveness of various packaging strategies to protect PV devices has also been investigated.

Jorgensen, G. J.; Terwilliger, K. M.; Kempe, M. D.; McMahon, T. J.

2005-02-01T23:59:59.000Z

24

Test plan/procedure for the shock limiting device of the radioisotope thermoelectric generator package mounting subsystem 145. Revision 1  

SciTech Connect (OSTI)

This document defines the procedure to be used in the 18 inch drop test to be used for design verification of the RTG Transportation System Package Mounting.

Satoh, J.A.

1995-05-25T23:59:59.000Z

25

Hazardous Material Packaging for Transport - Administrative Procedures  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

To establ1sh administrative procedures for the certification and use of radioactive and other hazardous materials packaging by the Department of Energy (DOE).

1986-09-30T23:59:59.000Z

26

Safety analysis report for packaging (onsite) sample pig transport system  

SciTech Connect (OSTI)

This Safety Analysis Report for Packaging (SARP) provides a technical evaluation of the Sample Pig Transport System as compared to the requirements of the U.S. Department of Energy, Richland Operations Office (RL) Order 5480.1, Change 1, Chapter III. The evaluation concludes that the package is acceptable for the onsite transport of Type B, fissile excepted radioactive materials when used in accordance with this document.

MCCOY, J.C.

1999-03-16T23:59:59.000Z

27

Packaging and Transportation Support at LANL CTMA 2012  

SciTech Connect (OSTI)

Operations Support Packaging and Transportation (OS-PT) supports LANL in various functions. Some highlights of the past year have been with the work relating to environmental remediation, type B packaging, non-DOT compliant transfers, and special permit training. The TA-21 remediation project was part of the ARRA funding that LANL received. The $212 million in funding was used to demolish 24 buildings at TA-21, excavate the lab's oldest waste disposal site, and install 16 groundwater monitoring wells. The project was completed ahead of schedule and under budget. More than 300 tons of metal was recycled and all the soil excavated from MDA-B was replaced with clean fill. OS-PT supported this projected by transporting more than 7 million pounds of waste to TA-54 Area G with an addendum to their TSD. Because of the public access on the transfer route, Los Alamos County restricted the transfer to happen from 2:00 AM to 4:00 AM. OS-PT conducted 8 transfers in support of this project. Some concerns included the contaminated trailers at receipt facilities when transferring filled Super Sacks. Future Super Sacks were over packed into new IP-2 Super Sacks before shipping. OS-PT is also supporting the remediation of TA-54 Area G. LANL has an agreement with the State of New Mexico to remove all TRU waste currently stored above ground from at Area G. OS-PT supports this initiative with transfers of TRU waste under LANL's TSD and support of TRU shipments to WIPP. Another project supported by our organization is gas cylinder/dewar recycling and remediation. We are focusing on reducing risk associated with unneeded gasses at LANL. To minimized excessive ordering, to save money and time, and to minimize hazards OS-PT is supporting a gas recycling program. This program will allow programmatic organization across LANL to share unused/unneeded gasses. Instead of old dewars being disposed of, OS-PT has began identifying these dewars and sending them for refurbishment. To date, this effort has saved LANL $450K and estimated saving for future efforts will be more than $1.5 million. Some Projects that are happening here at LANL are offsite source recovery, weapon component transfers, and isotope science production. There are specific packages that help support these projects for the shipment of related materials. OS-PT provides support to these packages to ensure they are and will be available to continue this support. The Areva 435-B Overpack will help the Offsite Source Recovery Project recover high activity gamma sources from various locations across the globe. The Safety Analysis for Packaging is scheduled for initial completion June of 2012. The DPP-1 package is designed to replace the Model FL, which was designed by Rocky Flats and began service in 1990. LANL has collaborated on package design with LLNL, Pantex, Y-12, and KCP. LANL is supporting LLNL on component fixture development. Testing to 10 CFR 71 is to be completed in the Fall of 2012 and scheduled for NA-174 approval in 2014. The SAFESHIELD package helps supports LANL's Isotope production projects. This package can transfer highly irradiated materials from LANL's accelerator to material processing facilities. LANL worked to renew the SAFESHEILD's Certification for 5 more years.

Salazar, Nick [Los Alamos National Laboratory

2012-06-08T23:59:59.000Z

28

ISSUES ASSOCIATED WITH SAFE PACKAGING AND TRANSPORT OF NANOPARTICLES  

SciTech Connect (OSTI)

Nanoparticles have long been recognized a hazardous substances by personnel working in the field. They are not, however, listed as a separate, distinct category of dangerous goods at present. As dangerous goods or hazardous substances, they require packaging and transportation practices which parallel the established practices for hazardous materials transport. Pending establishment of a distinct category for such materials by the Department of Transportation, existing consensus or industrial protocols must be followed. Action by DOT to establish appropriate packaging and transport requirements is recommended.

Gupta, N.; Smith, A.

2011-02-14T23:59:59.000Z

29

Order Module--DOE O 460.1C, PACKAGING AND TRANSPORTATION SAFETY...  

Office of Environmental Management (EM)

DOE O 460.1C, PACKAGING AND TRANSPORTATION SAFETY, DOE O 460.2A, DEPARTMENTAL MATERIALS TRANSPORTATION AND PACKAGING MANAGEMENT Order Module--DOE O 460.1C, PACKAGING AND...

30

DEVELOPMENT OF THE HS99 AIR TRANSPORT TYPE A FISSILE PACKAGE  

SciTech Connect (OSTI)

An air-transport Type A Fissile radioactive shipping package for the transport of special form uranium sources has been developed by the Savannah River National Laboratory (SRNL) for the Department of Homeland Security. The Package model number is HS99 for Homeland Security Model 99. This paper presents the major design features of the HS99 and highlights engineered materials necessary for meeting the design requirements for this light-weight Type AF packaging. A discussion is provided demonstrating how the HS99 complies with the regulatory safety requirements of the Nuclear Regulatory Commission. The paper summarizes the results of structural testing to specified in 10 CFR 71 for Normal Conditions of Transport and Hypothetical Accident Conditions events. Planned and proposed future missions for this packaging are also addressed.

Blanton, P.; Eberl, K.

2012-07-10T23:59:59.000Z

31

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

SciTech Connect (OSTI)

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)

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

2013-07-01T23:59:59.000Z

32

ANNUAL MAINTENANCE AND LEAK TESTING FOR THE 9975 SHIPPING PACKAGE  

SciTech Connect (OSTI)

The purpose of this document is to provide step-by-step instructions for the annual helium leak test certification and maintenance of the 9975 Shipping Package.

Trapp, D.

2014-08-25T23:59:59.000Z

33

LEVERAGING AGING MATERIALS DATA TO SUPPORT EXTENSION OF TRANSPORTATION SHIPPING PACKAGES SERVICE LIFE  

SciTech Connect (OSTI)

Nuclear material inventories are increasingly being transferred to interim storage locations where they may reside for extended periods of time. Use of a shipping package to store nuclear materials after the transfer has become more common for a variety of reasons. Shipping packages are robust and have a qualified pedigree for performance in normal operation and accident conditions but are only certified over an approved transportation window. The continued use of shipping packages to contain nuclear material during interim storage will result in reduced overall costs and reduced exposure to workers. However, the shipping package materials of construction must maintain integrity as specified by the safety basis of the storage facility throughout the storage period, which is typically well beyond the certified transportation window. In many ways, the certification processes required for interim storage of nuclear materials in shipping packages is similar to life extension programs required for dry cask storage systems for commercial nuclear fuels. The storage of spent nuclear fuel in dry cask storage systems is federally-regulated, and over 1500 individual dry casks have been in successful service up to 20 years in the US. The uncertainty in final disposition will likely require extended storage of this fuel well beyond initial license periods and perhaps multiple re-licenses may be needed. Thus, both the shipping packages and the dry cask storage systems require materials integrity assessments and assurance of continued satisfactory materials performance over times not considered in the original evaluation processes. Test programs for the shipping packages have been established to obtain aging data on materials of construction to demonstrate continued system integrity. The collective data may be coupled with similar data for the dry cask storage systems and used to support extending the service life of shipping packages in both transportation and storage.

Dunn, K. [Savannah River National Laboratory; Bellamy, S. [Savannah River National Laboratory; Daugherty, W. [Savannah River National Laboratory; Sindelar, R. [Savannah River National Laboratory; Skidmore, E. [Savannah River National Laboratory

2013-08-18T23:59:59.000Z

34

Radioisotope Thermoelectric Generator Transporation System licensed hardware second certification test series and package shock mount system test  

SciTech Connect (OSTI)

This paper presents a summary of two separate drop test a e performed in support of the Radioisotope Thermoelectric Generator (RTG) Transportation System (RTGTS). The first portion of this paper presents the second series of drop testing required to demonstrate that the RTG package design meets the requirements of Title 10, Code of Federal Regulations, ``Part 71`` (10 CFR 71). Results of the first test series, performed in July 1994, demonstrated that some design changes were necessary. The package design was modified to improve test performance and the design changes were incorporated into the Safety Analysis Report for Packaging (SARP). The second full-size certification test article (CTA-2) incorporated the modified design and was tested at the US Department of Energy`s (DOE) Hanford Site near Richland, Washington. With the successful completion of the test series, and pending DOE Office of Facility Safety Analysis approval of the SARP, a certificate of compliance will be issued for the RTG package allowing its use. The second portion of this paper presents the design and testing of the RTG Package Mount System. The RTG package mount was designed to protect the RTG from excessive vibration during transport, provide shock protection during on/off loading, and provide a mechanism for moving the RTG package with a forklift. Military Standard (MIL-STD) 810E, Transit Drop Procedure (DOE 1989), was used to verify that the shock limiting system limited accelerations in excess of 15 G`s at frequencies below 150 Hz. Results of the package mount drop tests indicate that an impact force of 15 G`s was not exceeded in any test from a free drop height of 457 mm (18 in.).

Ferrell, P.C.; Moody, D.A.

1995-10-01T23:59:59.000Z

35

Regulatory compliance in the design of packages used to transport radioactive materials  

SciTech Connect (OSTI)

Shipments of radioactive materials within the regulatory jurisdiction of the US Department of Energy (DOE) must meet the package design requirements contained in Title 10 of the Code of Federal Regulations, Part 71, and DOE Order 5480.3. These regulations do not provide design criteria requirements, but only detail the approval standards, structural performance criteria, and package integrity requirements that must be met during transport. The DOE recommended design criterion for high-level Category I radioactive packagings is Section III, Division 1, of the ASME Boiler and Pressure Vessel Code. However, alternative design criteria may be used if all the design requirements are satisfied. The purpose of this paper is to review alternatives to the Code criteria and discuss their applicability to the design of containment vessels in packages for high-level radioactive materials. Issues such as design qualification by physical testing, the use of scale models, and problems encountered using a non-ASME design approach are addressed.

Raske, D.T.

1993-06-01T23:59:59.000Z

36

DOT-7A Type A packaging test and evaluation procedure  

SciTech Connect (OSTI)

The purpose of this document is to provide guidance for qualifying a DOT-7A Type A packaging for use. WHC qualifies DOT-7A packaging for two purposes. The first is to provide packages for use by WHC (manufacturer-qualified). The second is to provide a contracted service in support of DOE/EM-76 (DOE-qualified). This document includes descriptions of the performance tests, the personnel involved and their qualifications, appropriate safety and quality assurance considerations, and the procedures to be followed when WHC performs the tests (either as the manufacturer, or on behalf of the DOE`s certification program).

Kelly, D.L., Westinghouse Hanford

1996-06-13T23:59:59.000Z

37

Radioisotope thermoelectric generator transportation system safety analysis report for packaging. Volumes 1 and 2  

SciTech Connect (OSTI)

This SARP describes the RTG Transportation System Package, a Type B(U) packaging system that is used to transport an RTG or similar payload. The payload, which is included in this SARP, is a generic, enveloping payload that specifically encompasses the General Purpose Heat Source (GPHS) RTG payload. The package consists of two independent containment systems mounted on a shock isolation transport skid and transported within an exclusive-use trailer.

Ferrell, P.C.

1996-04-18T23:59:59.000Z

38

2014-03-06 Issuance: Test Procedures for Packaged Terminal Air...  

Broader source: Energy.gov (indexed) [DOE]

Issuance: Test Procedures for Packaged Terminal Air Conditioners and Packaged Terminal Heat Pumps; Notice of Proposed Rulemaking 2014-03-06 Issuance: Test Procedures for...

39

Packaging, Transportation and Recycling of NPP Condenser Modules - 12262  

SciTech Connect (OSTI)

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 facility. If any of these issues were not adequately resolved prior to the start of the outage, costly delays would result and the re-start of the power plant could be impacted. The main focus of this project was to find successful methods for keeping this material out of the landfills and preserving the natural resources. In addition, this operation provided a significant cost savings to the public utility by minimizing landfill disposal. The onsite portion of the project has been completed without impact to the overall outage schedule. By the date of presentation, the majority of the waste from the condenser replacement project will have been processed and recycled. The goals for this project included helping Energy Northwest maintain the outage schedule, package and characterize waste compliantly, perform transportation activities in compliance with 49CFR (Ref-1), and minimize the waste disposal volume. During this condenser replacement project, over three millions pounds of waste was generated, packaged, characterized and transported without injury or incident. It is anticipated that 95% of the waste generated during this project will not require landfill disposal. All of the waste is scheduled to be processed, decontaminated and recycled by June of 2012. (authors)

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

2012-07-01T23:59:59.000Z

40

Test and evaluation document for DOT Specification 7A Type A Packaging. Revision 3  

SciTech Connect (OSTI)

The US Department of Energy (DOE) has been conducting, through several of its operating contractors, an evaluation and testing program to qualify Type A radioactive material packagings per US Department of Transportation (DOT) Specification 7A (DOT-7A) of the Code of Federal Regulations (CFR), Title 49, Part 178 (49 CFR 178). The program is currently administered by the DOE, Office of Facility Safety Analysis, DOE/EH-32, at DOE-Headquarters (DOE-HQ) in Germantown, Maryland. This document summarizes the evaluation and testing performed for all of the packagings successfully qualified in this program.

NONE

1996-01-30T23:59:59.000Z

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

Drop Tests of the Closure Ring for the 9975 Package  

SciTech Connect (OSTI)

The drop tests of the closure ring for 9975 packages, described here, were performed to answer questions raised by the regulatory authority as a result of deformation of the closure ring and drum rim observed during drop tests conducted in September 1998.

Smith, A.C

1999-09-29T23:59:59.000Z

42

AXIAL SOURCE PROFILE EFFECT ON WASTE PACKAGE TRANSPORTER SHIELDING  

SciTech Connect (OSTI)

The purpose of this scoping calculation is to support preliminary design of the Waste Package (WP) transporter radiation shield configuration. Spent Nuclear Fuel (SNF) is highly radioactive and site personnel must be protected during the period that the WPs are emplaced. Personnel protection is accomplished via a heavily shielded WP transporter that moves the waste from the surface to the emplacement drift. All previous WP transporter shielding calculations have assumed a Design Basis Fuel (DBF) in which the fuel burnup is uniform (e.g. Ref. 7.3, Ref. 7.4, and Ref. 7.12). In reality, SNF burnup varies significantly from one end of the fuel assembly to the other. Since source strengths are dependent upon fuel burnup, a model which varies the fuel burnup along the assembly axis will produce a more accurate depiction of the radiation field surrounding the WP transporter. The objective of this calculation is to determine the need for using the actual axial profile, as opposed to the uniform burnup assumption, in the WP transporter shield design. The scope of the calculation is as follows: (1) Determine the impact of axial source term variation on WP transporter contact dose rates. (2) Determine appropriate shielding modifications to account for expected dose rate peaking effects. Consistent with the previous subsurface shielding analyses, this calculation considers the bounding 21 Pressurized Water Reactor (PWR) WP only. The calculation will need to be revised and extended to Boiling Water Reactor (BWR) SNF upon selection of the WP design for the License Application (LA) and availability of the source terms from the WP Operations Group.

A. Nielsen

1999-06-30T23:59:59.000Z

43

DIGITAL RADIOGRAPHY OF SPECIAL NUCLEAR MATERIAL TEST PACKAGES  

SciTech Connect (OSTI)

The purpose of this document is to provide a brief introduction to digital radiography (DR), and a description of the DR configuration that was used to radiographically image the Special Nuclear Material (SNM) Test Packages before and after function tests that have been conducted. Also included are (1) Attachment 1, a comprehensive index that describes at which phase of the certification process that digital radiographic images were acquired, (2) digital radiographic images of each of the six packages at various stages of the certification process, and (3) Attachment 2, imaging instructions, that specify the setup procedures and detailed parameters of the DR imaging methodology that were used.

HOWARD, BOYD

2006-02-02T23:59:59.000Z

44

Savannah River Site Eastern Transportation Hub: A Concept For a DOE Eastern Packaging, Staging and Maintenance Center - 13143  

SciTech Connect (OSTI)

The Department of Energy (DOE) is working to de-inventory sites and consolidate hazardous materials for processing and disposal. The DOE administers a wide range of certified shipping packages for the transport of hazardous materials to include Special Nuclear Material (SNM), radioactive materials, sealed sources and radioactive wastes. A critical element to successful and safe transportation of these materials is the availability of certified shipping packages. There are over seven thousand certified packagings (i.e., Type B/Type AF) utilized within the DOE for current missions. The synergistic effects of consolidated maintenance, refurbishment, testing, certification, and costing of these services would allow for efficient management of the packagings inventory and to support anticipated future in-commerce shipping needs. The Savannah River Site (SRS) receives and ships radioactive materials (including SNM) and waste on a regular basis for critical missions such as consolidated storage, stabilization, purification, or disposition using H-Canyon and HB-Line. The Savannah River National Laboratory (SRNL) has the technical capability and equipment for all aspects of packaging management. SRS has the only active material processing facility in the DOE complex and is one of the sites of choice for nuclear material consolidation. SRS is a logical location to perform maintenance and periodic testing of the DOE fleet of certified packagings. This initiative envisions a DOE Eastern Packaging Staging and Maintenance Center (PSMC) at the SRS and a western hub at the Nevada National Security Site (NNSS), an active DOE Regional Disposal Site. The PSMC's would be the first place DOE would go to meet their radioactive packaging needs and the primary locations projects would go to disposition excess packaging for beneficial reuse. These two hubs would provide the centralized management of a packaging fleet rather than the current approach to design, procure, maintain and dispose of packagings on a project-by-project basis. This initiative provides significant savings in packaging costs and acceleration of project schedules. In addition to certified packaging, the PSMC would be well suited for select designs of 7A Type A packaging and Industrial Packaging. (authors)

England, Jeffery L. [Savannah River National Laboratory, Aiken, South Carolina (United States)] [Savannah River National Laboratory, Aiken, South Carolina (United States); Adams, Karen; Maxted, Maxcine; Ruff Jr, Clarence [U.S. Department of Energy, Savannah River Site, Aiken, SC (United States)] [U.S. Department of Energy, Savannah River Site, Aiken, SC (United States); Albenesius, Andrew; Bowers, Mark D.; Fountain, Geoffrey; Hughes, Michael [Savannah River Nuclear Solutions, Aiken, SC (United States)] [Savannah River Nuclear Solutions, Aiken, SC (United States); Gordon, Sydney [National Security Technologies, LLC, Las Vegas, NV (United States)] [National Security Technologies, LLC, Las Vegas, NV (United States); O'Connor, Stephen [U.S. Department of Energy, HQ DOE, EM-33, Germantown MD (United States)] [U.S. Department of Energy, HQ DOE, EM-33, Germantown MD (United States)

2013-07-01T23:59:59.000Z

45

Used Fuel Testing Transportation Model  

SciTech Connect (OSTI)

This report identifies shipping packages/casks that might be used by the Used Nuclear Fuel Disposition Campaign Program (UFDC) to ship fuel rods and pieces of fuel rods taken from high-burnup used nuclear fuel (UNF) assemblies to and between research facilities for purposes of evaluation and testing. Also identified are the actions that would need to be taken, if any, to obtain U.S. Nuclear Regulatory (NRC) or other regulatory authority approval to use each of the packages and/or shipping casks for this purpose.

Ross, Steven B.; Best, Ralph E.; Maheras, Steven J.; Jensen, Philip J.; England, Jeffery L.; LeDuc, Dan

2014-09-24T23:59:59.000Z

46

Retractable pin dual in-line package test clip  

DOE Patents [OSTI]

This invention is a Dual In-Line Package (DIP) test clip for use when troubleshooting circuits containing DIP integrated circuits. This test clip is a significant improvement over existing DIP test clips in that it has retractable pins which will permit troubleshooting without risk of accidentally shorting adjacent pins together when moving probes to different pins on energized circuits or when the probe is accidentally bumped while taking measurements.

Bandzuch, Gregory S. (Washington, PA); Kosslow, William J. (Jefferson Boro, PA)

1996-01-01T23:59:59.000Z

47

Physical test report for drop test of a 9974 radioactive material shipping packaging  

SciTech Connect (OSTI)

This report presents the drop test results for the 9974 radioactive material shipping package being dropped onto 6-inch diameter, 40-inch long puncture pin. Also reported are the drop test resuls for a 30-foot impact that failed the drum confinement boundary. The purpose of these drops was to show that the package lid would remain attached to the drum.

Blanton, P.S. [Westinghouse Savannah River Company, AIKEN, SC (United States)

1997-10-01T23:59:59.000Z

48

Type B package for the transport of large medical and industrial sources  

SciTech Connect (OSTI)

AREVA Federal Services LLC, under contract to the Los Alamos National Laboratory's Offsite Source Recovery Project, is developing a new Type B(U)-96 package for the transport of unwanted or abandoned high activity gamma and neutron radioactive sealed sources (sources). The sources were used primarily in medical or industrial devices, and are of domestic (USA) or foreign origin. To promote public safety and mitigate the possibility of loss or misuse, the Offsite Source Recovery Project is recovering and managing sources worldwide. The package, denoted the LANL-B, is designed to accommodate the sources within an internal gamma shield. The sources are located either in the IAEA's Long Term Storage Shield (LTSS), or within intact medical or industrial irradiation devices. As the sources are already shielded separately, the package does not include any shielding of its own. A particular challenge in the design of the LANL-B has been weight. Since the LTSS shield weighs approximately 5,000 lb [2,270 kg], and the total package gross weight must be limited to 10,000 lb [4,540 kg], the net weight of the package was limited to 5,000 lb, for an efficiency of 50% (i.e., the payload weight is 50% of the gross weight of the package). This required implementation of a light-weight bell-jar concept, in which the containment takes the form of a vertical bell which is bolted to a base. A single impact limiter is used on the bottom, to protect the elastomer seals and bolted joint. A top-end impact is mitigated by the deformation of a tori spherically-shaped head. Impacts in various orientations on the bottom end are mitigated by a cylindrical, polyurethane foam-filled impact limiter. Internally, energy is absorbed using honeycomb blocks at each end, which fill the torispherical head volumes. As many of the sources are considered to be in normal form, the LANL-B package offers leak-tight containment using an elastomer seal at the joint between the bell and the base, as well as on the single vent port. Leak testing prior to transport may be either using helium mass spectrometry or the pressure-rise concept.

Brown, Darrell Dwaine [Los Alamos National Laboratory; Noss, Philip W [AREVA FEDERAL SERVICES

2010-09-14T23:59:59.000Z

49

Accident Conditions versus Regulatory Test for NRC-Approved UF6 Packages  

SciTech Connect (OSTI)

The Nuclear Regulatory Commission (NRC) approves new package designs for shipping fissile quantities of UF{sub 6}. Currently there are three packages approved by the NRC for domestic shipments of fissile quantities of UF{sub 6}: NCI-21PF-1; UX-30; and ESP30X. For approval by the NRC, packages must be subjected to a sequence of physical tests to simulate transportation accident conditions as described in 10 CFR Part 71. The primary objective of this project was to relate the conditions experienced by these packages in the tests described in 10 CFR Part 71 to conditions potentially encountered in actual accidents and to estimate the probabilities of such accidents. Comparison of the effects of actual accident conditions to 10 CFR Part 71 tests was achieved by means of computer modeling of structural effects on the packages due to impacts with actual surfaces, and thermal effects resulting from test and other fire scenarios. In addition, the likelihood of encountering bodies of water or sufficient rainfall to cause complete or partial immersion during transport over representative truck routes was assessed. Modeled effects, and their associated probabilities, were combined with existing event-tree data, plus accident rates and other characteristics gathered from representative routes, to derive generalized probabilities of encountering accident conditions comparable to the 10 CFR Part 71 conditions. This analysis suggests that the regulatory conditions are unlikely to be exceeded in real accidents, i.e. the likelihood of UF{sub 6} being dispersed as a result of accident impact or fire is small. Moreover, given that an accident has occurred, exposure to water by fire-fighting, heavy rain or submersion in a body of water is even less probable by factors ranging from 0.5 to 8E-6.

MILLS, G. SCOTT; AMMERMAN, DOUGLAS J.; LOPEZ, CARLOS

2003-01-01T23:59:59.000Z

50

Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask  

SciTech Connect (OSTI)

This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required.

Romano, T.

1997-09-29T23:59:59.000Z

51

Classification of transportation packaging and dry spent fuel storage system components according to importance to safety  

SciTech Connect (OSTI)

This report provides a graded approach for classification of components used in transportation packaging and dry spent fuel storage systems. This approach provides a method for identifying, the classification of components according to importance to safety within transportation packagings and dry spent fuel storage systems. Record retention requirements are discussed to identify the documentation necessary to validate that the individual components were fabricated in accordance with their assigned classification. A review of the existing regulations pertaining to transportation packagings and dry storage systems was performed to identify current requirements The general types of transportation packagings and dry storage systems were identified. Discussions were held with suppliers and fabricators of packagings and storage systems to determine current practices. The methodology used in this report is based on Regulatory Guide 7.10, Establishing Quality Assurance Programs for Packaging Used in the Transport of Radioactive Material. This report also includes a list of generic components for each of the general types of transportation packagings and spent fuel storage systems. The safety importance of each component is discussed, and a classification category is assigned.

McConnell, J.W., Jr; Ayers, A.L. Jr; Tyacke, M.J. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

1996-02-01T23:59:59.000Z

52

Packaging and transportation system for K-Basin spent fuel  

SciTech Connect (OSTI)

This paper describes the cask/transportation system that was designed, procured and delivered to the Hanford K-Basin site at Richland, Washington. The performance requirements and design of the various components -- cask, trailer with cask tie-down system, and the cask operation equipment for the load-out pit -- will be discussed. The presentation will include the details of the factory acceptance testing and its results. The performance requirements for the cask/transportation system was dictated by the constraints imposed by the large number of high priority shipments and the spent fuel pool environment, and the complex interface requirements with other equipment and facility designs. The results of the testing form the basis for the conclusion that the system satisfies the site performance requirements. The cask/transportation system design was driven by the existing facility constraints and the limitations imposed by the large number of shipments over a short two-year period. This system may be useful information for other DOE facilities that may be or will be in a similar situation.

Kee, A.T.

1998-03-03T23:59:59.000Z

53

Implementation Guide for Use with DOE O 460.2 Departmental Materials Transportation and Packaging Management  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The purpose of this guide is to assist those responsible for transporting and packaging Department materials, and to provide an understanding of Department policies on activities which supplement regulatory requirements.

1996-11-15T23:59:59.000Z

54

Operating Experience and Lessons Learned in the Use of Soft-Sided Packaging for Transportation and Disposal of Low Activity Radioactive Waste  

SciTech Connect (OSTI)

This paper describes the operating experience and lessons learned at U.S. Department of Energy (DOE) sites as a result of an evaluation of potential trailer contamination and soft-sided packaging integrity issues related to the disposal of low-level and mixed low-level (LLW/MLLW) radioactive waste shipments. Nearly 4.3 million cubic meters of LLW/MLLW will have been generated and disposed of during fiscal year (FY) 2010 to FY 2015—either at commercial disposal sites or disposal sites owned by DOE. The LLW/MLLW is packaged in several different types of regulatory compliant packaging and transported via highway or rail to disposal sites safely and efficiently in accordance with federal, state, and local regulations and DOE orders. In 1999, DOE supported the development of LLW containers that are more volumetrically efficient, more cost effective, and easier to use as compared to metal or wooden containers that existed at that time. The DOE Idaho National Engineering and Environmental Laboratory (INEEL), working in conjunction with the plastic industry, tested several types of soft-sided waste packaging systems that meet U.S. Department of Transportation requirements for transport of low specific activity and surface contaminated objects. Since then, soft-sided packaging of various capacities have been used successfully by the decontamination and decommissioning (D&D) projects to package, transport, and dispose D&D wastes throughout the DOE complex. The joint team of experts assembled by the Energy Facility Contractors Group from DOE waste generating sites, DOE and commercial waste disposal facilities, and soft-sided packaging suppliers conducted the review of soft-sided packaging operations and transportation of these packages to the disposal sites. As a result of this evaluation, the team developed several recommendations and best practices to prevent or minimize the recurrences of equipment contamination issues and proper use of soft-sided packaging for transport and disposal of waste.

Kapoor, A. [DOE; Gordon, S. [NSTec; Goldston, W. [Energy Solutions

2013-07-08T23:59:59.000Z

55

MASSCLEAN - MASSive CLuster Evolution and ANalysis Package - Description and Tests  

E-Print Network [OSTI]

We present MASSCLEAN, a new, sophisticated and robust stellar cluster image and photometry simulation package. This visualization tool is able to create color-magnitude diagrams and standard FITS images in any of the traditional optical and near-infrared bands based on cluster characteristics input by the user, including but not limited to distance, age, mass, radius and extinction. At the limit of very distant, unresolved clusters, we have checked the integrated colors created in MASSCLEAN against those from other simple stellar population models with consistent results. We have also tested models which provide a reasonable estimate of the field star contamination in images and color-magnitude diagrams. We demonstrate the package by simulating images and color-magnitude diagrams of well known massive Milky Way clusters and compare their appearance to real data. Because the algorithm populates the cluster with a discrete number of tenable stars, it can be used as part of a Monte Carlo Method to derive the pr obabilistic range of characteristics (integrated colors, for example) consistent with a given cluster mass and age. Our simulation package is available for download and will run on any standard desktop running UNIX/Linux. Full documentation on installation and its use is also available. Finally, a web-based version of MASSCLEAN which can be immediately used and is sufficiently adaptable for most applications is available through a web interface.

Bogdan Popescu; M. M. Hanson

2009-05-28T23:59:59.000Z

56

Lessons Learned from Three Mile Island Packaging, Transportation and Disposition that Apply to Fukushima Daiichi Recovery  

SciTech Connect (OSTI)

Following the massive earthquake and resulting tsunami damage in March of 2011 at the Fukushima Daiichi nuclear power plant in Japan, interest was amplified for what was done for recovery at the Three Mile Island Unit 2 (TMI-2) in the United States following its meltdown in 1979. Many parallels could be drawn between to two accidents. This paper presents the results of research done into the TMI-2 recovery effort and its applicability to the Fukushima Daiichi cleanup. This research focused on three topics: packaging, transportation, and disposition. This research work was performed as a collaboration between Japan’s Central Research Institute of Electric Power Industry (CRIEPI) and the Idaho National Laboratory (INL). Hundreds of TMI-2 related documents were searched and pertinent information was gleaned from these documents. Other important information was also obtained by interviewing employees who were involved first hand in various aspects of the TMI-2 cleanup effort. This paper is organized into three main sections: (1) Transport from Three Mile Island to Central Facilities Area at INL, (2) Transport from INL Central Receiving Facility to INL Test Area North (TAN) and wet storage at TAN, and (3) Transport from TAN to INL Idaho Nuclear Technology and Engineering Center (INTEC) and Dry Storage at INTEC. Within each of these sections, lessons learned from performing recovery activities are presented and their applicability to the Fukushima Daiichi nuclear power plant cleanup are outlined.

Layne Pincock; Wendell Hintze; Dr. Koji Shirai

2012-07-01T23:59:59.000Z

57

Physical test report to drop test of a 9975 radioactive material shipping packaging  

SciTech Connect (OSTI)

This report presents the drop test results for the 9975 radioactive material shipping package being dropped 30 feet onto a unyielding surface followed by a 40-inch puncture pin drop. The purpose of these drops was to show that the package lid would remain attached to the drum. The 30-foot drop was designed to weaken the lid closure lug while still maintaining maximum extension of the lugs from the drum surface. This was accomplished by angling the drum approximately 30 degrees from horizontal in an inverted position. In this position, the drum was rotated slightly so as not to embed the closure lugs into the drum as a result of the 30-foot drop. It was determined that this orientation would maximize deformation to the closure ring around the closure lug while still maintaining the extension of the lugs from the package surface. The second drop was from 40 inches above a 40-inch tall 6-inch diameter puncture pin. The package was angled 10 degrees from vertical and aligned over the puncture pin to solidly hit the drum lug(s) in an attempt to disengage the lid when dropped.Tests were performed in response to DOE EM-76 review Q5 inquires that questioned the capability of the 9975 drum lid to remain in place under this test sequence. Two packages were dropped utilizing this sequence, a 9974 and 9975. Test results for the 9974 package are reported in WSRC-RP-97-00945. A series of 40-inch puncture pin tests were also performed on undamaged 9975 and 9974 packages.

Blanton, P.S.

1997-11-11T23:59:59.000Z

58

Packaging and Transfer or Transportation of Materials of National Security Interest  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

To establish requirements and responsibilities for the Transportation Safeguards System (TSS) packaging and transportation and onsite transfer of nuclear explosives, nuclear components, Naval nuclear fuel elements, Category I and Category II special nuclear materials, special assemblies, and other materials of national security interest. Cancels: DOE 5610.12 and DOE 5610.14.

2000-09-29T23:59:59.000Z

59

Implementation Guide for Use with DOE O 460.1A, Packaging and Transportation Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This Guide provides information concerning the use of current principles and practices, including regulatory guidance from the U. S. Department of Transportation and the U. S. Nuclear Regulatory Commission, where available, to establish and implement effective packaging and transportation safety programs.

1997-06-05T23:59:59.000Z

60

t -software package for numerical simulations of radioactive contaminant transport in groundwater  

E-Print Network [OSTI]

r3 t - software package for numerical simulations of radioactive contaminant transport equations that arise from the modelling of radioactive contaminant transport in porous media. It can solve, see [6]) can help to numerically simulate the spreading of radioactive contaminants in flowing ground

Frolkovic, Peter

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Work plan for the fabrication of the radioisotope thermoelectric generator transportation system package mounting  

SciTech Connect (OSTI)

The Radioisotope Thermoelectric Generator (RTG) has available a dedicated system for the transportation of RTG payloads. The RTG Transportation System (System 100) is comprised of four systems; the Package (System 120), the Semi-trailer (System 140), the Gas Management (System 160), and the Facility Transport (System 180). This document provides guidelines on the fabrication, technical requirements, and quality assurance of the Package Mounting (Subsystem 145), part of System 140. The description follows the Development Control Requirements of WHC-CM-6-1, EP 2.4, Rev. 3.

Satoh, J.A.

1994-11-09T23:59:59.000Z

62

Pressure Build-Up During the Fire Test in Type B(U) Packages Containing Water - 13280  

SciTech Connect (OSTI)

The safety assessment of packages for the transport of radioactive materials with content containing liquids requires special consideration. The main focus is on water as supplementary liquid content in Type B(U) packages. A typical content of a Type B(U) package is ion exchange resin, waste of a nuclear power plant, which is not dried, normally only drained. Besides the saturated ion exchange resin, a small amount of free water can be included in these contents. Compared to the safety assessment of packages with dry content, attention must be paid to some more specific issues. An overview of these issues is provided. The physical and chemical compatibility of the content itself and the content compatibility with the packages materials must be demonstrated for the assessment. Regarding the mechanical resistance the package has to withstand the forces resulting from the freezing liquid. The most interesting point, however, is the pressure build-up inside the package due to vaporization. This could for example be caused by radiolysis of the liquid and must be taken into account for the storage period. If the package is stressed by the total inner pressure, this pressure leads to mechanical loads to the package body, the lid and the lid bolts. Thus, the pressure is the driving force on the gasket system regarding the activity release and a possible loss of tightness. The total pressure in any calculation is the sum of partial pressures of different gases which can be caused by different effects. The pressure build-up inside the package caused by the regulatory thermal test (30 min at 800 deg. C), as part of the cumulative test scenario under accident conditions of transport is discussed primarily. To determine the pressure, the temperature distribution in the content must be calculated for the whole period from beginning of the thermal test until cooling-down. In this case, while calculating the temperature distribution, conduction and radiation as well as evaporation and condensation during the associated process of transport have to be considered. This paper discusses limiting amounts of water inside the cask which could lead to unacceptable pressure and takes into account saturated steam as well as overheated steam. However, the difficulties of assessing casks containing wet content will be discussed. From the authority assessment point of view, drying of the content could be an effective way to avoid the above described pressure build-up and the associated difficulties for the safety assessment. (authors)

Feldkamp, Martin; Nehrig, Marko; Bletzer, Claus; Wille, Frank [BAM Federal Institute for Materials Research and Testing, Unter den Eichen 44, 12205 Berlin (Germany)] [BAM Federal Institute for Materials Research and Testing, Unter den Eichen 44, 12205 Berlin (Germany)

2013-07-01T23:59:59.000Z

63

An issue paper on the use of hydrogen getters in transportation packaging  

SciTech Connect (OSTI)

The accumulation of hydrogen is usually an undesirable occurrence because buildup in sealed systems pose explosion hazards under certain conditions. Hydrogen scavengers, or getters, can avert these problems by removing hydrogen from such environments. This paper provides a review of a number of reversible and irreversible getters that potentially could be used to reduce the buildup of hydrogen gas in containers for the transport of radioactive materials. In addition to describing getters that have already been used for such purposes, novel getters that might find application in future transport packages are also discussed. This paper also discusses getter material poisoning, the use of getters in packaging, the effects of radiation on getters, the compatibility of getters with packaging, design considerations, regulatory precedents, and makes general recommendations for the materials that have the greatest applicability in transport packaging. At this time, the Pacific Northwest National Laboratory composite getter, DEB [1,4-(phenylethylene)benzene] or similar polymer-based getters, and a manganese dioxide-based getter appear to be attractive candidates that should be further evaluated. These getters potentially can help prevent pressurization from radiolytic reactions in transportation packaging.

NIGREY,PAUL J.

2000-02-01T23:59:59.000Z

64

2014-03-06 Issuance: Test Procedures for Packaged Terminal Air Conditioners and Packaged Terminal Heat Pumps; Notice of Proposed Rulemaking  

Broader source: Energy.gov [DOE]

This document is a pre-publication Federal Register notice of proposed rulemaking regarding test procedures for packaged terminal air conditioners and packaged terminal heat pumps, as issued by the Deputy Assistant Secretary on March 6, 2014.

65

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

SciTech Connect (OSTI)

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

Dyer, R.H. [USDOE Oak Ridge Operations Office, TN (United States); Kovac, F.M. [Oak Ridge National Lab., TN (United States); Pryor, W.A. [PAI Corp., Oak Ridge, TN (United States)

1993-10-01T23:59:59.000Z

66

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

SciTech Connect (OSTI)

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.

Smith, R.J.

1995-02-01T23:59:59.000Z

67

Hail Impact Testing on Crystalline Si Modules with Flexible Packaging  

Broader source: Energy.gov (indexed) [DOE]

% Semi-flexible packaging of silicon solar cells has potential applications in BIPV and consumer electronics. One of the more difficult reliability requirements for...

68

Spent Fuel Transportation Package Response to the Baltimore Tunnel Fire Scenario  

SciTech Connect (OSTI)

On July 18, 2001, a freight train carrying hazardous (non-nuclear) materials derailed and caught fire while passing through the Howard Street railroad tunnel in downtown Baltimore, Maryland. The United States Nuclear Regulatory Commission (USNRC), one of the agencies responsible for ensuring the safe transportation of radioactive materials in the United States, undertook an investigation of the train derailment and fire to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by railroad. Shortly after the accident occurred, the USNRC met with the National Transportation Safety Board (NTSB, the U.S. agency responsible for determining the cause of transportation accidents), to discuss the details of the accident and the ensuing fire. Following these discussions, the USNRC assembled a team of experts from the National Institute of Standards and Technology (NIST), the Center for Nuclear Waste Regulatory Analyses (CNWRA), and Pacific Northwest National Laboratory (PNNL) to determine the thermal conditions that existed in the Howard Street tunnel fire and analyze the effects of this fire on various spent fuel transportation package designs. The Fire Dynamics Simulator (FDS) code, developed by NIST, was used to determine the thermal environment present in the Howard Street tunnel during the fire. The FDS results were used as boundary conditions in the COBRA-SFS and ANSYS® computer codes to evaluate the thermal performance of different package designs. The staff concluded that larger transportation packages resembling the HOLTEC Model No. HI STAR 100 and TransNuclear Model No. TN-68 would withstand a fire with thermal conditions similar to those that existed in the Baltimore tunnel fire event with only minor damage to peripheral components. This is due to their sizable thermal inertia and design specifications in compliance with currently imposed regulatory requirements. The staff also concluded that some components of smaller transportation packages resembling the NAC Model No. LWT, despite placement within an ISO container, could degrade. USNRC staff evaluated the radiological consequences of the package responses to the Baltimore tunnel fire. Though components in some packages heated up beyond their service temperatures, the staff determined that there would be no significant dose as a result of the fire for any of these and similar packages.

Adkins, Harold E.; Cuta, Judith M.; Koeppel, Brian J.; Guzman, Anthony D.; Bajwa, Christopher S.

2006-11-15T23:59:59.000Z

69

Quality assurance guidance for TRUPACT-II (Transuranic Package Transporter-II) payload control  

SciTech Connect (OSTI)

The Transuranic Package Transporter-II (TRUPACT-II) Safety Analysis Report for Packaging (SARP) approved by the Nuclear Regulatory Commission (NRC), discusses authorized methods for payload control in Appendix 1.3.7 and the Quality Assurance (QA) requirements in Section 9.3. Subsection 9.3.2.1 covers maintenance and use of the TRUPACT-II and the specific QA requirements are given in DOE/WIPP 89-012. Subsection 9.3.2.2 covers payload compliance, for which this document was written. 6 refs.

Not Available

1989-10-01T23:59:59.000Z

70

Information management and collection for US DOE's packaging and transportation needs in the 90's  

SciTech Connect (OSTI)

The Transportation Assessment and Integration (TRAIN) Project (US DOE, 1992) was established to provide a systematic approach to identify the problems and needs that will affect the capability of the United States Department of Energy (US DOE) to provide itself with cost-effective, efficient, and coordinated transportation services during the 1990s. Eight issue areas were identified to be included in the TRAIN Project, with one principal investigator assigned to each. The eight areas are as follows: (1) Packaging and Transportation Needs (PATN) in the 1990s; (2) Institutional and Outreach Programs; (3) Regulatory Impacts on Transportation Management; (4) Traffic and Packaging Operations; (5) Research and Development Requirements; (6) Training Support; (7) Emergency Preparedness Requirements; and (8) US DOE-EM 561 Roles and Responsibilities. This paper focuses on the results of the PATN activity of TRAIN. The objective of PATN is to prepare the US DOE, in general, and US DOE-EM 561 (Environmental Restoration and Waste Management (EM), Office of Technology Development, Transportation) in particular, respond to the transportation needs of program elements in the Department. One of the first tasks in evaluating these needs was to formulate the potential for transportation of radioactive materials in the next decade.

Wheeler, T. A.; Luna, Robert E.; McClure, J. D. [Sandia National Labs., Albuquerque, NM (United States); Quinn, Geoffrey [Wastren, Inc., Germantown, MD (United States)

1991-01-01T23:59:59.000Z

71

Improvement of operational safety of dual-purpose transport packaging set for naval SNF in storage  

SciTech Connect (OSTI)

Available in abstract form only. Full text of publication follows: In recent ten years a new technology of management of irradiated nuclear fuel (SNF) at the final stage of fuel cycle has been intensely developing on a basis of a new type of casks used for interim storage of SNF and subsequent transportation therein to the place of processing, further storage or final disposal. This technology stems from the concept of a protective cask which provides preservation of its content (SNF) and fulfillment of all other safety requirements for storage and transportation of SNF. Radiation protection against emissions and non-distribution of activity outside the cask is ensured by physical barriers, i.e. all-metal or composite body, shells, inner cavities for irradiated fuel assemblies (SFA), lids with sealing systems. Residual heat release of SFA is discharged to the environment by natural way: through emission and convection of surrounding air. By now more than 100 dual purpose packaging sets TUK-108/1 are in operation in the mode of interim storage and transportation of SNF from decommissioned nuclear powered submarines (NPS). In accordance with certificate, spent fuel is stored in TUK-108/1 on the premises of plants involved in NPS dismantlement for 2 years, whereupon it is transported for processing to PO Mayak. At one Far Eastern plant Zvezda involved in NPS dismantlement there arose a complicated situation due to necessity to extend period of storage of SNF in TUK- 108/1. To ensure safety over a longer period of storage of SNF in TUK-108/1 it is essential to modify conditions of storage by removing of residual water and filling the inner cavity of the cask with an inert gas. Within implementation of the international 1.1- 2 project Development of drying technology for the cask TUK-108/1 intended for naval SNF under the Program, there has been developed the technology of preparation of the cask for long-term storage of SNF in TUK-108/1, the design of a mobile TUK-108/1 drying facility; a pilot facility has been manufactured. This report describes key issues of cask drying technology, justification of terms of dry storage of naval SNF in no-108/1, design features of the mobile drying facility, results of tests of the pilot facility at the Far Eastern plant Zvezda. (authors)

Guskov, Vladimir; Korotkov, Gennady [JSC 'KBSM' (Russian Federation); Barnes, Ella [US Environmental Protection Agency - EPA (United States); Snipes, Randy [Oak Ridge National Laboratory - ORNL, 1 Bethel Valley Rd, Oak Ridge, TN 37830 (United States)

2007-07-01T23:59:59.000Z

72

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

SciTech Connect (OSTI)

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.

McCormick, W.A.

1997-05-09T23:59:59.000Z

73

Packaging and Transportation for Offsite Shipment of Materials of National Security Interest  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The purpose of this Order is to make clear that the packaging and transportation of all offsite shipments of materials of national security interest for DOE must be conducted in accordance with DOT and Nuclear Regulatory Commission (NRC) regulations that would be applicable to comparable commercial shipments, except where an alternative course of action is identified in this Order. Cancels DOE O 461.1A.

2010-12-20T23:59:59.000Z

74

Testing hadronic-interaction packages at cosmic-ray energies  

SciTech Connect (OSTI)

A comparative analysis of the secondary particles output of the main hadronic interaction packages used in simulations of extensive air showers is presented. Special attention is given to the study of events with very energetic leading secondary particles, including diffractive interactions.

Canal, C. A. Garcia; Sciutto, S. J. [Departamento de Fisica, Universidad Nacional de La Plata, C.C. 67-1900 La Plata (Argentina); IFLP - CONICET, Universidad Nacional de La Plata, C.C. 67-1900 La Plata (Argentina); Tarutina, T. [Departamento de Fisica, Universidad Nacional de La Plata, C.C. 67-1900 La Plata (Argentina)

2009-03-01T23:59:59.000Z

75

NUHOWS - Storage and Transportation of Irradiated Reactor Components in Large Packages - 13439  

SciTech Connect (OSTI)

Most irradiated reactor components (hardware such as Control Rod Blades, Fuel Channels, Poison Curtains, etc.) generated at reactors previously required significant processing for size reduction due to the available transportation casks not being physically capable of containing unprocessed material. As of July 1, 2008, disposal for this typical waste class (B and C) became inaccessible (for the major part of the nation) due to the Barnwell, SC disposal facility being closed to all but its three compact states (CT, NJ and SC). Currently in the United States, most facilities are storing their irradiated hardware on-site in the spent fuel pools. Until recently with the opening of the Waste Control Specialists' Texas disposal facility, utilities faced the challenges of spent fuel pool space and capacity management. However, even with WCS's disposal availability, the site currently has annual Curie limitations for disposal, which will continue to promote interim on-site storage until such time as disposal is available. In response, Transnuclear Inc., (TN) an AREVA company, proceeded with designing a new large Radioactive Waste Container (RWC) that can be used to package irradiated hardware without the need for significant processing. The design features of the RWC allows for intermittent loadings of the hardware for better packaging efficiency, higher packaging density, space savings and reduced cost. This RWC is also compatible with TN's on-site modular vault storage system. Once completely loaded, the RWC can be transported to an on-site storage facility, an off-site storage facility and/or an available disposal facility. To accommodate the transportation, TN has designed a large transportation cask, the MP197HB. As the original design was for transporting fuel, it contains the necessary shielding to allow for the transport of unprocessed irradiated reactor components, while significantly reducing the amount of irradiated hardware shipments required with the use of the new RWC. This paper provides information on the unique design features of the RWC, storage module vaults, MP197HB Transportation Cask and cost saving benefits of using the large RWC for packaging, storage, transport and disposal. (authors)

Rae, Glen A. [Transnuclear, Inc., 7135 Minstrel Way, Columbia, MD 21045 (United States)] [Transnuclear, Inc., 7135 Minstrel Way, Columbia, MD 21045 (United States)

2013-07-01T23:59:59.000Z

76

9977 TYPE B PACKAGING INTERNAL DATA COLLECTION FEASIBILITY TESTING - MAGNETIC FIELD COMMUNICATIONS  

SciTech Connect (OSTI)

The objective of this report is to document the findings from proof-of-concept testing performed by the Savannah River National Laboratory (SRNL) R&D Engineering and Visible Assets, Inc. for the DOE Packaging Certification Program (PCP) to determine if RuBee (IEEE 1902.1) tags and readers could be used to provide a communication link from within a drum-style DOE certified Type B radioactive materials packaging. A Model 9977 Type B Packaging was used to test the read/write capability and range performance of a RuBee tag and reader. Testing was performed with the RuBee tags placed in various locations inside the packaging including inside the drum on the outside of the lid of the containment vessel and also inside of the containment vessel. This report documents the test methods and results. A path forward will also be recommended.

Shull, D.

2012-06-18T23:59:59.000Z

77

2014-02-07 Issuance: Test Procedure for Commercial Packaged Boilers; Request for Information  

Broader source: Energy.gov [DOE]

This document is a pre-publication Federal Register request for information regarding test procedures for commercial packaged boilers, as issued by the Deputy Assistant Secretary for Energy Efficiency on February 7, 2014.

78

Evaluation of Packaging Film Mechanical Integrity Using a Standardized Scratch Test  

E-Print Network [OSTI]

EVALUATION OF PACKAGING FILM MECHANICAL INTEGRITY USING A STANDARDIZED SCRATCH TEST A Thesis by BRIAN ANTHONY HARE Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment of the requirements... for the degree of MASTER OF SCIENCE August 2011 Major Subject: Materials Science and Engineering Evaluation of Packaging Film Mechanical Integrity Using a Standardized Scratch Test Copyright 2011 Brian...

Hare, Brian

2012-10-19T23:59:59.000Z

79

Definition of Small Gram Quantity Contents for Type B Radioactive Material Transportation Packages: Activity-Based Content Limitations  

SciTech Connect (OSTI)

Since the 1960's, the Department of Transportation Specification (DOT Spec) 6M packages have been used extensively for transportation of Type B quantities of radioactive materials between Department of Energy (DOE) facilities, laboratories, and productions sites. However, due to the advancement of packaging technology, the aging of the 6M packages, and variability in the quality of the packages, the DOT implemented a phased elimination of the 6M specification packages (and other DOT Spec packages) in favor of packages certified to meet federal performance requirements. DOT issued the final rule in the Federal Register on October 1, 2004 requiring that use of the DOT Specification 6M be discontinued as of October 1, 2008. A main driver for the change was the fact that the 6M specification packagings were not supported by a Safety Analysis Report for Packaging (SARP) that was compliant with Title 10 of the Code of Federal Regulations part 71 (10 CFR 71). Therefore, materials that would have historically been shipped in 6M packages are being identified as contents in Type B (and sometimes Type A fissile) package applications and addenda that are to be certified under the requirements of 10 CFR 71. The requirements in 10 CFR 71 include that the Safety Analysis Report for Packaging (SARP) must identify the maximum radioactivity of radioactive constituents and maximum quantities of fissile constituents (10 CFR 71.33(b)(1) and 10 CFR 71.33(b)(2)), and that the application (i.e., SARP submittal or SARP addendum) demonstrates that the external dose rate (due to the maximum radioactivity of radioactive constituents and maximum quantities of fissile constituents) on the surface of the packaging (i.e., package and contents) not exceed 200 mrem/hr (10 CFR 71.35(a), 10 CFR 71.47(a)). It has been proposed that a 'Small Gram Quantity' of radioactive material be defined, such that, when loaded in a transportation package, the dose rates at external points of an unshielded packaging not exceed the regulatory limits prescribed by 10 CFR 71 for non-exclusive shipments. The mass of each radioisotope presented in this paper is limited by the radiation dose rate on the external surface of the package, which per the regulatory limit should not exceed 200 mrem/hr. The results presented are a compendium of allowable masses of a variety of different isotopes (with varying impurity levels of beryllium in some of the actinide isotopes) that, when loaded in an unshielded packaging, do not result in an external dose rate on the surface of the package that exceeds 190 mrem/hr (190 mrem/hr was chosen to provide 5% conservatism relative to the regulatory limit). These mass limits define the term 'Small Gram Quantity' (SGQ) contents in the context of radioactive material transportation packages. The term SGQ is isotope-specific and pertains to contents in radioactive material transportation packages that do not require shielding and still satisfy the external dose rate requirements. Since these calculated mass limits are for contents without shielding, they are conservative for packaging materials that provide some limited shielding or if the contents are placed into a shielded package. The isotopes presented in this paper were chosen as the isotopes that Department of Energy (DOE) sites most likely need to ship. Other more rarely shipped isotopes, along with industrial and medical isotopes, are planned to be included in subsequent extensions of this work.

Sitaraman, S; Kim, S; Biswas, D; Hafner, R; Anderson, B

2010-10-27T23:59:59.000Z

80

Developing an Instrumentation Package for in-Water Testing of Marine Hydrokinetic Energy Devices: Preprint  

SciTech Connect (OSTI)

The ocean-energy industry is still in its infancy and device developers have provided their own equipment and procedures for testing. Currently, no testing standards exist for ocean energy devices in the United States. Furthermore, as prototype devices move from the test tank to in-water testing, the logistical challenges and costs grow exponentially. Development of a common instrumentation package that can be moved from device to device is one means of reducing testing costs and providing normalized data to the industry as a whole. As a first step, the U.S. National Renewable Energy Laboratory (NREL) has initiated an effort to develop an instrumentation package to provide a tool to allow common measurements across various ocean energy devices. The effort is summarized in this paper. First, we present the current status of ocean energy devices. We then review the experiences of the wind industry in its development of the instrumentation package and discuss how they can be applied in the ocean environment. Next, the challenges that will be addressed in the development of the ocean instrumentation package are discussed. For example, the instrument package must be highly adaptable to fit a large array of devices but still conduct common measurements. Finally, some possible system configurations are outlined followed by input from the industry regarding its measurement needs, lessons learned from prior testing, and other ideas.

Nelson, E.

2010-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

9978 AND 9975 TYPE B PACKAGING INTERNAL DATA COLLECTION FEASIBILITY TESTING  

SciTech Connect (OSTI)

The objective of this report is to document the findings from a series of proof-of-concept tests performed by Savannah River National Laboratory (SRNL) R and D Engineering, for the DOE Packaging Certification Program to determine if a viable radio link could be established from within the stainless steel confines of several drum-style DOE certified Type B radioactive materials packagings. Two in-hand, off-the-shelf radio systems were tested. The first system was a Wi-Fi Librestream Onsight{trademark} camera with a Fortress ES820 Access Point and the second was the On-Ramp Wireless Ultra-Link Processing{trademark} (ULP) radio system. These radio systems were tested within the Model 9975 and 9978 Type B packagings at the SRNL. This report documents the test methods and results. A path forward will also be recommended.

Fogle, R.

2012-05-07T23:59:59.000Z

82

Life and stability testing of packaged low-cost energy storage materials  

SciTech Connect (OSTI)

A low-cost laminated plastic film which is used to contain a Glauber's salt-based phase change thermal energy storage material in sausage-like containers called Chubs is discussed. The results of tests performed on the Chub packages themselves and on the thermal energy storage capacity of the packaged phase change material are described. From the test results, a set of specifications have been drawn up for a film material which will satisfactorily contain the phase change material under anticipated operating conditions. Calorimetric testing of the phase change material with thermal cycling indicates that a design capacity of 45 to 50 Btu/lb for a ..delta..T of 30/sup 0/F can be used for the packaged material.

Frysinger, G.R.

1980-07-01T23:59:59.000Z

83

Hail Impact Testing on Crystalline Si Modules with Flexible Packaging...  

Broader source: Energy.gov (indexed) [DOE]

Colorado pvmrw13ps2westpakbrown.pdf More Documents & Publications Test Procedure for UV Weathering Resistance of Backsheet Improved Reliability of PV Modules with Lexan PC...

84

RH Packaging Operations Manual  

SciTech Connect (OSTI)

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

Washington TRU Solutions LLC

2003-09-17T23:59:59.000Z

85

Radioisotope Thermoelectric Generator Package O-Ring Seal Material Validation Testing  

SciTech Connect (OSTI)

The Radioisotope Thermoelectric Generator Package O-Ring Seal Material Validation Test was conducted to validate the use of the Butyl material as a primary seal throughout the required temperature range. Three tests were performed at (1) 233 K ({minus}40 {degrees}F), (2) a specified operating temperature, and (3) 244 K ({minus}20 {degrees}F) before returning to room temperature. Helium leak tests were performed at each test point to determine seal performance. The two major test objectives were to establish that butyl rubber material would maintain its integrity under various conditions and within specified parameters and to evaluate changes in material properties.

Adkins, H.E.; Ferrell, P.C.; Knight, R.C.

1994-09-30T23:59:59.000Z

86

Review of waste package verification tests. Semiannual report, October 1983-March 1984. Volume 4  

SciTech Connect (OSTI)

The current study is part of an ongoing task to specify tests that may be used to verify that engineered waste package/repository systems comply with NRC radionuclide containment and controlled release performance objectives. Work covered in this report includes crushed tuff packing material for use in a high-level waste tuff repository. Ranges of repository conditions relevant to its testing and other factors important for its performance are discussed. 23 refs., 5 figs., 3 tabs.

Jain, H.; Veakis, E.; Soo, P.

1985-06-01T23:59:59.000Z

87

Transport Test Problems for Hybrid Methods Development  

SciTech Connect (OSTI)

This report presents 9 test problems to guide testing and development of hybrid calculations for the ADVANTG code at ORNL. These test cases can be used for comparing different types of radiation transport calculations, as well as for guiding the development of variance reduction methods. Cases are drawn primarily from existing or previous calculations with a preference for cases which include experimental data, or otherwise have results with a high level of confidence, are non-sensitive, and represent problem sets of interest to NA-22.

Shaver, Mark W.; Miller, Erin A.; Wittman, Richard S.; McDonald, Benjamin S.

2011-12-28T23:59:59.000Z

88

PATRAM '92: 10th international symposium on the packaging and transportation of radioactive materials [Papers presented by Sandia National Laboratories  

SciTech Connect (OSTI)

This document provides the papers presented by Sandia Laboratories at PATRAM '92, the tenth International symposium on the Packaging and Transportation of Radioactive Materials held September 13--18, 1992 in Yokohama City, Japan. Individual papers have been cataloged separately. (FL)

none,

1992-01-01T23:59:59.000Z

89

Notice of Intent to Revise Department of Energy Order 460.1C, Packaging and Transportation Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The purpose of this memorandum is to provide justification for the proposed revision of Department of Energy (DOE} Order (O} 460.lC, Packaging and Transportation Safety as part of the quadrennial review and recertification required by DOE O 251.lC, Departmental Directives Program.

2015-01-15T23:59:59.000Z

90

Feature test report for the Small Debris Collection and Packaging System  

SciTech Connect (OSTI)

The Spent Nuclear Fuel Equipment Engineering group performed feature testing of the Small Debris Collection and Packaging System (SDCPS) in the 305 Cold Test Facility from January 30, 1995, to February 1, 1995. Feature testing of the Small Debris Collection and Packaging System (SDCPS) was performed for the following reasons: To assess the feasibility of using ``drop-out`` vessels to collect small debris (<2.5 cm) in MK-II fuel canisters while transferring sludge to the Weasel Pit. To evaluate system performance under conditions similar to those in the K-Basins (e.g. submerged under 4.9 meters of water and operated with long handled tools) while using a surrogate sludge mixed with debris. To determine if canister weight could be used to predict the volume of sludge and/or debris contained within the canisters during system operation.

Brisbin, S.A.

1995-03-17T23:59:59.000Z

91

Aquifer testing data package for 1993 200-UP-1 Groundwater Operable Unit  

SciTech Connect (OSTI)

The following aquifer testing data supported 1993 Interim Remedial Measure field work for the U-1 and U-2 crib area near the uranium technetium and nitrate plumes beneath the U Plant Aggregate Area. The purpose of aquifer testing was to fill in hydraulic conductivity data gaps in the western portion of 200 West Area and help refine the hydrogeologic conceptual model. This data package reports data collected in accordance with the description of work released in 1993 by L.C. Swanson, entitled Description of Work for the 200-UP-1 Aquifer Testing Activity. These data are analyzed in the document Aquifer Test Analysis Results for 1993 200-UP-1 Groundwater Operable Unit. Slug tests were conducted at 7 existing wells, and pumping tests were conducted at 2 of those same existing wells.

Swanson, L.C.

1994-06-24T23:59:59.000Z

92

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

SciTech Connect (OSTI)

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 {sup 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.

Trennel, A.J.

1995-12-31T23:59:59.000Z

93

Radioisotope thermoelectric generator package o-ring seal material validation testing  

SciTech Connect (OSTI)

The Radioisotope Thermoelectric Generator Package O-Ring Seal Material Validation Test was conducted to validate the use of the Butyl material as a primary seal throughout the required temperature range. Three tests were performed at (I) 233 K ({minus}40 {degree}F), (2) a specified operating temperature, and (3) 244 K ({minus}20 {degree}F) before returning to room temperature. Helium leak tests were performed at each test point to determine seal performance. The two major test objectives were to establish that butyl rubber material would maintain its integrity under various conditions and within specified parameters and to evaluate changes in material properties. {copyright} {ital 1995} {ital American} {ital Institute} {ital of} {ital Physics}

Adkins, H.E.; Ferrell, P.C.; Knight, R.C. [Westinghouse Hanford Company, P. O. Box 1970, MSIN N1-25, Richland, Washington 99352 (United States)

1995-01-20T23:59:59.000Z

94

PERFORMANCE TESTING OF SPRING ENERGIZED C-RINGS FOR USE IN RADIOACTIVE MATERIAL PACKAGINGS CONTAINING TRITIUM  

SciTech Connect (OSTI)

This paper describes the sealing performance testing and results of silver-plated inconel Spring Energized C-Rings used for tritium containment in radioactive shipping packagings. The test methodology used follows requirements of the American Society of Mechanical Engineers (ASME) summarized in ASME Pressure Vessel Code (B&PVC), Section V, Article 10, Appendix IX (Helium Mass Spectrometer Test - Hood Technique) and recommendations by the American National Standards Institute (ANSI) described in ANSI N14.5-1997. The tests parameters bound the predicted structural and thermal responses from conditions defined in the Code of Federal Regulations 10 CFR 71. The testing includes an evaluation of the effects of pressure, temperature, flange deflection, surface roughness, permeation, closure torque, torque sequencing and re-use on performance of metal C-Ring seals.

Blanton, P; Kurt Eberl, K

2007-10-23T23:59:59.000Z

95

NREL: Transportation Research - Truck Platooning Testing  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the Contributions andData andFleet Test and Evaluation PhotoSystemsTransportationTruck

96

CH Packaging Program Guidance  

SciTech Connect (OSTI)

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

Washington TRU Solutions LLC

2003-04-30T23:59:59.000Z

97

CH Packaging Program Guidance  

SciTech Connect (OSTI)

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

Washington TRU Solutions LLC

2002-03-04T23:59:59.000Z

98

Over-the-road shock and vibration testing of the radioisotope thermoelectric generator transportation system  

SciTech Connect (OSTI)

Radioisotope Thermoelectric Generators (RTG) convert heat generated by radioactive decay into electricity through the use of thermocouples. The RTGs have a long operating life, are reasonably lightweight, and require little or no maintenance, which make them particularly attractive for use in spacecraft. However, because RTGs contain significant quantities of radioactive materials, normally plutonium-238 and its decay products, they must be transported in packages built in accordance with Title 10, Code of Federal Regulations, Part 71 (10 CFR 71). To meet these regulations, a RTG Transportation System (RTGTS) that fully complies with 10 CFR 71 has been developed, which protects RTGs from adverse environmental conditions during normal conditions of transport (e.g., shock, vibration, and heat). To ensure the protection of RTGs from shock and vibration loadings during transport, extensive over-the-road testing was conducted on the RTG`S to obtain real-time recordings of accelerations of the air-ride suspension system trailer floor, packaging, and support structure. This paper provides an overview of the RTG`S, a discussion of the shock and vibration testing, and a comparison of the test results to the specified shock response spectra and power spectral density acceleration criteria.

Becker, D.L.

1997-05-01T23:59:59.000Z

99

Evaluation and compilation of DOE waste package test data: Biannual report, August 1987--January 1988  

SciTech Connect (OSTI)

This report summarizes results of the National Bureau of Standards (NBS) evaluations on waste packages designed for containment of radioactive high-level nuclear waste (HLW). The waste package is a proposed engineered barrier that is part of a permanent repository for HLW. Metal alloys are the principal barriers within the engineered system. Since enactment of the Budget Reconciliation Act for Fiscal Year 1988, the Yucca Mountain, Nevada, site (in which tuff is the geologic medium) is the only site that will be characterized for use as high-level nuclear waste repository. During the reporting period of August 1987 to January 1988, five reviews were completed for tuff, and these were grouped into the categories: ferrous alloys, copper, groundwater chemistry, and glass. Two issues are identified for the Yucca Mountain site: the approach used to calculate corrosion rates for ferrous alloys, and crevice corrosion was observed in a copper-nickel alloy. Plutonium can form pseudo-colloids that may facilitate transport. NBS work related to the vitrification of HLW borosilicate glass at the West Valley Demonstration Project (WVDP) and the Defense Waste Processing Facility (DWPF) and activities of the DOE Materials Characterization Center (MCC) for the 6-month reporting period are also included. 27 refs., 3 figs.

Interrante, C.; Escalante, E.; Fraker, A.; Ondik, H.; Plante, E.; Ricker, R.; Ruspi, J.

1988-08-01T23:59:59.000Z

100

Safety evaluation for packaging (onsite) for the concrete-shielded RH TRU drum for the 327 Postirradiation Testing Laboratory  

SciTech Connect (OSTI)

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 HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments. The drum will be used for transport of 327 Building legacy waste from the 300 Area to a solid waste storage facility on the Hanford Site.

Smith, R.J.

1998-03-31T23:59:59.000Z

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
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101

Determining the Appropriate Package and Transportation Methodology for the Detroit Edison, Fermi II Msrs and Associated Components  

SciTech Connect (OSTI)

During the spring of 2005, Detroit Edison, Enrico Fermi II Nuclear Power Station (Fermi) decided to disposition two MSRs and associated components scheduled for replacement in the spring of 2006 during the MSR Replacement Outage. Of concern to Fermi was the proper packaging and transportation methodology when dis-positioning a component measuring approximately 110' in length and 13' in diameter and weighing over 300 tons. Upon removal from the Turbine Deck the retired MSRs and associated components were turned over to the Rad Waste Group for packaging and final disposition. Fermi requested quotations from vendors to package, transport, and disposition the MSRs and associated components. However, multiple Vendors informed Fermi that the size and weight of the MSRs were questionable in passing permitting requirements and would require segmentation and volume reduction on site or at a waste processor. Fermi contracted with MHF Logistical Solutions (MHF-LS) based on their ability to receive clearances for shipping the MSRs in one piece via two heavy haul rail conveyances acting as a bolstered load with professionally engineered blocking and bracing configured to support the retired MSRs. (authors)

Weber, B. [Detroit Edison Company/Enrico Fermi II Nuclear Power Station, Newport, MI (United States); Dempsey, S. [MHF Logistical Solutions, Oak Ridge, TN (United States)

2007-07-01T23:59:59.000Z

102

Evaluation and compilation of DOE waste package test data: Biannual report, February 1987--July 1987  

SciTech Connect (OSTI)

The waste package is a proposed engineering barrier that is part of a permanent repository for HLW. Metal alloys are the principal barriers within the engineered system. Technical discussions are given for the corrosion of metals proposed for the canister, particularly carbon steels, stainless steels, and copper. The current level of understanding of several canister materials is questioned for the candidate repository in tuff. Three issues are addressed, the possibility of the stress-induced failure of Zircaloy, the possible corrosion of copper and copper alloys, and the lack of site-specific characterization data. Discussions are given on problems concerning localized corrosion and environmentally assisted cracking of AISI 1020 steel at elevated temperatures (150{degree}C). For the proposed salt site, the importance of the duration of corrosion tests and some of the conditions that may preclude prompt initiation of needed long-term testing are two issues that are discussed. 31 refs., 5 figs.

Interrante, C.; Escalante, E.; Fraker, A.; Hall, D.; Harrison, S.; Liggett, W.; Linzer, M.; Ricker, R.; Ruspi, J.; Shull, R.

1988-05-01T23:59:59.000Z

103

RH Packaging Program Guidance  

SciTech Connect (OSTI)

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

Washington TRU Solutions LLC

2006-11-07T23:59:59.000Z

104

RH Packaging Program Guidance  

SciTech Connect (OSTI)

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

Washington TRU Solutions LLC

2008-01-12T23:59:59.000Z

105

Full-Scale Accident Testing in Support of Used Nuclear Fuel Transportation.  

SciTech Connect (OSTI)

The safe transport of spent nuclear fuel and high-level radioactive waste is an important aspect of the waste management system of the United States. The Nuclear Regulatory Commission (NRC) currently certifies spent nuclear fuel rail cask designs based primarily on numerical modeling of hypothetical accident conditions augmented with some small scale testing. However, NRC initiated a Package Performance Study (PPS) in 2001 to examine the response of full-scale rail casks in extreme transportation accidents. The objectives of PPS were to demonstrate the safety of transportation casks and to provide high-fidelity data for validating the modeling. Although work on the PPS eventually stopped, the Blue Ribbon Commission on America’s Nuclear Future recommended in 2012 that the test plans be re-examined. This recommendation was in recognition of substantial public feedback calling for a full-scale severe accident test of a rail cask to verify evaluations by NRC, which find that risk from the transport of spent fuel in certified casks is extremely low. This report, which serves as the re-assessment, provides a summary of the history of the PPS planning, identifies the objectives and technical issues that drove the scope of the PPS, and presents a possible path for moving forward in planning to conduct a full-scale cask test. Because full-scale testing is expensive, the value of such testing on public perceptions and public acceptance is important. Consequently, the path forward starts with a public perception component followed by two additional components: accident simulation and first responder training. The proposed path forward presents a series of study options with several points where the package performance study could be redirected if warranted.

Durbin, Samuel G.; Lindgren, Eric R.; Rechard, Rob P.; Sorenson, Ken B.

2014-09-01T23:59:59.000Z

106

Carbon Capture, Transport and Storage Regulatory Test Exercise...  

Open Energy Info (EERE)

Jump to: navigation, search Tool Summary LAUNCH TOOL Name: Carbon Capture, Transport and Storage Regulatory Test Exercise: Output Report Focus Area: Clean Fossil Energy Topics:...

107

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

SciTech Connect (OSTI)

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

Mourao, Rogerio P.; Leite da Silva, Luiz [Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte (Brazil)] [Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte (Brazil); Miranda, Carlos A.; Mattar Neto, Miguel [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil)] [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil); Quintana, Jose F.A.; Saliba, Roberto O. [Comision Nacional de Energia Atomica, Bariloche (Argentina)] [Comision Nacional de Energia Atomica, Bariloche (Argentina); Novara, Oscar E. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)] [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

2013-07-01T23:59:59.000Z

108

FIFTH STATUS REPORT: TESTING OF AGED SOFTWOOD FIBERBOARD MATERIAL FOR THE 9975 SHIPPING PACKAGE  

SciTech Connect (OSTI)

Samples have been prepared from a 9975 lower fiberboard subassembly fabricated from softwood fiberboard. Physical, mechanical and thermal properties have been measured following varying periods of conditioning in each of several environments. These tests have been conducted in the same manner as previous testing on cane fiberboard samples. Overall, similar aging trends are observed for softwood and cane fiberboard samples, with a few differences. Some softwood fiberboard properties tend to degrade faster in elevated humidity environments, while some cane fiberboard properties degrade faster in the hotter dry environments. As a result, it is premature to assume both materials will age at the same rates, and the preliminary aging models developed for cane fiberboard might not apply to softwood fiberboard. However, it is expected that both cane and softwood fiberboard assemblies will perform satisfactorily in conforming packages stored in a typical KAC storage environment for up to 15 years. Aging and testing of softwood fiberboard will continue and additional data will be collected. Additional samples will be added to each aging environment, to support development of an aging model specific to softwood fiberboard. Post-conditioning data have been measured on samples from a single softwood fiberboard assembly, and baseline data are also available from a limited number of vendor-provided samples. This provides minimal information on the possible sample-to-sample variation exhibited by softwood fiberboard. Data to date are generally consistent with the range seen in cane fiberboard, but some portions of the data trends are skewed toward the lower end of that range. Two additional softwood fiberboard source packages have been obtained and will begin to provide data on the range of variability of this material.

Daugherty, W.; Skidmore, E.; Dunn, K.

2014-04-15T23:59:59.000Z

109

CH Packaging Program Guidance  

SciTech Connect (OSTI)

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

Washington TRU Solutions LLC

2005-02-28T23:59:59.000Z

110

CH Packaging Operations Manual  

SciTech Connect (OSTI)

This procedure provides instructions for assembling the CH Packaging Drum payload assembly, Standard Waste Box (SWB) assembly, Abnormal Operations and ICV and OCV Preshipment Leakage Rate Tests on the packaging seals, using a nondestructive Helium (He) Leak Test.

Washington TRU Solutions LLC

2005-06-13T23:59:59.000Z

111

Underground Test Area Subproject Phase I Data Analysis Task. Volume VIII - Risk Assessment Documentation Package  

SciTech Connect (OSTI)

Volume VIII of the documentation for the Phase I Data Analysis Task performed in support of the current Regional Flow Model, Transport Model, and Risk Assessment for the Nevada Test Site Underground Test Area Subproject contains the risk assessment documentation. Because of the size and complexity of the model area, a considerable quantity of data was collected and analyzed in support of the modeling efforts. The data analysis task was consequently broken into eight subtasks, and descriptions of each subtask's activities are contained in one of the eight volumes that comprise the Phase I Data Analysis Documentation.

None

1996-12-01T23:59:59.000Z

112

Interfacial Transport Test section length = 4 m  

E-Print Network [OSTI]

Penetrations (e.g. modified back wall topology) Free Surface Interfacial Transport - Turbulence at Free Surface "cold" and "hot" strikes 20kW/m2 , 30°, 10 L/s flow Current data analysis and experiments are used ·NSTX environment simulation MTOR designed and constructed in collaboration between UCLA, PPPL and ORNL

Abdou, Mohamed

113

E-Print Network 3.0 - air transportable package Sample Search...  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

21st Century Air Law... Trading As Market-Based Options in Air Transport Air Law Emissions trading, tax, charge, environment, ICAO... , comptence Azzie, Ralph V p. 3 Second...

114

Results from simulated upper-plenum aerosol transport tests  

SciTech Connect (OSTI)

A series of eight aerosol transport experiments, designated as Aerosol Transport Tests (ATT) A101 through A108, has recently been completed at the Oak Ridge National Laboratory (ORNL). These tests provide a data base for validation of aerosol transport modeling used in the TRAP-MELT2 computer code (Jordan and Kuhlman, 1985), which was developed at Battelle Columbus Laboratories to calculate aerosol/fission-product transport in the reactor coolant system in postulated light-water reactor (LWR) core-melt accidents. Results from tests A103 and A104 have been summarized in a previous paper (Wright and Pattison, 1985a); the present paper discusses results from tests A105 through A108.

Wright, A.L.; Pattison, W.L.

1986-01-01T23:59:59.000Z

115

Transport Test Problems for Radiation Detection Scenarios  

SciTech Connect (OSTI)

This is the final report and deliverable for the project. It is a list of the details of the test cases for radiation detection scenarios.

Shaver, Mark W.; Miller, Erin A.; Wittman, Richard S.; McDonald, Benjamin S.

2012-09-30T23:59:59.000Z

116

Analysis of Alcove 8/Niche 3 Flow and Transport Tests  

SciTech Connect (OSTI)

The purpose of this report is to document analyses of the Alcove 8/Niche 3 flow and transport tests, with a focus on the large-infiltration-plot tests and compare pre-test model predictions with the actual test observations. The tests involved infiltration that originated from the floor of Alcove 8 (located in the Enhanced Characterization of Repository Block (ECRB) Cross Drift) and observations of seepage and tracer transport at Niche 3 (located in the Main Drift of the Exploratory Studies Facility (ESF)). The test results are relevant to drift seepage and solute transport in the unsaturated zone (UZ) of Yucca Mountain. The main objective of this analysis was to evaluate the modeling approaches used and the importance of the matrix diffusion process by comparing simulation and actual test observations. The pre-test predictions for the large plot test were found to differ from the observations and the reasons for the differences were documented in this report to partly address CR 6783, which concerns unexpected test results. These unexpected results are discussed and assessed with respect to the current baseline unsaturated zone radionuclide transport model in Sections 6.2.4, 6.3.2, and 6.4.

H.H. Liu

2006-09-01T23:59:59.000Z

117

SciTech Connect: Normal Conditions of Transport Truck Test of...  

Office of Scientific and Technical Information (OSTI)

Normal Conditions of Transport Truck Test of a Surrogate Fuel Assembly. Citation Details In-Document Search Title: Normal Conditions of Transport Truck Test of a Surrogate Fuel...

118

NREL: Transportation Research - Fleet Test and Evaluation  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the Contributions andData andFleet Test and Evaluation Photo of medium-duty truck with

119

NREL: Transportation Research - Truck Stop Electrification Testing  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the Contributions andData andFleet Test and Evaluation

120

Transuranic Package Transporter (TRUPACT-III) Content Codes (TRUCON-III) |  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNGInternationalTechnologyDepartmentStorage Interface TransportationDepartment of

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

Evaluation and compilation of DOE waste package test data: Biannual report, August 1986-January 1987  

SciTech Connect (OSTI)

This report summarizes results of the National Bureau of Standards (NBS) evaluations of Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW). The waste package is a proposed engineered barrier that is part of a permanent repository for HLW. Metal alloys are the principal barriers within the engineered system. Technical discussions are given for the corrosion of metals proposed for the canister, particularly carbon and stainless steels, and copper. In the section on tuff, the current level of understanding of several canister materials is questioned. Within the Basalt Waste Isolation Project (BWIP) section, discussions are given on problems concerning groundwater, materials for use in the metallic overpack, and diffusion through the packing. For the proposed salt site, questions are raised on the work on both ASTM A216 Steel and Ti-Code 12. NBS work related to the vitrification of HLW borosilicate glass at the West Valley Demonstration Project (WVDP) and the Defense Waste Processing Facility (DWPF) is covered. NBS reviews of selected DOE technical reports and a summary of current waste-package activities of the Materials Characterization Center (MCC) is presented. Using a database management system, a computerized database for storage and retrieval of reviews and evaluations of HLW data has been developed and is described. 17 refs., 2 figs., 2 tabs.

Interrante, C.; Escalante, E.; Fraker, A.; Harrison, S.; Shull, R.; Linzer, M.; Ricker, R.; Ruspi, J.

1987-10-01T23:59:59.000Z

122

FINITE ELEMENT ANALYSIS OF BULK TRITIUM SHIPPING PACKAGE  

SciTech Connect (OSTI)

The Bulk Tritium Shipping Package was designed by Savannah River National Laboratory. This package will be used to transport tritium. As part of the requirements for certification, the package must be shown to meet the scenarios of the Hypothetical Accident Conditions (HAC) defined in Code of Federal Regulations Title 10 Part 71 (10CFR71). The conditions include a sequential 30-foot drop event, 30-foot dynamic crush event, and a 40-inch puncture event. Finite Element analyses were performed to support and expand upon prototype testing. Cases similar to the tests were evaluated. Additional temperatures and orientations were also examined to determine their impact on the results. The peak stress on the package was shown to be acceptable. In addition, the strain on the outer drum as well as the inner containment boundary was shown to be acceptable. In conjunction with the prototype tests, the package was shown to meet its confinement requirements.

Jordan, J.

2010-06-02T23:59:59.000Z

123

2014-03-31 Issuance: Test Procedure for Commercial Packaged Boilers; Request for Information, Reopening of the Comment Period  

Broader source: Energy.gov [DOE]

This document is a pre-publication Federal Register notice reopening the comment period for the request for information regarding the commercial packaged boiler test procedure rulemaking, as issued by the Deputy Assistant Secretary for Energy Efficiency on March 31, 2014. Though it is not intended or expected, should any discrepancy occur between the document posted here and the document published in the Federal Register, the Federal Register publication controls. This document is being made available through the Internet solely as a means to facilitate the public's access to this document.

124

FABRICATION AND DEPLOYMENT OF THE 9979 TYPE AF RADIOACTIVE WASTE PACKAGING FOR THE DEPARTMENT OF ENERGY  

SciTech Connect (OSTI)

This paper summarizes the development, testing, and certification of the 9979 Type A Fissile Packaging that replaces the UN1A2 Specification Shipping Package eliminated from Department of Transportation (DOT) 49 CFR 173. The DOT Specification Package was used for many decades by the U.S. nuclear industry as a fissile waste container until its removal as an authorized container by DOT. This paper will discuss stream lining procurement of high volume radioactive material packaging manufacturing, such as the 9979, to minimize packaging production costs without sacrificing Quality Assurance. The authorized content envelope (combustible and non-combustible) as well as planned content envelope expansion will be discussed.

Blanton, P.; Eberl, K.

2013-10-10T23:59:59.000Z

125

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

SciTech Connect (OSTI)

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

Not Available

1991-10-18T23:59:59.000Z

126

DOT-7A Type A packaging design guide  

SciTech Connect (OSTI)

The purpose of this Design Guide is to provide instruction for designing a U.S. Department of Transportation Specification 7A (DOT-7A) Type A packaging. Another purpose for this Design Guide is to support the evaluation and testing activities that are performed on new designs by a U.S. Department of Energy (DOE) test facility. This evaluation and testing program is called the DOT-7A Program. When an applicant has determined that a DOT-7A packaging is needed and not commercially available, a design may be created according to this document. The design should include a packaging drawing, specifications, analysis report, operating instructions, and a Packaging Qualification Checklist; all of which should be forwarded to a DOE/HQ approved test facility for evaluation and testing. This report is being submitted through the Engineering Documentation System so that it may be used for reference and information purposes.

Kelly, D.L.

1995-01-23T23:59:59.000Z

127

DEVELOPMENT OF THE H1700 SHIPPING PACKAGE  

SciTech Connect (OSTI)

The H1700 Package is based on the DOE-EM Certified 9977 Packaging. The H1700 will be certified by the Packaging Certification Division of the National Nuclear Security Administration for the shipment of plutonium by air by the United Stated Military both within the United States and internationally. The H1700 is designed to ship radioactive contents in assemblies of Radioisotope Thermoelectric Generators (RTGs) or arrangements of nested food-pack cans. The RTG containers are designed and tested to remain leaktight during transport, handling, and storage; however, their ability to remain leaktight during transport in the H1700 is not credited. This paper discusses the design and special operation of the H1700.

Abramczyk, G.; Loftin, B.; Mann, P.

2009-06-05T23:59:59.000Z

128

WatSen: Design and testing of a prototype mid-IR spectrometer and microscope package for Mars exploration  

E-Print Network [OSTI]

We have designed and built a compact breadboard prototype instrument called WatSen: a combined ATR mid-IR spectrometer, fixed-focus microscope, and humidity sensor. The instrument package is enclosed in a rugged cylindrical casing only 26mm in diameter. The functionality, reliability and performance of the instrument was tested in an environment chamber set up to resemble martian surface conditions. The effective wavelength range of the spectrometer is 6.2 - 10.3 micron with a resolution delta-wavelength/wavelength = 0.015. This allows detection of silicates and carbonates, including an indication of the presence of water (ice). Spectra of clusters of grains < 1mm across were acquired that are comparable with spectra of the same material obtained using a commercial system. The microscope focuses through the diamond ATR crystal. Colour images of the grains being spectroscopically analysed are obtainable with a resolution of ~ 20 micron.

Wolters, Stephen D; Sund, Arnt T; Bohman, Axel; Guthery, William; Sund, Bjornar T; Hagermann, Axel; Tomkinson, Tim; Romstedt, Jens; Morgan, Geraint H; Grady, Monica M; 10.1007/s10686-012-9328-8

2013-01-01T23:59:59.000Z

129

CEPXS/ONELD Version 2. 0: A discrete ordinates code package for general one-dimensional coupled electron-photon transport  

SciTech Connect (OSTI)

CEPXS/ONELD is the only discrete ordinates code capable of modelling the fully-coupled electron-photon cascade at high energies. Quantities that are related to the particle flux such as dose and charge deposition can readily be obtained. This deterministic code is much faster than comparable Monte Carlo codes. The unique adjoint transport capability of CEPXS/ONELD also enables response functions to be readily calculated. Version 2.0 of the CEPXS/ONELD code package has been designed to allow users who are not expert in discrete ordinates methods to fully exploit the code's capabilities. 14 refs., 15 figs.

Lorence, L.J. Jr.

1991-01-01T23:59:59.000Z

130

Regulatory compliance guide for DOT-7A type A packaging design  

SciTech Connect (OSTI)

The purpose of this guide is to provide instruction for assuring that the regulatory design requirements for a DOT-7A Type A packaging are met. This guide also supports the testing and evaluation activities that are performed on new packaging designs by a DOE-approved test facility through the DOE`s DOT-7A Test Program. This Guide was updated to incorporate regulatory changes implemented by HM-169A (49 CFR, `Transportation`).

Kelly, D.L.

1996-06-04T23:59:59.000Z

131

Safety evaluation for packaging for the transport of K Basin sludge samples in the PAS-1 cask  

SciTech Connect (OSTI)

This safety evaluation for packaging authorizes the shipment of up to two 4-L sludge samples to and from the 325 Lab or 222-S Lab for characterization. The safety of this shipment is based on the current U.S. Department of Energy Certification of Compliance (CoC) for the PAS-1 cask, USA/9184/B(U) (DOE).

SMITH, R.J.

1998-11-17T23:59:59.000Z

132

June 2012 Groundwater Sampling at the Central Nevada Test Area (Data Validation Package)  

SciTech Connect (OSTI)

The U.S. Department of Energy Office of Legacy Management conducted annual sampling at the Central Nevada Test Area (CNTA) on June 26-27, 2012, in accordance with the 2004 Correction Action Decision Document/Corrective Action Plan for Corrective Action Unit 443: Central Nevada Test Area (CNTA)-Subsurface and the addendum to the "Corrective Action Decision Document/Corrective Action Plan" completed in 2008. Sampling and analysis were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PLN/S04351), continually updated).

None

2013-03-01T23:59:59.000Z

133

May 2011 Groundwater Sampling at the Central Nevada Test Area (Data Validation Package)  

SciTech Connect (OSTI)

The U.S. Department of Energy Office of Legacy Management conducted annual sampling at the Central Nevada Test Area (CNTA) on May 10-11, 2011, in accordance with the 2004 Correction Action Decision Document/Corrective Action Plan for Corrective Action Unit 443: Central Nevada Test Area (CNTA)-Subsurface and the addendum to the "Corrective Action Decision Document/Corrective Action Plan" completed in 2008. Sampling and analysis were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PLN/S04351), continually updated).

None

2011-11-01T23:59:59.000Z

134

May 2010 Groundwater Sampling at the Central Nevada Test Area (Data Validation Package)  

SciTech Connect (OSTI)

The U.S. Department of Energy Office of Legacy Management conducted annual sampling at the Central Nevada Test Area (CNTA) on June 7-9, 2010, in accordance with the 2004 Correction Action Decision Document/Corrective Action Plan for Corrective Action Unit 443: Central Nevada Test Area (CNTA)-Subsurface. Sampling and analysis were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PLN/S04351), continually updated).

None

2011-02-01T23:59:59.000Z

135

DOE-EM-45 PACKAGING OPERATIONS AND MAINTENANCE COURSE  

SciTech Connect (OSTI)

Savannah River National Laboratory - Savannah River Packaging Technology (SRNL-SRPT) delivered the inaugural offering of the Packaging Operations and Maintenance Course for DOE-EM-45's Packaging Certification Program (PCP) at the University of South Carolina Aiken on September 1 and 2, 2009. Twenty-nine students registered, attended, and completed this training. The DOE-EM-45 Packaging Certification Program (PCP) sponsored the presentation of a new training course, Packaging Maintenance and Operations, on September 1-2, 2009 at the University of South Carolina Aiken (USC-Aiken) campus in Aiken, SC. The premier offering of the course was developed and presented by the Savannah River National Laboratory, and attended by twenty-nine students across the DOE, NNSA and private industry. This training informed package users of the requirements associated with handling shipping containers at a facility (user) level and provided a basic overview of the requirements typically outlined in Safety Analysis Report for Packaging (SARP) Chapters 1, 7, and 8. The course taught packaging personnel about the regulatory nature of SARPs to help reduce associated and often costly packaging errors. Some of the topics covered were package contents, loading, unloading, storage, torque requirements, maintaining records, how to handle abnormal conditions, lessons learned, leakage testing (including demonstration), and replacement parts. The target audience for this course was facility operations personnel, facility maintenance personnel, and field quality assurance personnel who are directly involved in the handling of shipping containers. The training also aimed at writers of SARP Chapters 1, 7, and 8, package designers, and anyone else involved in radioactive material packaging and transportation safety. Student feedback and critiques of the training were very positive. SRNL will offer the course again at USC Aiken in September 2010.

Watkins, R.; England, J.

2010-05-28T23:59:59.000Z

136

RADIOACTIVE MATERIAL PACKAGING TORQUE REQUIREMENTS COMPLIANCE  

SciTech Connect (OSTI)

Shipping containers used to transport radioactive material (RAM) in commerce employ a variety of closure mechanisms. Often, these closure mechanisms require a specific amount of torque be applied to a bolt, nut or other threaded fastener. It is important that the required preload is achieved so that the package testing and analysis is not invalidated for the purpose of protecting the public. Torque compliance is a means of ensuring closure preload, is a major factor in accomplishing the package functions of confinement/containment, sub-criticality, and shielding. This paper will address the importance of applying proper torque to package closures, discuss torque value nomenclature, and present one methodology to ensure torque compliance is achieved.

Watkins, R.; Leduc, D.

2011-03-24T23:59:59.000Z

137

Azimuthal Anisotropies as Stringent Test for Nuclear Transport Models  

E-Print Network [OSTI]

Azimuthal distributions of charged particles and intermediate mass fragments emitted in Au+Au collisions at 600AMeV have been measured using the FOPI facility at GSI-Darmstadt. Data show a strong increase of the in-plane azimuthal anisotropy ratio with the charge of the detected fragment. Intermediate mass fragments are found to exhibit a strong momentum-space alignment with respect of the reaction plane. The experimental results are presented as a function of the polar center-of-mass angle and over a broad range of impact parameters. They are compared to the predictions of the Isospin Quantum Molecular Dynamics model using three different parametrisations of the equation of state. We show that such highly accurate data provide stringent test for microscopic transport models and can potentially constrain separately the stiffness of the nuclear equation of state and the momentum dependence of the nuclear interaction.

P. Crochet; F. Rami; R. Dona; the FOPI Collaboration

1997-09-15T23:59:59.000Z

138

Tank vapor sampling and analysis data package for tank 241-C-106 waste retrieval sluicing system process test phase III  

SciTech Connect (OSTI)

This data package presents sampling data and analytical results from the March 28, 1999, vapor sampling of Hanford Site single-shell tank 241-C-106 during active sluicing. Samples were obtained from the 296-C-006 ventilation system stack and ambient air at several locations. Characterization Project Operations (CPO) was responsible for the collection of all SUMMATM canister samples. The Special Analytical Support (SAS) vapor team was responsible for the collection of all triple sorbent trap (TST), sorbent tube train (STT), polyurethane foam (PUF), and particulate filter samples collected at the 296-C-006 stack. The SAS vapor team used the non-electrical vapor sampling (NEVS) system to collect samples of the air, gases, and vapors from the 296-C-006 stack. The SAS vapor team collected and analyzed these samples for Lockheed Martin Hanford Corporation (LMHC) and Tank Waste Remediation System (TWRS) in accordance with the sampling and analytical requirements specified in the Waste Retrieval Sluicing System Vapor Sampling and Analysis Plan (SAP) for Evaluation of Organic Emissions, Process Test Phase III, HNF-4212, Rev. 0-A, (LMHC, 1999). All samples were stored in a secured Radioactive Materials Area (RMA) until the samples were radiologically released and received by SAS for analysis. The Waste Sampling and Characterization Facility (WSCF) performed the radiological analyses. The samples were received on April 5, 1999.

LOCKREM, L.L.

1999-08-13T23:59:59.000Z

139

Neptunium Transport Behavior in the Vicinity of Underground Nuclear Tests at the Nevada Test Site  

SciTech Connect (OSTI)

We used short lived {sup 239}Np as a yield tracer and state of the art magnetic sector ICP-MS to measure ultra low levels of {sup 237}Np in a number of 'hot wells' at the Nevada National Security Site (NNSS), formerly known as the Nevada Test Site (NTS). The results indicate that {sup 237}Np concentrations at the Almendro, Cambric, Dalhart, Cheshire and Chancellor sites, are in the range of 3 x 10{sup -5} to 7 x 10{sup -2} pCi/L and well below the MCL for alpha emitting radionuclides (15 pCi/L) (EPA, 2009). Thus, while Np transport is believed to occur at the NNSS, activities are expected to be well below the regulatory limits for alpha-emitting radionuclides. We also compared {sup 237}Np concentration data to other radionuclides, including tritium, {sup 14}C, {sup 36}Cl, {sup 99}Tc, {sup 129}I, and plutonium, to evaluate the relative {sup 237}Np transport behavior. Based on isotope ratios relative to published unclassified Radiologic Source Terms (Bowen et al., 1999) and taking into consideration radionuclide distribution between melt glass, rubble and groundwater (IAEA, 1998), {sup 237}Np appears to be substantially less mobile than tritium and other non-sorbing radionuclides, as expected. However, this analysis also suggests that {sup 237}Np mobility is surprisingly similar to that of plutonium. The similar transport behavior of Np and Pu can be explained by one of two possibilities: (1) Np(IV) and Pu(IV) oxidation states dominate under mildly reducing NNSS groundwater conditions resulting in similar transport behavior or (2) apparent Np transport is the result of transport of its parent {sup 241}Pu and {sup 241}Am isotopes and subsequent decay to {sup 237}Np. Finally, measured {sup 237}Np concentrations were compared to recent Hydrologic Source Term (HST) models. The 237Np data collected from three wells in Frenchman Flat (RNM-1, RNM-2S, and UE-5n) are in good agreement with recent HST transport model predictions (Carle et al., 2005). The agreement provides confidence in the results of the predictive model. The comparison to Cheshire HST model predictions (Pawloski et al, 2001) is somewhat ambiguous due to the low concentration resolution of the particle transport model.

Zhao, P; Tinnacher, R M; Zavarin, M; Williams, R W; Kersting, A B

2010-12-03T23:59:59.000Z

140

Construction and early test results of waste transport in piping systems served by ULF water closets  

E-Print Network [OSTI]

The intent of this study was to determine if there is a correlation between discharge curves and venting on waste transport. Test stands were built to facilitate discharge curve and waste transport testing at the ESL Laboratory of Texas A&M. Tests...

Carrier, Jonathan Gerald

2003-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

DEVELOPMENT OF A NEW TYPE A(F)RADIOACTIVE MATERIAL PACKAGING FOR THE DEPARTMENT OF ENERGY  

SciTech Connect (OSTI)

In a coordinated effort, the Department of Transportation (DOT) and Nuclear Regulatory Commission (NRC) proposed the elimination of the Specification Packaging from 49 CFR 173.[1] In accordance with the Federal Register, issued on October 1, 2004, new fabrication of Specification Packages would no longer be authorized. In accordance with the NRC final rulemaking published January 26, 2004, Specification Packagings are mandated by law to be removed from service no later than October 1, 2008. This coordinated effort and resulting rulemaking initiated a planned phase out of Specification Type B and Type A fissile (F) material transportation packages within the Department of Energy (DOE) and its subcontractors. One of the Specification Packages affected by this regulatory change is the UN1A2 Specification Package, per DOT 49 CFR 173.417(a)(6). To maintain continuing shipments of DOE materials currently transported in UN1A2 Specification Package after the existing authorization expires, a replacement Type A(F) material packaging design is under development by the Savannah River National Laboratory. This paper presents a summary of the prototype design effort and testing of the new Type A(F) Package development for the DOE. This paper discusses the progress made in the development of a Type A Fissile Packaging to replace the expiring 49 CFR UN1A2 Specification Fissile Package. The Specification Package was mostly a single-use waste disposal container. The design requirements and authorized radioactive material contents of the UN1A2 Specification Package were defined in 49 CFR. A UN1A2 Specification Package was authorized to ship up to 350 grams of U-235 in any enrichment and in any non-pyrophoric form. The design was specified as a 55-gallon 1A2 drum overpack with a body constructed from 18 gauge steel with a 16 gauge drum lid. Drum closure was specified as a standard 12-gauge ring closure. The inner product container size was not specified but was listed as any container that met Specification 7A requirements per 49 CFR 178.350. Specification 7A containers were required to withstand Type A packaging tests required by 49CFR173.465 with compliance demonstrated through testing, analysis or similarity to other containers. The maximum weight of the 7A product container, the radioactive content, and any internal packaging was limited to 200 lbs. The total gross weight for the UN1A2 Specification Package was limited to 350 lbs. No additional restrictions were applied. Authorization for use did not require the UN1A2 Specification Package to be tested to the Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) required for performance based, Type A(F) packages certified by the NRC or DOE. The Type A(F) Packaging design discussed in this paper is required to be in compliance with the regulatory safety requirements defined in Code of Federal Regulations (CFR) 10 CFR 71.41 through 71.47 and 10 CFR71.71. Sub-criticality of content must be maintained under the Hypothetical Accident Conditions specified under 10 CFR71.73. These federal regulations, and other applicable DOE Orders and Guides, govern design requirements for a Type A(F) package. Type A(F) packages with less than an A2 quantity of radioactive material are not required to have a leak testable boundary. With this exception a Type A(F) package design is subject to the same test requirements set forth for the design of a performance based Type B packaging.

Blanton, P.; Eberl, K.

2008-09-14T23:59:59.000Z

142

CH Packaging Maintenance Manual  

SciTech Connect (OSTI)

This procedure provides instructions for performing inner containment vessel (ICV) and outer containment vessel (OCV) maintenance and periodic leakage rate testing on the following packaging seals and corresponding seal surfaces using a nondestructive helium (He) leak test. In addition, this procedure provides instructions for performing ICV and OCV structural pressure tests.

Washington TRU Solutions

2002-01-02T23:59:59.000Z

143

Transportable Heavy Duty Emissions Testing Laboratory and Research Program  

SciTech Connect (OSTI)

The objective of this program was to quantify the emissions from heavy-duty vehicles operating on alternative fuels or advanced fuel blends, often with novel engine technology or aftertreatment. In the first year of the program West Virginia University (WVU) researchers determined that a transportable chassis dynamometer emissions measurement approach was required so that fleets of trucks and buses did not need to be ferried across the nation to a fixed facility. A Transportable Heavy-Duty Vehicle Emissions Testing Laboratory (Translab) was designed, constructed and verified. This laboratory consisted of a chassis dynamometer semi-trailer and an analytic trailer housing a full scale exhaust dilution tunnel and sampling system which mimicked closely the system described in the Code of Federal Regulations for engine certification. The Translab was first used to quantify emissions from natural gas and methanol fueled transit buses, and a second Translab unit was constructed to satisfy research demand. Subsequent emissions measurement was performed on trucks and buses using ethanol, Fischer-Tropsch fuel, and biodiesel. A medium-duty chassis dynamometer was also designed and constructed to facilitate research on delivery vehicles in the 10,000 to 20,000lb range. The Translab participated in major programs to evaluate low-sulfur diesel in conjunction with passively regenerating exhaust particulate filtration technology, and substantial reductions in particulate matter were recorded. The researchers also participated in programs to evaluate emissions from advanced natural gas engines with closed loop feedback control. These natural gas engines showed substantially reduced levels of oxides of nitrogen. For all of the trucks and buses characterized, the levels of carbon monoxide, oxides of nitrogen, hydrocarbons, carbon dioxide and particulate matter were quantified, and in many cases non-regulated species such as aldehydes were also sampled. Particle size was also quantified during selected studies. A laboratory was established at WVU to provide for studies which supported and augmented the Translab research, and to provide for development of superior emissions measurement systems. This laboratory research focused on engine control and fuel sulfur issues. In recent years, as engine and aftertreatment technologies advanced, emissions levels were reduced such that they were at or below the Translab detectable limits, and in the same time frame the US Environmental Protection Agency required improved measurement methodologies for engine emissions certification. To remain current and relevant, the researchers designed a new Translab analytic system, housed in a container which can be transported on a semi-trailer. The new system's dilution tunnel flow was designed to use a subsonic venturi with closed loop control of blower speed, and the secondary dilution and particulate matter filter capture were designed to follow new EPA engine certification procedures. A further contribution of the program has been the development of techniques for creating heavy-duty vehicle test schedules, and the creation of schedules to mimic a variety of truck and bus vocations.

David Lyons

2008-03-31T23:59:59.000Z

144

The Packaging Handbook -- A guide to package design  

SciTech Connect (OSTI)

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

Shappert, L.B.

1995-12-31T23:59:59.000Z

145

Productivity Techniques and Quality Aspects in the Criticality Safety Evaluation of Y-12 Type-B Fissile Material Packages  

SciTech Connect (OSTI)

The inventory of certified Type-B fissile material packages consists of ten performance-based packages for offsite transportation purposes, serving transportation programs at the Y-12 National Security Complex. The containment vessels range from 5 to 19 in. in diameter and from 17 to 58 in. in height. The drum assembly external to the containment vessel ranges from 18 to 34 in. in diameter and from 26 to 71 in. in height. The weight of the packaging (drum assembly and containment vessel) ranges from 239 to 1550 lb. The older DT-nn series of Cellotex-based packages are being phased-out and replaced by a new generation of Kaolite-based ('Y-12 patented insulation') packages capable of withstanding the dynamic crush test 10 CFR 71.73(c)(2). Three replacement packages are in various stages of development; two are in use. The U.S. Department of Transportation (DOT) 6M specification package, which does not conform to the U.S. Nuclear Regulatory Commission requirements for Type-B packages, is no longer authorized for service on public roads. The ES-3100 shipping package is an example of a Kaolite-based Type-B fissile material package developed as a replacement package for the DOT 6M. With expanded utility, the ES-3100 is designed and licensed for transporting highly enriched uranium and plutonium materials on public roads. The ES-3100 provides added capability for air transport of up to 7-kg quantities of uranium material. This paper presents the productivity techniques and quality aspects in the criticality safety evaluation of Y-12 packages using the ES-3100 as an example.

DeClue, J. F.

2011-06-28T23:59:59.000Z

146

LONG TERM AGING AND SURVEILLANCE OF 9975 PACKAGE COMPONENTS  

SciTech Connect (OSTI)

The mission of the 9975 package, originally designed only for transportation of radioactive materials, has been broadened to include storage at the Savannah River Site. Two components of this package, namely the containment vessel O-rings and fiberboard overpack, require continued integrity assessment under the storage conditions. The performance of the components over time is being evaluated using accelerated-aging studies. Compression stress relaxation (CSR) and leak testing are being used to measure the performance of O-rings. The performance of the fiberboard is being evaluated using compression strength, thermal conductivity, specific heat capacity and other physical properties. Models developed from the data collected provide an initial prediction of service life for the two components, and support the conclusion that normal service conditions will not degrade the performance of the package beyond specified functional requirements for the first assessment interval. Increased confidence in this conclusion is derived from field surveillance data and destructive evaluation of packages removed from storage.

Hoffman, E.; Skidmore, E.; Daugherty, W.; Dunn, K.

2009-11-10T23:59:59.000Z

147

Technical Review Report for the Model 9978-96 Package Safety Analysis Report for Packaging (S-SARP-G-00002, Revision 1, March 2009)  

SciTech Connect (OSTI)

This Technical Review Report (TRR) documents the review, performed by Lawrence Livermore National Laboratory (LLNL) Staff, at the request of the Department of Energy (DOE), on the 'Safety Analysis Report for Packaging (SARP), Model 9978 B(M)F-96', Revision 1, March 2009 (S-SARP-G-00002). The Model 9978 Package complies with 10 CFR 71, and with 'Regulations for the Safe Transport of Radioactive Material-1996 Edition (As Amended, 2000)-Safety Requirements', International Atomic Energy Agency (IAEA) Safety Standards Series No. TS-R-1. The Model 9978 Packaging is designed, analyzed, fabricated, and tested in accordance with Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME B&PVC). The review presented in this TRR was performed using the methods outlined in Revision 3 of the DOE's 'Packaging Review Guide (PRG) for Reviewing Safety Analysis Reports for Packages'. The format of the SARP follows that specified in Revision 2 of the Nuclear Regulatory Commission's Regulatory Guide 7.9, i.e., 'Standard Format and Content of Part 71 Applications for Approval of Packages for Radioactive Material'. Although the two documents are similar in their content, they are not identical. Formatting differences have been noted in this TRR, where appropriate. The Model 9978 Packaging is a single containment package, using a 5-inch containment vessel (5CV). It uses a nominal 35-gallon drum package design. In comparison, the Model 9977 Packaging uses a 6-inch containment vessel (6CV). The Model 9977 and Model 9978 Packagings were developed concurrently, and they were referred to as the General Purpose Fissile Material Package, Version 1 (GPFP). Both packagings use General Plastics FR-3716 polyurethane foam as insulation and as impact limiters. The 5CV is used as the Primary Containment Vessel (PCV) in the Model 9975-96 Packaging. The Model 9975-96 Packaging also has the 6CV as its Secondary Containment Vessel (SCV). In comparison, the Model 9975 Packagings use Celotex{trademark} for insulation and as impact limiters. To provide a historical perspective, it is noted that the Model 9975-96 Packaging is a 35-gallon drum package design that has evolved from a family of packages designed by DOE contractors at the Savannah River Site. Earlier package designs, i.e., the Model 9965, the Model 9966, the Model 9967, and the Model 9968 Packagings, were originally designed and certified in the early 1980s. In the 1990s, updated package designs that incorporated design features consistent with the then-newer safety requirements were proposed. The updated package designs at the time were the Model 9972, the Model 9973, the Model 9974, and the Model 9975 Packagings, respectively. The Model 9975 Package was certified by the Packaging Certification Program, under the Office of Safety Management and Operations. The Model 9978 Package has six Content Envelopes: C.1 ({sup 238}Pu Heat Sources), C.2 ( Pu/U Metals), C.3 (Pu/U Oxides, Reserved), C.4 (U Metal or Alloy), C.5 (U Compounds), and C.6 (Samples and Sources). Per 10 CFR 71.59 (Code of Federal Regulations), the value of N is 50 for the Model 9978 Package leading to a Criticality Safety Index (CSI) of 1.0. The Transport Index (TI), based on dose rate, is calculated to be a maximum of 4.1.

West, M

2009-03-06T23:59:59.000Z

148

Strategy for experimental validation of waste package performance assessment  

SciTech Connect (OSTI)

A strategy for the experimental validation of waste package performance assessment has been developed as part of a program supported by the Repository Technology Program. The strategy was developed by reviewing the results of laboratory analog experiments, in-situ tests, repository simulation tests, and material interaction tests. As a result of the review, a listing of dependent and independent variables that influence the ingress of water into the near-field environment, the reaction between water and the waste form, and the transport of radionuclides from the near-field environment was developed. The variables necessary to incorporate into an experimental validation strategy were chosen by identifying those which had the greatest effect of each of the three major events, i.e., groundwater ingress, waste package reactions, and radionuclide transport. The methodology to perform validation experiments was examined by utilizing an existing laboratory analog approach developed for unsaturated testing of glass waste forms. 185 refs., 9 figs., 2 tabs.

Bates, J.K.; Abrajano, T.A. Jr.; Wronkiewicz, D.J.; Gerding, T.J.; Seils, C.A.

1990-07-01T23:59:59.000Z

149

DEVELOPMENT OF THE BULK TRITIUM SHIPPING PACKAGING  

SciTech Connect (OSTI)

A new radioactive shipping packaging for transporting bulk quantities of tritium, the Bulk Tritium Shipping Package (BTSP), has been designed for the Department of Energy (DOE) as a replacement for a package designed in the early 1970s. This paper summarizes significant design features and describes how the design satisfies the regulatory safety requirements of the Code of Federal Regulations and the International Atomic Energy Agency. The BTSP design incorporates many improvements over its predecessor by implementing improved testing, handling, and maintenance capabilities, while improving manufacturability and incorporating new engineered materials. This paper also discusses the results from testing of the BTSP to 10 CFR 71 Normal Conditions of Transport and Hypothetical Accident Condition events. The programmatic need of the Department of Energy (DOE) to ship bulk quantities of tritium has been satisfied since the late 1970s by the UC-609 shipping package. The current Certificate of Conformance for the UC-609, USA/9932/B(U) (DOE), will expire in late 2011. Since the UC-609 was not designed to meet current regulatory requirements, it will not be recertified and thereby necessitates a replacement Type B shipping package for continued DOE tritium shipments in the future. A replacement tritium packaging called the Bulk Tritium Shipping Package (BTSP) is currently being designed and tested by Savannah River National Laboratory (SRNL). The BTSP consists of two primary assemblies, an outer Drum Assembly and an inner Containment Vessel Assembly (CV), both designed to mitigate damage and to protect the tritium contents from leaking during the regulatory Hypothetical Accident Condition (HAC) events and during Normal Conditions of Transport (NCT). During transport, the CV rests on a silicone pad within the Drum Liner and is covered with a thermal insulating disk within the insulated Drum Assembly. The BTSP packaging weighs approximately 500 lbs without contents and is 50-1/2 inches high by 24-1/2 inches in outside diameter. With contents the gross weight of the BTSP is 650 lbs. The BTSP is designed for the safe shipment of 150 grams of tritium in a solid or gaseous state. To comply with the federal regulations that govern Type B shipping packages, the BTSP is designed so that it will not lose tritium at a rate greater than the limits stated in 10CFR 71.51 of 10{sup -6} A2 per hour for the 'Normal Conditions of Transport' (NCT) and an A2 in 1 week under 'Hypothetical Accident Conditions' (HAC). Additionally, since the BTSP design incorporates a valve as part of the tritium containment boundary, secondary containment features are incorporated in the CV Lid to protect against gas leakage past the valve as required by 10CFR71.43(e). This secondary containment boundary is designed to provide the same level of containment as the primary containment boundary when subjected to the HAC and NCT criteria.

Blanton, P.; Eberl, K.

2008-09-14T23:59:59.000Z

150

User's manual for ONEDANT: a code package for one-dimensional, diffusion-accelerated, neutral-particle transport  

SciTech Connect (OSTI)

ONEDANT is designed for the CDC-7600, but the program has been implemented and run on the IBM-370/190 and CRAY-I computers. ONEDANT solves the one-dimensional multigroup transport equation in plane, cylindrical, spherical, and two-angle plane geometries. Both regular and adjoint, inhomogeneous and homogeneous (k/sub eff/ and eigenvalue search) problems subject to vacuum, reflective, periodic, white, albedo, or inhomogeneous boundary flux conditions are solved. General anisotropic scattering is allowed and anisotropic inhomogeneous sources are permitted. ONEDANT numerically solves the one-dimensional, multigroup form of the neutral-particle, steady-state form of the Boltzmann transport equation. The discrete-ordinates approximation is used for treating the angular variation of the particle distribution and the diamond-difference scheme is used for phase space discretization. Negative fluxes are eliminated by a local set-to-zero-and-correct algorithm. A standard inner (within-group) iteration, outer (energy-group-dependent source) iteration technique is used. Both inner and outer iterations are accelerated using the diffusion synthetic acceleration method. (WHK)

O'Dell, R.D.; Brinkley, F.W. Jr.; Marr, D.R.

1982-02-01T23:59:59.000Z

151

COMPACTION OF FIBERBOARD IN A 9975 SHIPPING PACKAGE  

SciTech Connect (OSTI)

Compaction of lower layers in the fiberboard overpack has been observed in 9975 packages that contain elevated moisture. Lab testing has resulted in a better understanding of (1) the relationship between the fiberboard moisture level and compaction of the lower fiberboard assembly, and (2) the behavior of the fiberboard during transport. In laboratory tests, higher moisture content has been shown to correspond to higher total compaction of fiberboard material, greater rate of compaction, and continued compaction over a longer period of time. In addition, laboratory tests have shown that the application of a dynamic load results in higher fiberboard compaction. The test conditions and sample geometric/loading configurations were chosen to simulate the regulatory requirements for 9975 package input dynamic loading. Dynamic testing was conducted over a period of several months to acquire immediate and cumulative changes in geometric data for various moisture levels. Currently, one sample set has undergone a complete dynamic test regimen, while testing of another set is still in-progress. The dynamic input, data acquisition, test effects on sample dynamic parameters, and interim results from this test program will be summarized and compared to regulatory specifications for dynamic loading. This will provide a basis from which to evaluate the impact of moisture and fiberboard compaction on the safety basis for transportation (Safety Analysis Report for Packaging) and storage (facility Documented Safety Analysis) at the Savannah River Site (SRS).

Stefek, T.; Daugherty, W.; Estochen, E.; Leduc, D.

2011-05-11T23:59:59.000Z

152

Test particle study of ion transport in drift type turbulence  

SciTech Connect (OSTI)

Ion transport regimes in drift type turbulence are determined in the frame of a realistic model for the turbulence spectrum based on numerical simulations. The model includes the drift of the potential with the effective diamagnetic velocity, turbulence anisotropy, and dominant waves. The effects of the zonal flow modes are also analyzed. A semi-analytical method that is able to describe trajectory stochastic trapping or eddying is used for obtaining the transport coefficients as function of the parameters of the turbulence. Analytical approximations of the transport coefficients are derived from the results. They show the transition from Bohm to gyro-Bohm scaling as plasma size increases in very good agreement with the numerical simulations.

Vlad, M.; Spineanu, F. [National Institute of Laser, Plasma and Radiation Physics, Association EURATOM-MEdC, Atomistilor 409, 077125 Magurele, Bucharest (Romania)] [National Institute of Laser, Plasma and Radiation Physics, Association EURATOM-MEdC, Atomistilor 409, 077125 Magurele, Bucharest (Romania)

2013-12-15T23:59:59.000Z

153

MODEL 9977 B(M)F-96 SAFETY ANALYSIS REPORT FOR PACKAGING  

SciTech Connect (OSTI)

This Safety Analysis Report for Packaging (SARP) documents the analysis and testing performed on and for the 9977 Shipping Package, referred to as the General Purpose Fissile Package (GPFP). The performance evaluation presented in this SARP documents the compliance of the 9977 package with the regulatory safety requirements for Type B packages. Per 10 CFR 71.59, for the 9977 packages evaluated in this SARP, the value of ''N'' is 50, and the Transport Index based on nuclear criticality control is 1.0. The 9977 package is designed with a high degree of single containment. The 9977 complies with 10 CFR 71 (2002), Department of Energy (DOE) Order 460.1B, DOE Order 460.2, and 10 CFR 20 (2003) for As Low As Reasonably Achievable (ALARA) principles. The 9977 also satisfies the requirements of the Regulations for the Safe Transport of Radioactive Material--1996 Edition (Revised)--Requirements. IAEA Safety Standards, Safety Series No. TS-R-1 (ST-1, Rev.), International Atomic Energy Agency, Vienna, Austria (2000). The 9977 package is designed, analyzed and fabricated in accordance with Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, 1992 edition.

Abramczyk, G; Paul Blanton, P; Kurt Eberl, K

2006-05-18T23:59:59.000Z

154

NREL: Transportation Research - Innovative Way to Test Batteries...  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Innovative Way to Test Batteries Fills a Market Niche A square piece of machinery with a lid that opens upwards NETZSCH's Isothermal Battery Calorimeter (IBC 284), developed by...

155

Evaluation and compilation of DOE waste package test data; Volume 8: Biannual report, August 1989--January 1990  

SciTech Connect (OSTI)

This report summarizes evaluations by the National Institute of Standards and Technology (NIST) of some of the Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW) for the six-month period, August 1989--January 1990. This includes reviews of related materials research and plans, information on the Yucca Mountain, Nevada disposal site activities, and other information regarding supporting research and special assistance. Short discussions are given relating to the publications reviewed and complete reviews and evaluations are included. Reports of other work are included in the Appendices.

Interrante, C.G. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of High-Level Waste Management; Fraker, A.C.; Escalante, E. [National Inst. of Standards and Technology (MSEL), Gaithersburg, MD (United States). Metallurgy Div.

1993-06-01T23:59:59.000Z

156

Experiment Safety Assurance Package for the 40- to 50-GWd/MT Burnup Phase of Mixed Oxide Fuel Irradiation in Small I-Hole Positions in the Advanced Test Reactor  

SciTech Connect (OSTI)

This experiment safety assurance package (ESAP) is a revision of the last MOX ESAP issued in February 2001(Khericha 2001). The purpose of this revision is to identify the changes in the loading pattern and to provide a basis to continue irradiation up to {approx}42 GWd/MT burnup (+ 2.5%) as predicted by MCNP (Monte Carlo N-Particle) transport code before the preliminary postirradiation examination (PIE) results for 40 GWd/MT burnup are available. Note that the safety analysis performed for the last ESAP is still applicable and no additional analysis is required (Khericha 2001). In July 2001, it was decided to reconfigure the test assembly using the loading pattern for Phase IV, Part 3, at the end of Phase IV, Part 1, as the loading pattern for Phase IV, Parts 2 and 3. Three capsule assemblies will be irradiated until the highest burnup capsule assembly accumulates: {approx}50 GWd/MT burnup, based on the MCNP code predictions. The last ESAP suggests that at the end of Phase IV, Part 1, we remove the two highest burnup capsule assemblies ({at} {approx}40 GWd/MT burnup) and send them to ORNL for PIE. Then, irradiate the test assembly using the loading pattern for Phase IV, Part 2, until the highest burnup capsule reaches {approx}40 GWd/MT burnup per MCNP-predicted values.

Khericha, S.T.

2002-06-30T23:59:59.000Z

157

Experiment Safety Assurance Package for the 40- to 50-GWd/MT Burnup Phase of Mixed Oxide Fuel Irradiation in Small I-Hole Positions in the Advanced Test Reactor  

SciTech Connect (OSTI)

This experiment safety assurance package (ESAP) is a revision of the last MOX ESAP issued in February 2001(Khericha 2001). The purpose of this revision is to identify the changes in the loading pattern and to provide a basis to continue irradiation up to ~42 GWd/MT burnup (+ 2.5% as predicted by MCNP (Monte Carlo N-Particle) transport code before the preliminary postirradiation examination (PIE) results for 40 GWd/MT burnup are available. Note that the safety analysis performed for the last ESAP is still applicable and no additional analysis is required (Khericha 2001). In July 2001, it was decided to reconfigure the test assembly using the loading pattern for Phase IV, Part 3, at the end of Phase IV, Part 1, as the loading pattern for Phase IV, Parts 2 and 3. Three capsule assemblies will be irradiated until the highest burnup capsule assembly accumulates: ~50 GWd/MT burnup, based on the MCNP code predictions. The last ESAP suggests that at the end of Phase IV, Part 1, we remove the two highest burnup capsule assemblies (@ ~40 GWd/MT burnup) and send them to ORNL for PIE. Then, irradiate the test assembly using the loading pattern for Phase IV, Part 2, until the highest burnup capsule reaches ~40 GWd/MT burnup per MCNP-predicted values.

Khericha, Soli T

2002-06-01T23:59:59.000Z

158

Lessons Learned in the Design and Use of IP1 / IP2 Flexible Packaging - 13621  

SciTech Connect (OSTI)

For many years in the USA, Low Level Radioactive Waste (LLW), contaminated soils and construction debris, have been transported, interim stored, and disposed of, using IP1 / IP2 metal containers. The performance of these containers has been more than adequate, with few safety occurrences. The containers are used under the regulatory oversight of the US Department of Transportation (DOT), 49 Code of Federal Regulations (CFR). In the late 90's the introduction of flexible packaging for the transport, storage, and disposal of low level contaminated soils and construction debris was introduced. The development of flexible packaging came out of a need for a more cost effective package, for the large volumes of waste generated by the decommissioning of many of the US Department of Energy (DOE) legacy sites across the US. Flexible packaging had to be designed to handle a wide array of waste streams, including soil, gravel, construction debris, and fine particulate dust migration. The design also had to meet all of the IP1 requirements under 49CFR 173.410, and be robust enough to pass the IP2 testing 49 CFR 173.465 required for many LLW shipments. Tens of thousands of flexible packages have been safely deployed and used across the US nuclear industry as well as for hazardous non-radioactive applications, with no recorded release of radioactive materials. To ensure that flexible packages are designed properly, the manufacturer must use lessons learned over the years, and the tests performed to provide evidence that these packages are suitable for transporting low level radioactive wastes. The design and testing of flexible packaging for LLW, VLLW and other hazardous waste streams must be as strict and stringent as the design and testing of metal containers. The design should take into consideration the materials being loaded into the package, and should incorporate the right materials, and manufacturing methods, to provide a quality, safe product. Flexible packaging can be shown to meet the criteria for safe and fit for purpose packaging, by meeting the US DOT regulations, and the IAEA Standards for IP-1 and IP-2 including leak tightness. (authors)

Sanchez, Mike [VP Global Sales, PacTec, Inc. (United States)] [VP Global Sales, PacTec, Inc. (United States); Reeves, Wendall [National Sales Manager, PacTec, Inc. (United States)] [National Sales Manager, PacTec, Inc. (United States); Smart, Bill [Nuclear Sales Director, PacTec, Inc. (United States)] [Nuclear Sales Director, PacTec, Inc. (United States)

2013-07-01T23:59:59.000Z

159

Safety analysis report for packaging: the ORNL lithium hydroxide fire and impact shield  

SciTech Connect (OSTI)

The ORNL Lithium Hydroxide Fire and Impact Shield and its packaging were designed and fabricated at Oak Ridge National Laboratory to permit the transport of Type B quantities of radioactive material and limited quantities of fissionable material. The shield and its packaging were evaluated analytically and experimentally to determine its compliance with the applicable regulations governing containers in which radioactive and fissile materials are transported, and that evaluation is the subject of this report. Computational and test procedures were used to determine the structural integrity and thermal behavior of the shield relative to the general standards for normal conditions of transport and the standards for the hypothetical accident conditions. The results of the evaluation demonstrate that the shield and its packaging are in compliance with the applicable regulations. 16 references, 8 figures, 5 tables.

Evans, J.H.; Eversole, R.E.; Just, R.A.; Schaich, R.W.

1984-07-01T23:59:59.000Z

160

Ceremony Theming Packages Package One  

E-Print Network [OSTI]

will incur a surcharge of $50.00 per additional half hour Weddings at The University of Western Australia #12: _____________________________________________________________________________________________________ Weddings at The University of Western Australia #12;Ceremony Theming Packages Payment Details CREDIT CARD I, ______________________________________, authorise The University Club of Western Australia P/L, to debit the amount

Tobar, Michael

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
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161

High Efficiency Integrated Package  

SciTech Connect (OSTI)

Solid-state lighting based on LEDs has emerged as a superior alternative to inefficient conventional lighting, particularly incandescent. LED lighting can lead to 80 percent energy savings; can last 50,000 hours – 2-50 times longer than most bulbs; and contains no toxic lead or mercury. However, to enable mass adoption, particularly at the consumer level, the cost of LED luminaires must be reduced by an order of magnitude while achieving superior efficiency, light quality and lifetime. To become viable, energy-efficient replacement solutions must deliver system efficacies of ? 100 lumens per watt (LPW) with excellent color rendering (CRI > 85) at a cost that enables payback cycles of two years or less for commercial applications. This development will enable significant site energy savings as it targets commercial and retail lighting applications that are most sensitive to the lifetime operating costs with their extended operating hours per day. If costs are reduced substantially, dramatic energy savings can be realized by replacing incandescent lighting in the residential market as well. In light of these challenges, Cree proposed to develop a multi-chip integrated LED package with an output of > 1000 lumens of warm white light operating at an efficacy of at least 128 LPW with a CRI > 85. This product will serve as the light engine for replacement lamps and luminaires. At the end of the proposed program, this integrated package was to be used in a proof-of-concept lamp prototype to demonstrate the component’s viability in a common form factor. During this project Cree SBTC developed an efficient, compact warm-white LED package with an integrated remote color down-converter. Via a combination of intensive optical, electrical, and thermal optimization, a package design was obtained that met nearly all project goals. This package emitted 1295 lm under instant-on, room-temperature testing conditions, with an efficacy of 128.4 lm/W at a color temperature of ~2873K and 83 CRI. As such, the package’s performance exceeds DOE’s warm-white phosphor LED efficacy target for 2013. At the end of the program, we assembled an A19 sized demonstration bulb housing the integrated package which met Energy Star intensity variation requirements. With further development to reduce overall component cost, we anticipate that an integrated remote converter package such as developed during this program will find application in compact, high-efficacy LED-based lamps, particularly those requiring omnidirectional emission.

Ibbetson, James

2013-09-15T23:59:59.000Z

162

DWPF (Defense Waste Processing Facility) canister impact testing and analyses for the Transportation Technology Center  

SciTech Connect (OSTI)

A legal weight truck cask design has been developed for the US Department of Energy by GA Technologies, Inc. The cask will be used to transport defense high-level waste canisters produced by the Defense Waste Processing Facility (DWPF) at the Savannah River Plant. The development of the cask required the collection of impact data for the DWPF canisters. The Materials Characterization Center (MCC) performed this work under the guidance of the Transportation Technology Center (TTC) at Sandia National Laboratories. Two full-scale DWPF canisters filled with nonradioactive borosilicate glass were impacted under ''normal'' and ''hypothetical'' accident conditions. Two canisters, supplied by the DWPF, were tested. Each canister was vertically dropped on the bottom end from a height of either 0.3 m or 9.1 m (for normal or hypothetical accident conditions, respectively). The structural integrity of each canister was then examined using helium leak and dye penetrant testing. The canisters' diameters and heights, which had been previously measured, were then remeasured to determine how the canister dimensions had changed. Following structural integrity testing, the canisters were flaw leak tested. For transportation flaw leak testing, four holes were fabricated into the shell of canister A-27 (0.3 m drop height). The canister was then transported a total distance of 2069 miles. During transport, the waste form material that fell from each flaw was collected to determine the amount of size distribution of each flaw release. 2 refs., 8 figs., 12 tabs.

Farnsworth, R.K.; Mishima, J.

1988-12-01T23:59:59.000Z

163

NREL: Transportation Research - Fleet Test and Evaluation Publications  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the Contributions andData andFleet Test and Evaluation Photo of medium-duty truck

164

NREL: Transportation Research - Hybrid Electric Fleet Vehicle Testing  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the Contributions andData andFleet Test and Evaluation Photo of

165

NREL: Transportation Research - Hydraulic Hybrid Fleet Vehicle Testing  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the Contributions andData andFleet Test and Evaluation Photo ofHydraulic Hybrid Fleet

166

Determing Degradation Of Fiberboard In The 9975 Shipping Package By Measuring Axial Gap  

SciTech Connect (OSTI)

Currently, thousands of model 9975 transportation packages are in use by the US Department of Energy (DOE); the design of which has been certified by DOE for shipment of Type B radioactive and fissile materials in accordance with Part 71, Title 10 Code of Federal Regulations (CFR), or 10 CFR 71, Packaging and Transportation of Radioactive Material. These transportation packages are also approved for the storage of DOE-STD-3013 containers at the Savannah River Site (SRS). As such, the 9975 has been continuously exposed to the service environment for a period of time greater than the approved transportation service life. In order to ensure the material integrity as specified in the safety basis, an extensive surveillance program is in place in K-Area Complex (KAC) to monitor the structural and thermal properties of the fiberboard of the 9975 shipping packages. The surveillance approach uses a combination of Non-Destructive Examination (NDE) field surveillance and Destructive Examination (DE) lab testing to validate the 9975 performance assumptions. The fiberboard in the 9975 is credited with thermal insulation, criticality control and resistance to crushing. During surveillance monitoring in KAC, an increased axial gap of the fiberboard was discovered on selected items packaged at Rocky Flats Environmental Technology Site (RFETS). Many of these packages were later found to contain excess moisture. Savannah River National Laboratory (SRNL) testing has resulted in a better understanding of the relationship between the fiberboard moisture level and compaction of the fiberboard under storage conditions and during transport. In laboratory testing, the higher moisture content has been shown to correspond to higher total compaction of fiberboard material and compaction rate. The fiberboard height is reduced by compression of the layers. This change is observed directly in the axial gap between the flange and the air shield. The axial gap measurement is made during the pre-use inspection or during the annual recertification process and is a screening measurement for changes in the fiberboard.

Hackney, E. R.; Dougherty, W. L.; Dunn, K. A.; Stefek, T. M

2013-08-01T23:59:59.000Z

167

Regional groundwater flow and tritium transport modeling and risk assessment of the underground test area, Nevada Test Site, Nevada  

SciTech Connect (OSTI)

The groundwater flow system of the Nevada Test Site and surrounding region was evaluated to estimate the highest potential current and near-term risk to the public and the environment from groundwater contamination downgradient of the underground nuclear testing areas. The highest, or greatest, potential risk is estimated by assuming that several unusually rapid transport pathways as well as public and environmental exposures all occur simultaneously. These conservative assumptions may cause risks to be significantly overestimated. However, such a deliberate, conservative approach ensures that public health and environmental risks are not underestimated and allows prioritization of future work to minimize potential risks. Historical underground nuclear testing activities, particularly detonations near or below the water table, have contaminated groundwater near testing locations with radioactive and nonradioactive constituents. Tritium was selected as the contaminant of primary concern for this phase of the project because it is abundant, highly mobile, and represents the most significant contributor to the potential radiation dose to humans for the short term. It was also assumed that the predicted risk to human health and the environment from tritium exposure would reasonably represent the risk from other, less mobile radionuclides within the same time frame. Other contaminants will be investigated at a later date. Existing and newly collected hydrogeologic data were compiled for a large area of southern Nevada and California, encompassing the Nevada Test Site regional groundwater flow system. These data were used to develop numerical groundwater flow and tritium transport models for use in the prediction of tritium concentrations at hypothetical human and ecological receptor locations for a 200-year time frame. A numerical, steady-state regional groundwater flow model was developed to serve as the basis for the prediction of the movement of tritium from the underground testing areas on a regional scale. The groundwater flow model was used in conjunction with a particle-tracking code to define the pathlines followed by groundwater particles originating from 415 points associated with 253 nuclear test locations. Three of the most rapid pathlines were selected for transport simulations. These pathlines are associated with three nuclear test locations, each representing one of the three largest testing areas. These testing locations are: BOURBON on Yucca Flat, HOUSTON on Central Pahute Mesa, and TYBO on Western Pahute Mesa. One-dimensional stochastic tritium transport simulations were performed for the three pathlines using the Monte Carlo method with Latin hypercube sampling. For the BOURBON and TYBO pathlines, sources of tritium from other tests located along the same pathline were included in the simulations. Sensitivity analyses were also performed on the transport model to evaluate the uncertainties associated with the geologic model, the rates of groundwater flow, the tritium source, and the transport parameters. Tritium concentration predictions were found to be mostly sensitive to the regional geology in controlling the horizontal and vertical position of transport pathways. The simulated concentrations are also sensitive to matrix diffusion, an important mechanism governing the migration of tritium in fractured carbonate and volcanic rocks. Source term concentration uncertainty is most important near the test locations and decreases in importance as the travel distance increases. The uncertainty on groundwater flow rates is as important as that on matrix diffusion at downgradient locations. The risk assessment was performed to provide conservative and bounding estimates of the potential risks to human health and the environment from tritium in groundwater. Risk models were designed by coupling scenario-specific tritium intake with tritium dose models and cancer and genetic risk estimates using the Monte Carlo method. Estimated radiation doses received by individuals from chronic exposure to tritium, and the corre

None

1997-10-01T23:59:59.000Z

168

Praxis I/O package  

SciTech Connect (OSTI)

The Praxis language specification, like Algol and Ada, does not specify any I/O statements. The intent was to provide a standard I/O package as a companion to the compiler. This would allow the user to substitute, or supplement, the I/O package, as needed, for specialized applications. Like Algol, however, Praxis provided only limited (text) I/O for several years. Ada, in contrast, provided a comprehensive standard I/O package from its inception. Digital Equipment Corporation's (DEC's) implementation of Ada, on their VAX family of computers, further supplemented this package with other packages which exploit the I/O facilities available under the VMS operating system. The Praxis I/O package described in this document has been modeled after DEC's implementation of Ada and provides a similar set of I/O facilities. Currently, the I/O package is supported only under VAX/VMS. The design of the package, however, is essentially independent of any operating system (with the exception of the module COMMAND IO). The VAX/VMS version of the I/O package fully exploits the vast I/O facilities which are provided under VAX/VMS and makes them directly available to the Praxis programmer. The design, prototype implementation, and draft documentation of the Praxis I/O Package was done by Tim Sherman as part of a University project in computer science. Subsequent work by both Tim and Fred Holloway lead to a more complete implementation, testing and development of example programs, and inclusion of the package into the Praxis compilers as their principal interface to RMS and VMS.

Holloway, F.W.; Sherman, T.A.

1988-04-07T23:59:59.000Z

169

The Use of the Hanford Onsite Packaging and Transportation Safety Program to Meet Cleanup Milestones Under the Hanford Site Cleanup 2015 Vision and the American Recovery and Reinvestment Act of 2009 - 12403  

SciTech Connect (OSTI)

The Hanford Site presents unique challenges in meeting the U.S. Department of Energy Richland Operations Office (DOE-RL) 2015 Cleanup Vision. CH2M Hill Plateau Remediation Company (CHPRC), its subcontractors, and DOE-RL were challenged to retrieve, transport and remediate a wide range of waste materials. Through a collaborative effort by all Hanford Onsite Central Plateau Cleanup Team Members, disposition pathways for diverse and seemingly impossible to ship wastes were developed under a DOE Order 460.1C-compliant Hanford Onsite Transportation Safety Program. The team determined an effective method for transporting oversized compliant waste payloads to processing and disposition facilities. The use of the onsite TSD packaging authorizations proved to be vital to safely transporting these materials for processing and eventual final disposition. The American Recovery and Reinvestment Act of 2009 (ARRA) provided additional resources to expedite planning and execution of these important cleanup milestones. Through the innovative and creative use of the TSD, the Hanford Onsite Central Plateau Cleanup Team Members have developed and are executing an integrated project plan that enables the safe and compliant transport of a wide variety of difficult-to-transport waste items, accelerating previous cleanup schedules to meet cleanup milestones. (authors)

Lavender, John C. [CH2M HILL Plateau Remediation Company, Richland, WA 99354 (United States); Edwards, W. Scott [Areva Federal Services, Richland, WA 99354 (United States); Macbeth, Paul J.; Self, Richard J. [U.S. Department of Energy Richland Operations Office, Richland, WA 99352 (United States); West, Lori D. [Materials and Energy Corporation, Richland, WA 99354 (United States)

2012-07-01T23:59:59.000Z

170

WASTE PACKAGE REMEDIATION SYSTEM DESCRIPTION DOCUMENT  

SciTech Connect (OSTI)

The Waste Package Remediation System remediates waste packages (WPs) and disposal containers (DCs) in one of two ways: preparation of rejected DC closure welds for repair or opening of the DC/WP. DCs are brought to the Waste Package Remediation System for preparation of rejected closure welds if testing of the closure weld by the Disposal Container Handling System indicates an unacceptable, but repairable, welding flaw. DC preparation of rejected closure welds will require removal of the weld in such a way that the Disposal Container Handling System may resume and complete the closure welding process. DCs/WPs are brought to the Waste Package Remediation System for opening if the Disposal Container Handling System testing of the DC closure weld indicates an unrepairable welding flaw, or if a WP is recovered from the subsurface repository because suspected damage to the WP or failure of the WP has occurred. DC/WP opening will require cutting of the DC/WP such that a temporary seal may be installed and the waste inside the DC/WP removed by another system. The system operates in a Waste Package Remediation System hot cell located in the Waste Handling Building that has direct access to the Disposal Container Handling System. One DC/WP at a time can be handled in the hot cell. The DC/WP arrives on a transfer cart, is positioned within the cell for system operations, and exits the cell without being removed from the cart. The system includes a wide variety of remotely operated components including a manipulator with hoist and/or jib crane, viewing systems, machine tools for opening WPs, and equipment used to perform pressure and gas composition sampling. Remotely operated equipment is designed to facilitate DC/WP decontamination and hot cell equipment maintenance, and interchangeable components are provided where appropriate. The Waste Package Remediation System interfaces with the Disposal Container Handling System for the receipt and transport of WPs and DCs. The Waste Handling Building System houses the system, and provides the facility, safety, and auxiliary systems required to support operations. The system receives power from the Waste Handling Building Electrical System. The system also interfaces with the various DC systems.

N.D. Sudan

2000-06-22T23:59:59.000Z

171

Safety evaluation for packaging (onsite) product removal can containers  

SciTech Connect (OSTI)

This safety evaluation for packaging allows the transport of nine Product Removal (PR) Cans with their Containers from the PUREX Facility to the Plutonium Finishing Plant.

Boettger, J.S.

1997-04-29T23:59:59.000Z

172

Water Transport in PEM Fuel Cells: Advanced Modeling, Material Selection, Testing,  

E-Print Network [OSTI]

Optimization J. Vernon Cole and Ashok Gidwani CFDRC Prepared for: DOE Hydrogen Fuel Cell Kickoff MeetingWater Transport in PEM Fuel Cells: Advanced Modeling, Material Selection, Testing, and Design fuel cell design and operation; Demonstrate improvements in water management resulting in improved

173

Human Portable Radiation Detection System Communications Package Evaluation  

SciTech Connect (OSTI)

Testing and valuation of the Human Portable Radiation Detection System Communications Package for the US Coast Guard.

Morgen, Gerald P.; Peterson, William W.

2009-06-11T23:59:59.000Z

174

Experiment Safety Assurance Package for Mixed Oxide Fuel Irradiation in an Average Power Position (I-24) in the Advanced Test Reactor  

SciTech Connect (OSTI)

The Fissile Material Disposition Program Light Water Reactor Mixed Oxide Fuel Irradiation Test Project Plan details a series of test irradiations designed to investigate the use of weapons-grade plutonium in MOX fuel for light water reactors (LWR) (Cowell 1996a, Cowell 1997a, Thoms 1997a). Commercial MOX fuel has been successfully used in overseas reactors for many years; however, weapons-derived test fuel contains small amounts of gallium (about 2 parts per million). A concern exists that the gallium may migrate out of the fuel and into the clad, inducing embrittlement. For preliminary out-of-pile experiments, Wilson (1997) states that intermetallic compound formation is the principal interaction mechanism between zircaloy cladding and gallium. This interaction is very limited by the low mass of gallium, so problems are not expected with the zircaloy cladding, but an in-pile experiment is needed to confirm the out-of-pile experiments. Ryskamp (1998) provides an overview of this experiment and its documentation. The purpose of this Experiment Safety Assurance Package (ESAP) is to demonstrate the safe irradiation and handling of the mixed uranium and plutonium oxide (MOX) Fuel Average Power Test (APT) experiment as required by Advanced Test Reactor (ATR) Technical Safety Requirement (TSR) 3.9.1 (LMITCO 1998). This ESAP addresses the specific operation of the MOX Fuel APT experiment with respect to the operating envelope for irradiation established by the Upgraded Final Safety Analysis Report (UFSAR) Lockheed Martin Idaho Technologies Company (LMITCO 1997a). Experiment handling activities are discussed herein.

J. M . Ryskamp; R. C. Howard; R. C. Pedersen; S. T. Khericha

1998-10-01T23:59:59.000Z

175

Packaging Review Guide for Reviewing Safety Analysis Reports for Packagings  

SciTech Connect (OSTI)

This Packaging Review Guide (PRG) provides guidance for Department of Energy (DOE) review and approval of packagings to transport fissile and Type B quantities of radioactive material. It fulfills, in part, the requirements of DOE Order 460.1B for the Headquarters Certifying Official to establish standards and to provide guidance for the preparation of Safety Analysis Reports for Packagings (SARPs). This PRG is intended for use by the Headquarters Certifying Official and his or her review staff, DOE Secretarial offices, operations/field offices, and applicants for DOE packaging approval. This PRG is generally organized at the section level in a format similar to that recommended in Regulatory Guide 7.9 (RG 7.9). One notable exception is the addition of Section 9 (Quality Assurance), which is not included as a separate chapter in RG 7.9. Within each section, this PRG addresses the technical and regulatory bases for the review, the manner in which the review is accomplished, and findings that are generally applicable for a package that meets the approval standards. This Packaging Review Guide (PRG) provides guidance for DOE review and approval of packagings to transport fissile and Type B quantities of radioactive material. It fulfills, in part, the requirements of DOE O 460.1B for the Headquarters Certifying Official to establish standards and to provide guidance for the preparation of Safety Analysis Reports for Packagings (SARPs). This PRG is intended for use by the Headquarters Certifying Official and his review staff, DOE Secretarial offices, operations/field offices, and applicants for DOE packaging approval. The primary objectives of this PRG are to: (1) Summarize the regulatory requirements for package approval; (2) Describe the technical review procedures by which DOE determines that these requirements have been satisfied; (3) Establish and maintain the quality and uniformity of reviews; (4) Define the base from which to evaluate proposed changes in scope and requirements of reviews; and (5) Provide the above information to DOE organizations, contractors, other government agencies, and interested members of the general public. This PRG was originally published in September 1987. Revision 1, issued in October 1988, added new review sections on quality assurance and penetrations through the containment boundary, along with a few other items. Revision 2 was published October 1999. Revision 3 of this PRG is a complete update, and supersedes Revision 2 in its entirety.

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

2007-04-12T23:59:59.000Z

176

Evaluation of groundwater flow and transport at the Shoal underground nuclear test: An interim report  

SciTech Connect (OSTI)

Since 1962, all United States nuclear tests have been conducted underground. A consequence of this testing has been the deposition of large amounts of radioactive materials in the subsurface, sometimes in direct contact with groundwater. The majority of this testing occurred on the Nevada Test Site, but a limited number of experiments were conducted in other locations. One of these is the subject of this report, the Project Shoal Area (PSA), located about 50 km southeast of Fallon, Nevada. The Shoal test consisted of a 12-kiloton-yield nuclear detonation which occurred on October 26, 1963. Project Shoal was part of studies to enhance seismic detection of underground nuclear tests, in particular, in active earthquake areas. Characterization of groundwater contamination at the Project Shoal Area is being conducted by the US Department of Energy (DOE) under the Federal Facility Agreement and Consent Order (FFACO) with the State of Nevada Department of Environmental Protection and the US Department of Defense (DOD). This order prescribes a Corrective Action Strategy (Appendix VI), which, as applied to underground nuclear tests, involves preparing a Corrective Action Investigation Plan (CAIP), Corrective Action Decision Document (CADD), Corrective Action Plan, and Closure Report. The scope of the CAIP is flow and transport modeling to establish contaminant boundaries that are protective of human health and the environment. This interim report describes the current status of the flow and transport modeling for the PSA.

Pohll, G.; Chapman, J.; Hassan, A.; Papelis, C.; Andricevic, R.; Shirley, C.

1998-07-01T23:59:59.000Z

177

Final evaluation report for Lockheed Idaho Technologies Company, ARROW-PAK packaging, Docket 95-40-7A, Type A container  

SciTech Connect (OSTI)

The report documents the U.S. Department of Transportation Specification 7A Type A (DOT-7A) compliance test results of the ARROW-PAK packaging. The ARROW-PAK packaging system consists of Marlex M-8000 Driscopipe (Series 8000 [gas] or Series 8600 [industrial]) resin pipe, manufactured by Phillips-Driscopipe, Inc., and is sealed with two dome-shaped end caps manufactured from the same materials. The patented sealing process involves the use of electrical energy to heat opposing faces of the pipe and end caps, and hydraulic rams to press the heated surfaces together. This fusion process produces a homogeneous bonding of the end cap to the pipe. The packaging may be used with or without the two internal plywood spacers. This packaging was evaluated and tested in October 1995. The packaging configuration described in this report is designed to ship Type A quantities of solid radioactive materials, Form No. 1, Form No. 2, and Form No. 3.

Kelly, D.L.

1995-11-01T23:59:59.000Z

178

Transportation  

E-Print Network [OSTI]

Transportation in ancient Egypt entailed the use of boats2007 Land transport in Roman Egypt: A study of economics andDieter 1991 Building in Egypt: Pharaonic stone masonry. New

Vinson, Steve

2013-01-01T23:59:59.000Z

179

RH Packaging Program Guidance  

SciTech Connect (OSTI)

The purpose of this program guidance document is to provide technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the SARP and/or C of C shall govern. The C of C states: ''...each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, ''Operating Procedures,'' of the application.'' It further states: ''...each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, ''Acceptance Tests and Maintenance Program of the Application.'' Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M&O) contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC approved, users need to be familiar with 10 CFR {section} 71.11, ''Deliberate Misconduct.'' Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. This document details the instructions to be followed to operate, maintain, and test the RH-TRU 72-B packaging. This Program Guidance standardizes instructions for all users. Users shall follow these instructions. Following these instructions assures that operations are safe and meet the requirements of the SARP. This document is available on the Internet at: ttp://www.ws/library/t2omi/t2omi.htm. Users are responsible for ensuring they are using the current revision and change notices. Sites may prepare their own document using the word-for-word steps in th is document, in sequence, including Notes and cautions. Site specific information may be included as necessary. The document, and revisions, must then be submitted to CBFO at sitedocuments@wipp.ws for approval. A copy of the approval letter from CBFO shall be available for audit purposes. Users may develop site-specific procedures addressing preoperational activities, quality assurance (QA), hoisting and rigging, and radiation health physics to be used with the instructions contained in this document. Users may recommend changes to this document by submitting their recommendations (in writing) to the WIPP M&O Contractor RH Packaging Maintenance Engineer for evaluation. If approved, the change(s) will be incorporated into this document for use by ALL users. Before first use and every 12 months after, user sites will be audited to this document to ensure compliance. They will also be audited within one year from the effective date of revisions to this document.

Washington TRU Solutions, LLC

2003-08-25T23:59:59.000Z

180

Safety analysis report for packaging (onsite) steel drum  

SciTech Connect (OSTI)

This Safety Analysis Report for Packaging (SARP) provides the analyses and evaluations necessary to demonstrate that the steel drum packaging system meets the transportation safety requirements of HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments, for an onsite packaging containing Type B quantities of solid and liquid radioactive materials. The basic component of the steel drum packaging system is the 208 L (55-gal) steel drum.

McCormick, W.A.

1998-09-29T23:59:59.000Z

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

STATUS REPORT FOR MOISTURE EFFECTS ON COMPACTION OF FIBERBOARD IN A 9975 SHIPPING PACKAGE  

SciTech Connect (OSTI)

Compaction of lower layers in the fiberboard overpack has been observed in 9975 packages that contain elevated moisture. Lab testing has resulted in a better understanding of the relationship between the fiberboard moisture level and compaction of the lower fiberboard assembly, and the behavior of the fiberboard during transport. In laboratory tests, higher moisture content has been shown to correspond to higher total compaction of fiberboard material, greater rate of compaction, and continued compaction over a longer period of time. In addition, laboratory tests have shown that the application of a dynamic load results in higher fiberboard compaction. The test conditions and sample geometric/loading configurations were chosen to simulate the regulatory requirements for 9975 package input dynamic loading. Dynamic testing was conducted over a period of six months to acquire immediate and cumulative changes in geometric data for various moisture levels. Currently, one sample set has undergone a complete dynamic test regimen, while testing of another set is still in-progress. The dynamic input, data acquisition, test effects on sample dynamic parameters, and interim results from this test program are summarized and compared to regulatory specifications for dynamic loading. This will provide a basis from which to evaluate the impact of moisture and fiberboard compaction on the safety basis for transportation (Safety Analysis Report for Packaging) and storage (facility Documented Safety Analysis) at the Savannah River Site (SRS).

Stefek, T.; Daugherty, W.; Estochen, E.

2011-06-23T23:59:59.000Z

182

ASSESSING EXPOSURE TO THE PUBLIC FROM LOW LEVEL RADIOACTIVE WASTE (LLW) TRANSPORTATION TO THE NEVADA TEST SITE.  

SciTech Connect (OSTI)

The United States (U.S.) Department of Energy (DOE) Nevada Test Site (NTS) is one of two regional sites where low-level radioactive waste (LLW) from approved DOE and U.S. DOD generators across the United States is disposed. In federal fiscal year (FY) 2002, over 57,000 cubic meters of waste was transported to and disposed at the NTS. DOE and U.S. Department of Transportation (DOT) regulations ensure that radiation exposure from truck shipments to members of the public is negligible. Nevertheless, particularly in rural communities along transportation routes in Utah and Nevada, there is perceived risk from members of the public about incremental exposure from LLW trucks, especially when ''Main Street'' and the LLW transportation route are the same. To better quantify the exposure to gamma radiation, a stationary monitoring array of four pressurized ion chambers (PICs) have been set up in a pullout just before LLW trucks reach the entrance to the NTS. The PICs are positioned at a distance of one meter from the sides of the truck trailer and at a height appropriate for the design of the trucks that will be used in FY2003 to haul LLW to the NTS. The use of four PICs (two on each side of the truck) is to minimize and to correct for non-uniformity where radiation levels from waste packages vary from side to side, and from front to back in the truck trailer. The PIC array is being calibrated by collecting readings from each PIC exposed to a known 137Cs source that was positioned at different locations on a flatbed stationed in the PIC array, along with taking secondary readings from other known sources. Continuous data collection using the PICs, with and without a truck in the array, is being used to develop background readings. In addition, acoustic sensors are positioned on each side of the PIC array to record when a large object (presumably a truck) enters the array. In FY2003, PIC surveys from as many incoming LLW trucks as possible will be made and survey data recorded automatically by dataloggers that will be periodically downloaded. Solar panels provide power for the batteries to run both the dataloggers and PICs. Truck drivers have been asked to park their truck within the PIC array for only the time it takes to complete an information log before moving on to one of two Radioactive Waste Management Sites (RWMS) on the NTS. On the log, the truck drivers record their shipment identification number, the time of day, where the waste originated, and information on the route they used to reach the NTS. This data will facilitate comparison of PIC readings with waste manifests and other waste disposal operations data collected at the RWMSs. Gamma radiation measurements collected from the PICs will be analyzed using standard health physics and statistical methods for comparison to DOT standards, but with the added benefit of obtaining an improved understanding of the variability of readings that can occur in the near vicinity of a LLW truck. The data collected will be combined with measurements of street width and other information about transportation routes through towns to develop realistic dose scenarios for citizens in Nevada and Utah towns.

Miller, J.J.; Campbell, S.; Church, B.W.; Shafer, D. S.; Gillespie, D.; Sedano, S.; Cebe, J.J.

2003-02-27T23:59:59.000Z

183

Cost Estimation Package  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This chapter focuses on the components (or elements) of the cost estimation package and their documentation.

1997-03-28T23:59:59.000Z

184

The radioactive materials packaging handbook: Design, operations, and maintenance  

SciTech Connect (OSTI)

As part of its required activities in 1994, the US Department of Energy (DOE) made over 500,000 shipments. Of these shipments, approximately 4% were hazardous, and of these, slightly over 1% (over 6,400 shipments) were radioactive. Because of DOE`s cleanup activities, the total quantities and percentages of radioactive material (RAM) that must be moved from one site to another is expected to increase in the coming years, and these materials are likely to be different than those shipped in the past. Irradiated fuel will certainly be part of the mix as will RAM samples and waste. However, in many cases these materials will be of different shape and size and require a transport packaging having different shielding, thermal, and criticality avoidance characteristics than are currently available. This Handbook provides guidance on the design, testing, certification, and operation of packages for these materials.

Shappert, L.B.; Bowman, S.M. [Oak Ridge National Lab., TN (United States); Arnold, E.D. [Lockheed Martin Energy Systems, Oak Ridge, TN (United States)] [and others

1998-08-01T23:59:59.000Z

185

Transportation Management Workshop: Proceedings  

SciTech Connect (OSTI)

This report is a compilation of discussions presented at the Transportation Management Workshop held in Gaithersburg, Maryland. Topics include waste packaging, personnel training, robotics, transportation routing, certification, containers, and waste classification.

Not Available

1993-10-01T23:59:59.000Z

186

An improved environmental TLD field package  

SciTech Connect (OSTI)

This paper reports on the design of an environmental TAD field package which can have a significant impact on the accuracy and reliability of an environmental monitoring program. Ideally, a field package should protect its TAD(s) from light and moisture, but it should not expose the TAD(s) to elevated temperatures that could increase fading and invalidate calibration factors. Furthermore, a field package must be relatively strong and tamper-proof, while at the same time maintaining an open, unshielded configuration to minimize attenuation of incident radiation. These conflicting goals - protection without interference - can be satisfied with the improved field package design described her. This package has been tested with considerable success at the San Onofre Nuclear Generating Station, and a U.S. Patent Application has been filed in preparation for marketing the package as a commercial product.

Graham, B.D. (Southern California Edison Co., San Onofre Nuclear Generating Station, San Clemente, CA (US))

1986-01-01T23:59:59.000Z

187

Safety evaluation for packaging CPC metal boxes  

SciTech Connect (OSTI)

This Safety Evaluation for Packaging (SEP) provides authorization for the use of Container Products Corporation (CPC) metal boxes, as described in this document, for the interarea shipment of radioactive contaminated equipment and debris for storage in the Central Waste Complex (CWC) or T Plant located in the 200 West Area. Authorization is granted until November 30, 1995. The CPC boxes included in this SEP were originally procured as US Department of Transportation (DOT) Specification 7A Type A boxes. A review of the documentation provided by the manufacturer revealed the documentation did not adequately demonstrate compliance to the 4 ft drop test requirement of 49 CFR 173.465(c). Preparation of a SEP is necessary to document the equivalent safety of the onsite shipment in lieu of meeting DOT packaging requirements until adequate documentation is received. The equivalent safety of the shipment is based on the fact that the radioactive contents consist of contaminated equipment and debris which are not dispersible. Each piece is wrapped in two layers of no less than 4 mil plastic prior to being placed in the box which has an additional 10 mil liner. Pointed objects and sharp edges are padded to prevent puncture of the plastic liner and wrapping.

Romano, T.

1995-05-15T23:59:59.000Z

188

Development of a test system for verification and validation of nuclear transport simulations  

SciTech Connect (OSTI)

Verification and validation of nuclear data is critical to the accuracy of both stochastic and deterministic particle transport codes. In order to effectively test a set of nuclear data, the data must be applied to a wide variety of transport problems. Performing this task in a timely, efficient manner is tedious. The nuclear data team at Los Alamos National laboratory in collaboration with the University of Florida has developed a methodology to automate the process of nuclear data verification and validation (V and V). This automated V and V process can efficiently test a number of data libraries using well defined benchmark experiments, such as those in the International Criticality Safety Benchmark Experiment Project (ICSBEP). The process is implemented through an integrated set of Pyton scripts. Material and geometry data are read from an existing medium or given directly by the user to generate a benchmark experiment template file. The user specifies the choice of benchmark templates, codes, and libraries to form a V and V project. The Python scripts generate input decks for multiple transport codes from the templates, run and monitor individual jobs, and parse the relevant output automatically. The output can then be used to generate reports directly or can be stored into a database for later analysis. This methodology eases the burden on the user by reducing the amount of time and effort required for obtaining and compiling calculation results. The resource savings by using this automated methodology could potentially be an enabling technology for more sophisticated data studies, such as nuclear data uncertainty quantification. Once deployed, this tool will allow the nuclear data community to more thoroughly test data libraries leading to higher fidelity data in the future.

White, Morgan C [Los Alamos National Laboratory; Triplett, Brian S [GENERAL ELECTRIC; Anghaie, Samim [UNIV OF FL

2008-01-01T23:59:59.000Z

189

LOCFES-NL: a tool for testing nonlinear spatial approximations to neutron transport in plane-parallel geometry  

E-Print Network [OSTI]

of the requirements for the degree of MASTER OF SCIENCE December 1997 Major Subject:Nuclear Engineering LOCFES-NL: A TOOL FOR TESTING NONLINEAR SPATIAL APPROXIMATIONS TO NEUTRON TRANSPORT IN PLANE-PARALLEL GEOMETRY A Thesis by STEVEN DOUGLAS NOLEN Submitted...) John . Poston, Sr. (Head of Department) December 1997 Major Subject: Nuclear Engineering ABSTRACT LOCFES-NL: A Tool for Testing Nonlinear Spatial Approximations to Neutron Transport in Plane-Parallel Geometry. (December 1997) Steven Douglas Nolen...

Nolen, Steven Douglas

1997-01-01T23:59:59.000Z

190

Certification testing of the Los Alamos National Laboratory Heat Source/Radioisotopic Thermoelectric Generator shipping container  

SciTech Connect (OSTI)

The Heat Source/Radioisotopic Thermoelectric Generator shipping counter is a Type B packaging currently under development by Los Alamos National Laboratory. Type B packaging for transporting radioactive material is required to maintain containment and shielding after being exposed to normal and hypothetical accident environments defined in Title 10 of the Code of Federal Regulations Part 71. A combination of testing and analysis is used to verify the adequacy of this packaging design. This report documents the testing portion of the design verification. Six tests were conducted on a prototype package: a water spray test, a 4-foot normal conditions drop test, a 30-foot drop test, a 40-inch puncture test, a 30-minute thermal test, and an 8-hour immersion test.

Bronowski, D.R.; Madsen, M.M.

1991-09-01T23:59:59.000Z

191

A Validation Process for the Groundwater Flow and Transport Model of the Faultless Nuclear Test at Central Nevada Test Area  

SciTech Connect (OSTI)

Many sites of groundwater contamination rely heavily on complex numerical models of flow and transport to develop closure plans. This has created a need for tools and approaches that can be used to build confidence in model predictions and make it apparent to regulators, policy makers, and the public that these models are sufficient for decision making. This confidence building is a long-term iterative process and it is this process that should be termed ''model validation.'' Model validation is a process not an end result. That is, the process of model validation cannot always assure acceptable prediction or quality of the model. Rather, it provides safeguard against faulty models or inadequately developed and tested models. Therefore, development of a systematic approach for evaluating and validating subsurface predictive models and guiding field activities for data collection and long-term monitoring is strongly needed. This report presents a review of model validation studies that pertain to groundwater flow and transport modeling. Definitions, literature debates, previously proposed validation strategies, and conferences and symposia that focused on subsurface model validation are reviewed and discussed. The review is general in nature, but the focus of the discussion is on site-specific, predictive groundwater models that are used for making decisions regarding remediation activities and site closure. An attempt is made to compile most of the published studies on groundwater model validation and assemble what has been proposed or used for validating subsurface models. The aim is to provide a reasonable starting point to aid the development of the validation plan for the groundwater flow and transport model of the Faultless nuclear test conducted at the Central Nevada Test Area (CNTA). The review of previous studies on model validation shows that there does not exist a set of specific procedures and tests that can be easily adapted and applied to determine the validity of site-specific groundwater models. This is true for both deterministic and stochastic models, with the latter posing a more difficult and challenging problem when it comes to validation. This report then proposes a general validation approach for the CNTA model, which addresses some of the important issues recognized in previous validation studies, conferences, and symposia as crucial to the process. The proposed approach links model building, model calibration, model predictions, data collection, model evaluations, and model validation in an iterative loop. The approach focuses on use of collected validation data to reduce model uncertainty and narrow the range of possible outcomes of stochastic numerical models. It accounts for the stochastic nature of the numerical CNTA model, which used Monte Carlo simulation approach. The proposed methodology relies on the premise that absolute validity is not even a theoretical possibility and is not a regulatory requirement. Rather, it highlights the importance of testing as many aspects of the model as possible and using as many diverse statistical tools as possible for rigorous checking and confidence building in the model and its predictions. It is this confidence that will eventually allow for regulator and public acceptance of decisions based on the model predictions.

Ahmed Hassan

2003-01-01T23:59:59.000Z

192

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

SciTech Connect (OSTI)

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

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

2009-09-28T23:59:59.000Z

193

HYDROGEL TRACER BEADS: THE DEVELOPMENT, MODIFICATION, AND TESTING OF AN INNOVATIVE TRACER FOR BETTER UNDERSTANDING LNAPL TRANSPORT IN KARST AQUIFERS  

SciTech Connect (OSTI)

The goal of this specific research task is to develop proxy tracers that mimic contaminant movement to better understand and predict contaminant fate and transport in karst aquifers. Hydrogel tracer beads are transported as a separate phase than water and can used as a proxy tracer to mimic the transport of non-aqueous phase liquids (NAPL). They can be constructed with different densities, sizes & chemical attributes. This poster describes the creation and optimization of the beads and the field testing of buoyant beads, including sampling, tracer analysis, and quantitative analysis. The buoyant beads are transported ahead of the dissolved solutes, suggesting that light NAPL (LNAPL) transport in karst may occur faster than predicted from traditional tracing techniques. The hydrogel beads were successful in illustrating this enhanced transport.

Amanda Laskoskie, Harry M. Edenborn, and Dorothy J. Vesper

2012-01-01T23:59:59.000Z

194

Completion of the Radioactive Materials Packaging Handbook  

SciTech Connect (OSTI)

The Radioactive Materials Packaging Handbook: Design, Operation and Maintenance, which will serve as a replacement for the Cask Designers Guide (Shappert, 1970), has now been completed and submitted to the Oak Ridge National Laboratory (ORNL) electronics publishing group for layout and printing; it is scheduled to be printed in late spring 1998. The Handbook, written by experts in their particular fields, is a compilation of technical chapters that address the design aspects of a package intended for transporting radioactive material in normal commerce; it was prepared under the direction of M. E. Wangler of the US Department of Energy (DOE) and is intended to provide a wealth of technical guidance that will give designers a better understanding of the regulatory approval process, preferences of regulators on specific aspects of package design, and the types of analyses that should be considered when designing a package to carry radioactive materials.

Shappert, L.B.

1998-02-01T23:59:59.000Z

195

Standard practice for qualification and acceptance of boron based metallic neutron absorbers for nuclear criticality control for dry cask storage systems and transportation packaging  

E-Print Network [OSTI]

1.1 This practice provides procedures for qualification and acceptance of neutron absorber materials used to provide criticality control by absorbing thermal neutrons in systems designed for nuclear fuel storage, transportation, or both. 1.2 This practice is limited to neutron absorber materials consisting of metal alloys, metal matrix composites (MMCs), and cermets, clad or unclad, containing the neutron absorber boron-10 (10B). 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

American Society for Testing and Materials. Philadelphia

2007-01-01T23:59:59.000Z

196

Transportation  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLas ConchasPassiveSubmittedStatusButler TinaContact-Information-Transmission SignTransport

197

Atmospheric Data Package for the Composite Analysis  

SciTech Connect (OSTI)

The purpose of this data package is to summarize our conceptual understanding of atmospheric transport and deposition, describe how this understanding will be simplified for numerical simulation as part of the Composite Analysis (i.e., implementation model), and finally to provide the input parameters needed for the simulations.

Napier, Bruce A.; Ramsdell, James V.

2005-09-01T23:59:59.000Z

198

Flow and transport simulations using T2CG1, a package of conjugate gradient solvers for the TOUGH2 family of codes  

SciTech Connect (OSTI)

This report discusses the details of modifications made to the TOUGH2 family of codes to complement its direct solver which significantly increases the size of problems solved by the TOUGH2 code. With this modification, the TOUGH2 system is being tested in multiphase, multicomponent fluid and heat flow problems related to vadose zone hydrology, nuclear waste disposal, and environmental remediation.

Moridis, G.; Pruess, K.

1995-04-01T23:59:59.000Z

199

Experiment Safety Assurance Package for the 40- to 52-GWd/MT Burnup Phase of Mixed Oxide Fuel Irradiation in Small I-hole Positions in the Advanced Test Reactor  

SciTech Connect (OSTI)

This experiment safety assurance package (ESAP) is a revision of the last mixed uranium and plutonium oxide (MOX) ESAP issued in June 2002). The purpose of this revision is to provide a basis to continue irradiation up to 52 GWd/MT burnup [as predicted by MCNP (Monte Carlo N-Particle) transport code The last ESAP provided basis for irradiation, at a linear heat generation rate (LHGR) no greater than 9 kW/ft, of the highest burnup capsule assembly to 50 GWd/MT. This ESAP extends the basis for irradiation, at a LHGR no greater than 5 kW/ft, of the highest burnup capsule assembly from 50 to 52 GWd/MT.

S. T. Khericha; R. C. Pedersen

2003-09-01T23:59:59.000Z

200

CH Packaging Operations Manual  

SciTech Connect (OSTI)

This procedure provides instructions forassembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP)

Washington TRU Solutions LLC

2007-11-29T23:59:59.000Z

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

CH Packaging Operations Manual  

SciTech Connect (OSTI)

This procedure provides instructions forassembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP)

Washington TRU Solutions LLC

2007-08-22T23:59:59.000Z

202

CH Packaging Operations Manual  

SciTech Connect (OSTI)

Introduction - This procedure provides instructions for assembling the following CH packaging payload: -Drum payload assembly -Standard Waste Box (SWB) assembly -Ten-Drum Overpack (TDOP).

Washington TRU Solutions LLC

2003-06-26T23:59:59.000Z

203

CH Packaging Operations Manual  

SciTech Connect (OSTI)

Introduction - This procedure provides instructions forassembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP)

Washington TRU Solutions LLC

2007-05-15T23:59:59.000Z

204

CH Packaging Operations Manual  

SciTech Connect (OSTI)

Introduction - This procedure provides instructions forassembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP)

Washington TRU Solutions LLC

2006-12-18T23:59:59.000Z

205

Packaging Design Criteria for the MCO Cask  

SciTech Connect (OSTI)

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

FLANAGAN, B.D.

2000-08-01T23:59:59.000Z

206

FINAL REPORT FOR MOISTURE EFFECTS ON COMPACTION OF FIBERBOARD IN A 9975 SHIPPING PACKAGE  

SciTech Connect (OSTI)

Compaction of lower layers in the fiberboard assembly has been observed in 9975 packages that contain elevated moisture. Lab testing has resulted in a better understanding of the relationship between the fiberboard moisture level and compaction of the lower fiberboard assembly, and the behavior of the fiberboard during transport. In laboratory tests of cane fiberboard, higher moisture content has been shown to correspond to higher total compaction, greater rate of compaction, and continued compaction over a longer period of time. In addition, laboratory tests have shown that the application of a dynamic load results in higher fiberboard compaction compared to a static load. The test conditions and sample geometric/loading configurations were chosen to simulate the regulatory requirements for 9975 package input dynamic loading. Dynamic testing was conducted to acquire immediate and cumulative changes in geometric data for various moisture levels. Two sample sets have undergone a complete dynamic test regimen, one set for 27 weeks, and the second set for 47 weeks. The dynamic input, data acquisition, test effects on sample dynamic parameters, and results from this test program are summarized and compared to regulatory specifications for dynamic loading. Compaction of the bottom fiberboard layers due to the accumulation of moisture is one possible cause of an increase in the axial gap at the top of the package. The net compaction of the bottom layers will directly add to the axial gap. The moisture which caused this compaction migrated from the middle region of the fiberboard assembly (which is typically the hottest). This will cause the middle region to shrink axially, which will also contribute directly to the axial gap. Measurement of the axial gap provides a screening tool for identifying significant change in the fiberboard condition. The data in this report provide a basis to evaluate the impact of moisture and fiberboard compaction on 9975 package performance during storage at the Savannah River Site (SRS).

Stefek, T.; Daugherty, W.; Estochen, E.

2013-09-17T23:59:59.000Z

207

EBS Radionuclide Transport Abstraction  

SciTech Connect (OSTI)

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 from a breached waste package. Advective transport occurs when radionuclides that are dissolved or sorbed onto colloids (or both) are carried from the waste package by the portion of the seepage flux that passes through waste package breaches. Diffusive transport occurs as a result of a gradient in radionuclide concentration and may take place while advective transport is also occurring, as well as when no advective transport is occurring. Diffusive transport is addressed in detail because it is the sole means of transport when there is no flow through a waste package, which may dominate during the regulatory compliance period in the nominal and seismic scenarios. The advective transport rate, when it occurs, is generally greater than the diffusive transport rate. Colloid-facilitated advective and diffusive transport is also modeled and is presented in detail in Appendix B of this report.

J. Prouty

2006-07-14T23:59:59.000Z

208

Safety evaluation for packaging (onsite) SERF cask  

SciTech Connect (OSTI)

This safety evaluation for packaging (SEP) documents the ability of the Special Environmental Radiometallurgy Facility (SERF) Cask to meet the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B quantities (up to highway route controlled quantities) of radioactive material within the 300 Area of the Hanford Site. This document shall be used to ensure that loading, tie down, transport, and unloading of the SERF Cask are performed in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required.

Edwards, W.S.

1997-10-24T23:59:59.000Z

209

Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment  

SciTech Connect (OSTI)

This data package documents the experimentally derived input data on the representative waste glasses; LAWA44, LAWB45, and LAWC22. This data will be used for Subsurface Transport Over Reactive Multi-phases (STORM) simulations of the Integrated Disposal Facility (IDF) for immobilized low-activity waste (ILAW). The STORM code will be used to provide the near-field radionuclide release source term for a performance assessment to be issued in July 2005. Documented in this data package are data related to 1) kinetic rate law parameters for glass dissolution, 2) alkali (Na+)-hydrogen (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 (PUF) and product consistency (PCT) tests where used for accelerated weathering or aging of the glasses in order to determine a chemical reaction network of secondary phases that form. The majority of the thermodynamic data used in this data package were extracted from the thermody-namic database package shipped with the geochemical code EQ3/6, version 8.0. Because of the expected importance of 129I release from secondary waste streams being sent to IDF from various thermal treatment processes, parameter estimates for diffusional release and solubility-controlled release from cementitious waste forms were estimated from the available literature.

Pierce, Eric M.; McGrail, B. Peter; Rodriguez, Elsa A.; Schaef, Herbert T.; Saripalli, Prasad; Serne, R. Jeffrey; Krupka, Kenneth M.; Martin, P. F.; Baum, Steven R.; Geiszler, Keith N.; Reed, Lunde R.; Shaw, Wendy J.

2004-09-01T23:59:59.000Z

210

Annotated bibliography of literature relating to wind transport of plutonium-contaminated soils at the Nevada Test Site  

SciTech Connect (OSTI)

During the period from 1954 through 1963, a number of tests were conducted on the Nevada Test Site (NTS) and Tonopah Test Range (TTR) to determine the safety of nuclear devices with respect to storage, handling, transport, and accidents. These tests were referred to as ``safety shots.`` ``Safety`` in this context meant ``safety against fission reaction.`` The safety tests were comprised of chemical high explosive detonations with components of nuclear devices. The conduct of these tests resulted in the dispersion of plutonium, and some americium over areas ranging from several tens to several hundreds of hectares. Of the various locations used for safety tests, the site referred to as ``Plutonium Valley`` was subject to a significant amount of plutonium contamination. Plutonium Valley is located in Area 11 on the eastern boundary of the NTS at an elevation of about 1036 m (3400 ft). Plutonium Valley was the location of four safety tests (A,B,C, and D) conducted during 1956. A major environmental, health, and safety concern is the potential for inhalation of Pu{sup 239,240} by humans as a result of airborne dust containing Pu particles. Thus, the wind transport of Pu{sup 239,240} particles has been the subject of considerable research. This annotated bibliography was created as a reference guide to assist in the better understanding of the environmental characteristics of Plutonium Valley, the safety tests performed there, the processes and variables involved with the wind transport of dust, and as an overview of proposed clean-up procedures.

Lancaster, N.; Bamford, R.

1993-12-01T23:59:59.000Z

211

Overview of Advanced Technology Transportation, 2005 Update. Advanced Vehicle Testing Activity  

SciTech Connect (OSTI)

Document provides an overview of the transportation market in 2005. Areas covered include hybrid, fuel cell, hydrogen, and alternative fuel vehicles.

Barnitt, R.; Eudy, L.

2005-08-01T23:59:59.000Z

212

Phase II Transport Model of Corrective Action Unit 98: Frenchman Flat, Nevada Test Site, Nye County, Nevada, Revision 1  

SciTech Connect (OSTI)

This document, the Phase II Frenchman Flat transport report, presents the results of radionuclide transport simulations that incorporate groundwater radionuclide transport model statistical and structural uncertainty, and lead to forecasts of the contaminant boundary (CB) for a set of representative models from an ensemble of possible models. This work, as described in the Federal Facility Agreement and Consent Order (FFACO) Underground Test Area (UGTA) strategy (FFACO, 1996; amended 2010), forms an essential part of the technical basis for subsequent negotiation of the compliance boundary of the Frenchman Flat corrective action unit (CAU) by Nevada Division of Environmental Protection (NDEP) and National Nuclear Security Administration Nevada Site Office (NNSA/NSO). Underground nuclear testing via deep vertical shafts was conducted at the Nevada Test Site (NTS) from 1951 until 1992. The Frenchman Flat area, the subject of this report, was used for seven years, with 10 underground nuclear tests being conducted. The U.S. Department of Energy (DOE), NNSA/NSO initiated the UGTA Project to assess and evaluate the effects of underground nuclear tests on groundwater at the NTS and vicinity through the FFACO (1996, amended 2010). The processes that will be used to complete UGTA corrective actions are described in the “Corrective Action Strategy” in the FFACO Appendix VI, Revision No. 2 (February 20, 2008).

Gregg Ruskuaff

2010-01-01T23:59:59.000Z

213

An experimental test plan for the characterization of molten salt thermochemical properties in heat transport systems  

SciTech Connect (OSTI)

Molten salts are considered within the Very High Temperature Reactor program as heat transfer media because of their intrinsically favorable thermo-physical properties at temperatures starting from 300 C and extending up to 1200 C. In this context two main applications of molten salt are considered, both involving fluoride-based materials: as primary coolants for a heterogeneous fuel reactor core and as secondary heat transport medium to a helium power cycle for electricity generation or other processing plants, such as hydrogen production. The reference design concept here considered is the Advanced High Temperature Reactor (AHTR), which is a large passively safe reactor that uses solid graphite-matrix coated-particle fuel (similar to that used in gas-cooled reactors) and a molten salt primary and secondary coolant with peak temperatures between 700 and 1000 C, depending upon the application. However, the considerations included in this report apply to any high temperature system employing fluoride salts as heat transfer fluid, including intermediate heat exchangers for gas-cooled reactor concepts and homogenous molten salt concepts, and extending also to fast reactors, accelerator-driven systems and fusion energy systems. The purpose of this report is to identify the technical issues related to the thermo-physical and thermo-chemical properties of the molten salts that would require experimental characterization in order to proceed with a credible design of heat transfer systems and their subsequent safety evaluation and licensing. In particular, the report outlines an experimental R&D test plan that would have to be incorporated as part of the design and operation of an engineering scaled facility aimed at validating molten salt heat transfer components, such as Intermediate Heat Exchangers. This report builds on a previous review of thermo-physical properties and thermo-chemical characteristics of candidate molten salt coolants that was generated as part of the same project [1]. However, this work focuses on two materials: the LiF-BeF2 eutectic (67 and 33 mol%, respectively, also known as flibe) as primary coolant and the LiF-NaF-KF eutectic (46.5, 11.5, and 52 mol%, respectively, also known as flinak) as secondary heat transport fluid. At first common issues are identified, involving the preparation and purification of the materials as well as the development of suitable diagnostics. Than issues specific to each material and its application are considered, with focus on the compatibility with structural materials and the extension of the existing properties database.

Pattrick Calderoni

2010-09-01T23:59:59.000Z

214

2 IEEE TRANSACTIONS ON COMPONENTS, PACKAGING, AND MANUFACTURING TECHNOLOGY--PART B, VOL. 20, NO. 1, FEBRUARY 1997 A Novel Test Technique for MCM Substrates  

E-Print Network [OSTI]

, FEBRUARY 1997 A Novel Test Technique for MCM Substrates Bruce Kim, Member, IEEE, Madhavan Swaminathan-- This paper describes a novel and low-cost test technique that is capable of detecting process related defects such as opens and shorts in multichip module (MCM) substrates. This method is an alternative to existing test

Swaminathan, Madhavan

215

Nuclear Material Packaging Manual  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The manual provides detailed packaging requirements for protecting workers from exposure to nuclear materials stored outside of an approved engineered contamination barrier. No cancellation. Certified 11-18-10.

2008-03-07T23:59:59.000Z

216

Packaging and Transportation | Department of Energy  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 - September 2006 TheSteven Ashby Dr. Steven Para9 Revision: 0Science

217

DOE-Idaho's Packaging and Transportation Perspective  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergy Cooperation |South42.2ConsolidatedDepartment2-93 JANUARY6.1148-2002i5-975504-95Idaho's

218

Packaging design criteria modified fuel spacer burial box. Revision 1  

SciTech Connect (OSTI)

Various Hanford facilities must transfer large radioactively contaminated items to burial/storage. Presently, there are eighteen Fuel Spacer Burial Boxes (FSBBs) available on the Hanford Site for transport of such items. Previously, the FSBBS were transported from a rail car to the burial trench via a drag-off operation. To allow for the lifting of the boxes into the burial trench, it will be necessary to improve the packagings lifting attachments and provide structural reinforcement. Additional safety improvements to the packaging system will be provided by the addition of a positive closure system and package ventilation. FSBBs that are modified in such a manner are referred to as Modified Fuel Spacer Burial Boxes (MFSBs). The criteria provided by this PDC will be used to demonstrate that the transfer of the MFSB will provide an equivalent degree of safety as would be provided by a package meeting offsite transportation requirements. This fulfills the onsite transportation safety requirements implemented in WHC-CM-2-14, Hazardous Material Packaging and Shipping. A Safety Analysis Report for Packaging (SARP) will be prepared to evaluate the safety of the transfer operation. Approval of the SARP is required to authorize transfer. Criteria are also established to ensure burial requirements are met.

Stevens, P.F.

1994-09-13T23:59:59.000Z

219

Safety analysis report for packaging (onsite) doorstop samplecarrier system  

SciTech Connect (OSTI)

The Doorstop Sample Carrier System consists of a Type B certified N-55 overpack, U.S. Department of Transportation (DOT) specification or performance-oriented 208-L (55-gal) drum (DOT 208-L drum), and Doorstop containers. The purpose of the Doorstop Sample Carrier System is to transport samples onsite for characterization. This safety analysis report for packaging (SARP) provides the analyses and evaluation necessary to demonstrate that the Doorstop Sample Carrier System meets the requirements and acceptance criteria for both Hanford Site normal transport conditions and accident condition events for a Type B package. This SARP also establishes operational, acceptance, maintenance, and quality assurance (QA) guidelines to ensure that the method of transport for the Doorstop Sample Carrier System is performed safely in accordance with WHC-CM-2-14, Hazardous Material Packaging and Shipping.

Obrien, J.H.

1997-02-24T23:59:59.000Z

220

Conceptual design and development of high-activity radioactive liquid packaging (summary)  

SciTech Connect (OSTI)

Environmental remediation and disposal of US Department of Energy radioactive liquid waste require analytical support, characterization, process development, testing, demonstration, and stabilization. In support of these diverse activities, there is a need to transport varying quantities of Type B high-activity liquid (HAL). To date, except for quantities of 50 ml (1.7 oz), there has never been, a US, Nuclear Regulatory Commission-licensed liquid Type B package available to support these remediation activities. In an effort to develop suitable packaging for large volumes of HAL, an investigation into packaging alternatives that would facilitate such transfers is under way. In, past and present studies, a spent fuel shipping cask fitted with a high-integrity pressure vessel has been determined to be the most viable concept for large volume HAL shipments. One concept that was investigated utilized the Pacific Nuclear 125-B shipping container and has been shown to meet the strUctural, thermal, shielding, and criticality conditions for HAL. The results of these investigations are being extended to develop the concept into the HAL packaging system.

Riley, D.L.; McCoy, J.C.; Edwards, W.S.

1994-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

THERMAL EVALUATION OF DRUM TYPE RADIOACTIVE MATERIAL PACKAGING ARRAYS IN STORAGE  

SciTech Connect (OSTI)

Drum type packages are routinely used to transport radioactive material (RAM) in the U.S. Department of Energy (DOE) complex. These packages are designed to meet the federal regulations described in 10 CFR 71.[1] In recent years, there has been a greater need to use these packagings to store the excess fissile material, especially plutonium for long term storage. While the design requirements for safe transportation of these packagings are well defined, the requirements for safe long term storage are not well established. Since the RAM contents in the packagings produce decay heat, it is important that they are stored carefully to prevent overheating of the containment vessel (CV) seals to prevent any leakage and the impact limiter to maintain the package structural integrity. This paper analyzes different storage arrays for a typical 9977 packaging for thermal considerations and makes recommendations for their safe storage under normal operating conditions.

Gupta, N

2009-04-27T23:59:59.000Z

222

Technical Review Report for the Mound 1KW Package Safety Analysis Report for Packaging Waiver for the Use of Modified Primary Containment Vessel (PCV)  

SciTech Connect (OSTI)

This Technical Review Report (TRR) documents the review, performed by the Lawrence Livermore National Laboratory (LLNL) staff, at the request of the U.S. Department of Energy (DOE), on the Waiver for the Use of Modified Primary Containment Vessels (PCV). The waiver is to be used to support a limited number of shipments of fuel for the Multi-Mission Radioisotope Thermoelectric Generator (MMRTG) Project in support of the National Aeronautics and Space Administration's (NASA's) Mars Science Laboratory (MSL) mission. Under the waiver, an inventory of existing national security PCVs will be converted to standard PCVs. Both types of PCVs are currently approved for use by the Office of Nuclear Energy. LLNL has previously reviewed the national security PCVs under Mound 1KW Package Safety Analysis Report for Packaging, Addendum No. 1, Revision c, dated June 2007 (Addendum 1). The safety analysis of the package is documented in the Safety Analysis Report for Packaging (SARP) for the Mound 1KW Package (i.e., the Mound 1KW SARP, or the SARP) where the standard PCVs have been reviewed by LLNL. The Mound 1KW Package is certified by DOE Certificate of Compliance (CoC) number USA/9516/B(U)F-85 for the transportation of Type B quantities of plutonium heat source material. The waiver requests an exemption, claiming safety equivalent to the requirements specified in 10 CFR 71.12, Specific Exemptions, and will lead to a letter amendment to the CoC. Under the waiver, the Office of Radioisotope Power Systems, NE-34, is seeking an exemption from 10 CFR 71.19(d)(1), Previously Approved Package,[5] which states: '(d) NRC will approve modifications to the design and authorized contents of a Type B package, or a fissile material package, previously approved by NRC, provided--(1) The modifications of a Type B package are not significant with respect to the design, operating characteristics, or safe performance of the containment system, when the package is subjected to the tests specified in {section}71.71 and 71.73.' The LLNL staff had previously reviewed a request from Idaho National Laboratory (INL) to reconfigure national security PCVs to standard PCVs. With a nominal 50% reduction in both the height and the volume, the LLNL staff initially deemed the modifications to be significant, which would not be allowed under the provisions of 10 CFR 71.19(d)(1)--see above. As a follow-up, the DOE requested additional clarification from the Nuclear Regulatory Commission (NRC). The NRC concluded that the reconfiguration would be a new fabrication, and that an exemption to the regulations would be required to allow its use, as per the requirements specified in 10 CFR 71.19(c)(1), Previously Approved Package: '(c) A Type B(U) package, a Type B(M) package, or a fissile material package previously approved by the NRC with the designation '-85' in the identification number of the NRC CoC, may be used under the general license of {section}71.17 with the following additional conditions: (1) Fabrication of the package must be satisfactorily completed by December 31, 2006, as demonstrated by application of its model number in accordance with 71.85(c).' Although the preferred approach toward the resolution of this issue would be for the applicant to submit an updated SARP, the applicant has stated that the process of updating the Model Mound 1KW Package SARP is a work that is in progress, but that the updated SARP is not yet ready for submittal. The applicant has to provide a submittal, proving that the package meets the '-96' requirements of International Atomic Energy Agency (IAEA) Safety Standards Series No. TS-R-1, in order to fabricate approved packagings after December 31, 2006. The applicant has further stated that all other packaging features, as described in the currently approved Model Mound 1KW Package SARP, remain unchanged. This report documents the LLNL review of the waiver request. The specific review for each SARP Chapter is documented.

West, M; Hafner, R

2008-05-05T23:59:59.000Z

223

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

SciTech Connect (OSTI)

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.

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

2001-02-01T23:59:59.000Z

224

River Data Package for the 2004 Composite Analysis  

SciTech Connect (OSTI)

Beginning in fiscal year 2003, the DOE Richland Operations Office initiated activities, including the development of data packages, to support the 2004 Composite Analysis. The river data package provides calculations of flow and transport in the Columbia River system. This document presents the data assembled to run the river module components for the section of the Columbia River from Vernita Bridge to the confluence with the Yakima River.

Rakowski, Cynthia L.; Guensch, Gregory R.; Patton, Gregory W.

2004-08-01T23:59:59.000Z

225

Packaging and Transfer of Materials of National Security Interest Manual  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This Technical Manual establishes requirements for operational safety controls for onsite operations and provides Department of Energy (DOE) technical safety requirements and policy objectives for development of an Onsite Packaging and Transfer Program, pursuant to DOE O 461.1A, Packaging and Transfer or Transportation of Materials of National Security Interest. The DOE contractor must document this program in its Onsite Packaging and Transfer Manual/Procedures. Admin Chg 1, 7-26-05. Certified 2-2-07. Canceled by DOE O 461.2.

2000-09-29T23:59:59.000Z

226

Subsurface mass transport affects the radioxenon signatures that are used to identify clandestine nuclear tests  

E-Print Network [OSTI]

nuclear tests J. D. Lowrey,1 S. R. Biegalski,1 A. G. Osborne,1 and M. R. Deinert1 Received 14 September can provide critical information with which to verify that a belowground nuclear test has taken place and decay rate. The xenon signature of a nuclear test is then bounded by the signal from directly produced

Deinert, Mark

227

State DOT: Ohio Department of Transportation State Report Questions on NDT Testing  

E-Print Network [OSTI]

testing methods for concrete materials, concrete pavements, and overlays are you trying? If there are low cylinder strengths on Structure concrete, our QC/QA specification requires that the contractor test used for our QC/QA spec. On the occasion where the in-place concrete has been tested using NDT on our

228

DESTRUCTIVE EXAMINATION OF SHIPPING PACKAGE 9975-03431  

SciTech Connect (OSTI)

Destructive and non-destructive examinations have been performed on specified components of shipping package 9975-03431. For those attributes that were also measured during the field surveillance, no significant changes were observed. All observations and test results met identified criteria, or were collected for information and trending purposes. Except for modest corrosion of the lead shield (which is typical of these packages following several years service), no evidence of a degraded condition was found in this package. The Savannah River Site (SRS) stores packages containing plutonium (Pu) materials in the KArea Complex (KAC). The Pu materials are packaged per the DOE 3013 Standard and stored within Model 9975 shipping packages in KAC. The KAC facility DSA (Document Safety Analysis) credits the Model 9975 package to perform several safety functions, including criticality prevention, impact resistance, containment, and fire resistance to ensure the plutonium materials remain in a safe configuration during normal and accident conditions. The Model 9975 package is expected to perform its safety function for at least 12 years from initial packaging. The DSA recognizes the degradation potential for the materials of package construction over time in the KAC storage environment and requires an assessment of materials performance to validate the assumptions of the analysis and ultimately predict service life. As part of the comprehensive Model 9975 package surveillance program, destructive examination of package 9975-03431 was performed following field surveillance in accordance with Reference. Field surveillance of the Model 9975 package in KAC included nondestructive examination of the drum, fiberboard, lead shield and containment vessels. Results of the field surveillance are provided in Attachment 1.

Daugherty, W.

2012-05-30T23:59:59.000Z

229

Technical & Biosystems Engineering, Industrial Technology, and Packaging Services Organizations Hiring Students in Technical & Biosystems Engineering, Industrial Technology, and Packaging  

E-Print Network [OSTI]

Faurecia FCA Packaging Fischer Controls Fusion PKG Gavilon, LLC General Motors George W. Auch Geotex,000 57,000 12 Engineer, General 56,513 33,000 80,000 34 Equipment Test Technician 46,000 32,000 60,000 510 Technical & Biosystems Engineering, Industrial Technology, and Packaging Services Organizations

230

Waste disposal package  

DOE Patents [OSTI]

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.

Smith, M.J.

1985-06-19T23:59:59.000Z

231

Radioactive waste disposal package  

DOE Patents [OSTI]

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.

Lampe, Robert F. (Bethel Park, PA)

1986-01-01T23:59:59.000Z

232

Web Technology (elective package)  

E-Print Network [OSTI]

Web Technology (elective package) Offered by: Department of Mathematics and Computer Science? Computer Science-based approaches and enabling technologies for the web. Course descriptions Human and efficient. Web Technology The web has become the major source of information retrieval and is playing

Franssen, Michael

233

A Discussion of Conductivity Testing in High Temperature Membranes (lessons learned in assessing transport)  

Broader source: Energy.gov [DOE]

Presentation on conductivity testing in high temperature membranes given by Jim Boncella of Los Alamos National Laboratory at the High Temperature Membrane Working Group meeting in October 2005.

234

Testing and Evaluation Protocol for Mobile and Transportable Radiation Monitors Used for Homeland  

E-Print Network [OSTI]

................................................................................................................2 7. Guidance for testing ANSI N42.42 data format requirements ................................2 8. Test modifications from ANSI/IEEE N42.43-2006 requirements...........................3 9 on the performance requirements established in ANSI N42.43, "American National Standard for Evaluation

235

Modeling Groundwater Flow and Transport of Radionuclides at Amchitka Island's Underground Nuclear Tests: Milrow, Long Shot, and Cannikin  

SciTech Connect (OSTI)

Since 1963, all United States nuclear tests have been conducted underground. A consequence of this testing has been the deposition of large amounts of radioactive material in the subsurface, sometimes in direct contact with groundwater. The majority of this testing occurred on the Nevada Test Site (NTS), but a limited number of experiments were conducted in other locations. One of these locations, Amchitka Island, Alaska is the subject of this report. Three underground nuclear tests were conducted on Amchitka Island. Long Shot was an 80-kiloton-yield test conducted at a depth of 700 meters (m) on October 29, 1965 (DOE, 2000). Milrow had an announced yield of about 1,000 kilotons, and was detonated at a depth of 1,220 m on October 2, 1969. Cannikin had an announced yield less than 5,000 kilotons, and was conducted at a depth of 1,790 m on November 6, 1971. The purpose of this work is to provide a portion of the information needed to conduct a human-health risk assessment of the potential hazard posed by the three underground nuclear tests on Amchitka Island. Specifically, the focus of this work is the subsurface transport portion, including the release of radionuclides from the underground cavities and their movement through the groundwater system to the point where they seep out of the ocean floor and into the marine environment. This requires a conceptual model of groundwater flow on the island using geologic, hydrologic, and chemical information, a numerical model for groundwater flow, a conceptual model of contaminant release and transport properties from the nuclear test cavities, and a numerical model for contaminant transport. Needed for the risk assessment are estimates of the quantity of radionuclides (in terms of mass flux) from the underground tests on Amchitka that could discharge to the ocean, the time of possible discharge, and the location in terms of distance from shoreline. The radionuclide data presented here are all reported in terms of normalized masses to avoid presenting classified information. As only linear processes are modeled, the results can be readily scaled by the true classified masses for use in the risk assessment. The modeling timeframe for the risk assessment was set at 1,000 years, though some calculations are extended to 2,000 years. This first section of the report endeavors to orient the reader with the environment of Amchitka and the specifics of the underground nuclear tests. Of prime importance are the geologic and hydrologic conditions of the subsurface. A conceptual model for groundwater flow beneath the island is then developed and paired with an appropriate numerical modeling approach in section 2. The parameters needed for the model, supporting data for them, and data uncertainties are discussed at length. The calibration of the three flow models (one for each test) is then presented. At this point the conceptual radionuclide transport model is introduced and its numerical approach described in section 3. Again, the transport parameters and their supporting data and uncertainties are the focus. With all of the processes and parameters in place, the first major modeling phase can be discussed in section 4. In this phase, a parametric uncertainty analysis is performed to determine the sensitivity of the transport modeling results to the uncertainties present in the parameters. This analysis is motivated by the recognition of substantial uncertainty in the subsurface conditions on the island and the need to incorporate that uncertainty into the modeling. The conclusion of the first phase determines the parameters to hold as uncertain through the main flow and transport modeling. This second, main phase of modeling is presented in section 5, with the contaminant breakthrough behavior of each test site addressed. This is followed by a sensitivity analysis in section 6, regarding the importance of additional processes that could not be supported in the main modeling effort due to lack of data. Finally, the results for the individual sites are compared, the sensitivities discussed,

Ahmed Hassan; Karl Pohlmann; Jenny Chapman

2002-11-19T23:59:59.000Z

236

Reactor Pressure Vessel Head Packaging & Disposal  

SciTech Connect (OSTI)

Reactor Pressure Vessel (RPV) Head replacements have come to the forefront due to erosion/corrosion and wastage problems resulting from the susceptibility of the RPV Head alloy steel material to water/boric acid corrosion from reactor coolant leakage through the various RPV Head penetrations. A case in point is the recent Davis-Besse RPV Head project, where detailed inspections in early 2002 revealed significant wastage of head material adjacent to one of the Control Rod Drive Mechanism (CRDM) nozzles. In lieu of making ASME weld repairs to the damaged head, Davis-Besse made the decision to replace the RPV Head. The decision was made on the basis that the required weld repair would be too extensive and almost impractical. This paper presents the packaging, transport, and disposal considerations for the damaged Davis-Besse RPV Head. It addresses the requirements necessary to meet Davis Besse needs, as well as the regulatory criteria, for shipping and burial of the head. It focuses on the radiological characterization, shipping/disposal package design, site preparation and packaging, and the transportation and emergency response plans that were developed for the Davis-Besse RPV Head project.

Wheeler, D. M.; Posivak, E.; Freitag, A.; Geddes, B.

2003-02-26T23:59:59.000Z

237

PAT-2 (Plutonium Air-Transportable Model 2) safety analysis report  

SciTech Connect (OSTI)

The PAT-2 package is designed for the safe transport of plutonium and/or uranium in small quantities. The PAT-2 package is resistant to severe accidents, including that of a high-speed jet aircraft crash, and is designed to withstand such environments as extreme impact, crushing, puncturing and slashing loads, severe hydrocarbon-fueled fires, and deep underwater immersion, with no escape of contents. The package meets the requirements of 10 CFR 71 for Fissile Class I packages with a cargo of 15 grams of Pu-239, or other isotopic forms described herein, not to exceed 2 watts of thermal activity. This SAR presents design and oprational information including evaluations and analyses, test results, operating procedures, maintenance, and quality assurance information.

Andersen, J.A.; Davis, E.J.; Duffey, T.A.; Dupree, S.A.; George, O.L. Jr.; Ortiz, Z.

1981-07-01T23:59:59.000Z

238

Evaluation of pre-packaged agricultural drip irrigation kits  

E-Print Network [OSTI]

The purpose of this thesis is to conduct user testing and performance evaluation of two different agricultural pre-packaged drip irrigation kit (PDIK) systems: Chapin Bucket Kit and International Design Enterprises (IDE) ...

Huang, Shen, S.B. Massachusetts Institute of Technology

2012-01-01T23:59:59.000Z

239

Annual Transportation Report for Radioactive Waste Shipments to and from the Nevada Test Site, Fiscal Year 2009  

SciTech Connect (OSTI)

In February 1997, the U.S. Department of Energy (DOE), Nevada Operations Office (now known as the Nevada Site Office) issued the Mitigation Action Plan which addressed potential impacts described in the “Final Environmental Impact Statement for the Nevada Test Site and Off-Site Locations in the State of Nevada” (DOE/EIS 0243). The DOE, Nevada Operations Office committed to several actions, including the preparation of an annual report, which summarizes waste shipments to and from the Nevada Test Site (NTS) Radioactive Waste Management Site (RWMS) at Area 5 and Area 3. Since 2006, the Area 3 RWMS has been in cold stand-by. This document satisfies requirements regarding low-level radioactive waste (LLW) and mixed low-level radioactive waste (MLLW) transported to and from the NTS during FY 2009. In addition, this document provides shipment, volume, and route information on transuranic (TRU) waste shipped from the NTS to the Idaho National Laboratory, near Idaho Falls, Idaho.

U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office

2010-02-01T23:59:59.000Z

240

Technical Review Report for the Model 9975-96 Package Safety Analysis Report for Packaging (S-SARP-G-00003, Revision 0, January 2008)  

SciTech Connect (OSTI)

This Technical Review Report (TRR) documents the review, performed by the Lawrence Livermore National Laboratory (LLNL) Staff, at the request of the U.S. Department of Energy (DOE), on the Safety Analysis Report for Packaging, Model 9975, Revision 0, dated January 2008 (S-SARP-G-00003, the SARP). The review includes an evaluation of the SARP, with respect to the requirements specified in 10 CFR 71, and in International Atomic Energy Agency (IAEA) Safety Standards Series No. TS-R-1. The Model 9975-96 Package is a 35-gallon drum package design that has evolved from a family of packages designed by DOE contractors at the Savannah River Site. Earlier package designs, i.e., the Model 9965, the Model 9966, the Model 9967, and the Model 9968 Packagings, were originally designed and certified in the early 1980s. In the 1990s, updated package designs that incorporated design features consistent with the then newer safety requirements were proposed. The updated package designs at the time were the Model 9972, the Model 9973, the Model 9974, and the Model 9975 Packagings, respectively. The Model 9975 Package was certified by the Packaging Certification Program, under the Office of Safety Management and Operations. The safety analysis of the Model 9975-85 Packaging is documented in the Safety Analysis Report for Packaging, Model 9975, B(M)F-85, Revision 0, dated December 2003. The Model 9975-85 Package is certified by DOE Certificate of Compliance (CoC) package identification number, USA/9975/B(M)F-85, for the transportation of Type B quantities of uranium metal/oxide, {sup 238}Pu heat sources, plutonium/uranium metals, plutonium/uranium oxides, plutonium composites, plutonium/tantalum composites, {sup 238}Pu oxide/beryllium metal.

West, M

2009-05-22T23:59:59.000Z

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

Plutonium stabilization and packaging system  

SciTech Connect (OSTI)

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

NONE

1996-05-01T23:59:59.000Z

242

Optimal segmentation and packaging process  

DOE Patents [OSTI]

A process for improving packaging efficiency uses three dimensional, computer simulated models with various optimization algorithms to determine the optimal segmentation process and packaging configurations based on constraints including container limitations. The present invention is applied to a process for decontaminating, decommissioning (D&D), and remediating a nuclear facility involving the segmentation and packaging of contaminated items in waste containers in order to minimize the number of cuts, maximize packaging density, and reduce worker radiation exposure. A three-dimensional, computer simulated, facility model of the contaminated items are created. The contaminated items are differentiated. The optimal location, orientation and sequence of the segmentation and packaging of the contaminated items is determined using the simulated model, the algorithms, and various constraints including container limitations. The cut locations and orientations are transposed to the simulated model. The contaminated items are actually segmented and packaged. The segmentation and packaging may be simulated beforehand. In addition, the contaminated items may be cataloged and recorded.

Kostelnik, Kevin M. (Idaho Falls, ID); Meservey, Richard H. (Idaho Falls, ID); Landon, Mark D. (Idaho Falls, ID)

1999-01-01T23:59:59.000Z

243

Modeling of Groundwater Flow and Radionuclide Transport at the Climax Mine sub-CAU, Nevada Test Site  

SciTech Connect (OSTI)

The Yucca Flat-Climax Mine Corrective Action Unit (CAU) on the Nevada Test Site comprises 747 underground nuclear detonations, all but three of which were conducted in alluvial, volcanic, and carbonate rocks in Yucca Flat. The remaining three tests were conducted in the very different hydrogeologic setting of the Climax Mine granite stock located in Area 15 at the northern end of Yucca Flat. As part of the Corrective Action Investigation (CAI) for the Yucca Flat-Climax Mine CAU, models of groundwater flow and radionuclide transport will be developed for Yucca Flat. However, two aspects of these CAU-scale models require focused modeling at the northern end of Yucca Flat beyond the capability of these large models. First, boundary conditions and boundary flows along the northern reaches of the Yucca Flat-Climax Mine CAU require evaluation to a higher level of detail than the CAU-scale Yucca Flat model can efficiently provide. Second, radionuclide fluxes from the Climax tests require analysis of flow and transport in fractured granite, a unique hydrologic environment as compared to Yucca Flat proper. This report describes the Climax Mine sub-CAU modeling studies conducted to address these issues, with the results providing a direct feed into the CAI for the Yucca Flat-Climax Mine CAU. Three underground nuclear detonations were conducted for weapons effects testing in the Climax stock between 1962 and 1966: Hard Hat, Pile Driver, and Tiny Tot. Though there is uncertainty regarding the position of the water table in the stock, it is likely that all three tests were conducted in the unsaturated zone. In the early 1980s, the Spent Fuel Test-Climax (SFT-C) was constructed to evaluate the feasibility of retrievable, deep geologic storage of commercial nuclear reactor wastes. Detailed mapping of fractures and faults carried out for the SFT-C studies greatly expanded earlier data sets collected in association with the nuclear tests and provided invaluable information for subsequent modeling studies at Climax. The objectives of the Climax Mine sub-CAU work are to (1) provide simulated heads and groundwater flows for the northern boundaries of the Yucca Flat-Climax Mine CAU model, while incorporating alternative conceptualizations of the hydrogeologic system with their associated uncertainty, and (2) provide radionuclide fluxes from the three tests in the Climax stock using modeling techniques that account for groundwater flow in fractured granite. Meeting these two objectives required two different model scales. The northern boundary groundwater fluxes were addressed using the Death Valley Regional Flow System (DVRFS) model (Belcher, 2004) developed by the U.S. Geological Survey as a modeling framework, with refined hydrostratigraphy in a zone north of Yucca Flat and including Climax stock. Radionuclide transport was simulated using a separate model confined to the granite stock itself, but linked to regional groundwater flow through boundary conditions and calibration targets.

K. Pohlmann; M. Ye; D. Reeves; M. Zavarin; D. Decker; J. Chapman

2007-09-28T23:59:59.000Z

244

Reaction mechanisms in transport theories: a test of the nuclear effective interaction  

E-Print Network [OSTI]

We review recent results concerning collective excitations in neutron-rich systems and reactions between charge asymmetric systems at Fermi energies. Solving numerically self-consistent transport equations for neutrons and protons with specific initial conditions, we explore the structure of the different dipole vibrations in the $^{132}Sn$ system and investigate their dependence on the symmetry energy. We evidence the existence of a distinctive collective mode, that can be associated with the Pygmy Dipole Resonance, with an energy well below the standard Giant Dipole Resonance and isoscalar-like character, i.e. very weakly dependent on the isovector part of the nuclear effective interaction. At variance, the corresponding strength is rather sensitive to the behavior of the symmetry energy below saturation, which rules the number of excess neutrons in the nuclear surface. In reactions between charge asymmetric systems at Fermi energies, we investigate the interplay between dissipation mechanisms and isospin effects. Observables sensitive to the isospin dependent part of nuclear interaction are discussed, providing information on the symmetry energy density dependence below saturation.

M. Colonna; V. Baran; M. Di Toro; B. Frecus; Y. X. Zhang

2012-09-07T23:59:59.000Z

245

Safety Analysis Report for packaging (onsite) steel waste package  

SciTech Connect (OSTI)

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.

BOEHNKE, W.M.

2000-07-13T23:59:59.000Z

246

DESTRUCTIVE EXAMINATION OF SHIPPING PACKAGE 9975-06100  

SciTech Connect (OSTI)

Destructive and non-destructive examinations have been performed on specified components of shipping package 9975-06100. This package was selected for examination based on several characteristics: - This was the first destructively examined package in which the fiberboard assembly was fabricated from softwood fiberboard. - The package contained a relatively high heat load to contribute to internal temperature, which is a key environmental factor for fiberboard degradation. - The package has been stored in the middle or top of a storage array since its receipt in K- Area, positions that would contribute to increased service temperatures. No significant changes were observed for attributes that were measured during both field surveillance and destructive examination. Except for the axial gap, all observations and test results met identified criteria, or were collected for information and trending purposes. The axial gap met the 1 inch maximum criterion during field surveillance, but was just over the criterion during SRNL measurements. When re-measured at a later date, it again met the criterion. The bottom of the lower fiberboard assembly and the drum interior had two small stains at matching locations, suggestive of water intrusion. However, the fiberboard assembly did not contain any current evidence of excess moisture. No evidence of a degraded condition was found in this package. Despite exposure to the elevated temperatures of this higher-then-average wattage package, properties of the fiberboard and O-rings are consistent with those of new packages.

Daugherty, W.

2014-11-07T23:59:59.000Z

247

A groundwater flow and transport model of long-term radionuclide migration in central Frenchman flat, Nevada test site  

SciTech Connect (OSTI)

A set of groundwater flow and transport models were created for the Central Testing Area of Frenchman Flat at the former Nevada Test Site to investigate the long-term consequences of a radionuclide migration experiment that was done between 1975 and 1990. In this experiment, radionuclide migration was induced from a small nuclear test conducted below the water table by pumping a well 91 m away. After radionuclides arrived at the pumping well, the contaminated effluent was discharged to an unlined ditch leading to a playa where it was expected to evaporate. However, recent data from a well near the ditch and results from detailed models of the experiment by LLNL personnel have convincingly demonstrated that radionuclides from the ditch eventually reached the water table some 220 m below land surface. The models presented in this paper combine aspects of these detailed models with concepts of basin-scale flow to estimate the likely extent of contamination resulting from this experiment over the next 1,000 years. The models demonstrate that because regulatory limits for radionuclide concentrations are exceeded only by tritium and the half-life of tritium is relatively short (12.3 years), the maximum extent of contaminated groundwater has or will soon be reached, after which time the contaminated plume will begin to shrink because of radioactive decay. The models also show that past and future groundwater pumping from water supply wells within Frenchman Flat basin will have negligible effects on the extent of the plume.

Kwicklis, Edward Michael [Los Alamos National Laboratory; Becker, Naomi M [Los Alamos National Laboratory; Ruskauff, Gregory [NAVARRO-INTERA, LLC.; De Novio, Nicole [GOLDER AND ASSOC.; Wilborn, Bill [US DOE NNSA NSO

2010-11-10T23:59:59.000Z

248

Transportation Baseline Schedule  

SciTech Connect (OSTI)

The “1999 National Transportation Program - Transportation Baseline Report” presents data that form a baseline to enable analysis and planning for future Department of Energy (DOE) Environmental Management (EM) waste/material transportation. The companion “1999 Transportation ‘Barriers’ Analysis” analyzes the data and identifies existing and potential problems that may prevent or delay transportation activities based on the data presented. The “1999 Transportation Baseline Schedule” (this report) uses the same data to provide an overview of the transportation activities of DOE EM waste/materials. This report can be used to identify areas where stakeholder interface is needed, and to communicate to stakeholders the quantity/schedule of shipments going through their area. Potential bottlenecks in the transportation system can be identified; the number of packages needed, and the capacity needed at receiving facilities can be planned. This report offers a visualization of baseline DOE EM transportation activities for the 11 major sites and the “Geologic Repository Disposal” site (GRD).

Fawcett, Ricky Lee; John, Mark Earl

2000-01-01T23:59:59.000Z

249

A COMPARISON OF TWO THERMAL INSULATION AND STRUCTURAL MATERIALS FOR USE IN TYPE B PACKAGINGS  

SciTech Connect (OSTI)

This paper presents the summary of design features and test results of two Type B Shipping Package prototype configurations comprising different insulating materials developed by the Savannah River National Laboratory (SRNL) for the Department of Energy. The materials evaluated, a closed-cell polyurethane foam and a vacuformed ceramic fiber material, were selected to provide adequate structural protection to the package containment vessel during Normal Conditions of Transport (NCT) and Hypothetical Accident Condition (HAC) events and to provide thermal protection during the HAC fire. Polyurethane foam has been used in shipping package designs for many years because of the stiffness it provides to the structure and because of the thermal protection it provides during fire scenarios. This comparison describes how ceramic fiber material offers an alternative to the polyurethane foam in a specific overpack design. Because of the high operating temperature ({approx}2,300 F) of the ceramic material, it allows for contents with higher heat loads to be shipped than is possible with polyurethane foam. Methods of manufacturing and design considerations using the two materials will be addressed.

Blanton, P.; Eberl, K.

2010-07-16T23:59:59.000Z

250

Design package for vacuum wand for fuel retrieval system  

SciTech Connect (OSTI)

This is a design package that contains the details for the design, fabrication, and testing of a vacuum wand that will pick up sludge and corrosion products generated during fuel assembly handling operations at K-Basin. This document contains requirements, development design information, design calculations, tests, and test reports.

ROACH, H.L.

1999-07-28T23:59:59.000Z

251

General Corrosion and Localized Corrosion of Waste Package Outer Barrier  

SciTech Connect (OSTI)

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]).

K.G. Mon

2004-10-01T23:59:59.000Z

252

Environmental, operational and financial sustainability of packaging methods in delivery businesses  

E-Print Network [OSTI]

In retail delivery companies, packaging is used to transport goods to customers while preventing damage, shrinkage and loss of the contents. With consumer preferences reflecting the growing concern for the environment, ...

Ng, Joshua (Zi Jie Joshua)

2010-01-01T23:59:59.000Z

253

IN-PACKAGE CHEMISTRY ABSTRACTION  

SciTech Connect (OSTI)

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

E. Thomas

2005-07-14T23:59:59.000Z

254

EBS Radionuclide Transport Abstraction  

SciTech Connect (OSTI)

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 advective transport and diffusive transport from a breached waste package. Advective transport occurs when radionuclides that are dissolved or sorbed onto colloids (or both) are carried from the waste package by the portion of the seepage flux that passes through waste package breaches. Diffusive transport occurs as a result of a gradient in radionuclide concentration and may take place while advective transport is also occurring, as well as when no advective transport is occurring. Diffusive transport is addressed in detail because it is the sole means of transport when there is no flow through a waste package, which may dominate during the regulatory compliance period in the nominal and seismic scenarios. The advective transport rate, when it occurs, is generally greater than the diffusive transport rate. Colloid-facilitated advective and diffusive transport is also modeled and is presented in detail in Appendix B of this report.

J.D. Schreiber

2005-08-25T23:59:59.000Z

255

Grant Application Package CFDA Number  

E-Print Network [OSTI]

Grant Application Package CFDA Number: Opportunity Title: Offering Agency: Agency Contact: Opportunity Open Date: Opportunity Close Date: CFDA Description: Opportunity Number: Competition ID

Talley, Lynne D.

256

Naval Waste Package Design Report  

SciTech Connect (OSTI)

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.

M.M. Lewis

2004-03-15T23:59:59.000Z

257

Modeling Gas Transport in the Shallow Subsurface During the ZERT CO2 Release Test  

SciTech Connect (OSTI)

We used the multiphase and multicomponent TOUGH2/EOS7CA model to carry out predictive simulations of CO{sub 2} injection into the shallow subsurface of an agricultural field in Bozeman, Montana. The purpose of the simulations was to inform the choice of CO{sub 2} injection rate and design of monitoring and detection activities for a CO{sub 2} release experiment. The release experiment configuration consists of a long horizontal well (70 m) installed at a depth of approximately 2.5 m into which CO{sub 2} is injected to mimic leakage from a geologic carbon sequestration site through a linear feature such as a fault. We estimated the permeability of the soil and cobble layers present at the site by manual inversion of measurements of soil CO{sub 2} flux from a vertical-well CO{sub 2} release. Based on these estimated permeability values, predictive simulations for the horizontal well showed that CO{sub 2} injection just below the water table creates an effective gas-flow pathway through the saturated zone up to the unsaturated zone. Once in the unsaturated zone, CO{sub 2} spreads out laterally within the cobble layer, where liquid saturation is relatively low. CO{sub 2} also migrates upward into the soil layer through the capillary barrier and seeps out at the ground surface. The simulations predicted a breakthrough time of approximately two days for the 100kg d{sup -1} injection rate, which also produced a flux within the range desired for testing detection and monitoring approaches. The seepage area produced by the model was approximately five meters wide above the horizontal well, compatible with the detection and monitoring methods tested. For a given flow rate, gas-phase diffusion of CO{sub 2} tends to dominate over advection near the ground surface, where the CO{sub 2} concentration gradient is large, while advection dominates deeper in the system.

Oldenburg, Curtis M.; Lewicki, Jennifer L.; Dobeck, Laura; Spangler, Lee

2009-01-15T23:59:59.000Z

258

PACKAGING CERTIFICATION PROGRAM METHODOLOGY FOR DETERMINING DOSE RATES FOR SMALL GRAM QUANTITIES IN SHIPPING PACKAGINGS  

SciTech Connect (OSTI)

The Small Gram Quantity (SGQ) concept is based on the understanding that small amounts of hazardous materials, in this case radioactive materials (RAM), are significantly less hazardous than large amounts of the same materials. This paper describes a methodology designed to estimate an SGQ for several neutron and gamma emitting isotopes that can be shipped in a package compliant with 10 CFR Part 71 external radiation level limits regulations. These regulations require packaging for the shipment of radioactive materials, under both normal and accident conditions, to perform the essential functions of material containment, subcriticality, and maintain external radiation levels within the specified limits. By placing the contents in a helium leak-tight containment vessel, and limiting the mass to ensure subcriticality, the first two essential functions are readily met. Some isotopes emit sufficiently strong photon radiation that small amounts of material can yield a large dose rate outside the package. Quantifying the dose rate for a proposed content is a challenging issue for the SGQ approach. It is essential to quantify external radiation levels from several common gamma and neutron sources that can be safely placed in a specific packaging, to ensure compliance with federal regulations. The Packaging Certification Program (PCP) Methodology for Determining Dose Rate for Small Gram Quantities in Shipping Packagings provides bounding shielding calculations that define mass limits compliant with 10 CFR 71.47 for a set of proposed SGQ isotopes. The approach is based on energy superposition with dose response calculated for a set of spectral groups for a baseline physical packaging configuration. The methodology includes using the MCNP radiation transport code to evaluate a family of neutron and photon spectral groups using the 9977 shipping package and its associated shielded containers as the base case. This results in a set of multipliers for 'dose per particle' for each spectral group. For a given isotope, the source spectrum is folded with the response for each group. The summed contribution from all isotopes determines the total dose from the RAM in the container.

Nathan, S.; Loftin, B.; Abramczyk, G.; Bellamy, S.

2012-05-09T23:59:59.000Z

259

Building an R package Division of Biostatistics  

E-Print Network [OSTI]

Building an R package Yen-Yi Ho Division of Biostatistics School of Public Health, University of Minnesota Yen-Yi Ho Building an R package #12;Steps Prepare your functions, example data sets Build package in man subdirectory) Write a package vignette Build and install the R package (R CMD build) Check the R

Carlin, Bradley P.

260

Approaches to Quantify Potential Contaminant Transport in the Lower Carbonate Aquifer from Underground Nuclear Testing at Yucca Flat, Nevada National Security Site, Nye County, Nevada - 12434  

SciTech Connect (OSTI)

Quantitative modeling of the potential for contaminant transport from sources associated with underground nuclear testing at Yucca Flat is an important part of the strategy to develop closure plans for the residual contamination. At Yucca Flat, the most significant groundwater resource that could potentially be impacted is the Lower Carbonate Aquifer (LCA), a regionally extensive aquifer that supplies a significant portion of the water demand at the Nevada National Security Site, formerly the Nevada Test Site. Developing and testing reasonable models of groundwater flow in this aquifer is an important precursor to performing subsequent contaminant transport modeling used to forecast contaminant boundaries at Yucca Flat that are used to identify potential use restriction and regulatory boundaries. A model of groundwater flow in the LCA at Yucca Flat has been developed. Uncertainty in this model, as well as other transport and source uncertainties, is being evaluated as part of the Underground Testing Area closure process. Several alternative flow models of the LCA in the Yucca Flat/Climax Mine CAU have been developed. These flow models are used in conjunction with contaminant transport models and source term models and models of contaminant transport from underground nuclear tests conducted in the overlying unsaturated and saturated alluvial and volcanic tuff rocks to evaluate possible contaminant migration in the LCA for the next 1,000 years. Assuming the flow and transport models are found adequate by NNSA/NSO and NDEP, the models will undergo a peer review. If the model is approved by NNSA/NSO and NDEP, it will be used to identify use restriction and regulatory boundaries at the start of the Corrective Action Decision Document Corrective Action Plan (CADD/CAP) phase of the Corrective Action Strategy. These initial boundaries may be revised at the time of the Closure Report phase of the Corrective Action Strategy. (authors)

Andrews, Robert W.; Birdie, Tiraz [Navarro-INTERA LLC, Las Vegas, Nevada 89030 (United States); Wilborn, Bill; Mukhopadhyay, Bimal [National Nuclear Security Administration/Nevada Site Office, Las Vegas, Nevada 89030 (United States)

2012-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

Tritium waste package  

DOE Patents [OSTI]

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.

Rossmassler, Rich (Cranbury, NJ); Ciebiera, Lloyd (Titusville, NJ); Tulipano, Francis J. (Teaneck, NJ); Vinson, Sylvester (Ewing, NJ); Walters, R. Thomas (Lawrenceville, NJ)

1995-01-01T23:59:59.000Z

262

Tritium waste package  

DOE Patents [OSTI]

A containment and waste package system for processing and shipping tritium oxide waste received from a process gas includes an outer drum and an inner drum containing a disposable molecular sieve bed (DMSB) seated within the 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 and oxygen present in the process gas to recombine to form water vapor, which is then adsorbed onto the DMSB. 1 fig.

Rossmassler, R.; Ciebiera, L.; Tulipano, F.J.; Vinson, S.; Walters, R.T.

1995-11-07T23:59:59.000Z

263

Electro-Microfluidic Packaging  

SciTech Connect (OSTI)

Electro-microfluidics is experiencing explosive growth in new product developments. There are many commercial applications for electro-microfluidic devices such as chemical sensors, biological sensors, and drop ejectors for both printing and chemical analysis. The number of silicon surface micromachined electro-microfluidic products is likely to increase. Manufacturing efficiency and integration of microfluidics with electronics will become important. Surface micromachined microfluidic devices are manufactured with the same tools as IC's (integrated circuits) and their fabrication can be incorporated into the IC fabrication process. In order to realize applications for devices must be developed. An Electro-Microfluidic Dual In-line Package (EMDIP{trademark}) was developed surface micromachined electro-microfluidic devices, a practical method for getting fluid into these to be a standard solution that allows for both the electrical and the fluidic connections needed to operate a great variety of electro-microfluidic devices. The EMDIP{trademark} includes a fan-out manifold that, on one side, mates directly with the 200 micron diameter Bosch etched holes found on the device, and, on the other side, mates to lager 1 mm diameter holes. To minimize cost the EMDIP{trademark} can be injection molded in a great variety of thermoplastics which also serve to optimize fluid compatibility. The EMDIP{trademark} plugs directly into a fluidic printed wiring board using a standard dual in-line package pattern for the electrical connections and having a grid of multiple 1 mm diameter fluidic connections to mate to the underside of the EMDIP{trademark}.

BENAVIDES, GILBERT L.; GALAMBOS, PAUL C.

2002-06-01T23:59:59.000Z

264

MODEL 9975 SHIPPING PACKAGE: IMPACT OF CAPLUG REMOVAL ON FIBERBOARD MOISTURE LEVEL  

SciTech Connect (OSTI)

Two 9975 shipping packages were removed from KAC and provided to SRNL for test purposes, after both packages were found to exceed the 1 inch maximum criterion for the axial gap at the top of the package. Package 9975-01818 was found with an axial gap of 1.437 inch, and an estimated 2.5 liters of excess moisture in the lower fiberboard layers. Package 9975-02287 was found with an axial gap of 1.008 inch, and only slightly elevated moisture levels relative to typical packages. Prior data from the 9975 Surveillance Program has shown that the 9975 drum provides a degree of isolation, and will tend to preserve fiberboard moisture levels for an extended period of time. Both packages were provided to SRNL to identify whether removal of the 4 caplugs in each package would allow moisture to escape the package. Following testing with the caplugs removed for approximately 1 year, this report documents the findings from this effort. Two 9975 shipping packages removed from service in K-Area Complex (KAC) due to an excessive axial gap have been tested in SRNL to determine if caplug removal would facilitate the reduction of excess fiberboard moisture. An additional question to be answered through this testing was whether the resulting moisture loss would reduce the axial gap, reversing the effect seen during storage with excess moisture present. These packages have completed approximately 1 year in test, during which time the weight of each package has steadily decreased as a result of moisture migration out of the package. However, elevated moisture levels still remain in the packages. During this test period, the bottom fiberboard layers of package 9975-01818 (which contained the greater amount of excess moisture) experienced further compaction, and the axial gap of both packages has increased. This effort has shown that removal of the caplugs may not be a sufficient measure to rehabilitate packages with excess moisture or excess axial gaps in a timely manner. However, this measure might make a meaningful contribution in combination with other actions (to be determined). It is recommended that the caplug removal tests in SRNL be discontinued at this time.

Daugherty, W.

2011-06-23T23:59:59.000Z

265

Waste Package Materials Performance Peer Review | Department...  

Broader source: Energy.gov (indexed) [DOE]

Waste Package Materials Performance Peer Review Waste Package Materials Performance Peer Review A consensus peer review of the current technical basis and the planned experimental...

266

Optimal segmentation and packaging process  

DOE Patents [OSTI]

A process for improving packaging efficiency uses three dimensional, computer simulated models with various optimization algorithms to determine the optimal segmentation process and packaging configurations based on constraints including container limitations. The present invention is applied to a process for decontaminating, decommissioning (D and D), and remediating a nuclear facility involving the segmentation and packaging of contaminated items in waste containers in order to minimize the number of cuts, maximize packaging density, and reduce worker radiation exposure. A three-dimensional, computer simulated, facility model of the contaminated items are created. The contaminated items are differentiated. The optimal location, orientation and sequence of the segmentation and packaging of the contaminated items is determined using the simulated model, the algorithms, and various constraints including container limitations. The cut locations and orientations are transposed to the simulated model. The contaminated items are actually segmented and packaged. The segmentation and packaging may be simulated beforehand. In addition, the contaminated items may be cataloged and recorded. 3 figs.

Kostelnik, K.M.; Meservey, R.H.; Landon, M.D.

1999-08-10T23:59:59.000Z

267

K-Basins Sludge Treatment and Packaging at the Hanford Site - 13585  

SciTech Connect (OSTI)

Highly radioactive sludge resulting from the storage of degraded spent nuclear fuel has been consolidated in Engineered Containers (ECs) in the 105-K West Storage Basin located on the Hanford site near the Columbia River in Washington State. CH2M Hill Plateau Remediation Company (CHPRC) is proceeding with a project to retrieve the sludge, place it in Sludge Transport and Storage Containers (STSCs) and store those filled containers within the T Plant Canyon facility on the Hanford Site Central Plateau (Phase 1). Retrieval and transfer of the sludge material will enable removal of the 105-K West Basin and allow remediation of the subsurface contamination plumes under the basin. The U.S. Department of Energy (DOE) plans to treat and dispose of this K Basins sludge (Phase 2) as Remote Handled Transuranic Waste (RH TRU) at the Waste Isolation Pilot Plant (WIPP) located in New Mexico. The K Basin sludge currently contains uranium metal which reacts with water present in the stored slurry, generating hydrogen and other byproducts. The established transportation and disposal requirements require the transformation of the K Basins sludge to a chemically stable, liquid-free, packaged waste form. The Treatment and Packaging Project includes removal of the containerised sludge from T Plant, the treatment of the sludge as required, and packaging of all the sludge into a form that is certifiable for transportation to and disposal at WIPP. Completion of this scope will require construction and operation of a Sludge Treatment and Packaging Facility (STPF), which could be either a completely new facility or a modification of an existing Hanford Site facility. A Technology Evaluation and Alternatives Analysis (TEAA) for the STP Phase 2 was completed in 2011. A Request for Technology Information (RFI) had been issued in October 2009 to solicit candidate technologies for use in Phase 2. The RFI also included a preliminary definition of Phase 2 functions and requirements. Potentially applicable technologies were identified through a commercial procurement process, technical workshops, and review of the numerous previous sludge treatment technology studies. The identified technology approaches were screened using the criteria established in the Decision Plan, and focused bench top feasibility testing was conducted. Engineering evaluations of the costs, schedules, and technical maturity were developed and evaluated. Recommendations were developed based on technical evaluations. The criteria used in the evaluation process were as follows: (1) Safety, (2) Regulatory/stakeholder acceptance, (3) Technical maturity, (4) Operability and maintainability, (5) Life cycle cost and schedule, (6) Potential for beneficial integration with ongoing STP-Phase 1 activities, and (7) Integration with Site-wide RH-TRU processing/packaging, planning, schedule, and approach. The TEAA recommended Warm Water Oxidation (WWO) as the baseline treatment technology and two risk reduction enhancement options for further consideration during development of the process - size reduction and chemical oxidation (Fenton's reagent). The enhancement options would potentially allow a useful reduction in the total operating time required to process the K Basins sludge. The U.S. Department of Energy's Richland Field Office (DOE-RL) has approved this recommended technical approach. The baseline process can be broken down into the following main process steps: (1) STSC transfer from T Plant to the Sludge Treatment and Packaging Facility (STPF). (2) Retrieval of sludge from the STSCs and transfer to the Receipt and Reaction Tank (RRT). (3) Preparation for immobilization by oxidation using heated water (i.e., WWO) for those batches that require it and concentration by evaporating water at about atmospheric pressure in the RRT. (4) Immobilization by using additives to eliminate free liquids and packaging of the treated sludge into drums. (5) Inspection and handling of the filled drums prior to transfer to a separate storage and shipping facility. (6) Handling of vapor, condensate, and oth

Fogwell, Thomas W. [Fogwell Consulting, P.O. Box 20211, Piedmont, CA 94620 (United States)] [Fogwell Consulting, P.O. Box 20211, Piedmont, CA 94620 (United States); Honeyman, James O. [CH2M HILL Plateau Remediation Company, P.O. Box 1600 H7-30, Richland, WA (United States)] [CH2M HILL Plateau Remediation Company, P.O. Box 1600 H7-30, Richland, WA (United States); Stegen, Gary [Lucas Engineering and Management Services, Inc., 1201 Jadwin Avenue, Suite 102, Richland, WA 99352 (United States)] [Lucas Engineering and Management Services, Inc., 1201 Jadwin Avenue, Suite 102, Richland, WA 99352 (United States)

2013-07-01T23:59:59.000Z

268

Heart Rate Variability Analysis Using Threshold of Wavelet Package Coefficients  

E-Print Network [OSTI]

In this paper, a new efficient feature extraction method based on the adaptive threshold of wavelet package coefficients is presented. This paper especially deals with the assessment of autonomic nervous system using the background variation of the signal Heart Rate Variability HRV extracted from the wavelet package coefficients. The application of a wavelet package transform allows us to obtain a time-frequency representation of the signal, which provides better insight in the frequency distribution of the signal with time. A 6 level decomposition of HRV was achieved with db4 as mother wavelet, and the above two bands LF and HF were combined in 12 specialized frequencies sub-bands obtained in wavelet package transform. Features extracted from these coefficients can efficiently represent the characteristics of the original signal. ANOVA statistical test is used for the evaluation of proposed algorithm.

Kheder, G; Massoued, M Ben; Samet, M

2009-01-01T23:59:59.000Z

269

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

SciTech Connect (OSTI)

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.

Romano, T.

1997-09-25T23:59:59.000Z

270

Packaging of solid state devices  

DOE Patents [OSTI]

A package for one or more solid state devices in a single module that allows for operation at high voltage, high current, or both high voltage and high current. Low thermal resistance between the solid state devices and an exterior of the package and matched coefficient of thermal expansion between the solid state devices and the materials used in packaging enables high power operation. The solid state devices are soldered between two layers of ceramic with metal traces that interconnect the devices and external contacts. This approach provides a simple method for assembling and encapsulating high power solid state devices.

Glidden, Steven C.; Sanders, Howard D.

2006-01-03T23:59:59.000Z

271

TORT certification package  

SciTech Connect (OSTI)

The TORT code has been certified. TORT is a three-dimensional discrete ordinates transport theory code, than can solve neutron, photon, or coupled neutron/photon problems. The code will be used primarily for shielding and radiation field calculations SRS. As defined in this work, certification dies not imply validation. The code must be validated for a particular type of calculation before it can be used for critical applications.

Frost, R.L.

1993-10-01T23:59:59.000Z

272

Expanded Content Envelope For The Model 9977 Packaging  

SciTech Connect (OSTI)

An Addendum was written to the Model 9977 Safety Analysis Report for Packaging adding a new content consisting of DOE-STD-3013 stabilized plutonium dioxide materials to the authorized Model 9977 contents. The new Plutonium Oxide Content (PuO{sub 2}) Envelope will support the Department of Energy shipment of materials between Los Alamos National Laboratory and Savannah River Site facilities. The new content extended the current content envelope boundaries for radioactive material mass and for decay heat load and required a revision to the 9977 Certificate of Compliance prior to shipment. The Addendum documented how the new contents/configurations do not compromise the safety basis presented in the 9977 SARP Revision 2. The changes from the certified package baseline and the changes to the package required to safely transport this material is discussed.

Abramczyk, G. A.; Loftin, B. M.; Nathan, S. J.; Bellamy, J. S.

2013-07-30T23:59:59.000Z

273

Hydrologic transport of depleted uranium associated with open air dynamic range testing at Los Alamos National Laboratory, New Mexico, and Eglin Air Force Base, Florida  

SciTech Connect (OSTI)

Hydrologic investigations on depleted uranium fate and transport associated with dynamic testing activities were instituted in the 1980`s at Los Alamos National Laboratory and Eglin Air Force Base. At Los Alamos, extensive field watershed investigations of soil, sediment, and especially runoff water were conducted. Eglin conducted field investigations and runoff studies similar to those at Los Alamos at former and active test ranges. Laboratory experiments complemented the field investigations at both installations. Mass balance calculations were performed to quantify the mass of expended uranium which had transported away from firing sites. At Los Alamos, it is estimated that more than 90 percent of the uranium still remains in close proximity to firing sites, which has been corroborated by independent calculations. At Eglin, we estimate that 90 to 95 percent of the uranium remains at test ranges. These data demonstrate that uranium moves slowly via surface water, in both semi-arid (Los Alamos) and humid (Eglin) environments.

Becker, N.M. [Los Alamos National Lab., NM (United States); Vanta, E.B. [Wright Laboratory Armament Directorate, Eglin Air Force Base, FL (United States)

1995-05-01T23:59:59.000Z

274

Safety analysis report for packaging, Oak Ridge Y-12 Plant, model DC-1 package with HEU oxide contents. Change pages for Rev.1  

SciTech Connect (OSTI)

This Safety Analysis Report for Packaging for the Oak Ridge Y-12 Plant for the Model DC-1 package with highly enriched uranium (HEU) oxide contents has been prepared in accordance with governing regulations form the Nuclear Regulatory Commission and the Department of Transportation and orders from the Department of energy. The fundamental safety requirements addressed by these regulations and orders pertain to the containment of radioactive material, radiation shielding, and nuclear subcriticality. This report demonstrates how these requirements are met.

NONE

1995-01-18T23:59:59.000Z

275

DORT certification package  

SciTech Connect (OSTI)

The DORT code has been certified. DORT is a two-dimensional discrete ordinates transport theory code, that can solve neutron, photon, or coupled neutron/photon problems. It is anticipated that DORT will be used for criticality calculations as well as for shielding and radiation field analysis at SRS. In addition to the DORT module itself, 5 utility programs that are useful in certain DORT applications have been certified. These modules are: GIP, DOS, GRTUNCL, BNDRYS, and RTFLUM. As defined in this work, certification does not imply validation. These codes must be validated for a particular type of calculation before they can be used for critical applications.

Frost, R.L.

1994-01-01T23:59:59.000Z

276

AMPX-77 Phase 1 certification package  

SciTech Connect (OSTI)

The AMPX-77 Phase 1 modules have been certified. AMPX-77 is a modular code system for generating coupled multigroup neutron-gamma cross section libraries from Evaluated Nuclear Data Files (ENDF/B). All basic cross-section data are input from the formats used by the ENDF/B, and output can be obtained from a variety of formats, included in its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-ray data. The AMPX-77 code system will be used at SRS to perform critical calculations related to nuclear criticality safety. The AMPX-77 modular codes system contains forty-seven separate modules. For the certification process, the 47 modules have been divided into three groups or phases. This Certification Package is for the Phase 1 modules: BONAMI, LAPHNGAS, MALOCS, NITAWL, ROLAIDS, SMUG, and XSDRNPM.

Niemer, K.A.

1994-03-01T23:59:59.000Z

277

Contact-Handled and Remote-Handled Transuranic Waste Packaging  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

Provides specific instructions for packaging and/or repackaging contact-handled transuranic (CH-TRU) and remote-handled transuranic (RH-TRU) waste in a manner consistent with DOE O 435.1, Radioactive Waste Management, DOE M 435.1-1 Chg 1, Radioactive Waste Management Manual, CH-TRU and RH-TRU waste transportation requirements, and Waste Isolation Pilot Plant (WIPP) programmatic requirements. Does not cancel other directives.

2011-08-09T23:59:59.000Z

278

Initial Package Design Concepts Integrated Product Team (IPT) Summary Report  

SciTech Connect (OSTI)

Initially, the question of transporting TRU waste to WIPP was raised as part of the EM Integration activities. The issue was re-examined as part of the system-wide view to re-engineer the TRU waste program. Consequently, the National Transportation Program and the National TRU Waste Program, in a cooperative effort, made a commitment to EM-20 to examine the feasibility of using rail to transport TRU waste material to WIPP. In December of 1999 Mr. Philip Altomare assembled a team of subject matter experts (SME) to define initial concepts for a Type B package capable of shipping TRU waste by rail (see Attachment 1 for a list of team members). This same team of experts also provided input to a preliminary study to determine if shipping TRU waste by rail could offer cost savings or other significant advantages over the current mode of operation using TRUPACT-II packages loaded on truck. As part of the analysis, the team also identified barriers to implementing rail shipments to WIPP and outlined a path forward. This report documents the findings of the study and its initial set of recommendations. As the study progressed, it was expanded to include new packages for truck as well as rail in recognition of the benefits of shipping large boxes and contaminated equipment.

Moss, J.; Luke, Dale Elden

2000-03-01T23:59:59.000Z

279

Puncture evaluation of Shippingport package  

SciTech Connect (OSTI)

A puncture evaluation of a 900-ton type B category II shipping package was performed. The package consisted of the decommissioned Shippingport reactor pressure vessel (RPV) with its neutron shield tank (NST) in a concentric arrangement. The space inside the RPV and in annulus between the RPV and the 1-inch-thick NST was filled with concrete. The package was assumed to drop 40 inches into a 6-inch-diameter puncture bar of sufficient length to reach the RPV. The objective was to evaluate the puncture potential of the RPV. A nonlinear dynamic finite element analysis was performed. The NST and the concrete in the annulus were assumed to provide little resistance to puncture because the NST shell is thin and the concrete strength is low. In addition to the dynamic finite element evaluation of the package, a simple buckling analysis of the puncture bar was also performed. The buckling analysis was based on the tangent modulus theory of inelastic buckling. It was found that the puncture bar will not penetrate the RPV under the most severe stress state during the impact process. It was also found that the puncture bar will buckle long before this most severe stress state in the RPV can be reached. The package possesses so much kinetic energy before impact, a small fraction of this energy is sufficient to either buckle or overstress the puncture bar before the stresses in the RPV become critical. 5 refs., 3 figs., 1 tab.

Lo, Ting-Yu

1989-05-26T23:59:59.000Z

280

Package for integrated optic circuit and method  

DOE Patents [OSTI]

A structure and method for packaging an integrated optic circuit. The package comprises a first wall having a plurality of microlenses formed therein to establish channels of optical communication with an integrated optic circuit within the package. A first registration pattern is provided on an inside surface of one of the walls of the package for alignment and attachment of the integrated optic circuit. The package in one embodiment may further comprise a fiber holder for aligning and attaching a plurality of optical fibers to the package and extending the channels of optical communication to the fibers outside the package. In another embodiment, a fiber holder may be used to hold the fibers and align the fibers to the package. The fiber holder may be detachably connected to the package.

Kravitz, Stanley H. (26 Aspen Rd., Placitas, NM 87043); Hadley, G. Ronald (6012 Annapolis NE., Albuquerque, NM 87111); Warren, Mial E. (3825 Mary Ellen NE., Albuquerque, NM 87111); Carson, Richard F. (1036 Jewel Pl. NE., Albuquerque, NM 87123); Armendariz, Marcelino G. (1023 Oro Real NE., Albuquerque, NM 87123)

1998-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

MCNP certification package  

SciTech Connect (OSTI)

In response to a Department of Energy (DOE) request, Westinghouse Savannah River Company committed to certify all computer codes used in critical calculations at the site. Since the Monte Carlo Neutron Photon Transport (MCNP) code will be used to perform critical analyses involving criticality and shielding, the code must be certified. Certification as applied to existing computer codes includes the verification and validation process, placing the code in configuration control, and establishing user qualification standards and training requirements. All software intended for use in critical calculations must be certified. This report is intended to fulfill the requirements for the certification of the MCNP code, version 4.2, built June 11, 1992, by J.H. Hightower on the SRS CRAY. This report does not release MCNP for use under production status for any application for which a MCNP validation document does not exist. These validation documents will describe the specific range of applicability, limitations on use, results and biases for a particular MCNP application.

Trumble, E.F.

1992-08-01T23:59:59.000Z

282

Trans_package_Poster_Draft_8_7_12.indd  

National Nuclear Security Administration (NNSA)

Lead Shield Design and Testing Requirements: Plutonium, unirradiated mixed oxide (MOX) fuel assemblies, and transuranic (TRU) waste would be transported in U.S. Nuclear Regulatory...

283

DOE nuclear material packaging manual: storage container requirements for plutonium oxide materials  

SciTech Connect (OSTI)

Loss of containment of nuclear material stored in containers such as food-pack cans, paint cans, or taped slip lid cans has generated concern about packaging requirements for interim storage of nuclear materials in working facilities such as the plutonium facility at Los Alamos National Laboratory (LANL). In response, DOE has recently issued DOE M 441.1 'Nuclear Material Packaging Manual' with encouragement from the Defense Nuclear Facilities Safety Board. A unique feature compared to transportation containers is the allowance of filters to vent flammable gases during storage. Defining commonly used concepts such as maximum allowable working pressure and He leak rate criteria become problematic when considering vented containers. Los Alamos has developed a set of container requirements that are in compliance with 441.1 based upon the activity of heat-source plutonium (90% Pu-238) oxide, which bounds the requirements for weapons-grade plutonium oxide. The pre and post drop-test He leak rates depend upon container size as well as the material contents. For containers that are routinely handled, ease of handling and weight are a major consideration. Relatively thin-walled containers with flat bottoms are desired yet they cannot be He leak tested at a differential pressure of one atmosphere due to the potential for plastic deformation of the flat bottom during testing. The He leak rates and He leak testing configuration for containers designed for plutonium bearing materials will be presented. The approach to meeting the other manual requirements such as corrosion and thermal degradation resistance will be addressed. The information presented can be used by other sites to evaluate if their conditions are bounded by LANL requirements when considering procurement of 441.1 compliant containers.

Veirs, D Kirk [Los Alamos National Laboratory

2009-01-01T23:59:59.000Z

284

Safety analysis report for packaging (Onsite) for the Hanford Ecorok packaging  

SciTech Connect (OSTI)

This safety analysis report for packaging approves the Hanford Ecorok packaging for shipping contaminated water purification filters from K Basins to the Central Waste Complex.

Mercado, M.S.

1996-02-23T23:59:59.000Z

285

Radioisotope thermoelectric generator transport trailer system  

SciTech Connect (OSTI)

The Radioisotope Thermoelectric Generator (RTG) Transportation System, designated as System 100, comprises four major systems. The four major systems are designated as the Packaging System (System 120), Trailer System (System 140), Operations and Ancillary Equipment System (System 160), and Shipping and Receiving Facility Transport System (System 180). Packaging System (System 120), including the RTG packaging is licensed (regulatory) hardware; it is certified by the U.S. Department of Energy to be in accordance with Title 10, {ital Code} {ital of} {ital Federal} {ital Regulations}, Part 71 (10 CFR 71). System 140, System 160, and System 180 are nonlicensed (nonregulatory) hardware. {copyright} {ital 1995} {ital American} {ital Institute} {ital of} {ital Physics}

Ard, K.E.; King, D.A.; Leigh, H.; Satoh, J.A. [Westinghouse Hanford Company, P.O. Box 1970, MSIN N1-25, Richland, Washington 99352 (United States)

1995-01-20T23:59:59.000Z

286

Hanford Site radioactive hazardous materials packaging directory  

SciTech Connect (OSTI)

The Hanford Site Radioactive Hazardous Materials Packaging Directory (RHMPD) provides information concerning packagings owned or routinely leased by Westinghouse Hanford Company (WHC) for offsite shipments or onsite transfers of hazardous materials. Specific information is provided for selected packagings including the following: general description; approval documents/specifications (Certificates of Compliance and Safety Analysis Reports for Packaging); technical information (drawing numbers and dimensions); approved contents; areas of operation; and general information. Packaging Operations & Development (PO&D) maintains the RHMPD and may be contacted for additional information or assistance in obtaining referenced documentation or assistance concerning packaging selection, availability, and usage.

McCarthy, T.L.

1995-12-01T23:59:59.000Z

287

Investigations into High Temperature Components and Packaging  

SciTech Connect (OSTI)

The purpose of this report is to document the work that was performed at the Oak Ridge National Laboratory (ORNL) in support of the development of high temperature power electronics and components with monies remaining from the Semikron High Temperature Inverter Project managed by the National Energy Technology Laboratory (NETL). High temperature electronic components are needed to allow inverters to operate in more extreme operating conditions as required in advanced traction drive applications. The trend to try to eliminate secondary cooling loops and utilize the internal combustion (IC) cooling system, which operates with approximately 105 C water/ethylene glycol coolant at the output of the radiator, is necessary to further reduce vehicle costs and weight. The activity documented in this report includes development and testing of high temperature components, activities in support of high temperature testing, an assessment of several component packaging methods, and how elevated operating temperatures would impact their reliability. This report is organized with testing of new high temperature capacitors in Section 2 and testing of new 150 C junction temperature trench insulated gate bipolar transistor (IGBTs) in Section 3. Section 4 addresses some operational OPAL-GT information, which was necessary for developing module level tests. Section 5 summarizes calibration of equipment needed for the high temperature testing. Section 6 details some additional work that was funded on silicon carbide (SiC) device testing for high temperature use, and Section 7 is the complete text of a report funded from this effort summarizing packaging methods and their reliability issues for use in high temperature power electronics. Components were tested to evaluate the performance characteristics of the component at different operating temperatures. The temperature of the component is determined by the ambient temperature (i.e., temperature surrounding the device) plus the temperature increase inside the device due the internal heat that is generated due to conduction and switching losses. Capacitors and high current switches that are reliable and meet performance specifications over an increased temperature range are necessary to realize electronics needed for hybrid-electric vehicles (HEVs), fuel cell (FC) and plug-in HEVs (PHEVs). In addition to individual component level testing, it is necessary to evaluate and perform long term module level testing to ascertain the effects of high temperature operation on power electronics.

Marlino, L.D.; Seiber, L.E.; Scudiere, M.B.; M.S. Chinthavali, M.S.; McCluskey, F.P.

2007-12-31T23:59:59.000Z

288

U-003:RPM Package Manager security update | Department of Energy  

Broader source: Energy.gov (indexed) [DOE]

and updating software packages. Impact: Multiple flaws were found in the way the RPM library parsed package headers. An attacker could create a specially-crafted RPM package...

289

Packages in the `graphics' bundle D. P. Carlisle  

E-Print Network [OSTI]

Packages in the `graphics' bundle D. P. Carlisle The Graphics packages 7 4.1 Package Options Graphics Files . . . . . . . . . . . . . . . . . . . . . . .* * 9 4.5 Other commands in the graphics

290

Packages in the `graphics' bundle D. P. Carlisle  

E-Print Network [OSTI]

Packages in the `graphics' bundle D. P. Carlisle The Graphics packages 6 4.1 Package Options Graphics Files . . . . . . . . . . . . . . . . . . . . . . .* * 8 4.5 Other commands in the graphics

Stein, William

291

Examples of Cost Estimation Packages  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

Estimates can be performed in a variety of ways. Some of these are for projects for an undefined scope, a conventional construction project, or where there is a level of effort required to complete the work. Examples of cost estimation packages for these types of projects are described in this appendix.

1997-03-28T23:59:59.000Z

292

Qualification Status List (QSL Package)  

E-Print Network [OSTI]

qualification of Array B. These open items are summarized on page 2 and the equipment subsections which follow#12;ALSEP Qualification Status List (QSL Package) Flight 3 Configuration #12;NO. ltiY. NO. RI!V. NO. Qualification Status List ALSEP Array B Configuration 1. 0 INTRODUCTION ATM 825 PAGI! 1

Rathbun, Julie A.

293

Vadose Zone Hydrogeology Data Package for Hanford Assessments  

SciTech Connect (OSTI)

This data package documents the technical basis for selecting physical and geochemical parameters and input values that will be used in vadose zone modeling for Hanford assessments. This work was originally conducted as part of the Characterization of Systems Task of the Groundwater Remediation Project managed by Fluor Hanford, Inc., Richland, Washington, and revised as part of the Characterization of Systems Project managed by the Pacific Northwest National Laboratory (PNNL) for the U.S. Department of Energy, Richland Operations Office (DOE-RL). This data package describes the geologic framework, the physical, hydrologic, and contaminant transport properties of the geologic materials, and deep drainage (i.e., recharge) estimates, and builds on the general framework developed for the initial assessment conducted using the System Assessment Capability (SAC) (Bryce et al. 2002). The general approach for this work was to update and provide incremental improvements over the previous SAC data package completed in 2001. As with the previous SAC data package, much of the data and interpreted information were extracted from existing documents and databases. Every attempt was made to provide traceability to the original source(s) of the data or interpretations.

Last, George V.; Freeman, Eugene J.; Cantrell, Kirk J.; Fayer, Michael J.; Gee, Glendon W.; Nichols, William E.; Bjornstad, Bruce N.; Horton, Duane G.

2006-06-01T23:59:59.000Z

294

Packaging architecture: a critique of architectural commodification  

E-Print Network [OSTI]

's shared condition as packaged commodities provide the basis for Fridge, a computer animation that develops visual analogies between milk, the machine, packaging, and modem architecture. Constraints and architectural skills engaged in translating...

Pushpathadam, Thomas

1996-01-01T23:59:59.000Z

295

Harsh-Environment Packaging for Downhole Gas and Oil Exploration  

SciTech Connect (OSTI)

This research into new packaging materials and methods for elevated temperatures and harsh environment electronics focused on gaining a basic understanding of current state-of-the-art in electronics packaging used in industry today, formulating the thermal-mechanical models of the material interactions and developing test structures to confirm these models. Discussions were initiated with the major General Electric (GE) businesses that currently sell into markets requiring high temperature electronics and packaging. They related the major modes of failure they encounter routinely and the hurdles needed to be overcome in order to improve the temperature specifications of these products. We consulted with our GE business partners about the reliability specifications and investigated specifications and guidelines that from IPC and the SAE body that is currently developing guidelines for electronics package reliability. Following this, a risk analysis was conducted for the program to identify the critical risks which need to be mitigated in order to demonstrate a flex-based packaging approach under these conditions. This process identified metal/polyimide adhesion, via reliability for flex substrates and high temperature interconnect as important technical areas for reliability improvement.

Shubhra Bansal; Junghyun Cho; Kevin Durocher; Chris Kapusta; Aaron Knobloch; David Shaddock; Harry Schoeller; Hua Xia

2007-08-31T23:59:59.000Z

296

Final evaluation report for Westinghouse Hanford Company, WRAP-1,208 liter waste drum, docket 94-35-7A, type A packaging  

SciTech Connect (OSTI)

This report documents the U.S. Department of Transportation Specification 7A Type A (DOT-7A) compliance test results of the Westinghouse Hanford Company, Waste Receiving and Processing Facility, Module 1 (WRAP-1) Drum. The WRAP-1 Drum was tested for DOE-HQ in August 1994, by Los Alamos National Laboratory, under docket number 94-35-7A. Additionally, comparison and evaluation of the approved, as-tested packaging configuration was performed by WHC in September 1995. The WRAP-1 Drum was evaluated against the performance of the DOT-17C, 208 1 (55-gal) steel drums tested and evaluated under dockets 89-13-7A/90-18-7A and 94-37-7A.

Kelly, D.L., Westinghouse Hanford

1996-06-12T23:59:59.000Z

297

Borehole completion data package for well 199-N-81  

SciTech Connect (OSTI)

Well 199-N-81 was drilled in 1993 as a RCRA groundwater monitoring for the 1324-N network. The well is completed at the top of the uppermost aquifer, in the Ringold Formation. This data package includes information on drilling, construction, development, and aquifer testing. Copies of forms, notes, and diagrams completed in the field comprise the bulk of this document. Few interpretations are included. Lithologic contacts were picked by the site geologist. An attempt was made to interpret aquifer test data.

Hartman, M.J.

1994-05-05T23:59:59.000Z

298

Electrochemical Corrosion Testing of Borated Stainless Steel Alloys  

SciTech Connect (OSTI)

The Department of Energy Office of Civilian Radioactive Waste Management has specified borated stainless steel manufactured to the requirements of ASTM A 887-89, Grade A, UNS S30464, to be the material used for the fabrication of the fuel basket internals of the preliminary transportation, aging, and disposal canister system preliminary design. The long-term corrosion resistance performance of this class of borated materials must be verified when exposed to expected YMP repository conditions after a waste package breach. Electrochemical corrosion tests were performed on crevice corrosion coupons of Type 304 B4 and Type 304 B5 borated stainless steels exposed to single postulated in-package chemistry at 60°C. The results show low corrosion rates for the test period

lister, tedd e; Mizia, Ronald E

2007-09-01T23:59:59.000Z

299

Electrochemical Corrosion Testing of Borated Stainless Steel Alloys  

SciTech Connect (OSTI)

The Department of Energy Office of Civilian Radioactive Waste Management has specified borated stainless steel manufactured to the requirements of ASTM A 887-89, Grade A, UNS S30464, to be the material used for the fabrication of the fuel basket internals of the preliminary transportation, aging, and disposal canister system preliminary design. The long-term corrosion resistance performance of this class of borated materials must be verified when exposed to expected YMP repository conditions after a waste package breach. Electrochemical corrosion tests were performed on crevice corrosion coupons of Type 304 B4 and Type 304 B5 borated stainless steels exposed to single postulated in-package chemistry at 60°C. The results show low corrosion rates for the test period

lister, tedd e; Mizia, Ronald E

2007-05-01T23:59:59.000Z

300

NEW APPROACH TO ADDRESSING GAS GENERATION IN RADIOACTIVE MATERIAL PACKAGING  

SciTech Connect (OSTI)

Safety Analysis Reports for Packaging (SARP) document why the transportation of radioactive material is safe in Type A(F) and Type B shipping containers. The content evaluation of certain actinide materials require that the gas generation characteristics be addressed. Most packages used to transport actinides impose extremely restrictive limits on moisture content and oxide stabilization to control or prevent flammable gas generation. These requirements prevent some users from using a shipping container even though the material to be shipped is fully compliant with the remaining content envelope including isotopic distribution. To avoid these restrictions, gas generation issues have to be addressed on a case by case basis rather than a one size fits all approach. In addition, SARP applicants and review groups may not have the knowledge and experience with actinide chemistry and other factors affecting gas generation, which facility experts in actinide material processing have obtained in the last sixty years. This paper will address a proposal to create a Gas Generation Evaluation Committee to evaluate gas generation issues associated with Safety Analysis Reports for Packaging material contents. The committee charter could include reviews of both SARP approved contents and new contents not previously evaluated in a SARP.

Watkins, R; Leduc, D; Askew, N

2009-06-25T23:59:59.000Z

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

YUCCA MOUNTAIN WASTE PACKAGE CLOSURE SYSTEM  

SciTech Connect (OSTI)

The method selected for dealing with spent nuclear fuel in the US is to seal the fuel in waste packages and then to place them in an underground repository at the Yucca Mountain Site in Nevada. This article describes the Waste Package Closure System (WPCS) currently being designed for sealing the waste packages.

G. Housley; C. Shelton-davis; K. Skinner

2005-08-26T23:59:59.000Z

302

ASSISTANT PROFESSOR (SEMI-RIGID PLASTIC PACKAGING)  

E-Print Network [OSTI]

ASSISTANT PROFESSOR (SEMI-RIGID PLASTIC PACKAGING) IN FOOD, NUTRITION, & PACKAGING SCIENCES/rigid plastics packaging. The successful candidate will be responsible for teaching undergraduate and graduate in university teaching, experience in designing and implementing semi-rigid/rigid plastics research, a record

Stuart, Steven J.

303

LM337...KTE, KTP, OR KVU PACKAGE LM337...KTT (TO-263) PACKAGE  

E-Print Network [OSTI]

1FEATURES LM337...KTE, KTP, OR KVU PACKAGE (TOP VIEW) OUTPUT INPUT ADJUSTMENT INPUT INPUT LM337...KTT (TO-263) PACKAGE (TOP VIEW) OUTPUT INPUT ADJUSTMENT LM237, LM337...KC (TO-220) PACKAGE (TOP VIEW) INPUT OUTPUT ADJUSTMENT INPUT OUTPUT LM337...KCS (TO-220) PACKAGE (TOP VIEW) ADJUSTMENT INPUT INPUT

Ravikumar, B.

304

Packaging review guide for reviewing safety analysis reports for packagings: Revision 1  

SciTech Connect (OSTI)

The Department of Energy (DOE) has established procedures for obtaining certification of packagings used by DOE and its contractors for the transport of radioactive materials. The principal purpose of this document is to assure the quality and uniformity of PCS reviews and to present a well-defined base from which to evaluate proposed changes in the scope and requirements of reviews. The Packaging Review Guide (PRG) also sets forth solutions and approaches determined to be acceptable in the past in dealing with a specific safety issue or safety-related design area. These solutions and approaches are presented in this form so that reviewers can take consistent and well-understood positions as the same safety issues arise in future cases. An applicant submitting a SARP does not have to follow the solutions or approaches presented. It is also a purpose of the PRG to make information about DOE certification policy and procedures widely available to DOE field offices, DOE contractors, federal agencies, and interested members of the public. 77 refs., 16 figs., 15 tabs.

Fisher, L.E.; Chou, C.K.; Lloyd, W.R.; Mount, M.E.; Nelson, T.A.; Schwartz, M.W.; Witte, M.C.

1988-10-01T23:59:59.000Z

305

TRANSPORT AND EMPLACEMENT EQUIPMENT DESCRIPTIONS  

SciTech Connect (OSTI)

The objective and the scope of this document are to list and briefly describe the major mobile equipment necessary for waste package (WP) Transport and Emplacement in the proposed subsurface nuclear waste repository at Yucca Mountain. Primary performance characteristics and some specialized design features of the equipment are explained and summarized in the individual subsections of this document. The Transport and Emplacement equipment described in this document consists of the following: (1) WP Transporter; (2) Reusable Rail Car; (3) Emplacement Gantry; (4) Gantry Carrier; and (5) Transport Locomotive.

NA

1997-09-29T23:59:59.000Z

306

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

SciTech Connect (OSTI)

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-overpacked package to subsequently require overpacking. Therefore, the polyurea coating can not be credited to avoid the need to overpack a package or enhance the transportation safety of a structurally sound package that has polyurea on the exterior.

VAIL, T.S.

2007-03-30T23:59:59.000Z

307

RECERTIFICATION OF THE MODEL 9977 RADIOACTIVE MATERIAL PACKAGING  

SciTech Connect (OSTI)

The Model 9977 Packaging was initially issued a Certificate of Compliance (CoC) by the Department of Energy’s Office of Environmental Management (DOE-EM) for the transportation of radioactive material (RAM) in the Fall of 2007. This first CoC was for a single radioactive material and two packing configurations. In the five years since that time, seven Addendums have been written to the Safety Analysis Report for Packaging (SARP) and five Letter Amendments have been written that have authorized either new RAM contents or packing configurations, or both. This paper will discuss the process of updating the 9977 SARP to include all the contents and configurations, including the addition of a new content, and its submittal for recertification.

Abramczyk, G.; Bellamy, S.; Loftin, B.; Nathan, S.

2013-06-05T23:59:59.000Z

308

Safety evaluation for packaging (onsite) nitrogen trailers propane tanks  

SciTech Connect (OSTI)

The purpose of the Safety Evaluation for Packaging (SEP) is the evaluation and authorization of the onsite transport of propane tanks that are mounted on the Lockheed Martin Hanford Corporation Characterization Project`s nitrogen trailers. This SEP authorizes onsite transport of the nitrogen trailers, including the propane tanks, until May 31, 1998. The three nitrogen trailers (HO-64-4966, HO-64-4968, and HO-64-5170) are rated for 1,361 kg (30,000 lb) and are equipped with tandem axles and pintel hitches. Permanently mounted on each trailer is a 5,678 L (1,500 gal) cryogenic dewar that is filled with nitrogen, and a propane fired water bath vaporizer system, and a 454 L (1 20 gal) propane tank. The nitrogen trailer system is operated only when it is disconnected from the tow vehicle and is leveled and stabilized. When the trailers are transported, the propane tanks are isolated via closed supply valves.

Ferrell, P.C.

1998-01-28T23:59:59.000Z

309

WASTE PACKAGE OPERATIONS FY99 CLOSURE METHODS REPORT  

SciTech Connect (OSTI)

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.

M. C. Knapp

1999-09-23T23:59:59.000Z

310

EXAMINATION OF SHIPPING PACKAGE 9975-05050  

SciTech Connect (OSTI)

Shipping package 9975-05050 was examined in K-Area following its identification as a high wattage package. Elevated temperature and fiberboard moisture content are key parameters that impact the degradation rate of fiberboard within 9975 packages in a storage environment. The high wattage of this package contributes significantly to component temperatures. After examination in K-Area, the package was provided to SRNL for further examination of the fiberboard assembly. The moisture content of the fiberboard was relatively low (compared to packages examined previously), but the moisture gradient (between fiberboard ID and OD surfaces) was relatively high, as would be expected for the high heat load. The cane fiberboard appeared intact and displayed no apparent change in integrity relative to a new package.

Daugherty, W.

2014-11-06T23:59:59.000Z

311

Field studies of the potential for wind transport of plutonium- contaminated soils at sites in Areas 6 and 11, Nevada Test Site  

SciTech Connect (OSTI)

This report describes and documents a series of field experiments carried out in Areas 6 and 11 of the Nevada Test Site in June and July 1994 to determine parameters of boundary layer winds, surface characteristics, and vegetation cover that can be used to predict dust emissions from the affected sites. Aerodynamic roughness of natural sites is determined largely by the lateral cover of the larger and more permanent roughness elements (shrubs). These provide a complete protection of the surface from wind erosion. Studies using a field-portable wind tunnel demonstrated that natural surfaces in the investigated areas of the Nevada Test Site are stable except at very high wind speeds (probably higher than normally occur, except perhaps in dust devils). However, disturbance of silty-clay surfaces by excavation devices and vehicles reduces the entrainment threshold by approximately 50% and makes these areas potentially very susceptible to wind erosion and transport of sediments.

Lancaster, N.; Bamford, R.; Metzger, S. [University and Community Coll. System of Nevada, Reno, NV (United States). Quaternary Sciences Center, Desert Research Institute

1995-07-01T23:59:59.000Z

312

Performance-oriented packaging: A guide to identifying and designing. Identifying and designing hazardous materials packaging for compliance with post HM-181 DOT Regulations  

SciTech Connect (OSTI)

With the initial publication of Docket HM-181 (hereafter referred to as HM-181), the U.S. Department of Energy (DOE), Headquarters, Transportation Management Division decided to produce guidance to help the DOE community transition to performance-oriented packagings (POP). As only a few individuals were familiar with the new requirements, elementary guidance was desirable. The decision was to prepare the guidance at a level easily understood by a novice to regulatory requirements. This document identifies design development strategies for use in obtaining performance-oriented packagings that are not readily available commercially. These design development strategies will be part of the methodologies for compliance with post HM-181 U.S. Department of Transportation (DOT) packaging regulations. This information was prepared for use by the DOE and its contractors. The document provides guidance for making decisions associated with designing performance-oriented packaging, and not for identifying specific material or fabrication design details. It does provide some specific design considerations. Having a copy of the regulations handy when reading this document is recommended to permit a fuller understanding of the requirements impacting the design effort. While this document is not written for the packaging specialist, it does contain guidance important to those not familiar with the new POP requirements.

Not Available

1994-08-01T23:59:59.000Z

313

9975 SHIPPING PACKAGE LIFE EXTENSION SURVEILLANCE PROGRAM RESULTS SUMMARY  

SciTech Connect (OSTI)

Results from the 9975 Surveillance Program at the Savannah River Site (SRS) are summarized for justification to extend the life of the 9975 packages currently stored in the K-Area Materials Storage (KAMS) facility from 10 years to 15 years. This justification is established with the stipulation that surveillance activities will continue throughout this extended time to ensure the continued integrity of the 9975 materials of construction and to further understand the currently identified degradation mechanisms. The current 10 year storage life was developed prior to storage. A subsequent report was later used to extend the qualification of the 9975 shipping packages for 2 years for shipping plus 10 years for storage. However the qualification for the storage period was provided by the monitoring requirements of the Storage and Surveillance Program. This report summarizes efforts to determine a new safe storage limit for the 9975 shipping package based on the surveillance data collected since 2005 when the surveillance program began. KAMS is a zero-release facility that depends upon containment by the 9975 to meet design basis storage requirements. Therefore, to confirm the continued integrity of the 9975 packages while stored in KAMS, a 9975 Storage and Surveillance Program was implemented alongside the DOE required Integrated Surveillance Program (ISP) for 3013 plutonium-bearing containers. The 9975 Storage and Surveillance Program performs field surveillance as well as accelerated aging tests to ensure any degradation due to aging, to the extent that could affect packaging performance, is detected in advance of such degradation occurring in the field. The Program has demonstrated that the 9975 package has a robust design that can perform under a variety of conditions. As such the primary emphasis of the on-going 9975 Surveillance Program is an aging study of the 9975 Viton(reg.sign) GLT containment vessel O-rings and the Celotex(reg.sign) fiberboard thermal insulation at bounding conditions of radiation and elevated temperatures. Other materials of construction, however, are also discussed.

Daugherty, W.; Dunn, K.; Hackney, B.; Hoffman, E.; Skidmore, E.

2011-01-06T23:59:59.000Z

314

Pre-bid network analysis for transportation procurement auction under stochastic demand  

E-Print Network [OSTI]

Transportation procurement is one of the most critical sourcing decisions to be made in many companies. This thesis addresses a real-life industrial problem of creating package bids for a company's transportation procurement ...

Wang, Qian

2007-01-01T23:59:59.000Z

315

Safety analysis report for packaging for the Idaho National Engineering Laboratory TRA Type 1 Shipping Container and TRA Type 2 Shipping Capsule  

SciTech Connect (OSTI)

The TRA Type I Shipping Container and TRA Type II Shipping Capsule were designed and fabricated at the Idaho National Engineering Laboratory as special form containers for the transport of non-fissile radioisotopes and fissile radioisotopes in exempt quantities. The Type I container measures 0.75 in. outside diameter and 3.000 in long. The Type II capsule is 0.495 in. outside diameter 2.000 in. long. The container and capsule were tested and evaluated to determine their compliance with Title 49 Code of Federal Regulations 173, which governs packages for special form radioactive material. This report is based upon those tests and evaluations. The results of those tests and evaluations demonstrate the container and capsule are in full compliance with the special form shipping container regulations of 49 CFR 173.

Havlovick, B.J.

1992-07-27T23:59:59.000Z

316

Packaging  

E-Print Network [OSTI]

products to building brand images. An Infiniti automobilea clearly defined brand image. ” 114 The growth of theof presenting a meta-brand or image appealing to so many but

Galloway, Catherine Suzanne

2012-01-01T23:59:59.000Z

317

Challenges in the Packaging of MEMS  

SciTech Connect (OSTI)

Microelectromechanical Systems (MEMS) packaging is much different from conventional integrated circuit (IC) packaging. Many MEMS devices must interface to the environment in order to perform their intended function, and the package must be able to facilitate access with the environment while protecting the device. The package must also not interfere with or impede the operation of the MEMS device. The die attachment material should be low stress, and low outgassing, while also minimizing stress relaxation overtime which can lead to scale factor shifts in sensor devices. The fabrication processes used in creating the devices must be compatible with each other, and not result in damage to the devices. Many devices are application specific requiring custom packages that are not commercially available. Devices may also need media compatible packages that can protect the devices from harsh environments in which the MEMS device may operate. Techniques are being developed to handle, process, and package the devices such that high yields of functional packaged parts will result. Currently, many of the processing steps are potentially harmful to MEMS devices and negatively affect yield. It is the objective of this paper to review and discuss packaging challenges that exist for MEMS systems and to expose these issues to new audiences from the integrated circuit packaging community.

BROWN, WILLIAM D.; EATON, WILLIAM P.; MALSHE, AJAY P.; MILLER, WILLIAM M.; O'NEAL, CHAD; SINGH, SUSHILA B.

1999-09-24T23:59:59.000Z

318

Quality assurance for radioactive waste packages -- A general approach  

SciTech Connect (OSTI)

Radioactive waste packages must fulfill the requirements resulting from regulations concerning handling, treatment, conditioning, transportation, storage and disposal so that the goal of radioactive waste management can be achieved. Usually in different parts of waste management different quality systems are used, and different quality assurance measures are performed. In the paper, these problems ar elucidated and it is explained by means of the quality assurance performed for the disposal of radioactive waste in Germany how the fulfillment of the requirements of the repository can be ensured.

Martens, B.R. [Bundesamt fuer Strahlenschutz, Saltzgitter (Germany)

1993-12-31T23:59:59.000Z

319

FRAMES Software System: Linking to the Statistical Package R  

SciTech Connect (OSTI)

This document provides requirements, design, data-file specifications, test plan, and Quality Assurance/Quality Control protocol for the linkage between the statistical package R and the Framework for Risk Analysis in Multimedia Environmental Systems (FRAMES) Versions 1.x and 2.0. The requirements identify the attributes of the system. The design describes how the system will be structured to meet those requirements. The specification presents the specific modifications to FRAMES to meet the requirements and design. The test plan confirms that the basic functionality listed in the requirements (black box testing) actually functions as designed, and QA/QC confirms that the software meets the client’s needs.

Castleton, Karl J.; Whelan, Gene; Hoopes, Bonnie L.

2006-12-11T23:59:59.000Z

320

Packaging and the Supply Chain: A Look at Transportation  

E-Print Network [OSTI]

a producer may ship only a partial load by choice. However,of pallet utilization for partial load shipments will beare commonly shipped in a partial load, it is difficult to

Simon, Rachel; Chen, Yifen

2013-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

Packaging and the Supply Chain: A Look at Transportation  

E-Print Network [OSTI]

Distribution Distribution Use Recycle Incineration/ landfillUse Incineration/ landfill Fig. 10.3 The allocation

Simon, Rachel; Chen, Yifen

2013-01-01T23:59:59.000Z

322

Communication Is Key to Packaging and Transportation Safety and Compliance  

Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-Up fromDepartmentTieCelebrate Earth Codestheat Two Aluminum Sheet| Department of

323

Hazardous Materials Packaging and Transportation Safety - DOE Directives,  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation ProposedUsingFun with Bigfront.jpgcommunity200cell 9Harvey Brooks, 1960Options

324

Office of Packaging and Transportation Fiscal Year 2012 Annual Report |  

Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergyDepartment of Energy(National1EnergyFederal BureauDepartment of

325

EM Office of Packaging and Transportation | Department of Energy  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power SystemsResources DOE ZeroThreeEnergyDepartment0: DOE512: Alaska EM|ofTaxpayers |

326

DOE-Idaho's Packaging and Transportation Perspective | Department of Energy  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power SystemsResources DOE Zero Energy Ready Home16, 2009133-20147502-95

327

Using on-package memory  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLasDelivered energy consumption by sectorlongUpdates byUser GuideHadoopUsingon-package memory

328

Packaging of electro-microfluidic devices  

DOE Patents [OSTI]

A new architecture for packaging surface micromachined electro-microfluidic devices is presented. This architecture relies on two scales of packaging to bring fluid to the device scale (picoliters) from the macro-scale (microliters). The architecture emulates and utilizes electronics packaging technology. The larger package consists of a circuit board with embedded fluidic channels and standard fluidic connectors (e.g. Fluidic Printed Wiring Board). The embedded channels connect to the smaller package, an Electro-Microfluidic Dual-Inline-Package (EMDIP) that takes fluid to the microfluidic integrated circuit (MIC). The fluidic connection is made to the back of the MIC through Bosch-etched holes that take fluid to surface micromachined channels on the front of the MIC. Electrical connection is made to bond pads on the front of the MIC.

Benavides, Gilbert L. (Albuquerque, NM); Galambos, Paul C. (Albuquerque, NM); Emerson, John A. (Albuquerque, NM); Peterson, Kenneth A. (Albuquerque, NM); Giunta, Rachel K. (Albuquerque, NM); Watson, Robert D. (Tijeras, NM)

2002-01-01T23:59:59.000Z

329

Packaging of electro-microfluidic devices  

DOE Patents [OSTI]

A new architecture for packaging surface micromachined electro-microfluidic devices is presented. This architecture relies on two scales of packaging to bring fluid to the device scale (picoliters) from the macro-scale (microliters). The architecture emulates and utilizes electronics packaging technology. The larger package consists of a circuit board with embedded fluidic channels and standard fluidic connectors (e.g. Fluidic Printed Wiring Board). The embedded channels connect to the smaller package, an Electro-Microfluidic Dual-Inline-Package (EMDIP) that takes fluid to the microfluidic integrated circuit (MIC). The fluidic connection is made to the back of the MIC through Bosch-etched holes that take fluid to surface micromachined channels on the front of the MIC. Electrical connection is made to bond pads on the front of the MIC.

Benavides, Gilbert L. (Albuquerque, NM); Galambos, Paul C. (Albuquerque, NM); Emerson, John A. (Albuquerque, NM); Peterson, Kenneth A. (Albuquerque, NM); Giunta, Rachel K. (Albuquerque, NM); Zamora, David Lee (Albuquerque, NM); Watson, Robert D. (Tijeras, NM)

2003-04-15T23:59:59.000Z

330

Standardized DOE Spent Nuclear Fuel Canister and Transportation System for Shipping to the National Repository  

SciTech Connect (OSTI)

The U.S.Department of Energy’s (DOE) National Spent Nuclear Fuel Program (NSNFP), located at the Idaho National Engineering and Environmental Laboratory (INEEL), has been chartered with the responsibility for developing spent nuclear fuel (SNF) standardized canisters and a transportation cask system for shipping DOE SNF to the national repository. The mandate for this development is outlined in the Memorandum of Agreement for Acceptance of Department of Energy Spent Nuclear Fuel and High-Level Radioactive Waste that states, “EM shall design and fabricate … DOE SNF canisters for shipment to RW.” (1) It also states, “EM shall be responsible for the design, NRC certification, and fabrication of the transportation cask system for DOE SNF canisters or bare DOE SNF in accordance with 10 CFR Part 71.” (2) In fulfillment of these requirements, the NSNFP has developed four SNF standardized canister configurations and has conceptually designed a versatile transportation cask system for shipping the canisters to the national repository.1 The standardized canister sizes were derived from the national repository waste package design for co-disposal of SNF with high-level waste (HLW). One SNF canister can be placed in the center of the waste package or one can be placed in one of five radial positions, replacing a HLW canister. The internal cavity of the transportation cask was derived using the same logic, matching the size of the internal cavity of the waste package. The size of the internal cavity for the transportation cask allows the shipment of multiple canister configurations with the application of a removable basket design. The standardized canisters have been designed to be loaded with DOE SNF, placed into interim storage, shipped to the national repository, and placed in a waste package without having to be reopened. Significant testing has been completed that clearly demonstrates that the standardized canisters can safely achieve their intended design goals. The transportation cask system will include all of the standard design features, with the addition of dual containment for the shipment of failed fuel. The transportation cask system will also meet the rigorous licensing requirements of the Nuclear Regulatory Commission (NRC) to ensure that the design and the methods of fabrication employed will result in a shipping cask that will safely contain the radioactive materials under all credible accident scenarios. The standardization of the SNF canisters and the versatile design of the transportation cask system will eliminate a proliferation of designs and simplify the operations at the user sites and the national repository.

Pincock, David Lynn; Morton, Dana Keith; Lengyel, Arpad Leslie

2001-02-01T23:59:59.000Z

331

NFR TRIGA package design review report  

SciTech Connect (OSTI)

The purpose of this document is to compile, present and document the formal design review of the NRF TRIGA packaging. The contents of this document include: the briefing meeting presentations, package description, design calculations, package review drawings, meeting minutes, action item lists, review comment records, final resolutions, and released drawings. This design review required more than two meeting to resolve comments. Therefore, there are three meeting minutes and two action item lists.

Clements, M.D.

1994-08-26T23:59:59.000Z

332

Leadership Transition Program (LTP) Application Package | Department...  

Broader source: Energy.gov (indexed) [DOE]

and standard form-182. LTP Application Package More Documents & Publications ITP Aluminum: Technical Working Group on Inert Anode Technologies EIS-0333: Draft Environmental...

333

CH Packaging Operations for High Wattage Waste  

SciTech Connect (OSTI)

This document provides instructions for assembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP)

Washington TRU Solutions LLC

2006-01-06T23:59:59.000Z

334

PRIDE Surveillance Projects Data Packaging Project Information Package Specification Version 1.1  

SciTech Connect (OSTI)

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

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

2010-08-31T23:59:59.000Z

335

Compilation of current literature on seals, closures, and leakage for radioactive material packagings  

SciTech Connect (OSTI)

This report presents an overview of the features that affect the sealing capability of radioactive material packagings currently certified by the US Nuclear Regulatory Commission. The report is based on a review of current literature on seals, closures, and leakage for radioactive material packagings. Federal regulations that relate to the sealing capability of radioactive material packagings, as well as basic equations for leakage calculations and some of the available leakage test procedures are presented. The factors which affect the sealing capability of a closure, including the properties of the sealing surfaces, the gasket material, the closure method and the contents are discussed in qualitative terms. Information on the general properties of both elastomer and metal gasket materials and some specific designs are presented. A summary of the seal material, closure method, and leakage tests for currently certified packagings with large diameter seals is provided. 18 figs., 9 tabs.

Warrant, M.M.; Ottinger, C.A.

1989-01-01T23:59:59.000Z

336

SITE-SCALE SATURATED ZONE TRANSPORT  

SciTech Connect (OSTI)

This work provides a site-scale transport model for calculating radionuclide transport in the saturated zone (SZ) at Yucca Mountain, for use in the abstractions model in support of ''Total System Performance Assessment for License Application'' (TSPA-LA). The purpose of this model report is to provide documentation for the components of the site-scale SZ transport model in accordance with administrative procedure AP-SIII.10Q, Models. The initial documentation of this model report was conducted under the ''Technical Work Plan For: Saturated Zone Flow and Transport Modeling and Testing'' (BSC 2003 [DIRS 163965]). The model report has been revised in accordance with the ''Technical Work Plan For: Natural System--Saturated Zone Analysis and Model Report Integration'', Section 2.1.1.4 (BSC 2004 [DIRS 171421]) to incorporate Regulatory Integration Team comments. All activities listed in the technical work plan that are appropriate to the transport model are documented in this report and are described in Section 2.1.1.4 (BSC 2004 [DIRS 171421]). This report documents: (1) the advection-dispersion transport model including matrix diffusion (Sections 6.3 and 6.4); (2) a description and validation of the transport model (Sections 6.3 and 7); (3) the numerical methods for simulating radionuclide transport (Section 6.4); (4) the parameters (sorption coefficient, Kd ) and their uncertainty distributions used for modeling radionuclide sorption (Appendices A and C); (5) the parameters used for modeling colloid-facilitated radionuclide transport (Table 4-1, Section 6.4.2.6, and Appendix B); and (6) alternative conceptual models and their dispositions (Section 6.6). The intended use of this model is to simulate transport in saturated fractured porous rock (double porosity) and alluvium. The particle-tracking method of simulating radionuclide transport is incorporated in the finite-volume heat and mass transfer numerical analysis (FEHM) computer code, (FEHM V2.20, STN: 10086-2.20-00) (LANL 2003 [DIRS 161725]) and is described in Section 6.4 of this report. FEHM is a three-dimensional (3-D), finite-volume, finite-element, heat and mass flow-and-transport code. This report documents the features and capabilities of the site-scale transport model for calculating radionuclide transport in the SZ at Yucca Mountain in support of the TSPA-LA. Correlative flow-model calculations using FEHM are carried out and documented in the model report ''Saturated Zone Site-Scale Flow Model'' (BSC 2004 [DIRS 170037]). The velocity fields are calculated by the flow model independent of the transport processes and supplied as a part of the output package from the flow model, which is then used as inputs to the transport model. Several SZ analysis model reports provide information and data needed as feed-ins for this report, and this report in turn provides technical product outputs that feed into other SZ reports. The details of inputs to the site-scale transport model are provided in Section 4.

S. KELLER

2004-11-03T23:59:59.000Z

337

A review of the safety features of 6M packagings for DOE programs  

SciTech Connect (OSTI)

This report, prepared by a US Department of Energy (DOE) Task Force and organized for clarity into two-page modules, argues that the US Department of Transportation (DOT) Specification-6M packagings (hereafter referred to as 6M packaging, or simply 6M) merit continued DOE use and, if necessary, DOE certification. This report is designed to address the specific requirements of a Safety Analysis Report for Packaging (SARP). While not a SARP, this report constitutes a compilation of all available documentation on 6M packagings. The authors individually, and the Task Force collectively, believe their investigation provides justification for the continued use of 6M packagings because they meet criteria for quality assurance and for safety under normal and accident conditions as defined by the US Nuclear Regulatory Commission (NRC) regulations. This report may be used by DOE managers to assist in deliberations on future requirements for 6M packagings as they are required to support DOE programs. For the purpose of ready evaluation, this report includes categorical topics found in Nuclear Regulatory Guide 7.9, the topical guideline for SARPs. The format, however, will (it is hoped) pleasantly surprise customary reader expectations. For, while maintaining categorical headings and subheadings found in SARPs as a skeleton, the Task Force chose to adopt the document design principles developed by Hughes Aircraft in the 1960s, ''Sequential Thematic Organization of Publications'' (STOP). 37 figs.

Not Available

1988-12-01T23:59:59.000Z

338

RADIATION HEAT TRANSFER ENVIRONMENT IN FIRE AND FURNACE TESTS OF RADIOACTIVE MATERIALS PAKCAGES  

SciTech Connect (OSTI)

The Hypothetical Accident Conditions (HAC) sequential test of radioactive materials packages includes a thermal test to confirm the ability of the package to withstand a transportation fire event. The test specified by the regulations (10 CFR 71) consists of a 30 minute, all engulfing, hydrocarbon fuel fire, with an average flame temperature of at least 800 C. The requirements specify an average emissivity for the fire of at least 0.9, which implies an essentially black radiation environment. Alternate test which provide equivalent total heat input at the 800 C time averaged environmental temperature may also be employed. When alternate tests methods are employed, such as furnace or gaseous fuel fires, the equivalence of the radiation environment may require justification. The effects of furnace and open confinement fire environments are compared with the regulatory fire environment, including the effects of gases resulting from decomposition of package overpack materials. The results indicate that furnace tests can produce the required radiation heat transfer environment, i.e., equivalent to the postulated pool fire. An open enclosure, with transparent (low emissivity) fire does not produce an equivalent radiation environment.

Smith, A

2008-12-31T23:59:59.000Z

339

The pict2e package Hubert Galein  

E-Print Network [OSTI]

an apologetic error message. The new package extends the existing LATEX picture environment, using the familiar with the Standard LATEX picture environment. Contents 1 Introduction 2 2 Usage 2 2.1 Package options . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.8 Medium-level operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3

Mintmire, John W.

340

Effect of carbonate soil on transport and dose estimates from long-lived radionuclides at U. S. Pacific Test Site  

SciTech Connect (OSTI)

The United States conducted a series of nuclear tests from 1946 to 1958 at Bikini, a coral atoll, in the Marshall Islands (MI). The aquatic and terrestrial environments of the atoll are still contaminated with several long-lived radionuclides that were generated during testing. The four major radionuclides found in terrestrial plants and soils are Cesium-137 ({sup 137} Cs), Strontium-90 ({sup 90} Sr), Plutonium-239+ 240 ({sup 239+240}Pu) and Americium-241 ({sup 241}Am). {sup 137}Cs in the coral soils is more available for uptake by plants than {sup 137}Cs associated with continental soils of North America or Europe. Soil-to-plant {sup 137}Cs median concentration ratios (CR) (kBq kg{sup {minus}1} dry weight plant/kBq kg {sup {minus}1} dry weight soil) for tropical fruits and vegetables range between 0.8 and 36, much larger than the range of 0.005 to 0.5 reported for vegetation in temperate zones. Conversely, {sup 90}Sr median CRs range from 0.006 to 1.0 at the atoll versus a range from 0.02 to 3.0 for continental silica-based soils. Thus, the relative uptake of {sup 137}Cs and {sup 90}Sr by plants in carbonate soils is reversed from that observed in silica-based soils. The CRs for {sup 239+240}Pu and {sup 241}Am are very similar to those observed in continental soils. Values range from 10{sup {minus}6} to 10{sup {minus}4} for both {sup 239+240}Pu and {sup 241}Am. No significant difference is observed between the two in coral soil. The uptake of {sup 137}Cs by plants is enhanced because of the absence of mineral binding sites and the low concentration of potassium in the coral soil. {sup 137}Cs is bound to the organic fraction of the soil, whereas {sup 90}Sr, {sup 239+240}Pu and {sup 241}Am are primarily bound to soil particles. Assessment of plant uptake for {sup 137}Cs and {sup 90}Sr into locally grown food crops was a major contributing factor in (1) reliably predicting the radiological dose for returning residents, and (2) developing a strategy to limit the availability and uptake of {sup 137}Cs into locally g

Conrado, C.L.; Hamilton, T.F.; Robison, W.L.; Stoker, A.C.

1998-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

Flammability Analysis For Actinide Oxides Packaged In 9975 Shipping Containers  

SciTech Connect (OSTI)

Packaging options are evaluated for compliance with safety requirements for shipment of mixed actinide oxides packaged in a 9975 Primary Containment Vessel (PCV). Radiolytic gas generation rates, PCV internal gas pressures, and shipping windows (times to reach unacceptable gas compositions or pressures after closure of the PCV) are calculated for shipment of a 9975 PCV containing a plastic bottle filled with plutonium and uranium oxides with a selected isotopic composition. G-values for radiolytic hydrogen generation from adsorbed moisture are estimated from the results of gas generation tests for plutonium oxide and uranium oxide doped with curium-244. The radiolytic generation of hydrogen from the plastic bottle is calculated using a geometric model for alpha particle deposition in the bottle wall. The temperature of the PCV during shipment is estimated from the results of finite element heat transfer analyses.

Laurinat, James E.; Askew, Neal M.; Hensel, Steve J.

2013-03-21T23:59:59.000Z

342

Practical Thermal Evaluation Methods For HAC Fire Analysis In Type B Radiaoactive Material (RAM) Packages  

SciTech Connect (OSTI)

Title 10 of the United States Code of Federal Regulations Part 71 for the Nuclear Regulatory Commission (10 CFR Part 71.73) requires that Type B radioactive material (RAM) packages satisfy certain Hypothetical Accident Conditions (HAC) thermal design requirements to ensure package safety during accidental fire conditions. Compliance with thermal design requirements can be met by prototype tests, analyses only or a combination of tests and analyses. Normally, it is impractical to meet all the HAC using tests only and the analytical methods are too complex due to the multi-physics non-linear nature of the fire event. Therefore, a combination of tests and thermal analyses methods using commercial heat transfer software are used to meet the necessary design requirements. The authors, along with his other colleagues at Savannah River National Laboratory in Aiken, SC, USA, have successfully used this 'tests and analyses' approach in the design and certification of several United States' DOE/NNSA certified packages, e.g. 9975, 9977, 9978, 9979, H1700, and Bulk Tritium Shipping Package (BTSP). This paper will describe these methods and it is hoped that the RAM Type B package designers and analysts can use them for their applications.

Abramczyk, Glenn; Hensel, Stephen J; Gupta, Narendra K.

2013-03-28T23:59:59.000Z

343

Space Charge Compensation in the Linac4 Low Energy Beam Transport Line with Negative Hydrogen Ions  

E-Print Network [OSTI]

The space charge effect of low energy, unbunched ion beams can be compensated by the trapping of ions or electrons into the beam potential. This has been studied for the 45 keV negative hydrogen ion beam in the CERN Linac4 Low Energy Beam Tranport (LEBT) using the package IBSimu1, which allows the space charge calculation of the particle trajectories. The results of the beam simulations will be compared to emittance measurements of an H- beam at the CERN Linac4 3 MeV test stand, where the injection of hydrogen gas directly into the beam transport region has been used to modify the space charge compensation degree.

Valerio-Lizarraga, C; Leon-Monzon, I; Lettry, J; Midttun, O; Scrivens, R

2013-01-01T23:59:59.000Z

344

Simplification of Diesel Emission Control System Packaging Using...  

Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

Simplification of Diesel Emission Control System Packaging Using SCR Coated on DPF Simplification of Diesel Emission Control System Packaging Using SCR Coated on DPF Study...

345

2014-08-28 Issuance: Energy Conservation Standards for Packaged...  

Broader source: Energy.gov (indexed) [DOE]

Conservation Standards for Packaged Terminal Air Conditioners and Packaged Terminal Heat Pumps; Notice of Proposed Rulemaking and Public Meeting 2014-08-28 Issuance: Energy...

346

High-Temperature, Air-Cooled Traction Drive Inverter Packaging...  

Broader source: Energy.gov (indexed) [DOE]

High-Temperature, Air-Cooled Traction Drive Inverter Packaging High-Temperature, Air-Cooled Traction Drive Inverter Packaging 2010 DOE Vehicle Technologies and Hydrogen Programs...

347

Combined Heat and Power (CHP) Integrated with Burners for Packaged...  

Broader source: Energy.gov (indexed) [DOE]

Combined Heat and Power (CHP) Integrated with Burners for Packaged Boilers Combined Heat and Power (CHP) Integrated with Burners for Packaged Boilers Providing Clean, Low-Cost,...

348

Advanced Ceramic Materials and Packaging Technologies for Realizing...  

Broader source: Energy.gov (indexed) [DOE]

Advanced Ceramic Materials and Packaging Technologies for Realizing Sensors for Concentrating Solar Power Systems Advanced Ceramic Materials and Packaging Technologies for...

349

CRAD, Packaging and Transfer of Hazardous Materials and Materials...  

Office of Environmental Management (EM)

CRAD, Packaging and Transfer of Hazardous Materials and Materials of National Security Interest Assessment Plan CRAD, Packaging and Transfer of Hazardous Materials and Materials of...

350

assay package insert: Topics by E-print Network  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

disposal. An important task for waste management organizations is to translate general waste acceptance requirements into detailed waste package specifications. Waste package...

351

Low-Cost Packaged CHP System with Reduced Emissions - Presentation...  

Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

Low-Cost Packaged CHP System with Reduced Emissions - Presentation by Cummins Power Generation, June 2011 Low-Cost Packaged CHP System with Reduced Emissions - Presentation by...

352

Energy Management Systems Package for Small Commercial Buildings...  

Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

Energy Management Systems Package for Small Commercial Buildings Energy Management Systems Package for Small Commercial Buildings Commercial Buildings Integration Project for the...

353

Advanced Framing Systems and Packages - Building America Top...  

Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

Advanced Framing Systems and Packages - Building America Top Innovation Advanced Framing Systems and Packages - Building America Top Innovation This photo shows advanced framing...

354

Pre-Packaged Commercial Property-Accessed Clean Energy Financing...  

Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

Packaged Commercial Property-Accessed Clean Energy Financing Solutions - 2014 BTO Peer Review Pre-Packaged Commercial Property-Accessed Clean Energy Financing Solutions - 2014 BTO...

355

Status and Trend of Automotive Power Packaging  

SciTech Connect (OSTI)

Comprehensive requirements in aspects of cost, reliability, efficiency, form factor, weight, and volume for power electronics modules in modern electric drive vehicles have driven the development of automotive power packaging technology intensively. Innovation in materials, interconnections, and processing techniques is leading to enormous improvements in power modules. In this paper, the technical development of and trends in power module packaging are evaluated by examining technical details with examples of industrial products. The issues and development directions for future automotive power module packaging are also discussed.

Liang, Zhenxian [ORNL

2012-01-01T23:59:59.000Z

356

Phase I Contaminant Transport Parameters for the Groundwater Flow and Contaminant Transport Model of Corrective Action Unit 99: Rainier Mesa/Shoshone Mountain, Nevada Test Site, Nye County, Nevada, Revision 1  

SciTech Connect (OSTI)

This document presents a summary and framework of available transport data and other information directly relevant to the development of the Rainier Mesa/Shoshone Mountain (RMSM) Corrective Action Unit (CAU) 99 groundwater transport model. Where appropriate, data and information documented elsewhere are briefly summarized with reference to the complete documentation.

Nathan Bryant

2008-05-01T23:59:59.000Z

357

Phase I Contaminant Transport Parameters for the Groundwater Flow and Contaminant Transport Model of Corrective Action Unit 97: Yucca Flat/Climax Mine, Nevada Test Site, Nye County, Nevada, Revision 0  

SciTech Connect (OSTI)

This report documents transport data and data analyses for Yucca Flat/Climax Mine CAU 97. The purpose of the data compilation and related analyses is to provide the primary reference to support parameterization of the Yucca Flat/Climax Mine CAU transport model. Specific task objectives were as follows: • Identify and compile currently available transport parameter data and supporting information that may be relevant to the Yucca Flat/Climax Mine CAU. • Assess the level of quality of the data and associated documentation. • Analyze the data to derive expected values and estimates of the associated uncertainty and variability. The scope of this document includes the compilation and assessment of data and information relevant to transport parameters for the Yucca Flat/Climax Mine CAU subsurface within the context of unclassified source-term contamination. Data types of interest include mineralogy, aqueous chemistry, matrix and effective porosity, dispersivity, matrix diffusion, matrix and fracture sorption, and colloid-facilitated transport parameters.

John McCord

2007-09-01T23:59:59.000Z

358

An overview of the Radioisotope Thermoelectric Generator Transportation System Program  

SciTech Connect (OSTI)

Radioisotope Thermoelectric Generators (RTG) convert the heat generated by radioactive decay to electricity using thermocouples. RTGs have a long operating life, are reasonably lightweight, and require little or no maintenance once assembled and tested. These factors make RTGs particularly attractive for use in spacecraft. However, because RTGs contain significant quantities of radioactive materials, normally plutonium-238 and its decay products, they must be transported in packages built in accordance with Title 10, Code of Federal Regulations, Part 71. The U.S. Department of Energy assigned the Radioisotope Thermoelectric Generator Transportation System (RTGTS) Program to Westinghouse Hanford Company in 1988 to develop a system meeting the regulatory requirements. The program objective was to develop a transportation system that would fully comply with 10 CFR 71 while protecting RTGs from adverse environmental conditions during normal conditions of transport (e.g., shock and heat). The RTGTS is scheduled for completion in December 1996 and will be available to support the National Aeronautics and Space Administration{close_quote}s Cassini mission to Saturn in October 1997. This paper provides an overview of the RTGTS and discusses the hardware being produced. Additionally, various program management innovations mandated by recent major changes in the U.S. Department of Energy structure and resources will be outlined. {copyright} {ital 1996 American Institute of Physics.}

McCoy, J.C.; Becker, D.L. [Westinghouse Hanford Company, P.O. Box 1970, Richland, Washington 99352 (United States)

1996-03-01T23:59:59.000Z

359

Water Rights Analysis Package (WRAP) Users Manual  

E-Print Network [OSTI]

The Water Rights Analysis Package (WRAP) is documented by a Reference Manual and this Users Manual. The Reference Manual explains WRAP capabilities and methodologies. This Users Manual provides the operational logistics for applying the model...

Wurbs, Ralph A.

360

Water Rights Analysis Package (WRAP) Reference Manual  

E-Print Network [OSTI]

The Texas Water Resources Institute (TWRI), and many other agencies and organizations, have worked with Ralph Wurbs over the years to develop WRAP (the Water Rights Analysis Package). The WRAP model simulates management of the water resources of a...

Wurbs, Ralph A.

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

Waste Package Component Design Methodology Report  

SciTech Connect (OSTI)

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

D.C. Mecham

2004-07-12T23:59:59.000Z

362

ADIPLS -- the Aarhus adiabatic oscillation package  

E-Print Network [OSTI]

Development of the Aarhus adiabatic pulsation code started around 1978. Although the main features have been stable for more than a decade, development of the code is continuing, concerning numerical properties and output. The code has been provided as a generally available package and has seen substantial use at a number of installations. Further development of the package, including bringing the documentation closer to being up to date, is planned as part of the HELAS Coordination Action.

J. Christensen-Dalsgaard

2007-10-16T23:59:59.000Z

363

Packaging material for thin film lithium batteries  

DOE Patents [OSTI]

A thin film battery including components which are capable of reacting upon exposure to air and water vapor incorporates a packaging system which provides a barrier against the penetration of air and water vapor. The packaging system includes a protective sheath overlying and coating the battery components and can be comprised of an overlayer including metal, ceramic, a ceramic-metal combination, a parylene-metal combination, a parylene-ceramic combination or a parylene-metal-ceramic combination.

Bates, John B. (116 Baltimore Dr., Oak Ridge, TN 37830); Dudney, Nancy J. (11634 S. Monticello Rd., Knoxville, TN 37922); Weatherspoon, Kim A. (223 Wadsworth Pl., Oak Ridge, TN 37830)

1996-01-01T23:59:59.000Z

364

PHASE II VAULT TESTING OF THE ARGONNE RFID SYSTEM  

SciTech Connect (OSTI)

The U.S. Department of Energy (DOE) (Environmental Management [EM], Office of Packaging and Transportation [EM-45]) Packaging and Certification Program (DOE PCP) has developed a Radio Frequency Identification (RFID) tracking and monitoring system, called ARG-US, for the management of nuclear materials packages during transportation and storage. The performance of the ARG-US RFID equipment and system has been fully tested in two demonstration projects in April 2008 and August 2009. With the strong support of DOE-SR and DOE PCP, a field testing program was completed in Savannah River Site's K-Area Material Storage (KAMS) Facility, an active Category I Plutonium Storage Facility, in 2010. As the next step (Phase II) of continued vault testing for the ARG-US system, the Savannah River Site K Area Material Storage facility has placed the ARG-US RFIDs into the 910B storage vault for operational testing. This latest version (Mark III) of the Argonne RFID system now has the capability to measure radiation dose and dose rate. This paper will report field testing progress of the ARG-US RFID equipment in KAMS, the operability and reliability trend results associated with the applications of the system, and discuss the potential benefits in enhancing safety, security and materials accountability. The purpose of this Phase II K Area test is to verify the accuracy of the radiation monitoring and proper functionality of the ARG-US RFID equipment and system under a realistic environment in the KAMS facility. Deploying the ARG-US RFID system leads to a reduced need for manned surveillance and increased inventory periods by providing real-time access to status and event history traceability, including environmental condition monitoring and radiation monitoring. The successful completion of the testing program will provide field data to support a future development and testing. This will increase Operation efficiency and cost effectiveness for vault operation. As the next step (Phase II) of continued vault testing for the ARG-US system, the Savannah River Site K Area Material Storage facility has placed the ARG-US RFIDs into the 910B storage vault. Deploying the ARG-US RFID system lends to a reduced need for manned surveillance and increased inventory periods by providing real-time access to status and event history traceability, including radiation and environmental monitoring. The successful completion of the testing program will provide field data to support future development and testing.

Willoner, T.; Turlington, R.; Koenig, R.

2012-06-25T23:59:59.000Z

365

Experimental tests of paleoclassical transport  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation ProposedUsing ZirconiaPolicy and Assistance100Jefferson LabAuxiliarydetermined fortests of

366

CHP Integrated with Burners for Packaged Boilers  

SciTech Connect (OSTI)

The objective of this project was to engineer, design, fabricate, and field demonstrate a Boiler Burner Energy System Technology (BBEST) that integrates a low-cost, clean burning, gas-fired simple-cycle (unrecuperated) 100 kWe (net) microturbine (SCMT) with a new ultra low-NOx gas-fired burner (ULNB) into one compact Combined Heat and Power (CHP) product that can be retrofit on new and existing industrial and commercial boilers in place of conventional burners. The Scope of Work for this project was segmented into two principal phases: (Phase I) Hardware development, assembly and pre-test and (Phase II) Field installation and demonstration testing. Phase I was divided into five technical tasks (Task 2 to 6). These tasks covered the engineering, design, fabrication, testing and optimization of each key component of the CHP system principally, ULNB, SCMT, assembly BBEST CHP package, and integrated controls. Phase I work culminated with the laboratory testing of the completed BBEST assembly prior to shipment for field installation and demonstration. Phase II consisted of two remaining technical tasks (Task 7 and 8), which focused on the installation, startup, and field verification tests at a pre-selected industrial plant to document performance and attainment of all project objectives. Technical direction and administration was under the management of CMCE, Inc. Altex Technologies Corporation lead the design, assembly and testing of the system. Field demonstration was supported by Leva Energy, the commercialization firm founded by executives at CMCE and Altex. Leva Energy has applied for patent protection on the BBEST process under the trade name of Power Burner and holds the license for the burner currently used in the product. The commercial term Power Burner is used throughout this report to refer to the BBEST technology proposed for this project. The project was co-funded by the California Energy Commission and the Southern California Gas Company (SCG), a division of Sempra Energy. These match funds were provided via concurrent contracts and investments available via CMCE, Altex, and Leva Energy The project attained all its objectives and is considered a success. CMCE secured the support of GI&E from Italy to supply 100 kW Turbec T-100 microturbines for the project. One was purchased by the project’s subcontractor, Altex, and a second spare was purchased by CMCE under this project. The microturbines were then modified to convert from their original recuperated design to a simple cycle configuration. Replacement low-NOx silo combustors were designed and bench tested in order to achieve compliance with the California Air Resources Board (CARB) 2007 emission limits for NOx and CO when in CHP operation. The converted microturbine was then mated with a low NOx burner provided by Altex via an integration section that allowed flow control and heat recovery to minimize combustion blower requirements; manage burner turndown; and recover waste heat. A new fully integrated control system was designed and developed that allowed one-touch system operation in all three available modes of operation: (1) CHP with both microturbine and burner firing for boiler heat input greater than 2 MMBtu/hr; (2) burner head only (BHO) when the microturbine is under service; and (3) microturbine only when boiler heat input requirements fall below 2 MMBtu/hr. This capability resulted in a burner turndown performance of nearly 10/1, a key advantage for this technology over conventional low NOx burners. Key components were then assembled into a cabinet with additional support systems for generator cooling and fuel supply. System checkout and performance tests were performed in the laboratory. The assembled system and its support equipment were then shipped and installed at a host facility where final performance tests were conducted following efforts to secure fabrication, air, and operating permits. The installed power burner is now in commercial operation and has achieved all the performance goals.

Castaldini, Carlo; Darby, Eric

2013-09-30T23:59:59.000Z

367

CERAMIC WASTE FORM DATA PACKAGE  

SciTech Connect (OSTI)

The purpose of this data package is to provide information about simulated crystalline waste forms that can be used to select an appropriate composition for a Cold Crucible Induction Melter (CCIM) proof of principle demonstration. Melt processing, viscosity, electrical conductivity, and thermal analysis information was collected to assess the ability of two potential candidate ceramic compositions to be processed in the Idaho National Laboratory (INL) CCIM and to guide processing parameters for the CCIM operation. Given uncertainties in the CCIM capabilities to reach certain temperatures throughout the system, one waste form designated 'Fe-MP' was designed towards enabling processing and another, designated 'CAF-5%TM-MP' was designed towards optimized microstructure. Melt processing studies confirmed both compositions could be poured from a crucible at 1600{degrees}C although the CAF-5%TM-MP composition froze before pouring was complete due to rapid crystallization (upon cooling). X-ray diffraction measurements confirmed the crystalline nature and phase assemblages of the compositions. The kinetics of melting and crystallization appeared to vary significantly between the compositions. Impedance spectroscopy results indicated the electrical conductivity is acceptable with respect to processing in the CCIM. The success of processing either ceramic composition will depend on the thermal profiles throughout the CCIM. In particular, the working temperature of the pour spout relative to the bulk melter which can approach 1700{degrees}C. The Fe-MP composition is recommended to demonstrate proof of principle for crystalline simulated waste forms considering the current configuration of INL's CCIM. If proposed modifications to the CCIM can maintain a nominal temperature of 1600{degrees}C throughout the melter, drain, and pour spout, then the CAF-5%TM-MP composition should be considered for a proof of principle demonstration.

Amoroso, J.; Marra, J.

2014-06-13T23:59:59.000Z

368

AUTHORIZING THE DOT SPECIFICATION 6M PACKAGING FOR CONTINUED USE AT THE SAVANNAH RIVER SITE  

SciTech Connect (OSTI)

The U.S. Department of Transportation (DOT) Specification 6M packaging was in extensive use for more than 40 years for in-commerce shipments of Type B quantities of fissile and radioactive material (RAM) across the USA, among the Department of Energy (DOE) laboratories, and between facilities in the DOE production complex. In January 2004, the DOT Research and Special Programs Administration (RSPA) Agency issued a final rule in the Federal Register to ammend requirements in the Hazardous Materials Regulations (HMR) pertaining to the transportation of radioactive materials. The final rule became effective on October 1, 2004. One of those changes discontinued the use of the DOT specification 6M, along with other DOT specification packagings, on October 1, 2008. A main driver for the change was due to the fact that 6M specification packagings were not supported by a Safety Analysis Report for Packagings (SARP) that was compliant with Title 10 of the Code of Federal Regulations (CFR) Part 71 (10 CFR 71). The regulatory rules for the discontinued use have been edited in Title 49 of the CFR Parts 100-185, 2004 edition and thereafter. Prior to October 1, 2008, the use of the 6M within the boundaries of the Savannah River Site (SRS), called an onsite transfer, was governed by an onsite transportation document that referenced 49 CFR Parts 100-185. SRS had to develop an Onsite Safety Assessment (OSA) which was independent of 49 CFR in order to justify the continued use of the DOT Specification 6M for the transfer of radioactive material (RAM) at the SRS after October 1, 2008. This paper will discuss the methodology for and difficulties associated with authorizing the DOT Specification 6M Packaging for continued use at the Savannah River Site.

Watkins, R.; Loftin, B.; Hoang, D.

2010-03-04T23:59:59.000Z

369

Protection of microelectronic devices during packaging  

DOE Patents [OSTI]

The present invention relates to a method of protecting a microelectronic device during device packaging, including the steps of applying a water-insoluble, protective coating to a sensitive area on the device; performing at least one packaging step; and then substantially removing the protective coating, preferably by dry plasma etching. The sensitive area can include a released MEMS element. The microelectronic device can be disposed on a wafer. The protective coating can be a vacuum vapor-deposited parylene polymer, silicon nitride, metal (e.g. aluminum or tungsten), a vapor deposited organic material, cynoacrylate, a carbon film, a self-assembled monolayered material, perfluoropolyether, hexamethyldisilazane, or perfluorodecanoic carboxylic acid, silicon dioxide, silicate glass, or combinations thereof. The present invention also relates to a method of packaging a microelectronic device, including: providing a microelectronic device having a sensitive area; applying a water-insoluble, protective coating to the sensitive area; providing a package; attaching the device to the package; electrically interconnecting the device to the package; and substantially removing the protective coating from the sensitive area.

Peterson, Kenneth A. (Albuquerque, NM); Conley, William R. (Tijeras, NM)

2002-01-01T23:59:59.000Z

370

Temperature-package power correlations for open-mode geologic disposal concepts.  

SciTech Connect (OSTI)

Logistical simulation of spent nuclear fuel (SNF) management in the U.S. combines storage, transportation and disposal elements to evaluate schedule, cost and other resources needed for all major operations leading to final geologic disposal. Geologic repository reference options are associated with limits on waste package thermal power output at emplacement, in order to meet limits on peak temperature for certain key engineered and natural barriers. These package power limits are used in logistical simulation software such as CALVIN, as threshold requirements that must be met by means of decay storage or SNF blending in waste packages, before emplacement in a repository. Geologic repository reference options include enclosed modes developed for crystalline rock, clay or shale, and salt. In addition, a further need has been addressed for open modes in which SNF can be emplaced in a repository, then ventilated for decades or longer to remove heat, prior to permanent repository closure. For each open mode disposal concept there are specified durations for surface decay storage (prior to emplacement), repository ventilation, and repository closure operations. This study simulates those steps for several timing cases, and for SNF with three fuel-burnup characteristics, to develop package power limits at which waste packages can be emplaced without exceeding specified temperature limits many years later after permanent closure. The results are presented in the form of correlations that span a range of package power and peak postclosure temperature, for each open-mode disposal concept, and for each timing case. Given a particular temperature limit value, the corresponding package power limit for each case can be selected for use in CALVIN and similar tools.

Hardin, Ernest L.

2013-02-01T23:59:59.000Z

371

Packages in the `graphics' bundle D. P. Carlisle  

E-Print Network [OSTI]

Packages in the `graphics' bundle D. P. Carlisle 1999/01/13 Contents 1 Introduction 2 2 Driver Problems . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 4 The Graphics packages 7 4.1 Package Graphics Files . . . . . . . . . . . . . . . . . . . . . . . . 9 4.5 Other commands in the graphics package

Bloch, Ethan

372

Packages in the `graphics' bundle D. P. Carlisle  

E-Print Network [OSTI]

Packages in the `graphics' bundle D. P. Carlisle 1996/10/29 Contents 1 Introduction 1 2 Driver Problems . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4 The Graphics packages 6 4.1 Package Graphics Files . . . . . . . . . . . . . . . . . . . . . . . . 8 4.5 Other commands in the graphics package

Duncan, James S.

373

White LED with High Package Extraction Efficiency  

SciTech Connect (OSTI)

The goal of this project is to develop a high efficiency phosphor converting (white) Light Emitting Diode (pcLED) 1-Watt package through an increase in package extraction efficiency. A transparent/translucent monolithic phosphor is proposed to replace the powdered phosphor to reduce the scattering caused by phosphor particles. Additionally, a multi-layer thin film selectively reflecting filter is proposed between blue LED die and phosphor layer to recover inward yellow emission. At the end of the project we expect to recycle approximately 50% of the unrecovered backward light in current package construction, and develop a pcLED device with 80 lm/W{sub e} using our technology improvements and commercially available chip/package source. The success of the project will benefit luminous efficacy of white LEDs by increasing package extraction efficiency. In most phosphor-converting white LEDs, the white color is obtained by combining a blue LED die (or chip) with a powdered phosphor layer. The phosphor partially absorbs the blue light from the LED die and converts it into a broad green-yellow emission. The mixture of the transmitted blue light and green-yellow light emerging gives white light. There are two major drawbacks for current pcLEDs in terms of package extraction efficiency. The first is light scattering caused by phosphor particles. When the blue photons from the chip strike the phosphor particles, some blue light will be scattered by phosphor particles. Converted yellow emission photons are also scattered. A portion of scattered light is in the backward direction toward the die. The amount of this backward light varies and depends in part on the particle size of phosphors. The other drawback is that yellow emission from phosphor powders is isotropic. Although some backward light can be recovered by the reflector in current LED packages, there is still a portion of backward light that will be absorbed inside the package and further converted to heat. Heat generated in the package may cause a deterioration of encapsulant materials, affecting the performance of both the LED die and phosphor, leading to a decrease in the luminous efficacy over lifetime. Recent studies from research groups at Rensselaer Polytechnic Institute found that, under the condition to obtain a white light, about 40% of the light is transmitted outward of the phosphor layer and 60% of the light is reflected inward.1,2 It is claimed that using scattered photon extraction (SPE) technique, luminous efficacy is increased by 60%. In this project, a transparent/translucent monolithic phosphor was used to replace the powdered phosphor layer. In the normal pcLED package, the powdered phosphor is mixed with silicone either to be deposited on the top of LED die forming a chip level conversion (CLC) white LED or to be casted in the package forming a volume conversion white LED. In the monolithic phosphors there are no phosphor powder/silicone interfaces so it can reduce the light scattering caused by phosphor particles. Additionally, a multi-layer thin film selectively reflecting filter is inserted in the white LED package between the blue LED die and phosphor layer. It will selectively transmit the blue light from the LED die and reflect the phosphor's yellow inward emission outward. The two technologies try to recover backward light to the outward direction in the pcLED package thereby improving the package extraction efficiency.

Yi Zheng; Matthew Stough

2008-09-30T23:59:59.000Z

374

Directory of certificiates of compliance for radioactive materials packages: Report of NRC approved packages. Revision 19, Volume 1  

SciTech Connect (OSTI)

This directory provides information on packagings approved by the U.S. Nuclear Regulatory Commission.

NONE

1996-10-01T23:59:59.000Z

375

Procurement of a fully licensed radioisotope thermoelectric generator transportation system  

SciTech Connect (OSTI)

A fully licensed transportation system for Radioisotope Thermoelectric Generators and Light-Weight Radioisotope Heater Units is currently being designed and built. The system will comply with all applicable U.S. Department of Transportation regulations without the use of a DOE Alternative.'' The U.S. Department of Transportation has special double containment'' requirements for plutonium. The system packaging uses a doubly contained bell jar'' concept. A refrigerated trailer is used for cooling the high-heat payloads. The same packaging is used for both high- and low-heat payloads. The system is scheduled to be available for use by mid-1992.

Adkins, H.E.; Bearden, T.E. (Westinghouse Hanford Company, P.O. Box 1970, N1-42, Richland, Washington 99352 (US))

1991-01-01T23:59:59.000Z

376

Procurement of a fully licensed radioisotope thermoelectric generator transportation system  

SciTech Connect (OSTI)

A fully licensed transportation system for Radioisotope Thermoelectric Generators and Light-Weight Radioisotope Heater Units is currently being designed and built. The system will comply with all applicable US Department of Transportation regulations without the use of a DOE Alternative.'' The US Department of Transportation has special double containment'' requirements for plutonium. The system packaging uses a doubly contained bell jar'' concept. A refrigerated trailer is used for cooling the high-heat payloads. The same packaging is used for both high- and low-heat payloads. The system is scheduled to be available for use by mid-1992. 4 refs., 4 figs., 2 tabs.

Adkins, H.E.; Bearden, T.E.

1990-10-01T23:59:59.000Z

377

Defense waste transportation: cost and logistics studies  

SciTech Connect (OSTI)

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.

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

1982-08-01T23:59:59.000Z

378

KJRR-FAI Hydraulic Flow Testing Input Package  

SciTech Connect (OSTI)

The INL, in cooperation with the KAERI via Cooperative Research And Development Agreement (CRADA), undertook an effort in the latter half of calendar year 2013 to produce a conceptual design for the KJRR-FAI campaign. The outcomes of this effort are documented in further detail elsewhere [5]. The KJRR-FAI was designed to be cooled by the ATR’s Primary Coolant System (PCS) with no provision for in-pile measurement or control of the hydraulic conditions in the irradiation assembly. The irradiation assembly was designed to achieve the target hydraulic conditions via engineered hydraulic losses in a throttling orifice at the outlet of the irradiation vehicle.

N.E. Woolstenhulme; R.B. Nielson; D.B. Chapman

2013-12-01T23:59:59.000Z

379

Value Engineering Study for Closing Waste Packages Containing TAD Canisters  

SciTech Connect (OSTI)

The Office of Civilian Radioactive Waste Management announced their intention to have the commercial utilities package spent nuclear fuel in shielded, transportable, ageable, and disposable containers prior to shipment to the Yucca Mountain repository. This will change the conditions used as a basis for the design of the waste package closure system. The environment is now expected to be a low radiation, low contamination area. A value engineering study was completed to evaluate possible modifications to the existing closure system using the revised requirements. Four alternatives were identified and evaluated against a set of weighted criteria. The alternatives are (1) a radiation-hardened, remote automated system (the current baseline design); (2) a nonradiation-hardened, remote automated system (with personnel intervention if necessary); (3) a nonradiation-hardened, semi-automated system with personnel access for routine manual operations; and (4) a nonradiation-hardened, fully manual system with full-time personnel access. Based on the study, the recommended design is Alternative 2, a nonradiation-hardened, remote automated system. It is less expensive and less complex than the current baseline system, because nonradiation-hardened equipment can be used and some contamination control equipment is no longer needed. In addition, the inclusion of remote automation ensures throughput requirements are met, provides a more reliable process, and provides greater protection for employees from industrial accidents and radiation exposure than the semi-automated or manual systems. Other items addressed during the value engineering study as requested by OCRWM include a comparison to industry canister closure systems and corresponding lessons learned; consideration of closing a transportable, ageable, and disposable canister; and an estimate of the time required to perform a demonstration of the recommended closure system.

Colleen Shelton-Davis

2005-11-01T23:59:59.000Z

380

Real-time monitoring during transportation of a radioisotope thermoelectric generator (RTG) using the radioisotope thermoelectric generator transportation system (RTGTS)  

SciTech Connect (OSTI)

The Radioisotopic Thermoelectric Generators (RTGs) that will be used to support the Cassini mission will be transported in the Radioisotope Thermoelectric Generator Transportation System (RTGTS). To ensure that the RTGs will not be affected during transportation, all parameters that could adversely affect RTG's performance must be monitored. The Instrumentation and Data Acquisition System (IDAS) for the RTGTS displays, monitors, and records all critical packaging and trailer system parameters. The IDAS also monitors the package temperature control system, RTG package shock and vibration data, and diesel fuel levels for the diesel fuel tanks. The IDAS alarms if any of these parameters reach an out-of-limit condition. This paper discusses the real-time monitoring during transportation of the Cassini RTGs using the RTGTS IDAS.

Pugh, Barry K. [EG and G Mound Applied Technologies P.O. Box 3000 Miamisburg, Ohio 45343-3000 (United States)

1997-01-10T23:59:59.000Z

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

Real-time monitoring during transportation of a radioisotope thermoelectric generator (RTG) using the radioisotope thermoelectric generator transportation system (RTGTS)  

SciTech Connect (OSTI)

The Radioisotopic Thermoelectric Generators (RTGs) that will be used to support the Cassini mission will be transported in the Radioisotope Thermoelectric Generator Transportation System (RTGTS). To ensure that the RTGs will not be affected during transportation, all parameters that could adversely affect RTG{close_quote}s performance must be monitored. The Instrumentation and Data Acquisition System (IDAS) for the RTGTS displays, monitors, and records all critical packaging and trailer system parameters. The IDAS also monitors the package temperature control system, RTG package shock and vibration data, and diesel fuel levels for the diesel fuel tanks. The IDAS alarms if any of these parameters reach an out-of-limit condition. This paper discusses the real-time monitoring during transportation of the Cassini RTGs using the RTGTS IDAS. {copyright} {ital 1997 American Institute of Physics.}

Pugh, B.K. [EGG Mound Applied Technologies P.O. Box 3000 Miamisburg, Ohio45343-3000 (United States)

1997-01-01T23:59:59.000Z

382

Transportation Services  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Transportation Services Transporting nuclear materials within the United States and throughout the world is a complicated and sometimes highly controversial effort requiring...

383

Local Transportation  

E-Print Network [OSTI]

Local Transportation. Transportation from the Airport to Hotel. There are two types of taxi companies that operate at the airport: special and regular taxis (

384

Greening Transportation  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Transportation Goal 2: Greening Transportation LANL supports and encourages employees to reduce their personal greenhouse gas emissions by offering various commuting and work...

385

Space charge compensation in the Linac4 low energy beam transport line with negative hydrogen ions  

SciTech Connect (OSTI)

The space charge effect of low energy, unbunched ion beams can be compensated by the trapping of ions or electrons into the beam potential. This has been studied for the 45 keV negative hydrogen ion beam in the CERN Linac4 Low Energy Beam Transport using the package IBSimu [T. Kalvas et al., Rev. Sci. Instrum. 81, 02B703 (2010)], which allows the space charge calculation of the particle trajectories. The results of the beam simulations will be compared to emittance measurements of an H{sup ?} beam at the CERN Linac4 3 MeV test stand, where the injection of hydrogen gas directly into the beam transport region has been used to modify the space charge compensation degree.

Valerio-Lizarraga, Cristhian A., E-mail: cristhian.alfonso.valerio.lizarraga@cern.ch [CERN, Geneva (Switzerland); Departamento de Investigación en Física, Universidad de Sonora, Hermosillo (Mexico); Lallement, Jean-Baptiste; Lettry, Jacques; Scrivens, Richard [CERN, Geneva (Switzerland)] [CERN, Geneva (Switzerland); Leon-Monzon, Ildefonso [Facultad de Ciencias Fisico-Matematicas, Universidad Autónoma de Sinaloa, Culiacan (Mexico)] [Facultad de Ciencias Fisico-Matematicas, Universidad Autónoma de Sinaloa, Culiacan (Mexico); Midttun, Øystein [CERN, Geneva (Switzerland) [CERN, Geneva (Switzerland); University of Oslo, Oslo (Norway)

2014-02-15T23:59:59.000Z

386

Chamber transport  

SciTech Connect (OSTI)

Heavy ion beam transport through the containment chamber plays a crucial role in all heavy ion fusion (HIF) scenarios. Here, several parameters are used to characterize the operating space for HIF beams; transport modes are assessed in relation to evolving target/accelerator requirements; results of recent relevant experiments and simulations of HIF transport are summarized; and relevant instabilities are reviewed. All transport options still exist, including (1) vacuum ballistic transport, (2) neutralized ballistic transport, and (3) channel-like transport. Presently, the European HIF program favors vacuum ballistic transport, while the US HIF program favors neutralized ballistic transport with channel-like transport as an alternate approach. Further transport research is needed to clearly guide selection of the most attractive, integrated HIF system.

OLSON,CRAIG L.

2000-05-17T23:59:59.000Z

387

DEPLOYMENT OF THE BULK TRITIUM SHIPPING PACKAGE  

SciTech Connect (OSTI)

A new Bulk Tritium Shipping Package (BTSP) was designed by the Savannah River National Laboratory to be a replacement for a package that has been used to ship tritium in a variety of content configurations and forms since the early 1970s. The BTSP was certified by the National Nuclear Safety Administration in 2011 for shipments of up to 150 grams of Tritium. Thirty packages were procured and are being delivered to various DOE sites for operational use. This paper summarizes the design features of the BTSP, as well as associated engineered material improvements. Fabrication challenges encountered during production are discussed as well as fielding requirements. Current approved tritium content forms (gas and tritium hydrides), are reviewed, as well as, a new content, tritium contaminated water on molecular sieves. Issues associated with gas generation will also be discussed.

Blanton, P.

2013-10-10T23:59:59.000Z

388

Quality of vacuum packaged lamb retail cuts  

E-Print Network [OSTI]

) provision of an ideal environment for the ag1ng of meat. Materials wh1ch prevent the rapid exchange of gases result in longer storage 11fe of meat than is obta1ned by packaging with more permeable materials (Kraft and Ayres, 1952). Jaye et al. (1962... storage for up to 35 days. Hanna et al. (1976) reported that these coryneform bacteria were species of ~Cba t i d a o hath i ~th rmos ha t m. B. ~th rmos hactu has been reported to be a major constituent of the bacterial flora of vacuum packaged...

Wanstedt, Kristen Gae

1982-01-01T23:59:59.000Z

389

Engineered waste-package-system design specification  

SciTech Connect (OSTI)

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.

Not Available

1983-05-01T23:59:59.000Z

390

MEMS packaging efforts at Sandia National Laboratories.  

SciTech Connect (OSTI)

Sandia National Laboratories has programs covering a broad range of MEMS technologies from LIGA to bulk to surface micromachining. These MEMS technologies are being considered for an equally broad range of applications, including sensors, actuators, optics, and microfluidics. As these technologies have moved from the research to the prototype product stage, packaging has been required to develop new capabilities to integrated MEMS and other technologies into functional microsystems. This paper discusses several of Sandia's MEMS packaging efforts, focusing mainly on inserting Sandia's SUMMIT V (5-level polysilicon) surface micromachining technology into fieldable microsystems.

Custer, Jonathan Sloane

2003-02-01T23:59:59.000Z

391

Packaging Materials and Design for Improved PV Module Reliability  

SciTech Connect (OSTI)

A number of candidate alternative encapsulant and soft backsheet materials have been evaluated in terms of their suitability for photovoltaic (PV) module packaging applications. Relevant properties, including peel strength as a function of damp heat exposure and permeability, have been measured. Based on these tests, promising new encapsulants with adhesion-promoting primers have been identified that result in improved properties. Test results for backsheets provided by industry and prepared at the National Renewable Energy Laboratory (NREL) have suggested strategies to achieve significantly improved products. The ability of glass/glass and glass/breathable backsheet constructions laminated with various encapsulant and/or edge seal materials to protect thin-film aluminum coatings deposited onto glass substrates was assessed. Glass/glass laminate constructions can trap harmful compounds that catalyze moisture-driven corrosion of the aluminum. Constructions with breathable backsheets allow higher rates of moisture ingress, but also allow egress of deleterious substances that can result in decreased corrosion.

Jorgensen, G.; Terwilliger, K.; Kempe, M.; Pern, J.; Glick, S.; del Cueto, J.; Kennedy, C.; McMahon, T.

2005-01-01T23:59:59.000Z

392

Surface Micromachine Microfluidics: Design, Fabrication, Packaging, and Characterization  

SciTech Connect (OSTI)

The field of microfluidics is undergoing rapid growth in terms of new device and system development. Among the many methods of fabricating microfluidic devices and systems, surface micromachining is relatively underrepresented due to difficulties in the introduction of fluids into the very small channels produced, packaging problems, and difficulties in device and system characterization. The potential advantages of using surface micromachining including compatibility with the existing integrated circuit tool set, integration of electronic sensing and actuation with microfluidics, and fluid volume minimization. In order to explore these potential advantages we have developed first generation surface micromachined microfluidic devices (channels) using an adapted pressure sensor fabrication process to produce silicon nitride channels, and the SUMMiT process to produce polysilicon channels. The channels were characterized by leak testing and flow rate vs. pressure measurements. The fabrication processes used and results of these tests are reported in this paper.

Galambos, Paul; Eaton, William P.; Shul, Randy; Willison, Christi Gober; Sniegowski, Jeffrey J.; Miller, Samuel L.; Guttierez, Daniel

1999-06-30T23:59:59.000Z

393

Transportation capabilities study of DOE-owned spent nuclear fuel  

SciTech Connect (OSTI)

This study evaluates current capabilities for transporting spent nuclear fuel owned by the US Department of Energy. Currently licensed irradiated fuel shipping packages that have the potential for shipping the spent nuclear fuel are identified and then matched against the various spent nuclear fuel types. Also included are the results of a limited investigation into other certified packages and new packages currently under development. This study is intended to support top-level planning for the disposition of the Department of Energy`s spent nuclear fuel inventory.

Clark, G.L.; Johnson, R.A.; Smith, R.W. [Packaging Technology, Inc., Tacoma, WA (United States); Abbott, D.G.; Tyacke, M.J. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

1994-10-01T23:59:59.000Z

394

Measuring total longshore sediment transport with a LISST instrumented mini-sled.  

E-Print Network [OSTI]

A surf zone sediment transport study was conducted in Jamaica Beach, Texas, using new oceanographic equipment. A mini-sled was constructed and outfitted with an instrument package that consisted of two velocimeters, one current profiler, three OBS...

Huchzermeyer, Erick Karl

2006-04-12T23:59:59.000Z

395

Transport mode and network architecture : carbon footprint as a new decision metric  

E-Print Network [OSTI]

This thesis examines the tradeoffs between carbon footprint, cost, time and risk across three case studies of United States' perishable or consumer packaged goods firms and their transportation partners. Building upon ...

Andrieu, Nelly

2008-01-01T23:59:59.000Z

396

CANE FIBERBOARD DEGRADATION WITHIN THE 9975 SHIPPING PACKAGE DURING LONG-TERM STORAGE APPLICATION  

SciTech Connect (OSTI)

The 9975 shipping package is used as part of the configuration for long-term storage of special nuclear materials in the K Area Complex at the Savannah River Site. The cane fiberboard overpack in the 9975 package provides thermal insulation, impact absorption and criticality control functions relevant to this application. The Savannah River National Laboratory has conducted physical, mechanical and thermal tests on aged fiberboard samples to identify degradation rates and support the development of aging models and service life predictions in a storage environment. This paper reviews the data generated to date, and preliminary models describing degradation rates of cane fiberboard in elevated temperature – elevated humidity environments.

Daugherty, W.; Dunn, K.; Hackney, B.

2013-06-19T23:59:59.000Z

397

Uranium hexafluoride packaging tiedown systems overview at Portsmouth Gaseous Diffusion Plant, Piketon, Ohio. Revision 1  

SciTech Connect (OSTI)

The Portsmouth Gaseous Diffusion Plant (PORTS) in Piketon, Ohio, is operated by Martin Marietta Energy Systems, Inc., through the US Department of Energy-Oak Ridge Operations Office (DOE-ORO) for the US Department of Energy-Headquarters, Office of Nuclear Energy. The PORTS conducts those operations that are necessary for the production, packaging, and shipment of uranium hexafluoride (UF{sub 6}). Uranium hexafluoride enriched uranium than 1.0 wt percent {sup 235}U shall be packaged in accordance with the US Department of Transportation (DOT) regulations of Title 49 CFR Parts 173 (Reference 1) and 178 (Reference 2), or in US Nuclear Regulatory Commission (NRC) or US Department of Energy (DOE) certified package designs. Concerns have been expressed regarding the various tiedown methods and condition of the trailers being used by some shippers/carriers for international transport of the UF{sub 6} cylinders/overpacks. Because of the concerns about international shipments, the US Department of Energy-Headquarters (DOE-HQ) Office of Nuclear Energy, through DOE-HQ Transportation Management Division, requested Westinghouse Hanford Company (Westinghouse Hanford) to review UF{sub 6} packaging tiedown and shipping practices used by PORTS, and where possible and appropriate, provide recommendations for enhancing these practices. Consequently, a team of two individuals from Westinghouse Hanford visited PORTS on March 5 and 6, 1990, for the purpose of conducting this review. The paper provides a brief discussion of the review activities and a summary of the resulting findings and recommendations. A detailed reporting of the is documented in Reference 4.

Becker, D.L.; Green, D.J.; Lindquist, M.R.

1993-07-01T23:59:59.000Z

398

Assessment of microelectronics packaging for high temperature, high reliability applications  

SciTech Connect (OSTI)

This report details characterization and development activities in electronic packaging for high temperature applications. This project was conducted through a Department of Energy sponsored Cooperative Research and Development Agreement between Sandia National Laboratories and General Motors. Even though the target application of this collaborative effort is an automotive electronic throttle control system which would be located in the engine compartment, results of this work are directly applicable to Sandia`s national security mission. The component count associated with the throttle control dictates the use of high density packaging not offered by conventional surface mount. An enabling packaging technology was selected and thermal models defined which characterized the thermal and mechanical response of the throttle control module. These models were used to optimize thick film multichip module design, characterize the thermal signatures of the electronic components inside the module, and to determine the temperature field and resulting thermal stresses under conditions that may be encountered during the operational life of the throttle control module. Because the need to use unpackaged devices limits the level of testing that can be performed either at the wafer level or as individual dice, an approach to assure a high level of reliability of the unpackaged components was formulated. Component assembly and interconnect technologies were also evaluated and characterized for high temperature applications. Electrical, mechanical and chemical characterizations of enabling die and component attach technologies were performed. Additionally, studies were conducted to assess the performance and reliability of gold and aluminum wire bonding to thick film conductor inks. Kinetic models were developed and validated to estimate wire bond reliability.

Uribe, F.

1997-04-01T23:59:59.000Z

399

(Parallel Linear Algebra Package) Jess Cmara Moreno  

E-Print Network [OSTI]

álgebra lineal (Linear Algebra Objects). También permite la utilización de vistas (objetos referenciadosPLAPACK (Parallel Linear Algebra Package) Jesús Cámara Moreno Programación Paralela y Computación Reducción de Vectores Inicialización de PLAPACK. Funciones. Templates. Funciones. Linear Algebra Objects

Giménez, Domingo

400

Seer: An analysis package for LHCO files  

E-Print Network [OSTI]

Seer is a multipurpose package for performing trigger, signal determination and cuts of an arbitrary number of collider processes stored in the LHCO file format. This article details the use of Seer, including the necessary details for users to customize the code for investigating new kinematic variables.

Martin, Travis A W

2015-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


401

USING A CONTAINMENT VESSEL LIFTING APPARATUS FOR REMOTE OPERATIONS OF SHIPPING PACKAGES  

SciTech Connect (OSTI)

The 9977 and the 9975 shipping packages are used in various nuclear facilities within the Department of Energy. These shipping packages are often loaded in designated areas with designs using overhead cranes or A-frames with lifting winches. However, there are cases where loading operations must be performed in remote locations where these facility infrastructures do not exist. For these locations, a lifting apparatus has been designed to lift the containment vessels partially out of the package for unloading operations to take place. Additionally, the apparatus allows for loading and closure of the containment vessel and subsequent pre-shipment testing. This paper will address the design of the apparatus and the challenges associated with the design, and it will describe the use of the apparatus.

Loftin, Bradley [Savannah River National Laboratory; Koenig, Richard [Savannah River National Laboratory

2013-08-08T23:59:59.000Z

402

Determination of Fire Enviroment in Stacked Cargo Containers with Radioactive Materials Packages  

SciTech Connect (OSTI)

Results from a Fire Test with a three-by-three stack of standard 6 m long International Standards Organization shipping containers containing combustible fuels and empty radioactive materials packages are reported and discussed. The stack is intended to simulate fire conditions that could occur during on-deck stowage on container cargo ships. The fire is initated by locating the container stack adjacent to a 9.8 x 6 m pool fire. Temperatures of both cargoes (empty and simulated radioactive materials packages) and containers are recorded and reported. Observations on the duration, intensity and spread of the fire are discussed. Based on the results, models for simulation of fire exposure of radioactive materials packages in such fires are suggested.

Arviso, M.; Bobbe, J.G.; Dukart, R.D.; Koski, J.A.

1999-05-01T23:59:59.000Z

403

44-BWR WASTE PACKAGE LOADING CURVE EVALUATION  

SciTech Connect (OSTI)

The objective of this calculation is to evaluate the required minimum burnup as a function of initial boiling water reactor (BWR) assembly enrichment that would permit loading of spent nuclear fuel into the 44 BWR waste package configuration as provided in Attachment IV. This calculation is an application of the methodology presented in ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003). The scope of this calculation covers a range of enrichments from 0 through 5.0 weight percent (wt%) U-235, and a burnup range of 0 through 40 GWd/MTU. This activity supports the validation of the use of burnup credit for commercial spent nuclear fuel applications. The intended use of these results will be in establishing BWR waste package configuration loading specifications. Limitations of this evaluation are as follows: (1) The results are based on burnup credit for actinides and selected fission products as proposed in YMP (2003, Table 3-1) and referred to as the ''Principal Isotopes''. Any change to the isotope listing will have a direct impact on the results of this report. (2) The results of 100 percent of the current BWR projected waste stream being able to be disposed of in the 44-BWR waste package with Ni-Gd Alloy absorber plates is contingent upon the referenced waste stream being sufficiently similar to the waste stream received for disposal. (3) The results are based on 1.5 wt% Gd in the Ni-Gd Alloy material and having no tuff inside the waste package. If the Gd loading is reduced or a process to introduce tuff inside the waste package is defined, then this report would need to be reevaluated based on the alternative materials.

J.M. Scaglione

2004-08-25T23:59:59.000Z

404

An assessment of the value of retail ready packaging  

E-Print Network [OSTI]

Use of retail-ready packaging reduces the costs of replenishing store shelves by eliminating the labor of removing packaging materials and stocking individual items on shelves. While reducing costs for retailers, retail-ready ...

Jackson, Kathleen Anne

2008-01-01T23:59:59.000Z

405

The C-Cat Wordnet Package: An Open Source Package for modifying andapplying Wordnet  

SciTech Connect (OSTI)

We present the C-Cat Wordnet package, an open source library for using and modifying Wordnet. The package includes four key features: an API for modifying Synsets; implementations of standard similarity metrics, implementations of well known Word Sense Disambiguation algorithms, and an implementation of the Castanet algorithm. The library is easily extendible and usable in many runtime environments. We demonstrate it's use on two standard Word Sense Disambiguation tasks and apply the Castanet algorithm to a corpus.

Stevens, K; Huang, T; Buttler, D

2011-09-16T23:59:59.000Z

406

Motor Packaging with Consideration of Electromagnetic and Material...  

Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

Documents & Publications Motor Packaging with Consideration of Electromagnetic and Material Characteristics Alnico and Ferrite Hybrid Excitation Electric Machines Wireless Charging...

407

Surety applications in transportation  

SciTech Connect (OSTI)

Infrastructure surety can make a valuable contribution to the transportation engineering industry. The lessons learned at Sandia National Laboratories in developing surety principles and technologies for the nuclear weapons complex and the nuclear power industry hold direct applications to the safety, security, and reliability of the critical infrastructure. This presentation introduces the concepts of infrastructure surety, including identification of the normal, abnormal, and malevolent threats to the transportation infrastructure. National problems are identified and examples of failures and successes in response to environmental loads and other structural and systemic vulnerabilities are presented. The infrastructure surety principles developed at Sandia National Laboratories are described. Currently available technologies including (a) three-dimensional computer-assisted drawing packages interactively combined with virtual reality systems, (b) the complex calculational and computational modeling and code-coupling capabilities associated with the new generation of supercomputers, and (c) risk-management methodologies with application to solving the national problems associated with threats to the critical transportation infrastructure are discussed.

Matalucci, R.V.; Miyoshi, D.S.

1998-01-01T23:59:59.000Z

408

Documentation pckage for the RFID temperature monitoring system (Of Model 9977 packages at NTS).  

SciTech Connect (OSTI)

The technical basis for extending the Model 9977 shipping package periodic maintenance beyond the one-year interval to a maximum of five years is based on the performance of the O-ring seals and the environmental conditions. The DOE Packaging Certification Program (PCP) has tasked Argonne National Laboratory to develop a Radio-Frequency Identification (RFID) temperature monitoring system for use by the facility personnel at DAF/NTS. The RFID temperature monitoring system, depicted in the figure below, consists of the Mk-1 RFId tags, a reader, and a control computer mounted on a mobile platform that can operate as a stand-alone system, or it can be connected to the local IT network. As part of the Conditions of Approval of the CoC, the user must complete the prescribed training to become qualified and be certified for operation of the RFID temperature monitoring system. The training course will be administered by Argonne National Laboratory on behalf of the Headquarters Certifying Official. This is a complete documentation package for the RFID temperature monitoring system of the Model 9977 packagings at NTS. The documentation package will be used for training and certification. The table of contents are: Acceptance Testing Procedure of MK-1 RFID Tags for DOE/EM Nuclear Materials Management Applications; Acceptance Testing Result of MK-1 RFID Tags for DOE/EM Nuclear Materials Management Applications; Performance Test of the Single Bolt Seal Sensor for the Model 9977 Packaging; Calibration of Built-in Thermistors in RFID Tags for Nevada Test Site; Results of Calibration of Built-in Thermistors in RFID Tags; Results of Thermal Calibration of Second Batch of MK-I RFID Tags; Procedure for Installing and Removing MK-1 RFID Tag on Model 9977 Drum; User Guide for RFID Reader and Software for Temperature Monitoring of Model 9977 Drums at NTS; Software Quality Assurance Plan (SQAP) for the ARG-US System; Quality Category for the RFID Temperature Monitoring System; The Documentation Package for the RFID Temperature Monitoring System; Software Test Plan and Results for ARG-US OnSite; Configuration Management Plan (CMP) for the ARG-US System; Requirements Management Plan for the ARG-US System; and Design Management Plan for ARG-US.

Chen, K.; Tsai, H.; Decision and Information Sciences

2009-02-20T23:59:59.000Z

409

Trends in packaging of high power semiconductor laser bars  

SciTech Connect (OSTI)

Several different approaches to packaging high power diode laser bars for pumping solid state lasers or for direct diode laser applications are examined. The benefit and utility of each package is strongly dependent upon the fundamental optoelectronic properties of the individual diode laser bars. Factors which influence these properties are outlined and comparisons of packaging approaches for these materials are made.

Solarz, R.W.; Emanuel, M.A.; Skidmore, J.A.; Freitas, B.L.; Krupke, W.F.

1997-07-01T23:59:59.000Z

410

Film Badge Application Radioactive Material Package Receipt Log  

E-Print Network [OSTI]

;RADIOACTIVE MATERIAL PACKAGE RECEIPT LOG DATE: DELIVERED BY: AUTHORIZED BY: Contamination Check DPM/100 cm2APPENDIX A Film Badge Application Radioactive Material Package Receipt Log Radioactive Material Package Receipt Form (Off-Campus Locations) Radiation / Contamination Survey Form #12;PERSONNEL MONITORING

Slatton, Clint

411

Packages in the `graphics' bundle D. P. Carlisle  

E-Print Network [OSTI]

Packages in the `graphics' bundle D. P. Carlisle 1999/01/13 Contents 1 Introduction 2 2 Driver . . . . . . . . . . . . . . . . . . . . . . . . . . 6 4 The Graphics packages 7 4.1 Package Options . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 4.4 Including Graphics Files . . . . . . . . . . . . . . . . . . . . . . . 9 4.5 Other commands

412

APPLICATION FO FLOW FORMING FOR USE IN RADIOACTIVE MATERIAL PACKAGING DESIGNS  

SciTech Connect (OSTI)

This paper reports on the development and testing performed to demonstrate the use of flow forming as an alternate method of manufacturing containment vessels for use in radioactive material shipping packaging designs. Additionally, ASME Boiler and Pressure Vessel Code, Section III, Subsection NB compliance along with the benefits compared to typical welding of containment vessels will be discussed. SRNL has completed fabrication development and the testing on flow formed containment vessels to demonstrate the use of flow forming as an alternate method of manufacturing a welded 6-inch diameter containment vessel currently used in the 9975 and 9977 radioactive material shipping packaging. Material testing and nondestructive evaluation of the flow formed parts demonstrate compliance to the minimum material requirements specified in applicable parts of ASME Boiler and Pressure Vessel Code, Section II. Destructive burst testing shows comparable results to that of a welded design. The benefits of flow forming as compared to typical welding of containment vessels are significant: dimensional control is improved due to no weld distortion; less final machining; weld fit-up issues associated with pipes and pipe caps are eliminated; post-weld non-destructive testing (i.e., radiography and die penetrant tests) is not necessary; and less fabrication steps are required. Results presented in this paper indicate some of the benefits in adapting flow forming to design of future radioactive material shipping packages containment vessels.

Blanton, P.; Eberl, K.; Abramczyk, G.

2012-07-11T23:59:59.000Z

413

Copper vapor laser modular packaging assembly  

DOE Patents [OSTI]

A modularized packaging arrangement for one or more copper vapor lasers and associated equipment is disclosed herein. This arrangement includes a single housing which contains the laser or lasers and all their associated equipment except power, water and neon, and means for bringing power, water, and neon which are necessary to the operation of the lasers into the container for use by the laser or lasers and their associated equipment.

Alger, Terry W. (Tracy, CA); Ault, Earl R. (Dublin, CA); Moses, Edward I. (Castro Valley, CA)

1992-01-01T23:59:59.000Z

414

Healthy transportation - healthy communities: developing objective measures of built-environment using GIS and testing significance of pedestrian variables on walking to transit  

E-Print Network [OSTI]

-Transportation-Air Quality (LUTRAQ) study for Portland, Oregon (1000 Friends of Oregon, 1997) looked at the impact of specific built- environment variables on travel. The built-environment variables such as ease of street crossings, sidewalk continuity, local street... twenty-one pedestrian indices (Allan, 2001; Bandara et al., 1994; Bradshaw, 1993; Dixon 1996; USDOT; Landis et al., 2001; City of Ft. Collins, 2002; Khisty, 1994; Moudon, 2001; Moudon et al., 2002; City of Portland, 1998; Wellar; Gallin, 2001; Portland...

Maghelal, Praveen Kumar

2009-05-15T23:59:59.000Z

415

Plasmids and packaging cell lines for use in phage display  

DOE Patents [OSTI]

The invention relates to a novel phagemid display system for packaging phagemid DNA into phagemid particles which completely avoids the use of helper phage. The system of the invention incorporates the use of bacterial packaging cell lines which have been transformed with helper plasmids containing all required phage proteins but not the packaging signals. The absence of packaging signals in these helper plasmids prevents their DNA from being packaged in the bacterial cell, which provides a number of significant advantages over the use of both standard and modified helper phage. Packaged phagemids expressing a protein or peptide of interest, in fusion with a phage coat protein such as g3p, are generated simply by transfecting phagemid into the packaging cell line.

Bradbury, Andrew M.

2012-07-24T23:59:59.000Z

416

Module Design, Materials, and Packaging Research Team: Activities and Capabilities  

SciTech Connect (OSTI)

Our team activities are directed at improving PV module reliability by incorporating new, more effective, and less expensive packaging materials and techniques. New and existing materials or designs are evaluated before and during accelerated environmental exposure for the following properties: (1) Adhesion and cohesion: peel strength and lap shear. (2) Electrical conductivity: surface, bulk, interface and transients. (3) Water vapor transmission: solubility and diffusivity. (4) Accelerated weathering: ultraviolet, temperature, and damp heat tests. (5) Module and cell failure diagnostics: infrared imaging, individual cell shunt characterization, coring. (6) Fabrication improvements: SiOxNy barrier coatings and enhanced wet adhesion. (7) Numerical modeling: Moisture ingress/egress, module and cell performance, and cell-to-frame leakage current. (8) Rheological properties of polymer encapsulant and sheeting materials. Specific examples will be described.

McMahon, T. J.; del Cueto, J.; Glick, S.; Jorgensen, G.; Kempe, M.; Kennedy, C.; Pern, J.; Terwilliger, K

2005-01-01T23:59:59.000Z

417

Module Packaging Research and Reliability: Activities and Capabilities  

SciTech Connect (OSTI)

Our team activities are directed at improving PV module reliability by incorporating new, more effective, and less expensive packaging materials and techniques. New and existing materials or designs are evaluated before and during accelerated environmental exposure for the following properties: (1) Adhesion and cohesion: peel strength and lap shear. (2) Electrical conductivity: surface, bulk, interface and transients. (3) Water vapor transmission: solubility and diffusivity. (4) Accelerated weathering: ultraviolet, temperature, and damp heat tests. (5) Module and cell failure diagnostics: infrared imaging, individual cell shunt characterization, coring. (6) Fabrication improvements: SiOxNy barrier coatings and enhanced wet adhesion. (7) Numerical modeling: Moisture ingress/egress, module and cell performance, and cell-to-frame leakage current. (8) Rheological properties of polymer encapsulant and sheeting materials. Specific examples are described.

McMahon, T. J.; delCueto, J.; Glick, S.; Jorgensen, G.; Kempe, M.; Pern, J.; Terwilliger, K.

2005-11-01T23:59:59.000Z

418

Geologic selection methodology for transportation corridor routing  

E-Print Network [OSTI]

A lack of planning techniques and processes on long, linear, cut and cover-tunneling route transportation systems has resulted because of the advancement of transportation systems into underground corridors. The proposed methodology is tested...

Shultz, Karin Wilson

2002-01-01T23:59:59.000Z

419

Scale-up considerations relevant to experimental studies of nuclear waste-package behavior  

SciTech Connect (OSTI)

Results from a study that investigated whether testing large-scale nuclear waste-package assemblages was technically warranted are reported. It was recognized that the majority of the investigations for predicting waste-package performance to date have relied primarily on laboratory-scale experimentation. However, methods for the successful extrapolation of the results from such experiments, both geometrically and over time, to actual repository conditions have not been well defined. Because a well-developed scaling technology exists in the chemical-engineering discipline, it was presupposed that much of this technology could be applicable to the prediction of waste-package performance. A review of existing literature documented numerous examples where a consideration of scaling technology was important. It was concluded that much of the existing scale-up technology is applicable to the prediction of waste-package performance for both size and time extrapolations and that conducting scale-up studies may be technically merited. However, the applicability for investigating the complex chemical interactions needs further development. It was recognized that the complexity of the system, and the long time periods involved, renders a completely theoretical approach to performance prediction almost hopeless. However, a theoretical and experimental study was defined for investigating heat and fluid flow. It was concluded that conducting scale-up modeling and experimentation for waste-package performance predictions is possible using existing technology. A sequential series of scaling studies, both theoretical and experimental, will be required to formulate size and time extrapolations of waste-package performance.

Coles, D.G.; Peters, R.D.

1986-04-01T23:59:59.000Z

420

Coupling and Testing the Fate and Transport of Heavy Metals and Other Ionic Species in a Groundwater Setting at Oak Ridge, TN - 13498  

SciTech Connect (OSTI)

Historical data show that heavy metals (including mercury) were released from Y -12 National Security Complex (NSC) at Oak Ridge, Tennessee, to the surrounding environments during its operation in 1950's. Studies have also shown that metals accumulated in the soil, rock, and groundwater, and are, at the present time, sources of contamination to nearby rivers and creeks (e.g., East Fork Poplar Creek, Bear Creek). For instance, mercury (Hg), zinc (Zn), cadmium (Cd) and lead (Pb) have been found and reported on the site groundwater. The groundwater type at the site is Ca-Mg-HCO{sub 3}. This paper presents a modeling application of PHREEQC, a model that simulates geochemical processes and couples them to flow and transport settings. The objective was to assess the capability of PHREEQC to simulate the transport of ionic species in groundwater at Oak Ridge, Tennessee; data were available from core holes and monitoring wells over a 736-m distance, within 60-300 m depths. First, predictions of the transport of major ionic species (i.e., Ca{sup 2+} and Mg{sup 2+}) in the water were made between monitoring wells and for GW-131. Second, the model was used to assess hypotheses under two scenarios of transport for Zn, Cd, Pb and Hg, in Ca-Mg-HCO{sub 3} water, as influenced by the following solid-liquid interactions: a) the role of ion exchange and b) the role of both ion exchange and sorption, the latter via surface complexation with Fe(OH){sub 3}. The transport scenario with ion exchange suggests that significant ion exchange is expected to occur for Zn, Cd and Pb concentrations, with no significant impact on Hg, within the first 100 m. Predictions match the expected values of the exchange coefficients relative to Ca{sup 2+} and Mg{sup 2+} (e.g., K{sub Ca/Zn} = K{sub Ca/Cd} > K{sub Ca/Pb} > K{sub Ca/Hg}). The scenario with both ion exchange and sorption does affect the concentrations of Zn and Cd to a small extent within the first 100 m, but does more meaningfully reduce the concentration of Pb, within the same distance, and also decreases the concentration of Hg in between core holes. Analysis of the above results, in the light of available literature on the ions of this study, does fundamentally support the capability of PHREEQC to predict the transport of major ions in a groundwater setting; it also generally supports the hypothesized role of ion exchange and sorption. The results indicate the potential of the model as a tool in the screening, selection and monitoring of remediation technologies for contaminated groundwater sites. (authors)

Noosai, Nantaporn; Fuentes, Hector R. [CEE Florida International University, Miami, FL 33174 (United States)] [CEE Florida International University, Miami, FL 33174 (United States)

2013-07-01T23:59:59.000Z

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