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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

Thermal testing of packages for transport of radioactive wastes  

SciTech Connect

Shipping containers for radioactive materials must be shown capable of surviving tests specified by regulations such as Title 10, Code of Federal Regulations, Part 71 (called 10CFR71 in this paper) within the United States. Equivalent regulations hold for other countries such as Safety Series 6 issued by the International Atomic Energy Agency. The containers must be shown to be capable of surviving, in order, drop tests, puncture tests, and thermal tests. Immersion testing in water is also required, but must be demonstrated for undamaged packages. The thermal test is intended to simulate a 30 minute exposure to a fully engulfing pool fire that could occur if a transport accident involved the spill of large quantities of hydrocarbon fuels. Various qualification methods ranging from pure analysis to actual pool fire tests have been used to prove regulatory compliance. The purpose of this paper is to consider the alternatives for thermal testing, point out the strengths and weaknesses of each approach, and to provide the designer with the information necessary to make informed decisions on the proper test program for the particular shipping container under consideration. While thermal analysis is an alternative to physical testing, actual testing is often emphasized by regulators, and this report concentrates on these testing alternatives.

Koski, J.A.

1994-12-31T23:59:59.000Z

2

Materials Transportation Testing & Analysis at Sandia National...  

NLE Websites -- All DOE Office Websites (Extended Search)

Transportation Testing & Analysis Mission Sandia's Transportation Risk & Packaging Program develops innovative technologies and methodologies to solve transportation and packaging...

3

TRU waste transportation package development  

SciTech Connect

Inventories of the transuranic wastes buried or stored at various US DOE sites are tabulated. The leading conceptual design of Type-B packaging for contact-handled transuranic waste is the Transuranic Package Transporter (TRUPACT), a large metal container comprising inner and outer tubular steel frameworks which are separated by rigid polyurethane foam and sheathed with steel plate. Testing of TRUPACT is reported. The schedule for its development is given. 6 figures. (DLC)

Eakes, R. G.; Lamoreaux, G. H.; Romesberg, L. E.; Sutherland, S. H.; Duffey, T. A.

1980-01-01T23:59:59.000Z

4

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

SciTech Connect

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

5

Packaging and Transportation | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Packaging and Transportation Packaging and Transportation Packaging and Transportation Packaging and Transportation Radiological shipments are accomplished safely. Annually, about 400 million hazardous materials shipments occur in the United States by rail, air, sea, and land. Of these shipments, about three million are radiological shipments. Since Fiscal Year (FY) 2004, EM has completed over 150,000 shipments of radioactive material/waste. Please click here to see Office of Packaging and Transportation Fiscal Year 2012 Annual Report. SUPPORTING PROGRAMS SAFE TRANSPORTATION OF RADIOLOGICAL SHIPMENTS Transportation Emergency Preparedness Program (TEPP) TEPP provides the tools for planning, training and exercises, and technical assistance to assist State and Tribal authorities in preparing for response

6

Regulatory fire test requirements for plutonium air transport packages : JP-4 or JP-5 vs. JP-8 aviation fuel.  

Science Conference Proceedings (OSTI)

For certification, packages used for the transportation of plutonium by air must survive the hypothetical thermal environment specified in 10CFR71.74(a)(5). This regulation specifies that 'the package must be exposed to luminous flames from a pool fire of JP-4 or JP-5 aviation fuel for a period of at least 60 minutes.' This regulation was developed when jet propellant (JP) 4 and 5 were the standard jet fuels. However, JP-4 and JP-5 currently are of limited availability in the United States of America. JP-4 is very hard to obtain as it is not used much anymore. JP-5 may be easier to get than JP-4, but only through a military supplier. The purpose of this paper is to illustrate that readily-available JP-8 fuel is a possible substitute for the aforementioned certification test. Comparisons between the properties of the three fuels are given. Results from computer simulations that compared large JP-4 to JP-8 pool fires using Sandia's VULCAN fire model are shown and discussed. Additionally, the Container Analysis Fire (CAFE) code was used to compare the thermal response of a large calorimeter exposed to engulfing fires fueled by these three jet propellants. The paper then recommends JP-8 as an alternate fuel that complies with the thermal environment implied in 10CFR71.74.

Figueroa, Victor G.; Lopez, Carlos; Nicolette, Vernon F.

2010-10-01T23:59:59.000Z

7

Transportation and packaging resource guide  

Science Conference Proceedings (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

8

DOE-Idaho's Packaging and Transportation Perspective  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Idaho's Packaging and Idaho's Packaging and T t ti P ti Transportation Perspective Richard Provencher Manager DOE Idaho Operations Office DOE Idaho Operations Office Presented to the DOE National Transportation Stakeholders Forum Stakeholders Forum May 12, 2011 DOE's Idaho site ships and receives a wide variety of radioactive materials 2 Engineering Test Reactor vessel excavated, transported across the site and disposed 3 Navy SNF moved from wet to dry storage storage 4 5 Left: Contact-handled TRU shipments Right: A remote-handled TRU shipment Right: A remote handled TRU shipment 6 NAC spent nuclear fuel container on its trailer, prior to installation of its impact limiters 7 Examples of dry (CPP-603) and wet (CPP-666) storage in Idaho (CPP 666) storage in Idaho 8 INL's Materials and Fuels Complex Hot Fuel Examination

9

Development of an Air Transport Type A Fissile Package  

SciTech Connect

This paper presents the summary of testing by the Savannah River National Laboratory (SRNL) to support development of a light weight (<140 lbs) air transport qualified Type A Fissile Packaging. The package design incorporates features and materials specifically designed to minimize packaging weight. The light weight package is being designed to provide confinement to the contents when subjected to the normal and hypothetical conditions required of an air transportable Type A Fissile radioactive material shipping package. The objective of these tests was to provide design input to the final design for the LORX Type A Fissile Air Transport Packaging when subjected to the performance requirements of the drop, crush and puncture probe test of 10CFR71. The post test evaluation of the prototype packages indicates that all of the tested designs would satisfactorily confine the content within the packaging. The differences in the performance of the prototypes varied significantly depending on the core materials and their relative densities. Information gathered from these tests is being used to develop the final design for the Department of Homeland Security.

Blanton, P.; Ebert, K.

2011-07-13T23:59:59.000Z

10

Underground Test Area Subproject Phase I Data Analysis Task. Volume V - Transport Parameter and Source Term Data Documentation Package  

Science Conference Proceedings (OSTI)

Volume V 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 transport parameter and source term data. 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

11

Effects of mixed waste simulants on transportation packaging plastic components  

Science Conference Proceedings (OSTI)

The purpose of hazardous and radioactive materials packaging is to, enable these materials to be transported without posing a threat to the health or property of the general public. To achieve this aim, regulations have been written establishing general design requirements for such packagings. While no regulations have been written specifically for mixed waste packaging, regulations for the constituents of mixed wastes, i.e., hazardous and radioactive substances, have been codified. The design requirements for both hazardous and radioactive materials packaging specify packaging compatibility, i.e., that the materials of the packaging and any contents be chemically compatible with each other. Furthermore, Type A and Type B packaging design requirements stipulate that there be no significant chemical, galvanic, or other reaction between the materials and contents of the package. Based on these requirements, a Chemical Compatibility Testing Program was developed in the Transportation Systems Department at Sandia National Laboratories (SNL). The program, supported by the US Department of Energy`s (DOE) Transportation Management Division, EM-261 provides the means to assure any regulatory body that the issue of packaging material compatibility towards hazardous and radioactive materials has been addressed. In this paper, we describe the general elements of the testing program and the experimental results of the screening tests. The implications of the results of this testing are discussed in the general context of packaging development. Additionally, we present the results of the first phase of this experimental program. This phase involved the screening of five candidate liner and six seal materials against four simulant mixed wastes.

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

1994-12-31T23:59:59.000Z

12

Drop Tests for the 6M Specification Package Closure Investigation  

SciTech Connect

Results of tests of drum-type RAM packages employing conventional clamp-ring closures have caused concern within the DOE Complex over the Department of Transportation 6M Specification Package. To clarify these issues, the Savannah River Site's Radioactive Material Packaging Technology Group was commissioned to conduct a series of tests to determine the response of the clamp-ring closure to the regulatory Hypothetical Accident Condition drop tests, for packages at maximum allowable weight, 640 lb. Additionally, three enhanced closure designs were also tested: the Clamshell, plywood disk reinforcement, and J-Clip. The results of the tests showed that the standard closure was unable to retain its lid for both Center-of-Gravity-Over-Corner and Shallow-Angle cases, for the standard package, at its maximum allowed weight. Similar results were found for packages dropped from a reduced height. The Clamshell design provided the best performance of the enhanced closures.

Smith, A.C.

2003-10-02T23:59:59.000Z

13

Spent Fuel Transportation Package Performance Study - Experimental Design Challenges  

Science Conference Proceedings (OSTI)

Numerous studies of spent nuclear fuel transportation accident risks have been performed since the late seventies that considered shipping container design and performance. Based in part on these studies, NRC has concluded that the level of protection provided by spent nuclear fuel transportation package designs under accident conditions is adequate. [1] Furthermore, actual spent nuclear fuel transport experience showcase a safety record that is exceptional and unparalleled when compared to other hazardous materials transportation shipments. There has never been a known or suspected release of the radioactive contents from an NRC-certified spent nuclear fuel cask as a result of a transportation accident. In 1999 the United States Nuclear Regulatory Commission (NRC) initiated a study, the Package Performance Study, to demonstrate the performance of spent fuel and spent fuel packages during severe transportation accidents. NRC is not studying or testing its current regulations, a s the rigorous regulatory accident conditions specified in 10 CFR Part 71 are adequate to ensure safe packaging and use. As part of this study, NRC currently plans on using detailed modeling followed by experimental testing to increase public confidence in the safety of spent nuclear fuel shipments. One of the aspects of this confirmatory research study is the commitment to solicit and consider public comment during the scoping phase and experimental design planning phase of this research.

Snyder, A. M.; Murphy, A. J.; Sprung, J. L.; Ammerman, D. J.; Lopez, C.

2003-02-25T23:59:59.000Z

14

Package Safety Analysis Assessment for Sludge Transportation System  

SciTech Connect

This package safety analysis assessment demonstrates that the Sludge Transportation System meets the acceptance criteria for an equivalent package as specified in DOE/RL-2001-36, Hanford Sitewide Transportation Safety Document for onsite shipment.

ROMANO, T.

2003-03-19T23:59:59.000Z

15

DOE-Idaho's Packaging and Transportation Perspective | Department...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

and Transportation Perspective Presented by Richard Provencher, Manager for the DOE Idaho Operations Office. DOE-Idaho's Packaging and Transportation Perspective More Documents...

16

Testing of the CANDU Spent Fuel Storage Basket Package  

SciTech Connect

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

17

THE USE OF DIGITAL RADIOGRAPHY IN THE EVALUATION OF RADIOACTIVE MATERIALS PACKAGING PERFORMANCE TESTING  

Science Conference Proceedings (OSTI)

New designs of radioactive material shipping packages are required to be evaluated in accordance with 10 CFR Part 71, ''Packaging and Transportation of Radioactive Material''. This paper will discuss the use of digital radiography to evaluate the effects of the tests required by 10 CFR 71.71, Normal Conditions of Transport (NCT), and 10 CFR 71.73, Hypothetical Accident Conditions (HAC). One acceptable means of evaluating packaging performance is to subject packagings to the series of NCT and HAC tests. The evaluation includes a determination of the effect on the packaging by the conditions and tests. That determination has required that packagings be cut and sectioned to learn the actual effects on internal components. Digital radiography permits the examination of internal packaging components without sectioning a package. This allows a single package to be subjected to a series of tests. After each test, the package is digitally radiographed and the effects of particular tests evaluated. Radiography reduces the number of packages required for testing and also reduces labor and materials required to section and evaluate numerous packages. This paper will include a description of the digital radiography equipment used in the testing and evaluation of the 9977 and 9978 packages at SRNL. The equipment is capable of making a single radiograph of a full-sized package in one exposure. Radiographs will be compared to sectioned packages that show actual conditions compared to radiographic images.

May, C; Lawrence Gelder, L; Boyd Howard, B

2007-03-22T23:59:59.000Z

18

Initial waste package interaction tests: status report  

SciTech Connect

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

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

1980-12-01T23:59:59.000Z

19

THERMAL PERFORMANCE OF RADIOACTIVE MATERIAL PACKAGES IN TRANSPORT CONFIGURATION  

SciTech Connect

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 Part 71. The packages are transported in specially designed vehicles like Safe Secure Transport (SST) for safety and security. In the transport vehicles, the packages are placed close to each other to maximize the number of units in the vehicle. Since the RAM contents in the packagings produce decay heat, it is important that they are spaced sufficiently apart to prevent overheating of the containment vessel (CV) seals and the impact limiter to ensure the structural integrity of the package. This paper presents a simple methodology to assess thermal performance of a typical 9975 packaging in a transport configuration.

Gupta, N.

2010-03-04T23:59:59.000Z

20

DOE O 460.1C, Packaging and Transportation Safety  

Directives, Delegations, and Requirements

The order establishes safety requirements for the proper packaging and transportation of DOE, including NNSA, offsite shipments and onsite transfers of ...

2010-05-14T23:59:59.000Z

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

Type B plutonium transport package development that uses metallic filaments and composite materials  

Science Conference Proceedings (OSTI)

A new package was developed for transporting Pu and U quantities that are currently carried in DOT-6M packages. It uses double containment with threaded closures and elastomeric seals. A composite overpack of metallic wire mesh and ceramic or quartz cloth insulation is provided for protection in accidents. Two prototypes were subjected to dynamic crush tests. A thermal computer model was developed and benchmarked by test results to predict package behavior in fires. The material performed isotropically in a global fashion. A Type B Pu transport package can be developed for DOE Pu shipments for less than $5000 if manufactured in quantity. 5 figs, 6 refs. (DLC)

Pierce, J.D.; Moya, J.L.; McClure, J.D.; Hohnstreiter, G.F. (Sandia National Labs., Albuquerque, NM (United States)); Golliher, K.G. (USDOE Albuquerque Operations Office, NM (United States))

1991-01-01T23:59:59.000Z

22

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

O 460.1C, PACKAGING AND TRANSPORTATION SAFETY, 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 TRANSPORTATION SAFETY, DOE O 460.2A, DEPARTMENTAL MATERIALS TRANSPORTATION AND PACKAGING MANAGEMENT "The familiar level of this module is divided into two sections. The objectives and requirements of DOE O 460.1C and DOE O 460.2A will be discussed in the first and second sections, respectively. Several examples and practices throughout the module are provided to help familiarize you with the material. The practices will also help prepare you for the criterion test. Before continuing, you should obtain a copy of the Orders and implementation guides and manuals for this module. Copies of the Orders are available on the internet

23

Office of Packaging and Transportation Fiscal Year 2012 Annual Report |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Packaging and Transportation Fiscal Year 2012 Annual Packaging and Transportation Fiscal Year 2012 Annual Report Office of Packaging and Transportation Fiscal Year 2012 Annual Report The Office of Environmental Management (EM) was established to mitigate the risks and hazards posed by the legacy of nuclear weapons production and research. The most ambitious and far ranging of these missions is dealing with the environmental legacy of the Cold War. Many problems posed by its operations are unique, and include the transportation of unprecedented amounts of contaminated waste, water, and soil, and a vast number of contaminated structures during remediation of the contaminated sites. Since Fiscal Year (FY) 2004, EM has completed over 150,000 shipments of radioactive material and waste. The mission of the Department of Energy (DOE) Office of Packaging and

24

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

Science Conference Proceedings (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

25

Packaging and Transportation Support at LANL CTMA 2012  

SciTech Connect

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

26

Packaging and Transportation Support at LANL CTMA 2012  

SciTech Connect

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

27

ISSUES ASSOCIATED WITH SAFE PACKAGING AND TRANSPORT OF NANOPARTICLES  

SciTech Connect

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

28

Leveraging Available Data to Support Extension of Transportation Packages Service Life  

SciTech Connect

Data obtained from testing shipping package materials have been leveraged to support extending the service life of select shipping packages while in nuclear materials transportation. Increasingly, nuclear material inventories are being transferred to an interim storage location where they will reside for extended periods of time. Use of a shipping package to store nuclear materials in an interim storage location has become more attractive for a variety of reasons. Shipping packages are robust and have a qualified pedigree for their performance in normal operation and accident conditions within the approved shipment period and storing nuclear material within a shipping package results in reduced operations for the storage facility. However, the shipping package materials of construction must maintain a level of integrity as specified by the safety basis of the storage facility through the duration of the storage period, which is typically well beyond the one year transportation window. Test programs have been established to obtain aging data on materials of construction that are the most sensitive/susceptible to aging in certain shipping package designs. The collective data are being used to support extending the service life of shipping packages in both transportation and storage.

Dunn, K.; Abramczyk, G.; Bellamy, S.; Daugherty, W.; Hackney, B.; Hoffman, E.; Skidmore, E.; Stefek, T.

2012-06-12T23:59:59.000Z

29

Test concept for waste package environment tests at Yucca Mountain  

SciTech Connect

The Nevada Nuclear Waste Storage Investigations Project is characterizing a tuffaceous rock unit at Yucca Mountain, Nevada to evaluate its suitability for a repository for high level radioactive waste. The candidate repository horizon is a welded, devitrified tuff bed located at a depth of about 300 m in the unsaturated zone, over 100 m above the water table. As part of the project, Lawrence Livermore National Laboratory is responsible for designing the waste packages and for assessing their expected performance in the repository environment. The primary region of interest to package design and performance assessment is the portion of the rock mass within a few meters of waste emplacement holes. Hydrologic mechanisms active in this unsaturated near-field environment, along with thermal and mechanical phenomena that influence the hydrology, need to be understood well enough to confirm the basis of the waste package designs and performance assessment. Large scale in situ tests (called waste package environment tests) are being planned in order to develop this understanding and to provide data sets for performance assessment model validation (Yow, 1985). Exploratory shafts and limited underground facilities for in-situ testing will be constructed at Yucca Mountain during site characterization. Multiple waste package environment tests are being planned for these facilities to represent horizontal and vertical waste emplacement configurations in the repository target horizon. These approximately half-scale tests are being designed to investigate rock mass hydrologic conditions during a cycle of thermal loading.

Yow, J.L. Jr.

1987-06-01T23:59:59.000Z

30

DEVELOPMENT OF THE HS99 AIR TRANSPORT TYPE A FISSILE PACKAGE  

SciTech Connect

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

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

Science Conference Proceedings (OSTI)

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

Nigrey, P.J.

1995-12-01T23:59:59.000Z

32

Packaging and Transportation of Additional Neptunium Oxide  

Science Conference Proceedings (OSTI)

The Savannah River Site's HB-Line Facility completed a second neptunium oxide production campaign in which nine (9) additional cans of neptunium oxide were produced and shipped to the Idaho National Laboratory and Oak Ridge National Laboratory in the 9975 shipping container. These additional cans were from a different feed solution than the first fifty (50) cans of neptunium oxide that were previously produced and shipped via a Letter of Amendment to the 9975 Safety Analysis Report for Packaging (SARP) content table. This paper will address the challenges associated with demonstrating the neptunium oxide produced from the additional feed solution was equivalent to the original neptunium oxide and within the content description of the Letter of Amendment.

Watkins, R.; Jordan, J.; Hensel, S.

2010-05-05T23:59:59.000Z

33

Undulator Transportation Test Results  

SciTech Connect

A test was performed to determine whether transporting and handling the undulators makes any changes to their properties. This note documents the test. No significant changes to the test undulator were observed. After the LCLS undulators are tuned and fiducialized in the Magnetic Measurement Facility (MMF), they must be transported to storage buildings and transported to the tunnel. It has been established that the undulators are sensitive to temperature. We wish to know whether the undulators are also sensitive to the vibrations and shocks of transportation. To study this issue, we performed a test in which an undulator was measured in the MMF, transported to the tunnel, brought back to the MMF, and re-measured. This note documents the test and the results.

Wolf, Zachary

2010-11-17T23:59:59.000Z

34

BALLISTICS TESTING OF THE 9977 SHIPPING PACKAGE FOR STORAGE APPLICATIONS  

SciTech Connect

Radioactive materials are stored in a variety of locations throughout the DOE complex. At the Savannah River Site (SRS), materials are stored within dedicated facilities. Each of those facilities has a documented safety analysis (DSA) that describes accidents that the facility and the materials within it may encounter. Facilities at the SRS are planning on utilizing the certified Model 9977 Shipping Package as a long term storage package and one of these facilities required ballistics testing. Specifically, in order to meet the facility DSA, the radioactive materials (RAM) must be contained within the storage package after impact by a .223 caliber round. In order to qualify the Model 9977 Shipping Package for storage in this location, the package had to be tested under these conditions. Over the past two years, the Model 9977 Shipping Package has been subjected to a series of ballistics tests. The purpose of the testing was to determine if the 9977 would be suitable for use as a storage package at a Savannah River Site facility. The facility requirements are that the package must not release any of its contents following the impact in its most vulnerable location by a .223 caliber round. A package, assembled to meet all of the design requirements for a certified 9977 shipping configuration and using simulated contents, was tested at the Savannah River Site in March of 2011. The testing was completed and the package was examined. The results of the testing and examination are presented in this paper.

Loftin, B.; Abramczyk, G.; Koenig, R.

2012-06-06T23:59:59.000Z

35

Materials Transportation Testing & Analysis at Sandia National...  

NLE Websites -- All DOE Office Websites (Extended Search)

"How SAFE are radioactive material transportation packages?" RADCAT 2.0 Now Available RADCATRADTRAN Brochure pdf 237kb...

36

Air transport of plutonium metal : content expansion initiative for the Plutonium Air Transportable (PAT-1) packaging.  

SciTech Connect

The National Nuclear Security Administration (NNSA) has submitted an application to the Nuclear Regulatory Commission (NRC) for the air shipment of plutonium metal within the Plutonium Air Transportable (PAT-1) packaging. The PAT-1 packaging is currently authorized for the air transport of plutonium oxide in solid form only. The INMM presentation will provide a limited overview of the scope of the plutonium metal initiative and provide a status of the NNSA application to the NRC.

Mann, Paul T. (National Nuclear Security Administration); Caviness, Michael L. (Los Alamos National Laboratory); Yoshimura, Richard Hiroyuki

2010-06-01T23:59:59.000Z

37

Air transport of plutonium metal: content expansion initiative for the plutonium air transportable (PAT01) packaging  

Science Conference Proceedings (OSTI)

The National Nuclear Security Administration (NNSA) has submitted an application to the Nuclear Regulatory Commission (NRC) for the air shipment of plutonium metal within the Plutonium Air Transportable (PAT-1) packaging. The PAT-1 packaging is currently authorized for the air transport of plutonium oxide in solid form only. The INMM presentation will provide a limited overview of the scope of the plutonium metal initiative and provide a status of the NNSA application to the NRC.

Caviness, Michael L [Los Alamos National Laboratory; Mann, Paul T [NNSA/ALBUQUERQUE; Yoshimura, Richard H [SNL

2010-01-01T23:59:59.000Z

38

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

SciTech Connect

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

39

COMPARISON OF RESPONSE OF 9977 TEST PACKAGES TO ANALYTICAL RESULTS  

Science Conference Proceedings (OSTI)

Each of the hypothetical accident test cases for the 9977 prototypes was included in the battery of finite element structural analyses performed for the package. Comparison of the experimental and analytical results provides a means of confirming that the analytical model correctly represents the physical behavior of the package. The ability of the analytical model to correctly predict the performance of the foam overpack material for the crush test is of particular interest. The dissipation of energy in the crushing process determines the deceleration of the package upon impact and the duration of the impact. In addition, if the analytical model correctly models the foam behavior, the predicted deformation of the package will match that measured on the test articles. This study compares the deformations of the test packages with the analytical predictions. In addition, the impact acceleration and impact duration for the test articles are compared with those predicted by the analyses.

Smith, A; Tsu-Te Wu, T

2007-12-05T23:59:59.000Z

40

Drop Tests of 325 Pound 6M Specification Packages  

Science Conference Proceedings (OSTI)

Testing of 6M specification packages, performed in response to concerns over the integrity of the clamp-ring closure, showed that the clamp-ring was unable to retain the top in thirty foot drop tests of packages having the maximum allowed weight (290 kg or 640 lb). To determine if the clamp-ring closure was adequate for packages with lower contents weight, a series of tests were performed on packages weighing 147 kg (325 lb) at a range of impact angles. The results showed that the standard clamp-ring closure was unable to retain the top in tests of standard 6M packages weighing 147 kg (325 lb). A test employing a plywood disk enhanced closure with impact at 6.5 degrees retained its top successfully.

SMITH, AC

2004-04-30T23:59:59.000Z

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 for the 6M Specification Package Closure Investigation  

SciTech Connect

Results of tests of drum-type RAM packages employing conventional clamp-ring closures have caused concern over the DOT 6M Specification Package. To clarify these issues, a series of tests were performed to determine the response of the clamp-ring closure to the regulatory Hypothetical Accident Condition (9m) drop tests, for packages at maximum allowable weight. Three enhanced closure designs were also tested: the Clamshell, plywood disk reinforcement, and J-Clip. The results of the tests showed that the standard closure was unable to retain the top for both Center-of-Gravity-Over-Corner and Shallow Angle cases, for the standard package, at its maximum allowed weight. Similar results were found for packages dropped from a reduced height. The Clamshell design provided the best performance of the enhanced closures. It was concluded that the closure ring design employed on the 6M is inadequate to retain the top during the regulatory test sequence, for packages at the maximum allowed weight. For large heavy packages, the Center-of-Gravity- Over-Corner case is more challenging than the Shallow Angle case. The Clamshell design securely retained the top for all HAC test cases, and prevented formation of any opening which could compromise fire test performance.

SMITH, AC

2004-04-30T23:59:59.000Z

42

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

Science Conference Proceedings (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

43

Test procedure forms for sludge retrieval and packaging  

SciTech Connect

This document provides test procedure forms for sludge retrieval and packaging tests in the 305 Cold Test Facility. The completed and approved forms provide all descriptions, criteria and analysis to safely perform sludge equipment tests in the 305 Cold Test Facility.

Feigenbutz, L.V.

1994-09-27T23:59:59.000Z

44

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

Science Conference Proceedings (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

45

CRUSH TESTING OF 9977 GENERAL PURPOSE FISSILE PACKAGINGS  

SciTech Connect

The 9977 General Purpose Fissile Package (GPFP) was designed in response to the adoption of the crush test requirement in the US regulations for packages for radioactive materials (10 CFR 71). This presentation on crush testing of the 9977 GPFP Reviews origins of Crush Test Requirements and implementation of crush test requirements in 10 CFR 71. SANDIA testing performed to support the rule making is reviewed. The differences in practice, on the part of the US Department of Energy from those required by the NRC for commercial purposes, are explained. The design features incorporated into the 9977 GPFP to enable it to withstand the crush test and the crush tests performed on the 9977 are described. Lessons learned from crush testing of GPFP packagings are given.

Smith, A.

2010-07-28T23:59:59.000Z

46

Materials Transportation Testing & Analysis at Sandia National Laboratories  

NLE Websites -- All DOE Office Websites (Extended Search)

Testing Testing Doug Ammerman, (505) 845-8158 Type B packages that transport radioactive materials must survive a sequence of full-scale (actual physical size) impact, puncture, fire, and immersion tests designed to replicate transportation accident conditions. The Hypothetical Accident Conditions (six tests as defined in 10 CFR Part 71.73) tests 1 through 4 (Drop, Crush, Puncture and Fire) are sequential, test 5 (Immersion) is performed on either a previously tested or untested package. Free Drop Test Crush Test Puncture Test Thermal Test Immersion Test [drop] Click to view picture [crush] Click to view picture [puncture] Click to view picture [thermal] Click to view picture [immersion] Click to view picture Dropping a package from 30 feet onto an unyielding target. (the unyielding target forces all of the deformation to be in the package, none in the target). The speed on impact is 44 feet per second or 30 miles per hour. Dropping a 1100 pound steel plate from 30 feet onto a package. This test is only required for packages weighing less than 1100 pounds. The speed on impact is 44 feet per second or 30 miles per hour. Dropping a package from 40 inches onto a welded, 6 inch diameter, steel spike. The speed on impact is 14.6 feet per second or 10 miles per hour. Placing a package 40 inches above a pool of burning fuel for 30 minutes at 800 degrees Celsius (1475 degrees Fahrenheit). Placing a package under 50 feet of water for 8 hours. Fissile material packages are also immersed under 3 feet of water for 8 hours sequentially after tests 1 through 4

47

Materials Transportation Testing & Analysis at Sandia National Laboratories  

NLE Websites -- All DOE Office Websites (Extended Search)

Testing Testing Carlos Lopez, (505) 845-9545 Packages transporting the larger "Type B" quantities of radioactive materials must be qualified and certified under Title 10, Code of Federal Regulations, Part 71, or under the equivalent international standard ST-1 issued by the International Atomic Energy Agency. The principal thermal qualification test is the 30 minute pool fire. As part of the National Transportation Program, the Transportation Risk & Packaging Program at Sandia can plan and conduct these tests for DOE and other package suppliers. Test Plans, QA plans and other necessary test documents can be prepared for customer and regulatory approval. Tests may be conducted with a variety of available facilities at Sandia, including large pools, an indoor fire facility, and a radiant heat test

48

Materials Transportation Testing & Analysis at Sandia National Laboratories  

NLE Websites -- All DOE Office Websites (Extended Search)

Materials Characterization Materials Characterization Paul McConnell, (505) 844-8361 The purpose of hazardous and radioactive materials, i.e., mixed waste, packaging is to enable this waste type to be transported without posing a threat to the health or property of the general public. To achieve this goal, regulations have been written establishing general design requirement for such packagings. Based on these regulatory requirements, a Mixed Waste Chemical Compatibility Testing Program is intended to assure regulatory bodies that the issue of packaging compatibility towards hazardous and radioactive materials has been addressed. Such a testing program has been developed in the Transportation Systems Department at Sandia National Laboratories. Materials Characterization Capabilities

49

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

SciTech Connect

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

50

Materials Transportation Testing & Analysis at Sandia National Laboratories  

NLE Websites -- All DOE Office Websites (Extended Search)

Analysis Analysis Doug Ammerman, (505) 845-8158 Structural analysis utilizes computer design and analysis tools to provide package designers and certifiers with the most accurate method of determining package response to transportation environments. Computer analysis is an application of known engineering principles that take advantage of high-power computing capabilities in solving the response of computer models to various environments with complex mathematical calculations. It can be used for package certification by generating a computer model of a test object (package) and subjecting it to an accident environment to understand its response. A computer model must be constructed with the same weights, dimensions, hardnesses, specific heat, conduction, etc. as an

51

DOE Order Self Study Modules - DOE O 460.1C Packaging and Transportation Safety and DOE O 460.2A Departmental Materials Transportation and Packaging Management  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

60.1C 60.1C PACKAGING AND TRANSPORTATION SAFETY DOE O 460.2A DEPARTMENTAL MATERIALS TRANSPORTATION AND PACKAGING MANAGEMENT DOE O 460.1C and 460.2A Familiar Level June 2011 1 DOE O 460.1C PACKAGING AND TRANSPORTATION SAFETY DOE O 460.2A DEPARTMENTAL MATERIALS TRANSPORTATION AND PACKAGING MANAGEMENT FAMILIAR LEVEL _________________________________________________________________________ OBJECTIVES Given the familiar level of this module and the resources, you will be able to perform the following: 1. What are the objectives of U.S. Department of Energy (DOE) O 460.1C? 2. What is the DOE/National Nuclear Security Administration (NNSA) exemption process in DOE O 460.1C? 3. What are the onsite safety requirements specified by DOE O 460.1C? 4. What are the objectives of DOE O 460.2A?

52

THERMAL TESTING OF 9977 GENERAL PURPOSE FISSILE PACKAGE USING A POOL FIRE  

Science Conference Proceedings (OSTI)

The 9977/9978 General Purpose Fissile Package (GPFP), has been designed as a cost-effective, user-friendly replacement for the DOT 6M Specification Package for transporting Plutonium and Uranium metals and oxides. To ensure the capability of the 9977 GPFP to withstand the regulatory crush test, urethane foam was chosen for the impact absorbing overpack. As part of the package development it was necessary to confirm that the urethane foam overpack would provide the required protection for the containment vessel during the thermal test portion of the Hypothetical Accident Conditions Sequential Tests. Development tests of early prototypes were performed, using a furnace. Based on the results of the development tests, detailed design enhancements were incorporated into the final design. Examples of the definitive 9977 design configuration were subjected to an all-engulfing pool fire test, as part of the HAC Sequential Tests, to support the application for certification. Testing has confirmed the package's ability to withstand the HAC thermal tests.

Smith, A; Cecil May, C; Lawrence Gelder, L; Glenn Abramczyk, G

2007-02-15T23:59:59.000Z

53

Wafer-Package Test Mix for Optimal Defect Detection and Test Time Savings  

Science Conference Proceedings (OSTI)

Editor's note: For years, it has been common to run a test at wafer and then exactly thesame test again at package. This article shows how one company took adetailed look at the wafer/package test mix and adjusted it to reduce costwhile retaining quality.¿Rob ...

Peter C. Maxwell

2003-09-01T23:59:59.000Z

54

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

55

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

Science Conference Proceedings (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

56

Trans_package_Poster_Draft_8_7_12.indd  

National Nuclear Security Administration (NNSA)

breaking open or releasing radiation including: * Free-Drop Test * Puncture Test * Thermal Test Type B transportation packages that could be used to transport radioactive...

57

Transportation Safeguards & Security Test Bed (TSSTB) | ORNL  

NLE Websites -- All DOE Office Websites (Extended Search)

Transportation Safeguards and Security Test Bed May 30, 2013 The Transportation Safeguards and Security Test Bed consists of a test-bed vehicle and a monitoringlaboratorytraining...

58

Materials Transportation Testing & Analysis at Sandia National...  

NLE Websites -- All DOE Office Websites (Extended Search)

Send Your Comments andor Questions (Fill in blank fields and click on "Submit" to send) Send To: Transportation Risk & Packaging Your Name: Your E-mail Address: Topic of Interest:...

59

Radcalc for Windows 2.0 transportation packaging software to determine hydrogen generation and transportation classification  

DOE Green Energy (OSTI)

Radclac for Windows is a user friendly menu-driven Windows compatible software program with applications in the transportation of radioactive materials. It calculates the radiolytic generation of hydrogen gas in the matrix of low-level and high-level radioactive wastes. It also calculates pressure buildup due to hydrogen and the decay heat generated in a package at seal time. It computes the quantity of a radionuclide and its associated products for a given period of time. In addition, the code categorizes shipment quantities as reportable quantity (RQ), radioactive Type A or Type B, limited quality (LQ), low specific activity (LSA), highway road controlled quality (HRCQ), and fissile excepted using US Department of Transportation (DOT) definitions and methodologies.

Green, J.R.

1996-10-21T23:59:59.000Z

60

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

SciTech Connect

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

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

Materials Transportation Testing & Analysis at Sandia National Laboratories  

NLE Websites -- All DOE Office Websites (Extended Search)

Unique Solutions] Unique Solutions] [Working With Us] [Contacting Us] [News Center] [Search] [Home] [navigation panel] Materials Transportation Testing & Analysis Our Mission Our Contacts Write to Us Package Development Risk Assessment RADTRAN GIS Mapping Structural Analysis Thermal Analysis Structural Testing Thermal Testing MIDAS Data Aquisition System Concepts Materials Characterization Regulatory Development Certification Support RMIR Data Base Scientific Visualization Mobile Instrumentation Data Acquisition System (MIDAS) Doug Ammerman, (505) 845-8158 The Mobile Instrumentation Data Acquisition System (MIDAS), developed by Sandia National Laboratories for the U.S. Department of Energy, provides on-site data acquisition of containers that transport radioactive materials during impact, puncture, fire, and immersion tests.

62

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

DOE Green Energy (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

63

Packaging and transportation-related occurrence reports - January-March 1994  

SciTech Connect

The Oak Ridge National Lab. (ORNL) Packaging and Transportation Safety Program (PATS), which is sponsored by the US Department of Energy (DOE) Office of Environment, Safety and Health Transportation and Packaging Safety Division, EH-332, has been charged with the responsibility of retrieving reports and information pertaining to transportation or packaging incidents or accidents from the Occurrence Reporting and Processing System (ORPS). These selected reports are being analyzed for trends, impact on EH-332 policies and concerns, and lessons learned concerning transportation and packaging safety. This task is designed not only to keep EH-332 aware of what is occurring on DOE sites and potential transportation and packaging problems that may need attention, but also it is intended to allow future dissemination of lessons learned to the Operations Offices and subsequently to management and operating contractors. This report, which covers the first quarter of 1994, includes the weekly tabular reports OR-91-1 through OR-94-13, which were submitted to EH-332 for its information and use. Thirteen reports containing 43 selected occurrences were transmitted during this quarter.

Welch, M.J.; Dickerson, L.S.; Jennings, S.D.

1994-06-01T23:59:59.000Z

64

DOE O 461.1B, Packaging and Transportation for Offsite Shipment of Materials of National Security Interest  

Directives, Delegations, and Requirements

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 ...

2010-12-20T23:59:59.000Z

65

THERMAL TESTING OF PROTOTYPE GENERAL PURPOSE FISSILE PACKAGES USING A FURNACE  

SciTech Connect

The 9977/9978 General Purpose Fissile Package (GPFP) was designed by SRNL to replace the DOT 6M Specification Package and ship Plutonium and Uranium metals and oxides. Urethane foam was used for the overpack to ensure the package would withstand the 10CFR71.73(c)(2) crush test, which is a severe test for drum-type packages. In addition, it was necessary to confirm that the urethane foam configuration provided adequate thermal protection for the containment vessel during the subsequent 10CFR71.73(c)(4) thermal test. Development tests were performed on early prototype test specimens of different diameter overpacks and a range of urethane foam densities. The thermal test was performed using an industrial furnace. Test results were used to optimize the selection of package diameter and foam density, and provided the basis for design enhancements incorporated into the final package design.

Smith, A; Lawrence Gelder, L; Paul Blanton, P

2007-02-16T23:59:59.000Z

66

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

DOE Green Energy (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

67

Packaging- and transportation-related occurrence reports, January--March 1995  

Science Conference Proceedings (OSTI)

Reports on transportation/packaging incidents, from the Occurrence Reporting and Processing System, are being analyzed for trends, impact on DOE EH-32 policies and concerns, and lessons learned concerning transportation and packaging safety. Besides keeping EH-32 aware of current incidents and potential problems that may need attention on DOE sites, this task allows future dissemination of lessons learned to the Operations Offices and to management and operating contractors. This report covers the weekly tabular reports OR-95-01 through OR-95-13, which contained a total of 50 occurrence reports.

Dickerson, L.S.; Welch, M.J.; Armstrong, C.J.

1995-04-01T23:59:59.000Z

68

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

Science Conference Proceedings (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

69

Options for Pursuing Moderator Exclusion for Application to Spent-Fuel Transportation Packages  

Science Conference Proceedings (OSTI)

This report discusses options for pursuing moderator exclusion, either by itself or in combination with burnup credit, for application to the criticality evaluation of spent nuclear fuel transportation packages. Also, information is provided on how to proceed in developing a request for rulemaking if the industry determines that changes to the existing regulations for streamlining implementation of moderator exclusion are highly desirable.

2005-12-15T23:59:59.000Z

70

Options for Pursuing Moderator Exclusion for Application to Spent-Fuel Transportation Packages  

Science Conference Proceedings (OSTI)

This report discusses the concept of moderator exclusion, either by itself or in combination with burnup credit, for application to the criticality evaluation of spent nuclear fuel transportation packages. Information is also provided on how to proceed in the development of regulatory amendments if the industry determines that changes to the existing regulations for streamlining implementation of moderator exclusion are highly desirable.

2004-12-23T23:59:59.000Z

71

TRANSPORTATION RESEARCH BOARD Testing and Inspection Levels  

E-Print Network (OSTI)

TRANSPORTATION RESEARCH BOARD Testing and Inspection Levels for Hot-Mix Asphaltic Concrete Overlays, Editorial AssistantCHRISTOPHER HEDGES, Senior Program Officer TRANSPORTATION RESEARCH BOARD EXECUTIVE COMMITTEE 2000 OFFICERS Chair: Martin Wachs, Director, Institute of Transportation Studies, University

Sheridan, Jennifer

72

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

DOE Green Energy (OSTI)

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

Djordjevic, S.; Drez, P.; Murthy, D. (International Technology Corp., Albuquerque, NM (USA)); Temus, C. (Nuclear Packaging Corp., Federal Way, WA (USA))

1990-01-01T23:59:59.000Z

73

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

Science Conference Proceedings (OSTI)

The packaging, transportation and disposal of large, retired reactor components from operating or decommissioning nuclear plants pose unique challenges from a technical as well as regulatory compliance standpoint. In addition to the routine considerations associated with any radioactive waste disposition activity, such as characterization, ALARA, and manifesting, the technical challenges for large radioactively contaminated components, such as access, segmentation, removal, packaging, rigging, lifting, mode of transportation, conveyance compatibility, and load securing require significant planning and execution. In addition, the current regulatory framework, domestically in Titles 49 and 10 and internationally in TS-R-1, does not lend itself to the transport of these large radioactively contaminated components, such as reactor vessels, steam generators, reactor pressure vessel heads, and pressurizers, without application for a special permit or arrangement. This paper addresses the methods of overcoming the technical and regulatory challenges. The challenges and disposition decisions do differ during decommissioning versus component replacement during an outage at an operating plant. During decommissioning, there is less concern about critical path for restart and more concern about volume reduction and waste minimization. Segmentation on-site is an available option during decommissioning, since labor and equipment will be readily available and decontamination activities are routine. The reactor building removal path is also of less concern and there are more rigging/lifting options available. Radionuclide assessment is necessary for transportation and disposal characterization. Characterization will dictate the packaging methodology, transportation mode, need for intermediate processing, and the disposal location or availability. Characterization will also assist in determining if the large component can be transported in full compliance with the transportation and disposal regulations and criteria or if special authorizations must be granted to transport and/or dispose. The U.S. DOT routinely issues special permits for large components where compliance with regulatory or acceptance criteria is impractical or impossible to meet. Transportation and disposal safety must be maintained even under special permits or authorizations. For example, if transported un-packaged, performance analysis must still be performed to assess the ability of the large component's outer steel shell to contain the internal radioactive contamination under normal transportation conditions and possibly incidence normal to transportation. The dimensions and weight of a large component must be considered when determining the possible modes of transportation (rail, water, or highway). At some locations, rail and/or barge access is unavailable. Many locations that once had an active rail spur to deliver new construction materials and components have let the spur deteriorate to the point that repair and upgrade of the spur is no longer economically feasible. Barge slips that have not been used since new plant construction require significant repair and/or dredging. Short on-site haul routes must be assessed for surface and subsurface conditions, as well as longer off-site routes. Off-site routes require clearance approvals from the regulatory authorities or, in the case of rail transport, the rail lines. Significant engineering planning and analysis must be performed during the pre-mobilization. In conclusion, the packaging, transportation, and disposal of large, oversized radioactively contaminated components removed during plant decommissioning is complex. However, over the last 15 years, a 100 or more components have been safely and compliantly packaged and transported for processing and/or disposal.

Lewis, Mark S. [EnergySolutions: 140 Stoneridge Drive, Columbia, SC 29210 (United States)

2008-01-15T23:59:59.000Z

74

Excess Weapons Plutonium Disposition: Plutonium Packaging, Storage and Transportation and Waste Treatment, Storage and Disposal Activities  

SciTech Connect

A fifth annual Excess Weapons Plutonium Disposition meeting organized by Lawrence Livermore National Laboratory (LLNL) was held February 16-18, 2004, at the State Education Center (SEC), 4 Aerodromnya Drive, St. Petersburg, Russia. The meeting discussed Excess Weapons Plutonium Disposition topics for which LLNL has the US Technical Lead Organization responsibilities. The technical areas discussed included Radioactive Waste Treatment, Storage, and Disposal, Plutonium Oxide and Plutonium Metal Packaging, Storage and Transportation and Spent Fuel Packaging, Storage and Transportation. The meeting was conducted with a conference format using technical presentations of papers with simultaneous translation into English and Russian. There were 46 Russian attendees from 14 different Russian organizations and six non-Russian attendees, four from the US and two from France. Forty technical presentations were made. The meeting agenda is given in Appendix B and the attendance list is in Appendix C.

Jardine, L J; Borisov, G B

2004-07-21T23:59:59.000Z

75

Testing hadronic-interaction packages at cosmic-ray energies  

Science Conference Proceedings (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

76

DRFM: A new package for the evaluation of gas-phase transport properties  

DOE Green Energy (OSTI)

This report describes a complete and modernized procedure to evaluate pure species, binary and mixture transport properties of gases in the low density limit. This includes a description of the relationships used to calculate these quantities and the means used to obtain the necessary input data. The purpose of this work is to rectify certain limitations of previous transport packages, specifically: to employ collision integrals suitable for high temperatures, to modernize the mixture formula, and to modernize the input data base. This report includes a set of input parameters for: the species involved in H{sub 2}-, CO - air combustion, the noble gases, methane and the oxides of nitrogen.

Paul, P.H.

1997-11-01T23:59:59.000Z

77

Occurrence Classifications, Severity Weighting, and Normalization for the DOE Packaging and Transportation Safety Metrics Indicator Program  

SciTech Connect

The US Department of Energy (DOE) Occurrence Reporting and Processing System (ORPS) is an interactive computer system designed to support DOE-owned or -operated facilities in reporting and processing information concerning occurrences related to facility operations. The Oak Ridge National Laboratory has been charged by the DOE National Transportation Program Albuquerque (NTPA) with the responsibility of retrieving reports and information pertaining to packaging and transportation (P and T) incidents from the centralized ORPS database. These selected reports are analyzed for trends, impact on P and T operations and safety concerns, and ''lessons learned'' in P and T safety.

Dickerson, L.S.; Pope, R.B.; Michelhaugh, R.D.; Harrison, I.G.; Hermann, B.; Lester, P.B.

1999-06-01T23:59:59.000Z

78

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

Science Conference Proceedings (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

79

Evaluation of Packaging Film Mechanical Integrity Using a Standardized Scratch Test  

E-Print Network (OSTI)

Polymeric packaging films see widespread use in the food packaging industry, and their mechanical integrity is paramount to maintaining product appearance, freshness, and overall food safety. Current testing methods, such as tensile or puncture tests, do not necessarily correlate well with field damages that are observed to be scratch-like. The standardized linearly increasing load scratch test is investigated as a new means of evaluating the mechanical integrity of packaging films. Mechanical clamp and vacuum fixtures were considered for securing the films to a set of backing materials and tested under various testing rates and film orientation conditions. Film performance was evaluated according to their puncture load. Based on the above study, the vacuum fixture offers the most consistent and meaningful results by providing a more intimate contact between film and backing and minimizing uncontrolled buckling of the film during testing. Additional testing was also carried out on a commercial film to confirm similarity between damage observed in the scratched films and that from the field. The scratch test gives good correlation between field performance and scratch test results on a set of commercial films. The usefulness of the scratch test methodology for packaging film mechanical integrity evaluation is discussed. Scratch-induced damages on multi-layer commercial packaging films are investigated using cross- and longitudinal-sectioning. Scratch test results show clear distinction between the two tested systems on both the inside and outside surfaces. Microscopy was performed to investigate the feasibility of utilizing this methodology as a tool for packaging film structure evaluation by determining the effect each layer has on the resistance of scratch damages. It is shown that the film showing superior scratch test results also shows significantly better stress distribution through its layers during the scratch test, as well as better layer adhesion during severe deformation. The scratch test shows good ability to provide more in-depth film mechanical integrity testing by allowing for layer-by-layer analysis of damages and layer adhesion after testing.

Hare, Brian

2011-08-01T23:59:59.000Z

80

Field test of ultra-low head hydropower package based on marine thrusters. Final report  

DOE Green Energy (OSTI)

The project includes the design, fabrication, assembly, installation, and field test of the first full-scale operating hydropower package (turbine, transmission, and generator) based on a design which incorporates a marine-thruster as the hydraulic prime mover. Included here are: the project overview; engineering design; ultra-low head hydropower package fabrication; component procurement, cost control, and scheduling; thruster hydraulic section installation; site modeling and resulting recommended modifications; testing; and baseline environmental conditions at Stone Drop. (MHR)

Not Available

1983-12-01T23:59:59.000Z

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
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to obtain the most current and comprehensive results.


81

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

SciTech Connect

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

82

Review of the Lawrence Livermore Nationa Laboratory Identiified Defective Department of Transportation Hazardous Material Packages  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

5 5 Site Visit Report - Review of the Lawrence Livermore National Laboratory Identified Defective Department of Transportation Hazardous Material Packages This site visit report documents the results of Office of Health, Safety and Security's review of the Lawrence Livermore National Laboratory (LLNL) identification, immediate actions, communications, documentation, evaluation, reporting and follow-up to the discovery of defective Department of Transportation (DOT) UN1A2 55- and 30-gallon open head single bolt closure steel drums intended for storage and transportation of hazardous waste and materials. This review, conducted on January 26-29, 2010, was sponsored by the DOE Livermore Site Office (LSO) to support interface with the lab and this report is intended to support follow-up

83

DEVELOPMENT OF BURN TEST SPECIFICATIONS FOR FIRE PROTECTION MATERIALS IN RAM PACKAGES  

SciTech Connect

The regulations in 10 CFR 71 require that the radioactive material (RAM) packages must be able to withstand specific fire conditions given in 10 CFR 71.73 during Hypothetical Accident Conditions (HAC). This requirement is normally satisfied by extensive testing of full scale test specimens under required test conditions. Since fire test planning and execution is expensive and only provides a single snapshot into a package performance, every effort is made to minimize testing and supplement tests with results from computational thermal models. However, the accuracy of such thermal models depends heavily on the thermal properties of the fire insulating materials that are rarely available at the regulatory fire temperatures. To the best of authors knowledge no test standards exist that could be used to test the insulating materials and derive their thermal properties for the RAM package design. This paper presents a review of the existing industry fire testing standards and proposes testing methods that could serve as a standardized specification for testing fire insulating materials for use in RAM packages.

Gupta, N.

2010-03-03T23:59:59.000Z

84

Drop Test Results for the Combustion Engineering Model No. ABB-2901 Fuel Pellet Package  

SciTech Connect

The U.S. Nuclear Regulatory Commission (USNRC) contracted with the Packaging Review Group (PRG) at Lawrence Livermore National Laboratory (LLNL) to conduct a single, 30-ft shallow-angle drop test on the Combustion Engineering ABB-2901 drum-type shipping package. The purpose of the test was to determine if bolted-ring drum closures could fail during shallow-angle drops. The PRG at LLNL planned the test, and Defense Technologies Engineering Division (DTED) personnel from LLNL's Site-300 Test Group executed the plan. The test was conducted in November 2001 using the drop-tower facility at LLNL's Site 300. Two representatives from Westinghouse Electric Company in Columbia, South Carolina (WEC-SC); two USNRC staff members; and three PRG members from LLNL witnessed the preliminary test runs and the final test. The single test clearly demonstrated the vulnerability of the bolted-ring drum closure to shallow-angle drops-the test package's drum closure was easily and totally separated from the drum package. The results of the preliminary test runs and the 30-ft shallow-angle drop test offer valuable qualitative understandings of the shallow-angle impact.

Hafner, R S; Mok, G C; Hagler, L G

2004-04-23T23:59:59.000Z

85

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

DOE Green Energy (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

86

Development and feasibility of a waste package coupled reactive transport model (AREST-CT)  

Science Conference Proceedings (OSTI)

Most models that analyze the waste package and engineered barrier system (near-field) of an underground geologic repository assume constant boundary conditions at the waste form surface and constant chemical properties of the groundwater. These models are useful for preliminary modeling, iterative modeling to estimate uncertainties, and as a source for a total systems analysis. However, the chemical behavior of the system is a very important factor in the containment and release of radionuclides, and one needs to understand the underlying processes involved. Therefore, the authors are developing a model to couple the calculation of the chemical properties with the reactive transport which can be used to assess the near-field. This report describes the models being implemented and presents some simple analyses demonstrating the feasibility of the chemical and coupled transport models.

Engel, D.W.; McGrail, B.P.; Fort, J.A.; Roberts, J.S.

1994-05-01T23:59:59.000Z

87

9978 AND 9975 TYPE B PACKAGING INTERNAL DATA COLLECTION FEASIBILITY TESTING  

SciTech Connect

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

88

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

DOE Green Energy (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

89

DOE G 460.2-1, Implementation Guide for Use with DOE O 460.2 Departmental Materials Transportation and Packaging Management  

Directives, Delegations, and Requirements

The purpose of this guide is to assist those responsible for transporting and packaging Department materials, and to provide an understanding of Department ...

1996-11-15T23:59:59.000Z

90

Thermal Analysis of the 9975 Package with the 3013 Configuration During Normal Conditions of Transport  

DOE Green Energy (OSTI)

Thermal analysis of the 9975 package with three configurations of the BNFL 3013 outer container (with Rocky Flats, SRS, and BNFL inner containers) have been performed for Normal Conditions of Transport (NCT) of plutonium oxide and metal. The NCT is defined in 10 CFR 71.71(c)(1) s an ambient of 100 F (38 C) in still air with 800 W/m{sup 2} and 400 W/m{sup 2} of solar heating on the drum top and sides, respectively. The 9975 drum package is considered to be in an upright position, and the drum bottom is adiabatic. The Rocky and SRS 3013 configurations with Pu metal contents (19 watts) result in acceptable (similar) packaging temperatures, however the plutonium metal temperatures are lower for the SRS design (SRS has helium fill gas whereas Rocky is essentially air filled). The BNFL configuration for Pu oxide contents (19 watts) result in acceptable temperatures and pressures based on limits in the 9975 Safety Analysis Report (SARP). However, for 30 watts of Pu oxide, the fiberboard peak temperatures are very near the SARP allowable. The pressure in the 3013 container is 688.4 psig for the 30 watt Pu oxide content and 569.5 psig for the 19 watt Pu oxide content. Gas species in the pressure computations include air, helium and hydrogen from complete radiolysis of the moisture. The oxygen generated from the radiolysis of the water is not included as a pressure contributor. The gas temperature in the pressure computations was very conservatively assumed to be the maximum plutonium temperature.

Hensel, S

1999-02-22T23:59:59.000Z

91

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

Science Conference Proceedings (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 WHC-CM-2-14, Hazardous Material Packaging and Shipping. The drum will be used for transport of 327 Building legacy waste from the 300 Area to the Transuranic Waste Storage and Assay Facility in the 200 West Area and on to a Solid Waste Storage Facility, also in the 200 Area.

Adkins, H.E.

1996-10-29T23:59:59.000Z

92

Evaluation of Basic Parameters for Packaging, Storage and Transportation of Biomass Material from Field to Biorefinery  

E-Print Network (OSTI)

The universal adoption of biomass materials as an alternate fuel source to fossil fuels for transportation and electricity has been hindered by the high transportation costs involved in fuel production. Optimization of these initial costs will make the eco-friendly fuels more economically viable. Biomass is a promising feedstock for biofuels primarily because it is a renewable and sustainable resource. Among the most studied grassland crops, switchgrass is a perennial warm-season grass and has been identified as a potential energy crop. This research focuses on evaluating various physical parameters which affect the economic feasibility of packaging and transporting switchgrass from the field to the biorefinery. The switchgrass was harvested using a mower conditioner followed by field chopping after varying drying periods. The first harvesting period spanned from early November to mid December 2007 and the second was August to October 2008. Densification properties of chopped switchgrass were studied under compression. The effects of compressive stresses (41 to 101 kPa), number of strokes (1 to 10), moisture content (9 to 62 percent) and chopping length (63 and 95 mm) on the densification of chopped switchgrass were studied. The final dry matter density (DMD) increased with the compressive stresses and the number of strokes, small chop length and low moisture content. The maximum free-standing DMD obtained was 245 kg/m^3.

Paliwal, Richa

2010-12-01T23:59:59.000Z

93

Qualification and testing process to implement anti-counterfeiting technologies into IC packages  

Science Conference Proceedings (OSTI)

Counterfeiting is no longer limited to just fashion or luxury goods, the phenomenon has now reached electronics components which failure represents a high risk to the safety and security of human communities. One way for the semiconductor (SC) industry ... Keywords: anti-counterfeiting technologies, authentication, component counterfeiting, failure analysis, failure prevention, re-packaging, reliability testing, remarking

Nathalie Kae-Nune, Stephanie Pesseguier

2013-03-01T23:59:59.000Z

94

Radioisotope Thermoelectric Generator Package O-Ring Seal Material Validation Testing  

DOE Green Energy (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

95

Instrumentation package for a shrouded SWERVE/EPW rocket sled test  

SciTech Connect

A shrouded SWERVE test vehicle was instrumented and tested at MACH 3. The instrumentation consisted of accelerometers, pressure transducers, and event markers arranged in a nine-channel configuration. Data from the instrumentation were transmitted from the vehicle through an FM/FM telemetry system. Two fire sets and their associated command and controls were also included in this package. The event was recorded on high-speed color film, and the trajectory data were recorded by a laser tracker.

Garavaglia, E.G.

1987-08-01T23:59:59.000Z

96

Performance-oriented packaging testing of wood box for M83769/4-1 battery. Final report  

SciTech Connect

The current packaging configuration for the M873769/4-1 Battery was tested for conformance to Performance Oriented Packaging regulations. The cleated plywood box was tested with a gross weight of 214 pounds and met the requirements and retained its contents.

Libbert, K.J.

1991-05-01T23:59:59.000Z

97

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

SciTech Connect

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

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

2003-09-15T23:59:59.000Z

98

Transport Test Problems for Hybrid Methods Development  

Science Conference Proceedings (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

99

Review of waste package verification tests. Semiannual report, April 1985-September 1985  

Science Conference Proceedings (OSTI)

Several studies were completed this period to evaluate experimental and analytical methodologies being used in the DOE waste package program. The first involves a determination of the relevance of the test conditions being used by DOE to characterize waste package component behavior in a salt repository system. Another study focuses on the testing conditions and procedures used to measure radionuclide solubility and colloid formation in repository groundwaters. An attempt was also made to evaluate the adequacy of selected waste package performance codes. However, the latter work was limited by an inability to obtain several codes from DOE. Nevertheless, it was possible to comment briefly on the structures and intents of the codes based on publications in the open literature. The final study involved an experimental program to determine the likelihood of stress-corrosion cracking of austenitic stainless steels and Incoloy 825 in simulated tuff repository environments. Tests for six-month exposure periods in water and air-steam conditions are described. 52 figs., 48 tabs.

Soo, P. (ed.)

1986-01-01T23:59:59.000Z

100

Materials Transportation Testing & Analysis at Sandia National...  

NLE Websites -- All DOE Office Websites (Extended Search)

P-Thermal (Qtran) is a commercial resistor-capacitor network code Applications NUREG 0170 rework Modal study revision CLWR Tritium study Package design (ONC, EONC, BUSS...

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.


101

Materials Transportation Testing & Analysis at Sandia National...  

NLE Websites -- All DOE Office Websites (Extended Search)

Response Guidebook (ERG2000) was developed jointly by the U.S. Department of Transportation, Transport Canada, and the Secretariat of Communications and Transportation of...

102

CH Packaging Program Guidance  

SciTech Connect

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

103

CH Packaging Program Guidance  

Science Conference Proceedings (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

104

DOE Safety Metrics Indicator Program (SMIP) Fiscal Year 2000 Annual Report of Packaging- and Transportation-related Occurrences  

Science Conference Proceedings (OSTI)

The Oak Ridge National Laboratory (ORNL) has been charged by the DOE National Transportation Program (NTP) with the responsibility of retrieving reports and information pertaining to packaging and transportation (P&T) incidents from the centralized Occurrence Reporting and Processing System (ORPS) database. These selected reports have been analyzed for trends, impact on P&T operations and safety concerns, and lessons learned (LL) in P&T operations. This task is designed not only to keep the NTP aware of what is occurring at DOE sites on a periodic basis, but also to highlight potential P&T problems that may need management attention and allow dissemination of LL to DOE Operations Offices, with the subsequent flow of information to contractors. The Safety Metrics Indicator Program (SMIP) was established by the NTP in fiscal year (FY) 1998 as an initiative to develop a methodology for reporting occurrences with the appropriate metrics to show rates and trends. One of its chief goals has been to augment historical reporting of occurrence-based information and present more meaningful statistics for comparison of occurrences. To this end, the SMIP established a severity weighting system for the classification of the occurrences, which would allow normalization of the data and provide a basis for trending analyses. The process for application of this methodology is documented in the September 1999 report DOE Packaging and Transportation Measurement Methodology for the Safety Metrics Indicator Program (SMIP). This annual report contains information on those P&T-related occurrences reported to the ORPS during the period from October 1, 1999, through September 30, 2000. Only those incidents that occur in preparation for transport, during transport, and during unloading of hazardous material are considered as packaging- or transportation-related occurrences. Other incidents with P&T significance, but not involving hazardous material (such as vehicle accidents or empty packagings), are not rated by the SMIP criteria but are archived in the SMIP Subsidiary Database of occurrences, a sub-database of the main SMIP P&T Occurrence Database. A total of 146 reports were classified by the SMIP criteria, of which 144 have been finalized. Trending comparisons were made with these reports and the 851 other occurrence reports (ORs) accumulated in the SMIP P&T Occurrence Database since FY 1994, all of which were also evaluated according to the SMIP criteria. Additionally, information on the number of shipments made by DOE carriers and the types of materials transported was obtained from the Enterprise Transportation Analysis System (ETAS), formerly the Shipment Mobility Accountability Collection. This information was used in conjunction with the Transportation Routing Analysis Geographic Information System (TRAGIS, a GIS-based transportation and analysis model that replaces the older HIGHWAY and INTERLINE models) to estimate point-to-point mileage, yielding a metric of vehicle-miles or package-miles. This information was subsequently used to develop indicators for (1) determining the relative safety of DOE contractors who package and ship hazardous materials and (2) comparing of DOE P&T safety with that of private industry.

Dickerson, L.S.

2001-07-26T23:59:59.000Z

105

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

SciTech Connect

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

106

Testing and standards related to fluid and chemical transport  

Testing and standards related to fluid and chemical transport R. Doug Hooton UNIVERSITY OF TORONTO DEPT. CIVIL ENGINEERING

107

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

SciTech Connect

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

108

Development of Onsite Transportation Safety Documents for Nevada Test Site  

Science Conference Proceedings (OSTI)

Department of Energy (DOE) Orders require each DOE site to develop onsite transportation safety documents (OTSDs). The Nevada Test Site approach divided all onsite transfers into two groups with each group covered by a standalone OTSD identified as Non-Nuclear and Nuclear. The Non-Nuclear transfers involve all radioactive hazardous material in less than Hazard Category (HC)-3 quantities and all chemically hazardous materials. The Nuclear transfers involve all radioactive material equal to or greater than HC-3 quantities and radioactive material mated with high explosives regardless of quantity. Both OTSDs comply with DOE O 460.1B requirements. The Nuclear OTSD also complies with DOE O 461.1A requirements and includes a DOE-STD-3009 approach to hazard analysis (HA) and accident analysis as needed. All Nuclear OTSD proposed transfers were determined to be non-equivalent and a methodology was developed to determine if “equivalent safety” to a fully compliant Department of Transportation (DOT) transfer was achieved. For each HA scenario, three hypothetical transfers were evaluated: a DOT-compliant, uncontrolled, and controlled transfer. Equivalent safety is demonstrated when the risk level for each controlled transfer is equal to or less than the corresponding DOT-compliant transfer risk level. In this comparison the typical DOE-STD-3009 risk matrix was modified to reflect transportation requirements. Design basis conditions (DBCs) were developed for each non-equivalent transfer. Initial DBCs were based solely upon the amount of material present. Route-, transfer-, and site-specific conditions were evaluated and the initial DBCs revised as needed. Final DBCs were evaluated for each transfer’s packaging and its contents.

Frank Hand, Willard Thomas, Frank Sciacca, Manny Negrete, Susan Kelley

2008-05-08T23:59:59.000Z

109

Department of Transportation -- Exemption for using the Transuranic Package Transporter-I (TRUPACT-I) at the Idaho National Engineering Laboratory (Code of Federal Regulations, Title 49, Part 107, Subpart B -- Exemptions, 107-103 Application for Exemption)  

Science Conference Proceedings (OSTI)

Exemption from specific regulations is being sought for the Transuranic Package Transporter Model I (TRUPACT-I) container. The design has successfully undergone extensive testing of a quarter-scale model and a full-scale prototype of the container. Results from the analysis and testing are in the TRUPACT-1 Safely Analysis Report for Packaging (SARP), GA-Al8695/SAND 87-7104 (TTC0735), April 1987 (see Attachment 1). The container was never certified or used because of questions raised during the certification process. Two features of the container design failed to satisfy the regulations for Type B packaging. First, the design utilizes a venting system to control internal and external pressures; this venting system is not allowed by the Code of Federal Regulations, Title 10, Parts 71(h) and 71.51(b) [10 CFR 71.(h) and 71.51(b)]. Second, the maximum quantity fissile material proposed to be hauled in TRUPACT-I exceeded the limits in 10 CFR 71.63(b) for a single-containment container. To correct these design deficiencies, the vents would be plugged during transport, and the maximum quantity of fissile material would be limited to the allowables for a single-containment container. An engineering analysis showed that the container could safely transport radioactive material within the boundaries of the Idaho National Engineering Laboratory (INEL) with the vent system plugged (see Attachment 2). However, some of the requirements for determining pressure on a container need to be changed (i.e., exempted) to reflect conditions unique to the INEL. The following are the requirements needing to be changed for INEL conditions, variances being sought, and justifications for the variances.

Tyacke, M.J.; Macdonald, R.J.

1992-08-01T23:59:59.000Z

110

DOE Safety Metrics Indicator Program (SMIP) Fiscal Year 2001 Fourth Quarter Report of Packaging- and Transportation-related Occurrences  

Science Conference Proceedings (OSTI)

The Safety Metrics Indicator Program (SMIP) retrieved 44 packaging- or transportation-related occurrences from the Occurrence Reporting and Processing System (ORPS) during the period from July 1 through September 30, 2001. Only those incidents that occur in preparation for transport, during transport, and during unloading of hazardous material are considered as packaging- or transportation-related occurrences. Other incidents with packaging and transportation (P and T) significance but not involving hazardous material (such as vehicle accidents or empty packagings) are not rated to the SMIP criteria, but are archived in the SMIP Subsidiary Database of occurrences, a sub-database of the main SMIP P and T Occurrence Database. Thirty-two of the originally-selected 44 occurrences were appropriate for classification to the SMIP criteria, only 7 of which have offsite applicability. Eight of the original 44 reports are archived in a subsidiary database because they either do not involve the transport of hazardous material or do not involve transport by vehicle, plane, boat, or rail. The others either were deleted because more thorough review revealed that they were not strictly related to P and T or were canceled by the reporting site and removed from the ORPS. These occurrences have not been normalized as in the Annual Report of Occurrences because the necessary information is not yet available. The number and severity of the selected occurrence reports (ORs) are consistent with historical reporting. Contamination events continue to be among the most common type of occurrences; however, ''Shipping Preparation'' events decreased this quarter to only 4 events from the 21 reported last quarter. None of the 32 ORs that were rated had event consequence measures (W{sub EC}) greater than 2; 14 of them were categorized as having a W{sub EC} of 1. This means that all of the fourth-quarter FY 2001 ORs had only slight consequences at worst (i.e., resulting in minimal safety consequences with little potential for ultimately leading to suspected endangerment of people or environmental contamination). Because the event consequence measure is low, the overall hazard significance ratings (HSRs) are relatively low, indicating that the actual risks posed by the occurrences are not highly threatening. In fact, even the one reported emergency OR and the two unusual ORs had HSRs of 24 or less. The ORs are summarized in the appendix and listed along with their HSRs and stakeholder interest [currently termed the stakeholder and publicity rating (SPR)]. This enables one to get a feel for how the nature of an occurrence and its P and T significance translate into a severity rating. During the quarter, it was noticed that the number of events pertaining to either dropping packages or damaging packages while moving material by forklift had increased from previous years. Only four such events were reported in FY 1999 and seven events reported during FY 2000. However, 13 events have been reported during this fiscal year. Therefore a lessons learned (LL) pertaining to minimizing accidents while moving material by forklift was developed and posted to the U.S. Department of Energy LL list server to mitigate this trend.

Dickerson, L.S.

2001-11-30T23:59:59.000Z

111

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

DOE Green Energy (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

112

RH Packaging Program Guidance  

Science Conference Proceedings (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

113

RH Packaging Program Guidance  

Science Conference Proceedings (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

114

Materials Transportation Testing & Analysis at Sandia National Laboratories  

NLE Websites -- All DOE Office Websites (Extended Search)

Transportation Risk & Packaging Contacts Transportation Risk & Packaging Contacts Ken Sorenson Program Manager (505) 844-0074 kbsoren@sandia.gov David Miller Operations Manager (505) 284-2574 drmille@sandia.gov Administrative Assistant Pat Tode (505) 845-8339, 845-7800 pprippl@sandia.gov Financial Analyst Laurel Taylor (505) 845-8598 ljtaylo@sandia.gov Risk Assessment Ruth Weiner (505) 284-8406 rfweine@sandia.gov Jeremy Sprung (505) 844-0314 jlsprun@sandia.gov Doug Osborn (505) 284-6416 dosborn@sandia.gov RADTRAN Ruth Weiner (505) 284-8406 rfweine@sandia.gov GIS Mapping Doug Osborn (505) 284-6416 dosborn@sandia.gov Structural Analysis Doug Ammerman (505) 845-8158 djammer@sandia.gov Bob Kalan (505) 844-0244 rjkalan@sandia.gov Jeff Smith (505) 845-0299 jasmith@sandia.gov Thermal Analysis Carlos Lopez

115

CH Packaging Program Guidance  

Science Conference Proceedings (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

116

Safety analysis report for packaging: the ORNL DOT specification 6M - special form package  

SciTech Connect

The ORNL DOT Specification 6M - Special Form Package was fabricated at the Oak Ridge Nation al Laboratory (ORNL) for the transport of Type B solid non-fissile radioactive materials in special form. The package was evaluated on the basis of tests performed by the Dow Chemical Company, Rocky Flats Division, on the DOT-6M container and special form tests performed on a variety of stainless steel capsules at ORNL by Operations Division personnel. The results of these evaluations demonstrate that the package is in compliance with the applicable regulations for the transport of Type B quantities in special form of non-fissile radioactive materials.

Schaich, R.W.

1982-07-01T23:59:59.000Z

117

A class of ejecta transport test problems  

Science Conference Proceedings (OSTI)

Hydro code implementations of ejecta dynamics at shocked interfaces presume a source distribution function ofparticulate masses and velocities, f{sub 0}(m, v;t). Some of the properties of this source distribution function have been determined from extensive Taylor and supported wave experiments on shock loaded Sn interfaces of varying surface and subsurface morphology. Such experiments measure the mass moment of f{sub o} under vacuum conditions assuming weak particle-particle interaction and, usually, fully inelastic capture by piezo-electric diagnostic probes. Recently, planar Sn experiments in He, Ar, and Kr gas atmospheres have been carried out to provide transport data both for machined surfaces and for coated surfaces. A hydro code model of ejecta transport usually specifies a criterion for the instantaneous temporal appearance of ejecta with source distribution f{sub 0}(m, v;t{sub 0}). Under the further assumption of separability, f{sub 0}(m,v;t{sub 0}) = f{sub 1}(m)f{sub 2}(v), the motion of particles under the influence of gas dynamic forces is calculated. For the situation of non-interacting particulates, interacting with a gas via drag forces, with the assumption of separability and simplified approximations to the Reynolds number dependence of the drag coefficient, the dynamical equation for the time evolution of the distribution function, f(r,v,m;t), can be resolved as a one-dimensional integral which can be compared to a direct hydro simulation as a test problem. Such solutions can also be used for preliminary analysis of experimental data. We report solutions for several shape dependent drag coefficients and analyze the results of recent planar dsh experiments in Ar and Xe.

Hammerberg, James E [Los Alamos National Laboratory; Buttler, William T [Los Alamos National Laboratory; Oro, David M [Los Alamos National Laboratory; Rousculp, Christopher L [Los Alamos National Laboratory; Morris, Christopher [Los Alamos National Laboratory; Mariam, Fesseha G [Los Alamos National Laboratory

2011-01-31T23:59:59.000Z

118

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

Science Conference Proceedings (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 KAMS environment for up to 15 years. Aging and testing of softwood fiberboard will continue and additional data will be collected. 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. Further understanding of the variability of softwood fiberboard properties will require testing of additional material.

Daugherty, W.

2013-03-05T23:59:59.000Z

119

CH Packaging Operations Manual  

Science Conference Proceedings (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

120

Materials Transportation Testing & Analysis at Sandia National...  

NLE Websites -- All DOE Office Websites (Extended Search)

and 25 years of transportation experience and documented accident data by demonstrating RAM information, regulations, requirements, safety issues, emergency response, regulatory...

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

Underground Test Area Subproject Phase I Data Analysis Task. Volume IV - Hydrologic Parameter Data Documentation Package  

Science Conference Proceedings (OSTI)

Volume IV 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 hydrologic parameter data. 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

1996-09-01T23:59:59.000Z

122

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

Science Conference Proceedings (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

123

Planning and design of additional East Mesa Geothermal Test Facilities. Phase 1B. Volume II. Procurement package  

DOE Green Energy (OSTI)

Procurement packages of technical specifications and construction drawings for eleven test facility additions to the ERDA East Mesa Geothermal Component Test Facility are presented. Each of the specifications includes all of the technical requirements needed for procurement and construction starting with Division 2. The information is presented under the following subject headings: injection pump system: 52-2 injection pipeline; control and instrumentation spools; calibration test bench; test pad modifications; test pad piping headers; production and injection wells; well 5-2 modifications; well 8-1 down-hole pump; well 6-1 down-hole pump; and well 8-1 booster pump. (JGB)

Pearson, R.O.

1976-10-15T23:59:59.000Z

124

Review of SAR for Packaging Report  

Energy.gov (U.S. Department of Energy (DOE))

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.

125

Soy Package  

Science Conference Proceedings (OSTI)

Contains four (4) titles. Soy Package Food Science Health Nutrition Biochemistry Processing Soybeans Food Science & Technology Health - Nutrition - Biochemistry Processing Value Packages This Value Package includes:

126

Virtual Testing of Concrete Transport Properties  

Science Conference Proceedings (OSTI)

... Ph.D. in Civil Engineering from Texas A&M University ... of 10 %, a value intermediate between the ... for Testing and Materials, West Conshohocken, PA ...

2009-05-18T23:59:59.000Z

127

Transport Test Problems for Radiation Detection Scenarios  

Science Conference Proceedings (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

128

Testing of Biomass in a Transport Reactor Gasifier  

Science Conference Proceedings (OSTI)

A 200-hour gasification test was undertaken on biomass fuels from sources that include wood waste and a potential energy crop such as switchgrass. The test involved the design and construction of a feed system to allow 100% biomass to be continuously fed to the pilot-scale transport reactor development unit (TRDU) at the Energy & Environmental Research Center. Biomass performance was also assessed in a high-efficiency transport reactor gasifier, the centerpiece of an advanced biomass integrated ...

2012-11-28T23:59:59.000Z

129

Argonne Transportation - Advanced Powertrain Test Facility  

NLE Websites -- All DOE Office Websites (Extended Search)

Powertrain Test Cell A hybrid electric vehicle (HEV) has both an electric motor and a fuel-using device, such as a small gasoline engine. The two power sources can work together in...

130

Carbon Capture, Transport and Storage Regulatory Test Exercise: Output  

Open Energy Info (EERE)

Carbon Capture, Transport and Storage Regulatory Test Exercise: Output Carbon Capture, Transport and Storage Regulatory Test Exercise: Output Report 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: Market Analysis Website: cdn.globalccsinstitute.com/sites/default/files/publications/7326/carbo Equivalent URI: cleanenergysolutions.org/content/carbon-capture-transport-and-storage- Policies: Regulations Regulations: Emissions Mitigation Scheme The Scottish Government published this report to identify regulatory gaps or overlaps in the nation's framework for regulating carbon capture and storage (CCS). The report aims to streamline and better manage CCS regulation. It focuses on evaluating the risks, barriers, information gaps,

131

Demands placed on waste package performance testing and modeling by some general results on reliability analysis  

Science Conference Proceedings (OSTI)

Waste packages for a US nuclear waste repository are required to provide reasonable assurance of maintaining substantially complete containment of radionuclides for 300 to 1000 years after closure. The waiting time to failure for complex failure processes affecting engineered or manufactured systems is often found to be an exponentially-distributed random variable. Assuming that this simple distribution can be used to describe the behavior of a hypothetical single barrier waste package, calculations presented in this paper show that the mean time to failure (the only parameter needed to completely specify an exponential distribution) would have to be more than 10{sub 7} years in order to provide reasonable assurance of meeting this requirement. With two independent barriers, each would need to have a mean time to failure of only 10{sup 5} years to provide the same reliability. Other examples illustrate how multiple barriers can provide a strategy for not only achieving but demonstrating regulatory compliance.

Chesnut, D.A.

1991-09-01T23:59:59.000Z

132

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

Science Conference Proceedings (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

133

Reversible Bending Fatigue Test System for Investigating Vibration Integrity of Spent Nuclear Fuel during Transportation  

SciTech Connect

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading during road or rail shipment. Oak Ridge National Laboratory (ORNL) has been developing testing capabilities that can be used to improve the understanding of the impacts on SNF integrity due to vibration loading, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet the nuclear industry and U.S. Nuclear Regulatory Commission needs in the area of safety and security of spent nuclear fuel storage and transport operations. The ORNL developed test system can perform reversible-bending fatigue testing to evaluate both the static and dynamic mechanical response of SNF rods under simulated loads. The testing apparatus is also designed to meet the challenges of hot-cell operation, including remote installation and detachment of the SNF test specimen, in-situ test specimen deformation measurement, and implementation of a driving system suitable for use in a hot cell. The system contains a U-frame set-up equipped with uniquely designed grip rigs, to protect SNF rod and to ensure valid test results, and use of 3 specially designed LVDTs to obtain the in-situ curvature measurement. A variety of surrogate test rods have been used to develop and calibrate the test system as well as in performing a series of systematic cyclic fatigue tests. The surrogate rods include stainless steel (SS) cladding, SS cladding with cast epoxy, and SS cladding with alumina pellets inserts simulating fuel pellets. Testing to date has shown that the interface bonding between the SS cladding and the alumina pellets has a significant impact on the bending response of the test rods as well as their fatigue strength. The failure behaviors observed from tested surrogate rods provides a fundamental understanding of the underlying failure mechanisms of the SNF surrogate rod under vibration which has not been achieved previously. The newly developed device is scheduled to be installed in the hot-cell in summer 2013 to test high burnup SNF.

Wang, Jy-An John [ORNL; Wang, Hong [ORNL; Bevard, Bruce Balkcom [ORNL; Howard, Rob L [ORNL; Flanagan, Michelle [U.S. Nuclear Regulatory Commission

2013-01-01T23:59:59.000Z

134

THERMAL UPGRADING OF 9977 RADIOACTIVE MATERIAL (RAM) TYPE B PACKAGE  

Science Conference Proceedings (OSTI)

The 9977 package is a radioactive material package that was originally certified to ship Heat Sources and RTG contents up to 19 watts and it is now being reviewed to significantly expand its contents in support of additional DOE missions. Thermal upgrading will be accomplished by employing stacked 3013 containers, a 3013 aluminum spacer and an external aluminum sleeve for enhanced heat transfer. The 7th Addendum to the original 9977 package Safety Basis Report describing these modifications is under review for the DOE certification. The analyses described in this paper show that this well-designed and conservatively analyzed package can be upgraded to carry contents with decay heat up to 38 watts with some simple design modifications. The Model 9977 package has been designed as a replacement for the Department of Transportation (DOT) Fissile Specification 6M package. The 9977 package is a very versatile Type B package which is certified to transport and store a wide spectrum of radioactive materials. The package was analyzed quite conservatively to increase its usefulness and store different payload configurations. Its versatility is evident from several daughter packages such as the 9978 and H1700, and several addendums where the payloads have been modified to suit the Shipper's needs without additional testing.

Gupta, N.; Abramczyk, G.

2012-03-26T23:59:59.000Z

135

FFTF railroad tank car safety evaluation for packaging  

SciTech Connect

This Safety Evaluation for Packaging (SEP) provides evaluations necessary to approve transfer of the 8,000 gallon Liquid Waste Tank Car (LWTC) from the Fast Flux Test Facility (FFTF) to the 200 Areas. This SEP will demonstrate that the transfer cif the LWTC will provide an equivalent degree of safety as would be provided by packages meeting U.S. Department of Transportation (DOT) requirements. This fulfills onsite transportation requirements implemented in the Hazardous Material Packaging and Shipping, WHC-CM-2-14.

Romano, T.

1996-10-25T23:59:59.000Z

136

Biodiesel Package  

Science Conference Proceedings (OSTI)

A special collection of books and CD-ROMS on the topic of biodiesel. Biodiesel Package Biofuels and Bioproducts and Biodiesel Biofuels - Bioproducts Value Packages AOCS Press This Value Package includes: ...

137

Proceedings of the 6th Annual Meeting for Excess Weapons Plutonium Disposition: Plutonium Packaging, Storage and Transportation and WasteTreatment, Storage and Disposal Activities  

SciTech Connect

The sixth annual Excess Weapons Plutonium Disposition meeting organized by Lawrence Livermore National Laboratory (LLNL) was held November 15-17, 2004, at the State Education Center (SEC), 4 Aerodromnya Drive, St. Petersburg, Russia. The meeting discussed Excess Weapons Plutonium Disposition topics for which LLNL has the US Technical Lead Organization responsibilities. The technical areas discussed included Radioactive Waste Treatment, Storage, and Disposal, and Plutonium Oxide and Plutonium Metal Packaging, Storage and Transportation and Spent Fuel Packaging, Storage and Transportation. The meeting was conducted with a conference format using technical presentations of papers with simultaneous translation into English and Russian. There were 55 Russian attendees from 16 different Russian organizations and four non-Russian attendees from the US. Forty technical presentations were made. The meeting agenda is given in Appendix B and the attendance list is in Appendix C. The 16 different Russian design, industrial sites, and scientific organizations in attendance included staff from Rosatom/Minatom, Federal Nuclear and Radiation Safety Authority of Russia (GOSATOMNADZOR, NIERA/GAN), All Russian Designing & Scientific Research Institute of Complex Power Technology (VNIPIET), Khlopin Radium Institute (KRI), A. A. Bochvar All Russian Scientific Research Institute of Inorganic Materials (VNIINM), All Russian & Design Institute of Production Engineering (VNIPIPT), Ministry of Atomic Energy of Russian Federation Specialized State Designing Institute (GSPI), State Scientific Center Research Institute of Atomic Reactors (RIAR), Siberian Chemical Combine Tomsk (SCC), Mayak PO, Mining Chemical Combine (MCC K-26), Institute of Biophysics (IBPh), Sverdlosk Scientific Research Institute of Chemical Machine Building (SNIIChM), Kurchatov Institute (KI), Institute of Physical Chemistry Russian Academy of Science (IPCh RAS) and Radon PO-Moscow. The four non-Russian attendees included one representative from DOE NNSA, and LLNL, and two from Duratek, The meeting was organized into three major sessions: (1) Waste Treatment, Storage and Disposal; (2) Plutonium Packaging, Storage and Transportation; (3) Spent Fuel Packaging, Storage and Transportation. Twenty presentations were made on the topic of Waste Treatment, Storage and Disposal (Session II), ten presentations on Plutonium Packaging, Storage and Transportation (Session III), and four presentations on Spent Fuel Packaging, Storage and Transportation (Session IV). In addition, DOE/NNSA, Minatom/Rosatom and TVEL summarized the bases for the conference at the beginning of the meeting (Session I). Nine months had passed since the last LLNL contracts review meeting. During that time period, LLNL and TVEL have been able to sign six contracts for a total of $1,700,000 in the areas of: (1) Waste treatment, storage and disposal; and (2) Plutonium packaging, storage and transportation. The scope of several other work projects are now in various stages of development in these areas. It is anticipated that more contracts will be signed before the next meeting of this type. These events have allowed us to start work in our technical activities under new direction from TVEL, which is now the single Russian organization to coordinate and conclude contracts with LLNL. The meeting presentations and discussions have defined where we are and where we are going in the near term in regard to our joint interests in excess weapons plutonium disposition. Each topical section of this Proceedings is introduced by a summary of the presentations in that section.

Jardine, L J

2005-06-30T23:59:59.000Z

138

DOE - Safety of Radioactive Material Transportation  

NLE Websites -- All DOE Office Websites (Extended Search)

SAFE are radioactive material transportations packages? SAFE are radioactive material transportations packages? RAM PACKAGES TESTING & CERTIFICATION REGULATIONS & GUIDANCE SITE MAP This graphic was generated from a computer analysis and shows the results from a regulatory puncture test of a stainless steel packaging dropping 40 inches (10 MPH) onto a 6 inch diameter steel spike. U.S. DOE | Office of Civilian Radioactive Waste Management (OCRWM) Sandia National Laboratories | Nuclear Energy & Fuel Cucle Programs © Sandia Corporation | Site Contact | Sandia Site Map | Privacy and Security An internationally recognized web-site from PATRAM 2001 - the 13th International Symposium on the Packaging and Transportation of Radioactive Material. Recipient of the AOKI AWARD. PATRAM, sponsored by the U.S. Department of Energy in cooperation with the International Atomic Energy Agency brings government and industry leaders together to share information on innovations, developments, and lessons learned about radioactive materials packaging and transportation.

139

FINITE ELEMENT ANALYSIS OF BULK TRITIUM SHIPPING PACKAGE  

SciTech Connect

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

140

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

SciTech Connect

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

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 THE H1700 SHIPPING PACKAGE  

SciTech Connect

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

142

Packaging Research and Routing Optimization - Research Area - National  

NLE Websites -- All DOE Office Websites (Extended Search)

Packaging Research Packaging Research The Packaging Research Facility conducts research for DOE National Nuclear Security Administration, as well as for commercial customers. The facility specializes in testing packaging systems for the transportation of nuclear fuel. Once a package design has been successfully tested, it can then be certified by one of several regulatory authorities, usually DOE, the National Nuclear Security Administration, or the Nuclear Regulatory Commission, for use on the road. Welcome EESD Programs EES Directorate ORNL Web Contact Disclaimer Oak Ridge National Laboratory is a national multi-program research and development facility managed by UT-Battelle, LLC for the U.S. Department of Energy UT-Battelle, LLC U.S. Department of Energy Contact Scott Ludwig

143

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

SciTech Connect

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

144

Performance oriented packaging testing of nine Mk 3 Mod 0 signal containers in PPP-B-621 wood box for packing group II solid hazardous materials. Final report  

SciTech Connect

A PPP-B-621 wood box containing nine Mk 3 Mod 0 Signal containers was tested for conformance to Performance Oriented Packaging criteria established by Code of Federal Regulations Title 49 CFR. The container was tested with a gross weight of 123.3 pounds (56 kilograms) and met all requirements.

Libbert, K.J.

1992-10-01T23:59:59.000Z

145

Performance oriented packaging testing of the six-foot flexible linear shaped charge box for packing group II hazardous materials. Final report  

SciTech Connect

The wood box (Drawing 53711-6665109) for six-foot flexible linear shaped charges was tested for conformance to Performance Oriented Packaging standards specified by the Code of Federal Regulations, Title 49 CFR, Parts 107 through 178, dated 31 December 1991. The box was tested with a gross weight of 14 kilograms and met all the requirements.

Libbert, K.J.

1992-10-01T23:59:59.000Z

146

Transportation Organization and Functions  

Energy.gov (U.S. Department of Energy (DOE))

Office of Packaging and Transportation list of organizations and functions, with a list of acronyms.

147

Testing in support of transportation of residues in the pipe overpack container  

Science Conference Proceedings (OSTI)

The disposition of the large back-log of plutonium residues at the Rocky Flats Environmental Technology Site (Rocky Flats) will require interim storage and subsequent shipment to a waste repository. Current plants call for disposal at the Waste Isolation Pilot Plant (WIPP) and the transportation to WIPP in the TRUPACT-II. The transportation phase will require the residues to be packaged in a container that is more robust than a standard 55-gallon waste drum. Rocky Flats has designed the Pipe Overpack Container to meet this need. The tests described here were performed to qualify the Pipe Overpack Container as a waste container for shipment in the TRUPACT-II. Using a more robust container will assure the fissile materials in each container can not be mixed with the fissile material from the other containers and will provide criticality control. This will allow an increase in the payload of the TRUPACT-II from 325 fissile gram equivalents to 2,800 fissile gram equivalents.

Ammerman, D.J.; Bobbe, J.G.; Arviso, M.; Bronowski, D.R. [Sandia National Labs., Albuquerque, NM (United States). Transportation Systems Dept.

1997-04-01T23:59:59.000Z

148

The transportable heavy-duty engine emissions testing laboratory  

DOE Green Energy (OSTI)

West Virginia University has designed and constructed a Transportable Emissions Testing Laboratory for measuring emissions from heavy duty vehicles, such as buses and trucks operating on conventional and alternative fuels. The laboratory facility can be transported to a test site located at, or nearby, the home base of the vehicles to be tested. The laboratory has the capability of measuring vehicle emissions as the vehicle is operated under either transient or steady state loads and speeds. The exhaust emissions from the vehicle is sampled and the levels of the constituents of the emission are measured. The laboratory consists of two major units; a power absorber unit and an emissions measurement unit. A power absorber unit allows for the connection of a dynamic load to the drive train of the vehicle so that the vehicle can be driven'' through a test cycle while actually mounted on a stationary test bed. The emissions unit contains instrumentation and equipment which allows for the dilution of the vehicle's exhaust with air. The diluteed exhaust is sampled and analyzed to measure the level of concentration of those constituents which have been identified to have impact on the clean environment. Sampling probes withdraw diluted exhaust which is supplied to a number of different exhaust gas analysis instruments. The exhaust gas analysis instruments have the capability to measure the levels of the following exhaust gas constituents: carbon monoxide (CO), carbon dioxide (CO{sub 2}), oxides of nitrogen (NO{sub x}), unburned hydrocarbons (HC), formaldehyde (HCHO), methane and particulate matter. Additional instruments or sampling devices can be installed whenever measurements of additional constituents are desired. A computer based, data acquisition system is used to continuously monitor a wide range of parameters important to the operation of the test and to record the test results.

Not Available

1991-05-01T23:59:59.000Z

149

The transportable heavy-duty engine emissions testing laboratory  

SciTech Connect

West Virginia University has designed and constructed a Transportable Emissions Testing Laboratory for measuring emissions from heavy duty vehicles, such as buses and trucks operating on conventional and alternative fuels. The laboratory facility can be transported to a test site located at, or nearby, the home base of the vehicles to be tested. The laboratory has the capability of measuring vehicle emissions as the vehicle is operated under either transient or steady state loads and speeds. The exhaust emissions from the vehicle is sampled and the levels of the constituents of the emission are measured. The laboratory consists of two major units; a power absorber unit and an emissions measurement unit. A power absorber unit allows for the connection of a dynamic load to the drive train of the vehicle so that the vehicle can be driven'' through a test cycle while actually mounted on a stationary test bed. The emissions unit contains instrumentation and equipment which allows for the dilution of the vehicle's exhaust with air. The diluteed exhaust is sampled and analyzed to measure the level of concentration of those constituents which have been identified to have impact on the clean environment. Sampling probes withdraw diluted exhaust which is supplied to a number of different exhaust gas analysis instruments. The exhaust gas analysis instruments have the capability to measure the levels of the following exhaust gas constituents: carbon monoxide (CO), carbon dioxide (CO{sub 2}), oxides of nitrogen (NO{sub x}), unburned hydrocarbons (HC), formaldehyde (HCHO), methane and particulate matter. Additional instruments or sampling devices can be installed whenever measurements of additional constituents are desired. A computer based, data acquisition system is used to continuously monitor a wide range of parameters important to the operation of the test and to record the test results.

1991-05-01T23:59:59.000Z

150

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

Science Conference Proceedings (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. There is no clear trend thus far to indicate one material ages in a manner significantly different from the other material. Some softwood fiberboard properties degrade faster in some environments, while cane fiberboard degrades faster with regards to other properties and environments. Given the limited aging time accumulated to date in the elevated humidity environments, it is recommended that aging and testing of softwood fiberboard continue for another year. 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. Further understanding of the variability of softwood fiberboard properties will require testing of additional material.

Daugherty, W.

2011-12-13T23:59:59.000Z

151

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

SciTech Connect

Samples have been prepared from a softwood fiberboard lower subassembly. 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. On the positive side, the softwood fiberboard data to date shows less sample-to-sample variation in physical properties than cane fiberboard, and the thermal conductivity decreases at a slower rate at 250F for softwood fiberboard than for cane fiberboard. On the other hand, the softwood fiberboard physical property samples generally show degradation rates greater than cane fiberboard samples in the 185F 30%RH environment. Testing following additional conditioning will continue and the addition of samples in other elevated humidity environment(s) will be pursued to identify the extent of these trends. 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. Further understanding of the variability of softwood fiberboard properties will require testing of additional material.

Daugherty, W.

2010-12-27T23:59:59.000Z

152

Testing of Hydro-Quebec Cases with Dynamic Security Assessment Package  

Science Conference Proceedings (OSTI)

Transient stability contingencies provided by Hydro-Quebec (HQ) were evaluated by Siemens using two EPRI Dynamic Security Assessment (DSA) software components: Contingency Screening and Ranking and Time Domain Simulation. The main purpose of the HQ collaboration was to ascertain the performance of the EPRI on-line DSA system for the Hydro-Quebec network and test the contingency ranking methodology. During 1995 and 1996, comparing and matching transient stability modeling and ranking a list of HQ continge...

2000-12-11T23:59:59.000Z

153

AUTHORIZATION OF THE TROJAN REACTOR VESSEL PACKAGE FOR ONE-TIME SHIPMENT FOR DISPOSAL (1)  

E-Print Network (OSTI)

Request Commission approval, by negative consent, for the staff to grant two specific exemptions from package test requirements specified in 10 CFR Part 71 for the Trojan Reactor Vessel Package (TRVP), and to authorize the TRVP for one-time transport for disposal. BACKGROUND: Portland General Electric Company (PGE) has requested approval of the TRVP (including internals) for transport to the disposal facility operated by US

L. Joseph; Callan Executive; Director Operations; John R. Cook

1998-01-01T23:59:59.000Z

154

Vadose Zone Transport Field Study: Detailed Test Plan for Simulated Leak Tests  

Science Conference Proceedings (OSTI)

This report describes controlled transport experiments at well-instrumented field tests to be conducted during FY 2000 in support of DOE?s Vadose Zone Transport Field Study (VZTFS). The VZTFS supports the Groundwater/Vadose Zone Integration Project Science and Technology Initiative. The field tests will improve understanding of field-scale transport and lead to the development or identification of efficient and cost-effective characterization methods. These methods will capture the extent of contaminant plumes using existing steel-cased boreholes. Specific objectives are to 1) identify mechanisms controlling transport processes in soils typical of the hydrogeologic conditions of Hanford?s waste disposal sites; 2) reduce uncertainty in conceptual models; 3) develop a detailed and accurate data base of hydraulic and transport parameters for validation of three-dimensional numerical models; and 4) identify and evaluate advanced, cost-effective characterization methods with the potential to assess changing conditions in the vadose zone, particularly as surrogates of currently undetectable high-risk contaminants. Pacific Northwest National Laboratory (PNNL) manages the VZTFS for DOE.

Ward, Anderson L.; Gee, Glendon W.

2000-06-23T23:59:59.000Z

155

RADIOACTIVE MATERIAL PACKAGING TORQUE REQUIREMENTS COMPLIANCE  

Science Conference Proceedings (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

156

Vadose zone transport field study: Detailed test plan for simulated leak tests  

Science Conference Proceedings (OSTI)

The US Department of Energy (DOE) Groundwater/Vadose Zone Integration Project Science and Technology initiative was created in FY 1999 to reduce the uncertainty associated with vadose zone transport processes beneath waste sites at DOE's Hanford Site near Richland, Washington. This information is needed not only to evaluate the risks from transport, but also to support the adoption of measures for minimizing impacts to the groundwater and surrounding environment. The principal uncertainties in vadose zone transport are the current distribution of source contaminants and the natural heterogeneity of the soil in which the contaminants reside. Oversimplified conceptual models resulting from these uncertainties and limited use of hydrologic characterization and monitoring technologies have hampered the understanding contaminant migration through Hanford's vadose zone. Essential prerequisites for reducing vadose transport uncertainly include the development of accurate conceptual models and the development or adoption of monitoring techniques capable of delineating the current distributions of source contaminants and characterizing natural site heterogeneity. The Vadose Zone Transport Field Study (VZTFS) was conceived as part of the initiative to address the major uncertainties confronting vadose zone fate and transport predictions at the Hanford Site and to overcome the limitations of previous characterization attempts. Pacific Northwest National Laboratory (PNNL) is managing the VZTFS for DOE. The VZTFS will conduct field investigations that will improve the understanding of field-scale transport and lead to the development or identification of efficient and cost-effective characterization methods. Ideally, these methods will capture the extent of contaminant plumes using existing infrastructure (i.e., more than 1,300 steel-cased boreholes). The objectives of the VZTFS are to conduct controlled transport experiments at well-instrumented field sites at Hanford to: identify mechanisms controlling transport processes in soils typical of the hydrogeologic conditions of Hanford's waste disposal sites; reduce uncertainty in conceptual models; develop a detailed and accurate database of hydraulic and transport parameters for validation of three-dimensional numerical models; identify and evaluate advanced, cost-effective characterization methods with the potential to assess changing conditions in the vadose zone, particularly as surrogates of currently undetectable high-risk contaminants. This plan provides details for conducting field tests during FY 2000 to accomplish these objectives. Details of additional testing during FY 2001 and FY 2002 will be developed as part of the work planning process implemented by the Integration Project.

AL Ward; GW Gee

2000-06-23T23:59:59.000Z

157

Type B Drum packages  

Science Conference Proceedings (OSTI)

The Type B Drum package is a container in which a single drum containing Type B quantities of radioactive material will be packaged for shipment. The Type B Drum containers are being developed to fill a void in the packaging and transportation capabilities of the US Department of Energy (DOE), as no double containment packaging for single drums of Type B radioactive material is currently available. Several multiple-drum containers and shielded casks presently exist. However, the size and weight of these containers present multiple operational challenges for single-drum shipments. The Type B Drum containers will offer one unshielded version and, if needed, two shielded versions, and will provide for the option of either single or double containment. The primary users of the Type B Drum container will be any organization with a need to ship single drums of Type B radioactive material. Those users include laboratories, waste retrieval facilities, emergency response teams, and small facilities.

Edwards, W.S.

1995-11-01T23:59:59.000Z

158

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

Science Conference Proceedings (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

159

Automated Transportation Management System (ATMS) | Department...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Waste Management Packaging and Transportation Automated Transportation Management System (ATMS) Automated Transportation Management System (ATMS) The Department of Energy's...

160

Transportation Emergency Preparedness Program (TEPP) | Department...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Packaging and Transportation Transportation Emergency Preparedness Program (TEPP) Transportation Emergency Preparedness Program (TEPP) In an effort to address responder...

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

Crush Testing at Oak Ridge National Laboratory  

SciTech Connect

The dynamic crush test is required in the certification testing of some small Type B transportation packages. International Atomic Energy Agency regulations state that the test article must be 'subjected to a dynamic crush test by positioning the specimen on the target so as to suffer maximum damage.' Oak Ridge National Laboratory (ORNL) Transportation Technologies Group performs testing of Type B transportation packages, including the crush test, at the National Transportation Research Center in Knoxville, Tennessee (United States). This paper documents ORNL's experiences performing crush tests on several different Type B packages. ORNL has crush tested five different drum-type package designs, continuing its 60 year history of RAM package testing. A total of 26 crush tests have been performed in a wide variety of package orientations and crush plate CG alignments. In all cases, the deformation of the outer drum created by the crush test was significantly greater than the deformation damage caused by the 9 m drop test. The crush test is a highly effective means for testing structural soundness of smaller nondense Type B shipping package designs. Further regulatory guidance could alleviate the need to perform the crush test in a wide range of orientations and crush plate CG alignments.

Feldman, Matthew R [ORNL

2011-01-01T23:59:59.000Z

162

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

163

The Packaging Handbook -- A guide to package design  

Science Conference Proceedings (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

164

CH Packaging Maintenance Manual  

SciTech Connect

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

165

Lipid Analysis Package  

Science Conference Proceedings (OSTI)

A Lipid Value package of 5 different books. Lipid Analysis Package Methods and Analyses Methods - Analyses Books Value Packages Methods - Analyses Books Methods This Value Package includes:   ...

166

Materials Transportation Testing & Analysis at Sandia National Laboratories  

NLE Websites -- All DOE Office Websites (Extended Search)

RMIR (Radioactive Materials Incident Report) Database Transportation RMIR (Radioactive Materials Incident Report) Database Transportation Accident and Incident Experience,1971-1999 Access Hazardous Materials Information System (HMIS) the primary source of national data for the Federal, state, and local governmental agencies responsible for the safety of hazardous materials transportation. Rail Transport Highway Transport Air Transport The Radioactive Material Incident Report (RMIR) Database was developed in 1981 at the Transportation Technology Center of Sandia National Laboratories (SNL) to support its research and development activities for the U.S. Department of Energy (DOE). This database contains information about radioactive materials transportation incidents that have occurred in the U.S. from 1971 through 1999. These data were drawn from the U.S.

167

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

SciTech Connect

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

168

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

Science Conference Proceedings (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

169

LONG TERM AGING AND SURVEILLANCE OF 9975 PACKAGE COMPONENTS  

Science Conference Proceedings (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

170

Transportable Heavy Duty Emissions Testing Laboratory and Research Program  

DOE Green Energy (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

171

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

Science Conference Proceedings (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

172

DOE - Safety of Radioactive Material Transportation  

NLE Websites -- All DOE Office Websites (Extended Search)

What are full-scale tests? What are scale-model tests? What is computer analysis? What are examples of severe testing? How do the certification tests compare to real-life accidents? Demonstrating target hardness. A packaging is certified when it can survive a sequence of impact, crush, puncture, fire, and immersion tests designed to replicate transport accident conditions. Type B Packages must meet the testing requirements of: Compliance Testing, as defined in 10 CFR Part 71.85 and 10 CFR Part 71.87 Normal Conditions of Transport, Ten tests as defined in 10 CFR Part 71.71 Hypothetical Accident Conditions, Six tests as defined in 10 CFR Part 71.73 The ability of radioactive material packages to withstand testing environments can be demonstrated by full-scale testing, scale-model

173

WIPP Transportation  

NLE Websites -- All DOE Office Websites (Extended Search)

Transuranic Waste Transportation Container Documents Documents related to transuranic waste containers and packages. CBFO Tribal Program Information about WIPP shipments across...

174

DEVELOPMENT OF THE BULK TRITIUM SHIPPING PACKAGING  

SciTech Connect

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

175

Strategy for experimental validation of waste package performance assessment  

SciTech Connect

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

176

Reactive transport modeling of leaching tests and long-term ...  

Test-case B = macro-fractures Test-case C = micro-crack network + dual porosity @ANDRA. Cementitious materials workshop, Dec. 12-14 (2006), SRNL (USA) ...

177

Evaluation and compilation of DOE [Department of Energy] waste package test data; Biannual report, February 1988--July 1988  

Science Conference Proceedings (OSTI)

This report summarizes evaluations by the National Institute of Standards and Technology (NIST) of Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW) for the six month period February 1988 through July 1988. Activities for the DOE Materials Characterization Center are reviewed for the period January 1988 through June 1988. A summary is given of the Yucca Mountain, Nevada disposal site activities. Short discussions relating to the reviewed publications are given and complete reviews and evaluations are included. 20 refs., 1 fig., 1 tab.

Interrante, C.; Escalante, E.; Fraker, A.; Plante, E.

1989-10-01T23:59:59.000Z

178

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

Science Conference Proceedings (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

179

Evaluation and compilation of DOE waste package test data; Biannual report, August 1988--January 1989: Volume 6  

Science Conference Proceedings (OSTI)

This report summarizes evaluations by the National Institute of Standards and Technology (NIST) of Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW) for the six-month period August 1988 through January 1989. Included are reviews of related materials research and plans, activities for the DOE Materials Characterization Center, information on the Yucca Mountain Project, and other information regarding supporting research and special assistance. NIST comments are given on the Yucca Mountain Consultation Draft Site Characterization Plan (CDSCP) and on the Waste Compliance Plan for the West Valley Demonstration Project (WVDP) High-Level Waste (HLW) Form. 3 figs.

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

1990-11-01T23:59:59.000Z

180

Functional Foods Package  

Science Conference Proceedings (OSTI)

Contains five (5) titles regarding functional foods. Functional Foods Package Health - Nutrition - Biochemistry Value Packages Nutrition Health Food Science Biochemistry This Value Package includes: ...

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

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

Science Conference Proceedings (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

182

AGING PERFORMANCE OF VITON GLT O-RINGS IN RADIOACTIVE MATERIAL PACKAGES  

Science Conference Proceedings (OSTI)

Radioactive material packages used for transportation of plutonium-bearing materials often contain multiple O-ring seals for containment. Packages such as the Model 9975 are also being used for interim storage of Pu-bearing materials at the Savannah River Site (SRS). One of the seal materials used in such packages is Viton{reg_sign} GLT fluoroelastomer. The aging behavior of containment vessel O-rings based on Viton{reg_sign} GLT at long-term containment term storage conditions is being characterized to assess its performance in such applications. This paper summarizes the program and test results to date.

Skidmore, E; Kerry Dunn, K; Elizabeth Hoffman, E; Elise Fox, E; Kathryn Counts, K

2007-05-07T23:59:59.000Z

183

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

Open Energy Info (EERE)

Test Exercise: Output Report Focus Area: Clean Fossil Energy Topics: Market Analysis Website: cdn.globalccsinstitute.comsitesdefaultfilespublications7326carbo...

184

Evaluation and compilation of DOE waste package test data; Biannual report, February 1989--July 1989: Volume 7  

Science Conference Proceedings (OSTI)

This report summarizes evaluations by the National Institute of Standards and Technology (NIST) of Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW) for the six-month period, February through July 1989. 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. Outlines for planned interpretative reports on the topics of aqueous corrosion of copper, mechanisms of stress corrosion cracking and internal failure modes of Zircaloy cladding are included. For the publications reviewed during this reporting period, short discussions are given to supplement the completed reviews and evaluations. Included in this report is an overall review of a 1984 report on glass leaching mechanisms, as well as reviews for each of the seven chapters of this report.

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 (IMSE), Gaithersburg, MD (United States). Metallurgy Div.

1991-12-01T23:59:59.000Z

185

Packaging and Labeling  

Science Conference Proceedings (OSTI)

Packaging and Labeling. The Fair Packaging and Labeling Act (FPLA) and other Federal laws and regulations govern the ...

2013-05-17T23:59:59.000Z

186

Argonne Transportation Technology R&D Center - Battery Test Facility...  

NLE Websites -- All DOE Office Websites (Extended Search)

Research and Analysis Computing Center Working With Argonne Contact TTRDC Battery Test Facility Argonne researcher Lee Walker Argonne researcher Lee Walker examines a...

187

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

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

188

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

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

189

DOE - Safety of Radioactive Material Transportation  

NLE Websites -- All DOE Office Websites (Extended Search)

What are examples of severe testing? What are examples of severe testing? How do the certification tests compare to real-life accidents? Demonstrating target hardness. These full-scale tests, conducted at Sandia National Laboratories Transportation Programs, demonstrate how spent fuel casks perform in accident environments that are more similar to what may happen during actual shipments. Each of the tests included the transportation vehicle as well as the cask. The damage to the casks from these tests was less than the damage during the regulatory hypothetical accident tests, demonstrating that the regulatory tests are more severe. DESCRIPTION PHOTO DURING TEST PHOTO AFTER TEST PHOTO OF PACKAGE AFTER TEST VIDEO OF TEST CRASH TEST Cask rail car with a 74 ton Type B Package on it crashing into a 690 ton concrete block at 81 miles per hour [photo]

190

Hot-Gas Filter Testing with a Transport Reactor Gasifier  

Science Conference Proceedings (OSTI)

Today, coal supplies over 55% of the electricity consumed in the United States and will continue to do so well into the next century. One of the technologies being developed for advanced electric power generation is an integrated gasification combined cycle (IGCC) system that converts coal to a combustible gas, cleans the gas of pollutants, and combusts the gas in a gas turbine to generate electricity. The hot exhaust from the gas turbine is used to produce steam to generate more electricity from a steam turbine cycle. The utilization of advanced hot-gas particulate and sulfur control technologies together with the combined power generation cycles make IGCC one of the cleanest and most efficient ways available to generate electric power from coal. One of the strategic objectives for U.S. Department of Energy (DOE) IGCC research and development program is to develop and demonstrate advanced gasifiers and second-generation IGCC systems. Another objective is to develop advanced hot-gas cleanup and trace contaminant control technologies. One of the more recent gasification concepts to be investigated is that of the transport reactor gasifier, which functions as a circulating fluid-bed gasifier while operating in the pneumatic transport regime of solid particle flow. This gasifier concept provides excellent solid-gas contacting of relatively small particles to promote high gasification rates and also provides the highest coal throughput per unit cross-sectional area of any other gasifier, thereby reducing capital cost of the gasification island.

Swanson, M.L.; Hajicek, D.R.

2002-09-18T23:59:59.000Z

191

Effects of simulant Hanford tank waste on plastic packaging components  

Science Conference Proceedings (OSTI)

In this paper, the authors describe a chemical compatibility testing program for packaging components which might be used to transport mixed wastes. They mention the results of the screening phase of this program and then present the results of the second phase of this experimental program. This effort involved the comprehensive testing of five plastic liner materials in the aqueous mixed waste simulant. The testing protocol involved exposing the respective materials to {approximately} 140, 290, 570, and 3,670 krads of gamma radiation followed by 7, 14, 28, 180 day exposures to the waste simulant at 18, 50, and 60 C. From the data analysis performed to date in this study, they have identified the fluorocarbon Kel-F{trademark} as having the greatest chemical compatibility after being exposed to gamma radiation followed by exposure to the Hanford Tank simulant mixed waste. The most striking observation from this study was the poor performance of Teflon under these conditions. The data obtained from this testing program will be available to packaging designers for the development of mixed waste packagings. The implications of the testing results on the selection of appropriate materials as packaging components are discussed.

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

1996-07-01T23:59:59.000Z

192

Handling and Packaging a Potentially Radiologically Contaminated Patient |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Handling and Packaging a Potentially Radiologically Contaminated Handling and Packaging a Potentially Radiologically Contaminated Patient Handling and Packaging a Potentially Radiologically Contaminated Patient The purpose of this procedure is to provide guidance to EMS care providers for properly handling and packaging potentially radiologically contaminated patients. This procedure applies to Emergency Medical Service care providers who respond to a radioactive material transportation incident that involves potentially contaminated injuries. Handling and Packaging a Potentially Radiologically Contaminated Patient.docx More Documents & Publications Pre-Hospital Practices for Handling a Radiologically Contaminated Patient Emergency Response to a Transportation Accident Involving Radioactive Material Radioactive Materials Transportation and Incident Response

193

In-Package Chemistry Abstraction  

SciTech Connect

This report was developed in accordance with the requirements in ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The purpose of the in-package chemistry model is to predict the bulk chemistry inside of a breached waste package and to provide simplified expressions of that chemistry as function of time after breach to Total Systems Performance Assessment for the License Application (TSPA-LA). The scope of this report is to describe the development and validation of the in-package chemistry model. The in-package model is a combination of two models, a batch reactor model that uses the EQ3/6 geochemistry-modeling tool, and a surface complexation model that is applied to the results of the batch reactor model. The batch reactor model considers chemical interactions of water with the waste package materials and the waste form for commercial spent nuclear fuel (CSNF) waste packages and codisposed waste packages that contain both high-level waste glass (HLWG) and DOE spent fuel. The surface complexation model includes the impact of fluid-surface interactions (i.e., surface complexation) on the resulting fluid composition. The model examines two types of water influx: (1) the condensation of water vapor that diffuses into the waste package, and (2) seepage water that enters the waste package from the drift as a liquid. (1) Vapor Influx Case: The condensation of vapor onto the waste package internals is simulated as pure H2O and enters at a rate determined by the water vapor pressure for representative temperature and relative humidity conditions. (2) Water Influx Case: The water entering a waste package from the drift is simulated as typical groundwater and enters at a rate determined by the amount of seepage available to flow through openings in a breached waste package. TSPA-LA uses the vapor influx case for the nominal scenario for simulations where the waste package has been breached but the drip shield remains intact, so all of the seepage flow is diverted from the waste package. The chemistry from the vapor influx case is used to determine the stability of colloids and the solubility of radionuclides available for transport by diffusion, and to determine the degradation rates for the waste forms. TSPA-LA uses the water influx case for the seismic scenario, where the waste package has been breached and the drip shield has been damaged such that seepage flow is actually directed into the waste package. The chemistry from the water influx case that is a function of the flow rate is used to determine the stability of colloids and the solubility of radionuclides available for transport by diffusion and advection, and to determine the degradation rates for the CSNF and HLW glass. TSPA-LA does not use this model for the igneous scenario. Outputs from the in-package chemistry model implemented inside TSPA-LA include pH, ionic strength, and total carbonate concentration. These inputs to TSPA-LA will be linked to the following principle factors: dissolution rates of the CSNF and HLWG, dissolved concentrations of radionuclides, and colloid generation.

E. Thomas

2004-11-09T23:59:59.000Z

194

Innovative Approaches to Large Component Packaging  

Science Conference Proceedings (OSTI)

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

Freitag, A.; Hooper, M.; Posivak, E.; Sullivan, J. [WMG, Inc., Peekskill, NY 10566 (United States)

2006-07-01T23:59:59.000Z

195

Safety evaluation for packaging two plywood boxes  

Science Conference Proceedings (OSTI)

This safety evaluation for packaging evaluates and documents the ability of the plywood boxes listed below to meet the packaging requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for the onsite transfer of Type B radioactive material. Onsite transfer is the transport of hazardous materials on controlled routes confined to established limited areas and to portions of federally owned roadways to which public access is prohibited during transfer. The plywood boxes being used for this transport are PIN number PNLD-95-322 and PNLD-95-385. The contents being transported are wood, plastic, piping, rubber, and gloves. The source term was determined by nondestructive analysis and obtained from the solid waste storage/disposal record. Before the nondestructive analysis, the intention was to transport the boxes under WHC-SD-TP-SEP-020, Safety Evaluation for Packaging (Onsite) Plywood Box (WHC 1994), but Type B shipments are not included.

Flanagan, B.D.

1996-09-26T23:59:59.000Z

196

ElectronicPackaging  

NLE Websites -- All DOE Office Websites (Extended Search)

Packaging Packaging Manufacturing Technologies The Electronic Packaging technologies in the Thin Film, Vacuum, and Packaging Department are a resource for all aspects of microelectronic packag- ing. From design and layout to fabrication of proto- type samples, the staff offers partners the opportu- nity for concurrent engineering and development of a variety of electronic packaging concepts. This includes assistance in selecting the most appropri- ate technology for manufacturing, analysis of per- formance characteristics and development of new and unique processes. Capabilities 1. Network Fabrication * Low Temperature Co-Fired Ceramic (LTCC) * Thick Film * Thin Film 2. Packaging and Assembly * Chip Level Packaging * MEMs Packaging * Hermetic Sealing * Surface Mount Technology

197

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

SciTech Connect

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.

2013-08-01T23:59:59.000Z

198

Packaging Review Guide for Reviewing Safety Analysis Reports for Packagings  

SciTech Connect

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

199

Trans Labeling Package  

Science Conference Proceedings (OSTI)

A special collection of books and CD-ROMS on the topic of trans fat. Trans Labeling Package Health Nutrition Biochemistry Trans Health - Nutrition - Biochemistry Value Packages This Value Package includes: ...

200

Transportation  

NLE Websites -- All DOE Office Websites (Extended Search)

Transportation banner Home Agenda Awards Exhibitors Lodging Posters Registration T-Shirt Contest Transportation Workshops Contact Us User Meeting Archives Users' Executive...

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

Transportation  

NLE Websites -- All DOE Office Websites (Extended Search)

Transportation Print banner Home Agenda Awards Exhibitors Lodging Posters Registration T-Shirt Contest Transportation Workshops Contact Us User Meeting Archives Users' Executive...

202

Transportation  

NLE Websites -- All DOE Office Websites (Extended Search)

Links Transportation and Air Quality Transportation Energy Policy Analysis Batteries and Fuel Cells Buildings Energy Efficiency Electricity Grid Energy Analysis Appliance Energy...

203

Assessment of hydrologic transport of radionuclides from the Rulison Underground Nuclear Test Site, Colorado  

SciTech Connect

The U.S. Department of Energy (DOE) is operating an environmental restoration program to characterize, remediate, and close non-Nevada Test Site locations that were used for nuclear testing. Evaluation of radionuclide transport by groundwater from these sites is an important part of the preliminary risk analysis. These evaluations are undertaken to allow prioritization of the test areas in terms of risk, provide a quantitative basis for discussions with regulators and the public about future work at the sites, and provide a framework for assessing data needs to be filled by site characterization. The Rulison site in west-central Colorado was the location of an underground detonation of a 40-kiloton nuclear device in 1969. The test took place 2,568 m below ground surface in the Mesaverde Formation. Though located below the regional water table, none of the bedrock formations at the site yielded water during hydraulic tests, indicating extremely low permeability conditions. The scenario evaluated was the migration of radionuclides from the blast-created cavity through the Mesaverde Formation. Transport calculations were performed using the solute flux method, with input based on the limited data available for the site. Model results suggest that radionuclides from the test are contained entirely within the area currently administered by DOE. The transport calculations are most sensitive to changes in the mean groundwater velocity and the correlation scale of hydraulic conductivity, with transport of strontium and cesium also sensitive to the sorption coefficient.

Earman, S.; Chapman, J.; Andricevic, R.

1996-09-01T23:59:59.000Z

204

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

Science Conference Proceedings (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

205

Omega-3 Package  

Science Conference Proceedings (OSTI)

Contain two(2) titles. Omega-3 Package Health Nutrition Biochemistry Health - Nutrition - Biochemistry Value Packages Fish, Omega-3 and Human Health ...

206

DOE - Safety of Radioactive Material Transportation  

NLE Websites -- All DOE Office Websites (Extended Search)

Who makes rules? What are the requirements? Safety Record USERS OF PACKAGINGS CARRIER PACKAGE TYPE Hospitals and their suppliers common carrier Type A Industrial radiography companies private carrier Type B Soil testing laboratories private carrier Type B Food irradiators contract carrier Type B Medical supply sterilizers contract carrier Type B Academic research institutes common & contract carrier all types Nuclear energy fuel cycle facilities common & contract carrier all types Nuclear weapons complex contract & government carrier all types An agency or company that wants to ship RAM (shipper) often makes arrangements with a common or contract carrier or (where appropriate) a private carrier may transport the material. Packagings may be procured or

207

RH Packaging Program Guidance  

Science Conference Proceedings (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

208

Transportation  

NLE Websites -- All DOE Office Websites (Extended Search)

Transportation Transportation Transportation of Depleted Uranium Materials in Support of the Depleted Uranium Hexafluoride Conversion Program Issues associated with transport of depleted UF6 cylinders and conversion products. Conversion Plan Transportation Requirements The DOE has prepared two Environmental Impact Statements (EISs) for the proposal to build and operate depleted uranium hexafluoride (UF6) conversion facilities at its Portsmouth and Paducah gaseous diffusion plant sites, pursuant to the National Environmental Policy Act (NEPA). The proposed action calls for transporting the cylinder at ETTP to Portsmouth for conversion. The transportation of depleted UF6 cylinders and of the depleted uranium conversion products following conversion was addressed in the EISs.

209

Assessment of hydrologic transport of radionuclides from the Rio Blanco underground nuclear test site, Colorado  

SciTech Connect

DOE is operating an environmental restoration program to characterize, remediate, and close non-Nevada Test Site locations used for nuclear testing. Evaluation of radionuclide transport by groundwater is part of preliminary risk analysis. These evaluations allow prioritization of test areas in terms of risk, provide a basis for discussions with regulators and the public about future work, and provide a framework for assessing site characterization data needs. The Rio Blanco site in Colorado was the location of the simultaneous detonation of three 30-kiloton nuclear devices. The devices were located 1780, 1899, and 2039 below ground surface in the Fort Union and Mesaverde formations. Although all the bedrock formations at the site are thought to contain water, those below the Green River Formation (below 1000 in depth) are also gas-bearing, and have very low permeabilities. The transport scenario evaluated was the migration of radionuclides from the blast-created cavity through the Fort Union Formation. Transport calculations were performed using the solute flux method, with input based on the limited data available for the site. Model results suggest that radionuclides from the test are contained entirely within the area currently administered by DOE. This modeling was performed to investigate how the uncertainty in various physical parameters affect radionuclide transport at the site, and to serve as a starting point for discussion regarding further investigation; it was not intended to be a definitive simulation of migration pathways or radionuclide concentration values. Given the sparse data, the modeling results may differ significantly from reality. Confidence in transport predictions can be increased by obtaining more site data, including the amount of radionuclides which would have been available for transport (i.e., not trapped in melt glass or vented during gas flow testing), and the hydraulic properties of the formation. 38 refs., 6 figs., 1 tab.

Chapman, J.; Earman, S.; Andricevic, R.

1996-10-01T23:59:59.000Z

210

Transportation  

NLE Websites -- All DOE Office Websites (Extended Search)

Health Risks » Transportation Health Risks » Transportation DUF6 Health Risks line line Accidents Storage Conversion Manufacturing Disposal Transportation Transportation A discussion of health risks associated with transport of depleted UF6. Transport Regulations and Requirements In the future, it is likely that depleted uranium hexafluoride cylinders will be transported to a conversion facility. For example, it is currently anticipated that the cylinders at the ETTP Site in Oak Ridge, TN, will be transported to the Portsmouth Site, OH, for conversion. Uranium hexafluoride has been shipped safely in the United States for over 40 years by both truck and rail. Shipments of depleted UF6 would be made in accordance with all applicable transportation regulations. Shipment of depleted UF6 is regulated by the

211

Safety analysis report for packaging (onsite) steel drum  

SciTech Connect

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

212

DOE - Safety of Radioactive Material Transportation  

NLE Websites -- All DOE Office Websites (Extended Search)

When are they used? How are they moved? What's their construction? Who uses them? Who makes rules? What are the requirements? Safety Record A radioactive material (RAM) packaging is a container that is used to safely transport radioactive material from one location to another. In RAM transportation the container alone is called the Packaging. The packaging together with its contents is called the Package. Basic types of radioactive material packagings are: Excepted Packaging Industrial Packaging Type A Packaging Type B Packaging [EXCEPTED] Click to view picture [IP] Click to view picture [TYPE A] Click to view picture [TYPE B] Click to view picture Excepted Packagings are designed to survive normal conditions of transport. Excepted packagings are used for transportation of materials that are either Low Specific Activity (LSA) or Surface Contaminated Objects (SCO) and that are limited quantity shipments, instruments or articles, articles manufactured from natural or depleted uranium or natural thorium; empty packagings are also excepted (49CFR 173.421-428).

213

? Thermal Packaging  

E-Print Network (OSTI)

? Final devices will be lab bench and field tested for reliability. Recent Results ? Fuel delivery system fabricated ? 500um thick NiFe structures plated ? 900um Rotor/Housing fabrication process improved Next Six Months ? Fuel / air mixture manifold

Dorian Liepmann Ph. D; David C. Walther; Prof Al Pisano

2003-01-01T23:59:59.000Z

214

CRAD, Packaging and Transfer of Hazardous Materials and Materials of  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Packaging and Transfer of Hazardous Materials and Materials Packaging and Transfer of Hazardous Materials and Materials of National Security Interest Assessment Plan CRAD, Packaging and Transfer of Hazardous Materials and Materials of National Security Interest Assessment Plan Performance Objective: Verify that packaging and transportation safety requirements of hazardous materials and materials of national security interest have been established and are in compliance with DOE Orders 461.1 and 460.1B Criteria: Verify that safety requirements for the proper packaging and transportation of DOE/NNSA offsite shipments and onsite transfers of hazardous materials and for modal transport have been established [DOE O 460.1B, 1, "Objectives"]. Verify that the contractor transporting a package of hazardous materials is in compliance with the requirements of the Hazardous Materials

215

Shippers. general requirements for shipments and packagings. transportation of methane in specification DOT 3AX, 3AAX, and 3T cylinders. an extension of time to file comments  

SciTech Connect

An extension of time to file comments on regulations for transportation of methane in specification DOT 3AX, 3AAX, and 3T cylinders is issued by the U.S. Materials Transportation Bureau. The time for filing comments is extended from 5/3/78 to 7/17/78.

1978-05-18T23:59:59.000Z

216

Transportation  

Science Conference Proceedings (OSTI)

Transportation systems are an often overlooked critical infrastructure component. These systems comprise a widely diverse elements whose operation impact all aspects of society today. This chapter introduces the key transportation sectors and illustrates ...

Mark Hartong; Rajn Goel; Duminda Wijesekera

2012-01-01T23:59:59.000Z

217

Safety Evaluation for Packaging 101-SY Hydrogen Mitigation Mixer Pump package  

DOE Green Energy (OSTI)

This Safety Evaluation for Packaging (SEP) provides analysis and considered necessary to approve a one-time transfer of the 101-SY Hydrogen Mitigation Mixer Pump (HMMP). This SEP will demonstrate that the transfer of the HMMP in a new shipping container will provide an equivalent degree of safety as would be provided by packages meeting US Department of Transportation (DOT)/US Nuclear Regulatory Commission (NRC) requirements. This fulfills onsite, transportation requirements implemented by WHC-CM-2-14.

Carlstrom, R.F.

1994-10-05T23:59:59.000Z

218

WIPP transportation exercise to test emergency response capablities for Midland-Odessa  

NLE Websites -- All DOE Office Websites (Extended Search)

Transportation Exercise to Test Transportation Exercise to Test Emergency Response Capabilities for Midland-Odessa CARLSBAD, N.M., January 10, 2000 - Emergency response agencies from Midland and Odessa, Texas, will take part in a 1 p.m. (CST) training exercise Jan. 12 at the Ector County Coliseum. The graded exercise will help agencies determine whether emergency personnel are prepared to respond to a possible accident involving a shipment of transuranic radioactive waste headed for the U.S. Department of Energy's (DOE) Waste Isolation Pilot Plant (WIPP). "This is an excellent opportunity for emergency responders to test the skills they've learned," said Dale Childers, assistant chief of the Odessa Fire Department and emergency management coordinator for Ector County. "It will also help us determine what improvements,

219

Spent Nuclear Fuel Transportation: An Overview  

Science Conference Proceedings (OSTI)

Spent nuclear fuel comprises a fraction of the hazardous materials packages shipped annually in the United States. In fact, at the present time, fewer than 100 packages of spent nuclear fuel are shipped annually. At the onset of spent fuel shipments to the proposed Yucca Mountain, Nevada, repository, the U.S. Department of Energy (DOE) expects to ship 400 - 500 spent fuel transport casks per year over the life of the facility. This study summarizes work on transportation cask design and testing, regulato...

2004-02-18T23:59:59.000Z

220

Surveillance Guides - PTS 13.2 Packaging and Preparation for Shipment  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

PACKAGING AND PREPARATION FOR SHIPMENT PACKAGING AND PREPARATION FOR SHIPMENT 1.0 Objective The objective of this surveillance is to evaluate the effectiveness of the contractor's programs for packaging radioactive and hazardous wastes for shipment. The Facility Representative examines packages ready for shipment, observes preparation of packages, and reviews documents that establish the acceptability of packages. The Facility Representative verifies compliance with DOE requirements including requirements established by the Department of Transportation and the U.S. Nuclear Regulatory Commission. 2.0 References 2.1 DOE 5480.3, Safety Requirements for the Packaging and Transportation of Hazardous Materials, Hazardous Substances, and Hazardous Wastes

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

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

Science Conference Proceedings (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

222

terms Package — FiPy 3.0.1-dev157-g518df83 documentation  

Science Conference Proceedings (OSTI)

... steppers Package. Next topic. tests Package. This Page. Show Source. Quick search. Enter search terms or a module, class or function name. Contact ...

2013-08-30T23:59:59.000Z

223

Relevance of biotic pathways to the long-term regulation of nuclear waste disposal. Estimation of radiation dose to man resulting from biotic transport: the BIOPORT/MAXI1 software package. Volume 5  

Science Conference Proceedings (OSTI)

BIOPORT/MAXI1 is a collection of five computer codes designed to estimate the potential magnitude of the radiation dose to man resulting from biotic transport processes. Dose to man is calculated for ingestion of agricultural crops grown in contaminated soil, inhalation of resuspended radionuclides, and direct exposure to penetrating radiation resulting from the radionuclide concentrations established in the available soil surface by the biotic transport model. This document is designed as both an instructional and reference document for the BIOPORT/MAXI1 computer software package and has been written for two major audiences. The first audience includes persons concerned with the mathematical models of biological transport of commercial low-level radioactive wastes and the computer algorithms used to implement those models. The second audience includes persons concerned with exercising the computer program and exposure scenarios to obtain results for specific applications. The report contains sections describing the mathematical models, user operation of the computer programs, and program structure. Input and output for five sample problems are included. In addition, listings of the computer programs, data libraries, and dose conversion factors are provided in appendices.

McKenzie, D.H.; Cadwell, L.L.; Gano, K.A.; Kennedy, W.E. Jr.; Napier, B.A.; Peloquin, R.A.; Prohammer, L.A.; Simmons, M.A.

1985-10-01T23:59:59.000Z

224

ALTERNATE MATERIALS IN DESIGN OF RADIOACTIVE MATERIAL PACKAGES  

SciTech Connect

This paper presents a summary of design and testing of material and composites for use in radioactive material packages. These materials provide thermal protection and provide structural integrity and energy absorption to the package during normal and hypothetical accident condition events as required by Title 10 Part 71 of the Code of Federal Regulations. Testing of packages comprising these materials is summarized.

Blanton, P.; Eberl, K.

2010-07-09T23:59:59.000Z

225

The reduction of packaging waste  

Science Conference Proceedings (OSTI)

Nationwide, packaging waste comprises approximately one-third of the waste disposed in sanitary landfills. the US Department of Energy (DOE) generated close to 90,000 metric tons of sanitary waste. With roughly one-third of that being packaging waste, approximately 30,000 metric tons are generated per year. The purpose of the Reduction of Packaging Waste project was to investigate opportunities to reduce this packaging waste through source reduction and recycling. The project was divided into three areas: procurement, onsite packaging and distribution, and recycling. Waste minimization opportunities were identified and investigated within each area, several of which were chosen for further study and small-scale testing at the Hanford Site. Test results, were compiled into five ``how-to`` recipes for implementation at other sites. The subject of the recipes are as follows: (1) Vendor Participation Program; (2) Reusable Containers System; (3) Shrink-wrap System -- Plastic and Corrugated Cardboard Waste Reduction; (4) Cardboard Recycling ; and (5) Wood Recycling.

Raney, E.A.; Hogan, J.J.; McCollom, M.L.; Meyer, R.J.

1994-04-01T23:59:59.000Z

226

Validation Analysis of the Groundwater Flow and Transport Model of the Central Nevada Test Area  

Science Conference Proceedings (OSTI)

The Central Nevada Test Area (CNTA) is a U.S. Department of Energy (DOE) site undergoing environmental restoration. The CNTA is located about 95 km northeast of Tonopah, Nevada, and 175 km southwest of Ely, Nevada (Figure 1.1). It was the site of the Faultless underground nuclear test conducted by the U.S. Atomic Energy Commission (DOE's predecessor agency) in January 1968. The purposes of this test were to gauge the seismic effects of a relatively large, high-yield detonation completed in Hot Creek Valley (outside the Nevada Test Site [NTS]) and to determine the suitability of the site for future large detonations. The yield of the Faultless underground nuclear test was between 200 kilotons and 1 megaton (DOE, 2000). A three-dimensional flow and transport model was created for the CNTA site (Pohlmann et al., 1999) and determined acceptable by DOE and the Nevada Division of Environmental Protection (NDEP) for predicting contaminant boundaries for the site.

A. Hassan; J. Chapman; H. Bekhit; B. Lyles; K. Pohlmann

2006-09-30T23:59:59.000Z

227

Transportation  

NLE Websites -- All DOE Office Websites (Extended Search)

Meier AKMeier@lbl.gov (510) 486-4740 Links Transportation and Air Quality Batteries and Fuel Cells Buildings Energy Efficiency Electricity Grid Energy Analysis Energy...

228

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

Science Conference Proceedings (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

229

DOE - Safety of Radioactive Material Transportation  

NLE Websites -- All DOE Office Websites (Extended Search)

Crush Comparison Puncture Comparison Fire Comparison Immersion Comparison Demonstrating target hardness. Comparison of the Free Drop Test to a Passenger Train and Semi-truck Trailer Collision Free Drop Test 3,000,000 lbs of force present in this package certification test. [DROP test] Click to view picture Real-life Accident Comparison 1,000,000 lbs of force present in this real-life accident. [DROP scenario] Click to view picture Real-life scenarios that are encompassed by the above test include: the package being struck by a train traveling 60 MPH the package falling off of a 30-foot high bridge onto solid rock or from a higher bridge onto a highway or railroad the package running into a bridge support or rock slope at 45 MPH. Packages are transported onboard trucks or rail cars, which absorb some of the impact energy, reducing the resulting damage to the packages from the accident. On May 2, 1995, an O&J Gordon Trucking Company truck consisting of a tractor and a lowbed semitrailer became lodged on a high-profile (hump) railroad grade crossing near Sycamore, South Carolina. About 35 minutes later, the truck was struck by southbound Amtrak train No. 81, Silver Star, en route from New York City to Tampa, Florida.

230

Safety evaluation for packaging CPC metal boxes  

Science Conference Proceedings (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

231

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

Science Conference Proceedings (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

232

Handling and Packaging a Potentially Radiologically Contaminated Patient |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Handling and Packaging a Potentially Radiologically Contaminated Handling and Packaging a Potentially Radiologically Contaminated Patient Handling and Packaging a Potentially Radiologically Contaminated Patient The purpose of this procedure is to provide guidance to EMS care providers for properly handling and packaging potentially radiologically contaminated patients. This procedure applies to Emergency Medical Service care providers who respond to a radioactive material transportation incident that involves potentially contaminated injuries. Handling and Packaging a Potentially Radiologically Contaminated Patient.docx More Documents & Publications Pre-Hospital Practices for Handling a Radiologically Contaminated Patient Medical Examiner/Coroner on the Handling of a Body/Human Remains that are Potentially Radiologically Contaminated

233

Edible Oils Package  

Science Conference Proceedings (OSTI)

Contains four (4) titles regarding frying and edible oils. Edible Oils Package Food Science & Technology Health - Nutrition - Biochemistry Value Packages 1766A8D5F05863694E128DE1C47D07C3 This Value Package includes: ...

234

Transportation Management Workshop: Proceedings  

Science Conference Proceedings (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

235

Assessment of hydrologic transport of radionuclides from the Gnome underground nuclear test site, New Mexico  

SciTech Connect

The U.S. Department of Energy (DOE) is operating an environmental restoration program to characterize, remediate, and close non-Nevada Test Site locations that were used for nuclear testing. Evaluation of radionuclide transport by groundwater from these sites is an important part of the preliminary site risk analysis. These evaluations are undertaken to allow prioritization of the test areas in terms of risk, provide a quantitative basis for discussions with regulators and the public about future work at the sites, and provide a framework for assessing data needs to be filled by site characterization. The Gnome site in southeastern New Mexico was the location of an underground detonation of a 3.5-kiloton nuclear device in 1961, and a hydrologic tracer test using radionuclides in 1963. The tracer test involved the injection of tritium, {sup 90}Sr, and {sup 137}Cs directly into the Culebra Dolomite, a nine to ten-meter-thick aquifer located approximately 150 in below land surface. The Gnome nuclear test was carried out in the Salado Formation, a thick salt deposit located 200 in below the Culebra. Because salt behaves plastically, the cavity created by the explosion is expected to close, and although there is no evidence that migration has actually occurred, it is assumed that radionuclides from the cavity are released into the overlying Culebra Dolomite during this closure process. Transport calculations were performed using the solute flux method, with input based on the limited data available for the site. Model results suggest that radionuclides may be present in concentrations exceeding drinking water regulations outside the drilling exclusion boundary established by DOE. Calculated mean tritium concentrations peak at values exceeding the U.S. Environmental Protection Agency drinking water standard of 20,000 pCi/L at distances of up to almost eight kilometers west of the nuclear test.

Earman, S.; Chapman, J.; Pohlmann, K.; Andricevic, R.

1996-09-01T23:59:59.000Z

236

The transportable heavy-duty engine emissions testing laboratory. Annual progress report, April 1990--April 1991  

SciTech Connect

West Virginia University has designed and constructed a Transportable Emissions Testing Laboratory for measuring emissions from heavy duty vehicles, such as buses and trucks operating on conventional and alternative fuels. The laboratory facility can be transported to a test site located at, or nearby, the home base of the vehicles to be tested. The laboratory has the capability of measuring vehicle emissions as the vehicle is operated under either transient or steady state loads and speeds. The exhaust emissions from the vehicle is sampled and the levels of the constituents of the emission are measured. The laboratory consists of two major units; a power absorber unit and an emissions measurement unit. A power absorber unit allows for the connection of a dynamic load to the drive train of the vehicle so that the vehicle can be ``driven`` through a test cycle while actually mounted on a stationary test bed. The emissions unit contains instrumentation and equipment which allows for the dilution of the vehicle`s exhaust with air. The diluteed exhaust is sampled and analyzed to measure the level of concentration of those constituents which have been identified to have impact on the clean environment. Sampling probes withdraw diluted exhaust which is supplied to a number of different exhaust gas analysis instruments. The exhaust gas analysis instruments have the capability to measure the levels of the following exhaust gas constituents: carbon monoxide (CO), carbon dioxide (CO{sub 2}), oxides of nitrogen (NO{sub x}), unburned hydrocarbons (HC), formaldehyde (HCHO), methane and particulate matter. Additional instruments or sampling devices can be installed whenever measurements of additional constituents are desired. A computer based, data acquisition system is used to continuously monitor a wide range of parameters important to the operation of the test and to record the test results.

1991-05-01T23:59:59.000Z

237

Assessment of hydrologic transport of radionuclides from the Gasbuggy underground nuclear test site, New Mexico  

SciTech Connect

The U.S. Department of Energy (DOE) is operating an environmental restoration program to characterize, remediate, and close non-Nevada Test Site locations that were used for nuclear testing. Evaluation of radionuclide transport by groundwater from these sites is an important part of the preliminary risk analysis. These evaluations are undertaken to allow prioritization of the test areas in terms of risk, provide a quantitative basis for discussions with regulators and the public about future work at the sites, and provide a framework for assessing data needs to be filled by site characterization. The Gasbuggy site in northwestern New Mexico was the location of an underground detonation of a 29-kiloton nuclear device in 1967. The test took place in the Lewis Shale, approximately 182 m below the Ojo Alamo Sandstone, which is the aquifer closest to the detonation horizon. The conservative assumption was made that tritium was injected from the blast-created cavity into the Ojo Alamo Sandstone by the force of the explosion, via fractures created by the shot. Model results suggest that if radionuclides produced by the shot entered the Ojo Alamo, they are most likely contained within the area currently administered by DOE. The transport calculations are most sensitive to changes in the mean groundwater velocity, followed by the variance in hydraulic conductivity, the correlation scale of hydraulic conductivity, the transverse hydrodynamic dispersion coefficient, and uncertainty in the source size. This modeling was performed to investigate how the uncertainty in various physical parameters affects calculations of radionuclide transport at the Gasbuggy site, and to serve as a starting point for discussion regarding further investigation at the site; it was not intended to be a definitive simulation of migration pathways or radionuclide concentration values.

Earman, S.; Chapman, J.; Andricevic, R.

1996-09-01T23:59:59.000Z

238

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

NLE Websites -- All DOE Office Websites (Extended Search)

in PEM Fuel Cells: in PEM Fuel Cells: Advanced Modeling, Material Selection, Testing, and Design Optimization J. Vernon Cole and Ashok Gidwani CFDRC Prepared for: DOE Hydrogen Fuel Cell Kickoff Meeting February 13, 2007 This presentation does not contain any proprietary or confidential information. Background Water Management Issues Arise From: ƒ Generation of water by cathodic reaction ƒ Membrane humidification requirements ƒ Capillary pressure driven transport through porous MEA and GDL materials ƒ Scaling bipolar plate channel dimensions J.H. Nam and M. Kaviany, Int. J. Heat Mass Transfer, 46, pp. 4595-4611 (2003) Relevant Barriers and Targets ƒ Improved Gas Diffusion Layer, Flow Fields, Membrane Electrode Assemblies Needed to Improve Water Management: * Flooding blocks reactant transport

239

Why packages? The Windows tools  

E-Print Network (OSTI)

Why packages? The Windows tools A sample package Going further Package Development in Windows from August 13, 2008; updated November 23, 2012 1 of 45 #12;Why packages? The Windows tools A sample of packages 2 The Windows tools The main tools Missing pieces Installing the tools 3 A sample package Getting

Murdoch, Duncan

240

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

SciTech Connect

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

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

Safety Analysis Report for Packaging (SARP): Model AL-M1 nuclear packaging (DOE C of C No. USA/9507/BLF)  

Science Conference Proceedings (OSTI)

This Safety Analysis Report for Packaging (SARP) satisfies the request of the US Department of Energy for a formal safety analysis of the shipping container identified as USA/9507/BLF, also called AL-M1, configuration 5. This report makes available to all potential users the technical information and the limits pertinent to the construction and use of the shipping containers. It includes discussions of structural integrity, thermal resistance, radiation shielding and radiological safety, nuclear criticality safety, and quality control. A complete physical and technical description of the package is presented. The package consists of an inner container centered within an insulated steel drum. The configuration-5 package contains tritiated water held on sorbent material. There are two other AL-M1 packages, designated configurations 1 and 3. These use the same insulated outer drum, but licensing of these containers will not be addressed in this SARP. Design and development considerations, the tests and evaluations required to prove the ability of the container to withstand normal transportation conditions, and the sequence of four hypothetical accident conditions (free drop, puncture, thermal, and water immersion) are discussed. Tables, graphs, dimensional sketches, photographs, technical references, loading and shipping procedures, Monsanto Research Corporation-Mound experience in using the containers, and a copy of the DOE/OSD/ALO Certificate of Compliance are included.

Coleman, H.L.; Whitney, M.A.; Williams, M.A.; Alexander, B.M.; Shapiro, A.

1987-11-24T23:59:59.000Z

242

Radiation Level Changes at RAM Package Surfaces  

Science Conference Proceedings (OSTI)

This paper will explore design considerations required to meet the regulations that limit radiation level variations at external surfaces of radioactive material (RAM) packages. The radiation level requirements at package surfaces (e.g. TS-R-1 paragraphs 531 and 646) invoke not only maximum radiation levels, but also strict limits on the allowable increase in the radiation level during transport. This paper will explore the regulatory requirements by quantifying the amount of near surface movement and/or payload shifting that results in a 20% increase in the radiation level at the package surface. Typical IP-2, IP-3, Type A and Type B packaging and source geometries will be illustrated. Variations in surface radiation levels are typically the result of changes in the geometry of the surface due to an impact, puncture or crush event, or shifting and settling of radioactive contents.

Opperman, Erich [Washington Savannah River Company; Hawk, Mark B [ORNL; Kapoor, Ashok [U.S. Department of Energy, Office of Packaging and Transportation; Natali, Ronald [R. B. Natali Consulting, Inc.

2010-01-01T23:59:59.000Z

243

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

Science Conference Proceedings (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

244

Transportation Issues and Resolutions Compilation of Laboratory  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Transportation Issues and Resolutions Compilation of Laboratory Transportation Issues and Resolutions Compilation of Laboratory Transportation Work Package Reports Transportation Issues and Resolutions Compilation of Laboratory Transportation Work Package Reports The Transportation Team identified the retrievability and subcriticality safety functions to be of primary importance to the transportation of UNF after extended storage and to transportation of high burnup fuel. The tasks performed and described herein address issues related to retrievability and subcriticality; integrity of cladding (embrittled, high burnup cladding, loads applied to cladding during transport), criticality analyses of failed UNF within transport packages, moderator exclusion concepts, stabilization of cladding with canisters for criticality control;

245

Hot-Gas Filter Testing with a Transport Reactor Development Unit  

Science Conference Proceedings (OSTI)

The objective of the hot-gas cleanup (HGC) work on the transport reactor demonstration unit (TRDU) located at the Environmental Research Center is to demonstrate acceptable performance of hot-gas filter elements in a pilot-scale system prior to long-term demonstration tests. The primary focus of the experimental effort in the 2-year project will be the testing of hot- gas filter elements as a function of particulate collection efficiency, filter pressure differential, filter cleanability, and durability during relatively short-term operation (100-200 hours). A filter vessel will be used in combination with the TRDU to evaluate the performance of selected hot- gas filter elements under gasification operating conditions. This work will directly support the Power Systems Development Facility utilizing the M.W. Kellogg transport reactor located at Wilsonville, Alabama and indirectly the Foster Wheeler advanced pressurized fluid-bed combustor, also located at Wilsonville and the Clean Coal IV Pinon Pine IGCC Power Project. This program has a phased approach involving modification and upgrades to the TRDU and the fabrication, assembly, and operation of a hot-gas filter vessel (HGFV) capable of operating at the outlet design conditions of the TRDU. Phase 1 upgraded the TRDU based upon past operating experiences. Additions included a nitrogen supply system upgrade, upgraded LASH auger and 1807 coal feed lines, the addition of a second pressurized coal feed hopper and a dipleg ash hopper, and modifications to spoil the performance of the primary cyclone. Phase 2 included the HGFV design, procurement, and installation. Phases 3 through 5 consist of 200-hour hot-gas filter tests under gasification conditions using the TRDU at temperatures of 540-650{degrees}C (1000-1200{degrees}F), 9.3 bar, and face velocities of 1.4, 2. and 3.8 cm/s, respectively. The increased face velocities are achieved by removing candles between each test.

Swanson, M.L.; Ness, R.O., Jr. [North Dakota Univ., Grand Forks, ND (United States). Energy and Environmental Research Center

1996-12-31T23:59:59.000Z

246

Hot-gas filter testing with the transport reactor demonstration unit  

Science Conference Proceedings (OSTI)

The objectives of the hot-gas cleanup (HGC) work on the transport reactor demonstration unit (TRDU) located at the Energy & Environmental Research Center (EERC) is to demonstrate acceptable performance of hot-gas filter elements in a pilot-scale system prior to long-term demonstration tests. The primary focus of the experimental effort in the 2-year project will be the testing of hot-gas filter element performance (particulate collection efficiency, filter pressure differential, filter cleanability, and durability) as a function of temperature and filter face velocity during short-term operation (100-200 hours). This filter vessel will be utilized in combination with the TRDU to evaluate the performance of selected hot-gas filter elements under gasification operating conditions. This work will directly support the power systems development facility (PSDF) utilizing the M.W. Kellogg transport reactor located at Wilsonville, Alabama and, indirectly, the Foster Wheeler advanced pressurized fluid-bed combustor, also located at Wilsonville.

Mann, M.D.; Swanson, M.L.; Ness, R.O.; Haley, J.S.

1995-11-01T23:59:59.000Z

247

Performance testing of hydrogen transport membranes at elevated temperatures and pressures.  

DOE Green Energy (OSTI)

The development of hydrogen transport ceramic membranes offers increased opportunities for hydrogen gas separation and utilization. Commercial application of such membranes will most likely take place under conditions of elevated temperature and pressure, where industrial processes producing and or utilizing hydrogen occur, and where such membranes are theoretically expected to have the greatest permeability. Hydrogen separation membrane performance data at elevated temperature is quite limited, and data at elevated pressures is conspicuously lacking. This paper will describe the design, construction, and recent experimental results obtained from a membrane testing unit located at the U.S. Department of Energy's Federal Energy Technology Center (FETC). The membrane testing unit is capable of operating at temperatures up to 900 C and pressures up to 500 psi. Mixed-oxide ceramic ion-transport membranes, fabricated at Argonne National Laboratory (ANL), were evaluated for hydrogen permeability and characterized for surface changes and structural integrity using scanning electron microscopy/X-ray microanalysis (SEM/EDS), X-ray photoelectron spectroscopy (XPS) and atomic force microscopy (AFM), as a function of temperature, pressure, and hydrogen exposure.

Balachandran, U.; Cugini, A. V.; Dorris, S. E.; Fisher, E. P.; Graham, W. J.; Martello, D. V.; Poston, J. A.; Rothenberger, K. S.; Siriwardane, R. W.

1999-06-16T23:59:59.000Z

248

CH Packaging Operations Manual  

Science Conference Proceedings (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

249

CH Packaging Operations Manual  

SciTech Connect

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

2002-03-04T23:59:59.000Z

250

CH Packaging Operations Manual  

SciTech Connect

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

251

CH Packaging Operations Manual  

SciTech Connect

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

252

CH Packaging Operations Manual  

Science Conference Proceedings (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

253

CH Packaging Operations Manual  

SciTech Connect

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

254

Next Generation Packaging  

Science Conference Proceedings (OSTI)

Feb 28, 2011 ... Creation and Manipulation of Aligned Nanowires for Packaging and Circuit ... Plastic deformation plays an important role in the control of ...

255

Installation package for the solaron solar subsystems  

DOE Green Energy (OSTI)

This package contains information that is intended to be a guide for installation, operation, and maintenance of the various Solaron Solar Subsystems. The subsystems consist of the following: collectors, storage, transport (air handler) and controller for heat pump and off-peak storage. Two prototype residential systems have been installed at Akron, Ohio, and Duffield, Virginia.

Not Available

1979-04-01T23:59:59.000Z

256

Interconnect Design and Reliability in Electronic Packages III  

Science Conference Proceedings (OSTI)

More than 400 test vehicles were assembled using ceramic and plastic BGAs, LCCs, .... still draws big attention among the electronic package manufacturers.

257

Operational guidance for using DOT-6M/2R packaging  

SciTech Connect

The purpose of this paper is to describe a new US Department of Energy (DOE), Transportation Management Division task to create a US Department of Transportation (DOT) Specification 6M/2R packaging configuration user`s guide. The need for a user`s guide was identified because the DOT-6M/2R packaging configuration is widely used by DOE site contractors, and DOE receives many questions about the approved packaging configurations. Currently, two DOE organizations have the authority to approve new DOT-6M/2R configurations. For Defense Programs, the Transportation and Packaging Safety Division (EH-332) administers the program. For Environmental Restoration and Waste Management, the Transportation Management Division (EM-261) administers the program.

Kelly, D.L.; Hummer, J.H.

1994-03-01T23:59:59.000Z

258

Package downsizing: is it ethical?  

Science Conference Proceedings (OSTI)

Package downsizing is a practice where the package content is reduced without changing the package or the price of the product. In a market that is defined by ‘hyper-competition,’ package downsizing is often practiced by marketers to effect ... Keywords: Downsizing, Ethical, Package, Principle of equivalence

Omprakash K. Gupta; Sudhir Tandon; Sukumar Debnath; Anna S. Rominger

2007-04-01T23:59:59.000Z

259

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

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

260

Packaged CHP System Assessment  

Science Conference Proceedings (OSTI)

The Packaged CHP System Assessment report provides an analysis of packaged combined heat and power (CHP) systems. The report summarizes and compares the technical characteristics of commercial product lines with electric power output up to 3,000 kWe.

2004-03-22T23:59:59.000Z

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.


261

TRNSYS for windows packages  

SciTech Connect

TRNSYS 14.1 was released in 1994. This package represents a significant step forward in usability due to several graphical utility programs for DOS. These programs include TRNSHELL, which encapsulates TRNSYS functions, PRESIM, which allows the graphical creation of a simulation system, and TRNSED, which allows the easy sharing of simulations. The increase in usability leads to a decrease in the time necessary to prepare the simulation. Most TRNSYS users operate on PC computers with the Windows operating system. Therefore, the next logical step in increased usability was to port the current TRNSYS package to the Windows operating system. Several organizations worked on this conversion that has resulted in two distinct Windows packages. One package closely resembles the DOS version and includes TRNSHELL for Windows and PRESIM for Windows. The other package incorporates a general front-end, called IISIBat, that is a general simulation tool front-end. 8 figs.

Blair, N.J.; Beckman, W.A.; Klein, S.A.; Mitchell, J.W.

1996-09-01T23:59:59.000Z

262

Chemical compatibility screening results of plastic packaging to mixed waste simulants  

Science Conference Proceedings (OSTI)

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

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

1995-12-01T23:59:59.000Z

263

Transportation  

NLE Websites -- All DOE Office Websites (Extended Search)

Due to limited parking, all visitors are strongly encouraged to: Due to limited parking, all visitors are strongly encouraged to: 1) car-pool, 2) take the Lab's special conference shuttle service, or 3) take the regular off-site shuttle. If you choose to use the regular off-site shuttle bus, you will need an authorized bus pass, which can be obtained by contacting Eric Essman in advance. Transportation & Visitor Information Location and Directions to the Lab: Lawrence Berkeley National Laboratory is located in Berkeley, on the hillside directly above the campus of University of California at Berkeley. The address is One Cyclotron Road, Berkeley, California 94720. For comprehensive directions to the lab, please refer to: http://www.lbl.gov/Workplace/Transportation.html Maps and Parking Information: On Thursday and Friday, a limited number (15) of barricaded reserved parking spaces will be available for NON-LBNL Staff SNAP Collaboration Meeting participants in parking lot K1, in front of building 54 (cafeteria). On Saturday, plenty of parking spaces will be available everywhere, as it is a non-work day.

264

The ENSDF Java Package  

Science Conference Proceedings (OSTI)

A package of computer codes has been developed to process and display nuclear structure and decay data stored in the ENSDF (Evaluated Nuclear Structure Data File) library. The codes were written in an object-oriented fashion using the java language. This allows for an easy implementation across multiple platforms as well as deployment on web pages. The structure of the different java classes that make up the package is discussed as well as several different implementations.

Sonzogni, A.A. [National Nuclear Data Center, Brookhaven National Laboratory, Upton, NY 11973-5000 (United States)

2005-05-24T23:59:59.000Z

265

9975 SHIPPING PACKAGE PERFORMANCE OF ALTERNATE MATERIALS FOR LONG-TERM STORAGE APPLICATION  

SciTech Connect

The Model 9975 shipping package specifies the materials of construction for its various components. With the loss of availability of material for two components (cane fiberboard overpack and Viton{reg_sign} GLT O-rings), alternate materials of construction were identified and approved for use for transport (softwood fiberboard and Viton{reg_sign} GLT-S O-rings). As these shipping packages are part of a long-term storage configuration at the Savannah River Site, additional testing is in progress to verify satisfactory long-term performance of the alternate materials under storage conditions. The test results to date can be compared to comparable results on the original materials of construction to draw preliminary conclusions on the performance of the replacement materials.

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

2010-02-24T23:59:59.000Z

266

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

SciTech Connect

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

267

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

E-Print Network (OSTI)

Research Forum (PERF) Dense Gas Dispersion Modeling Project,Atmospheric dispersion of dense gases, Ann. Rev. Fluidgas (LNG) terminals and transport, and emphasize atmospheric dispersion

Oldenburg, Curtis M.

2009-01-01T23:59:59.000Z

268

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

DOE Green Energy (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

269

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

SciTech Connect

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

270

Packaging design criteria modified fuel spacer burial box. Revision 1  

Science Conference Proceedings (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

271

TRANSPORTATION TRANSPORTATION  

E-Print Network (OSTI)

TEXASTRANS TEXAS TRANSPORTATION HALL HONOR OF HALL HONOR OF TEXASTRAN HALL HONOR OF TEXASTRAN HALL HONOR OF Inductees #12;2 TEXAS TRANSPORTATION HALL HONOR OF L NOR OF Texas is recognized as having one of the finest multimodal transportation systems in the world. The existence of this system has been key

272

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

Science Conference Proceedings (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

273

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

SciTech Connect

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

274

Development of an Automated Pit Packaging System for Pantex  

E-Print Network (OSTI)

Sandia National Laboratories is developing a system that uses robots to package pits at Pantex in the AT-400A pit storage and transportation container. This report will give an overview of the AT-400A packaging process, and the parts of the overall AT-400A packaging operation that will be performed robotically. The process employed to move from development in the laboratory at Sandia to production use at Pantex will be described. Finally, important technology components being developed for and incorporated into the robotic system will be described. ____________________________________________________________________________ 2 Development of an Automated Pit Packaging System for Pantex Intentionally Left Blank ______________________________________________________________________________ Development of an Automated Pit Packaging System for Pantex 3 Contents 1. Introduction........................................................................................................... 7 ...

Jill Fahrenholtz Manufacturing; System For Pantex; Jill C. Fahrenholtz

1997-01-01T23:59:59.000Z

275

PTS 13.2 Packaging and Preparation for Shipment 4/10/95 | Department of  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

PTS 13.2 Packaging and Preparation for Shipment 4/10/95 PTS 13.2 Packaging and Preparation for Shipment 4/10/95 PTS 13.2 Packaging and Preparation for Shipment 4/10/95 The objective of this surveillance is to evaluate the effectiveness of the contractor's programs for packaging radioactive and hazardous wastes for shipment. The Facility Representative examines packages ready for shipment, observes preparation of packages, and reviews documents that establish the acceptability of packages. The Facility Representative verifies compliance with DOE requirements including requirements established by the Department of Transportation and the U.S. Nuclear Regulatory Commission. PTS13-02.doc More Documents & Publications PTS 13.1 Radioactive And Hazardous Material Transportation 4/13/00 CMS 3.4 Temporary Changes, 4/10/95

276

THERMAL EVALUATION OF DRUM TYPE RADIOACTIVE MATERIAL PACKAGING ARRAYS IN STORAGE  

SciTech Connect

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

277

MODELING ASSUMPTIONS FOR THE ADVANCED TEST REACTOR FRESH FUEL SHIPPING CONTAINER  

SciTech Connect

The Advanced Test Reactor Fresh Fuel Shipping Container (ATR FFSC) is currently licensed per 10 CFR 71 to transport a fresh fuel element for either the Advanced Test Reactor, the University of Missouri Research Reactor (MURR), or the Massachusetts Institute of Technology Research Reactor (MITR-II). During the licensing process, the Nuclear Regulatory Commission (NRC) raised a number of issues relating to the criticality analysis, namely (1) lack of a tolerance study on the fuel and packaging, (2) moderation conditions during normal conditions of transport (NCT), (3) treatment of minor hydrogenous packaging materials, and (4) treatment of potential fuel damage under hypothetical accident conditions (HAC). These concerns were adequately addressed by modifying the criticality analysis. A tolerance study was added for both the packaging and fuel elements, full-moderation was included in the NCT models, minor hydrogenous packaging materials were included, and fuel element damage was considered for the MURR and MITR-II fuel types.

Rick J. Migliore

2009-09-01T23:59:59.000Z

278

Work Package Templates  

Science Conference Proceedings (OSTI)

Work Package Templates provides fossil plant maintenance personnel with assorted inspection, minor repair or overhaul templates for various pieces of plant equipment. This guide will assist plant maintenance personnel in improving the efficiency, reliability and reducing the maintenance costs for associated with maintenance on selected pieces of equipment.

2007-02-15T23:59:59.000Z

279

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

280

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

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

Package for fragile objects  

SciTech Connect

A package for fragile objects such as radioactive fusion pellets of micron size shipped in mounted condition or unmounted condition with a frangible inner container which is supported in a second inner container which in turn is supported in a final outer container, the second inner container having recesses for supporting alternate design inner containers.

Burgeson, Duane A. (Ann Arbor, MI)

1977-01-01T23:59:59.000Z

282

River Data Package for the 2004 Composite Analysis  

Science Conference Proceedings (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

283

Development of an automated pit packaging system for Pantex  

Science Conference Proceedings (OSTI)

Sandia National Laboratories is developing a system that uses robots to package pits at Pantex in the AT-400A pit storage and transportation container. This report will give an overview of the AT-400A packaging process, and the parts of the overall AT-400A packaging operation that will be performed robotically. The process employed to move from development in the laboratory at Sandia to production use at Pantex will be described. Finally, important technology components being developed for and incorporated into the robotic system will be described. 7 refs., 9 figs.

Fahrenholtz, J.C.

1997-09-01T23:59:59.000Z

284

DESTRUCTIVE EXAMINATION OF SHIPPING PACKAGE 9975-03431  

SciTech Connect

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

285

Mining IC Test Data to Optimize VLSI Testing Tony Fountain Thomas Dietterich Bill Sudyka  

E-Print Network (OSTI)

contacts are bonded in place, and then a protective covering is added. 5. Package test Once the ICs are assembled into packages and connected to the package pins. Finally, the resulting packages are tested is to avoid the expense of packaging bad die and to provide rapid feedback to the fabrication process

Dietterich, Thomas G.

286

Vehicle Testing and Analysis Group: Center for Transportation Technologies and Systems (CTTS) (Brochure)  

DOE Green Energy (OSTI)

Describes NREL's Vehicle Testing and Analysis Group's work in vehicle and fleet evaluations, testing, data, and analysis for government and industry partners.

Not Available

2008-10-01T23:59:59.000Z

287

ORISE Contract, PART 1 Â… THE SCHEDULE, Section D Packaging and Marking  

NLE Websites -- All DOE Office Websites (Extended Search)

THE SCHEDULE THE SCHEDULE SECTION D PACKAGING AND MARKING D.1 PACKAGING (NOV 2004) .................................................................................................. 3 D.2 MARKING (MAY 1997)...................................................................................................... 3 Section D - Page 1 of 4 DE-AC05-06OR23100 Blank Page Section D - Page 2 of 4 DE-AC05-06OR23100 PART I - THE SCHEDULE SECTION D PACKAGING AND MARKING D.1 PACKAGING (NOV 2004) Preservation, packaging, and packing for shipment or mailing of all work delivered hereunder shall be in accordance with good commercial practice and adequate to insure acceptance by common carrier and safe transportation at the most economical rate(s). The Contractor shall not utilize certified or registered

288

Transportation and Program Management Services  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Atlanta, Georgia Atlanta, Georgia Transportation and Program Management Services Secured Transportation Services, LLC Founded: December, 2003 ff Staff: 7 Experience: Over 145 years combined experience in Nuclear Transportation, Security, HP & Operations Services Transportation The largest Transportation Coordinators of Spent Nuclear Fuel in North America On-Site, Hands-On Assistance (Before & During both Loading & Transport) P d A i t (W iti d/ R i ) Procedure Assistance (Writing and/or Review) Package Handling, Loading Services Certificate of Compliance and Competent Authority Reviews & Requests Carrier Coordination (Empty Packages & Equipment, Loaded, & Returns) Vessel Charters, Special Trains, Dedicated Exclusive Use Trucks p

289

PORFLOW TESTING AND VERIFICATION DOCUMENT  

Science Conference Proceedings (OSTI)

The PORFLOW software package is a comprehensive mathematical model for simulation of multi-phase fluid flow, heat transfer and mass transport in variably saturated porous and fractured media. PORFLOW can simulate transient or steady-state problems in Cartesian or cylindrical geometry. The porous medium may be anisotropic and heterogeneous and may contain discrete fractures or boreholes with the porous matrix. The theoretical models within the code provide a unified treatment of concepts relevant to fluid flow and transport. The main features of PORFLOW that are relevant to Performance Assessment modeling at the Savannah River National Laboratory (SRNL) include variably saturated flow and transport of parent and progeny radionuclides. This document involves testing a relevant sample of problems in PORFLOW and comparing the outcome of the simulations to analytical solutions or other commercial codes. The testing consists of the following four groups. Group 1: Groundwater Flow; Group 2: Contaminant Transport; Group 3: Numerical Dispersion; and Group 4: Keyword Commands.

Aleman, S

2007-06-18T23:59:59.000Z

290

Energy Conservation Tests of a Coupled Kinetic-kinetic Plasma-neutral Transport Code  

SciTech Connect

A Monte Carlo neutral transport routine, based on DEGAS2, has been coupled to the guiding center ion-electron-neutral neoclassical PIC code XGC0 to provide a realistic treatment of neutral atoms and molecules in the tokamak edge plasma. The DEGAS2 routine allows detailed atomic physics and plasma-material interaction processes to be incorporated into these simulations. The spatial pro le of the neutral particle source used in the DEGAS2 routine is determined from the uxes of XGC0 ions to the material surfaces. The kinetic-kinetic plasma-neutral transport capability is demonstrated with example pedestal fueling simulations.

D.P. Stotler, C.S. Chang, S.H. Ku, J. Lang and G. Park

2012-08-29T23:59:59.000Z

291

Technical Review Report for the Mound 1KW Package Safety Analysis Report for Packaging Addendum No. 1, through Revision b  

SciTech Connect

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 'Mound 1KW Package Safety Analysis Report for Packaging, Addendum No. 1, Revision b', dated May 2007 (Addendum 1). 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 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). Addendum 1 incorporates a new fueled capsule assembly payload. The following changes have been made to add this payload: (1) The primary containment vessel (PCV) will be of the same design, but will increase in height to 11.16 in.; (2) A new graphite support block will be added to support up to three fueled capsule assemblies per package; (3) The cutting groove height on the secondary containment vessel (SCV) will be heightened to accommodate the taller PCV; and (4) A 3.38 in. high graphite filler block will be placed on top of the PCV. All other packaging features, as described in the Mound 1KW SARP [3], remain unchanged. This report documents the LLNL review of Addendum 1[1]. The specific review for each SARP Chapter is documented herein.

DiSabatino, A; West, M; Hafner, R; Russell, E

2007-10-04T23:59:59.000Z

292

NUCLEAR MATERIAL PACKAGING MANUAL  

E-Print Network (OSTI)

The enclosed copy ofdraft DOE Manual M44I.I, Nuclear Material Packaging Manual, is forwarded for your review and comment. This satisfies commitment 5.1-3 in Appendix o ofthe implementation plan (IP) for recommendation 2005-1, Nuclear Material Packaging. The next milestone in Section 5.1 ofthe 2005-1 IP is forwarding the manual to the DOE 2005-1 Technical Review Board (TRB) by April 30, 2006 to begin the final TRB review. Therefore, your comments are requested by April 21, 2006, in order to allow one week for resolution and updating the manual before it is sent to the TRB. Please contact me at 301-903-4407 ifyou have any questions. t

The Honorable; A. J. Eggenberger; M. Whitaker Dr-i

2006-01-01T23:59:59.000Z

293

Aquaculture information package  

DOE Green Energy (OSTI)

This package of information is intended to provide background information to developers of geothermal aquaculture projects. The material is divided into eight sections and includes information on market and price information for typical species, aquaculture water quality issues, typical species culture information, pond heat loss calculations, an aquaculture glossary, regional and university aquaculture offices and state aquaculture permit requirements. A bibliography containing 68 references is also included.

Boyd, T.; Rafferty, K.

1998-08-01T23:59:59.000Z

294

Plutonium stabilization and packaging system  

Science Conference Proceedings (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

295

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

296

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

SciTech Connect

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

297

Performance oriented packaging report for fuse, blasting, time, M700. Final report  

Science Conference Proceedings (OSTI)

This POP report is for the Fuse, Blasting, Time, M700 which is packaged 4000 feet/ Mil-B-2427 wood box. This report describes the results of testing conducted on a similar packaging which is used as an analogy for this item....Performance oriented packaging, POP, Fuse, Blasting, Time, M700, Mil-B-2427 Wood box.

Sniezek, F.M.

1992-11-02T23:59:59.000Z

298

Design package for programmable controller and hydronic subsystem  

DOE Green Energy (OSTI)

This report contains information used in the evaluation of design of Sunkeeper Control's electronic controllers and hydronic packages. Some of the information includes system performance specification, design data brochure, drawings, and qualification and acceptance test procedures.

Not Available

1979-03-01T23:59:59.000Z

299

Effects of Hanford tank simulant waste on plastic packaging to components  

Science Conference Proceedings (OSTI)

We have developed a chemical compatibility program for the evaluation of plastic packaging components which may be incorporated in packaging for transporting mixed waste forms. Consistent with the methodology outlined in this paper, we have performed the second phase of this experimental program to determine the effects of simulant Hanford Tank mixed wastes on packaging materials. This effort involved the comprehensive testing of five plastic liner materials in the aqueous mixed waste simulant. The testing protocol involved exposing the respective materials to {approximately}1, 3, 6, and 40 kGy of gamma radiation followed by 7, 14, 28, 180 day exposures to the waste simulant at 18, 50, and 60{degree}C. From the limited data analyses performed to date in this study, we have identified the fluorocarbon Kel-F{trademark} as having the greatest chemical compatibility after having been exposed to 40 kGy gamma radiation followed by exposure to the Hanford Tank simulant mixed waste at 60{degree}C. The most stricking observation from this study was the poor performance of Teflon under these conditions.

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

1995-12-01T23:59:59.000Z

300

Computer-assisted comparison of analysis and test results in transportation experiments  

SciTech Connect

As a part of its ongoing research efforts, Sandia National Laboratories` Transportation Surety Center investigates the integrity of various containment methods for hazardous materials transport, subject to anomalous structural and thermal events such as free-fall impacts, collisions, and fires in both open and confined areas. Since it is not possible to conduct field experiments for every set of possible conditions under which an actual transportation accident might occur, accurate modeling methods must be developed which will yield reliable simulations of the effects of accident events under various scenarios. This requires computer software which is capable of assimilating and processing data from experiments performed as benchmarks, as well as data obtained from numerical models that simulate the experiment. Software tools which can present all of these results in a meaningful and useful way to the analyst are a critical aspect of this process. The purpose of this work is to provide software resources on a long term basis, and to ensure that the data visualization capabilities of the Center keep pace with advancing technology. This will provide leverage for its modeling and analysis abilities in a rapidly evolving hardware/software environment.

Knight, R.D. [Gram, Inc., Albuquerque, NM (United States); Ammerman, D.J.; Koski, J.A. [Sandia National Labs., Albuquerque, NM (United States)

1998-05-10T23:59:59.000Z

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.


301

CH Packaging Operations for High Wattage Waste at LANL  

SciTech Connect

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

Washington TRU Solutions LLC

2003-03-21T23:59:59.000Z

302

CH Packaging Operations for High Wattage Waste at LANL  

Science Conference Proceedings (OSTI)

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

Washington TRU Solutions LLC

2002-12-18T23:59:59.000Z

303

CH Packaging Operations for High Wattage Waste at LANL  

Science Conference Proceedings (OSTI)

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

Washington TRU Solutions LLC

2003-08-28T23:59:59.000Z

304

CH Packaging Operations for High Wattage Waste at LANL  

SciTech Connect

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

Washington TRU Solutions LLC

2003-05-06T23:59:59.000Z

305

Identification of unique reciprocal and non reciprocal cross packaging relationships between HIV-1, HIV-2 and SIV reveals an efficient SIV/HIV-2 lentiviral vector system with highly favourable features for in vivo testing and clinical usage  

E-Print Network (OSTI)

, transduction of neural stem cells with lentiviral and adeno associated viral vectors express- ing therapeutic genes that will affect differentiation and Gag-Pol packaging constructsFigure 1 Gag-Pol packaging constructs. gag Poly ApolLTR vpx vprvif rev tat ?Env... (1153bp) ?? ? ? gag Poly A polCMV rev tat ?Env ?? HIV-1 Gag-Pol ?8.9 (2nd Generation) gag LTRpolLTR vpx vprvif rev tat nef ?Env (550bp) ?? ? ? RRE RRE RRE SIV Gag-Pol (SgpDelta2) Gag-Pol packaging constructs HIV-2 Gag-PolPage 2 of 14 (page number...

Strappe, Padraig M; Hampton, David W; Brown, Douglas; Gonzales, Begona-cachon; Caldwell, Maeve; Fawcett, James W; Lever, Andrew M L

2005-09-16T23:59:59.000Z

306

Aging Tests of Neutron-Shielding Materials for Transport of Storage Casks  

Science Conference Proceedings (OSTI)

Special Issue Technical Paper / Second Seminar on Accelerated Testing of Materials in Spent Nuclear Fuel and High-Level Waste Storage Systems / Materials for Nuclear Systems

Herve Issard; Pascale Abadie

307

Modeling and analysis of a heat transport transient test facility for space nuclear systems.  

E-Print Network (OSTI)

??The purpose of this thesis is to design a robust test facility for a small space nuclear power system and model its physical behavior under… (more)

[No author

2013-01-01T23:59:59.000Z

308

LAC Regional Platform Workshop Participant Package | Open Energy  

Open Energy Info (EERE)

Participant Package Participant Package Jump to: navigation, search LEDSGP Logo.png Advancing climate-resilient low emission development around the world Home About Tools Expert Assistance Events Publications Join Us LAC Workshop Announcement Agenda Participant Package Accommodations Location & Transportation Insurance & Visas Participants Presentations Outcomes Report Links Contact Us General Information (Español) Participant Package.pdf Powered by OpenEI ledsgp.org is built on the same platform as the popular Wikipedia site. Like Wikipedia, it is a "wiki" or website developed collaboratively by a community of users. Thanks to our unique relationship with OpenEI.org, you can add or edit most content on ledsgp.org. For more information about this unique collaboration, contact us. View or edit this page on OpenEI.org.

309

The Fireball integrated code package  

Science Conference Proceedings (OSTI)

Many deep-space satellites contain a plutonium heat source. An explosion, during launch, of a rocket carrying such a satellite offers the potential for the release of some of the plutonium. The fireball following such an explosion exposes any released plutonium to a high-temperature chemically-reactive environment. Vaporization, condensation, and agglomeration processes can alter the distribution of plutonium-bearing particles. The Fireball code package simulates the integrated response of the physical and chemical processes occurring in a fireball and the effect these processes have on the plutonium-bearing particle distribution. This integrated treatment of multiple phenomena represents a significant improvement in the state of the art for fireball simulations. Preliminary simulations of launch-second scenarios indicate: (1) most plutonium vaporization occurs within the first second of the fireball; (2) large non-aerosol-sized particles contribute very little to plutonium vapor production; (3) vaporization and both homogeneous and heterogeneous condensation occur simultaneously; (4) homogeneous condensation transports plutonium down to the smallest-particle sizes; (5) heterogeneous condensation precludes homogeneous condensation if sufficient condensation sites are available; and (6) agglomeration produces larger-sized particles but slows rapidly as the fireball grows.

Dobranich, D.; Powers, D.A.; Harper, F.T.

1997-07-01T23:59:59.000Z

310

The reduction of packaging waste  

Science Conference Proceedings (OSTI)

Nationwide, packaging waste comprises approximately one third of the waste being sent to our solid waste landfills. These wastes range from product and shipping containers made from plastic, glass, wood, and corrugated cardboard to packaging fillers and wraps made from a variety of plastic materials such as shrink wrap and polystyrene peanuts. The amount of packaging waste generated is becoming an important issue for manufacturers, retailers, and consumers. Elimination of packaging not only conserves precious landfill space, it also reduces consumption of raw materials and energy, all of which result in important economic and environmental benefits. At the US Department of Energy-Richland Field Office's (DOE-RL) Hanford Site as well as other DOE sites the generation of packaging waste has added importance. By reducing the amount of packaging waste, DOE also reduces the costs and liabilities associated with waste handling, treatment, storage, and disposal.

Raney, E.A.; McCollom, M.; Hogan, J.

1993-04-01T23:59:59.000Z

311

The reduction of packaging waste  

Science Conference Proceedings (OSTI)

Nationwide, packaging waste comprises approximately one third of the waste being sent to our solid waste landfills. These wastes range from product and shipping containers made from plastic, glass, wood, and corrugated cardboard to packaging fillers and wraps made from a variety of plastic materials such as shrink wrap and polystyrene peanuts. The amount of packaging waste generated is becoming an important issue for manufacturers, retailers, and consumers. Elimination of packaging not only conserves precious landfill space, it also reduces consumption of raw materials and energy, all of which result in important economic and environmental benefits. At the US Department of Energy-Richland Field Office`s (DOE-RL) Hanford Site as well as other DOE sites the generation of packaging waste has added importance. By reducing the amount of packaging waste, DOE also reduces the costs and liabilities associated with waste handling, treatment, storage, and disposal.

Raney, E.A.; McCollom, M.; Hogan, J.

1993-04-01T23:59:59.000Z

312

DESTRUCTIVE EXAMINATION OF SHIPPING PACKAGE 9975-02168  

Science Conference Proceedings (OSTI)

The Savannah River Site (SRS) stores packages containing plutonium (Pu) materials in the K-Area 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 9975-02028 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. Destructive and non-destructive examinations have been performed on specified components of shipping package 9975-02168. For those attributes that were also measured during the field surveillance, no significant changes were observed. Two conditions were identified that do not meet inspection criteria. These conditions are subject to additional investigation and disposition by the Surveillance Program Authority. The conditions include: (1) The lead shield was covered with a white corrosion layer, and (2) Fiberboard thermal conductivity in the axial direction exceeded the specified range. The Surveillance Program Authority was notified of these conditions and will document the findings by surveillance report. All other observations and test results met identified criteria, or were collected for information and trending purposes.

Daugherty, W.

2010-11-18T23:59:59.000Z

313

IN-PACKAGE CHEMISTRY ABSTRACTION  

Science Conference Proceedings (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

314

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

Science Conference Proceedings (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

315

About the ZOOM minimization package  

SciTech Connect

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

Fischler, M.; Sachs, D.; /Fermilab

2004-11-01T23:59:59.000Z

316

Convective Transport Theory for Surface Fluxes Tested over the Western Pacific Warm Pool  

Science Conference Proceedings (OSTI)

Turbulent flux measurements from five flights of the National Center for Atmospheric Research Electra aircraft during the Tropical Oceans and Global Atmosphere Coupled Ocean–Atmosphere Response Experiment (TOGA COARE) are used to test convective ...

Lawrence Greischar; Roland Stull

1999-07-01T23:59:59.000Z

317

Quality Assurance requirements - Safety Analysis Reports for Packaging. An effective approach in developing QA requirements  

SciTech Connect

Application of QA requirements for packaging and transportation of radioactive materials should not be solely based on safety-related considerations. The operability of items, components, and systems must be considered as equally important. The nuclear industry has begun to recognize operability considerations along with safety concerns. This has resulted in a new approach in establishing QA requirements for packaging.

Fabian, R.R.

1986-01-01T23:59:59.000Z

318

Annual Transportation Report for Radioactive Waste Shipments to and from the Nevada Test Site, Fiscal Year 2006  

SciTech Connect

In February 1997, the U.S. Department of Energy, Nevada Operations 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 U.S. Department of Energy, 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 Sites (RWMS) at Area 3 and Area 5. This document satisfies requirements with regard to low-level radioactive waste (LLW) and mixed low-level radioactive waste (MLLW) transported to or from the NTS during fiscal year (FY) 2006.

DOE /NNSA NSO

2007-01-01T23:59:59.000Z

319

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

320

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 outer drum. The DMSB includes an inlet diffuser assembly, an outlet diffuser assembly, and a hydrogen catalytic recombiner. The DMSB adsorbs 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.

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

1994-12-31T23:59:59.000Z

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

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

322

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

Science Conference Proceedings (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

323

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

324

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

SciTech Connect

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

325

TEPP Training - Modular Emergency Response Radiological Transportation  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Services » Waste Management » Packaging and Transportation » Services » Waste Management » Packaging and Transportation » Transportation Emergency Preparedness Program » TEPP Training - Modular Emergency Response Radiological Transportation Training (MERRTT) TEPP Training - Modular Emergency Response Radiological Transportation Training (MERRTT) Once the jurisdiction has completed an evaluation of their plans and procedures, they will need to address any gaps in training. To assist, TEPP has developed the Modular Emergency Response Radiological Transportation Training (MERRTT) program. MERRTT provides fundamental knowledge for responding to transportation incidents involving radiological material and builds on training in existing hazardous materials curricula. MERRTT satisfies the training requirements outlined in the Waste Isolation Pilot

326

Status of PERST-5 package  

SciTech Connect

The methods and algorithms used in the PERST-5 package are described. This package is part of the MCU-5 code and is intended for neutron-physical calculation of the cells and parts of nuclear reactors using a generalized method of first collision probabilities.

Gomin, E. A.; Gurevich, M. I.; Kalugin, M. A.; Lazarenko, A. P.; Pryanichnikov, A. V., E-mail: prianik@adis.vver.kiae.ru; Sidorenko, V. D. [National Research Centre Kurchatov Institute (Russian Federation); Druzhinin, V. E. [Scientific and Research Institute of Nuclear Power Plant Operation (VNIIAES) (Russian Federation); Zhirnov, A. P.; Rozhdestvenskiy, I. M. [Scientific Research and Design Institute of Electrical Engineering (NIKIET) (Russian Federation)

2012-12-15T23:59:59.000Z

327

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

328

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

SciTech Connect

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

329

Solar test of an integrated sodium reflux heat pipe receiver/reactor for thermochemical energy transport  

DOE Green Energy (OSTI)

A chemical reactor for carbon dioxide reforming of methane was integrated into a sodium reflux heat pipe receiver and tested in the solar furnace of the Weizmann Institute of Science, Rehovot, Israel. The receiver/reactor was a heat pipe with seven tubes inside an evacuated metal box containing sodium. The catalyst, 0.5 wt% Rh on alumina, filled two of the tubes with the front surface of the box serving as the solar absorber. In operation, concentrated sunlight heated the front plate and vaporized sodium from a wire mesh wick attached to other side. Sodium vapor condensed on the reactor tubes, releasing latent heat and returning to the wick by gravity. The receiver system performed satisfactorily in many tests under varying flow conditions. The maximum power absorbed was 7.5 kW at temperatures above 800C. The feasibility of operating a heat pipe receiver/reactor under solar conditions was proven, and the advantages of reflux devices confirmed.

Diver, R.B.; Fish, J.D. (Sandia National Labs., Albuquerque, NM (United States)); Levitan, R.; Levy, M.; Meirovitch, E.; Rosin, H. (Weizmann Inst. of Science, Rehovot (Israel)); Paripatyadar, S.A.; Richardson, J.T. (Univ. of Houston, TX (United States))

1992-01-01T23:59:59.000Z

330

CDIAC catalog of numeric data packages and computer model packages  

Science Conference Proceedings (OSTI)

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

Boden, T.A. [Oak Ridge National Lab., TN (United States). Carbon Dioxide Information Analysis Center; O`Hara, F.M. Jr. [O`Hara (Fred M., Jr.), Oak Ridge, TN (US); Stoss, F.W. [Univ. of Tennessee, Knoxville, TN (US). Energy, Environment, and Resources Center

1993-05-01T23:59:59.000Z

331

Pre-Packaged Data Set List  

NLE Websites -- All DOE Office Websites (Extended Search)

Pre-Packaged Data Sets You will need to RegisterSign In to Order Pre-Packaged Products. You must RegisterSign In to order Pre-Packaged Products. Close window....

332

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

E-Print Network (OSTI)

The Linear One-cell Functional Experimental Studies (LOCFES) suite of codes provides a user with a means of testing and comparing spatial approximations to the monoenergetic, azimuthally symmetric, steady-state, one dimensional transport equation. Prior to the latest addendum described in this thesis, the structure of the codes in this suite necessitated that the functional form for the volumetric flux values depend linearly on the cell's edge flux and source moments. A user supplied set of coefficients for these terms could then be used to specify any linear spatial approximation. While such linear methods represent the bulk of closure approximations, some methods that permit this dependence to be nonlinear have proven to be very effective as well. This thesis is concerned with an addition (LOCFES-NL) to the Linear One-Cell Experimental Studies (LOCFES) suite of transport codes to include support for nonlinear spatial approximations. This latest addition to the suite lets the user implement any method and evaluate the method's performance on user-supplied problems. For illustration, sample computations compare the nonlinear Exponential Method (EM) of Barbucci and Pasquantonio to the Nonlinear Characteristics (NC) Method of Walters and Wareing and to other common linear methods.

Nolen, Steven Douglas

1997-01-01T23:59:59.000Z

333

Analytical Solution of a Simplified Transport Problem with Continuous Energy for Testing Monte Carlo Procedures,” this conference  

E-Print Network (OSTI)

A simplified transport problem is presented with continuous energy and neutrons moving only into the +X and –X direction. An exact analytical solution is given with a 1/E energy dependence of the space, direction and energy dependent neutron flux. The source function, which also behaves basically as 1/E has to be modified for a certain energy range if a maximum energy is introduced in the problem. At the cost of more complicated mathematics the total and scattering cross sections and the anisotropy of scattering may very with energy. This model can be implemented in a general purpose Monte Carlo code like MCNP5 without modification, but needs a specially prepared cross section library file. The model can be applied to test Monte Carlo procedures, like the generation of multi-group cross section and scattering matrices which can be calculated analytically from the continuous-energy cross section data. From the adjoint equation the optimum importance function can be derived, which can be used to devise a continuous-energy zero-variance Monte Carlo scheme. Key Words: transport problem, analytical solution, continuous energy, Monte Carlo 1.

J. Eduard Hoogenboom

2007-01-01T23:59:59.000Z

334

EBS Radionuclide Transport Abstraction  

Science Conference Proceedings (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

335

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

Science Conference Proceedings (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

336

Preliminary testing of turbulence and radionuclide transport modeling in deep ocean environment  

Science Conference Proceedings (OSTI)

Pacific Northwest Laboratory (PNL) performed a study for the US Environmental Protection Agency's Office of Radiation Programs to (1) identify candidate models for regional modeling of low-level waste ocean disposal sites in the mid-Atlantic ocean; (2) evaluate mathematical representation of the model's eddy viscosity/dispersion coefficients; and (3) evaluate the adequacy of the k-{epsilon} turbulence model and the feasibility of one of the candidate models, TEMPEST{copyright}/FLESCOT{copyright}, to deep-ocean applications on a preliminary basis. PNL identified the TEMPEST{copyright}/FLESCOT{copyright}, FLOWER, Blumberg's, and RMA 10 models as appropriate candidates for the regional radionuclide modeling. Among these models, TEMPEST/FLESCOT is currently the only model that solves distributions of flow, turbulence (with the k-{epsilon} model), salinity, water temperature, sediment, dissolved contaminants, and sediment-sorbed contaminants. Solving the Navier-Stokes equations using higher order correlations is not practical for regional modeling because of the prohibitive computational requirements; therefore, the turbulence modeling is a more practical approach. PNL applied the three-dimensional code, TEMPEST{copyright}/FLESCOT{copyright} with the k-{epsilon} model, to a very simple, hypothetical, two-dimensional, deep-ocean case, producing at least qualitatively appropriate results. However, more detailed testing should be performed for the further testing of the code. 46 refs., 39 figs., 6 tabs.

Onishi, Y.; Dummuller, D.C.; Trent, D.S. (Pacific Northwest Lab., Richland, WA (USA); Washington State Univ., Pullman, WA (USA); Pacific Northwest Lab., Richland, WA (USA))

1989-03-01T23:59:59.000Z

337

Shipment of Small Quantities of Unspecified Radioactive Material in Chalfant Packagings  

SciTech Connect

In the post 6M era, radioactive materials package users are faced with the disciplined operations associated with use of Certified Type B packagings. Many DOE, commercial and academic programs have a requirement to ship and/or store small masses of poorly characterized or unspecified radioactive material. For quantities which are small enough to be fissile exempt and have low radiation levels, the materials could be transported in a package which provides the required containment level. Because their Chalfant type containment vessels meet the highest standard of containment (helium leak-tight), the 9975, 9977, and 9978 are capable of transporting any of these contents. The issues associated with certification of a high-integrity, general purpose package for shipping small quantities of unspecified radioactive material are discussed and certification of the packages for this mission is recommended.

Smith, Allen; Abramczyk, Glenn; Nathan, Steven; Bellamy, Steve

2009-06-12T23:59:59.000Z

338

Some effects of packaging materials on critical arrays of fissile materials  

SciTech Connect

The surface density representation of array criticality provides a comprehensive display of criticality parameters of arrays of packaged fissile materials. The study leads to the following conclusions: (1) The mass limits established by the N 16.5 standard for air-spaced spherical units in water-reflected arrays may be adequate for transportation packages; (2) criticality assessments made for one fissile material can be extended to other materials which have defined equivalent masses for array criticality of air-spaced units; and (3) a uniform minimum margin of subcriticality can be established for transportation of packaged fissile materials.

Thomas, J.T.; Tang, J.S.

1977-01-01T23:59:59.000Z

339

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

340

Development of a container for the transportation and storage of plutonium bearing materials  

Science Conference Proceedings (OSTI)

There is a large backlog of plutonium contaminated materials at the Rocky Flats Environmental Technology Site near Denver, Colorado, USA. The clean-up of this site requires this material to be packaged in such a way as to allow for efficient transportation to other sites or to a permanent geologic repository. Prior to off-site shipment of the material, it may be stored on-site for a period of time. For this reason, it is desirable to have a container capable of meeting the requirements for storage as well as the requirements for transportation. Most of the off-site transportation is envisioned to take place using the TRUPACT-II Type B package, with the Waste Isolation Pilot Plant (WIPP) as the destination. Prior to the development of this new container, the TRUPACT-II had a limit of 325 FGE (fissile gram equivalents) of plutonium due to criticality control concerns. Because of the relatively high plutonium content in the material to be transported, transporting 325 FGE per TRUPACT-II is uneconomical. Thus, the purpose of the new containers is to provide criticality control to increase the allowed TRUPACT-II payload and to provide a safe method for on-site storage prior to transport. This paper will describe the analysis and testing used to demonstrate that the Pipe Overpack Container provides safe on-site storage of plutonium bearing materials in unhardened buildings and provides criticality control during transportation within the TRUPACT-II. Analyses included worst-case criticality analyses, analyses of fork-lift time impacts, and analyses of roof structure collapse onto the container. Testing included dynamic crush tests, bare pipe impact tests, a 30-minute totally engulfing pool-fire test, and multiple package impact tests in end-on and side-on orientations.

Ammerman, D. [Sandia National Labs., Albuquerque, NM (United States); Geinitz, R.; Thorp, D. [Safe Sites of Colorado, Golden, CO (United States); Rivera, M. [Los Alamos Technology Associates, Golden, CO (United States)

1998-03-01T23:59:59.000Z

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

June 2005 Proper Packaging Required to Maintain ...  

Science Conference Proceedings (OSTI)

... as follows: 1. Wrap in multiple layers of plastic bubble packaging material; 2 ... are another type of standard that is difficult to safely package. ...

2010-12-16T23:59:59.000Z

342

TORT certification package  

Science Conference Proceedings (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

343

Water Transport in PEM Fuel Cells: Advanced Modeling, Material...  

NLE Websites -- All DOE Office Websites (Extended Search)

against * steady state and transient operational cell data. Complete fuel cell water transport model improvements * and code package development to include two phase flow....

344

AGING PERFORMANCE OF MODEL 9975 PACKAGE FLUOROELASTOMER O-RINGS  

Science Conference Proceedings (OSTI)

The influence of temperature and radiation on Viton{reg_sign} GLT and GLT-S fluoroelastomer O-rings is an ongoing research focus at the Savannah River National Laboratory. The O-rings are credited for leaktight containment in the Model 9975 shipping package used for transportation of plutonium-bearing materials. At the Savannah River Site, the Model 9975 packages are being used for interim storage. Primary research efforts have focused on surveillance of O-rings from actual packages, leak testing of seals at bounding aging conditions and the effect of aging temperature on compression stress relaxation behavior, with the goal of service life prediction for long-term storage conditions. Recently, an additional effort to evaluate the effect of aging temperature on the oxidation of the materials has begun. Degradation in the mechanical properties of elastomers is directly related to the oxidation of the polymer. Sensitive measurements of the oxidation rate can be performed in a more timely manner than waiting for a measurable change in mechanical properties, especially at service temperatures. Measuring the oxidation rate therefore provides a means to validate the assumption that the degradation mechanisms(s) do not change from the elevated temperatures used for accelerated aging and the lower service temperatures. Monitoring the amount of oxygen uptake by the material over time at various temperatures can provide increased confidence in lifetime predictions. Preliminary oxygen consumption analysis of a Viton GLT-based fluoroelastomer compound (Parker V0835-75) using an Oxzilla II differential oxygen analyzer in the temperature range of 40-120 C was performed. Early data suggests oxygen consumption rates may level off within the first 100,000 hours (10-12 years) at 40 C and that sharp changes in the degradation mechanism (stress-relaxation) are not expected over the temperature range examined. This is consistent with the known long-term heat aging resistance of fluoroelastomers relative to hydrocarbon-based elastomers, and in absence of antioxidants that may be consumed over time. Additional experimental effort will be undertaken in the short term range within the first 100 hours of thermal aging to capture further details of the oxygen consumption rate.

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

2011-05-31T23:59:59.000Z

345

Voting System Testing  

Science Conference Proceedings (OSTI)

... required to be accredited to all core test methods involving: 1) technical data package review; 2) physical configuration audit (including examination ...

2013-09-06T23:59:59.000Z

346

Solar test of an integrated sodium reflux heat-pipe receiver/reactor for thermochemical energy transport  

DOE Green Energy (OSTI)

In October 1987, a chemical reactor integrated into a sodium reflux heat-pipe receiver was tested in the solar furnace at the Weizmann Institute of Science, Rehovot, Israel. The reaction carried out was the carbon dioxide reforming of methane. This reaction is one of the leading candidates for thermochemical energy transport either within a distributed solar receiver system or over long distances. The Schaeffer Solar Furnace consists of a 96 square meter heliostat and a 7.3 meter diameter dish concentrator with a 65-degree rim angle and a 3.5 meter focal length. Measurements have shown a peak concentration ratio of over 10,000 and a total power of 15 kW at an insolation of 800 w/square meter. The receiver/reactor contains seven catalyst-filled tubes inside an evacuated metal box containing sodium. The front surface of this box serves as the solar absorber of the receiver. In operation, concentrated sunlight heats the 1/8-inch Inconel plate and vaporizes sodium from the wire-mesh wick attached to the back of it. The sodium vapor condenses on the reactor tubes, releases its latent heat, and returns by gravity to the wick. Test results and areas for future development are discussed.

Diver, R.B.; Fish, J.D.; Levitan, R.; Levy, M.; Rosin, H.; Richardson, J.T.

1988-01-01T23:59:59.000Z

347

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

348

DORT certification package  

SciTech Connect

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

349

Expanded Content Envelope For The Model 9977 Packaging  

SciTech Connect

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.

2013-07-30T23:59:59.000Z

350

Transportation Issues  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Issues Issues and Resolutions - Compilation of Laboratory Transportation Work Package Reports Prepared for U.S. Department of Energy Used Fuel Disposition Campaign Compiled by Paul McConnell Sandia National Laboratories September 30, 2012 FCRD-UFD-2012-000342 Transportation Issues and Resolutions ii September 2012 Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000. DISCLAIMER This information was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any

351

Theoretical and experimental determination of matrix diffusion and related solute transport properties of fractured tuffs from the Nevada Test Site  

SciTech Connect

Theoretical and experimental studies of the chemical and physical factors which affect molecular diffusion of dissolved substances from fractures into a tuffaceous rock matrix have been made on rocks from G-Tunnel and Yucca Mountain at the Nevada Test Site (NTS). A variety of groundwater tracers, which may be useful in field tests at the NTS, have also been developed and tested. Although a number of physical/chemical processes may cause nonconvective transport of dissolved species from fractures into the tuff matrix, molecular diffusion seems to be the most important process. Molecular diffusion in these rocks is controlled by the composition of the groundwater through multicomponent effects and several rock properties. The porosities of the samples studied ranged from about 0.1 to 0.4. The constrictivity-tortuosity parameter ranged from 0.1 and 0.3 and effective matrix-diffusion coefficients were measured to be between 2 to 17. x 10{sup -7} c,{sup 2}/s for sodium halides and sodium pentafluorobenzoate. Total porosity was found to be the principle factor accounting for the variation in effective diffusion coefficients. The constrictivity-tortuosity factor was found to have a fair correlation (r = 0.75) with the median pore diameters measured by mercury intrusion. Measurements of bulk-rock electrical impedance changes with frequency indicate that the constrictivity factor has a maximum value of 0.8 to 1, but may be smaller. If the larger values are correct, then the diffusion paths in tuff are more tortuous than in granular media. Computation of the full diffusion-coefficient matrix for various tracers in J-13 well water from the NTS indicates coupling of the diffusion fluxes of all ionic species. These effects are being incorporated into a numerical model of multicomponent-matrix diffusion.

Walter, G.R.

1982-10-01T23:59:59.000Z

352

Safety Analysis Report for Packaging (SARP) for USA/5790/BLF (DOE-AL) and USA/5791/BLF (DOE-AL)  

Science Conference Proceedings (OSTI)

This revised Safety Analysis Report for Packaging (SARP) includes discussions of structural integrity, thermal resistance, radiation shielding and radiological safety, nuclear criticality safety, and quality control of shipping containers. Much of the information was previously submitted to AEC/OSD/ALO and the Department of Transportation (DOT) and provided the basis for obtaining special permits DOT-SP-5790 and DOT-SP-5791 as well as the Interim Certificates of Compliance until the original SARP could be prepared and Certificates of Compliance issued by ERDA. This SARP revision incorporates information on certain design changes, the most significant of which relate to the inner container for the type 5790 package. Complete physical and technical descriptions of the packages are presented. Each package consists of a cylindrical steel inner container centered within an insulating steel drum assembly. The contents may be any radioactive materials which satisfy the requirements established in this SARP. A shipment of plutonium-238 in the form of a solid oxide is evaluated in this SARP as an example. The results of the nuclear criticality safety analysis show how much of the fissile isotopes may be shipped as Fissile Class I, II, or III for each container. Design and development considerations, the tests and evaluations required to prove the ability of the containers to withstand normal transportation conditions, and the sequence of four hypothetical accident conditions (free drop, puncture, thermal, and water immersion) are discussed. Tables, graphs, dimensional sketches, photographs, technical references, loading and shipping procedures, Mound Facility experience in using the containers, and copies of the DOE Certificates of Compliance are included. Internal reviews of the original and revised SARP's have been performed in compliance with the requirement of DOEM 5201-Part V.

Roome, L.G.; Watkins, R.A.; Bertram, R.E.; Kreider, H.B.

1980-01-25T23:59:59.000Z

353

MODEL 9975 LIFE EXTENSION PACKAGE 1 - FINAL REPORT  

Science Conference Proceedings (OSTI)

Life extension package LE1 (9975-03382) was instrumented and subjected to a temperature/humidity environment that bounds KAMS package storage conditions for 92 weeks. During this time, the maximum fiberboard temperature was {approx}180 F, and was established by a combination of internal heat (12 watts) and external heat ({approx}142 F). The relative humidity external to the package was maintained at 80 %RH. This package was removed from test in November 2010 after several degraded conditions were observed during a periodic examination. These conditions included degraded fiberboard (easily broken, bottom layer stuck to the drum), corrosion of the drum, and separation of the air shield from the upper fiberboard assembly. Several tests and parameters were used to characterize the package components. Results from these tests generally indicate agreement between this full-scale shipping package and small-scale laboratory tests on fiberboard and O-ring samples. These areas of agreement include the rate of fiberboard weight loss, change in fiberboard thermal conductivity, fiberboard compression strength, and O-ring compression set. In addition, this package provides an example of the extent to which moisture within the fiberboard can redistribute in the presence of a temperature gradient such as might be created by a 12 watt internal heat load. Much of the moisture near the fiberboard ID surface migrated towards the OD surface, but there was not a significant axial moisture gradient during most of the test duration. Only during the last inspection period (i.e. after 92 weeks exposure during the second phase) did enough moisture migrate to the bottom fiberboard layers to cause saturation. A side effect of moisture migration is the leaching of soluble compounds from the fiberboard. In particular, the corrosion observed on the drum appears related primarily to the leaching and concentration of chlorides. In most locations, this attack appears to be general corrosion, with shallow attack of the drum surface. The primary areas susceptible to corrosion are weld/heat-affected zones. However, one corrosion location not immediately associated with a weld was tested for, and confirmed as having, throughwall penetration. An increase in the axial gap at the top of the package is also related to the migration of moisture within the fiberboard. As moisture redistributes within the package, a majority of the fiberboard loses moisture (on average) and shrinks axially. In addition, an increased moisture content of the bottom fiberboard layers locally reduces its compression strength, leading to compaction of those layers under the weight of the package internal components. In addition to these moisture-driven phenomena, the fiberboard will shrink due to slow pyrolysis in an elevated temperature/humidity environment. Under the collective influence of these effects, the axial gap in the LE1 package increased from an initial value of 0.58 inch and exceeded the 1 inch (maximum) criterion after approximately 18 weeks conditioning. The axial gap eventually reached 1.86 inches. Despite the degradation seen in several of the package components (drum, fiberboard, shield and O-rings), the package appears to have retained sufficient integrity to meet the functional requirements for storage in KAMS. This demonstrates a degree of robustness in the package design relative to the storage environment.

Daugherty, W.

2011-03-04T23:59:59.000Z

354

Assessing Potential Exposure from Truck Transport of Low-level Radioactive Waste to the Nevada Test Site  

Science Conference Proceedings (OSTI)

Since 1980, over 651,558 m{sup 3} (23,000,000 ft{sup 3}) of low-level radioactive waste (LLW) have been disposed of at the Nevada Test Site (NTS) by shallow land burial. Since 1988, the majority of this waste has been generated at other United States (U.S.) Department of Energy (DOE) and Department of Defense (DoD) sites and facilities in the U.S. Between fiscal year (FY) 2002 and the publication date, the volumes of LLW being shipped by truck to the NTS increased sharply with the accelerated closure of DOE Environmental Management (EM) Program sites (DOE, 2002). The NTS is located 105 km (65 mi) northwest of Las Vegas, Nevada, in the U.S. There continue to be public concerns over the safety of LLW shipments to the NTS. They can be broadly divided into two categories: (1) the risk of accidents involving trucks traveling on public highways; and (2) whether residents along transportation routes receive cumulative exposure from individual LLW shipments that pose a long-term health risk. The 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 a perceived risk from members of the public about cumulative exposure, particularly when ''Main Street'' and the routes being used by LLW trucks are one in the same. To provide an objective assessment of gamma radiation exposure to members of the public from LLW transport by truck, the Desert Research Institute (DRI) and the DOE, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) established a stationary and automated array of four pressurized ion chambers (PICs) in a vehicle pullout for LLW trucks to pass through just outside the entrance to the NTS. The PICs were positioned at a distance of 1.0 m (3.3 ft) from the sides of the truck trailer and at a height of 1.5 m (5.0 ft) to simulate conditions that a member of the public (Turner, 1995) might experience if a truck were to pass while the person was on the side of the road, or if a truck were to come to a stop at a stoplight in one of the smaller towns along the transportation routes. The 1.0-m (3.3-ft) distance also allowed for comparison with gamma readings of trucks taken with portable, hand-held instruments at the two LLW disposal sites at the NTS: the Area 5 Radioactive Waste Management Complex (RWMC) and the Area 3 Radioactive Waste Management Site (RWMS). The purpose in automating the system was to provide the most objective and consistent measurement and calculation of radiation exposure from the trucks possible. The array was set up in November 2002 and equipment was tested and calibrated over the next two months. Data collection on trucks began on February 13, 2003, and continued to the end of December 2003. In all, external gamma readings were collected from 1,012 of the 2,260 trucks that delivered LLW to the NTS during this period. Because DOE could not contractually require waste generators to participate in the study, the database is biased toward voluntary participants; however, data were collected from the 10 generators that represented 92 percent of the LLW shipments to the NTS during the study period, with another eight generators accounting for the balance of the shipments. Because of the voluntary nature of the participation, the identity of the waste generators is not used in the report. Previous studies on potential exposure to the public from transporting LLW to the NTS either relied on calculated exposures (Davis et al., 2002) or was based on a small population of trucks (e.g., 88) where a relatively high-background value of 50 microRoentgens per hour (R/h) (background value measured at the LLW disposal sites) were subtracted from the gross reading of the truck trailer as measured by portable, handheld instruments (Gertz, 2001). The dataset that resulted from the DRI study is the largest collection of measurements of LLW trucks in transit of which the authors are aware.

J. Miller; D. Shafer; K. Gray; B. Church; S. Campbell; B. Holz

2005-08-01T23:59:59.000Z

355

Assessing Potential Exposure from Truck Transport of Low-level Radioactive Waste to the Nevada Test Site  

Science Conference Proceedings (OSTI)

Since 1980, over 651,558 m{sup 3} (23,000,000 ft{sup 3}) of low-level radioactive waste (LLW) have been disposed of at the Nevada Test Site (NTS) by shallow land burial. Since 1988, the majority of this waste has been generated at other United States (U.S.) Department of Energy (DOE) and Department of Defense (DoD) sites and facilities in the U.S. Between fiscal year (FY) 2002 and the publication date, the volumes of LLW being shipped by truck to the NTS increased sharply with the accelerated closure of DOE Environmental Management (EM) Program sites (DOE, 2002). The NTS is located 105 km (65 mi) northwest of Las Vegas, Nevada, in the U.S. There continue to be public concerns over the safety of LLW shipments to the NTS. They can be broadly divided into two categories: (1) the risk of accidents involving trucks traveling on public highways; and (2) whether residents along transportation routes receive cumulative exposure from individual LLW shipments that pose a long-term health risk. The 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 a perceived risk from members of the public about cumulative exposure, particularly when ''Main Street'' and the routes being used by LLW trucks are one in the same. To provide an objective assessment of gamma radiation exposure to members of the public from LLW transport by truck, the Desert Research Institute (DRI) and the DOE, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) established a stationary and automated array of four pressurized ion chambers (PICs) in a vehicle pullout for LLW trucks to pass through just outside the entrance to the NTS. The PICs were positioned at a distance of 1.0 m (3.3 ft) from the sides of the truck trailer and at a height of 1.5 m (5.0 ft) to simulate conditions that a member of the public (Turner, 1995) might experience if a truck were to pass while the person was on the side of the road, or if a truck were to come to a stop at a stoplight in one of the smaller towns along the transportation routes. The 1.0-m (3.3-ft) distance also allowed for comparison with gamma readings of trucks taken with portable, hand-held instruments at the two LLW disposal sites at the NTS: the Area 5 Radioactive Waste Management Complex (RWMC) and the Area 3 Radioactive Waste Management Site (RWMS). The purpose in automating the system was to provide the most objective and consistent measurement and calculation of radiation exposure from the trucks possible. The array was set up in November 2002 and equipment was tested and calibrated over the next two months. Data collection on trucks began on February 13, 2003, and continued to the end of December 2003. In all, external gamma readings were collected from 1,012 of the 2,260 trucks that delivered LLW to the NTS during this period. Because DOE could not contractually require waste generators to participate in the study, the database is biased toward voluntary participants; however, data were collected from the 10 generators that represented 92 percent of the LLW shipments to the NTS during the study period, with another eight generators accounting for the balance of the shipments. Because of the voluntary nature of the participation, the identity of the waste generators is not used in the report. Previous studies on potential exposure to the public from transporting LLW to the NTS either relied on calculated exposures (Davis et al., 2002) or was based on a small population of trucks (e.g., 88) where a relatively high-background value of 50 microRoentgens per hour ({micro}R/h) (background value measured at the LLW disposal sites) were subtracted from the gross reading of the truck trailer as measured by portable, handheld instruments (Gertz, 2001). The dataset that resulted from the DRI study is the largest collection of measurements of LLW trucks in transit of which the authors are aware.

Miller, J; Shafer, D; Gray, K; Church, B; Campbell, S; Holtz, B.

2005-08-15T23:59:59.000Z

356

National Transportation Stakeholders Forum  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Transportation Stakeholders Forum Transportation Stakeholders Forum May 14-16, 2013 Tuesday, May 14 7:00 am - 5:00 pm Registration Niagara Foyer 7:00 am - 7:45 am Breakfast and Networking Grand A 8:00 am - 10:00 am National Updates for Transportation Stakeholder Groups and Guests - Panel Grand BC Moderator: John Giarrusso Jr., MA Emergency Management Agency / Northeast High-Level Radioactive Waste Transportation Task Force Co-Chair US Department of Energy, Office of Environmental Management - Steve O'Connor, Director, Office of Packaging & Transportation US Nuclear Regulatory Commission - Earl P. Easton, Senior Level Advisor (retired) and David W. Pstrak, Transportation and Storage Specialist, Division of Spent Fuel Storage and Transportation

357

AMPX-77 Phase 1 certification package  

Science Conference Proceedings (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

358

Spring 2012 National Transportation Stakeholder Forum Meetings, Tennessee |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Services » Waste Management » Packaging and Transportation » Services » Waste Management » Packaging and Transportation » National Transportation Stakeholders Forum » Spring 2012 National Transportation Stakeholder Forum Meetings, Tennessee Spring 2012 National Transportation Stakeholder Forum Meetings, Tennessee NTSF Registration Website Save The Date! NTSF Spring 2012 Agenda NTSF Agenda Midwestern Radioactive Materials Transportation Committee Agenda Northeast High-Level Radioactive Waste Transportation Task Force Agenda Transuranic Waste Transportation Working Group Agenda Western Governor's Association Agenda NTSF Presentations Session Newcomers' Orientation Plenary Sessions Keynote Address Oak Ridge Operations Office of Environmental Management Overview Global Threat Reduction Initiative Task Force for Strategic Developments to Blue Ribbon Commission

359

MCNP certification package  

Science Conference Proceedings (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

360

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.


361

Package for integrated optic circuit and method  

DOE Patents (OSTI)

A structure and method are disclosed 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. 6 figs.

Kravitz, S.H.; Hadley, G.R.; Warren, M.E.; Carson, R.F.; Armendariz, M.G.

1998-08-04T23:59:59.000Z

362

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

Science Conference Proceedings (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

363

DOE - Safety of Radioactive Material Transportation  

NLE Websites -- All DOE Office Websites (Extended Search)

Immersion Comparison Demonstrating target hardness. Comparison of the Fire Test to a Gasoline Tanker and Sedan collision under an Overpass Fire Test [FIRE test] Click to view picture Real-life Accident Comparison [FIRE scenario] Click to view picture Real-life scenarios that the above test is designed to protect against include being involved in an accident with a gasoline tanker truck, causing the gasoline contents to burn the package. The amount of fuel being burned is approximately 5000 gallons in a pool 30 feet in diameter. During this test, the package is fully engulfed in the fire and is not protected by a transporting vehicle. On October 9, 1997, a truck tractor pulling a cargo tank semitrailer was going under an overpass of the New York State Thruway in Yonkers, New York when it was struck by a sedan. The car hit the right side of the cargo tank in the area of the tank's external loading/unloading lines, releasing the 8800 gallons of gasoline they contained.

364

INVESTIGATION OF THE PRESENCE OF DRUGSTORE BEETLES WITHIN CELOTEX ASSEMBLIES IN RADIOACTIVE MATERIAL PACKAGINGS  

Science Conference Proceedings (OSTI)

During normal operations at the Department of Energy's Hanford Site in Hanford, WA, drugstore beetles, (Stegobium paniceum (L.) Coleoptera: Anobiidae), were found within the fiberboard subassemblies of two 9975 Shipping Packages. Initial indications were that the beetles were feeding on the Celotex{trademark} assemblies within the package. Celotex{trademark} fiberboard is used in numerous radioactive material packages serving as both a thermal insulator and an impact absorber for both normal conditions of transport and hypothetical accident conditions. The Department of Energy's Packaging Certification Program (EM-63) directed a thorough investigation to determine if the drugstore beetles were causing damage that would be detrimental to the safety performance of the Celotex{trademark}. The Savannah River National Laboratory is conducting the investigation with entomological expertise provided by Clemson University. The two empty 9975 shipping packages were transferred to the Savannah River National Laboratory in the fall of 2007. This paper will provide details and results of the ongoing investigation.

Loftin, B; Glenn Abramczyk, G

2008-06-04T23:59:59.000Z

365

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

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

366

Microelectronics plastic molded packaging  

Science Conference Proceedings (OSTI)

The use of commercial off-the-shelf (COTS) microelectronics for nuclear weapon applications will soon be reality rather than hearsay. The use of COTS for new technologies for uniquely military applications is being driven by the so-called Perry Initiative that requires the U.S. Department of Defense (DoD) to accept and utilize commercial standards for procurement of military systems. Based on this philosophy, coupled with several practical considerations, new weapons systems as well as future upgrades will contain plastic encapsulated microelectronics. However, a conservative Department of Energy (DOE) approach requires lifetime predictive models. Thus, the focus of the current project is on accelerated testing to advance current aging models as well as on the development of the methodology to be used during WR qualification of plastic encapsulated microelectronics. An additional focal point involves achieving awareness of commercial capabilities, materials, and processes. One of the major outcomes of the project has been the definition of proper techniques for handling and evaluation of modern surface mount parts which might be used in future systems. This program is also raising the familiarity level of plastic within the weapons complex, allowing subsystem design rules accommodating COTS to evolve. A two year program plan is presented along with test results and commercial interactions during this first year.

Johnson, D.R. [Ktech Corp., Albuquerque, NM (United States); Palmer, D.W.; Peterson, D.W. [Sandia National Lab., Albuquerque, NM (United States)] [and others

1997-02-01T23:59:59.000Z

367

Hanford Site radioactive hazardous materials packaging directory  

SciTech Connect

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

368

Transportation technology at Sandia  

SciTech Connect

Industrial and military activities in the US produce large amounts of hazardous mixed waste, which includes both radioactive and toxic substances. The already overburdened environment is faced with the task of safely disposing of these complex wastes. A very important aspect of this effort is the safe and economical transportation of radioactive and toxic chemical wastes to projected repositories. Movement of wastes to the repository sites is accomplished by a combination of truck, rail, ship, and air. The DOE directs transportation activities including cask development technology for use in single or multimode transport. Sandia National Laboratories` Transportation Technology programs provide the technology and know-how to support DOE in achieving safe, efficient, and economical packaging and transportation of nuclear and other hazardous waste materials. This brochure describes the Transportation Technology programs and the specialized techniques and capabilities they offer to prospective users.

1994-12-31T23:59:59.000Z

369

Performance oriented packaging report for M6 electric blasting cap. Final report  

SciTech Connect

This POP report is for the M6 Electric Blasting Cap which is packaged 180/ Mil-B-2427 wood box. This report describes the results of testing conducted. Performance Oriented Packaging, POP, M6 Electric Blasting Cap, Mil-B-2427 wood box.

Sniezek, F.M.

1992-11-02T23:59:59.000Z

370

Performance oriented packaging report for M7 non-electric blasting cap. Final report  

Science Conference Proceedings (OSTI)

This POP report is for the M7 Non-Electric Blasting Cap which is packaged 480/Mil-B-2427 wood box. This report describes the results of testing conducted. Performance Oriented Packaging, POP, M7 Non-Electric Blasting Cap, Mil-B-2427 Wood box.

Sniezek, F.M.

1992-11-02T23:59:59.000Z

371

Performance oriented packaging report for ignitor, time blasting fuse, weatherproof: M60. Final report  

Science Conference Proceedings (OSTI)

This POP report is for the Time Blasting Fuse, Weatherproof: M60 which is packaged 300/ Mil-B-2427 wood box. This report describes the results of testing conducted.... Performance oriented packaging, POP, Time blasting fuse, Weatherproof: M60 Mil-B-2427 wood box.

Sniezek, F.

1992-11-02T23:59:59.000Z

372

NIST Handbook 133 Checking the Net Contents of Packaged ...  

Science Conference Proceedings (OSTI)

... 4. Except for aerosol or other pressurized packages, open the sample packages, empty, clean, and dry them as appropriate for the packaging ...

2012-11-14T23:59:59.000Z

373

Investigations into High Temperature Components and Packaging  

SciTech Connect

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

374

Technical Data Package Standards Development Summit  

Science Conference Proceedings (OSTI)

Technical Data Package Standards Development Summit. Purpose: ... Breakout Groups (session will begin after lunch): · Data Delivery/Content. ...

2011-10-11T23:59:59.000Z

375

One-to-Many Multimodal Fusion Package  

Science Conference Proceedings (OSTI)

The One-to-many Multimodal Fusion Package. Participants from the Iris Exchange (IREX) III Evaluation and the Multibiometrics ...

2012-04-05T23:59:59.000Z

376

Model Based Enterprise / Technical Data Package Summit ...  

Science Conference Proceedings (OSTI)

Page 1. NIST Technical Note 1753 Model Based Enterprise / Technical Data Package Summit Report Joshua Lubell Kenway ...

2012-10-22T23:59:59.000Z

377

MST: Organizations: Thin Film, Vacuum, and Packaging  

NLE Websites -- All DOE Office Websites (Extended Search)

Processes & Services Electronic Fabrication Manufacturing Process Science & Technology Thin Film, Vacuum, & Packaging Organic Materials Ceramic & Glass Meso Manufacturing &...

378

Vadose Zone Hydrogeology Data Package for Hanford Assessments  

SciTech Connect

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

379

Biodegradable hydrogel film for food packaging  

Science Conference Proceedings (OSTI)

Disposal of waste plastic packaging materials has raised a serious problem worldwide leading to environmental pollution due to the fact that most of the plastic packaging materials are generally non-biodegradable. In this article we have reported about ... Keywords: biodegradable, biopolymer, breathable, compost, hydrogel, packaging

Niladri Roy; Nabanita Saha; Petr Saha

2011-07-01T23:59:59.000Z

380

Principles of Package Design Bertrand Meyer  

E-Print Network (OSTI)

List(lastpage) Fig. 4. Structure of the Manuals. USER / \\ APPLICATION / PROGRAM / ~ PACKAGE ~ , ~ SYSTEM / (compiler Issues The programming language for writing a package should offer a structure corresponding a program structure ("class" in Simula and "package" in Ada) with three cate- gories of elements: data

Meyer, Bertrand

Note: This page contains sample records for the topic "transport packaging tests" from the National Library of EnergyBeta (NLEBeta).
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We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

The todonotes package Henrik Skov Midtiby  

E-Print Network (OSTI)

The todonotes package Henrik Skov Midtiby henrikmidtiby@gmail.com December 25, 2009 Abstract The todonotes package allows you to insert to­do items in your docu- ment. At any point in the document a list.1 Usage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.2 Package options

Hoffmann, Rolf

382

Intelligent Transportation Systems - Center for Transportation Analysis  

NLE Websites -- All DOE Office Websites (Extended Search)

Intelligent Transportation Systems Intelligent Transportation Systems The Center for Transportation Analysis does specialty research and development in intelligent transportation systems. Intelligent Transportation Systems (ITS) are part of the national strategy for improving the operational safety, efficiency, and security of our nation's highways. Since the early 1990s, ITS has been the umbrella under which significant efforts have been conducted in research, development, testing, deployment and integration of advanced technologies to improve the measures of effectiveness of our national highway network. These measures include level of congestion, the number of accidents and fatalities, delay, throughput, access to transportation, and fuel efficiency. A transportation future that includes ITS will involve a significant improvement in these

383

SHIPMENT OF TWO DOE-STD-3013 CONTAINERS IN A 9977 TYPE B PACKAGE  

Science Conference Proceedings (OSTI)

The 9977 is a certified Type B Packaging authorized to ship uranium and plutonium in metal and oxide forms. Historically, the standard container for these materials has been the DOE-STD-3013 which was specifically designed for the long term storage of plutonium bearing materials. The Department of Energy has used the 9975 Packaging containing a single 3013 container for the transportation and storage of these materials. In order to reduce container, shipping, and storage costs, the 9977 Packaging is being certified for transportation and storage of two 3013 containers. The challenges and risks of this content and the 9977s ability to meet the Code of Federal Regulations for the transport of these materials are presented.

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

2011-06-06T23:59:59.000Z

384

Packaging and Transfer of Hazardous Materials and Materials of National Security Interest Assessment plan - Developed By NNSA/Nevada Site Office Facility Representative Division  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

PACKAGING AND TRANSFER PACKAGING AND TRANSFER OF HAZARDOUS MATERIALS AND MATERIALS OF NATIONAL SECURITY INTEREST Assessment Plan NNSA/Nevada Site Office Facility Representative Division Performance Objective: Verify that packaging and transportation safety requirements of hazardous materials and materials of national security interest have been established and are in compliance with DOE Orders 461.1 and 460.1B Criteria: Verify that safety requirements for the proper packaging and transportation of DOE/NNSA offsite shipments and onsite transfers of hazardous materials and for modal transport have been established [DOE O 460.1B, 1, "Objectives"]. Verify that the contractor transporting a package of hazardous materials is in compliance with the requirements of the Hazardous Materials Regulations

385

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

Science Conference Proceedings (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

386

UML's Package Extension Mechanism: Taking a Closer Look at Package Merge  

E-Print Network (OSTI)

. Since package merge was designed specically to be able to structure the UML metamodel, it is not clearUML's Package Extension Mechanism: Taking a Closer Look at Package Merge by Alanna Pauline Zito #12;Abstract The UML 2 specication introduced the notion of package merge as a means of den- ing

387

Harsh-Environment Packaging for Downhole Gas and Oil Exploration  

Science Conference Proceedings (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

388

Transportation of RCRA hazardous wastes. RCRA Information Brief  

Science Conference Proceedings (OSTI)

The Resource Conservation and Recovery Act (RCRA) and the Hazardous Materials Transportation Act (HMTA) regulate the transport of hazardous wastes. Under these statutes, specific pretransport regulatory requirements must be met by DOE before the shipment of hazardous wastes, including radioactive mixed wastes. The pretransport requirements are designed to help reduce the risk of loss, leakage, or exposure during shipment of hazardous materials and to communicate information on potential hazards posed by the hazardous material in transport. These goals are accomplished through the tracking of shipments, correctly packaging and labeling containers, and communicating potential hazards. Specific requirements include manifesting, packaging, marking and labeling waste packages; placarding transport vehicles; choosing appropriate waste transporters and shipment destinations; and record keeping and reporting. This information Brief focuses primarily on the transporter requirements both for transportation within a DOE facility and using a commercial transporter to transport RCRA hazardous wastes off-site.

Not Available

1994-04-01T23:59:59.000Z

389

Waste Package Design Methodology Report  

Science Conference Proceedings (OSTI)

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

D.A. Brownson

2001-09-28T23:59:59.000Z

390

RECLAMATION OF RADIOACTIVE MATERIAL PACKAGING COMPONENTS  

SciTech Connect

Radioactive material packages are withdrawn from use for various reasons; loss of mission, decertification, damage, replacement, etc. While the packages themselves may be decertified, various components may still be able to perform to their required standards and find useful service. The Packaging Technology and Pressurized Systems group of the Savannah River National Laboratory has been reducing the cost of producing new Type B Packagings by reclaiming, refurbishing, and returning to service the containment vessels from older decertified packagings. The program and its benefits are presented.

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

2011-06-06T23:59:59.000Z

391

Technical Review Report for the Application for Contents Amendment for Shipping Isentropic Compression Experiment (ICE) Apparatus in 9977 Packaging  

Science Conference Proceedings (OSTI)

This report documents the review of Application for Contents Amendment for Shipping Isentropic Compression Experiment (ICE) Apparatus in 9977 Packaging, prepared by Savannah River Packaging Technology (SRPT) of Savannah River National Laboratory (SRNL) of Savannah River Nuclear Solutions, LLC, -- the Submittal -- at the request of the Department of Energy's (DOE) National Nuclear Security Agency's (NNSA) Albuquerque Facility Operations Division, for the shipment of the ICE apparatus from Los Alamos National Laboratory (LANL), to Sandia National Laboratory (SNL). The ICE apparatus consists of a stainless steel assembly containing about 8 grams of {sup 239}Pu or its dose equivalent as noted in Table 1, Comparison of 9977 Content C.1 and the ICE Radioactive Contents, of the Submittal. The ICE target is mounted on the transport container assembly base. A Viton{sup R} O-ring seals the transport container base to the transport container body. Another Viton{sup R} O-ring seals the transport container handle to the transport container body. The ICE apparatus weighs less than 30 pounds and has less than 0.6 watts decay heat rate. For the Model 9977 Package, the maximum payload weight is 100 pounds and the maximum decay heat rate is 19 watts. Thus, the maximum payload weight and the maximum decay heat rate for the Model 9977 Package easily bound those for the ICE apparatus. This Addendum supplements the Safety Analysis Report for Packaging (SARP), Revision 2, for the Model 9977 Package and Addendum 1, Revision 2, to Revision 2 of the Model 9977 Package SARP. The ICE apparatus is considered as part of Content Envelope C.6, Samples and Sources, under the submittal for the Model 9978 Package SARP currently under review. The Staff at Lawrence Livermore National Laboratory (LLNL) recommends that the Submittal be approved by the DOE-Headquarters Certifying Official (EM-60), and incorporated into a subsequent revision to the current Certificate of Compliance (CoC), to the Model 9977-96 Packaging.

West, M

2009-04-16T23:59:59.000Z

392

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

DOE Green Energy (OSTI)

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

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

1995-11-01T23:59:59.000Z

393

RECERTIFICATION OF THE MODEL 9977 RADIOACTIVE MATERIAL PACKAGING  

SciTech Connect

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

394

Fleet servicing facilities for servicing, maintaining, and testing rail and truck radioactive waste transport systems: functional requirements, technical design concepts and options cost estimates and comparisons  

Science Conference Proceedings (OSTI)

This is a resource document which examines feasibility design concepts and feasibility studies of a Fleet Servicing Facility (FSF). Such a facility is intended to be used for routine servicing, preventive maintenance, and for performing requalification license compliance tests and inspections, minor repairs, and decontamination of both the transportation casks and their associated rail cars or tractor-trailers. None of the United States' waste handling plants presently receiving radioactive wastes have an on-site FSF, nor is there an existing third party facility providing these services. This situation has caused the General Accounting Office to express concern regarding the quality of waste transport system maintenance once the system is placed into service. Thus, a need is indicated for FSF's, or their equivalent, at various radioactive materials receiving sites. In this report, three forms of FSF's solely for spent fuel transport systems were examined: independent, integrated, and colocated. The independent concept was already the subject of a detailed report and is extensively referenced in this document so that capital cost comparisons of the three concepts could be made. These facilities probably could service high-level, intermediate-level, low-level, or other waste transportation systems with minor modification, but this study did not include any system other than spent fuel. Both the Integrated and Colocated concepts were assumed to be associated with some radioactive materials handling facility such as an AFR repository.

Watson, C.D.; Hudson, B.J.; Keith, D.A.; Preston, M.K. Jr.; McCreery, P.N.; Knox, W.; Easterling, E.M.; Lamprey, A.S.; Wiedemann, G.

1980-05-01T23:59:59.000Z

395

Safety evaluation for packaging (onsite) nitrogen trailers propane tanks  

SciTech Connect

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

396

TEST  

Science Conference Proceedings (OSTI)

This is an abstract. TEST Lorem ipsum dolor sit amet, consectetur adipiscing elit. Cras lacinia dui et est venenatis lacinia. Vestibulum lacus dolor, adipiscing id mattis sit amet, ultricies sed purus. Nulla consectetur aliquet feugiat. Maecenas ips

397

AGC-2 Graphite Preirradiation Data Package  

SciTech Connect

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

David Swank; Joseph Lord; David Rohrbaugh; William Windes

2012-10-01T23:59:59.000Z

398

Transportation of Commercial Spent Nuclear Fuel  

Science Conference Proceedings (OSTI)

The U.S. industrys limited efforts at licensing transportation packages characterized as high-capacity, or containing high-burnup (>45 GWd/MTU) commercial spent nuclear fuel (CSNF), or both, have not been successful considering existing spent-fuel inventories that will have to be eventually transported. A holistic framework is proposed for resolving several CSNF transportation issues. The framework considers transportation risks, spent-fuel and cask-design features, and defense-in-depth in context of pre...

2010-12-10T23:59:59.000Z

399

Subgrid Scale Physics in 1-Month Forecasts. Part I: Experiment with Four Parameterization Packages  

Science Conference Proceedings (OSTI)

Four packages of subgrid scale (SGS) physics parameterization are tested by including them in a general circulation model and by applying the four models to 1-month forecasts. The four models are formulated by accumulatively increasing the ...

J. Sirutis; K. Miyakoda

1990-05-01T23:59:59.000Z

400

9975 SHIPPING PACKAGE LIFE EXTENSION SURVEILLANCE PROGRAM RESULTS SUMMARY  

SciTech Connect

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

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401

Transportation Infrastructure  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Infrastructure Infrastructure New Technologies * Potential need for dual-use casks * DOE should look toward industry & international communities for innovations * Industry unclear about delivery & receipt locations * Advances in physical & tracking technologies need to be factored in * Cost-benefit analysis of new technology Training & Dry Runs * Begin as soon as possible * Suggested order: #1-demonstrations, #2-training, #3-dry-runs * Don't re-invent the wheel- look at international programs * Allows DOE to test POC info/training * Standardization of training & materials * DOE should consider centralized training center * Use real equipment in dry- runs * Need for regionalized dry runs Packages * Full-scale Testing - Funds requested in 2003, potential use of

402

Software realization problems of mathematical models of pollutants transport in rivers  

Science Conference Proceedings (OSTI)

A software package of realization of mathematical models of pollutants transport in rivers is offered. This package is designed as a up-to-date convenient, reliable tool for specialists of various areas of knowledge such as ecology, hydrology, building, ... Keywords: Difference scheme, Mathematical model, Software package, Solution accuracy, Time of solution

K. J. Kachiashvili; D. I. Melikdzhanian

2009-10-01T23:59:59.000Z

403

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

SciTech Connect

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

404

Communication Is Key to Packaging and Transportation Safety and Compliance  

Energy.gov (U.S. Department of Energy (DOE))

Presentation made by Steve O'Connor for the NTSF annual meeting held from May 14-16, 2013 in Buffalo, NY

405

NFR TRIGA package design review report  

SciTech Connect

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

406

Safety evaluation for packaging 222-S laboratory cargo tank for onetime type B material shipment  

Science Conference Proceedings (OSTI)

The purpose of this Safety Evaluation for Packaging (SEP) is to evaluate and document the safety of the onetime shipment of bulk radioactive liquids in the 222-S Laboratory cargo tank (222-S cargo tank). The 222-S cargo tank is a US Department of Transportation (DOT) MC-312 specification (DOT 1989) cargo tank, vehicle registration number HO-64-04275, approved for low specific activity (LSA) shipments in accordance with the DOT Title 49, Code of Federal Regulations (CFR). In accordance with the US Department of Energy, Richland Operations Office (RL) Order 5480.1A, Chapter III (RL 1988), an equivalent degree of safety shall be provided for onsite shipments as would be afforded by the DOT shipping regulations for a radioactive material package. This document demonstrates that this packaging system meets the onsite transportation safety criteria for a onetime shipment of Type B contents.

Nguyen, P.M.

1994-08-19T23:59:59.000Z

407

The Underground Test Area Project of the Nevada Test Site: Building Confidence in Groundwater Flow and Transport Models at Pahute Mesa Through Focused Characterization Studies  

SciTech Connect

Pahute Mesa at the Nevada Test Site contains about 8.0E+07 curies of radioactivity caused by underground nuclear testing. The Underground Test Area Subproject has entered Phase II of data acquisition, analysis, and modeling to determine the risk to receptors from radioactivity in the groundwater, establish a groundwater monitoring network, and provide regulatory closure. Evaluation of radionuclide contamination at Pahute Mesa is particularly difficult due to the complex stratigraphy and structure caused by multiple calderas in the Southwestern Nevada Volcanic Field and overprinting of Basin and Range faulting. Included in overall Phase II goals is the need to reduce the uncertainty and improve confidence in modeling results. New characterization efforts are underway, and results from the first year of a three-year well drilling plan are presented.

Pawloski, G A; Wurtz, J; Drellack, S L

2009-12-29T23:59:59.000Z

408

FAQS Qualification Card - NNSA Package Certification Engineer |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

NNSA Package Certification Engineer NNSA Package Certification Engineer FAQS Qualification Card - NNSA Package Certification Engineer A key element for the Department's Technical Qualification Programs is a set of common Functional Area Qualification Standards (FAQS) and associated Job Task Analyses (JTA). These standards are developed for various functional areas of responsibility in the Department, including oversight of safety management programs identified as hazard controls in Documented Safety Analyses (DSA). For each functional area, the FAQS identify the minimum technical competencies and supporting knowledge and skills for a typical qualified individual working in the area. FAQC-NNSAPackageCertificationEngineer.docx Description NNSA Package Certification Engineer Qualification Card

409