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Sample records for sites verification package

  1. Cleanup Verification Package for the 300-18 Waste Site

    International Nuclear Information System (INIS)

    Capron, J.M.

    2005-01-01

    This cleanup verification package documents completion of remedial action for the 300-18 waste site. This site was identified as containing radiologically contaminated soil, metal shavings, nuts, bolts, and concrete

  2. Cleanup Verification Package for the 300 VTS Waste Site

    International Nuclear Information System (INIS)

    Clark, S.W.; Mitchell, T.H.

    2006-01-01

    This cleanup verification package documents completion of remedial action for the 300 Area Vitrification Test Site, also known as the 300 VTS site. The site was used by Pacific Northwest National Laboratory as a field demonstration site for in situ vitrification of soils containing simulated waste

  3. Cleanup Verification Package for the 300 VTS Waste Site

    Energy Technology Data Exchange (ETDEWEB)

    S. W. Clark and T. H. Mitchell

    2006-03-13

    This cleanup verification package documents completion of remedial action for the 300 Area Vitrification Test Site, also known as the 300 VTS site. The site was used by Pacific Northwest National Laboratory as a field demonstration site for in situ vitrification of soils containing simulated waste.

  4. Cleanup Verification Package for the 600-47 Waste Site

    International Nuclear Information System (INIS)

    Cutlip, M.J.

    2005-01-01

    This cleanup verification package documents completion of interim remedial action for the 600-47 waste site. This site consisted of several areas of surface debris and contamination near the banks of the Columbia River across from Johnson Island. Contaminated material identified in field surveys included four areas of soil, wood, nuts, bolts, and other metal debris

  5. Cleanup Verification Package for the 600-259 Waste Site

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2006-02-09

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

  6. Cleanup Verification Package for the 600-259 Waste Site

    International Nuclear Information System (INIS)

    Capron, J.M.

    2006-01-01

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

  7. Remaining Sites Verification Package for the 128-B-3 Burn Pit Site, Waste Site Reclassification Form 2006-058

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2006-11-17

    The 128-B-3 waste site is a former burn and disposal site for the 100-B/C Area, located adjacent to the Columbia River. The 128-B-3 waste site has been remediated to meet the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results of sampling at upland areas of the site also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  8. Remaining Sites Verification Package for 100-F-38 Stained Soil Site. Attachment to Waste Site Reclassification Form 2004-093

    International Nuclear Information System (INIS)

    Carlson, R.A.

    2006-01-01

    The 100-F-38 Stained Soil site was an area of yellow stained soil that was discovered while excavating a trench for the placement of electrical conduit. The 100-F-38 Stained Soil site meets the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling show demonstrate that residual contaminant concentrations support future unrestricted land uses that can be represented by a rural-residential scenario. The results also show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils and the contaminant concentrations remaining in the soil are protective of groundwater and the Columbia River

  9. Remaining Sites Verification Package for the 1607-F3 Sanitary Sewer System, Waste Site Reclassification Form 2006-047

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2007-04-26

    The 1607-F3 waste site is the former location of the sanitary sewer system that supported the 182-F Pump Station, the 183-F Water Treatment Plant, and the 151-F Substation. The sanitary sewer system included a septic tank, drain field, and associated pipeline, all in use between 1944 and 1965. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  10. Packaged low-level waste verification system

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  11. Remaining Sites Verification Package for the 126-F-2, 183-F Clearwells, Waste Site Reclassification Form 2006-017

    Energy Technology Data Exchange (ETDEWEB)

    R. A. Carlson

    2006-05-04

    The 126-F-2 site is the clearwell facility formerly used as part of the reactor cooling water treatment at the 183-F facility. During demolition operations in the 1970s, potentially contaminated debris was disposed in the eastern clearwell structure. The site has been remediated by removing all debris in the clearwell structure to the Environmental Restoration Disposal Facility. The results of radiological surveys and visual inspection of the remediated clearwell structure show neither residual contamination nor the potential for contaminant migration beyond the clearwell boundaries. The results of verification sampling at the remediation waste staging area demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  12. Remaining Sites Verification Package for the 126-F-2, 183-F Clearwells, Waste Site Reclassification Form 2006-017

    International Nuclear Information System (INIS)

    Carlson, R.A.

    2006-01-01

    The 126-F-2 site is the clearwell facility formerly used as part of the reactor cooling water treatment at the 183-F facility. During demolition operations in the 1970s, potentially contaminated debris was disposed in the eastern clearwell structure. The site has been remediated by removing all debris in the clearwell structure to the Environmental Restoration Disposal Facility. The results of radiological surveys and visual inspection of the remediated clearwell structure show neither residual contamination nor the potential for contaminant migration beyond the clearwell boundaries. The results of verification sampling at the remediation waste staging area demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River

  13. Packaged low-level waste verification system

    International Nuclear Information System (INIS)

    Tuite, K.T.; Winberg, M.; Flores, A.Y.; Killian, E.W.; McIsaac, C.V.

    1996-01-01

    Currently, states and low-level radioactive waste (LLW) disposal site operators have no method of independently verifying the radionuclide content of packaged LLW that arrive at disposal sites for disposal. At this time, disposal sites rely on LLW generator shipping manifests and accompanying records to insure that LLW received meets the waste acceptance criteria. An independent verification system would provide a method of checking generator LLW characterization methods and help ensure that LLW disposed of at disposal facilities meets requirements. The Mobile Low-Level Waste Verification System (MLLWVS) provides the equipment, software, and methods to enable the independent verification of LLW shipping records to insure that disposal site waste acceptance criteria are being met. The MLLWVS system was developed under a cost share subcontract between WMG, Inc., and Lockheed Martin Idaho Technologies through the Department of Energy's National Low-Level Waste Management Program at the Idaho National Engineering Laboratory (INEL)

  14. Remaining Sites Verification Package for the 100-D-2 Lead Sheeting Waste Site. Attachment to Waste Site Reclassification Form 2007-030

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2008-01-01

    The 100-D-2 Lead Sheeting waste site was located approximately 50 m southwest of the 185-D Building and approximately 16 m north of the east/west oriented road. The site consisted of a lead sheet covering a concrete pad. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  15. Remaining Sites Verification Package for the 126-B-2, 183-B Clearwells

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2007-01-01

    The 126-B-2, 183-B Clearwells were built as part of the 183-B Water Treatment Facility and are composed of 2 covered concrete reservoirs. The bulk of the water stored in the clearwells was used as process water to cool the 105-B Reactor and as a source of potable water. Residual conditions were determined to meet the remedial action objectives specified in the Remaining Sites ROD through an evaluation of the available process knowledge. The results of the evaluation do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also indicate that residual concentrations are protective of groundwater and the Columbia River.

  16. Remaining Sites Verification Package for the 600-233 Waste Site, Vertical Pipe Near 100-B Electrical Laydown Area. Attachment to Waste Site Reclassification Form 2005-041

    International Nuclear Information System (INIS)

    Carlson, R.A.

    2005-01-01

    The 600-233 waste site consisted of three small-diameter pipelines within the 600-232 waste site, including previously unknown diesel fuel supply lines discovered during site remediation. The 600-233 waste site has been remediated to achieve the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  17. Remaining Sites Verification Package for the 128-F-2, 100-F Burning Pit Waste Site. Attachment to Waste Site Reclassification Form 2008-031

    International Nuclear Information System (INIS)

    Capron, J.M.

    2008-01-01

    The 128-F-2 waste site consisted of multiple burn and debris filled pits located directly east of the 107-F Retention Basin and approximately 30.5 m east of the northeast corner of the 100-F Area perimeter road that runs along the riverbank. The burn pits were used for incinerating nonradioactive, combustible materials from 1945 to 1965. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  18. Remaining Sites Verification Package for the 128-F-2, 100-F Burning Pit Waste Site, Waste Site Reclassification Form 2008-031

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2008-12-01

    The 128-F-2 waste site consisted of multiple burn and debris filled pits located directly east of the 107-F Retention Basin and approximately 30.5 m east of the northeast corner of the 100-F Area perimeter road that runs along the riverbank. The burn pits were used for incinerating nonradioactive, combustible materials from 1945 to 1965. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  19. Remaining Sites Verification Package for the 1607-D4 Septic System. Attachment to Waste Site Reclassification Form 2005-036

    International Nuclear Information System (INIS)

    Carlson, R.A.

    2006-01-01

    The 1607-D4 Septic System was a septic tank and tile field that received sanitary sewage from the 115-D/DR Gas Recirculation Facility. This septic system operated from 1944 to 1968. Decommissioning took place in 1985 and 1986 when all above-grade features were demolished and the tank backfilled. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River

  20. Remaining Sites Verification Package for the 120-F-1 Glass Dump Waste Site. Attachment to Waste Site Reclassification Form 2008-028

    International Nuclear Information System (INIS)

    Capron, J.M.

    2008-01-01

    The 120-F-1 waste site consisted of two dumping areas located 660 m southeast of the 105-F Reactor containing laboratory equipment and bottles, demolition debris, light bulbs and tubes, small batteries, small drums, and pesticide contaminated soil. It is probable that 108-F was the source of the debris but the material may have come from other locations within the 100-F Area. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  1. Remaining Sites Verification Package for the 128-B-2, 100-B Burn Pit No.2 Waste Site. Attchment to Waste Site Reclassification Form 2005-038

    International Nuclear Information System (INIS)

    Carlson, R.A.

    2005-01-01

    The 128-B-2 waste site was a burn pit historically used for the disposal of combustible and noncombustible wastes, including paint and solvents, office waste, concrete debris, and metallic debris. This site has been remediated by removing approximately 5,627 bank cubic meters of debris, ash, and contaminated soil to the Environmental Restoration Disposal Facility. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River

  2. Remaining Sites Verification Package for the 100-B-26 Spillway. Attachment to Waste Site Reclassification Form 2006-052

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2006-01-01

    The 100-B-26 Spillway waste site is a spillway that served as an emergency discharge point for the 132-C-2 outfall in the event that the 100-B-15 river effluent pipelines were blocked, damaged, or undergoing maintenance. The selected action involved demonstrating through confirmatory sampling that cleanup goals have been met and proposing a reclassification of this site to No Action. The results of the confirmatory sampling demonstrate that residual contaminant concentrations remaining in the soil beneath the riprap are more protective of groundwater and the Columbia River than the risk they would pose if the site was remediated

  3. Remaining Sites Verification Package for the 1607-B1 Septic System. Attachment to Waste Site Reclassification Form 2007-015

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2007-01-01

    The 1607-B1 Septic System includes a septic tank, drain field, and associated connecting pipelines and influent sanitary sewer lines. This septic system serviced the former 1701-B Badgehouse, 1720-B Patrol Building/Change Room, and the 1709-B Fire Headquarters. The 1607-B1 waste site received unknown amounts of nonhazardous, nonradioactive sanitary sewage from these facilities during its operational history from 1944 to approximately 1970. In accordance with this evaluation, the confirmatory sampling results support a reclassification of this site to No Action. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  4. Remaining Sites Verification Package for the 100-B-24 Spillway. Attachment to Waste Site Reclassification Form 2006-051

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2006-01-01

    The 100-B-24 Spillway is a spillway that was designed to serve as an emergency discharge point for the 116-B-7 outfall in the event that the 100-B-15 river effluent pipelines were blocked, damaged, or undergoing maintenance. The site meets the remedial action objectives specified in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  5. Remaining Sites Verification Package for the 600-243 Petroleum-Contaminated Soil Bioremediation Pad. Attachment to Waste Site Reclassification Form 2007-033

    International Nuclear Information System (INIS)

    Capron, J.M.

    2008-01-01

    The 600-243 waste site consisted of a bioremediation pad for petroleum-contaminated soils resulting from the 1100 Area Underground Storage Tank (UST) upgrades in 1994. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  6. Cleanup Verification Package for the 618-2 Burial Ground

    Energy Technology Data Exchange (ETDEWEB)

    W. S. Thompson

    2006-12-28

    This cleanup verification package documents completion of remedial action for the 618-2 Burial Ground, also referred to as Solid Waste Burial Ground No. 2; Burial Ground No. 2; 318-2; and Dry Waste Burial Site No. 2. This waste site was used primarily for the disposal of contaminated equipment, materials and laboratory waste from the 300 Area Facilities.

  7. Cleanup Verification Package for the 618-2 Burial Ground

    International Nuclear Information System (INIS)

    Thompson, W.S.

    2006-01-01

    This cleanup verification package documents completion of remedial action for the 618-2 Burial Ground, also referred to as Solid Waste Burial Ground No. 2; Burial Ground No. 2; 318-2; and Dry Waste Burial Site No. 2. This waste site was used primarily for the disposal of contaminated equipment, materials and laboratory waste from the 300 Area Facilities

  8. Remaining Sites Verification Package for the 100-F-31, 144-F Sanitary Sewer System, Waste Site Reclassification Form 2006-033

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2006-08-24

    The 100-F-31 waste site is a former septic system that supported the inhalation laboratories, also referred to as the 144-F Particle Exposure Laboratory (132-F-2 waste site), which housed animals exposed to particulate material. The 100-F-31 waste site has been remediated to achieve the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  9. Remaining Sites Verification Package for the 100-F-26:9, 1607-F2 Sanitary Sewer Pipelines. Attachment to Waste Site Reclassification Form 2008-029

    International Nuclear Information System (INIS)

    Capron, J.M.

    2008-01-01

    The 100-F-26:9 underground pipeline subsite consists of the sanitary sewers servicing the 105-F, 108-F, 184-F, 185-F, and 190-F buildings, and the 1700-F administration and service buildings (1704-F, 1707-F, 1707-FA, 1713-F, 1717-F, 1719-F, and 1722-F). In accordance with this evaluation, the confirmatory and verification sampling results support a reclassification of this site to Interim Closed Out. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  10. Remaining Sites Verification Package for the 116-C-3, 105-C Chemical Waste Tanks. Attachment to Waste Site Reclassification Form 2008-002

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2008-01-01

    The 116-C-3 waste site consisted of two underground storage tanks designed to receive mixed waste from the 105-C Reactor Metals Examination Facility chemical dejacketing process. Confirmatory evaluation and subsequent characterization of the site determined that the southern tank contained approximately 34,000 L (9,000 gal) of dejacketing wastes, and that the northern tank was unused. In accordance with this evaluation, the verification sampling and modeling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrate that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also show that residual contaminant concentrations are protective of groundwater and the Columbia River

  11. Remaining Sites Verification Package for the 1607-F4 Sanitary Sewer System. Attachment to Waste Site Reclassification Form 2004-131

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2007-01-01

    The 1607-F4 waste site is the former location of the sanitary sewer system that serviced the former 115-F Gas Recirculation Building. The system included a septic tank, drain field, and associated pipeline that were in use from 1944 to 1965. The 1607-F4 waste site received unknown amounts of sanitary sewage from the 115-F Gas Recirculation Building and may have potentially contained hazardous and radioactive contamination. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River

  12. Remaining Sites Verification Package for the 100-F-26:13, 108-F Drain Pipelines. Attchment to Waste Site Reclassification Form 2005-011

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2008-01-01

    The 100-F-26:13 waste site is the network of process sewer pipelines that received effluent from the 108-F Biological Laboratory and discharged it to the 188-F Ash Disposal Area (126-F-1 waste site). The pipelines included one 0.15-m (6-in.)-, two 0.2-m (8-in.)-, and one 0.31-m (12-in.)-diameter vitrified clay pipe segments encased in concrete. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River

  13. Remaining Sites Verification Package for the 126-B-3, 184-B Coal Pit Dumping Area, Waste Site Reclassification Form 2005-028

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2006-08-07

    The 126-B-3 waste site is the former coal storage pit for the 184-B Powerhouse. During demolition operations in the 1970s, the site was used for disposal of demolition debris from 100-B/C Area facilities. The site has been remediated by removing debris and contaminated soils. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  14. Remaining Sites Verification Package for the 100-F-33, 146-F Aquatic Biology Fish Ponds, Waste Site Reclassification Form 2006-021

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2006-08-25

    The 100-F-33, 146-F Aquatice Biology Fish Ponds waste site was an area with six small rectangular ponds and one large circular pond used to conduct tests on fish using various mixtures of river and reactor effluent water. The current site conditions achieve the remedial action objectives specified in the Remaining Sites ROD. The results of verification and applicable confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  15. Remaining Sites Verification Package for the 100-F-33, 146-F Aquatic Biology Fish Ponds. Attachment to Waste Site Reclassification Form 2006-021

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2006-01-01

    The 100-F-33, 146-F Aquatice Biology Fish Ponds waste site was an area with six small rectangular ponds and one large circular pond used to conduct tests on fish using various mixtures of river and reactor effluent water. The current site conditions achieve the remedial action objectives specified in the Remaining Sites ROD. The results of verification and applicable confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  16. Remaining Sites Verification Package for the 126-B-3, 184-B Coal Pit Dumping Area. Attachment to Waste Site Reclassification Form 2005-028

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2006-01-01

    The 126-B-3 waste site is the former coal storage pit for the 184-B Powerhouse. During demolition operations in the 1970s, the site was used for disposal of demolition debris from 100-B/C Area facilities. The site has been remediated by removing debris and contaminated soils. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River

  17. Cleanup Verification Package for the 118-F-6 Burial Ground

    International Nuclear Information System (INIS)

    Sulloway, H.M.

    2008-01-01

    This cleanup verification package documents completion of remedial action for the 118-F-6 Burial Ground located in the 100-FR-2 Operable Unit of the 100-F Area on the Hanford Site. The trenches received waste from the 100-F Experimental Animal Farm, including animal manure, animal carcasses, laboratory waste, plastic, cardboard, metal, and concrete debris as well as a railroad tank car

  18. Remaining Sites Verification Package for the 120-B-1, 105-B Battery Acid Sump. Attachment to Waste Site Reclassification Form 2006-057

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2006-01-01

    The 120-B-1 waste site, located in the 100-BC-1 Operable Unit of the Hanford Site, consisted of a concrete battery acid sump that was used from 1944 to 1969 to neutralize the spent sulfuric acid from lead cell batteries of emergency power packs and the emergency lighting system. The battery acid sump was associated with the 105-B Reactor Building and was located adjacent to the building's northwest corner. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River

  19. Remaining Sites Verification Package for the 100-B-18, 184-B Powerhouse Debris Pile. Attachment to Waste Site Reclassification Form 2007-020

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2007-01-01

    The 100-B-18 Powerhouse Debris Pile contained miscellaneous demolition waste from the decommissioning activities of the 184-B Powerhouse. The debris covered an area roughly 15 m by 30 m and included materials such as concrete blocks, mixed aggregate/concrete slabs, stone rubble, asphalt rubble, traces of tar/coal, broken fluorescent lights, brick chimney remnants, and rubber hoses. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  20. Remaining Sites Verification Package for the 1607-B2 Septic System and 100-B-14:2 Sanitary Sewer System, Waste Site Reclassification Form 2006-055

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2007-03-21

    The 1607-B2 waste site is a former septic system associated with various 100-B facilities, including the 105-B, 108-B, 115-B/C, and 185/190-B buildings. The site was evaluated based on confirmatory results for feeder lines within the 100-B-14:2 subsite and determined to require remediation. The 1607-B2 waste site has been remediated to achieve the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  1. Remaining Sites Verification Package for the 1607-F7, 141-M Building Septic Tank. Attachment to Waste Site Reclassification Form 2006-040

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2006-01-01

    The 1607-F7, 141-M Building Septic Tank waste site was a septic tank and drain field that received sanitary sewage from the former 141-M Building. Remedial action was performed in August and November 2005. The results of verification sampling demonstrate that residual contaminant concentrations support future unrestricted land uses that can be represented by a rural-residential scenario. These results also show that residual concentrations support unrestricted future use of shallow zone soil and that contaminant levels remaining in the soil are protective of groundwater and the Columbia River

  2. Remaining Sites Verification Package for the 100-C-9:2 Sanitary Sewer Pipelines, Waste Site Reclassification Form 2004-013

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2007-07-11

    The 100-C-9:2 sanitary sewer pipelines include the feeder pipelines associated with the 1607-B8, the 1607-B9, the 1607-B10 and the 1607-B11 septic systems. Contaminated soil and piping from the feeder lines to the septic systems were removed and disposed of. The remaining soil in the excavations has been shown to meet the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  3. Remaining Sites Verification Package for the 100-C-9:2 Sanitary Sewer Pipelines. Attachment to Waste Site Reclassification Form 2004-013

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2007-01-01

    The 100-C-9:2 sanitary sewer pipelines include the feeder pipelines associated with the 1607-B8, the 1607-B9, the 1607-B10 and the 1607-B11 septic systems. Contaminated soil and piping from the feeder lines to the septic systems were removed and disposed of. The remaining soil in the excavations has been shown to meet the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  4. Inspection and verification of waste packages for near surface disposal

    International Nuclear Information System (INIS)

    2000-01-01

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

  5. SCALE criticality safety verification and validation package

    International Nuclear Information System (INIS)

    Bowman, S.M.; Emmett, M.B.; Jordan, W.C.

    1998-01-01

    Verification and validation (V and V) are essential elements of software quality assurance (QA) for computer codes that are used for performing scientific calculations. V and V provides a means to ensure the reliability and accuracy of such software. As part of the SCALE QA and V and V plans, a general V and V package for the SCALE criticality safety codes has been assembled, tested and documented. The SCALE criticality safety V and V package is being made available to SCALE users through the Radiation Safety Information Computational Center (RSICC) to assist them in performing adequate V and V for their SCALE applications

  6. Remaining Sites Verification Package for the 1607-F5 Sanitary Sewer System (124-F-5). Attachment to Waste Site Reclassification Form 2006-043

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2006-01-01

    The 1607-F5 waste site is a former septic tank, tile field, and associated pipeline located within the 100-FR-1 Operable Unit that received sewage from the former 181-F Pumphouse. Lead, gamma-chlordane, and heptachlor epoxide were identified within or around the septic system at concentrations exceeding the direct exposure cleanup criteria. Multiple metal and pesticide constituents were also identified as exceeding the groundwater and river protection cleanup criteria. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River

  7. Remaining Sites Verification Package for the 100-D-9 Boiler Fuel Oil Tank Site. Attachment to Waste Site Reclassification Form 2006-030

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2006-01-01

    The 100-D-9 site is the former location of an underground storage tank used for holding fuel for the 184-DA Boiler House. Results of soil-gas samples taken from six soil-gas probes in a rectangle around the site the tank had been removed from concluded that there were no volatile organic compounds at detectable levels in the area. The 100-D-9 Boiler Fuel Oil Tank Site meets the remedial action objectives specified in the Remaining Sites ROD. The results demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  8. Cleanup Verification Package for the 618-8 Burial Ground

    International Nuclear Information System (INIS)

    Appel, M.J.

    2006-01-01

    This cleanup verification package documents completion of remedial action for the 618-8 Burial Ground, also referred to as the Solid Waste Burial Ground No. 8, 318-8, and the Early Solid Waste Burial Ground. During its period of operation, the 618-8 site is speculated to have been used to bury uranium-contaminated waste derived from fuel manufacturing, and construction debris from the remodeling of the 313 Building

  9. Cleanup Verification Package for the 618-3 Burial Ground

    International Nuclear Information System (INIS)

    Appel, M.J.

    2006-01-01

    This cleanup verification package documents completion of remedial action for the 618-3 Solid Waste Burial Ground, also referred to as Burial Ground Number 3 and the Dry Waste Burial Ground Number 3. During its period of operation, the 618-3 site was used to dispose of uranium-contaminated construction debris from the 311 Building and construction/demolition debris from remodeling of the 313, 303-J and 303-K Buildings

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    Capron, J.M.

    2006-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    E. J. Farris and H. M. Sulloway

    2008-01-10

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

  13. 300 Area Process Trenches Verification Package

    International Nuclear Information System (INIS)

    Lerch, J.A.

    1998-03-01

    The purpose of this verification package is to document achievement of the remedial action objectives for the 300 Area Process Trenches (300 APT) located within the 300-FF-1 Operable Unit (OU). The 300 APT became active in 1975 as a replacement for the North and South Process Pond system that is also part of the 300-FF-1 OU. The trenches received 300 Area process effluent from the uranium fuel fabrication facilities. Waste from the 300 Area laboratories that was determined to be below discharge limits based on monitoring performed at the 307 retention basin was also released to the trenches. Effluent flowed through the headworks sluice gates, down a concrete apron, and into the trenches. From the beginning of operations in 1975 until 1993, a continuous, composite sampler was located at the headwork structure to analyze process effluent at the point of discharge to the environment

  14. Remaining Sites Verification Package for the 1607-B2 Septic System and 100-B-14:2 Sanitary Sewer System, Waste Site Reclassification Form 2004-006

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2007-03-21

    The 100-B-14:2 subsite encompasses the former sanitary sewer feeder lines associated with the 1607-B2 and 1607-B7 septic systems. Feeder lines associated with the 185/190-B building have also been identified as the 100-B-14:8 subsite, and feeder lines associated with the 1607-B7 septic system have also been identified as the 100-B-14:9 subsite. These two subsites have been administratively cancelled to resolve the redundancy. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  15. Remaining Sites Verification Package for 132-DR-1, 1608-DR Effluent Pumping Station. Attachment to Waste Site Reclassification Form 2005-035

    International Nuclear Information System (INIS)

    Carlson, R.A.

    2005-01-01

    Radiological characterization, decommissioning and demolition of the 132-DR-1 site, 1608-DR Effluent Pumping Station was performed in 1987. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. Residual concentrations support future land uses that can be represented by a rural-residential scenario and pose no threat to groundwater or the Columbia River based on RESRAD modeling

  16. Remaining Sites Verification Package for the 100-B-1 Surface Chemical and Solid Waste Dumping Area,. Attachment to Waste Site Reclassification Form 2006-003

    International Nuclear Information System (INIS)

    Carlson, R.A.

    2006-01-01

    The 100-B-1 waste site was a dumping site that was divided into two areas. One area was used as a laydown area for construction materials, and the other area was used as a chemical dumping area. The 100-B-1 Surface Chemical and Solid Waste Dumping Area site meets the remedial action objectives specified in the Remaining Sites ROD. The results demonstrate that residual contaminant concentrations support future unrestricted land uses that can be represented by a rural-residential scenario. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  17. Remaining Sites Verification Package for the 600-111, P-11 Critical Mass Laboratory Crib, and UPR-600-16, Fire and Contamination Spread Waste Sites. Attachment to Waste Site Reclassification Form 2004-065

    International Nuclear Information System (INIS)

    Capron, J.M.

    2008-01-01

    The 600-111, P-11 Critical Mass Laboratory Crib waste site, also referred to as the P-11 Facility, included the 120 Experimental Building, the 123 Control Building, and the P-11 Crib. The facility was constructed in 1949 and was used as a laboratory for plutonium criticality studies. In accordance with this evaluation, the confirmatory and verification sampling results support a reclassification of this site to Interim Closed Out. The results of confirmatory and verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  18. Remaining Sites Verification Package for the 100-F-52, 146-FR Radioecology and Aquatic Biology Laboratory Soil. Attachment to Waste Site Reclassification Form 2008-022

    International Nuclear Information System (INIS)

    Capron, J.M.

    2008-01-01

    The 100-F-52 waste site consisted of the soil under and around the former 146-FR Radioecology and Aquatic Biology Laboratory. The laboratory was used for studies of the effects of pre-reactor and post-reactor process water on fish eggs, young fish, and other small river creatures of interest. In accordance with this evaluation, the confirmatory sampling results support a reclassification of this site to No Action. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  19. Remaining Sites Verification Package for the 100-B-20, 1716-B Maintenance Garage Underground Tank, Waste Site Reclassification Form 2006-019

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2006-09-27

    The 100-B-20 waste site, located in the 100-BC-1 Operable Unit of the Hanford Site, consisted of an underground oil tank that once serviced the 1716-B Maintenance Garage. The selected action for the 100-B-20 waste site involved removal of the oil tanks and their contents and demonstrating through confirmatory sampling that all cleanup goals have been met. In accordance with this evaluation, a reclassification status of interim closed out has been determined. The results demonstrate that the site will support future unrestricted land uses that can be represented by a rural-residential scenario. These results also show that residual concentrations support unrestricted future use of shallow zone soil and that contaminant levels remaining in the soil are protective of groundwater and the Columbia River.

  20. Remaining Sites Verification Package for the 100-F-52, 146-FR Radioecology and Aquatic Biology Laboratory Soil, Waste Site Reclassification Form 2008-022

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2008-06-27

    The 100-F-52 waste site consisted of the soil under and around the former 146-FR Radioecology and Aquatic Biology Laboratory. The laboratory was used for studies of the effects of pre-reactor and post-reactor process water on fish eggs, young fish, and other small river creatures of interest. In accordance with this evaluation, the confirmatory sampling results support a reclassification of this site to No Action. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  1. Remaining Sites Verification Package for the 100-B-20, 1716-B Maintenance Garage Underground Tank. Attachment to Waste Site Reclassification Form 2006-019

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2006-01-01

    The 100-B-20 waste site, located in the 100-BC-1 Operable Unit of the Hanford Site, consisted of an underground oil tank that once serviced the 1716-B Maintenance Garage. The selected action for the 100-B-20 waste site involved removal of the oil tanks and their contents and demonstrating through confirmatory sampling that all cleanup goals have been met. In accordance with this evaluation, a reclassification status of interim closed out has been determined. The results demonstrate that the site will support future unrestricted land uses that can be represented by a rural-residential scenario. These results also show that residual concentrations support unrestricted future use of shallow zone soil and that contaminant levels remaining in the soil are protective of groundwater and the Columbia River

  2. Remaining Sites Verification Package for the 100-F-46, 119-F Stack Sampling French Drain. Attachment to Waste Site Reclassification Form 2008-021

    International Nuclear Information System (INIS)

    Capron, J.M.

    2008-01-01

    The 100-F-46 french drain consisted of a 1.5 to 3 m long, vertically buried, gravel-filled pipe that was approximately 1 m in diameter. Also included in this waste site was a 5 cm cast-iron pipeline that drained condensate from the 119-F Stack Sampling Building into the 100-F-46 french drain. In accordance with this evaluation, the confirmatory sampling results support a reclassification of this site to No Action. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  3. Remaining Sites Verification Package for the 100-F-44:2, Discovery Pipeline Near 108-F Building. Attachment to Waste Site Reclassification Form 2007-006

    International Nuclear Information System (INIS)

    Capron, J.M.

    2008-01-01

    The 100-F-44:2 waste site is a steel pipeline that was discovered in a junction box during confirmatory sampling of the 100-F-26:4 pipeline from December 2004 through January 2005. The 100-F-44:2 pipeline feeds into the 100-F-26:4 subsite vitrified clay pipe (VCP) process sewer pipeline from the 108-F Biology Laboratory at the junction box. In accordance with this evaluation, the confirmatory sampling results support a reclassification of this site to No Action. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  4. Remaining Sites Verification Package for the 331 Life Sciences Laboratory Drain Field Septic System. Attachment to Waste Site Reclassification Form 2008-020

    International Nuclear Information System (INIS)

    Capron, J.M.

    2008-01-01

    The 331 Life Sciences Laboratory Drain Field (LSLDF) septic system waste site consists of a diversion chamber, two septic tanks, a distribution box, and a drain field. This septic system was designed to receive sanitary waste water, from animal studies conducted in the 331-A and 331-B Buildings, for discharge into the soil column. However, field observations and testing suggest the 331 LSLDF septic system did not receive any discharges. In accordance with this evaluation, the confirmatory sampling results support a reclassification of the 331 LSLDF waste site to No Action. This site does not have a deep zone or other condition that would warrant an institutional control in accordance with the 300-FF-2 ROD under the industrial land use scenario

  5. Remaining Sites Verification Package for the 100-B-21:2 Subsite (100-B/C Discovery Pipeline DS-100BC-002). Attachment Waste Site Reclassification Form 2008-003

    International Nuclear Information System (INIS)

    Capron, J.M.

    2008-01-01

    The 100-B-21:2 waste site consists of the immediate area of the DS-100BC-02 pipeline. In accordance with this evaluation, the confirmatory and verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  6. Remaining Sites Verification Package for the 100-B-22:1 Pipelines and Associated Soils. Attachment to Waste Site Reclassification Form 2005-042

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2006-01-01

    The 100-B-22:1 pipelines and associated soils were part of the 100-B Area water treatment facilities. The 100-B-22:1 waste site is limited to those pipelines that interconnected the 185-B Filter House, the 126-B-2 Clearwells, the 185-B Deaeration Plant, and the 190-B Process Pumphouse. None of the 100-B-22:1 pipelines carried environmentally significant contamination. In accordance with the historical information and field observations of this evaluation, the results support a reclassification of this site to No Action required to meet future rural-residential uses and be protective of groundwater and the Columbia River

  7. Remaining Sites Verification Package for 132-D-3, 1608-D Effluent Pumping Station. Attchment to Waste Site Reclassification Form 2005-033

    International Nuclear Information System (INIS)

    Carlson, R.A.

    2006-01-01

    Decommissioning and demolition of the 132-D-3 site, 1608-D Effluent Pumping Station was performed in 1986. Decommissioning included removal of equipment, water, and sludge for disposal as radioactive waste. The at- and below-grade structure was demolished to at least 1 m below grade and the resulting rubble buried in situ. The area was backfilled to grade with at least 1 m of clean fill and contoured to the surrounding terrain. Residual concentrations support future land uses that can be represented by a rural-residential scenario and pose no threat to groundwater or the Columbia River based on RESRAD modeling

  8. Remaining Sites Verification Package for the 100-F-36, 108-F Biological Laboratory, and for the 116-F-15, 108-F Radiation Crib. Attachment to Waste Site Reclassification Form 2007-002 and 2007-003

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2007-01-01

    The 116-F-15 waste site is the former location of the 108-F Radiation Crib that was located in the first floor of the 108-F Biological Laboratory. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  9. Remaining Sites Verification Package for the 100-F-44:4, Discovery Pipeline in Silica Gel Pit. Attachment to Waste Site Reclassification Form 2008-030

    International Nuclear Information System (INIS)

    Capron, J.M.

    2008-01-01

    The 100-F-44:4, Discovery Pipeline in Silica Gel Pit subsite is located in the 100-FR-1 Operable Unit of the Hanford Site, near the location of the former 110-F Gas Storage Tanks structure. The 100-F-44:4 subsite is a steel pipe discovered October 17, 2004, during trenching to locate the 118-F-4 Silica Gel Pit. Based on visual inspection and confirmatory investigation sampling data, the 100-F-44:4 subsite is a piece of non-hazardous electrical conduit debris. The 100-F-44:4 subsite supports unrestricted future use of shallow zone soil and is protective of groundwater and the Columbia River. No residual contamination exists within the deep zone. Therefore, no deep zone institutional controls are required

  10. Cleanup Verification Package for the 118-B-6, 108-B Solid Waste Burial Ground

    International Nuclear Information System (INIS)

    Proctor, M.L.

    2006-01-01

    This cleanup verification package documents completion of remedial action for the 118-B-6, 108-B Solid Waste Burial Ground. The 118-B-6 site consisted of 2 concrete pipes buried vertically in the ground and capped by a concrete pad with steel lids. The site was used for the disposal of wastes from the 'metal line' of the P-10 Tritium Separation Project.

  11. Cleanup Verification Package for the 100-F-20, Pacific Northwest Laboratory Parallel Pits

    International Nuclear Information System (INIS)

    Appel, M.J.

    2007-01-01

    This cleanup verification package documents completion of remedial action for the 100-F-20, Pacific Northwest Laboratory Parallel Pits waste site. This waste site consisted of two earthen trenches thought to have received both radioactive and nonradioactive material related to the 100-F Experimental Animal Farm

  12. Remaining Sites Verification Package for the 100-F-26:10, 1607-F3 Sanitary Sewer Pipelines (182-F, 183-F, and 151-F Sanitary Sewer Lines), Waste Site Reclassification Form 2007-028

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2007-12-03

    The 100-F-26:10 waste site includes sanitary sewer lines that serviced the former 182-F, 183-F, and 151-F Buildings. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  13. Remaining Sites Verification Package for the 100-F-26:15 Miscellaneous Pipelines Associated with the 132-F-6, 1608-F Waste Water Pumping Station. Attachment to Waste Site Reclassification Form 2007-031

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2008-01-01

    The 100-F-26:15 waste site consisted of the remnant portions of underground process effluent and floor drain pipelines that originated at the 105-F Reactor. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  14. Remaining Sites Verification Package for the 100-F-26:10, 1607-F3 Sanitary Sewer Pipelines (182-F, 183-F, and 151-F Sanitary Sewer Lines). Attachment to Waste Site Reclassification Form 2007-028

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2007-01-01

    The 100-F-26:10 waste site includes sanitary sewer lines that serviced the former 182-F, 183-F, and 151-F Buildings. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  15. Remaining Sites Verification Package for the 600-111, P-11 Critical Mass Laboratory Crib, and UPR-600-16, Fire and Contamination Spread Waste Sites. Attachment to Waste Site Reclassification Form 2004-065 and 2008-045

    International Nuclear Information System (INIS)

    Capron, J.M.

    2008-01-01

    The UPR-600-16, Fire and Contamination Spread waste site is an unplanned release that occurred on December 4, 1951, when plutonium contamination was spread by a fire that ignited inside the 120 Experimental Building. The 120 Experimental Building was a laboratory building that was constructed in 1949 and used for plutonium criticality studies as part of the P-11 Project. In November 1951, a criticality occurred in the 120 Experimental Building that resulted in extensive plutonium contamination inside the building. The confirmatory evaluation supports a reclassification of this site to Interim Closed Out. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of the extensive radiological survey of the surface soil and the confirmatory and verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

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

    Energy Technology Data Exchange (ETDEWEB)

    M. J. Appel and J. M. Capron

    2007-07-25

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

  17. Cleanup Verification Package for the 126-F-1, 184-F Powerhouse Ash Pit

    International Nuclear Information System (INIS)

    Clark, S.W.; Sulloway, H.M.

    2007-01-01

    This cleanup verification package documents completion of remedial action for the 126-F-1, 184-F Powerhouse Ash Pit. This waste site received coal ash from the 100-F Area coal-fired steam plant. Leakage of process effluent from the 116-F-14 , 107-F Retention Basins flowed south into the ash pit, contaminating the northern portion

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

    International Nuclear Information System (INIS)

    Appel, M.J.

    2006-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    M. J. Appel

    2006-11-02

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

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

    International Nuclear Information System (INIS)

    Appel, M.J.

    2007-01-01

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

  2. Remaining Sites Verification Package for the 100-F-26:12, 1.8-m (72-in.) Main Process Sewer Pipeline. Attachment to Waste Site Reclassification Form 2007-034

    International Nuclear Information System (INIS)

    Capron, J.M.

    2008-01-01

    The 100-F-26:12 waste site was an approximately 308-m-long, 1.8-m-diameter east-west-trending reinforced concrete pipe that joined the North Process Sewer Pipelines (100-F-26:1) and the South Process Pipelines (100-F-26:4) with the 1.8-m reactor cooling water effluent pipeline (100-F-19). In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  3. Remaining Sites Verification Package for the 116-F-8, 1904-F Outfall Structure and the 100-F-42, 1904-F Spillway. Attachment to Waste Site Reclassification Form 2006-038 and 2006-045

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2006-01-01

    The 116-F-8 waste site is the former 1904-F Outfall Structure used to discharge reactor cooling water effluent from the 107-F Retention Basin to the Columbia River. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  4. Remaining Sites Verification Package for the 116-F-16, PNL Outfall and the 100-F-43, PNL Outfall Spillway. Attachment to Waste Site Reclassification Form 2006-039 and 2006-096

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2006-01-01

    The 116-F-16 waste site is the former Pacific National Laboratories (PNL) Outfall, used to discharge waste effluents from the 100-F Experimental Animal Farm. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  5. Remaining Sites Verification Package for the 1607-B2 Septic System and 100-B-14:2 Sanitary Sewer System. Attachment to Waste Site Reclassification Form 2006-055 and 2004-006

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2007-01-01

    The 1607-B2 waste site is a former septic system associated with various 100-B facilities, including the 105-B, 108-B, 115-B/C, and 185/190-B buildings. The site was evaluated based on confirmatory results for feeder lines within the 100-B-14:2 subsite and determined to require remediation. The 1607-B2 waste site has been remediated to achieve the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  6. Remaining Sites Verification Package for the 116-F-8, 1904-F Outfall Structure and the 100-F-42, 1904-F Spillway. Attachment to Waste Site Reclassification Form 2006-038 and 2006-045

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2006-01-01

    The 100-F-42 waste site is the portion of the former emergency overflow spillway for the 1904-F Outfall Structure formerly existing above the ordinary high water mark of the Columbia River. The spillway consisted of a concrete flume designed to discharge effluent from the 107-F Retention Basin in the event that flows could not be completely discharged via the river outfall pipelines. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  7. Remaining Sites Verification Package for the 116-F-16, PNL Outfall and the 100-F-43, PNL Outfall Spillway. Attachment to Waste Site Reclassification Form 2006-039 and 2006-046

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2006-01-01

    The 100-F-43 waste site is the portion of the former discharge spillway for the PNL Outfall formerly existing above the ordinary high water mark of the Columbia River. The spillway consisted of a concrete flume used to discharge waste effluents from the 100-F Experimental Animal Farm. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  8. Hanford Site radioactive hazardous materials packaging directory

    International Nuclear Information System (INIS)

    McCarthy, T.L.

    1995-12-01

    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 ampersand Development (PO ampersand 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

  9. Hanford Site radioactive hazardous materials packaging directory

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, T.L.

    1995-12-01

    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.

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

    International Nuclear Information System (INIS)

    Capron, J.M.

    2008-01-01

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

  11. Validation and verification of the MTRPC thermohydraulic package

    International Nuclear Information System (INIS)

    Doval, Alicia

    1998-01-01

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

  12. Remaining Sites Verification Package for the 1607-F1 Sanitary Sewer System (124-F-1) and the 100-F-26:8 (1607-F1) Sanitary Sewer Pipelines Waste Sites. Attachment to Waste Site Reclassification Forms 2004-130 and 2005-004

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2008-01-01

    The 100-F-26:8 waste site consisted of the underground pipelines that conveyed sanitary waste water from the 1701-F Gatehouse, 1709-F Fire Station, and the 1720-F Administrative Office to the 1607-F1 septic tank. The site has been remediated and presently exists as an open excavation. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River

  13. Verification test calculations for the Source Term Code Package

    International Nuclear Information System (INIS)

    Denning, R.S.; Wooton, R.O.; Alexander, C.A.; Curtis, L.A.; Cybulskis, P.; Gieseke, J.A.; Jordan, H.; Lee, K.W.; Nicolosi, S.L.

    1986-07-01

    The purpose of this report is to demonstrate the reasonableness of the Source Term Code Package (STCP) results. Hand calculations have been performed spanning a wide variety of phenomena within the context of a single accident sequence, a loss of all ac power with late containment failure, in the Peach Bottom (BWR) plant, and compared with STCP results. The report identifies some of the limitations of the hand calculation effort. The processes involved in a core meltdown accident are complex and coupled. Hand calculations by their nature must deal with gross simplifications of these processes. Their greatest strength is as an indicator that a computer code contains an error, for example that it doesn't satisfy basic conservation laws, rather than in showing the analysis accurately represents reality. Hand calculations are an important element of verification but they do not satisfy the need for code validation. The code validation program for the STCP is a separate effort. In general the hand calculation results show that models used in the STCP codes (e.g., MARCH, TRAP-MELT, VANESA) obey basic conservation laws and produce reasonable results. The degree of agreement and significance of the comparisons differ among the models evaluated. 20 figs., 26 tabs

  14. Remaining Sites Verification Package for the 100-F-36, 108-F Biological Laboratory, and for the 116-F-15, 108-F Radiation Crib. Attachment to Waste Site Reclassification Form 2007-002

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2007-01-01

    The 100-F-36 waste site is the location of the former 108-F Biological Laboratory. The building was closed in 1973, decontaminated, decommissioned, and eventually demolished in 1999. In accordance with this evaluation, the confirmatory sampling results support a reclassification of this site to No Action. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  15. Verification of RRC Ki code package for neutronic calculations of WWER core with GD

    International Nuclear Information System (INIS)

    Aleshin, S.S.; Bolshagin, S.N.; Lazarenko, A.P.; Markov, A.V.; Pavlov, V.I.; Pavlovitchev, A.M.; Sidorenko, V.D.; Tsvetkov, V.M.

    2001-01-01

    The report presented is concerned with verification results of TVS-M/PERMAK-A/BIPR-7A code package for WWERs neutronic calculation as applied to calculation of systems containing U-GD pins. The verification is based on corresponded benchmark calculations, data critical experiments and on operation data obtained WWER units with Gd. The comparison results are discussed (Authors)

  16. Verification of 3-D generation code package for neutronic calculations of WWERs

    International Nuclear Information System (INIS)

    Sidorenko, V.D.; Aleshin, S.S.; Bolobov, P.A.; Bolshagin, S.N.; Lazarenko, A.P.; Markov, A.V.; Morozov, V.V.; Syslov, A.A.; Tsvetkov, V.M.

    2000-01-01

    Materials on verification of the 3 -d generation code package for WWERs neutronic calculations are presented. The package includes: - spectral code TVS-M; - 2-D fine mesh diffusion code PERMAK-A for 4- or 6-group calculation of WWER core burnup; - 3-D coarse mesh diffusion code BIPR-7A for 2-group calculations of quasi-stationary WWERs regimes. The materials include both TVS-M verification data and verification data on PERMAK-A and BIPR-7A codes using constant libraries generated with TVS-M. All materials are related to the fuel without Gd. TVS-M verification materials include results of comparison both with benchmark calculations obtained by other codes and with experiments carried out at ZR-6 critical facility. PERMAK-A verification materials contain results of comparison with TVS-M calculations and with ZR-6 experiments. BIPR-7A materials include comparison with operation data for Dukovany-2 and Loviisa-1 NPPs (WWER-440) and for Balakovo NPP Unit 4 (WWER-1000). The verification materials demonstrate rather good accuracy of calculations obtained with the use of code package of the 3 -d generation. (Authors)

  17. Documentation and verification required for type A packaging use

    Energy Technology Data Exchange (ETDEWEB)

    O`Brien, J.H.

    1997-07-30

    This document furnishes knowledge and methods for verifying compliance with the U.S. Department of Transportation (DOT) packaging requirements for shipping Type A quantities of radioactive material. The primary emphasis is on the requirements identified in 49 CFR 173.415(a), which states, ``Each offeror of a Specification 7A package must maintain on file for at least one year after the shipment, and shall provide to DOT on request, complete documentation of tests and an engineering evaluation of comparative data showing that the construction methods, packaging design, and materials of construction comply with that specification.`` This guidance document uses a checklist to show compliance.

  18. Documentation and verification required for type A packaging use

    International Nuclear Information System (INIS)

    O'Brien, J.H.

    1997-01-01

    This document furnishes knowledge and methods for verifying compliance with the U.S. Department of Transportation (DOT) packaging requirements for shipping Type A quantities of radioactive material. The primary emphasis is on the requirements identified in 49 CFR 173.415(a), which states, ''Each offeror of a Specification 7A package must maintain on file for at least one year after the shipment, and shall provide to DOT on request, complete documentation of tests and an engineering evaluation of comparative data showing that the construction methods, packaging design, and materials of construction comply with that specification.'' This guidance document uses a checklist to show compliance

  19. Study of applicable methods on safety verification of disposal facilities and waste packages

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    Three subjects about safety verification on the disposal of low level radioactive waste were investigated in FY. 2012. For radioactive waste disposal facilities, specs and construction techniques of covering with soil to prevent possible destruction caused by natural events (e.g. earthquake) were studied to consider verification methods for those specs. For waste packages subject to near surface pit disposal, settings of scaling factor and average radioactivity concentration (hereafter referred to as ''SF'') on container-filled and solidified waste packages generated from Kashiwazaki Kariwa Nuclear Power Station Unit 1-5, setting of cesium residual ratio of molten solidified waste generated from Tokai and Tokai No.2 Power Stations, etc. were studied. Those results were finalized in consideration of the opinion from advisory panel, and publicly opened as JNES-EV reports. In FY 2012, five JNES reports were published and these have been used as standards of safety verification on waste packages. The verification method of radioactive wastes subject to near-surface trench disposal and intermediate depth disposal were also studied. For radioactive wastes which will be returned from overseas, determination methods of radioactive concentration, heat rate and hydrogen generation rate of CSD-C were established. Determination methods of radioactive concentration and heat rate of CSD-B were also established. These results will be referred to verification manuals. (author)

  20. Cleanup Verification Package for the 118-F-5 PNL Sawdust Pit

    International Nuclear Information System (INIS)

    Habel, L.D.

    2008-01-01

    This cleanup verification package documents completion of remedial action, sampling activities, and compliance with cleanup criteria for the 118-F-5 Burial Ground, the PNL (Pacific Northwest Laboratory) Sawdust Pit. The 118-F-5 Burial Ground was an unlined trench that received radioactive sawdust from the floors of animal pens in the 100-F Experimental Animal Farm

  1. Caliver: An R package for CALIbration and VERification of forest fire gridded model outputs.

    Science.gov (United States)

    Vitolo, Claudia; Di Giuseppe, Francesca; D'Andrea, Mirko

    2018-01-01

    The name caliver stands for CALIbration and VERification of forest fire gridded model outputs. This is a package developed for the R programming language and available under an APACHE-2 license from a public repository. In this paper we describe the functionalities of the package and give examples using publicly available datasets. Fire danger model outputs are taken from the modeling components of the European Forest Fire Information System (EFFIS) and observed burned areas from the Global Fire Emission Database (GFED). Complete documentation, including a vignette, is also available within the package.

  2. ENDVER-ENDVER/GUI, The ENDF File Verification Support Package

    International Nuclear Information System (INIS)

    Trkov, Andrej; Zerkin, V.; Cullen, Dermott E.

    2005-01-01

    1 - Description of program or function: Experimental and evaluated nuclear reaction data are compiled world-wide in EXFOR and in ENDF format, respectively. The ENDVER package can be used to convert EXFOR data into computational C4 format, display them and compare graphically with the contents of a specified evaluated data file. The package also contains utilities to retrieve selected materials from a master library in ENDF format, extract cross sections (including differential and double differential data) and output them in two-column PLOTTAB 'curves' format. IAEA1402/03: The ENVER/GUI version contains in addition to ENDVER, also the EXFOR and CINDA Databases (version 1.70 of January 2005) and software providing data search and presentation, the PREPRO ENDF pre-processing codes by D.E. Cullen (PREPRO-2002) and the ZVVIEW interactive graphic data display package. A new GUI (Graphics User Interface) is presented. EndVer and EXFOR/CINDA are directly accessible by the user through the GUI. 2 - Restrictions on the complexity of the problem: None

  3. Cleanup Verification Package for the 107-D5 Trench

    International Nuclear Information System (INIS)

    Corpuz, F.M.; Fancher, J.D.; Blumenkranz, D.B.

    1998-03-01

    This document presents the results of remedial action objectives performed at the 107-D5 Sludge Trench, located at the 100-DR-1 Operable Unit in the 100 Area of the Hanford Site in southeastern Washington State. The 107-D5 Sludge Trench is also identified in the Hanford Waste Information Data System as Waste Site 100-D-4 (site code). The selected remedial action was (1) excavation of the site to the extent required to meet specified soil cleanup levels, (2) disposal of contaminated excavation materials at the Environmental Restoration and Disposal Facility at the 200 Area of the Hanford Site, and (3) backfilling the site with clean soil to adjacent grade elevations

  4. Verification of a BWR code package by gamma scan measurements

    International Nuclear Information System (INIS)

    Nakajima, Tsuyoshi; Iwamoto, Tatsuya; Kumanomido, Hironori

    1996-01-01

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

  5. Verification of the CONPAS (CONtainment Performance Analysis System) code package

    International Nuclear Information System (INIS)

    Kim, See Darl; Ahn, Kwang Il; Song, Yong Man; Choi, Young; Park, Soo Yong; Kim, Dong Ha; Jin, Young Ho.

    1997-09-01

    CONPAS is a computer code package to integrate the numerical, graphical, and results-oriented aspects of Level 2 probabilistic safety assessment (PSA) for nuclear power plants under a PC window environment automatically. For the integrated analysis of Level 2 PSA, the code utilizes four distinct, but closely related modules: (1) ET Editor, (2) Computer, (3) Text Editor, and (4) Mechanistic Code Plotter. Compared with other existing computer codes for Level 2 PSA, and CONPAS code provides several advanced features: computational aspects including systematic uncertainty analysis, importance analysis, sensitivity analysis and data interpretation, reporting aspects including tabling and graphic as well as user-friendly interface. The computational performance of CONPAS has been verified through a Level 2 PSA to a reference plant. The results of the CONPAS code was compared with an existing level 2 PSA code (NUCAP+) and the comparison proves that CONPAS is appropriate for Level 2 PSA. (author). 9 refs., 8 tabs., 14 figs

  6. Transportation Packages to Support Savannah River Site Missions

    International Nuclear Information System (INIS)

    Opperman, E.

    2001-01-01

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

  7. Yucca Mountain Site Characterization Project Waste Package Plan

    International Nuclear Information System (INIS)

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

    1991-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

  9. A history of solid waste packaging at the Hanford Site

    International Nuclear Information System (INIS)

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

    1995-02-01

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

  10. VIPEX (Vital-area Identification Package EXpert) Software Verification and Validation

    International Nuclear Information System (INIS)

    Jung, Woo Sik; Suh, Jae Seung

    2010-06-01

    The purposes of this report are (1) to perform a Verification and Validation (V and V) test for the VIPEX(Vital-area Identification Package EXpert) software and (2) to improve a software quality through the V and V test. The VIPEX was developed in Korea Atomic Energy Research Institute (KAERI) for the Vital Area Identification (VAI) of nuclear power plants. The version of the VIPEX which was distributed is 3.2.0.0. The VIPEX was revised based on the first V and V test and the second V and V test was performed. We have performed the following tasks for the V and V test on Windows XP and VISTA operating systems: Ο Testing basic functions including fault tree editing Ο Testing all kind of functions Ο Research for update from Visual BASIC 6.0 to Visual BASIC 2008

  11. Independent verification in operations at nuclear power plants: Summaries of site visits

    International Nuclear Information System (INIS)

    Donderi, D.C.; Smiley, A.; Ostry, D.J.; Moray, N.P.

    1995-09-01

    A critical review of approaches to independent verification in operations used in nuclear power plant quality assurance programs in other countries was conducted and are detailed in volume 1. This paper is a compilation of the visits to nuclear power plant sites to study independent verification in operations at sites in Canada, USA, Japan, United Kingdom, France and Germany. 3 tabs., 22 figs

  12. Production controls (PC) and technical verification testing (TVT). A methodology for the control and tracking of LILW waste package conditioning

    International Nuclear Information System (INIS)

    Leon, A.M.; Nieto, J.L.L.; Garrido, J.G.

    2003-01-01

    As part of its low and intermediate level radioactive waste (LILW) characterisation and acceptance activities, ENRESA has set up a quality control programme that covers the different phases of radioactive waste package production and implies different levels of tracking in generation, assessment of activity and control of the documentation associated therewith. Furthermore, ENRESA has made available the mechanisms required for verification, depending on the results of periodic sampling, of the quality of the end product delivered by the waste producers. Both processes are included within the framework of two programmes of complementary activities: production controls (PC) and technical verification testing (TVT). (orig.)

  13. Low-Level Radioactive Waste siting simulation information package

    International Nuclear Information System (INIS)

    1985-12-01

    The Department of Energy's National Low-Level Radioactive Waste Management Program has developed a simulation exercise designed to facilitate the process of siting and licensing disposal facilities for low-level radioactive waste. The siting simulation can be conducted at a workshop or conference, can involve 14-70 participants (or more), and requires approximately eight hours to complete. The exercise is available for use by states, regional compacts, or other organizations for use as part of the planning process for low-level waste disposal facilities. This information package describes the development, content, and use of the Low-Level Radioactive Waste Siting Simulation. Information is provided on how to organize a workshop for conducting the simulation. 1 ref., 1 fig

  14. Independent Verification Final Summary Report for the David Witherspoon, Inc. 1630 Site Knoxville, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    P.C. Weaver

    2009-04-29

    The primary objective of the independent verification was to determine if BJC performed the appropriate actions to meet the specified “hot spot” cleanup criteria of 500 picocuries per gram (pCi/g) uranium-238 (U-238) in surface soil. Specific tasks performed by the independent verification team (IVT) to satisfy this objective included: 1) performing radiological walkover surveys, and 2) collecting soil samples for independent analyses. The independent verification (IV) efforts were designed to evaluate radioactive contaminants (specifically U-238) in the exposed surfaces below one foot of the original site grade, given that the top one foot layer of soil on the site was removed in its entirety.

  15. Independent Verification Final Summary Report for the David Witherspoon, Inc. 1630 Site Knoxville, Tennessee

    International Nuclear Information System (INIS)

    Weaver, P.C.

    2009-01-01

    The primary objective of the independent verification was to determine if BJC performed the appropriate actions to meet the specified 'hot spot' cleanup criteria of 500 picocuries per gram (pCi/g) uranium-238 (U-238) in surface soil. Specific tasks performed by the independent verification team (IVT) to satisfy this objective included: (1) performing radiological walkover surveys, and (2) collecting soil samples for independent analyses. The independent verification (IV) efforts were designed to evaluate radioactive contaminants (specifically U-238) in the exposed surfaces below one foot of the original site grade, given that the top one foot layer of soil on the site was removed in its entirety.

  16. Numerical Analysis and Experimental Verification of Stresses Building up in Microelectronics Packaging

    NARCIS (Netherlands)

    Rezaie Adli, A.R.

    2017-01-01

    This thesis comprises a thorough study of the microelectronics packaging process by means of various experimental and numerical methods to estimate the process induced residual stresses. The main objective of the packaging is to encapsulate the die, interconnections and the other exposed internal

  17. Evaluation factors for verification and validation of low-level waste disposal site models

    International Nuclear Information System (INIS)

    Moran, M.S.; Mezga, L.J.

    1982-01-01

    The purpose of this paper is to identify general evaluation factors to be used to verify and validate LLW disposal site performance models in order to assess their site-specific applicability and to determine their accuracy and sensitivity. It is intended that the information contained in this paper be employed by model users involved with LLW site performance model verification and validation. It should not be construed as providing protocols, but rather as providing a framework for the preparation of specific protocols or procedures. A brief description of each evaluation factor is provided. The factors have been categorized according to recommended use during either the model verification or the model validation process. The general responsibilities of the developer and user are provided. In many cases it is difficult to separate the responsibilities of the developer and user, but the user is ultimately accountable for both verification and validation processes. 4 refs

  18. Guidelines for the verification and validation of expert system software and conventional software: Volume 5, Rationale and description of verification and validation guideline packages and procedures. Final report

    International Nuclear Information System (INIS)

    Miller, L.A.; Hayes, J.E.; Mirsky, S.M.

    1995-05-01

    This report is the fifth volume in a series of reports describing the results of the Expert System Verification and Validation (V ampersand V) project which is jointly funded by US NRC and EPRI toward formulating guidelines for V ampersand V of expert systems for use in nuclear power applications. This report provides the rationale for and description of those guidelines. The actual guidelines themselves (and the accompanying 11 step by step Procedures) are presented in Volume 7, User's Manual. Three factors determine what V ampersand V is needed: (1) the stage, of the development life cycle (requirements, design, or implementation), (2) whether the overall system or a specialized component needs be tested (knowledge base component, inference engine or other highly reusable element, or a component involving conventional software), and (3) the stringency of V ampersand V that is needed (as judged from an assessment of the system's complexity and the requirement for its integrity to form three Classes). A V ampersand V guideline package is provided for each of the combinations of these three variables. The package specifies the V ampersand V methods recommended and the order in which they should be administered, the assurances each method provides, the qualifications needed by the V ampersand V team to employ each Particular method, the degree to which the methods should be applied, the performance measures that should be taken, and the decision criteria for accepting, conditionally accepting, or rejecting an evaluated system. In addition to the guideline packages, highly detailed step-by-step procedures are provided for 11 of the more important methods, to ensure that they Can be implemented correctly. The guidelines can apply to conventional procedural software systems as well as all kinds of AI systems

  19. On site PWR fuel inspection measurements for operational and design verification

    International Nuclear Information System (INIS)

    1996-01-01

    The on-site inspection of irradiated Pressurized Water Reactor (PWR) fuel and Non-Fuel Bearing Components (NFBC) is typically limited to visual inspections during refuelings using underwater TV cameras and is intended primarily to confirm whether the components will continue in operation. These inspections do not normally provide data for design verification nor information to benefit future fuel designs. Japanese PWR utilities and Nuclear Fuel Industries Ltd. designed, built, and performed demonstration tests of on-site inspection equipment that confirms operational readiness of PWR fuel and NFBC and also gathers data for design verification of these components. 4 figs, 3 tabs

  20. Design/licensing of on-site package for core component

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  1. Code Package to Analyze Parameters of the WWER Fuel Rod. TOPRA-2 Code - Verification Data

    International Nuclear Information System (INIS)

    Scheglov, A.; Proselkov, V.; Passage, G.; Stefanova, S.

    2009-01-01

    Presented are the data for computer codes to analyze WWER fuel rods, used in the WWER department of RRC 'Kurchatov Institute'. Presented is the description of TOPRA-2 code intended for the engineering analysis of thermophysical and strength parameters of the WWER fuel rod - temperature distributions along the fuel radius, gas pressures under the cladding, stresses in the cladding, etc. for the reactor operation in normal conditions. Presented are some results of the code verification against test problems and the data obtained in the experimental programs. Presented are comparison results of the calculations with TOPRA-2 and TRANSURANUS (V1M1J06) codes. Results obtained in the course of verification demonstrate possibility of application of the methodology and TOPRA-2 code for the engineering analysis of the WWER fuel rods

  2. Package

    Directory of Open Access Journals (Sweden)

    Arsić Zoran

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Soo, P.

    1986-01-01

    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

  4. Review of waste package verification tests. Semiannual report, October 1984-March 1985

    International Nuclear Information System (INIS)

    Soo, P.

    1985-07-01

    The potential of WAPPA, a second-generation waste package system code, to meet the needs of the regulatory community is analyzed. The analysis includes an indepth review of WAPPA's individual process models and a review of WAPPA's operation. It is concluded that the code is of limited use to the NRC in the present form. Recommendations for future improvement, usage, and implementation of the code are given. This report also describes the results of a testing program undertaken to determine the chemical environment that will be present near a high-level waste package emplaced in a basalt repository. For this purpose, low carbon 1020 steel (a current BWIP reference container material), synthetic basaltic groundwater and a mixture of bentonite and basalt were exposed, in an autoclave, to expected conditions some period after repository sealing (150 0 C, approx. =10.4 MPa). Parameters measured include changes in gas pressure with time and gas composition, variation in dissolved oxygen (DO), pH and certain ionic concentrations of water in the packing material across an imposed thermal gradient, mineralogic alteration of the basalt/bentonite mixture, and carbon steel corrosion behavior. A second testing program was also initiated to check the likelihood of stress corrosion cracking of austenitic stainless steels and Incoloy 825 which are being considered for use as waste container materials in the tuff repository program. 82 refs., 70 figs., 27 tabs

  5. Experience in verification regimes. United States On-Site Inspection Agency

    International Nuclear Information System (INIS)

    Reppert, J.

    1998-01-01

    Experiences are described of the United States On-site Inspection Agency in verification regimes all over the world where it has been applied in the last 30 years. The challenge for the future is to extend the benefits of the applied tools to all states in all regions to enhance stability and to create conditions for peace at lower levels of armaments than currently exist. The USA need to engage states currently caught in cycles of violence and arms escalation. They must examine technologies which together with on-site aspects of verification or transparency regimes can provide a comprehensive picture at affordable costs. They foresee a growth in combined training with new states entering for the first time into regime that include arms control and transparency measure

  6. Verification and validation of the SAPHIRE Version 4.0 PRA software package

    International Nuclear Information System (INIS)

    Bolander, T.W.; Calley, M.B.; Capps, E.L.

    1994-02-01

    A verification and validation (V ampersand V) process has been performed for the System Analysis Programs for Hands-on Integrated Reliability Evaluation (SAPHIRE). SAPHIRE is a set of four computer programs that the Nuclear Regulatory Commission (NRC) developed to perform probabilistic risk assessments (PRAs). These programs allow an analyst to create, quantify, and evaluate the risk associated with a facility or process being analyzed. The programs included in this set are Integrated Reliability and Risk Analysis System (IRRAS), System Analysis and Risk Assessment (SARA), Models and Results Database (MAR-D), and Fault Tree/Event Tree/Piping and Instrumentation Diagram (FEP) graphical editor. The V ampersand V steps included a V ampersand V plan to describe the process and criteria by which the V ampersand V would be performed; a software requirements documentation review to determine the correctness, completeness, and traceability of the requirements; a user survey to determine the usefulness of the user documentation, identification and testing of vital and non-vital features, and documentation of the test results

  7. On-site detection of packaged squid freshness

    Science.gov (United States)

    Ahmad, Noor Azizah; Heng, Lee Yook; Salam, Faridah; Hanifah, Sharina Abu

    2018-04-01

    The development of indicator label for detection of total volatile basic nitrogen (TVB-N) is described. Dye extract from edible plants containing anthocyanins was immobilized onto iota-carrageenan as polymer matrix. TVB-N detection worked based on pH increase as the basic deterioration volatile amines generated in the package headspace. Results showed that the indicator label has changed color from blue to green after 12 hours of storage at ambient conditions. The TVB-N value was 38.9648 mg /100 g which is exceeded of acceptability level for seafood products. The pH value of squid flesh has also increased during storage. The colour values of L * and a * negative increases while b* negative decrease with increasing storage time. The indicator label is potentially used as freshness indicator for squid at ambient conditions.

  8. Cleanup Verification Package for the 100-K-55:1 and 100-K-56:1 Pipelines and the 116-KW-4 and 116-KE-5 Heat Recovery Stations

    International Nuclear Information System (INIS)

    Capron, J.M.

    2005-01-01

    This cleanup verification package documents completion of remedial action for the 100-K-55:1 and 100-K-56:1 reactor cooling effluent underground pipelines and for the 116-KW-4 and 116-KE-5 heat recovery stations. The 100-K-55 and 100-K-56 sites consisted of those process effluent pipelines that serviced the 105-KW and 105-KE Reactors. This cleanup verification package documents completion of remedial action for the 100-K-55: 1 and 100-K-56: 1 reactor cooling effluent underground pipelines, referred to herein as the 100-K-55:1 and 100-K-56:l sites, as well as for the 116-KW-4 and 116-KE-5 heat recovery stations, referred to herein as the 116-KW-4 and 116-KE-5 sites. The 116-KW-4 and 116-KE-5 heat recovery stations were co-located and remediated with the 100-K-55:1 and 100-K-56:1 pipelines, respectively. These sites are located in the 100-KR-2 Operable Unit in the 100-K Area of the Hanford Site in southeastern Washington State. The 100-K-55 and 100-K-56 sites consisted of those process effluent pipelines that serviced the 105-KW and 105-KE Reactors, respectively. Both of these sites have been administratively divided into subunits based on the current extent of remediation. Portions of the pipelines remaining within the reactor security fencing and in proximity to active utility features have been delineated as the 100-K-55:2 and 100-K-56:2 pipelines, with the portions of the pipelines excluded from these boundaries identified as the 100-K-55:1 and 100-K-56:1 pipelines. This cleanup verification package addresses only the 100-K-55:1 and 100-K-56:I subunits; the 100-K-55:2 and 100-K-56:2 subunits will be addressed within a separate cleanup verification package. Site excavation and waste disposal are complete, and the exposed surfaces have been sampled and analyzed to verify attainment of the remedial action goals. Results of the sampling, laboratory analyses, and data evaluations for the 100-K-55:1, 100-K-56:1, 116-KW-4, and 116-KE-5 sites indicate that all remedial

  9. Verification of the 2.00 WAPPA-B [Waste Package Performance Assessment-B version] code

    International Nuclear Information System (INIS)

    Tylock, B.; Jansen, G.; Raines, G.E.

    1987-07-01

    The old version of the Waste Package Performance Assessment (WAPPA) code has been modified into a new code version, 2.00 WAPPA-B. The input files and the results for two benchmarks at repository conditions are fully documented in the appendixes of the EA reference report. The 2.00 WAPPA-B version of the code is suitable for computation of barrier failure due to uniform corrosion; however, an improved sub-version, 2.01 WAPPA-B, is recommended for general use due to minor errors found in 2.00 WAPPA-B during its verification procedures. The input files and input echoes have been modified to include behavior of both radionuclides and elements, but the 2.00 WAPPA-B version of the WAPPA code is not recommended for computation of radionuclide releases. The 2.00 WAPPA-B version computes only mass balances and the initial presence of radionuclides that can be released. Future code development in the 3.00 WAPPA-C version will include radionuclide release computations. 19 refs., 10 figs., 1 tab

  10. bcROCsurface: an R package for correcting verification bias in estimation of the ROC surface and its volume for continuous diagnostic tests.

    Science.gov (United States)

    To Duc, Khanh

    2017-11-18

    Receiver operating characteristic (ROC) surface analysis is usually employed to assess the accuracy of a medical diagnostic test when there are three ordered disease status (e.g. non-diseased, intermediate, diseased). In practice, verification bias can occur due to missingness of the true disease status and can lead to a distorted conclusion on diagnostic accuracy. In such situations, bias-corrected inference tools are required. This paper introduce an R package, named bcROCsurface, which provides utility functions for verification bias-corrected ROC surface analysis. The shiny web application of the correction for verification bias in estimation of the ROC surface analysis is also developed. bcROCsurface may become an important tool for the statistical evaluation of three-class diagnostic markers in presence of verification bias. The R package, readme and example data are available on CRAN. The web interface enables users less familiar with R to evaluate the accuracy of diagnostic tests, and can be found at http://khanhtoduc.shinyapps.io/bcROCsurface_shiny/ .

  11. In-field inspection support software: A status report on the Common Inspection On-site Software Package (CIOSP) project

    International Nuclear Information System (INIS)

    Novatchev, Dimitre; Titov, Pavel; Siradjov, Bakhtiiar; Vlad, Ioan; Xiao Jing

    2001-01-01

    Full text: IAEA has invested much thought and effort into developing software that can assist inspectors during their inspection work. Experience with such applications has been steadily growing and IAEA has recently commissioned a next-generation software package. This kind of software accommodates inspection tasks that can vary substantially in function depending on the type of installation being inspected as well as ensures that the resulting software package has a wide range of usability and can preclude excessive development of plant-specific applications. The Common Inspection On-site Software Package is being developed in the Department of Safeguards to address the limitations of the existing software and to expand its coverage of the inspection process. CIOSP is 'common' in that it is aimed at providing support for as many facilities as possible with the minimum re-configuration. At the same time it has to cater to varying needs of individual facilities, different instrumentation and verification methods used. A component-based approach was taken to successfully tackle the challenges that the development of this software presented. CIOSP consists of the following major components: A framework into which individual plug-ins supporting various inspection activities can integrate at run-time; A central data store containing all facility configuration data and all data collected during inspections; A local data store, which resides on the inspector's computer, where the current inspection's data is stored; A set of services used by all plug-ins (i.e. data transformation, authentication, replication services etc.). This architecture allows for incremental development and extension of the software with plug-ins that support individual inspection activities. The core set of components along with the framework, the Inventory Verification, Book Examination and Records and Reports Comparison plug-ins have been developed. The development of the Short Notice Random

  12. Remaining Sites Verification Package for the 182-F Reservoir Waste Site. Attachment to Waste Site Reclassification Form 2005-025

    International Nuclear Information System (INIS)

    Carlson, R.A.

    2005-01-01

    The 182-F Reservoir was a rectangular-shaped concrete basin consisting of two sections divided by a concrete wall. The reservoir provided reserve water from the Columbia River for reactor cooling water and raw water for the 100 Area and had a storage capacity of 94.6 million liters (25 million gallons). The 182-F Reservoir was later used as a landfill for decontaminated rubble from buildings that were decommissioned in the 100-F Area. The results of the 182-F Reservoir evaluation showed that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River

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

    International Nuclear Information System (INIS)

    Smith, R.J.

    1995-02-01

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

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

    International Nuclear Information System (INIS)

    England, Jeffery L.; Adams, Karen; Maxted, Maxcine; Ruff Jr, Clarence; Albenesius, Andrew; Bowers, Mark D.; Fountain, Geoffrey; Hughes, Michael; Gordon, Sydney; O'Connor, Stephen

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  16. Packaging and transportation of depleted uranium for disposition from the Savannah River Site

    International Nuclear Information System (INIS)

    Gillas, D.L.; Berg, J.

    2009-01-01

    The Savannah River Site (SRS) produced a large inventory of depleted uranium trioxide (DUO) in a powder form packaged in approximately 36,000 55-gallon drums that required final disposition. Each drum weighs an average of 680 kg (1,500 pounds) with some as much as 820 kg (1,800 pounds). The weight, and the fact that the material is in a powder form, requires detailed planning concerning the packaging and transportation (P and T) that must be used. Four disposition campaigns have been completed with the first in Fiscal Year 2003 (FY03), the second in FY04/05, and the most recent two campaigns being completed in early FY09. The remaining inventory of approximately 16,000 drums will likely follow similar paths in the future. This paper will describe the DUO inventory and the thought process behind determining the appropriate P and T for each campaign, very briefly covering the first two campaigns and emphasizing the most recent campaigns. In FY03, SRS completed a pilot project that disposed of 3,270 55-gallon drums of DUO. The shipping method used 110-ton mill gondola rail-cars with a polypropylene coated fabric liner as the DOT 'strong, tight' package. These rail-cars were shipped to the EnergySolutions low level waste (LLW) disposal facility in Clive, UT (previously Envirocare of Utah now referred to in this paper as the Clive Facility) for final disposition of the DUO as LLW. In FY04/05, an additional 7,296 drums that were over-packed in 85-gallon drums were shipped in boxcars (not part of the packaging) since the overpacks were qualified as IP-2 containers due to the excessive weight of the drums (over 680 kg each) to the Clive Facility. The two most recent campaigns consisted of: 1) 5,408 55-gallon drums that were shipped to the Clive Facility in 52.5-foot gondola rail-cars with fiberglass lids; the rail-car itself was the package as well as the conveyance, and 2) 4014 55-gallon drums that were shipped to the Nevada Test Site (NTS) in 20-foot modified cargo

  17. Guidelines for the verification and validation of expert system software and conventional software: Rationale and description of V ampersand V guideline packages and procedures. Volume 5

    International Nuclear Information System (INIS)

    Mirsky, S.M.; Hayes, J.E.; Miller, L.A.

    1995-03-01

    This report is the fifth volume in a series of reports describing the results of the Expert System Verification C, and Validation (V ampersand V) project which is jointly funded by the U.S. Nuclear Regulatory Commission and the Electric Power Research Institute toward the objective of formulating Guidelines for the V ampersand V of expert systems for use in nuclear power applications. This report provides the rationale for and description of those guidelines. The actual guidelines themselves are presented in Volume 7, open-quotes User's Manual.close quotes Three factors determine what V ampersand V is needed: (1) the stage of the development life cycle (requirements, design, or implementation); (2) whether the overall system or a specialized component needs to be tested (knowledge base component, inference engine or other highly reusable element, or a component involving conventional software); and (3) the stringency of V ampersand V that is needed (as judged from an assessment of the system's complexity and the requirement for its integrity to form three Classes). A V ampersand V Guideline package is provided for each of the combinations of these three variables. The package specifies the V ampersand V methods recommended and the order in which they should be administered, the assurances each method provides, the qualifications needed by the V ampersand V team to employ each particular method, the degree to which the methods should be applied, the performance measures that should be taken, and the decision criteria for accepting, conditionally accepting, or rejecting an evaluated system. In addition to the Guideline packages, highly detailed step-by-step procedures are provided for 11 of the more important methods, to ensure that they can be implemented correctly. The Guidelines can apply to conventional procedural software systems as well as all kinds of Al systems

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

    Highly radioactive sludge resulting from the storage of degraded spent nuclear fuel has been consolidated in Engineered Containers (ECs) in the 105-K West Storage Basin located on the Hanford site near the Columbia River in Washington State. CH2M Hill Plateau Remediation Company (CHPRC) is proceeding with a project to retrieve the sludge, place it in Sludge Transport and Storage Containers (STSCs) and store those filled containers within the T Plant Canyon facility on the Hanford Site Central Plateau (Phase 1). Retrieval and transfer of the sludge material will enable removal of the 105-K West Basin and allow remediation of the subsurface contamination plumes under the basin. The U.S. Department of Energy (DOE) plans to treat and dispose of this K Basins sludge (Phase 2) as Remote Handled Transuranic Waste (RH TRU) at the Waste Isolation Pilot Plant (WIPP) located in New Mexico. The K Basin sludge currently contains uranium metal which reacts with water present in the stored slurry, generating hydrogen and other byproducts. The established transportation and disposal requirements require the transformation of the K Basins sludge to a chemically stable, liquid-free, packaged waste form. The Treatment and Packaging Project includes removal of the containerised sludge from T Plant, the treatment of the sludge as required, and packaging of all the sludge into a form that is certifiable for transportation to and disposal at WIPP. Completion of this scope will require construction and operation of a Sludge Treatment and Packaging Facility (STPF), which could be either a completely new facility or a modification of an existing Hanford Site facility. A Technology Evaluation and Alternatives Analysis (TEAA) for the STP Phase 2 was completed in 2011. A Request for Technology Information (RFI) had been issued in October 2009 to solicit candidate technologies for use in Phase 2. The RFI also included a preliminary definition of Phase 2 functions and requirements. Potentially

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

    International Nuclear Information System (INIS)

    Fogwell, Thomas W.; Honeyman, James O.; Stegen, Gary

    2013-01-01

    Highly radioactive sludge resulting from the storage of degraded spent nuclear fuel has been consolidated in Engineered Containers (ECs) in the 105-K West Storage Basin located on the Hanford site near the Columbia River in Washington State. CH2M Hill Plateau Remediation Company (CHPRC) is proceeding with a project to retrieve the sludge, place it in Sludge Transport and Storage Containers (STSCs) and store those filled containers within the T Plant Canyon facility on the Hanford Site Central Plateau (Phase 1). Retrieval and transfer of the sludge material will enable removal of the 105-K West Basin and allow remediation of the subsurface contamination plumes under the basin. The U.S. Department of Energy (DOE) plans to treat and dispose of this K Basins sludge (Phase 2) as Remote Handled Transuranic Waste (RH TRU) at the Waste Isolation Pilot Plant (WIPP) located in New Mexico. The K Basin sludge currently contains uranium metal which reacts with water present in the stored slurry, generating hydrogen and other byproducts. The established transportation and disposal requirements require the transformation of the K Basins sludge to a chemically stable, liquid-free, packaged waste form. The Treatment and Packaging Project includes removal of the containerised sludge from T Plant, the treatment of the sludge as required, and packaging of all the sludge into a form that is certifiable for transportation to and disposal at WIPP. Completion of this scope will require construction and operation of a Sludge Treatment and Packaging Facility (STPF), which could be either a completely new facility or a modification of an existing Hanford Site facility. A Technology Evaluation and Alternatives Analysis (TEAA) for the STP Phase 2 was completed in 2011. A Request for Technology Information (RFI) had been issued in October 2009 to solicit candidate technologies for use in Phase 2. The RFI also included a preliminary definition of Phase 2 functions and requirements. Potentially

  20. Geochemical Data Package for Performance Assessment Calculations Related to the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, Daniel I. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-07-22

    The Savannah River Site (SRS) disposes of low-level radioactive waste (LLW) and stabilizes high-level radioactive waste (HLW) tanks in the subsurface environment. Calculations used to establish the radiological limits of these facilities are referred to as Performance Assessments (PA), Special Analyses (SA), and Composite Analyses (CA). The objective of this document is to revise existing geochemical input values used for these calculations. This work builds on earlier compilations of geochemical data (2007, 2010), referred to a geochemical data packages. This work is being conducted as part of the on-going maintenance program of the SRS PA programs that periodically updates calculations and data packages when new information becomes available. Because application of values without full understanding of their original purpose may lead to misuse, this document also provides the geochemical conceptual model, the approach used for selecting the values, the justification for selecting data, and the assumptions made to assure that the conceptual and numerical geochemical models are reasonably conservative (i.e., bias the recommended input values to reflect conditions that will tend to predict the maximum risk to the hypothetical recipient). This document provides 1088 input parameters for geochemical parameters describing transport processes for 64 elements (>740 radioisotopes) potentially occurring within eight subsurface disposal or tank closure areas: Slit Trenches (ST), Engineered Trenches (ET), Low Activity Waste Vault (LAWV), Intermediate Level (ILV) Vaults, Naval Reactor Component Disposal Areas (NRCDA), Components-in-Grout (CIG) Trenches, Saltstone Facility, and Closed Liquid Waste Tanks. The geochemical parameters described here are the distribution coefficient, Kd value, apparent solubility concentration, ks value, and the cementitious leachate impact factor.

  1. PROGRESS WITH K BASINS SLUDGE RETRIEVAL STABILIZATION & PACKAGING AT THE HANFORD NUCLEAR SITE

    Energy Technology Data Exchange (ETDEWEB)

    KNOLLMEYER, P.M.; PHILLIPS, C; TOWNSON, P.S.

    2006-01-30

    This paper shows how Fluor Hanford and BNG America have combined nuclear plant skills from the U.S. and the U.K. to devise methods to retrieve and treat the sludge that has accumulated in K Basins at the Hanford Site over many years. Retrieving the sludge is the final stage in removing fuel and sludge from the basins to allow them to be decontaminated and decommissioned, so as to remove the threat of contamination of the Columbia River. A description is given of sludge retrieval using vacuum lances and specially developed nozzles and pumps into Consolidation Containers within the basins. The special attention that had to be paid to the heat generation and potential criticality issues with the irradiated uranium-containing sludge is described. The processes developed to re-mobilize the sludge from the Consolidation Containers and pump it through flexible and transportable hose-in-hose piping to the treatment facility are explained with particular note made of dealing with the abrasive nature of the sludge. The treatment facility, housed in an existing Hanford building, is described, and the uranium-corrosion and grout packaging processes explained. The uranium corrosion process is a robust, tempered process very suitable for dealing with a range of differing sludge compositions. Optimization and simplification of the original sludge corrosion process design is described and the use of transportable and reusable equipment is indicated. The processes and techniques described in the paper are shown to have wide applicability to nuclear cleanup.

  2. Results of the radiological and beryllium verification survey at the Peek Street Site, Schenectady, New York (SY001V)

    International Nuclear Information System (INIS)

    Foley, R.D.; Johnson, C.A.; Carrier, R.F.; Allred, J.F.

    1994-10-01

    At the request of the U.S. Department of Energy (DOE), a team from Oak Ridge National Laboratory conducted an independent verification radiological and non-radioactive beryllium survey at the Peek Street site, located at 425 Peek Street, Schenectady, New York. The purpose of the survey, conducted during 1993 and continuing through January 1994, was to confirm the success of the remedial actions performed to remove any beryllium concentrations or radioactive materials in excess of the identified guidelines. The verification survey included surface gamma scans and gamma readings at one meter indoors and outdoors, alpha and beta scans inside the structure, and the collection of soil, dust and debris samples and smears for radionuclide and beryllium analyses. Results of the survey demonstrated that all radiological and beryllium measurements on the property were within applicable DOE guidelines. Based on all data collected, the industrial property at 425 Peek Street and the adjacent state-owned bike path in Schenectady, New York, conforms to all applicable radiological and non-radioactive beryllium guidelines established for this site by DOE and approved by the State of New York

  3. Verification of criticality safety in on-site spent fuel storage systems

    International Nuclear Information System (INIS)

    Rasmussen, R.W.

    1989-01-01

    On February 15, 1984, Duke Power Company received approval for a two-region, burnup credit, spent fuel storage rack design at both Units 1 and 2 of the McGuire Nuclear Station. Duke also hopes to obtain approval by January of 1990 for a dry spent fuel storage system at the Oconee Nuclear Station, which will incorporate the use of burnup credit in the criticality analysis governing the design of the individual storage units. While experiences in burnup verification for criticality safety for their dry storage system at Oconee are in the future, the methods proposed for burnup verification will be similar to those currently used at the McGuire Nuclear Station in the two-region storage racks installed in both pools. In conclusion, the primary benefit of the McGuire rerack effort has obviously been the amount of storage expansion it provided. A total increase of about 2,000 storage cells was realized, 1,000 of which were the result of pursuing the two-region rather than the conventional poison rack design. Less impacting, but equally as important, however, has been the experience gained during the planning, installation, and operation of these storage racks. This experience should prove useful for future rerack efforts likely to occur at Duke's Catawba Nuclear Station as well as for the current dry storage effort underway for the Oconee Nuclear Station

  4. US monitoring and verification technology: on-site inspection experience and future challenges

    International Nuclear Information System (INIS)

    Gullickson, R.L.; Carlson, D.; Ingraham, J.; Laird, B.

    2013-01-01

    The United States has a long and successful history of cooperation with treaty partners in monitoring and verification. For strategic arms reduction treaties, our collaboration has resulted in the development and application of systems with limited complexity and intrusiveness. As we progress beyond New START (NST) along the 'road to zero', the reduced number of nuclear weapons is likely to require increased confidence in monitoring and verification techniques. This may place increased demands on the technology to verify the presence of a nuclear weapon and even confirm the presence of a certain type. Simultaneously, this technology must include the ability to protect each treaty partner's sensitive nuclear weapons information. Mutual development of this technology by treaty partners offers the best approach for acceptance in treaty negotiations. This same approach of mutual cooperation and development is essential for developing nuclear test monitoring technology in support of the Comprehensive Nuclear Test Ban Treaty (CTBT). Our ability to detect low yield and evasive testing will be enhanced through mutually developed techniques and experiments using laboratory laser experiments and high explosives tests in a variety of locations and geologies. (authors)

  5. Analysis and evaluation of a radioactive waste package retrieved from the Farallon Islands 900-meter disposal site

    International Nuclear Information System (INIS)

    Colombo, P.; Kendig, M.W.

    1990-09-01

    The Environmental Protection Agency (EPA) was given a Congressional mandate to develop criteria and regulations governing the ocean disposal of all forms of waste. The EPA taken an active role both nationally and within the international nuclear regulatory community to develop the effective controls necessary to protect the health and safety of man and the marine environment. The EPA Office of Radiation Programs (ORP) first initiated feasibility studies to determine whether current technologies could be applied toward determining the fate of radioactive waste disposed of in the past. After successfully locating actual radioactive waste packages in formerly used disposal sites, in the United States, the Office of Radiation Programs developed an intensive program of site characterization studies to examine biological, chemical and physical characteristics including evaluations of the concentration and distribution of radionuclides within these sites, and has conducted a performance evaluation of past packaging techniques and materials. Brookhaven National Laboratory (BNL) has performed container corrosion and matrix analysis studies on the recovered radioactive waste packages. This report presents the final results of laboratory analyses performed. 17 refs., 40 figs., 7 tabs

  6. Analysis and evaluation of a radioactive waste package retrieved from the Farallon Islands 900-meter disposal site

    Energy Technology Data Exchange (ETDEWEB)

    Colombo, P.; Kendig, M.W.

    1990-09-01

    The Environmental Protection Agency (EPA) was given a Congressional mandate to develop criteria and regulations governing the ocean disposal of all forms of waste. The EPA taken an active role both nationally and within the international nuclear regulatory community to develop the effective controls necessary to protect the health and safety of man and the marine environment. The EPA Office of Radiation Programs (ORP) first initiated feasibility studies to determine whether current technologies could be applied toward determining the fate of radioactive waste disposed of in the past. After successfully locating actual radioactive waste packages in formerly used disposal sites, in the United States, the Office of Radiation Programs developed an intensive program of site characterization studies to examine biological, chemical and physical characteristics including evaluations of the concentration and distribution of radionuclides within these sites, and has conducted a performance evaluation of past packaging techniques and materials. Brookhaven National Laboratory (BNL) has performed container corrosion and matrix analysis studies on the recovered radioactive waste packages. This report presents the final results of laboratory analyses performed. 17 refs., 40 figs., 7 tabs.

  7. Combining Space Geodesy, Seismology, and Geochemistry for Monitoring Verification and Accounting of CO2 in Sequestration Sites

    Energy Technology Data Exchange (ETDEWEB)

    Swart, Peter K. [Univ. of Miami, Key Biscayne, FL (United States); Dixon, Tim [Univ. of South Florida, Tampa, FL (United States)

    2014-09-30

    A series of surface geophysical and geochemical techniques are tested in order to demonstrate and validate low cost approaches for Monitoring, Verification and Accounting (MVA) of the integrity of deep reservoirs for CO2 storage. These techniques are (i) surface deformation by GPS; ii) surface deformation by InSAR; iii) passive source seismology via broad band seismometers; and iv) soil gas monitoring with a cavity ring down spectrometer for measurement of CO2 concentration and carbon isotope ratio. The techniques were tested at an active EOR (Enhanced Oil Recovery) site in Texas. Each approach has demonstrated utility. Assuming Carbon Capture, Utilization and Storage (CCUS) activities become operational in the future, these techniques can be used to augment more expensive down-hole techniques.

  8. Geology Data Package for the Single-Shell Tank Waste Management Areas at the Hanford Site

    Energy Technology Data Exchange (ETDEWEB)

    Reidel, Steve P.; Chamness, Mickie A.

    2007-01-01

    This data package discusses the geology of the single-shell tank (SST) farms and the geologic history of the area. The focus of this report is to provide the most recent geologic information available for the SST farms. This report builds upon previous reports on the tank farm geology and Integrated Disposal Facility geology with information available after those reports were published.

  9. Analysis of capsid portal protein and terminase functional domains: interaction sites required for DNA packaging in bacteriophage T4.

    Science.gov (United States)

    Lin, H; Rao, V B; Black, L W

    1999-06-04

    Bacteriophage DNA packaging results from an ATP-driven translocation of concatemeric DNA into the prohead by the phage terminase complexed with the portal vertex dodecamer of the prohead. Functional domains of the bacteriophage T4 terminase and portal gene 20 product (gp20) were determined by mutant analysis and sequence localization within the structural genes. Interaction regions of the portal vertex and large terminase subunit (gp17) were determined by genetic (terminase-portal intergenic suppressor mutations), biochemical (column retention of gp17 and inhibition of in vitro DNA packaging by gp20 peptides), and immunological (co-immunoprecipitation of polymerized gp20 peptide and gp17) studies. The specificity of the interaction was tested by means of a phage T4 HOC (highly antigenicoutercapsid protein) display system in which wild-type, cs20, and scrambled portal peptide sequences were displayed on the HOC protein of phage T4. Binding affinities of these recombinant phages as determined by the retention of these phages by a His-tag immobilized gp17 column, and by co-immunoprecipitation with purified terminase supported the specific nature of the portal protein and terminase interaction sites. In further support of specificity, a gp20 peptide corresponding to a portion of the identified site inhibited packaging whereas the scrambled sequence peptide did not block DNA packaging in vitro. The portal interaction site is localized to 28 residues in the central portion of the linear sequence of gp20 (524 residues). As judged by two pairs of intergenic portal-terminase suppressor mutations, two separate regions of the terminase large subunit gp17 (central and COOH-terminal) interact through hydrophobic contacts at the portal site. Although the terminase apparently interacts with this gp20 portal peptide, polyclonal antibody against the portal peptide appears unable to access it in the native structure, suggesting intimate association of gp20 and gp17 possibly

  10. High-throughput verification of transcriptional starting sites by Deep-RACE

    DEFF Research Database (Denmark)

    Olivarius, Signe; Plessy, Charles; Carninci, Piero

    2009-01-01

    We present a high-throughput method for investigating the transcriptional starting sites of genes of interest, which we named Deep-RACE (Deep–rapid amplification of cDNA ends). Taking advantage of the latest sequencing technology, it allows the parallel analysis of multiple genes and is free...

  11. Seismic imaging for an ocean drilling site survey and its verification in the Izu rear arc

    Science.gov (United States)

    Yamashita, Mikiya; Takahashi, Narumi; Tamura, Yoshihiko; Miura, Seiichi; Kodaira, Shuichi

    2018-01-01

    To evaluate the crustal structure of a site proposed for International Ocean Discovery Program drilling, the Japan Agency for Marine-Earth Science and Technology carried out seismic surveys in the Izu rear arc between 2006 and 2008, using research vessels Kaiyo and Kairei. High-resolution dense grid surveys, consisting of three kinds of reflection surveys, generated clear seismic profiles, together with a seismic velocity image obtained from a seismic refraction survey. In this paper, we compare the seismic profiles with the geological column obtained from the drilling. Five volcaniclastic sedimentary units were identified in seismic reflection profiles above the 5 km/s and 6 km/s contours of P-wave velocity obtained from the velocity image from the seismic refraction survey. However, some of the unit boundaries interpreted from the seismic images were not recognised in the drilling core, highlighting the difficulties of geological target identification in volcanic regions from seismic images alone. The geological core derived from drilling consisted of seven lithological units (labelled I to VII). Units I to V were aged at 0-9 Ma, and units VI and VII, from 1320-1806.5 m below seafloor (mbsf) had ages from 9 to ~15 Ma. The strong heterogeneity of volcanic sediments beneath the drilling site U1437 was also identified from coherence, calculated using cross-spectral analysis between grid survey lines. Our results suggest that use of a dense grid configuration is important in site surveys for ocean drilling in volcanic rear-arc situations, in order to recognise heterogeneous crustal structure, such as sediments from different origins.

  12. Data Validation Package May 2016 Groundwater Sampling at the Bluewater, New Mexico, Disposal Site, September 2016

    International Nuclear Information System (INIS)

    Johnson, Dick; Tsosie, Bernadette

    2016-01-01

    Groundwater samples were collected from monitoring wells at the Bluewater, New Mexico, Disposal Site to monitor groundwater contaminants as specified in the 1997 Long-Term Surveillance Plan for the DOE Bluewater (UMTRCA Title II) Disposal Site Near Grants, New Mexico (LTSP). Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). A duplicate sample was collected from location 16(SG).

  13. Data Validation Package May 2016 Groundwater Sampling at the Bluewater, New Mexico, Disposal Site, September 2016

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, Dick [Navarro Nevada Environmental Services (NNES), Las Vegas, NV (United States); Tsosie, Bernadette [US Department of Energy, Washington, DC (United States)

    2016-09-01

    Groundwater samples were collected from monitoring wells at the Bluewater, New Mexico, Disposal Site to monitor groundwater contaminants as specified in the 1997 Long-Term Surveillance Plan for the DOE Bluewater (UMTRCA Title II) Disposal Site Near Grants, New Mexico (LTSP). Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). A duplicate sample was collected from location 16(SG).

  14. Phase two of Site 300's ecological risk assessment: Model verification and risk management

    International Nuclear Information System (INIS)

    Carlson, T.M.; Gregory, S.D.

    1995-01-01

    The authors completed the baseline ecological risk assessment (ERA) for Lawrence Livermore National Laboratory's Site 300 in 1993. Using data collection and modeling techniques adapted from the human health risk assessment (HRA), they evaluated the potential hazard of contaminants in environmental media to ecological receptors. They identified potential hazards to (1) aquatic invertebrates from heavy metal contaminants in surface water, (2) burrowing vertebrates from contaminants volatilizing from subsurface soil into burrow air, and (3) grazing deer and burrowing vertebrates from cadmium contamination in surface soil. They recently began collecting data to refine the estimates of potential hazard to these ecological receptors. Bioassay results form the surface water failed to verify a hazard to aquatic invertebrates. Soil vapor surveys of subsurface burrows did verify the presence of high concentrations of volatile organic compounds (VOCs). However, they have not yet verified a true impact on the burrowing populations. The authors also completed an extensive surface soil sampling program, which identified local hot spots of cadmium contamination. In addition, they have been collecting data on the land use patterns of the deer population. Their data indicate that deer do not typically use those areas with cadmium surface soil contamination. Information from this phase of the ERA, along with the results of the HRA, will direct the selection of remedial alternatives for the site. For the ecological receptors, remedial alternatives include developing a risk management program which includes ensuring that (1) sensitive burrowing species (such as rare or endangered species) do not use areas of surface or subsurface contamination, and (2) deer populations do not use areas of surface soil contamination

  15. Preliminary thermal/thermomechanical analyses of the Site Characterization Plan's Conceptual Design for a repository containing horizontally emplaced waste packages at the Deaf Smith County site

    International Nuclear Information System (INIS)

    Ghantous, N.Y.; Raines, G.E.

    1987-10-01

    This report presents thermal/thermomechanical analyses of the Site Characterization Plan Conceptual Design for horizontal package emplacement at the Deaf Smith County site, Texas. The repository was divided into three geometric regions. Then two-dimensional finite-element models were set up to approximate the three-dimensional nature of each region. Thermal and quasistatic thermomechanical finite-element analyses were performed to evaluate the thermal/thermomechanical responses of the three regions. The exponential-time creep law was used to represent the creep behavior of salt rock. The repository design was evaluated by comparing the thermal/thermomechanical responses obtained for the three regions with interim performance constraints. The preliminary results show that all the performance constraints are met except for those of the waste package. The following factors were considered in interpreting these results: (1) the qualitative description of the analytical responses; (2) the limitations of the analyses; and (3) either the conclusions based on overall evaluation of limitations and analytical results or the conclusions based on the fact that the repository design may be evaluated only after further analyses. Furthermore, a parametric analysis was performed to estimate the effect of material parameters on the predicted thermal/thermomechanical response. 23 refs., 34 figs., 9 tabs

  16. AATSR land surface temperature product algorithm verification over a WATERMED site

    Science.gov (United States)

    Noyes, E. J.; Sòria, G.; Sobrino, J. A.; Remedios, J. J.; Llewellyn-Jones, D. T.; Corlett, G. K.

    A new operational Land Surface Temperature (LST) product generated from data acquired by the Advanced Along-Track Scanning Radiometer (AATSR) provides the opportunity to measure LST on a global scale with a spatial resolution of 1 km2. The target accuracy of the product, which utilises nadir data from the AATSR thermal channels at 11 and 12 μm, is 2.5 K for daytime retrievals and 1.0 K at night. We present the results of an experiment where the performance of the algorithm has been assessed for one daytime and one night time overpass occurring over the WATERMED field site near Marrakech, Morocco, on 05 March 2003. Top of atmosphere (TOA) brightness temperatures (BTs) are simulated for 12 pixels from each overpass using a radiative transfer model, with the LST product and independent emissivity values and atmospheric data as inputs. We have estimated the error in the LST product over this biome for this set of conditions by applying the operational AATSR LST retrieval algorithm to the modelled BTs and comparing the results with the original AATSR LSTs input into the model. An average bias of -1.00 K (standard deviation 0.07 K) for the daytime data, and -1.74 K (standard deviation 0.02 K) for the night time data is obtained, which indicates that the algorithm is yielding an LST that is too cold under these conditions. While these results are within specification for daytime retrievals, this suggests that the target accuracy of 1.0 K at night is not being met within this biome.

  17. Geochemical Data Package for Performance Assessment Calculations Related to the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, Daniel I. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2010-03-15

    The Savannah River Site disposes of low-activity radioactive waste within subsurface-engineered facilities. One of the tools used to establish the capacity of a given site to safely store radioactive waste (i.e., that a site does not exceed its Waste Acceptance Criteria) is the Performance Assessment (PA). The objective of this document is to provide the geochemical values for the PA calculations. This work is being conducted as part of the on-going maintenance program that permits the PA to periodically update existing calculations when new data become available.

  18. May 2012 Groundwater and Surface Water Sampling at the Rio Blanco, Colorado, Site (Data Validation Package)

    International Nuclear Information System (INIS)

    2012-01-01

    Annual sampling was conducted at the Rio Blanco, Colorado, site for the Long-Term Hydrologic Monitoring Program May 9-10, 2012, to monitor groundwater and surface water for potential radionuclide contamination. Sampling and analyses were conducted as specified in Sampling and Analysis Plan for the U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). A duplicate sample was collected from location Johnson Artesian WL. Samples were analyzed for gamma-emitting radionuclides by high-resolution gamma spectrometry and for tritium using the conventional and enrichment methods. Results of this monitoring at the Rio Blanco site demonstrate that groundwater and surface water outside the site boundaries have not been affected by project-related contaminants.

  19. January 2011 Groundwater Sampling at the Gnome-Coach, New Mexico, Site (Data Validation Package)

    International Nuclear Information System (INIS)

    2011-01-01

    Annual sampling was conducted January 19, 2011, to monitor groundwater for potential radionuclide contamination at the Gnome-Coach site in New Mexico. The sampling was performed as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PLN/S04351, continually updated). Well LRL-7 was not sampled per instruction from the lead. A duplicate sample was collected from well USGS-1.Water levels were measured in the monitoring wells onsite.

  20. May 2011 Groundwater and Surface Water Sampling at the Rio Blanco, Colorado, Site (Data Validation Package)

    International Nuclear Information System (INIS)

    2011-01-01

    Annual sampling was conducted at the Rio Blanco, Colorado, site for the Long-Term Hydrologic Monitoring Program May 16-17, 2011, to monitor groundwater and surface water for potential radionuclide contamination. Sampling and analyses were conducted as specified in Sampling and Analysis Plan for the U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). A duplicate sample was collected from location Johnson Artesian WL. Samples were analyzed by the U.S. Environmental Protection Agency (EPA) Radiation&Indoor Environments National Laboratory in Las Vegas, Nevada. Samples were analyzed for gamma-emitting radionuclides by high-resolution gamma spectrometry, and for tritium using the conventional method. Tritium was not measured using the enrichment method because the EPA laboratory no longer offers that service. Results of this monitoring at the Rio Blanco site demonstrate that groundwater and surface water outside the boundaries have not been affected by project-related contaminants.

  1. January 2012 Groundwater Sampling at the Gnome-Coach, New Mexico, Site (Data Validation Package)

    International Nuclear Information System (INIS)

    2012-01-01

    Annual sampling was conducted January 18, 2012, to monitor groundwater for potential radionuclide contamination at the Gnome-Coach site in New Mexico. The sampling was performed as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PLN/S04351, continually updated). Well LRL-7 was not sampled per instruction from the lead. A duplicate sample was collected from well USGS-1 and water levels were measured in the monitoring wells onsite.

  2. January 2012 Groundwater Sampling at the Gnome-Coach, New Mexico, Site (Data Validation Package)

    Energy Technology Data Exchange (ETDEWEB)

    Findlay, Richard C. [S.M. Stoller Corporation, Broomfield, CO (United States)

    2012-12-01

    Annual sampling was conducted January 18, 2012, to monitor groundwater for potential radionuclide contamination at the Gnome-Coach site in New Mexico. The sampling was performed as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PLN/S04351, continually updated). Well LRL-7 was not sampled per instruction from the lead. A duplicate sample was collected from well USGS-1 and water levels were measured in the monitoring wells onsite.

  3. May 2013 Groundwater and Surface Water Sampling at the Rio Blanco, Colorado, Site (Data Validation Package)

    Energy Technology Data Exchange (ETDEWEB)

    Hutton, Rick [S.M. Stoller Corporation, Broomfield, CO (United States)

    2013-10-01

    Annual sampling was conducted at the Rio Blanco, Colorado, site for the Long-Term Hydrologic Monitoring Program May 14-16, 2013, to monitor groundwater and surface water for potential radionuclide contamination. Sampling and analyses were conducted as specified in Sampling and Analysis Plan for the U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). A duplicate sample was collected from location CER #1 Black Sulphur. Samples were analyzed for gamma-emitting radionuclides by high-resolution gamma spectrometry and for tritium using the conventional and enrichment methods.

  4. Data Validation Package, December 2015, Groundwater Sampling at the Bluewater, New Mexico, Disposal Site, September 2016

    Energy Technology Data Exchange (ETDEWEB)

    Tsosie, Bernadette [U. S. Department of Energy, Washington, DC (United States). Office of Legacy Management; Johnson, Richard [Navarro Research and Engineering, Oak Ridge, TN (United States)

    2016-09-01

    Groundwater samples were collected from monitoring wells at the Bluewater, New Mexico, Disposal Site to monitor groundwater contaminants as specified in the 1997 Long-Term Surveillance Plan for the DOE Bluewater (UMTRCA Title II) Disposal Site Near Grants, New Mexico (LTSP). Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). A duplicate sample was collected from location HMC-951. Alluvium wells are completed in the alluvial sediments in the former channel of the Rio San Jose, which was covered by basalt lava flows known as the El Malpais, and are identified by the suffix (M). Bedrock wells are completed in the San Andres Limestone/Glorieta Sandstone hydrologic unit (San Andres aquifer) and are identified by the suffix (SG). Wells HMC-951 and OBS-3 are also completed in the San Andres aquifer. The LTSP requires monitoring for molybdenum, selenium, uranium, and polychlorinated biphenyls (PCBs); PCB monitoring occurs only during November sampling events. This event included sampling for an expanded list of analytes to characterize the site aquifers and to support a regional groundwater investigation being conducted by the New Mexico Environment Department.

  5. Data Validation Package - July 2016 Groundwater Sampling at the Gunnison, Colorado, Disposal Site

    Energy Technology Data Exchange (ETDEWEB)

    Linard, Joshua [USDOE Office of Legacy Management, Washington, DC (United States); Campbell, Sam [Navarro Research and Engineering, Inc., Las Vegas, NV (United States)

    2016-10-25

    Groundwater sampling at the Gunnison, Colorado, Disposal Site is conducted every 5 years to monitor disposal cell performance. During this event, samples were collected from eight monitoring wells as specified in the 1997 Long-Term Surveillance Plan for the Gunnison, Colorado, Disposal Site. Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for US Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated, http://energy.gov/lm/downloads/sampling-and­ analysis-plan-us-department-energy-office-legacy-management-sites). Planned monitoring locations are shown in Attachment 1, Sampling and Analysis Work Order. A duplicate sample was collected from location 0723. Water levels were measured at all monitoring wells that were sampled and seven additional wells. The analytical data and associated qualifiers can be viewed in environmental database reports and are also available for viewing with dynamic mapping via the GEMS (Geospatial Environmental Mapping System) website at http://gems.lm.doe.gov/#. No issues were identified during the data validation process that require additional action or follow-up.

  6. Data Validation Package, June 2016 Groundwater Sampling at the Hallam, Nebraska, Decommissioned Reactor Site, August 2016

    Energy Technology Data Exchange (ETDEWEB)

    Surovchak, Scott [USDOE Office of Legacy Management, Washington, DC (United States); Miller, Michele [Navarro Research and Engineering, Oak Ridge, TN (United States)

    2016-08-01

    The 2008 Long-Term Surveillance Plan [LTSP] for the Decommissioned Hallam Nuclear Power Facility, Hallam, Nebraska (http://www.lm.doe.gov/Hallam/Documents.aspx) requires groundwater monitoring once every 2 years. Seventeen monitoring wells at the Hallam site were sampled during this event as specified in the plan. Planned monitoring locations are shown in Attachment 1, Sampling and Analysis Work Order. Water levels were measured at all sampled wells and at two additional wells (6A and 6B) prior to the start of sampling. Additionally, water levels of each sampled well were measured at the beginning of sampling. See Attachment 2, Trip Report, for additional details. Sampling and analysis were conducted as specified in Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated, http://energy.gov/lm/downloads/sampling-and-analysis-plan-us-department- energy-office-legacy-management-sites). Gross alpha and gross beta are the only parameters that were detected at statistically significant concentrations. Time/concentration graphs of the gross alpha and gross beta data are included in Attachment 3, Data Presentation. The gross alpha and gross beta activity concentrations observed are consistent with values previously observed and are attributed to naturally occurring radionuclides (e.g., uranium and uranium decay chain products) in the groundwater.

  7. Data Validation Package: April 2016 Groundwater Sampling at the Falls City, Texas, Disposal Site

    Energy Technology Data Exchange (ETDEWEB)

    Jasso, Tashina [USDOE Office of Legacy Management, Washington, DC (United States); Widdop, Michael [Navarro Research and Engineering, Inc., Las Vegas, NV (United States)

    2016-09-29

    Nine groundwater samples were collected at the Falls City, Texas, Disposal Site as specified in the March 2008 Long-Term Surveillance Plan for the US Department of Energy Falls City Uranium Mill Tailings Disposal Site, Falls City, Texas (DOE-LM/1602-2008). Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for US Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). The wells sampled included the cell performance monitoring wells (0709, 0858, 0880, 0906, and 0921) and the groundwater monitoring wells (0862, 0886, 0891, 0924, and 0963). A duplicate sample was collected from location 0891. Water levels were measured at each sampled well. Historically, cell performance monitoring wells 0908 and 0916 have not produced water and were confirmed as dry during this sampling event. These wells are completed above the saturated interval in the formation. Notable observations for time-concentration graphs in this report include: (1) uranium concentrations in well 0891 continue to increase; (2) the uranium concentration in well 0880 is higher than the 2015 value and lower than the 2014 value, and it remains within the range of historical values; and (3) uranium concentrations in the other sampled wells are below 2 mg/L and consistent with previous results.

  8. Data Validation Package May 2016 Groundwater Sampling at the Lakeview, Oregon, Processing Site August 2016

    Energy Technology Data Exchange (ETDEWEB)

    Linard, Joshua [USDOE Office of Legacy Management, Washington, DC (United States); Hall, Steve [Navarro Research and Engineering, Inc., Oak Ridge, TN (United States)

    2016-08-01

    This biennial event includes sampling five groundwater locations (four monitoring wells and one domestic well) at the Lakeview, Oregon, Processing Site. For this event, the domestic well (location 0543) could not be sampled because no one was in residence during the sampling event (Note: notification was provided to the resident prior to the event). Per Appendix A of the Groundwater Compliance Action Plan, sampling is conducted to monitor groundwater quality on a voluntary basis. Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). One duplicate sample was collected from location 0505. Water levels were measured at each sampled monitoring well. The constituents monitored at the Lakeview site are manganese and sulfate. Monitoring locations that exceeded the U.S. Environmental Protection Agency (EPA) Secondary Maximum Contaminant Levels for these constituents are listed in Table 1. Review of time-concentration graphs included in this report indicate that manganese and sulfate concentrations are consistent with historical measurements.

  9. Development of a design package for a viscous barrier at the Savannah River Site

    International Nuclear Information System (INIS)

    Moridis, G.J.; James, A.; Oldenburg, C.

    1996-10-01

    This paper describes elements of a design for a pilot-scale field demonstration of a new subsurface containment technology for waste isolation developed at the Lawrence Berkeley National Laboratory (LBNL), which uses a new generation of barrier liquids for permeation grouting. The demonstration site was Retention Basin 281-3H, a shallow catchment basin at the Savannah River Site (SRS), originally built to control contaminated runoff for the H Reactor, and which has been contaminated mainly by radionuclides. The LBNL viscous barrier technology employs barrier liquids which, when injected into the subsurface, produce chemically benign nearly impermeable barriers through a very large increase in viscosity. The initially low-viscosity liquids are emplaced through multiple injection points in the subsurface and the intersecting plumes merge and completely surround the contaminant source and/or plume. Once in place, they gel or cure to form a nearly impermeable barrier. The barrier liquid to be used in this application is Colloidal Silica (CS), an aqueous suspension of silica microspheres in a stabilizing electrolyte. It has excellent durability characteristics, poses no health hazard, is practically unaffected by filtration, and is chemically and biologically benign

  10. Data Validation Package May 2016 Groundwater Sampling at the Sherwood, Washington, Disposal Site August 2016

    Energy Technology Data Exchange (ETDEWEB)

    Kreie, Ken [USDOE Office of Legacy Management, Washington, DC (United States); Traub, David [Navarro Research and Engineering, Inc., Oak Ridge, TN (United States)

    2016-08-04

    The 2001 Long-Term Surveillance Plan (LTSP) for the US. Department of Energy Sherwood Project (UMI'RCA Title II) Reclamation Cell, Wellpinit, Washington, does not require groundwater compliance monitoring at the Sherwood site. However, the LTSP stipulates limited groundwater monitoring for chloride and sulfate (designated indicator parameters) and total dissolved solids (TDS) as a best management practice. Samples were collected from the background well, MW-2B, and the two downgradient wells, MW-4 and MW-10, in accordance with the LTSP. Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for US. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). Water levels were measured in all wells prior to sampling and in four piezometers completed in the tailings dam. Time-concentration graphs included in this report indicate that the chloride, sulfate, and TDS concentrations are consistent with historical measurements. The concentrations of chloride and sulfate are well below the State of Washington water quality criteria value of 250 milligrams per liter (mg/L) for both parameters.

  11. M3 version 3.0: Verification and validation; Hydrochemical model of ground water at repository site

    Energy Technology Data Exchange (ETDEWEB)

    Gomez, Javier B. (Dept. of Earth Sciences, Univ. of Zaragoza, Zaragoza (Spain)); Laaksoharju, Marcus (Geopoint AB, Sollentuna (Sweden)); Skaarman, Erik (Abscondo, Bromma (Sweden)); Gurban, Ioana (3D-Terra (Canada))

    2009-01-15

    Hydrochemical evaluation is a complex type of work that is carried out by specialists. The outcome of this work is generally presented as qualitative models and process descriptions of a site. To support and help to quantify the processes in an objective way, a multivariate mathematical tool entitled M3 (Multivariate Mixing and Mass balance calculations) has been constructed. The computer code can be used to trace the origin of the groundwater, and to calculate the mixing proportions and mass balances from groundwater data. The M3 code is a groundwater response model, which means that changes in the groundwater chemistry in terms of sources and sinks are traced in relation to an ideal mixing model. The complexity of the measured groundwater data determines the configuration of the ideal mixing model. Deviations from the ideal mixing model are interpreted as being due to reactions. Assumptions concerning important mineral phases altering the groundwater or uncertainties associated with thermodynamic constants do not affect the modelling because the calculations are solely based on the measured groundwater composition. M3 uses the opposite approach to that of many standard hydrochemical models. In M3, mixing is evaluated and calculated first. The constituents that cannot be described by mixing are described by reactions. The M3 model consists of three steps: the first is a standard principal component analysis, followed by mixing and finally mass balance calculations. The measured groundwater composition can be described in terms of mixing proportions (%), while the sinks and sources of an element associated with reactions are reported in mg/L. This report contains a set of verification and validation exercises with the intention of building confidence in the use of the M3 methodology. At the same time, clear answers are given to questions related to the accuracy and the precision of the results, including the inherent uncertainties and the errors that can be made

  12. Data Validation Package May 2015, Groundwater Sampling at the Shoal, Nevada, Site

    International Nuclear Information System (INIS)

    Findlay, Rick; Kautsky, Mark

    2016-01-01

    The U.S. Department of Energy Office of Legacy Management conducted annual sampling at the Shoal, Nevada, Site (Shoal) in May 2015. Groundwater samples were collected from wells MV-1, MV-2, MV-3, MV-4, MV-5, H-3, HC-1, HC-2d, HC-3, HC-4, HC-5, HC-6, HC-7, HC-8, and HS-1. Sampling was conducted as specified in the Sampling and Analysis Plan for US. Department of Energy Office of Legacy Management Sites: LMS/PRO/S04351, continually updated, http://energy.gov/lm/downloads/sampling-and-analysis-plan-us-department- energy-office-legacy-management-sites. Monitoring wells MV-1, MV-2, MV-3, MV-4, MV-5, HC-2d, HC-4, HC-5, HC-7, HC-8, and HS-1 were purged prior to sampling using dedicated submersible pumps. At least one well casing volume was removed, and field parameters (temperature, pH, and specific conductance) were allowed to stabilize before samples were collected. Samples were collected from wells H-3, HC-1, HC-3, and HC-6 using a depth-specific bailer because these wells are not completed with dedicated submersible pumps. Samples were submitted under Requisition Index Number (RIN) 15057042 to ALS Laboratory Group in Fort Collins, Colorado, for the determination of bromide, gross alpha, gross beta, tritium, uranium isotopes, and total uranium (by mass); and under RIN 15057043 to the University of Arizona for the determination of carbon-14 and iodine-129. A duplicate sample from location MV-2 was included with RIN 15057042. The laboratory results from the 2015 sampling event are consistent with those of previous years with the exception of sample results from well HC-4. This well continues to be the only well with tritium concentrations above the laboratory's minimum detectable concentration which is attributed to the wells proximity to the nuclear detonation. The tritium concentration (731 picocuries per liter [pCi/L]) is consistent with past results and is below the U.S. Environmental Protection Agency's (EPA) maximum contaminant level (MCL) of 20,000 p

  13. Data Validation Package May 2015, Groundwater Sampling at the Shoal, Nevada, Site

    Energy Technology Data Exchange (ETDEWEB)

    Findlay, Rick [Navarro Research and Engineering, Oak Ridge, TN (United States); Kautsky, Mark [US Department of Energy, Washington, DC (United States). Office of Legacy Management

    2016-03-01

    The U.S. Department of Energy Office of Legacy Management conducted annual sampling at the Shoal, Nevada, Site (Shoal) in May 2015. Groundwater samples were collected from wells MV-1, MV-2, MV-3, MV-4, MV-5, H-3, HC-1, HC-2d, HC-3, HC-4, HC-5, HC-6, HC-7, HC-8, and HS-1. Sampling was conducted as specified in the Sampling and Analysis Plan for US. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated, http://energy.gov/lm/downloads/sampling-and-analysis-plan-us-department-energy­ office-legacy-management-sites). Monitoring wells MV-1, MV-2, MV-3, MV-4, MV-5, HC-2d, HC-4, HC-5, HC-7, HC-8, and HS-1 were purged prior to sampling using dedicated submersible pumps. At least one well casing volume was removed, and field parameters (temperature, pH, and specific conductance) were allowed to stabilize before samples were collected. Samples were collected from wells H-3, HC-1, HC-3, and HC-6 using a depth-specific bailer because these wells are not completed with dedicated submersible pumps. Samples were submitted under Requisition Index Number (RIN) 15057042 to ALS Laboratory Group in Fort Collins, Colorado, for the determination of bromide, gross alpha, gross beta, tritium, uranium isotopes, and total uranium (by mass); and under RIN 15057043 to the University of Arizona for the determination of carbon-14 and iodine-129. A duplicate sample from location MV-2 was included with RIN 15057042. The laboratory results from the 2015 sampling event are consistent with those of previous years with the exception of sample results from well HC-4. This well continues to be the only well with tritium concentrations above the laboratory’s minimum detectable concentration which is attributed to the wells proximity to the nuclear detonation. The tritium concentration (731 picocuries per liter [pCi/L]) is consistent with past results and is below the U.S. Environmental Protection Agency's (EPA) maximum contaminant level (MCL) of 20,000 p

  14. Remaining Sites Verification Package for the 100-D-50:5 Process Sewers (183-DR Sedimentation Basin Drains). Attachment to Waste Site Reclassification Form 2006-025

    International Nuclear Information System (INIS)

    Dittmer, L.M.

    2007-01-01

    The 100-D-50:5 subsite encompasses the southern process sewers formerly servicing the 183-DR coagulation and sedimentation basins and proximate surface runoff collection drains. The results of confirmatory sampling of pipeline sediments and underlying soils at the 100-D-50:5 subsite demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River

  15. Verification study on technology for preliminary investigation for HLW geological disposal. Part 2. Verification of surface geophysical prospecting through establishing site descriptive models

    International Nuclear Information System (INIS)

    Kondo, Hirofumi; Suzuki, Koichi; Hasegawa, Takuma; Goto, Keiichiro; Yoshimura, Kimitaka; Muramoto, Shigenori

    2012-01-01

    The Yokosuka demonstration and validation project using Yokosuka CRIEPI site has been conducted since FY 2006 as a cooperative research between NUMO (Nuclear Waste Management Organization of Japan) and CRIEPI. The objectives of this project are to examine and to refine the basic methodology of the investigation and assessment of properties of geological environment in the stage of Preliminary Investigation for HLW geological disposal. Within Preliminary Investigation technologies, surface geophysical prospecting is an important means of obtaining information from deep geological environment for planning borehole surveys. In FY 2010, both seismic prospecting (seismic reflection and vertical seismic profiling methods) for obtaining information about geological structure and electromagnetic prospecting (magneto-telluric and time domain electromagnetic methods) for obtaining information about resistivity structure reflecting the distribution of salt water/fresh water boundary to a depth of over several hundred meters were conducted in the Yokosuka CRIEPI site. Through these surveys, the contribution of geophysical prospecting methods in the surface survey stage to improving the reliability of site descriptive models was confirmed. (author)

  16. Results of the independent radiological verification survey at the former Associate Aircraft Tool and Manufacturing Company site, Fairfield, Ohio (FOH001)

    International Nuclear Information System (INIS)

    Rice, D.E.; Murray, M.E.; Brown, K.S.

    1996-01-01

    The former Associate Aircraft Tool and Manufacturing Company site is located at 3550 Dixie Highway, Fairfield, Ohio. Associate Aircraft Tool and Manufacturing Company produced hollow uranium slugs in a machine shop at the site in 1956. The work was performed for National Lead of Ohio in a contract with the Atomic Energy Commission to augment the capacity of the Feed Materials Production Center at Fernald in the development of nuclear energy for defense-related projects. The current occupant of the building, Force Control, operates a multipurpose machine shop. At the request of the US Department of Energy (DOE), a team from Oak Ridge National Laboratory conducted an independent radiological verification survey at the former Associate Aircraft Tool and Manufacturing Company Site, Fairfield, Ohio. The survey was performed from February to May of 1995. The purpose of the survey was to verify that radioactivity from residues of 238 U was remediated to a level below acceptable DOE guidelines levels

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

    International Nuclear Information System (INIS)

    1987-06-01

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

  18. Evaluation of three state-of-the-art metabolite prediction software packages (Meteor, MetaSite, and StarDrop) through independent and synergistic use.

    Science.gov (United States)

    T'jollyn, H; Boussery, K; Mortishire-Smith, R J; Coe, K; De Boeck, B; Van Bocxlaer, J F; Mannens, G

    2011-11-01

    The aim of this study was to evaluate three different metabolite prediction software packages (Meteor, MetaSite, and StarDrop) with respect to their ability to predict loci of metabolism and suggest relative proportions of metabolites. A chemically diverse test set of 22 compounds, for which in vivo human mass balance studies and metabolic schemes were available, was used as basis for the evaluation. Each software package was provided with structures of the parent compounds, and predicted metabolites were compared with experimentally determined human metabolites. The evaluation consisted of two parts. First, different settings within each software package were investigated and the software was evaluated using those settings determined to give the best prediction. Second, the three different packages were combined using the optimized settings to see whether a synergistic effect concerning the overall metabolism prediction could be established. The performance of the software was scored for both sensitivity and precision, taking into account the capabilities/limitations of the particular software. Varying results were obtained for the individual packages. Meteor showed a general tendency toward overprediction, and this led to a relatively low precision (∼35%) but high sensitivity (∼70%). MetaSite and StarDrop both exhibited a sensitivity and precision of ∼50%. By combining predictions obtained with the different packages, we found that increased precision can be obtained. We conclude that the state-of-the-art individual metabolite prediction software has many advantageous features but needs refinement to obtain acceptable prediction profiles. Synergistic use of different software packages could prove useful.

  19. FMCT verification: Case studies

    International Nuclear Information System (INIS)

    Hui Zhang

    2001-01-01

    for states that have traditionally had 'less transparency' in their military sectors. As case studies, first we investigate how to applied verification measures including remote sensing, off-site environmental sampling and on-site inspections to monitor the shutdown status of plutonium production facilities, and what measures could be taken to prevent the disclosure of sensitive information at the site. We find the most effective verification measure to monitor the status of the reprocessing plant would be on-site environmental sampling. Some countries may worry that sample analysis could disclose sensitive information about their past plutonium production activities. However, we find that sample analysis at the reprocessing site need not reveal such information. Sampling would not reveal such information as long as inspectors are not able to measure total quantities of Cs-137 and Sr-90 from HLW produced at former military plutonium production facilities. Secondly, we consider verification measures for shutdown gaseous diffusion uranium-enrichment plants (GDPs). The GDPs could be monitored effectively by satellite imagery, as one telltale operational signature of the GDP would be the water-vapor plume coming from the cooling tower, which should be easy to detect with satellite images. Furthermore, the hot roof of the enrichment building could be detectable using satellite thermal-infrared images. Finally, some on-site verification measures should be allowed, such as visual observation, surveillance and tamper-indicating seals. Finally, FMCT verification regime would have to be designed to detect undeclared fissile material production activities and facilities. These verification measures could include something like special or challenge inspections or complementary access. There would need to be provisions to prevent the abuse of such inspections, especially at sensitive and non-proscribed military and nuclear activities. In particular, to protect sensitive

  20. Verification test for three WindCube WLS7 LiDARs at the Høvsøre test site

    DEFF Research Database (Denmark)

    Gottschall, Julia; Courtney, Michael

    The report describes the procedure of testing ground-based WindCube lidars (manufactured by the French company Leosphere) at the Høvsøre test site in comparison to reference sensors mounted at a meteorological mast. Results are presented for three tested units – in detail for unit WLS7-0062, and ......-0062, and in a summary for units WLS7-0064 and WLS7-0066. The verification test covers the evaluation of measured mean wind speeds, wind directions and wind speed standard deviations. The data analysis is basically performed in terms of different kinds of regression analyses.......The report describes the procedure of testing ground-based WindCube lidars (manufactured by the French company Leosphere) at the Høvsøre test site in comparison to reference sensors mounted at a meteorological mast. Results are presented for three tested units – in detail for unit WLS7...

  1. A hydrochemical data base for the Hanford Site, Washington

    International Nuclear Information System (INIS)

    Early, T.O.; Mudd, R.D.; Spice, G.D.; Starr, D.L.

    1985-02-01

    This data package contains a complete listing of the Site Hydrochemical Data Base for water samples associated with the Basalt Waste Isolation Project (BWIP). In addition to the detailed chemical analyses are a summary description of the data base format, detailed descriptions of verification procedures used to check data entries, and detailed descriptions of validation procedures used to evaluate data quality

  2. Thermal Pollution Math Model. Volume 1. Thermal Pollution Model Package Verification and Transfer. [environment impact of thermal discharges from power plants

    Science.gov (United States)

    Lee, S. S.; Sengupta, S.

    1980-01-01

    Two three dimensional, time dependent models, one free surface, the other rigid lid, were verified at Anclote Anchorage and Lake Keowee respectively. The first site is a coastal site in northern Florida; the other is a man-made lake in South Carolina. These models describe the dispersion of heated discharges from power plants under the action of ambient conditions. A one dimensional, horizontally-averaged model was also developed and verified at Lake Keowee. The data base consisted of archival in situ measurements and data collected during field missions. The field missions were conducted during winter and summer conditions at each site. Each mission consisted of four infrared scanner flights with supporting ground truth and in situ measurements. At Anclote, special care was taken to characterize the complete tidal cycle. The three dimensional model results compared with IR data for thermal plumes on an average within 1 C root mean square difference. The one dimensional model performed satisfactorily in simulating the 1971-1979 period.

  3. Global Structure of a Three-Way Junction in a Phi29 Packaging RNA Dimer Determined Using Site-Directed Spin Labeling

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Xiaojun; Tung, Chang-Shung; Sowa, Glenna; Hatmal, Ma' mon M.; Haworth, Ian S.; Qin, Peter Z.

    2012-02-08

    The condensation of bacteriophage phi29 genomic DNA into its preformed procapsid requires the DNA packaging motor, which is the strongest known biological motor. The packaging motor is an intricate ring-shaped protein/RNA complex, and its function requires an RNA component called packaging RNA (pRNA). Current structural information on pRNA is limited, which hinders studies of motor function. Here, we used site-directed spin labeling to map the conformation of a pRNA three-way junction that bridges binding sites for the motor ATPase and the procapsid. The studies were carried out on a pRNA dimer, which is the simplest ring-shaped pRNA complex and serves as a functional intermediate during motor assembly. Using a nucleotide-independent labeling scheme, stable nitroxide radicals were attached to eight specific pRNA sites without perturbing RNA folding and dimer formation, and a total of 17 internitroxide distances spanning the three-way junction were measured using Double Electron-Electron Resonance spectroscopy. The measured distances, together with steric chemical constraints, were used to select 3662 viable three-way junction models from a pool of 65 billion. The results reveal a similar conformation among the viable models, with two of the helices (HT and HL) adopting an acute bend. This is in contrast to a recently reported pRNA tetramer crystal structure, in which HT and HL stack onto each other linearly. The studies establish a new method for mapping global structures of complex RNA molecules, and provide information on pRNA conformation that aids investigations of phi29 packaging motor and developments of pRNA-based nanomedicine and nanomaterial.

  4. International ENEA/ISMES/ENS specialist meeting on 'On-site experimental verification of the seismic behaviour of nuclear reactor structures and components'. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-07-01

    The seismic verification of nuclear plants is a subject of increasing interest in all the industrial countries, with respect to both the safety aspects and the impact of the seismic event on the design and the costs of a nuclear reactor. This topic is especially of great interest for a country like Italy, whose territory is unfortunately characterized by non - negligible seismicity: we remember, not too many years ago, the catastrophic earthquakes of Frioul and Irpinia, that caused thousands of dead people. The meeting aimed at establishing the state-of-the-art on on-site testing of nuclear reactors structures and components, with particular attention to experiences and research programmes concerning: methodologies of on-site tests and interpretation of the experimental data; seismic monitoring systems, recorded data, their use and interpretation; calibration and validation of numerical analyses. Six technical sessions were held, during which 23 high papers were presented and discussed, and six panel discussions were held (the importance of discussion was emphasized in the meeting). The technical contributions consisted of: an introduction paper, summarizing the seismic studies performed in Italy for PEC reactor and explaining the reasons why on-site tests had been performed on this reactor; 6 invited lectures, one for each of the countries that are more deeply involved in seismic analysis, providing the state-of-the-art on the topics of interest for the meeting; 16 contributed papers dealing with more specific technical items, related to the various countries and international organizations.

  5. International ENEA/ISMES/ENS specialist meeting on 'On-site experimental verification of the seismic behaviour of nuclear reactor structures and components'. Proceedings

    International Nuclear Information System (INIS)

    1988-01-01

    The seismic verification of nuclear plants is a subject of increasing interest in all the industrial countries, with respect to both the safety aspects and the impact of the seismic event on the design and the costs of a nuclear reactor. This topic is especially of great interest for a country like Italy, whose territory is unfortunately characterized by non - negligible seismicity: we remember, not too many years ago, the catastrophic earthquakes of Frioul and Irpinia, that caused thousands of dead people. The meeting aimed at establishing the state-of-the-art on on-site testing of nuclear reactors structures and components, with particular attention to experiences and research programmes concerning: methodologies of on-site tests and interpretation of the experimental data; seismic monitoring systems, recorded data, their use and interpretation; calibration and validation of numerical analyses. Six technical sessions were held, during which 23 high papers were presented and discussed, and six panel discussions were held (the importance of discussion was emphasized in the meeting). The technical contributions consisted of: an introduction paper, summarizing the seismic studies performed in Italy for PEC reactor and explaining the reasons why on-site tests had been performed on this reactor; 6 invited lectures, one for each of the countries that are more deeply involved in seismic analysis, providing the state-of-the-art on the topics of interest for the meeting; 16 contributed papers dealing with more specific technical items, related to the various countries and international organizations

  6. Technical specifications for waste packages conditioned in a durable confining shell, with an hydraulic binder basis, intended to a ground disposal site

    International Nuclear Information System (INIS)

    1995-06-01

    The aim of this document is to precise the general and particular conditions for the acceptance on a ground disposal site of a low- and middle-level radioactive waste package conditioned in a durable confining shell. This specification concerns the wastes that contain beta and gamma decay radionuclides and/or long life alpha decay radionuclides in higher quantities than accepted for the protective coatings. Physico-chemical and mechanical specifications are given for the wastes, the fixing material, the confining shell and the container. Accepted limits for degassing and dose rates, surface contamination, dimensions and weight are given. The agreement is delivered by the ANDRA after the package has satisfied the different mechanical, chemical, fire, moisture and radiation resistance tests. (J.S.). 1 fig., 3 tabs., 1 glossary

  7. Verification of geomechanical integrity and prediction of long-term mineral trapping for the Ketzin CO2 storage pilot site

    Science.gov (United States)

    Kempka, Thomas; De Lucia, Marco; Kühn, Michael

    2014-05-01

    Static and dynamic numerical modelling generally accompany the entire CO2 storage site life cycle. Thereto, it is required to match the employed models with field observations on a regular basis in order to predict future site behaviour. We investigated the coupled processes at the Ketzin CO2 storage pilot site [1] using a model coupling concept focusing on the temporal relevance of processes involved (hydraulic, chemical and mechanical) at given time-scales (site operation, abandonment and long-term stabilization). For that purpose, long-term dynamic multi-phase flow simulations [2], [3] established the basis for all simulations discussed in the following. Hereby, pressure changes resulting in geomechanical effects are largest during site operation, whereas geochemical reactions are governed by slow kinetics resulting in a long-term stabilization. To account for mechanical integrity, which may be mainly affected during site operation, we incorporated a regional-scale coupled hydro-mechanical model. Our simulation results show maximum ground surface displacements of about 4 mm, whereas shear and tensile failure are not observed. Consequently, the CO2 storage operation at the Ketzin pilot site does not compromise reservoir, caprock and fault integrity. Chemical processes responsible for mineral trapping are expected to mainly occur during long-term stabilization at the Ketzin pilot site [4]. Hence, our previous assessment [3] was extended by integrating two long-term mineral trapping scenarios. Thereby, mineral trapping contributes to the trapping mechanisms with 11.7 % after 16,000 years of simulation in our conservative and with 30.9 % in our maximum reactivity scenarios. Dynamic flow simulations indicate that only 0.2 % of the CO2 injected (about 67,270 t CO2 in total) is in gaseous state, but structurally trapped after 16,000 years. Depending on the studied long-term scenario, CO2 dissolution is the dominating trapping mechanism with 68.9 % and 88

  8. Lively package

    International Nuclear Information System (INIS)

    Jaremko, G.

    1997-01-01

    Progress on the Lloydminster Heavy Oil Interpretive Centre, sponsored by the Lloydminster Oilfield Technical Society and expected to open in late 1998, was discussed. Some $150,000 of the $750,000 budget is already in the bank, and another $150,000 is in the pipeline. The Centre will be added to an existing and well-established visitor's site. It is reported to contain a lively and imaginatively-designed exhibit package, and promises to become a combination of educational tool and tourist attraction for the town of Lloydminster, Saskatchewan, in the heart of heavy oil country

  9. Data Validation Package September 2016 Groundwater and Surface Water Sampling at the Slick Rock, Colorado, Processing Sites January 2017

    Energy Technology Data Exchange (ETDEWEB)

    Traub, David [Navarro Research and Engineering, Inc., Oak Ridge, TN (United States); Nguyen, Jason [US Department of Energy, Washington, DC (United States)

    2017-01-04

    The Slick Rock, Colorado, Processing Sites are referred to as the Slick Rock West Processing Site (SRK05) and the Slick Rock East Processing Site (SRK06). This annual event involved sampling both sites for a total of 16 monitoring wells and 6 surface water locations as required by the 2006 Draft Final Ground Water Compliance Action Plan for the Slick Rock, Colorado, Processing Sites (GCAP). A domestic well was also sampled at a property adjacent to the Slick Rock East site at the request of the landowner.

  10. International Benchmark on Numerical Simulations for 1D, Nonlinear Site Response (PRENOLIN) : Verification Phase Based on Canonical Cases

    NARCIS (Netherlands)

    Régnier, Julie; Bonilla, Luis-Fabian; Bard, Pierre-Yves; Bertrand, Etienne; Hollender, Fabrice; Kawase, Hiroshi; Sicilia, Deborah; Arduino, Pedro; Amorosi, Angelo; Asimaki, Dominiki; Pisano, F.

    2016-01-01

    PREdiction of NOn‐LINear soil behavior (PRENOLIN) is an international benchmark aiming to test multiple numerical simulation codes that are capable of predicting nonlinear seismic site response with various constitutive models. One of the objectives of this project is the assessment of the

  11. HOW THE ROCKY FLATS ENVIRONMENTAL TECHNOLOGY SITE DEVELOPED A NEW WASTE PACKAGE USING A POLYUREA COATING THAT IS SAFELY AND ECONOMICALLY ELIMINATING SIZE REDUCTION OF LARGE ITEMS

    International Nuclear Information System (INIS)

    Dorr, Kent A.; Hogue, Richard S.; Kimokeo, Margaret K.

    2003-01-01

    One of the major challenges involved in closing the Rocky Flats Environmental Technology Site (RFETS) is the disposal of extremely large pieces of contaminated production equipment and building debris. Past practice has been to size reduce the equipment into pieces small enough to fit into approved, standard waste containers. Size reducing this equipment is extremely expensive, and exposes workers to high-risk tasks, including significant industrial, chemical, and radiological hazards. RFETS has developed a waste package using a Polyurea coating for shipping large contaminated objects. The cost and schedule savings have been significant

  12. Geochemical Processes Data Package for the Vadose Zone in the Single-Shell Tank Waste Management Areas at the Hanford Site

    Energy Technology Data Exchange (ETDEWEB)

    Cantrell, Kirk J.; Zachara, John M.; Dresel, P. Evan; Krupka, Kenneth M.; Serne, R. Jeffrey

    2007-09-28

    This data package discusses the geochemistry of vadose zone sediments beneath the single-shell tank farms at the U.S. Department of Energy’s (DOE’s) Hanford Site. The purpose of the report is to provide a review of the most recent and relevant geochemical process information available for the vadose zone beneath the single-shell tank farms and the Integrated Disposal Facility. Two companion reports to this one were recently published which discuss the geology of the farms (Reidel and Chamness 2007) and groundwater flow and contamination beneath the farms (Horton 2007).

  13. FEFTRA {sup TM} verification. Update 2013

    Energy Technology Data Exchange (ETDEWEB)

    Loefman, J. [VTT Technical Research Centre of Finland, Espoo (Finland); Meszaros, F. [The Relief Lab., Harskut, (Hungary)

    2013-12-15

    FEFTRA is a finite element program package developed at VTT for the analyses of groundwater flow in Posiva's site evaluation programme that seeks a final repository for spent nuclear fuel in Finland. The code is capable of modelling steady-state or transient groundwater flow, solute transport and heat transfer as coupled or separate phenomena. Being a typical research tool used only by its developers, the FEFTRA code lacked long of a competent testing system and precise documentation of the verification of the code. In 2006 a project was launched, in which the objective was to reorganise all the material related to the existing verification cases and place them into the FEFTRA program path under the version-control system. The work also included development of a new testing system, which automatically calculates the selected cases, checks the new results against the old approved results and constructs a summary of the test run. All the existing cases were gathered together, checked and added into the new testing system. The documentation of each case was rewritten with the LATEX document preparation system and added into the testing system in a way that the whole test documentation (this report) could easily be generated in a postscript or pdf-format. The current report is the updated version of the verification report published in 2007. At the moment the report includes mainly the cases related to the testing of the primary result quantities (i.e. hydraulic head, pressure, salinity concentration, temperature). The selected cases, however, represent typical hydrological applications, in which the program package has been and will be employed in the Posiva's site evaluation programme, i.e. the simulations of groundwater flow, solute transport and heat transfer as separate or coupled phenomena. The comparison of the FEFTRA results to the analytical, semianalytical and/or other numerical solutions proves the capability of FEFTRA to simulate such problems

  14. Verification study on technology for site investigation for geological disposal. Confirmation of the applicability of survey methods through establishing site descriptive models in accordance with stepwise investigation approach

    International Nuclear Information System (INIS)

    Kondo, Hirofumi; Suzuki, Koichi; Hasegawa, Takuma; Hamada, Takaomi; Yoshimura, Kimitaka

    2014-01-01

    The Yokosuka Demonstration and Validation Project, which uses the Yokosuka Central Research Institute of Electric Power Industry (CRIEPI) site, a Neogene sedimentary and coastal environment, has been conducted since the 2006 fiscal year as a cooperative research project between NUMO (Nuclear Waste Management Organization of Japan) and CRIEPI. The objectives of this project were to examine and refine the basic methodology of the investigation and assessment in accordance with the conditions of geological environment at each stage of investigations from the surface (Preliminary Investigation and the first half of Detailed Investigation conducted by NUMO) for high level radioactive waste geological disposal. Within investigation technologies at these early stages, a borehole survey is an important means of directly obtaining various properties of the deep geological environment. On the other hand, surface geophysical prospecting data provide information about the geological and resistivity structures at depth for planning borehole surveys. During the 2006-2009 fiscal years, a series of on-site surveys and tests, including borehole surveys of YDP-1 (depth: 350 m) and YDP-2 (depth: 500 m), were conducted in this test site. Furthermore, seismic surveys (including seismic reflection method) and electromagnetic surveys (including magnetotelluric method) were conducted within the expanded CRIEPI site in the 2010 fiscal year to obtain information about the geological structure, and the resistivity structure reflecting the distribution of the salt water/fresh water boundary, respectively, to a depth of over several hundred meters. The validity of existing survey and testing methods for stepwise investigations (from surface to borehole surveys) for obtaining properties of the geological environment (in various conditions relating to differences in the properties of the Miura and the Hayama Groups at this site) was confirmed through establishing site descriptive models based on

  15. Analysis and evaluation of the Gorleben site characteristics. Report on the working package 2. Preliminary safety analysis Gorleben

    International Nuclear Information System (INIS)

    Kukla, Peter; Pechnig, Renate; Urai, Janos

    2011-10-01

    The report on analysis and evaluation of the Gorleben site characteristics covers the following chapters: Site characteristics on the hydrology of the cover rock: hydrology and hydraulic properties, geothermal studies, ground water properties and motion, model calculations. Site characteristics on geology of the cover and adjoining rock: borehole explorations, geologic development, 3D geometry of the salt dome Gorleben, geologic development of the salt structure Gorleben. Site characteristics on the saliniferous formation: geophysical exploration, modeling of the internal structure of the salt dome Gorleben, fluids in the salt dome Gorleben. Geotechnical studies at the site: geothermal studies, hydraulic data on the saliniferous formation, thermal model calculations, geomechanical in-situ measurements, mechanical and transport properties, thermo-mechanical modeling. Summarized evaluation of the site characteristics and further research requirement.

  16. Multi-site Field Verification of Laboratory Derived FDOM Sensor Corrections: The Good, the Bad and the Ugly

    Science.gov (United States)

    Saraceno, J.; Shanley, J. B.; Aulenbach, B. T.

    2014-12-01

    Fluorescent dissolved organic matter (FDOM) is an excellent proxy for dissolved organic carbon (DOC) in natural waters. Through this relationship, in situ FDOM can be utilized to capture both high frequency time series and long term fluxes of DOC in small streams. However, in order to calculate accurate DOC fluxes for comparison across sites, in situ FDOM data must be compensated for matrix effects. Key matrix effects, include temperature, turbidity and the inner filter effect due to color. These interferences must be compensated for to develop a reasonable relationship between FDOM and DOC. In this study, we applied laboratory-derived correction factors to real time data from the five USGS WEBB headwater streams in order to gauge their effectiveness across a range of matrix effects. The good news is that laboratory derived correction factors improved the predicative relationship (higher r2) between DOC and FDOM when compared to uncorrected data. The relative importance of each matrix effect (i.e. temperature) varied by site and by time, implying that each and every matrix effect should be compensated for when available. In general, temperature effects were more important on longer time scales, while corrections for turbidity and DOC inner filter effects were most prevalent during hydrologic events, when the highest instantaneous flux of DOC occurred. Unfortunately, even when corrected for matrix effects, in situ FDOM is a weaker predictor of DOC than A254, a common surrogate for DOC, implying that either DOC fluoresces at varying degrees (but should average out over time), that some matrix effects (e.g. pH) are either unaccounted for or laboratory-derived correction factors do not encompass the site variability of particles and organics. The least impressive finding is that the inherent dependence on three variables in the FDOM correction algorithm increases the likelihood of record data gaps which increases the uncertainty in calculated DOC flux values.

  17. Remediation of cadmium-contaminated paddy soils by washing with calcium chloride: Verification of on-site washing

    International Nuclear Information System (INIS)

    Makino, Tomoyuki; Kamiya, Takashi; Takano, Hiroyuki; Itou, Tadashi; Sekiya, Naoki; Sasaki, Kouta; Maejima, Yuji; Sugahara, Kazuo

    2007-01-01

    We developed a new, three-step soil-wash method to remediate Cd-contaminated paddy fields. The method comprises (1) chemically washing the field soil with a CaCl 2 solution; (2) washing the treated soil with water to eliminate residual Cd and CaCl 2 ; and (3) on-site treatment of wastewater using a portable wastewater treatment system. Cd concentrations in the treated water were below Japan's environmental quality standard (0.01 mg Cd L -1 ), and the removal of Cd from the exchangeable fraction was 55% and from the acid-soluble fraction 15%. While soil fertility properties were affected by the soil washing, adverse effects were not crucial and could be corrected. The washing had no affect on rice growth, and reduced the average Cd concentration in rice grains by about two-thirds compared to a control plot. These results confirmed the effectiveness of the soil-wash method in remediating Cd-contaminated paddy fields. - In situ soil washing in a paddy field using an on-site wastewater treatment system resulted in an effective decrease of Cd in soil and rice grains without affecting rice yield

  18. Verification of the both hydrogeological and hydrogeochemical code results by an on-site test in granitic rocks

    Directory of Open Access Journals (Sweden)

    Michal Polák

    2007-01-01

    Full Text Available The project entitled “Methods and tools for the evaluation of the effect of engeneered barriers on distant interactions in the environment of a deep repository facility” deals with the ability to validate the behavior of applied engeneered barriers on hydrodynamic and migration parameters in the water-bearing granite environment of a radioactive waste deep repository facility. A part of the project represents a detailed mapping of the fracture network by means of geophysical and drilling surveys on the test-site (active granite quarry, construction of model objects (about 100 samples with the shape of cylinders, ridges and blocks, and the mineralogical, petrological and geochemical description of granite. All the model objects were subjected to migration and hydrodynamic tests with the use of fluorescein and NaCl as tracers. The tests were performed on samples with simple fractures, injected fractures and with an undisturbed integrity (verified by ultrasonic. The gained hydrodynamic and migration parameters of the model objects were processed with the modeling software NAPSAC and FEFLOW. During the following two years, these results and parameters will be verified (on the test-site by means of a long-term field test including the tuning of the software functionality.

  19. Packaging fluency

    DEFF Research Database (Denmark)

    Mocanu, Ana; Chrysochou, Polymeros; Bogomolova, Svetlana

    2011-01-01

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

  20. Data Validation Package, December 2015, Groundwater and Surface Water Sampling at the Monument Valley, Arizona, Processing Site March 2016

    Energy Technology Data Exchange (ETDEWEB)

    Tyrrell, Evan [Navarro Research and Engineering, Inc., Oak Ridge, NV (United States); Denny, Angelita [USDOE Office of Legacy Management, Washington, DC (United States)

    2016-03-23

    Fifty-two groundwater samples and one surface water sample were collected at the Monument Valley, Arizona, Processing Site to monitor groundwater contaminants for evaluating the effectiveness of the proposed compliance strategy as specified in the 1999 Final Site Observational Work Plan for the UMTRA Project Site at Monument Valley, Arizona. Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated, http://energy.gov/lm/downloads/sampling-and-analysis-plan-us-department- energy-office-legacy-management-sites). Samples were collected for metals, anions, nitrate + nitrite as N, and ammonia as N analyses at all locations.

  1. NRF TRIGA packaging

    International Nuclear Information System (INIS)

    Clements, M.D.

    1995-11-01

    Training Reactor Isotopes, General Atomics (TRIGA reg-sign) Reactors are in use at four US Department of Energy (DOE) complex facilities and at least 23 university, commercial, or government facilities. The development of the Neutron Radiography Facility (NRF) TRIGA packaging system began in October 1993. The Hanford Site NRF is being shut down and requires an operationally user-friendly transportation and storage packaging system for removal of the TRIGA fuel elements. The NRF TRIGA packaging system is designed to remotely remove the fuel from the reactor and transport the fuel to interim storage (up to 50 years) on the Hanford Site. The packaging system consists of a cask and an overpack. The overpack is used only for transport and is not necessary for storage. Based upon the cask's small size and light weight, small TRIGA reactors will find it versatile for numerous refueling and fuel storage needs. The NRF TRIGA packaging design also provides the basis for developing a certifiable and economical packaging system for other TRIGA reactor facilities. The small size of the NRF TRIGA cask also accommodates placing the cask into a larger certified packaging for offsite transport. The Westinghouse Hanford Company NRF TRIGA packaging, as described herein can serve other DOE sites for their onsite use, and the design can be adapted to serve university reactor facilities, handling a variety of fuel payloads

  2. Microelectronic packaging

    CERN Document Server

    Datta, M; Schultze, J Walter

    2004-01-01

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

  3. Data Validation Package October 2016 Groundwater and Surface Water Sampling at the Monticello, Utah, Disposal and Processing Sites January 2017

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Jason [USDOE Office of Legacy Management (LM), Washington, DC (United States); Smith, Fred [Navarro Research and Engineering, Inc., Grand Junction, CO (United States)

    2017-02-01

    Sampling Period: October 10–12, 2016. This semiannual event includes sampling groundwater and surface water at the Monticello Disposal and Processing Sites. Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated) and Program Directive MNT-2016-01. Samples were collected from 54 of 64 planned locations (16 of 17 former mill site wells, 15 of 18 downgradient wells, 7 of 9 downgradient permeable reactive barrier wells, 3 of 3 bedrock wells, 4 of 7 seeps and wetlands, and 9 of 10 surface water locations).

  4. TU-FG-201-03: Automatic Pre-Delivery Verification Using Statistical Analysis of Consistencies in Treatment Plan Parameters by the Treatment Site and Modality

    International Nuclear Information System (INIS)

    Liu, S; Wu, Y; Chang, X; Li, H; Yang, D

    2016-01-01

    Purpose: A novel computer software system, namely APDV (Automatic Pre-Delivery Verification), has been developed for verifying patient treatment plan parameters right prior to treatment deliveries in order to automatically detect and prevent catastrophic errors. Methods: APDV is designed to continuously monitor new DICOM plan files on the TMS computer at the treatment console. When new plans to be delivered are detected, APDV checks the consistencies of plan parameters and high-level plan statistics using underlying rules and statistical properties based on given treatment site, technique and modality. These rules were quantitatively derived by retrospectively analyzing all the EBRT treatment plans of the past 8 years at authors’ institution. Therapists and physicists will be notified with a warning message displayed on the TMS computer if any critical errors are detected, and check results, confirmation, together with dismissal actions will be saved into database for further review. Results: APDV was implemented as a stand-alone program using C# to ensure required real time performance. Mean values and standard deviations were quantitatively derived for various plan parameters including MLC usage, MU/cGy radio, beam SSD, beam weighting, and the beam gantry angles (only for lateral targets) per treatment site, technique and modality. 2D-based rules of combined MU/cGy ratio and averaged SSD values were also derived using joint probabilities of confidence error ellipses. The statistics of these major treatment plan parameters quantitatively evaluate the consistency of any treatment plans which facilitates automatic APDV checking procedures. Conclusion: APDV could be useful in detecting and preventing catastrophic errors immediately before treatment deliveries. Future plan including automatic patient identify and patient setup checks after patient daily images are acquired by the machine and become available on the TMS computer. This project is supported by the

  5. TU-FG-201-03: Automatic Pre-Delivery Verification Using Statistical Analysis of Consistencies in Treatment Plan Parameters by the Treatment Site and Modality

    Energy Technology Data Exchange (ETDEWEB)

    Liu, S; Wu, Y; Chang, X; Li, H; Yang, D [Washington University School of Medicine, St. Louis, MO (United States)

    2016-06-15

    Purpose: A novel computer software system, namely APDV (Automatic Pre-Delivery Verification), has been developed for verifying patient treatment plan parameters right prior to treatment deliveries in order to automatically detect and prevent catastrophic errors. Methods: APDV is designed to continuously monitor new DICOM plan files on the TMS computer at the treatment console. When new plans to be delivered are detected, APDV checks the consistencies of plan parameters and high-level plan statistics using underlying rules and statistical properties based on given treatment site, technique and modality. These rules were quantitatively derived by retrospectively analyzing all the EBRT treatment plans of the past 8 years at authors’ institution. Therapists and physicists will be notified with a warning message displayed on the TMS computer if any critical errors are detected, and check results, confirmation, together with dismissal actions will be saved into database for further review. Results: APDV was implemented as a stand-alone program using C# to ensure required real time performance. Mean values and standard deviations were quantitatively derived for various plan parameters including MLC usage, MU/cGy radio, beam SSD, beam weighting, and the beam gantry angles (only for lateral targets) per treatment site, technique and modality. 2D-based rules of combined MU/cGy ratio and averaged SSD values were also derived using joint probabilities of confidence error ellipses. The statistics of these major treatment plan parameters quantitatively evaluate the consistency of any treatment plans which facilitates automatic APDV checking procedures. Conclusion: APDV could be useful in detecting and preventing catastrophic errors immediately before treatment deliveries. Future plan including automatic patient identify and patient setup checks after patient daily images are acquired by the machine and become available on the TMS computer. This project is supported by the

  6. The development of an on-site container

    International Nuclear Information System (INIS)

    Glass, R.E.; McAllaster, M.E.; Jones, P.L.; McKinney, A.L.

    1992-01-01

    Sandia National Laboratories (SNL) has developed a package for the on-site transport of chemical munitions for the US Army. This package was designed to prevent the release of lethal quantities of chemical agents during transportation of munitions to the demilitarization facilities on-site. The packaging prevents auto-ignition of the munitions by limiting the thermal and structural assault on the munitions during an accident. This package, with some modifications to account for contents, may be suitable for the on-site transport of mixed wastes at United States Department of Energy facilities. This paper discusses the design and verification testing of the package. The safety criteria for the package were modeled after the International Atomic Energy Agency (IAEA) hypothetical accident sequence and modified to take credit for operational controls. The modified accident sequence consisted of drop, puncture, and thermal events. The post-accident leak rate was established to prevent harm to an exposed worker. The packaging has a mass of 8600 kg and can accommodate up to 3600 kg of contents. The interior of the package is 188 cm in diameter and 232 cm long. Two sample ports can be used to sample the interior of the package prior to opening the closure and an o-ring test port can be used to determine the leak rates prior to and after transport

  7. Standard Guide for On-Site Inspection and Verification of Operation of Solar Domestic Hot Water Systems

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1987-01-01

    1.1 This guide covers procedures and test methods for conducting an on-site inspection and acceptance test of an installed domestic hot water system (DHW) using flat plate, concentrating-type collectors or tank absorber systems. 1.2 It is intended as a simple and economical acceptance test to be performed by the system installer or an independent tester to verify that critical components of the system are functioning and to acquire baseline data reflecting overall short term system heat output. 1.3 This guide is not intended to generate accurate measurements of system performance (see ASHRAE standard 95-1981 for a laboratory test) or thermal efficiency. 1.4 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only. 1.5 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine th...

  8. Data Validation Package - April and July 2015 Groundwater and Surface Water Sampling at the Gunnison, Colorado, Processing Site

    Energy Technology Data Exchange (ETDEWEB)

    Linard, Joshua [Dept. of Energy (DOE), Washington, DC (United States). Office of Legacy Management; Campbell, Sam [Navarro Research and Engineering, Inc., Oak Ridge, TN (United States)

    2016-02-01

    This event included annual sampling of groundwater and surface water locations at the Gunnison, Colorado, Processing Site. Sampling and analyses were conducted as specified in Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites. Samples were collected from 28 monitoring wells, three domestic wells, and six surface locations in April at the processing site as specified in the 2010 Ground Water Compliance Action Plan for the Gunnison, Colorado, Processing Site. Domestic wells 0476 and 0477 were sampled in July because the homes were unoccupied in April, and the wells were not in use. Duplicate samples were collected from locations 0113, 0248, and 0477. One equipment blank was collected during this sampling event. Water levels were measured at all monitoring wells that were sampled. No issues were identified during the data validation process that requires additional action or follow-up.

  9. Data Validation Package February 2016 Groundwater and Surface Water Sampling at the Tuba City, Arizona, Disposal Site April 2016

    Energy Technology Data Exchange (ETDEWEB)

    Bush, Richard [USDOE Office of Legacy Management, Washington, DC (United States); Lemke, Peter [Navarro Research and Engineering, Inc., Oak Ridge, TN (United States)

    2016-04-01

    The groundwater compliance strategy for the Tuba City, Arizona, Disposal Site is defined in the 1999 Phase I Ground Water Compliance Action Plan for the Tuba City, Arizona, UMTRA Site. Samples are collected and analyzed on a semiannual basis to evaluate the performance of the Phase I remediation system. Sampling and analyses were conducted as specified in Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). U.S. Environmental Protection Agency (EPA) groundwater standards were exceeded in samples collected from monitoring wells as listed in Table 1. The data from this sampling event are generally consistent with previously obtained values and are acceptable for general use as qualified. Data anomalies are not significant with respect to the known nature and extent of contamination and progress of remedial action at the site. The data from this sampling event will be incorporated into the annual performance evaluation report that will present a comprehensive hydrologic summary and evaluation of groundwater remedial action performance at the Tuba City site through March 2016.

  10. Data Validation Package - June 2016 Groundwater and Surface Water Sampling at the Green River, Utah, Disposal Site

    Energy Technology Data Exchange (ETDEWEB)

    Linard, Joshua [USDOE Office of Legacy Management, Washington, DC (United States); Price, Jeffrey [Navarro Research and Engineering, Inc., Las Vegas, NV (United States)

    2016-10-10

    This event included annual sampling of groundwater and surface water locations at the Green River, Utah, Disposal Site. Sampling and analyses were conducted as specified in Sampling and Analysis Plan for US. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated, http://energy.gov/lrnldownloads/sampling-and- analysis-plan-us-department-energy-office-legacy-management-sites). Samples were collected from 15 monitoring wells and two surface locations at the disposal site as specified in the draft 2011 Ground Water Compliance Action Plan for the Green River, Utah, Disposal Site. Planned monitoring locations are shown in Attachment 1, Sampling and Analysis Work Order. A duplicate sample was collected from location 0179. One equipment blank was collected during this sampling event. Water levels were measured at all monitoring wells that were sampled. See Attachment 2, Trip Reports for additional details. The analytical data and associated qualifiers can be viewed in environmental database reports and are also available for viewing with dynamic mapping via the GEMS (Geospatial Environmental Mapping System) website at http://gems.lm.doe.gov/#. No issues were identified during the data validation process that requires additional action or follow-up.

  11. Data Validation Package August 2015 Groundwater and Surface Water Sampling at the Tuba City, Arizona, Disposal Site November 2015

    Energy Technology Data Exchange (ETDEWEB)

    Bush, Richard [USDOE Office of Legacy Management, Washington, DC (United States); Lemke, Peter [Navarro Research and Engineering, Inc., Oak Ridge, TN (United States)

    2015-11-01

    The groundwater compliance strategy for the Tuba City, Arizona, Disposal Site is defined in the 1999 Phase I Ground Water Compliance Action Plan for the Tuba City, Arizona, UMTRA Site. Samples are collected and analyzed on a semiannual basis to evaluate the performance of the Phase I remediation system. Sampling and analyses were conducted as specified in Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). U.S. Environmental Protection Agency (EPA) groundwater standards were exceeded in samples collected from monitoring wells and extraction wells as listed in Table 1. The data from this sampling event are generally consistent with previously obtained values and are acceptable for general use as qualified. Data anomalies are not significant with respect to the known nature and extent of contamination and progress of remedial action at the site. The data from this sampling event will be incorporated into the annual performance evaluation report that will present a comprehensive hydrologic summary and evaluation of groundwater remedial action performance at the Tuba City site through March 2016.

  12. Acceptance and tracking of waste packages from nuclear power plants at the Centre de l'Aube

    International Nuclear Information System (INIS)

    Errera, J.; Tison, J.L.

    2001-01-01

    For 30 years, the French National Agency for Radioactive Waste Management (ANDRA) is in charge of the radioactive waste management and acquired a good knowledge relating to the control of low and intermediate level waste produced by nuclear power plants (NPP), the waste characteristics and the waste conditioning. The integrated waste management system for low-level radioactive waste in France implemented by ANDRA covers all stages from waste generation to final disposal at the Centre de I'Aube near surface facility. ANDRA defined a quality assurance program for waste management that specifies the level of quality to be achieved by solidification and packaging processes, defines quality control requirements and defines waste tracking requirements, from waste generation through final disposal. Verification of quality of waste packages is implemented at three levels of the waste management system. The first one consists of inspections of waste packages at the generator's premises and audits of the quality assurance organization of the waste generator. The second level of verification consists of the waste tracking system. It allows identifying and tracking each waste package from the step it is fabricated to its final disposal at the ANDRA site. The third level of verification is obtained by mean of non-destructive and destructive assays of waste packages. These assays allow to verify generator compliance with ANDRA's technical specifications and to investigate the accuracy of physical and radioactive characteristics reported to ANDRA by the generator. (author)

  13. Plutonium stabilization and packaging system

    International Nuclear Information System (INIS)

    1996-01-01

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

  14. Data Validation Package, July 2016 Groundwater Sampling at the Shirley Basin South, Wyoming, Disposal Site November 2016

    Energy Technology Data Exchange (ETDEWEB)

    Frazier, William [USDOE Office of Legacy Management, Washington, DC (United States); Price, Jeffrey [Navarro Research and Engineering, Inc., Oak Ridge, TN (United States)

    2016-11-01

    Sampling Period: July 14-15, 2016 The 2004 Long-Term Surveillance Plan for the Shirley Basin South (UMTRCA Title II) Disposal Site, Carbon County, Wyoming, requires annual monitoring to verify continued compliance with the pertinent alternate concentration limits (ACLs) and Wyoming Class III (livestock use) groundwater protection standards. Planned monitoring locations are shown in Attachment 1, Sampling and Analysis Work Order. Point-of-compliance (POC) wells 19-DC, 5-DC, and 5-SC, and monitoring wells 10-DC, 110-DC, 112-DC, 113-DC, 40-SC, 54-SC, 100-SC, 102-SC, and K.G.S.#3 were sampled. POC well 51-SC and downgradient well 101-SC were dry at the time of sampling. The water level was measured at each sampled well. See Attachment 2, Trip Report for additional details. Sampling and analyses were conducted in accordance with the Sampling and Analysis Plan for the U S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated, http://energy.gov/lm/downloads/sampling-and­ analysis-plan-us-department-energy-office-legacy-management-sites). ACLs are approved for cadmium, chromium, lead, nickel, radium-226, radium-228, selenium, thorium-230, and uranium in site groundwater. Time-concentration graphs of the contaminants of concern in POC wells are included in Attachment 3, Data Presentation. The only ACL exceedance in a POC well was radium-228 in well 5-DC where the concentration was 30.7 picocuries per liter (pCi/L), exceeding the ACL of 25.7 pCi/L. Concentrations of sulfate and total dissolved solids continue to exceed their respective Wyoming Class III groundwater protection standards for livestock use in wells 5-DC, 5-SC, and 54-SC as they have done throughout the sampling history; however, there is no livestock use of the water from these aquifers at the site, and no constituent concentrations exceed groundwater protection standards at the wells near the site boundary.

  15. Survey package: Technical and contracting strategies for single-shell tank waste retrieval on the Hanford Site

    International Nuclear Information System (INIS)

    Ramsower, D.C.

    1995-01-01

    Westinghouse Hanford Company is interested in innovative, commercially available or adaptable retrieval system equipment, concepts, and contracting strategies that will ad to existing Hanford Site technology and significantly reduce cost and/or risk from the baseline retrieval approach of sluicing (hydraulically mining) the waste from the SSTs onsite. The objective of this request is to gather information from industry to identify and summarize a suite of retrieval-related components, systems, and contracting approaches. This information will be used to ensure that WHC understands the various waste retrieval alternative approaches, their risks, and their application on the Hanford Site tanks for those occasions when sluicing is not sufficiently effective, appropriate, or cost-effective. An additional objective is to facilitate industry's understanding of the tank and site interface requirements for SST waste retrieval and the complex statutory, legal, regulatory, labor, and other institutional standards being applied to the Hanford Site. This effort will identify and summarize retrieval solutions by the end of September 1996 so that a clear basis for future retrieval program decisions can be established

  16. Data Validation Package October 2015 Groundwater and Surface Water Sampling at the Monticello, Utah, Processing Site January 2016

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Jason [U.S. Dept. of Energy, Washington, DC (United States). Office of Legacy Management; Smith, Fred [Navarro Research and Engineering, Inc., Oak Ridge, TN (United States)

    2016-01-21

    Sampling Period: October 12–14, 2015. This semiannual event includes sampling groundwater and surface water at the Monticello Mill Tailings Site. Sampling and analyses were conducted as specified in the 2004 Monticello Mill Tailings Site Operable Unit III Post-Record of Decision Monitoring Plan, Draft Final and Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). Samples were collected from 52 of 61 planned locations (15 of 17 former mill site wells, 17 of 18 downgradient wells, 9 of 9 downgradient permeable reactive barrier wells, 2 of 7 seeps and wetlands, and 9 of 10 surface water locations). Locations MW00-07, Seep 1, Seep 2, Seep 3, Seep 5, Seep 6, SW00-01, T01-13, and T01-19 were not sampled because of insufficient water availability. All samples were filtered as specified in the monitoring plan. Duplicate samples were collected from surface water location W3-04 and from monitoring wells 82-08, 92-09, and 92-10. Water levels were measured at all but one sampled well and an additional set of wells. The contaminants of concern (COCs) for the Monticello Mill Tailings Site are arsenic, manganese, molybdenum, nitrate + nitrite as nitrogen (nitrate + nitrite as N), selenium, uranium, and vanadium. Time-concentration graphs of the COCs for all groundwater and surface water locations are included in this report. Locations with COCs that exceeded remediation goals are listed.

  17. Radiological consequence analysis. Report on the work package 10. Preliminary safety analysis for the site Gorleben; Radiologische Konsequenzanalyse. Bericht zum Arbeitspaket 10. Vorlaeufige Sicherheitsanalyse fuer den Standort Gorleben

    Energy Technology Data Exchange (ETDEWEB)

    Larue, Juergen; Baltes, Bruno; Fischer, Heidemarie; Frieling, Gerd; Kock, Ingo; Navarro, Martin; Seher, Holger

    2013-02-15

    Work package 10 of the preliminary safety analysis for the site Gorleben concerns the analysis of release scenarios and the radiological consequences. The report includes the description of the applied methodology, the concept for the mathematical description of scenarios, the variety of scenarios for the radiological consequence analysis with the definition of the reference scenario. The database for the calculation covers geology and hydrology, the final repository concept and the process of the final disposal system. The used models were the one-phase transport model MARINE and the two-phase transport model TOUGH2. The report summarizes the results including an evaluation of the realization of the demonstration concept with respect to solution transport, fluid transport, robustness of the modeling results and analogy discussion to other disposal variants (i.e. borehole disposal and storage of transport or storage casks).

  18. MEMS packaging

    CERN Document Server

    Hsu , Tai-Ran

    2004-01-01

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

  19. Safety assessment document for spent fuel handling, packaging, and storage demonstrations at the E-MAD facility on the Nevada Test Site

    International Nuclear Information System (INIS)

    1985-04-01

    The objectives for spent fuel handling and packaging demonstration are to develop the capability to satisfactorily encapsulate typical commercial nuclear reactor spent fuel assemblies and to establish the suitability of interim dry surface and near surface storage concepts. To accomplish these objectives, spent fuel assemblies from a pressurized water reactor have been received, encapsulated in steel canisters, and emplaced in on-site storage facilities and subjected to other tests. As an essential element of these demonstrations, a thorough safety assessment of the demonstration activities conducted at the E-MAD facility has been completed. This document describes the site location and characteristics, the existing E-MAD facility, and the facility modifications and equipment additions made specifically for the demonstrations. The document also summarizes the Quality Assurance Program utilized, and specifies the principal design criteria applicable to the facility modifications, equipment additions, and process operations. Evaluations have been made of the radiological impacts of normal operations, abnormal operations, and postulated accidents. Analyses have been performed to determine the affects on nuclear criticality safety of postulated accidents and credible natural phenomena. The consequences of postulated accidents resulting in fission product gas release have also been estimated. This document identifies the engineered safety features, procedures, and site characteristics that (1) prevent the occurrence of potential accidents or (2) assure that the consequences of postulated accidents are either insignificant or adequately mitigated

  20. Data Validation Package April 2016 Groundwater and Surface Water Sampling at the Monticello, Utah, Disposal and Processing Sites August 2016

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Jason [USDOE Office of Legacy Management, Washington, DC (United States); Smith, Fred [Navarro Research and Engineering, Oak Ridge, TN (United States)

    2016-08-01

    This semiannual event includes sampling groundwater and surface water at the Monticello Disposal and Processing Sites. Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated) and Program Directive MNT-2016-01. Complete sample sets were collected from 42 of 48 planned locations (9 of 9 former mill site wells, 13 of 13 downgradient wells, 7 of 9 downgradient permeable reactive barrier wells, 4 of 7 seeps and wetlands, and 9 of 10 surface water locations). Planned monitoring locations are shown in Attachment 1, Sampling and Analysis Work Order. Locations R6-M3, SW00-01, Seep 1, Seep 2, and Seep 5 were not sampled due to insufficient water availability. A partial sample was collected at location R4-M3 due to insufficient water. All samples from the permeable reactive barrier wells were filtered as specified in the program directive. Duplicate samples were collected from surface water location Sorenson and from monitoring wells 92-07 and RlO-Ml. Water levels were measured at all sampled wells and an additional set of wells. See Attachment2, Trip Report for additional details. The contaminants of concern (COCs) for the Monticello sites are arsenic, manganese, molybdenum, nitrate+ nitrite as nitrogen (nitrate+ nitrite as N), selenium, uranium, and vanadium. Locations with COCs that exceeded remediation goals are listed in Table 1 and Table 2. Time-concentration graphs of the COCs for all groundwater and surface water locations are included in Attachment 3, Data Presentation. An assessment of anomalous data is included in Attachment 4.

  1. Data Validation Package - June 2015 Groundwater and Surface Water Sampling at the Green River, Utah, Disposal Site

    Energy Technology Data Exchange (ETDEWEB)

    Linard, Joshua [USDOE Office of Legacy Management, Washington, DC (United States); Price, Jeffrey [Navarro Research and Engineering, Inc., Las Vegas, NV (United States)

    2015-08-01

    Groundwater samples were collected during the 2015 sampling event from point-of-compliance (POC) wells 0171, 0173, 0176, 0179, 0181, and 0813 to monitor the disposition of contaminants in the middle sandstone unit of the Cedar Mountain Formation. Groundwater samples also were collected from alluvium monitoring wells 0188, 0189, 0192, 0194, and 0707, and basal sandstone monitoring wells 0182, 0184, 0185, and 0588 as a best management practice. Surface locations 0846 and 0847 were sampled to monitor for degradation of water quality in the backwater area of Brown’s Wash and in the Green River immediately downstream of Brown’s Wash. The Green River location 0801 is upstream from the site and is sampled to determine background-threshold values (BTVs). Sampling and analyses were conducted as specified in Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated, http://energy.gov/lm/downloads/sampling-and- analysis-plan-us-department-energy-office-legacy-management-sites). Water levels were measured at each sampled well. The analytical data and associated qualifiers can be viewed in environmental database reports and are also available for viewing with dynamic mapping via the GEMS (Geospatial Environmental Mapping System) website at http://gems.lm.doe.gov/#. All six POC wells are completed in the middle sandstone unit of the Cedar Mountain Formation and are monitored to measure contaminant concentrations for comparison to proposed alternate concentration limits (ACLs), as provided in Table 1. Contaminant concentrations in the POC wells remain below their respective ACLs.

  2. Data Validation Package December 2015 Groundwater Sampling at the Ambrosia Lake, New Mexico, Disposal Site March 2016

    Energy Technology Data Exchange (ETDEWEB)

    Tsosie, Bernadette [USDOE Office of Legacy Management, Washington, DC (United States); Johnson, Dick [Navarro Research and Engineering, Inc., Oak Ridge, TN (United States)

    2016-03-01

    The Long-Term Surveillance Plan for the Ambrosia Lake, New Mexico, Disposal Site does not require groundwater monitoring because groundwater in the uppermost aquifer is of limited use, and supplemental standards have been applied to the aquifer. However, at the request of the New Mexico Environment Department, the U.S. Department of Energy conducts annual monitoring at three locations: monitoring wells 0409, 0675, and 0678. Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for US. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). Monitoring Well 0409 was not sampled during this event because it was dry. Water levels were measured at each sampled well. One duplicate sample was collected from location 0675. Groundwater samples from the two sampled wells were analyzed for the constituents listed in Table 1. Time-concentration graphs for selected analytes are included in this report. At well 0675, the duplicate results for total dissolved solids and for most metals (magnesium, molybdenum, potassium, selenium, sodium, and uranium) were outside acceptance criteria, which may indicate non-homogeneous conditions at this location. November 2014 results for molybdenum and uranium at well 0675 also were outside acceptance criteria. The well condition will be evaluated prior to the next sampling event.

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

    International Nuclear Information System (INIS)

    SWENSON, C.E.

    2000-01-01

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

  4. Data Packages for the Hanford Immobilized Low-Activity Tank Waste Performance Assessment: 2001 Version

    International Nuclear Information System (INIS)

    MANN, F.M.

    2000-01-01

    Data package supporting the 2001 Immobilized Low-Activity Waste Performance Analysis. Geology, hydrology, geochemistry, facility, waste form, and dosimetry data based on recent investigation are provided. Verification and benchmarking packages for selected software codes are provided

  5. On-site treatment of a motorway service area wastewater using a package sequencing batch reactor (SBR).

    Science.gov (United States)

    Del Solar, J; Hudson, S; Stephenson, T

    2005-01-01

    A sequencing batch reactor (SBR) treating the effluent of a motorway service station in the south of England situated on a major tourist route was investigated. Wastewater from the kitchens, toilets and washrooms facilities was collected from the areas on each side of the motorway for treatment on-site. The SBR was designed for a population equivalent (p.e.) of 500, assuming an average flow of 100 m3/d, influent biochemical oxygen demand (BOD) of 300 mg/l, and influent suspended solids (SS) of 300 mg/l. Influent monitoring over 8 weeks revealed that the average flow was only 65 m3/d and the average influent BOD and SS were 480 mg/l and 473 mg/l respectively. This corresponded to a high sludge loading rate (F:M) of 0.42 d(-1) which accounted for poor performance. Therefore the cycle times were extended from 6 h to 7 h and effluent BOD improved from 79 to 27 mg/l.

  6. Packaging microservices

    DEFF Research Database (Denmark)

    Montesi, Fabrizio; Thrane, Dan Sebastian

    2017-01-01

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

  7. Science packages

    Science.gov (United States)

    1997-01-01

    Primary science teachers in Scotland have a new updating method at their disposal with the launch of a package of CDi (Compact Discs Interactive) materials developed by the BBC and the Scottish Office. These were a response to the claim that many primary teachers felt they had been inadequately trained in science and lacked the confidence to teach it properly. Consequently they felt the need for more in-service training to equip them with the personal understanding required. The pack contains five disks and a printed user's guide divided up as follows: disk 1 Investigations; disk 2 Developing understanding; disks 3,4,5 Primary Science staff development videos. It was produced by the Scottish Interactive Technology Centre (Moray House Institute) and is available from BBC Education at £149.99 including VAT. Free Internet distribution of science education materials has also begun as part of the Global Schoolhouse (GSH) scheme. The US National Science Teachers' Association (NSTA) and Microsoft Corporation are making available field-tested comprehensive curriculum material including 'Micro-units' on more than 80 topics in biology, chemistry, earth and space science and physics. The latter are the work of the Scope, Sequence and Coordination of High School Science project, which can be found at http://www.gsh.org/NSTA_SSandC/. More information on NSTA can be obtained from its Web site at http://www.nsta.org.

  8. Material integrity verification radar

    International Nuclear Information System (INIS)

    Koppenjan, S.K.

    1999-01-01

    The International Atomic Energy Agency (IAEA) has the need for verification of 'as-built' spent fuel-dry storage containers and other concrete structures. The IAEA has tasked the Special Technologies Laboratory (STL) to fabricate, test, and deploy a stepped-frequency Material Integrity Verification Radar (MIVR) system to nondestructively verify the internal construction of these containers. The MIVR system is based on previously deployed high-frequency, ground penetrating radar (GPR) systems that have been developed by STL for the U.S. Department of Energy (DOE). Whereas GPR technology utilizes microwave radio frequency energy to create subsurface images, MTVR is a variation for which the medium is concrete instead of soil. The purpose is to nondestructively verify the placement of concrete-reinforcing materials, pipes, inner liners, and other attributes of the internal construction. The MIVR system underwent an initial field test on CANDU reactor spent fuel storage canisters at Atomic Energy of Canada Limited (AECL), Chalk River Laboratories, Ontario, Canada, in October 1995. A second field test at the Embalse Nuclear Power Plant in Embalse, Argentina, was completed in May 1996. The DOE GPR also was demonstrated at the site. Data collection and analysis were performed for the Argentine National Board of Nuclear Regulation (ENREN). IAEA and the Brazilian-Argentine Agency for the Control and Accounting of Nuclear Material (ABACC) personnel were present as observers during the test. Reinforcing materials were evident in the color, two-dimensional images produced by the MIVR system. A continuous pattern of reinforcing bars was evident and accurate estimates on the spacing, depth, and size were made. The potential uses for safeguard applications were jointly discussed. The MIVR system, as successfully demonstrated in the two field tests, can be used as a design verification tool for IAEA safeguards. A deployment of MIVR for Design Information Questionnaire (DIQ

  9. Swarm Verification

    Science.gov (United States)

    Holzmann, Gerard J.; Joshi, Rajeev; Groce, Alex

    2008-01-01

    Reportedly, supercomputer designer Seymour Cray once said that he would sooner use two strong oxen to plow a field than a thousand chickens. Although this is undoubtedly wise when it comes to plowing a field, it is not so clear for other types of tasks. Model checking problems are of the proverbial "search the needle in a haystack" type. Such problems can often be parallelized easily. Alas, none of the usual divide and conquer methods can be used to parallelize the working of a model checker. Given that it has become easier than ever to gain access to large numbers of computers to perform even routine tasks it is becoming more and more attractive to find alternate ways to use these resources to speed up model checking tasks. This paper describes one such method, called swarm verification.

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

    Science.gov (United States)

    Beaudin, Bart; And Others

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

  11. Challenges for effective WMD verification

    International Nuclear Information System (INIS)

    Andemicael, B.

    2006-01-01

    Effective verification is crucial to the fulfillment of the objectives of any disarmament treaty, not least as regards the proliferation of weapons of mass destruction (WMD). The effectiveness of the verification package depends on a number of factors, some inherent in the agreed structure and others related to the type of responses demanded by emerging challenges. The verification systems of three global agencies-the IAEA, the Comprehensive Nuclear-Test-Ban Treaty Organization (CTBTO, currently the Preparatory Commission), and the Organization for the Prohibition of Chemical Weapons (OPCW)-share similarities in their broad objectives of confidence-building and deterrence by assuring members that rigorous verification would deter or otherwise detect non-compliance. Yet they are up against various constraints and other issues, both internal and external to the treaty regime. These constraints pose major challenges to the effectiveness and reliability of the verification operations. In the nuclear field, the IAEA safeguards process was the first to evolve incrementally from modest Statute beginnings to a robust verification system under the global Treaty on the Non-Proliferation of Nuclear Weapons (NPT). The nuclear non-proliferation regime is now being supplemented by a technology-intensive verification system of the nuclear test-ban treaty (CTBT), a product of over three decades of negotiation. However, there still remain fundamental gaps and loopholes in the regime as a whole, which tend to diminish the combined effectiveness of the IAEA and the CTBT verification capabilities. He three major problems are (a) the lack of universality of membership, essentially because of the absence of three nuclear weapon-capable States-India, Pakistan and Israel-from both the NPT and the CTBT, (b) the changes in US disarmament policy, especially in the nuclear field, and (c) the failure of the Conference on Disarmament to conclude a fissile material cut-off treaty. The world is

  12. Leak testing United States Nuclear Regulatory Commission type b packaging

    International Nuclear Information System (INIS)

    Lacy, K.A.

    1995-01-01

    The Waste Isolation Pilot Plant (WTPP) is a one of its kind research and development facility operated by the Department of Energy, Carlsbad Area Office. Located in southeastern New Mexico, the WTPP is designed to demonstrate the safe, permanent disposal of transuranic (TRU) radioactive nuclear waste, accumulated from 40 years of nuclear weapons production. Before the waste can be disposed of, it must first be safely transported from generator storage sites to the WIPP. To accomplish this, the TRUPACT-II was designed and fabricated. This double containment, non-vented waste packaging successfully completed a rigorous testing program, and in 1989 received a Certificate of Compliance (C of C) from the Nuclear Regulatory Commission (NRC). Currently, the TRUPACT-II is in use at Idaho National Engineering Laboratory to transport waste on site for characterization. The DOE/CAO is responsible for maintaining the TRUPACT-II C of C. The C of C requires performance of nondestructive examination (NDE), e.g., visual testing (VT), dimensional inspections, Liquid Dye Penetrant testing (PT), and Helium Leak Detection (HLD). The Waste Isolation Division (WID) uses HLD for verification of the containment integrity. The following HLD tests are performed on annual basis or when required, i.e. repairs or component replacement: (1) fabrication verification leak tests on both the outer containment vessel (OCV) and the inner containment vessel (ICV); (2) assembly verification leak tests on the OCV and ICV upper main o-rings; and (3) assembly verification leak tests on the OCV and the ICV vent port plugs. These tests are addressed in detail as part of this presentation

  13. Independent Verification Survey Report for the Offsite Portion of the Potential Release Site-7 Abandoned Sanitary Line, Miamisburg Closure Project, Miamisburg, Ohio

    International Nuclear Information System (INIS)

    Weaver, P.C.

    2008-01-01

    The ORISE objective was to confirm that the remedial action process implemented by the contractor was in accordance with the PRS-7 Work Package. Following removal of the sanitary line, the soil beneath the line would be sampled to determine if remediation was required (ARC 2007a).

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

    Energy Technology Data Exchange (ETDEWEB)

    Romano, T.

    1997-09-25

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

  15. Data Packages for the Hanford Immobilized Low Activity Tank Waste Performance Assessment 2001 Version [SEC 1 THRU 5

    Energy Technology Data Exchange (ETDEWEB)

    MANN, F.M.

    2000-03-02

    Data package supporting the 2001 Immobilized Low-Activity Waste Performance Analysis. Geology, hydrology, geochemistry, facility, waste form, and dosimetry data based on recent investigation are provided. Verification and benchmarking packages for selected software codes are provided.

  16. Demonstration of impact performance of the nuclear transport package in on-site hypothetical collision scenarios by a heavy goods vehicle

    International Nuclear Information System (INIS)

    Tso, C.F.; Izatt, C.

    2004-01-01

    Spent fuel modules are contained in Module Removal Container (MRC) during on-site transport at the D154 facilities in the Devonport Naval Dockyard in the United Kingdom. The container is transported on its own on a Low Level Transfer Trolley (LLTT) and accommodated within a Transfer Frame. The LLTT travels on rails and moves either under its own power or towed by a Rail Tug Unit. The Transfer Frame provides a secure means of support to the MRC during transit and provides impact protection in the event of collision. The MRC is accommodated within the Transfer Frame by way of a sub-frame assembly. It rests on its sub-frame and is held in a vertical position by a number of support arms bolted to the Frame. The Transfer Frame is attached to the Low Level Transfer Trolley by a combination of bolts and shear pins. The combination of LLTT, Transfer Frame, sub-frame and a MRC is known as a Nuclear Transport Package (NTP). The design basis vehicle impact accident specifies a collision from a 20 tonne vehicle travelling at 20 mph from any direction. In order to satisfy the safety functional requirements, the NTP is required to meet the following conditions: The NTP should not overturn as a complete assembly following the impact. The Transfer Frame should not detach from the LLTT, and with the attachments remaining within the Level D stress limits specified in the ASME Boiler and Pressure Vessel Code Section 3. The MRC should be shown to withstand any potential impacts of the vehicle in the event of failure of any of the frame members. The frame must not transmit as a result of the vehicle impact, to either container, loads that would compromise their shielding and containment boundaries. The performance of the NTP was substantiated by finite element (FE) analysis, using the explicit non-linear transient code LS-DYNA. The work formed part of the site license application for the D154 facilities

  17. Disarmament Verification - the OPCW Experience

    International Nuclear Information System (INIS)

    Lodding, J.

    2010-01-01

    The Chemical Weapons Convention is the only multilateral treaty that bans completely an entire category of weapons of mass destruction under international verification arrangements. Possessor States, i.e. those that have chemical weapons stockpiles at the time of becoming party to the CWC, commit to destroying these. All States undertake never to acquire chemical weapons and not to help other States acquire such weapons. The CWC foresees time-bound chemical disarmament. The deadlines for destruction for early entrants to the CWC are provided in the treaty. For late entrants, the Conference of States Parties intervenes to set destruction deadlines. One of the unique features of the CWC is thus the regime for verifying destruction of chemical weapons. But how can you design a system for verification at military sites, while protecting military restricted information? What degree of assurance is considered sufficient in such circumstances? How do you divide the verification costs? How do you deal with production capability and initial declarations of existing stockpiles? The founders of the CWC had to address these and other challenges in designing the treaty. Further refinement of the verification system has followed since the treaty opened for signature in 1993 and since inspection work was initiated following entry-into-force of the treaty in 1997. Most of this work concerns destruction at the two large possessor States, Russia and the United States. Perhaps some of the lessons learned from the OPCW experience may be instructive in a future verification regime for nuclear weapons. (author)

  18. Verification of Chemical Weapons Destruction

    International Nuclear Information System (INIS)

    Lodding, J.

    2010-01-01

    The Chemical Weapons Convention is the only multilateral treaty that bans completely an entire category of weapons of mass destruction under international verification arrangements. Possessor States, i.e. those that have chemical weapons stockpiles at the time of becoming party to the CWC, commit to destroying these. All States undertake never to acquire chemical weapons and not to help other States acquire such weapons. The CWC foresees time-bound chemical disarmament. The deadlines for destruction for early entrants to the CWC are provided in the treaty. For late entrants, the Conference of States Parties intervenes to set destruction deadlines. One of the unique features of the CWC is thus the regime for verifying destruction of chemical weapons. But how can you design a system for verification at military sites, while protecting military restricted information? What degree of assurance is considered sufficient in such circumstances? How do you divide the verification costs? How do you deal with production capability and initial declarations of existing stockpiles? The founders of the CWC had to address these and other challenges in designing the treaty. Further refinement of the verification system has followed since the treaty opened for signature in 1993 and since inspection work was initiated following entry-into-force of the treaty in 1997. Most of this work concerns destruction at the two large possessor States, Russia and the United States. Perhaps some of the lessons learned from the OPCW experience may be instructive in a future verification regime for nuclear weapons. (author)

  19. Results of the independent radiological verification survey of the remedial action performed at the former Alba Craft Laboratory site, Oxford, Ohio, (OXO001)

    International Nuclear Information System (INIS)

    Kleinhans, K.R.; Murray, M.E.; Carrier, R.F.

    1996-04-01

    Between October 1952 and February 1957, National Lead of Ohio (NLO), a primary contractor for the Atomic Energy Commission (AEC), subcontracted certain uranium machining operations to Alba Craft Laboratory, Incorporated, located at 10-14 West Rose Avenue, Oxford, Ohio. In 1992, personnel from Oak Ridge National Laboratory (ORNL) confirmed the presence of residual radioactive materials from the AEC-related operations in and around the facility in amounts exceeding the applicable Department of Energy (DOE) guidelines. Although the amount of uranium found on the property posed little health hazard if left undisturbed, the levels were sufficient to require remediation to bring radiological conditions into compliance with current guidelines, thus ensuring that the public and the environment are protected. A team from ORNL conducted a radiological verification survey of the former Alba Craft Laboratory property between December 1994 and February 1995. The survey was conducted at the request of DOE and included directly measured radiation levels, the collection and analysis of soil samples to determine concentrations of uranium and certain other radionuclides, and comparison of these data to the guidelines. This document reports the findings of this survey. The results of the independent verification survey of the former Alba Craft Laboratory property demonstrate that all contaminated areas have been remediated to radionuclide concentrations and activity levels below the applicable guideline limits set by DOE

  20. Radioactive material packaging performance testing

    International Nuclear Information System (INIS)

    Romano, T.; Cruse, J.M.

    1991-02-01

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

  1. RASPLAV package

    International Nuclear Information System (INIS)

    1990-01-01

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

  2. Multi-canister overpack project - verification and validation, MCNP 4A

    International Nuclear Information System (INIS)

    Goldmann, L.H.

    1997-01-01

    This supporting document contains the software verification and validation (V and V) package used for Phase 2 design of the Spent Nuclear Fuel Multi-Canister Overpack. V and V packages for both ANSYS and MCNP are included. Description of Verification Run(s): This software requires that it be compiled specifically for the machine it is to be used on. Therefore to facilitate ease in the verification process the software automatically runs 25 sample problems to ensure proper installation and compilation. Once the runs are completed the software checks for verification by performing a file comparison on the new output file and the old output file. Any differences between any of the files will cause a verification error. Due to the manner in which the verification is completed a verification error does not necessarily indicate a problem. This indicates that a closer look at the output files is needed to determine the cause of the error

  3. msgbsR: An R package for analysing methylation-sensitive restriction enzyme sequencing data.

    Science.gov (United States)

    Mayne, Benjamin T; Leemaqz, Shalem Y; Buckberry, Sam; Rodriguez Lopez, Carlos M; Roberts, Claire T; Bianco-Miotto, Tina; Breen, James

    2018-02-01

    Genotyping-by-sequencing (GBS) or restriction-site associated DNA marker sequencing (RAD-seq) is a practical and cost-effective method for analysing large genomes from high diversity species. This method of sequencing, coupled with methylation-sensitive enzymes (often referred to as methylation-sensitive restriction enzyme sequencing or MRE-seq), is an effective tool to study DNA methylation in parts of the genome that are inaccessible in other sequencing techniques or are not annotated in microarray technologies. Current software tools do not fulfil all methylation-sensitive restriction sequencing assays for determining differences in DNA methylation between samples. To fill this computational need, we present msgbsR, an R package that contains tools for the analysis of methylation-sensitive restriction enzyme sequencing experiments. msgbsR can be used to identify and quantify read counts at methylated sites directly from alignment files (BAM files) and enables verification of restriction enzyme cut sites with the correct recognition sequence of the individual enzyme. In addition, msgbsR assesses DNA methylation based on read coverage, similar to RNA sequencing experiments, rather than methylation proportion and is a useful tool in analysing differential methylation on large populations. The package is fully documented and available freely online as a Bioconductor package ( https://bioconductor.org/packages/release/bioc/html/msgbsR.html ).

  4. 19 CFR 134.53 - Examination packages.

    Science.gov (United States)

    2010-04-01

    ... 19 Customs Duties 1 2010-04-01 2010-04-01 false Examination packages. 134.53 Section 134.53... TREASURY COUNTRY OF ORIGIN MARKING Articles Found Not Legally Marked § 134.53 Examination packages. (a) Site of marking—(1) Customs custody. Articles (or containers) in examination packages may be marked by...

  5. Tamper indicating packaging

    International Nuclear Information System (INIS)

    Baumann, M.J.; Bartberger, J.C.; Welch, T.D.

    1994-01-01

    Protecting sensitive items from undetected tampering in an unattended environment is crucial to the success of non-proliferation efforts relying on the verification of critical activities. Tamper Indicating Packaging (TIP) technologies are applied to containers, packages, and equipment that require an indication of a tamper attempt. Examples include: the transportation and storage of nuclear material, the operation and shipment of surveillance equipment and monitoring sensors, and the retail storage of medicine and food products. The spectrum of adversarial tampering ranges from attempted concealment of a pin-hole sized penetration to the complete container replacement, which would involve counterfeiting efforts of various degrees. Sandia National Laboratories (SNL) has developed a technology base for advanced TIP materials, sensors, designs, and processes which can be adapted to various future monitoring systems. The purpose of this technology base is to investigate potential new technologies, and to perform basic research of advanced technologies. This paper will describe the theory of TIP technologies and recent investigations of TIP technologies at SNL

  6. Software verification for nuclear industry

    International Nuclear Information System (INIS)

    Wilburn, N.P.

    1985-08-01

    Why verification of software products throughout the software life cycle is necessary is considered. Concepts of verification, software verification planning, and some verification methodologies for products generated throughout the software life cycle are then discussed

  7. Verification and disarmament

    Energy Technology Data Exchange (ETDEWEB)

    Blix, H. [IAEA, Vienna (Austria)

    1998-07-01

    The main features are described of the IAEA safeguards verification system that non-nuclear weapon states parties of the NPT are obliged to accept. Verification activities/problems in Iraq and North Korea are discussed.

  8. Verification and disarmament

    International Nuclear Information System (INIS)

    Blix, H.

    1998-01-01

    The main features are described of the IAEA safeguards verification system that non-nuclear weapon states parties of the NPT are obliged to accept. Verification activities/problems in Iraq and North Korea are discussed

  9. On-site verification trials using fly ash for reclamation behind bulkheads; Sekitanbai wo gogan uraumezai ni riyosuru genba jissho chosa

    Energy Technology Data Exchange (ETDEWEB)

    Kozawa, K [Center for Coal Utilization, Japan, Tokyo (Japan); Yoshida, T [Toyo Construction Co. Ltd., Tokyo (Japan); Miyagawa, H; Kobayashi, M

    1996-09-01

    As a method to utilize coal ash generated from coal burning power plants more effectively in bulk, its use has been studied as a reclamation material behind bulkhead structures in harbors and airports. Verification trials for the study results were performed at the Hekinan power plant of the Chubu Electric Power Company. The trials included the following: an experiment to verify horizontal soil pressure and active earth pressure when slurry made of fly ash added with cement and seawater was placed in frameworks installed behind bulkheads of a harbor, a slurry hardening test, environmental impact investigation, and constructibility investigation. As a result, a large number of findings were obtained, including the following matters: earth pressure of slurry which has been placed in a soil tank in about ten minutes would be measured as pressure of liquid, but it shifts to behavior as a soil in a relatively short time; the earth pressure after three hours agreed with static earth pressure calculated under provision of K{sub o} = 0.2; and a hardened body made with cement under a certain mixing ratio was obtained, which stands by itself at a height of 7.5 m at compression strength of 1.77 kgf/cm {sup 2}. 11 figs., 2 tabs.

  10. TRU waste transportation package development

    International Nuclear Information System (INIS)

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

    1980-01-01

    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

  11. Reflective Packaging

    Science.gov (United States)

    1994-01-01

    The aluminized polymer film used in spacecraft as a radiation barrier to protect both astronauts and delicate instruments has led to a number of spinoff applications. Among them are aluminized shipping bags, food cart covers and medical bags. Radiant Technologies purchases component materials and assembles a barrier made of layers of aluminized foil. The packaging reflects outside heat away from the product inside the container. The company is developing new aluminized lines, express mailers, large shipping bags, gel packs and insulated panels for the building industry.

  12. K Basin sludge packaging design criteria (PDC) and safety analysis report for packaging (SARP) approval plan

    International Nuclear Information System (INIS)

    Brisbin, S.A.

    1996-01-01

    This document delineates the plan for preparation, review, and approval of the Packaging Design Crieteria for the K Basin Sludge Transportation System and the Associated on-site Safety Analysis Report for Packaging. The transportation system addressed in the subject documents will be used to transport sludge from the K Basins using bulk packaging

  13. CATS Deliverable 5.1 : CATS verification of test matrix and protocol

    OpenAIRE

    Uittenbogaard, J.; Camp, O.M.G.C. op den; Montfort, S. van

    2016-01-01

    This report summarizes the work conducted within work package (WP) 5 "Verification of test matrix and protocol" of the Cyclist AEB testing system (CATS) project. It describes the verification process of the draft CATS test matrix resulting from WP1 and WP2, and the feasibility of meeting requirements set by CATS consortium based on requirements in Euro NCAP AEB protocols regarding accuracy, repeatability and reproducibility using the developed test hardware. For the cases where verification t...

  14. CATS Deliverable 5.1 : CATS verification of test matrix and protocol

    NARCIS (Netherlands)

    Uittenbogaard, J.; Camp, O.M.G.C. op den; Montfort, S. van

    2016-01-01

    This report summarizes the work conducted within work package (WP) 5 "Verification of test matrix and protocol" of the Cyclist AEB testing system (CATS) project. It describes the verification process of the draft CATS test matrix resulting from WP1 and WP2, and the feasibility of meeting

  15. RH Packaging Program Guidance

    International Nuclear Information System (INIS)

    Washington TRU Solutions, LLC

    2003-01-01

    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 and O) contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC approved, users need to be familiar with 10 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

  16. Physics Verification Overview

    Energy Technology Data Exchange (ETDEWEB)

    Doebling, Scott William [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-09-12

    The purpose of the verification project is to establish, through rigorous convergence analysis, that each ASC computational physics code correctly implements a set of physics models and algorithms (code verification); Evaluate and analyze the uncertainties of code outputs associated with the choice of temporal and spatial discretization (solution or calculation verification); and Develop and maintain the capability to expand and update these analyses on demand. This presentation describes project milestones.

  17. Complex-Wide Waste Flow Analysis V1.0 verification and validation report

    International Nuclear Information System (INIS)

    Hsu, K.M.; Lundeen, A.S.; Oswald, K.B.; Shropshire, D.E.; Robinson, J.M.; West, W.H.

    1997-01-01

    The complex-wide waste flow analysis model (CWWFA) was developed to assist the Department of Energy (DOE) Environmental Management (EM) Office of Science and Technology (EM-50) to evaluate waste management scenarios with emphasis on identifying and prioritizing technology development opportunities to reduce waste flows and public risk. In addition, the model was intended to support the needs of the Complex-Wide Environmental Integration (EMI) team supporting the DOE's Accelerating Cleanup: 2006 Plan. CWWFA represents an integrated environmental modeling system that covers the life cycle of waste management activities including waste generation, interim process storage, retrieval, characterization and sorting, waste preparation and processing, packaging, final interim storage, transport, and disposal at a final repository. The CWWFA shows waste flows through actual site-specific and facility-specific conditions. The system requirements for CWWFA are documented in the Technical Requirements Document (TRD). The TRD is intended to be a living document that will be modified over the course of the execution of CWWFA development. Thus, it is anticipated that CWWFA will continue to evolve as new requirements are identified (i.e., transportation, small sites, new streams, etc.). This report provides a documented basis for system verification of CWWFA requirements. System verification is accomplished through formal testing and evaluation to ensure that all performance requirements as specified in the TRD have been satisfied. A Requirement Verification Matrix (RVM) was used to map the technical requirements to the test procedures. The RVM is attached as Appendix A. Since February of 1997, substantial progress has been made toward development of the CWWFA to meet the system requirements. This system verification activity provides a baseline on system compliance to requirements and also an opportunity to reevaluate what requirements need to be satisfied in FY-98

  18. Inspector measurement verification activities

    International Nuclear Information System (INIS)

    George, R.S.; Crouch, R.

    e most difficult and complex activity facing a safeguards inspector involves the verification of measurements and the performance of the measurement system. Remeasurement is the key to measurement verification activities. Remeasurerements using the facility's measurement system provide the bulk of the data needed for determining the performance of the measurement system. Remeasurements by reference laboratories are also important for evaluation of the measurement system and determination of systematic errors. The use of these measurement verification activities in conjunction with accepted inventory verification practices provides a better basis for accepting or rejecting an inventory. (U.S.)

  19. RISKIND verification and benchmark comparisons

    International Nuclear Information System (INIS)

    Biwer, B.M.; Arnish, J.J.; Chen, S.Y.; Kamboj, S.

    1997-08-01

    This report presents verification calculations and benchmark comparisons for RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the population from exposures associated with the transportation of spent nuclear fuel and other radioactive materials. Spreadsheet calculations were performed to verify the proper operation of the major options and calculational steps in RISKIND. The program is unique in that it combines a variety of well-established models into a comprehensive treatment for assessing risks from the transportation of radioactive materials. Benchmark comparisons with other validated codes that incorporate similar models were also performed. For instance, the external gamma and neutron dose rate curves for a shipping package estimated by RISKIND were compared with those estimated by using the RADTRAN 4 code and NUREG-0170 methodology. Atmospheric dispersion of released material and dose estimates from the GENII and CAP88-PC codes. Verification results have shown the program to be performing its intended function correctly. The benchmark results indicate that the predictions made by RISKIND are within acceptable limits when compared with predictions from similar existing models

  20. RISKIND verification and benchmark comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Biwer, B.M.; Arnish, J.J.; Chen, S.Y.; Kamboj, S.

    1997-08-01

    This report presents verification calculations and benchmark comparisons for RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the population from exposures associated with the transportation of spent nuclear fuel and other radioactive materials. Spreadsheet calculations were performed to verify the proper operation of the major options and calculational steps in RISKIND. The program is unique in that it combines a variety of well-established models into a comprehensive treatment for assessing risks from the transportation of radioactive materials. Benchmark comparisons with other validated codes that incorporate similar models were also performed. For instance, the external gamma and neutron dose rate curves for a shipping package estimated by RISKIND were compared with those estimated by using the RADTRAN 4 code and NUREG-0170 methodology. Atmospheric dispersion of released material and dose estimates from the GENII and CAP88-PC codes. Verification results have shown the program to be performing its intended function correctly. The benchmark results indicate that the predictions made by RISKIND are within acceptable limits when compared with predictions from similar existing models.

  1. Requirements Verification Report AN Farm to 200E Waste Transfer System for Project W-314, Tank Farm Restoration and Safe Operations

    International Nuclear Information System (INIS)

    MCGREW, D.L.

    1999-01-01

    This Requirements Verification Report (RVR) for Project W-314 ''AN Farm to 200E Waste Transfer System'' package provides documented verification of design compliance to all the applicable Project Development Specification (PDS) requirements. Additional PDS requirements verification will be performed during the project's procurement, construction, and testing phases, and the RVR will be updated to reflect this information as appropriate

  2. Verification and the safeguards legacy

    International Nuclear Information System (INIS)

    Perricos, Demetrius

    2001-01-01

    of Iraq was a case of late detection of undeclared activities, the case of DPRK was a case of prompt detection of discrepancies in the initial declaration through implementation of modem detection techniques, such as environmental sampling, and access to information. Access to the Security Council became important in view of the protracted process of non-compliance. The Model Additional Protocol (INFCIRC 540) agreed in 1997 incorporates the results of the efforts to strengthen the safeguards system and as such provides the possibility for more transparency by the States and more access to locations by the inspectors on the basis of information. It does not provide the broad and intrusive access rights as in the case of Iraq, since such rights are unprecedented and the result of a cease-fire arrangement involving the Security Council. But the expectations are that the broad implementation of the Additional Protocol will result in an effective and efficient safeguards verification system for the future. The on-site verification systems on a national, regional or multinational basis that have been put into operation in the past or are being discussed by States for the implementation of disarmament and non-proliferation conventions related to weapons of mass destruction whether nuclear, chemical or biological, have benefited and will benefit in the future from the guiding experience - both from the strengths and weaknesses -of the IAEA verification system. This is hopefully a legacy for the future of verification

  3. Profiling a multiplex short tandem repeat loci from human urine with use of low cost on-site technology for verification of sample authenticity.

    Science.gov (United States)

    Pires, Nuno M M; Tao Dong; Berntzen, Lasse; Lonningdal, Torill

    2017-07-01

    This work focuses on the development of a sophisticated technique via STR typing to unequivocally verify the authenticity of urine samples before sent to laboratories. STR profiling was conducted with the CSF1PO, TPOX, TH01 Multiplex System coupled with a smartphone-based detection method. The promising capability of the method to identify distinct STR profiles from urine of different persons opens the possibility to conduct sample authenticity tests. On-site STR profiling could be realized with a self-contained autonomous device with an integrated PCR microchip shown hereby.

  4. The development of an on-site container

    International Nuclear Information System (INIS)

    Glass, R.E.; McAllaster, M.E.; Jones, P.L.; McKinney, A.L.

    1993-01-01

    This paper describes the results of the development of a packaging designed for on-site transportation of chemical munitions. The criteria for the package were patterned after the requirements for Type B packagings, but were modified to take credit for the operational controls that could be applied on-site. The design phase has been completed and a contract has been placed for fabrication of 165 units. In addition to the tests completed during the design phase, a complete sequence of verification tests will be performed on the first fleet unit. This sequence will consist of: (1) a 22,700-kg load placed on the packaging for 24 hours, (2) the flat side free drop, (3) the center-of-gravity over closure joint puncture test, and (4) a 15-min all-engulfing JP-4 pool-fire test. At the completion of the test sequence, the package will be required to meet the less than 1 x 10 -1 std cc/s leak test. (J.P.N.)

  5. Data Validation Package June 2016 Groundwater and Surface Water Sampling at the Old and New Rifle, Colorado, Processing Sites September 2016

    Energy Technology Data Exchange (ETDEWEB)

    Bush, Richard [USDOE Office of Legacy Management (LM), Washington, DC (United States); Lemke, Peter [Navarro Research and Engineering, Inc., Oak Ridge, TN (United States)

    2016-10-17

    Sampling Period: June 14–17 and July 7, 2016. Water samples were collected from 36 locations at New Rifle and Old Rifle, Colorado, Disposal/Processing Sites. Planned monitoring locations are shown in Attachment 1, Sampling and Analysis Work Order. Duplicate samples were collected from New Rifle locations 0216 and 0855, and Old Rifle location 0655. One equipment blank was collected after decontamination of non-dedicated equipment used to collect one surface water sample. See Attachment 2, Trip Report for additional details. Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated, http://energy.gov/lm/downloads/sampling-and- analysis-plan-us-department-energy-office-legacy-management-sites). New Rifle Site Samples were collected at the New Rifle site from 16 monitoring wells and 7 surface locations in compliance with the December 2008 Groundwater Compliance Action Plan [GCAP] for the New Rifle, Colorado, Processing Site (LMS/RFN/S01920). Monitoring well 0216 could not be sampled in June because it was surrounded by standing water due to the high river stage from spring runoff, it was later sampled in July. Monitoring well 0635 and surface location 0322 could not be sampled because access through the elk fence along Interstate 70 has not been completed at this time. Old Rifle Site Samples were collected at the Old Rifle site from eight monitoring wells and five surface locations in compliance with the December 2001 Ground Water Compliance Action Plan for the Old Rifle, Colorado, UMTRA Project Site (GJ0-2000-177-TAR).

  6. Data Validation Package May and June 2015 Groundwater and Surface Water Sampling at the Bluewater, New Mexico, Disposal Site, August 2015

    International Nuclear Information System (INIS)

    Johnson, Dick; Tsosie, Bernadette

    2015-01-01

    Groundwater samples were collected from monitoring wells at the Bluewater, New Mexico, Disposal Site to monitor groundwater contaminants as specified in the 1997 Long-Term Surveillance Plan for the DOE Bluewater (UMTRCA Title II) Disposal Site Near Grants, New Mexico (LTSP). Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). Duplicate samples were collected from locations 14(SG) and 21(M). Sampling originally scheduled for the week of May 11, 2015 was interrupted by heavy rainfall and later completed in June.

  7. Data Validation Package May and June 2015 Groundwater and Surface Water Sampling at the Bluewater, New Mexico, Disposal Site, August 2015

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, Dick [Navarro Research and Engineering, Inc., Oak Ridge, TN (United States); Tsosie, Bernadette [US Department of Energy, Washington, DC (United States)

    2015-08-01

    Groundwater samples were collected from monitoring wells at the Bluewater, New Mexico, Disposal Site to monitor groundwater contaminants as specified in the 1997 Long-Term Surveillance Plan for the DOE Bluewater (UMTRCA Title II) Disposal Site Near Grants, New Mexico (LTSP). Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). Duplicate samples were collected from locations 14(SG) and 21(M). Sampling originally scheduled for the week of May 11, 2015 was interrupted by heavy rainfall and later completed in June.

  8. Management of dumping of packaged low-level wastes in the deep ocean with emphasis on the North East Atlantic dump site

    International Nuclear Information System (INIS)

    Templeton, W.L.

    1981-08-01

    The following aspects are discussed: radiological principles as applied to disposal to the environment; international regulations; historical dumping practices; assessment of the Northeast Atlantic dump site; IAEA generic studies; and implications of issues on US needs and policies

  9. Verification Account Management System (VAMS)

    Data.gov (United States)

    Social Security Administration — The Verification Account Management System (VAMS) is the centralized location for maintaining SSA's verification and data exchange accounts. VAMS account management...

  10. Waste package materials selection process

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  11. PRIDE Surveillance Projects Data Packaging Project Information Package Specification Version 1.1

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-08-31

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

  12. SHIELD verification and validation report

    International Nuclear Information System (INIS)

    Boman, C.

    1992-02-01

    This document outlines the verification and validation effort for the SHIELD, SHLDED, GEDIT, GENPRT, FIPROD, FPCALC, and PROCES modules of the SHIELD system code. Along with its predecessors, SHIELD has been in use at the Savannah River Site (SRS) for more than ten years. During this time the code has been extensively tested and a variety of validation documents have been issued. The primary function of this report is to specify the features and capabilities for which SHIELD is to be considered validated, and to reference the documents that establish the validation

  13. Advanced verification topics

    CERN Document Server

    Bhattacharya, Bishnupriya; Hall, Gary; Heaton, Nick; Kashai, Yaron; Khan Neyaz; Kirshenbaum, Zeev; Shneydor, Efrat

    2011-01-01

    The Accellera Universal Verification Methodology (UVM) standard is architected to scale, but verification is growing and in more than just the digital design dimension. It is growing in the SoC dimension to include low-power and mixed-signal and the system integration dimension to include multi-language support and acceleration. These items and others all contribute to the quality of the SOC so the Metric-Driven Verification (MDV) methodology is needed to unify it all into a coherent verification plan. This book is for verification engineers and managers familiar with the UVM and the benefits it brings to digital verification but who also need to tackle specialized tasks. It is also written for the SoC project manager that is tasked with building an efficient worldwide team. While the task continues to become more complex, Advanced Verification Topics describes methodologies outside of the Accellera UVM standard, but that build on it, to provide a way for SoC teams to stay productive and profitable.

  14. Data Validation Package November 2015 Groundwater and Surface Water Sampling at the Old and New Rifle, Colorado, Processing Sites February 2016

    Energy Technology Data Exchange (ETDEWEB)

    Bush, Richard [USDOE Office of Legacy Management, Washington, DC (United States); Lemke, Peter [Navarro Research and Engineering, Inc., Oak Ridge, TN (United States)

    2016-02-01

    Water samples were collected from 36 locations at New Rifle and Old Rifle, Colorado, Processing Sites. Duplicate samples were collected from New Rifle locations 0659 and 0855, and Old Rifle location 0304. One equipment blank was collected after decontamination of non-dedicated equipment used to collect one surface water sample. Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). New Rifle Site Samples were collected at the New Rifle site from 16 monitoring wells and 7 surface locations in compliance with the December 2008 Groundwater Compliance Action Plan [GCAP] for the New Rifle, Colorado, Processing Site (LMS/RFN/S01920), with one exception: New Rifle location 0635 could not be sampled because it was inaccessible; a fence installed by the Colorado Department of Transportation prevents access to this location. DOE is currently negotiating access with the Colorado Department of Transportation. Analytes measured at the New Rifle site included contaminants of concern (COCs) (arsenic, molybdenum, nitrate + nitrite as nitrogen, selenium, uranium, and vanadium) ammonia as nitrogen, major cations, and major anions. Field measurements of total alkalinity, oxidation- reduction potential, pH, specific conductance, turbidity, and temperature were made at each location, and the water level was measured at each sampled well. A proposed alternate concentration limit (ACL) for vanadium of 50 milligrams per liter (mg/L), specific to the compliance (POC) wells (RFN-0217, -0659, -0664, and -0669) is included in the New Rifle GCAP. Vanadium concentrations in the POC wells were below the proposed ACL as shown in the time-concentration graphs in the Data Presentation section (Attachment 2). Time-concentration graphs from all other locations sampled are also included in Attachment 2. Sampling location RFN-0195 was misidentified for the June/August 2014

  15. The drift-flux correlation package MDS

    International Nuclear Information System (INIS)

    Hoeld, A.

    2001-01-01

    Based on the SONNENBURG drift-flux correlation, developed at GRS/Garching (Germany), a comprehensive drift-flux correlation package (MDS) has been established. Its aim is to support thermal-hydraulic mixture-fluid models, models being used for the simulation of the steady state and transient behaviour of characteristic thermal-hydraulic parameters of single- or two-phase fluids flowing along coolant channels of different types (being, e.g., parts of NPP-s, steam generators etc.). The characteristic properties of this package with respect to the behaviour at co- and counter-current flow, its inverse solutions needed for steady state simulations, its behaviour when approaching the lower or upper boundary of a two-phase region, its verification and behaviour with respect to other correlations will be discussed. An adequate driver code, MDSDRI, has been established too, allowing to test the package very thoroughly out of the complex thermal-hydraulic codes. (author)

  16. The drift-flux correlation package MDS

    Energy Technology Data Exchange (ETDEWEB)

    Hoeld, A. [Bernaysstr. 16A, Munich, F.R. (Germany)

    2001-07-01

    Based on the SONNENBURG drift-flux correlation, developed at GRS/Garching (Germany), a comprehensive drift-flux correlation package (MDS) has been established. Its aim is to support thermal-hydraulic mixture-fluid models, models being used for the simulation of the steady state and transient behaviour of characteristic thermal-hydraulic parameters of single- or two-phase fluids flowing along coolant channels of different types (being, e.g., parts of NPP-s, steam generators etc.). The characteristic properties of this package with respect to the behaviour at co- and counter-current flow, its inverse solutions needed for steady state simulations, its behaviour when approaching the lower or upper boundary of a two-phase region, its verification and behaviour with respect to other correlations will be discussed. An adequate driver code, MDSDRI, has been established too, allowing to test the package very thoroughly out of the complex thermal-hydraulic codes. (author)

  17. Nuclear test ban verification

    International Nuclear Information System (INIS)

    Chun, Kin-Yip

    1991-07-01

    This report describes verification and its rationale, the basic tasks of seismic verification, the physical basis for earthquake/explosion source discrimination and explosion yield determination, the technical problems pertaining to seismic monitoring of underground nuclear tests, the basic problem-solving strategy deployed by the forensic seismology resarch team at the University of Toronto, and the scientific significance of the team's research. The research carried out at the Univeristy of Toronto has two components: teleseismic verification using P wave recordings from the Yellowknife Seismic Array (YKA), and regional (close-in) verification using high-frequency L g and P n recordings from the Eastern Canada Telemetered Network. Major differences have been found in P was attenuation among the propagation paths connecting the YKA listening post with seven active nuclear explosion testing areas in the world. Significant revisions have been made to previously published P wave attenuation results, leading to more interpretable nuclear explosion source functions. (11 refs., 12 figs.)

  18. Standard Verification System (SVS)

    Data.gov (United States)

    Social Security Administration — SVS is a mainframe program that accesses the NUMIDENT to perform SSN verifications. This program is called by SSA Internal applications to verify SSNs. There is also...

  19. Formal Verification -26 ...

    Indian Academy of Sciences (India)

    by testing of the components and successful testing leads to the software being ... Formal verification is based on formal methods which are mathematically based ..... scenario under which a similar error could occur. There are various other ...

  20. SSN Verification Service

    Data.gov (United States)

    Social Security Administration — The SSN Verification Service is used by Java applications to execute the GUVERF02 service using the WebSphere/CICS Interface. It accepts several input data fields...

  1. Environmental technology verification methods

    CSIR Research Space (South Africa)

    Szewczuk, S

    2016-03-01

    Full Text Available Environmental Technology Verification (ETV) is a tool that has been developed in the United States of America, Europe and many other countries around the world to help innovative environmental technologies reach the market. Claims about...

  2. Verification of RADTRAN

    International Nuclear Information System (INIS)

    Kanipe, F.L.; Neuhauser, K.S.

    1995-01-01

    This document presents details of the verification process of the RADTRAN computer code which was established for the calculation of risk estimates for radioactive materials transportation by highway, rail, air, and waterborne modes

  3. Studies relating to human intrusion into a repository. Report pertaining to work package 11. Preliminary safety case of the Gorleben site (VSG)

    Energy Technology Data Exchange (ETDEWEB)

    Beuth, Thomas; Buhmann, Dieter; Fischer-Appelt, Klaus; Moenig, Joerg; Ruebel, Andre; Wolf, Jens [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany); Bollingerfehr, Wilhelm; Filbert, Wolfgang [DBE Technology GmbH, Peine (Germany); Charlier, Frank [international nuclear safety engineering gmbh (nse), Aachen (Germany); Baltes, Bruno

    2014-10-15

    The question of the long-term safety of a repository system is inseparably linked with the intensive technical examination of the possible future evolution of the site and the repository system e. g. as a result of climatic, geologic, waste-related and repository-related processes. Here, the possible evolutions to be considered are those that have the potential to have a negative impact on the intended, furthest-possible, immediate, and lasting isolation of the radioactive waste in a defined area around the underground workings of the repository mine in salt rock, which is referred to as the containment-providing rock zone (CPRZ).

  4. A waste package strategy for regulatory compliance

    International Nuclear Information System (INIS)

    Stahl, D.; Cloninger, M.O.

    1990-01-01

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

  5. Multilateral disarmament verification

    International Nuclear Information System (INIS)

    Persbo, A.

    2013-01-01

    Non-governmental organisations, such as VERTIC (Verification Research, Training and Information Centre), can play an important role in the promotion of multilateral verification. Parties involved in negotiating nuclear arms accords are for the most part keen that such agreements include suitable and robust provisions for monitoring and verification. Generally progress in multilateral arms control verification is often painstakingly slow, but from time to time 'windows of opportunity' - that is, moments where ideas, technical feasibility and political interests are aligned at both domestic and international levels - may occur and we have to be ready, so the preparatory work is very important. In the context of nuclear disarmament, verification (whether bilateral or multilateral) entails an array of challenges, hurdles and potential pitfalls relating to national security, health, safety and even non-proliferation, so preparatory work is complex and time-greedy. A UK-Norway Initiative was established in order to investigate the role that a non-nuclear-weapon state such as Norway could potentially play in the field of nuclear arms control verification. (A.C.)

  6. TFE design package final report, TFE Verification Program

    International Nuclear Information System (INIS)

    1994-06-01

    The program objective is to demonstrate the technology readiness of a TFE suitable for use as the basic element in a thermionic reactor with electric power output in the 0.5 to 5.0 MW(e) range, and a full-power life of 7 years. A TFE for a megawatt class system is described. Only six cells are considered for simplicity; a megawatt class TFE would have many more cells, the exact number dependent on optimization trade studies

  7. Emergency response packaging: A conceptual outline

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  8. Geochemical analysis of the sealing system. Technical Report to work package 9.1.2. Preliminary safety case of the Gorleben site (VSG)

    Energy Technology Data Exchange (ETDEWEB)

    Xie, Mingliang; Herbert, Horst-Juergen

    2012-02-15

    The geochemical reaction simulation of ground water and brines with sealing materials for a proposed repository at the Gorleben site is a task under the framework of the VSG project. The calculations presented in this report are aimed to provide a preliminary evaluation of the geochemical stability of three potential engineering barrier materials foreseen in the shaft sealing system in case of groundwater and brine intrusion. The long-term stability of these materials is a key issue for the sealing function of the shaft seals. This is governed by many factors such as geotechnical, hydraulic and geochemical processes. In order to better understand the potential effect of geochemical processes on the long-term properties of these sealing materials, geochemical simulations of the potential interactions between groundwater and brine and shaft sealing materials were performed.

  9. Embedded software verification and debugging

    CERN Document Server

    Winterholer, Markus

    2017-01-01

    This book provides comprehensive coverage of verification and debugging techniques for embedded software, which is frequently used in safety critical applications (e.g., automotive), where failures are unacceptable. Since the verification of complex systems needs to encompass the verification of both hardware and embedded software modules, this book focuses on verification and debugging approaches for embedded software with hardware dependencies. Coverage includes the entire flow of design, verification and debugging of embedded software and all key approaches to debugging, dynamic, static, and hybrid verification. This book discusses the current, industrial embedded software verification flow, as well as emerging trends with focus on formal and hybrid verification and debugging approaches. Includes in a single source the entire flow of design, verification and debugging of embedded software; Addresses the main techniques that are currently being used in the industry for assuring the quality of embedded softw...

  10. Low-level waste packaging--a managerial perspective

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  11. Stabilizing And Packaging Pu Materials Per 3013 At SRS

    International Nuclear Information System (INIS)

    STEVE, HENSEL

    2005-01-01

    The Savannah River Site (SRS) began packaging Pu metals into 3013 containers in April, 2003 and oxides in October, 2003. A total of 919 outer 3013 containers were made in the FB-Line at SRS when stabilization and packaging was completed in January, 2005. Experiences, lessons learned, and an overview of packaging activities are presented

  12. Waste package performance assessment

    International Nuclear Information System (INIS)

    Lester, D.H.

    1981-01-01

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

  13. Reference waste package environment report

    International Nuclear Information System (INIS)

    Glassley, W.E.

    1986-01-01

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

  14. Tritium waste package

    Science.gov (United States)

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

    1995-01-01

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

  15. Modeling of radionuclide and heavy metal sorption around low and high pH waste disposal sites at Oak Ridge, Tennessee: Classification review package

    International Nuclear Information System (INIS)

    Saunders, J.A.

    1994-10-01

    Modeling of mineral precipitation and metal sorption reactions using MINTEQA2 and the iron oxyhydroxide diffuse-layer model has provided insights into geochemical processes governing contaminant migration from low-level radioactive waste disposal sites at the US Department of Energy's Oak Ridge National Laboratory and Y-12 Plant at Oak Ridge, Tennessee. Both acidic and basic nuclear-fuel reprocessing wastes, locally mixed with decontamination solvents, were disposed of in unlined trenches and lagoons. Model results show that as wastes move toward neutral pH due to reactions with surrounding soils and saprolite, mineral precipitation and sorption can limit the solubility of heavy metals and radionuclides. However, observed contaminant levels in monitoring wells indicate that at least locally, wastes are moving in faults and fractures and are not retarded by sorption reactions along such flow paths. Model results also support previous studies that have indicated organic complexing agents used in decontamination procedures can enhance radionuclide and heavy metal solubility when mixed with nuclear fuel reprocessing wastes. However, complex interactions between metal-organic complexes and mineral surfaces and natural organic matter, biodegradation, and fracture flow complicate the interpretation of contaminant mobility

  16. Packaging for Food Service

    Science.gov (United States)

    Stilwell, E. J.

    1985-01-01

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

  17. Waste Package Lifting Calculation

    International Nuclear Information System (INIS)

    H. Marr

    2000-01-01

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

  18. Code verification and Y2K testing, calibration, testing, and installation of the radionuclide assay system-photon (RAS-P) at multiple sites for the Savannah River Site

    International Nuclear Information System (INIS)

    Hodge, C.A.

    2000-01-01

    The Radionuclide Assay System - Photon (RAS-P) is a near-field, transmission-corrected assay system developed for measurement of the actinide content of relatively homogeneous waste generated by facility operations. It is intended for use by facility operations personnel, and has features to enhance its usefulness and efficiency. These include multinuclide assay capability, automatic (off-shift) collection of background and straight-through transmission source data, enforcement of measurement control requirements, Go-NoGo or Assay modes, password protection, and reporting of total fissile gram equivalent values. System hardware consists of a shielded high-resolution germanium detector, a turntable, a shielded transmission source and shutter assembly, and a desktop computer and laser printer mounted on a compact frame. RAS-P was designed to assay the contents of cylindrical containers up to 30 inches diameter by 32 inches high, boxes up to 30 inches diagonal by 32 inches high, and HE PA filters up to 2 x 2 x 1 feet. Prior to installation at the Savannah River Site (SRS), code validation, system performance, and assurance against Y2K effects all were confirmed. Code validation was accomplished using spreadsheet calculations that were independent of the original code to calculate intermediate and final result produced by RAS-P. System testing was performed by repeated operation of the instrument under all required circumstances. Y2K testing was performed simultaneously with code validation following a protocol prescribed by the SRS Y2K subcommittee that required assays with dates varying throughout the expected useful life of the RAS-P, particularly those bracketing Y2K boundaries. Performance history has been compiled demonstrating reliability (system availability), diversability (the ability to alter assay parameters and obtain results quickly), and measurement control characteristics

  19. RNA secondary structures of the bacteriophage phi6 packaging regions.

    OpenAIRE

    Pirttimaa, M J; Bamford, D H

    2000-01-01

    Bacteriophage phi6 genome consists of three segments of double-stranded RNA. During maturation, single-stranded copies of these segments are packaged into preformed polymerase complex particles. Only phi6 RNA is packaged, and each particle contains only one copy of each segment. An in vitro packaging and replication assay has been developed for phi6, and the packaging signals (pac sites) have been mapped to the 5' ends of the RNA segments. In this study, we propose secondary structure models ...

  20. Procedure generation and verification

    International Nuclear Information System (INIS)

    Sheely, W.F.

    1986-01-01

    The Department of Energy has used Artificial Intelligence of ''AI'' concepts to develop two powerful new computer-based techniques to enhance safety in nuclear applications. The Procedure Generation System, and the Procedure Verification System, can be adapted to other commercial applications, such as a manufacturing plant. The Procedure Generation System can create a procedure to deal with the off-normal condition. The operator can then take correct actions on the system in minimal time. The Verification System evaluates the logic of the Procedure Generator's conclusions. This evaluation uses logic techniques totally independent of the Procedure Generator. The rapid, accurate generation and verification of corrective procedures can greatly reduce the human error, possible in a complex (stressful/high stress) situation

  1. Nuclear disarmament verification

    International Nuclear Information System (INIS)

    DeVolpi, A.

    1993-01-01

    Arms control treaties, unilateral actions, and cooperative activities -- reflecting the defusing of East-West tensions -- are causing nuclear weapons to be disarmed and dismantled worldwide. In order to provide for future reductions and to build confidence in the permanency of this disarmament, verification procedures and technologies would play an important role. This paper outlines arms-control objectives, treaty organization, and actions that could be undertaken. For the purposes of this Workshop on Verification, nuclear disarmament has been divided into five topical subareas: Converting nuclear-weapons production complexes, Eliminating and monitoring nuclear-weapons delivery systems, Disabling and destroying nuclear warheads, Demilitarizing or non-military utilization of special nuclear materials, and Inhibiting nuclear arms in non-nuclear-weapons states. This paper concludes with an overview of potential methods for verification

  2. Merganser Download Package

    Data.gov (United States)

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

  3. Creative Thinking Package

    Science.gov (United States)

    Jones, Clive

    1972-01-01

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

  4. Verification of Ceramic Structures

    Science.gov (United States)

    Behar-Lafenetre, Stephanie; Cornillon, Laurence; Rancurel, Michael; De Graaf, Dennis; Hartmann, Peter; Coe, Graham; Laine, Benoit

    2012-07-01

    In the framework of the “Mechanical Design and Verification Methodologies for Ceramic Structures” contract [1] awarded by ESA, Thales Alenia Space has investigated literature and practices in affiliated industries to propose a methodological guideline for verification of ceramic spacecraft and instrument structures. It has been written in order to be applicable to most types of ceramic or glass-ceramic materials - typically Cesic®, HBCesic®, Silicon Nitride, Silicon Carbide and ZERODUR®. The proposed guideline describes the activities to be performed at material level in order to cover all the specific aspects of ceramics (Weibull distribution, brittle behaviour, sub-critical crack growth). Elementary tests and their post-processing methods are described, and recommendations for optimization of the test plan are given in order to have a consistent database. The application of this method is shown on an example in a dedicated article [7]. Then the verification activities to be performed at system level are described. This includes classical verification activities based on relevant standard (ECSS Verification [4]), plus specific analytical, testing and inspection features. The analysis methodology takes into account the specific behaviour of ceramic materials, especially the statistical distribution of failures (Weibull) and the method to transfer it from elementary data to a full-scale structure. The demonstration of the efficiency of this method is described in a dedicated article [8]. The verification is completed by classical full-scale testing activities. Indications about proof testing, case of use and implementation are given and specific inspection and protection measures are described. These additional activities are necessary to ensure the required reliability. The aim of the guideline is to describe how to reach the same reliability level as for structures made of more classical materials (metals, composites).

  5. Verification and validation benchmarks.

    Energy Technology Data Exchange (ETDEWEB)

    Oberkampf, William Louis; Trucano, Timothy Guy

    2007-02-01

    Verification and validation (V&V) are the primary means to assess the accuracy and reliability of computational simulations. V&V methods and procedures have fundamentally improved the credibility of simulations in several high-consequence fields, such as nuclear reactor safety, underground nuclear waste storage, and nuclear weapon safety. Although the terminology is not uniform across engineering disciplines, code verification deals with assessing the reliability of the software coding, and solution verification deals with assessing the numerical accuracy of the solution to a computational model. Validation addresses the physics modeling accuracy of a computational simulation by comparing the computational results with experimental data. Code verification benchmarks and validation benchmarks have been constructed for a number of years in every field of computational simulation. However, no comprehensive guidelines have been proposed for the construction and use of V&V benchmarks. For example, the field of nuclear reactor safety has not focused on code verification benchmarks, but it has placed great emphasis on developing validation benchmarks. Many of these validation benchmarks are closely related to the operations of actual reactors at near-safety-critical conditions, as opposed to being more fundamental-physics benchmarks. This paper presents recommendations for the effective design and use of code verification benchmarks based on manufactured solutions, classical analytical solutions, and highly accurate numerical solutions. In addition, this paper presents recommendations for the design and use of validation benchmarks, highlighting the careful design of building-block experiments, the estimation of experimental measurement uncertainty for both inputs and outputs to the code, validation metrics, and the role of model calibration in validation. It is argued that the understanding of predictive capability of a computational model is built on the level of

  6. Verification and validation benchmarks

    International Nuclear Information System (INIS)

    Oberkampf, William Louis; Trucano, Timothy Guy

    2007-01-01

    Verification and validation (V and V) are the primary means to assess the accuracy and reliability of computational simulations. V and V methods and procedures have fundamentally improved the credibility of simulations in several high-consequence fields, such as nuclear reactor safety, underground nuclear waste storage, and nuclear weapon safety. Although the terminology is not uniform across engineering disciplines, code verification deals with assessing the reliability of the software coding, and solution verification deals with assessing the numerical accuracy of the solution to a computational model. Validation addresses the physics modeling accuracy of a computational simulation by comparing the computational results with experimental data. Code verification benchmarks and validation benchmarks have been constructed for a number of years in every field of computational simulation. However, no comprehensive guidelines have been proposed for the construction and use of V and V benchmarks. For example, the field of nuclear reactor safety has not focused on code verification benchmarks, but it has placed great emphasis on developing validation benchmarks. Many of these validation benchmarks are closely related to the operations of actual reactors at near-safety-critical conditions, as opposed to being more fundamental-physics benchmarks. This paper presents recommendations for the effective design and use of code verification benchmarks based on manufactured solutions, classical analytical solutions, and highly accurate numerical solutions. In addition, this paper presents recommendations for the design and use of validation benchmarks, highlighting the careful design of building-block experiments, the estimation of experimental measurement uncertainty for both inputs and outputs to the code, validation metrics, and the role of model calibration in validation. It is argued that the understanding of predictive capability of a computational model is built on the

  7. Verification and validation benchmarks

    International Nuclear Information System (INIS)

    Oberkampf, William L.; Trucano, Timothy G.

    2008-01-01

    Verification and validation (V and V) are the primary means to assess the accuracy and reliability of computational simulations. V and V methods and procedures have fundamentally improved the credibility of simulations in several high-consequence fields, such as nuclear reactor safety, underground nuclear waste storage, and nuclear weapon safety. Although the terminology is not uniform across engineering disciplines, code verification deals with assessing the reliability of the software coding, and solution verification deals with assessing the numerical accuracy of the solution to a computational model. Validation addresses the physics modeling accuracy of a computational simulation by comparing the computational results with experimental data. Code verification benchmarks and validation benchmarks have been constructed for a number of years in every field of computational simulation. However, no comprehensive guidelines have been proposed for the construction and use of V and V benchmarks. For example, the field of nuclear reactor safety has not focused on code verification benchmarks, but it has placed great emphasis on developing validation benchmarks. Many of these validation benchmarks are closely related to the operations of actual reactors at near-safety-critical conditions, as opposed to being more fundamental-physics benchmarks. This paper presents recommendations for the effective design and use of code verification benchmarks based on manufactured solutions, classical analytical solutions, and highly accurate numerical solutions. In addition, this paper presents recommendations for the design and use of validation benchmarks, highlighting the careful design of building-block experiments, the estimation of experimental measurement uncertainty for both inputs and outputs to the code, validation metrics, and the role of model calibration in validation. It is argued that the understanding of predictive capability of a computational model is built on the

  8. Reload core safety verification

    International Nuclear Information System (INIS)

    Svetlik, M.; Minarcin, M.

    2003-01-01

    This paper presents a brief look at the process of reload core safety evaluation and verification in Slovak Republic. It gives an overview of experimental verification of selected nuclear parameters in the course of physics testing during reactor start-up. The comparison of IAEA recommendations and testing procedures at Slovak and European nuclear power plants of similar design is included. An introduction of two level criteria for evaluation of tests represents an effort to formulate the relation between safety evaluation and measured values (Authors)

  9. Packaging Printing Today

    OpenAIRE

    Stanislav Bolanča; Igor Majnarić; Kristijan Golubović

    2015-01-01

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

  10. Genome packaging in viruses

    OpenAIRE

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

    2010-01-01

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

  11. Trends in Food Packaging.

    Science.gov (United States)

    Ott, Dana B.

    1988-01-01

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

  12. Transparencies used in describing the CTBT verification regime and its four monitoring technologies

    International Nuclear Information System (INIS)

    Basham, P.

    1999-01-01

    This presentation includes description of the CTBT verification regime and its four monitoring technologies, (namely, seismic monitoring, hydro acoustic monitoring, infrasound monitoring and radionuclides monitoring) CTBT global verification system, sequence of steps needed for installing an international monitoring system station which includes: site survey, site preparation and construction, equipment procurement and installation, final tests and certification

  13. Radioactive material package testing capabilities at Sandia National Laboratories

    International Nuclear Information System (INIS)

    Uncapher, W.L.; Hohnstreiter, G.F.

    1995-01-01

    Evaluation and certification of radioactive and hazardous material transport packages can be accomplished by subjecting these packages to normal transport and hypothetical accident test conditions. The regulations allow package designers to certify packages using analysis, testing, or a combination of analysis and testing. Testing can be used to substantiate assumptions used in analytical models and to demonstrate package structural and thermal response. Regulatory test conditions include impact, puncture, crush, penetration, water spray, immersion, and thermal environments. Testing facilities are used to simulate the required test conditions and provide measurement response data. Over the past four decades, comprehensive testing facilities have been developed at Sandia National Laboratories to perform a broad range of verification and certification tests on hazardous and radioactive material packages or component sections. Sandia's facilities provide an experience base that has been established during the development and certification of many package designs. These unique facilities, along with innovative instrumentation data collection capabilities and techniques, simulate a broad range of testing environments. In certain package designs, package testing can be an economical alternative to complex analysis to resolve regulatory questions or concerns

  14. WE-D-BRA-04: Online 3D EPID-Based Dose Verification for Optimum Patient Safety

    International Nuclear Information System (INIS)

    Spreeuw, H; Rozendaal, R; Olaciregui-Ruiz, I; Mans, A; Mijnheer, B; Herk, M van; Gonzalez, P

    2015-01-01

    Purpose: To develop an online 3D dose verification tool based on EPID transit dosimetry to ensure optimum patient safety in radiotherapy treatments. Methods: A new software package was developed which processes EPID portal images online using a back-projection algorithm for the 3D dose reconstruction. The package processes portal images faster than the acquisition rate of the portal imager (∼ 2.5 fps). After a portal image is acquired, the software seeks for “hot spots” in the reconstructed 3D dose distribution. A hot spot is in this study defined as a 4 cm 3 cube where the average cumulative reconstructed dose exceeds the average total planned dose by at least 20% and 50 cGy. If a hot spot is detected, an alert is generated resulting in a linac halt. The software has been tested by irradiating an Alderson phantom after introducing various types of serious delivery errors. Results: In our first experiment the Alderson phantom was irradiated with two arcs from a 6 MV VMAT H&N treatment having a large leaf position error or a large monitor unit error. For both arcs and both errors the linac was halted before dose delivery was completed. When no error was introduced, the linac was not halted. The complete processing of a single portal frame, including hot spot detection, takes about 220 ms on a dual hexacore Intel Xeon 25 X5650 CPU at 2.66 GHz. Conclusion: A prototype online 3D dose verification tool using portal imaging has been developed and successfully tested for various kinds of gross delivery errors. The detection of hot spots was proven to be effective for the timely detection of these errors. Current work is focused on hot spot detection criteria for various treatment sites and the introduction of a clinical pilot program with online verification of hypo-fractionated (lung) treatments

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

    International Nuclear Information System (INIS)

    Romano, T.

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    1998-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

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

  18. DOT-7A packaging test procedure

    International Nuclear Information System (INIS)

    Kelly, D.L.

    1995-01-01

    This test procedure documents the steps involved with performance testing of Department of Transportation Specification 7A (DOT-7A) Type A packages. It includes description of the performance tests, the personnel involved, appropriate safety considerations, and the procedures to be followed while performing the tests. Westinghouse Hanford Company (WHC) is conducting the evaluation and testing discussed herein for the Department of Energy-Headquarters, Division of Quality Verification and Transportation Safety (EH-321). Please note that this report is not in WHC format. This report is being submitted through the Engineering Documentation System so that it may be used for reference and information purposes

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  20. Active Packaging Coatings

    Directory of Open Access Journals (Sweden)

    Luis J. Bastarrachea

    2015-11-01

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

  1. Is flow verification necessary

    International Nuclear Information System (INIS)

    Beetle, T.M.

    1986-01-01

    Safeguards test statistics are used in an attempt to detect diversion of special nuclear material. Under assumptions concerning possible manipulation (falsification) of safeguards accounting data, the effects on the statistics due to diversion and data manipulation are described algebraically. A comprehensive set of statistics that is capable of detecting any diversion of material is defined in terms of the algebraic properties of the effects. When the assumptions exclude collusion between persons in two material balance areas, then three sets of accounting statistics are shown to be comprehensive. Two of the sets contain widely known accountancy statistics. One of them does not require physical flow verification - comparisons of operator and inspector data for receipts and shipments. The third set contains a single statistic which does not require physical flow verification. In addition to not requiring technically difficult and expensive flow verification, this single statistic has several advantages over other comprehensive sets of statistics. This algebraic approach as an alternative to flow verification for safeguards accountancy is discussed in this paper

  2. Integrated Java Bytecode Verification

    DEFF Research Database (Denmark)

    Gal, Andreas; Probst, Christian; Franz, Michael

    2005-01-01

    Existing Java verifiers perform an iterative data-flow analysis to discover the unambiguous type of values stored on the stack or in registers. Our novel verification algorithm uses abstract interpretation to obtain definition/use information for each register and stack location in the program...

  3. ITER interim design report package and relevant documents

    International Nuclear Information System (INIS)

    1996-01-01

    This publication documents the technical basis which underlay the Interim Design Report, Cost Review and Safety Analysis submitted to the ITER Councils (IC-8 and IC-9) Records of decisions and the ''ITER Interim Design Report Package''. This publication contains ITER Site Requirements and ITER Site Design Assumptions, TAC-8 Report, SRG Report, CP's Report on Tentative Sequence of Events and Parties' Views on the IDR Package and Parties' Technical Comments on the IDR Package. Figs, tabs

  4. Advancements in meat packaging.

    Science.gov (United States)

    McMillin, Kenneth W

    2017-10-01

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

  5. Computerized waste-accountability shipping and packaging system

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  6. Edible packaging materials.

    Science.gov (United States)

    Janjarasskul, Theeranun; Krochta, John M

    2010-01-01

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

  7. Verification Games: Crowd-Sourced Formal Verification

    Science.gov (United States)

    2016-03-01

    additional paintbrushes. Additionally, in Paradox , human players are never given small optimization problems (for example, toggling the values of 50...were developed by the Center for Game Science: Pipe Jam, Traffic Jam, Flow Jam and Paradox . Verification tools and games were integrated to verify...4 4. Paradox …………………………………………………......5 5. MyClass ………………………………………………….....7 6. Results …………………………………………………......11 7. Time to

  8. Verification Survey of Uranium Mine Remediation

    International Nuclear Information System (INIS)

    Ron, Stager

    2009-01-01

    The Canadian Nuclear Safety Commission (CNSC) contracted an independent verification of an intensive gamma radiation survey conducted by a mining company to demonstrate that remediation of disturbed areas was complete. This site was the first of the recent mines being decommissioned in Canada and experience gained here may be applied to other mines being decommissioned in the future. The review included examination of the site-specific basis for clean-up criteria and ALARA as required by CNSC guidance. A paper review of the company report was conducted to determine if protocols were followed and that the summarized results could be independently reproduced. An independent verification survey was conducted on parts of the site and comparisons were made between gamma radiation measurements from the verification survey and the original company survey. Some aspects of data collection using rate meters linked to GPS data loggers are discussed as are aspects for data management and analyses methods required for the large amount of data collected during these surveys. Recommendations were made for implementation of future surveys and reporting the data from those surveys in order to ensure that remediation was complete. (authors)

  9. Innovative Approaches to Large Component Packaging

    International Nuclear Information System (INIS)

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

    2006-01-01

    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)

  10. The design of verification regimes

    International Nuclear Information System (INIS)

    Gallagher, N.W.

    1991-01-01

    Verification of a nuclear agreement requires more than knowledge of relevant technologies and institutional arrangements. It also demands thorough understanding of the nature of verification and the politics of verification design. Arms control efforts have been stymied in the past because key players agreed to verification in principle, only to disagree radically over verification in practice. In this chapter, it is shown that the success and stability of arms control endeavors can be undermined by verification designs which promote unilateral rather than cooperative approaches to security, and which may reduce, rather than enhance, the security of both sides. Drawing on logical analysis and practical lessons from previous superpower verification experience, this chapter summarizes the logic and politics of verification and suggests implications for South Asia. The discussion begins by determining what properties all forms of verification have in common, regardless of the participants or the substance and form of their agreement. Viewing verification as the political process of making decisions regarding the occurrence of cooperation points to four critical components: (1) determination of principles, (2) information gathering, (3) analysis and (4) projection. It is shown that verification arrangements differ primarily in regards to how effectively and by whom these four stages are carried out

  11. Generalized waste package containment model

    International Nuclear Information System (INIS)

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

    1985-02-01

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

  12. Verification of the thermal design of electronic equipment

    Energy Technology Data Exchange (ETDEWEB)

    Hienonen, R.; Karjalainen, M.; Lankinen, R. [VTT Automation, Espoo (Finland). ProTechno

    1997-12-31

    The project `Verification of the thermal design of electronic equipment` studied the methodology to be followed in the verification of thermal design of electronic equipment. This project forms part of the `Cool Electronics` research programme funded by TEKES, the Finnish Technology Development Centre. This project was carried out jointly by VTT Automation, Lappeenranta University of Technology, Nokia Research Center and ABB Industry Oy VSD-Technology. The thermal design of electronic equipment has a significant impact on the cost, reliability, tolerance to different environments, selection of components and materials, and ergonomics of the product. This report describes the method for verification of thermal design. It assesses the goals set for thermal design, environmental requirements, technical implementation of the design, thermal simulation and modelling, and design qualification testing and the measurements needed. The verification method covers all packaging levels of electronic equipment from the system level to the electronic component level. The method described in this report can be used as part of the quality system of a corporation. The report includes information about the measurement and test methods needed in the verification process. Some measurement methods for the temperature, flow and pressure of air are described. (orig.) Published in Finnish VTT Julkaisuja 824. 22 refs.

  13. Lithium niobate packaging challenges

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  14. Packaged die heater

    Science.gov (United States)

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

    2011-06-21

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

  15. Packaging for Sustainability

    CERN Document Server

    Lewis, Helen; Fitzpatrick, Leanne

    2012-01-01

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

  16. Distorted Fingerprint Verification System

    Directory of Open Access Journals (Sweden)

    Divya KARTHIKAESHWARAN

    2011-01-01

    Full Text Available Fingerprint verification is one of the most reliable personal identification methods. Fingerprint matching is affected by non-linear distortion introduced in fingerprint impression during the image acquisition process. This non-linear deformation changes both the position and orientation of minutiae. The proposed system operates in three stages: alignment based fingerprint matching, fuzzy clustering and classifier framework. First, an enhanced input fingerprint image has been aligned with the template fingerprint image and matching score is computed. To improve the performance of the system, a fuzzy clustering based on distance and density has been used to cluster the feature set obtained from the fingerprint matcher. Finally a classifier framework has been developed and found that cost sensitive classifier produces better results. The system has been evaluated on fingerprint database and the experimental result shows that system produces a verification rate of 96%. This system plays an important role in forensic and civilian applications.

  17. RESRAD-BUILD verification

    International Nuclear Information System (INIS)

    Kamboj, S.; Yu, C.; Biwer, B. M.; Klett, T.

    2002-01-01

    The results generated by the RESRAD-BUILD code (version 3.0) were verified with hand or spreadsheet calculations using equations given in the RESRAD-BUILD manual for different pathways. For verification purposes, different radionuclides--H-3, C-14, Na-22, Al-26, Cl-36, Mn-54, Co-60, Au-195, Ra-226, Ra-228, Th-228, and U-238--were chosen to test all pathways and models. Tritium, Ra-226, and Th-228 were chosen because of the special tritium and radon models in the RESRAD-BUILD code. Other radionuclides were selected to represent a spectrum of radiation types and energies. Verification of the RESRAD-BUILD code was conducted with an initial check of all the input parameters for correctness against their original source documents. Verification of the calculations was performed external to the RESRAD-BUILD code with Microsoft Excel to verify all the major portions of the code. In some cases, RESRAD-BUILD results were compared with those of external codes, such as MCNP (Monte Carlo N-particle) and RESRAD. The verification was conducted on a step-by-step basis and used different test cases as templates. The following types of calculations were investigated: (1) source injection rate, (2) air concentration in the room, (3) air particulate deposition, (4) radon pathway model, (5) tritium model for volume source, (6) external exposure model, (7) different pathway doses, and (8) time dependence of dose. Some minor errors were identified in version 3.0; these errors have been corrected in later versions of the code. Some possible improvements in the code were also identified

  18. Methods of Software Verification

    Directory of Open Access Journals (Sweden)

    R. E. Gurin

    2015-01-01

    Full Text Available This article is devoted to the problem of software verification (SW. Methods of software verification designed to check the software for compliance with the stated requirements such as correctness, system security and system adaptability to small changes in the environment, portability and compatibility, etc. These are various methods both by the operation process and by the way of achieving result. The article describes the static and dynamic methods of software verification and paid attention to the method of symbolic execution. In its review of static analysis are discussed and described the deductive method, and methods for testing the model. A relevant issue of the pros and cons of a particular method is emphasized. The article considers classification of test techniques for each method. In this paper we present and analyze the characteristics and mechanisms of the static analysis of dependencies, as well as their views, which can reduce the number of false positives in situations where the current state of the program combines two or more states obtained both in different paths of execution and in working with multiple object values. Dependences connect various types of software objects: single variables, the elements of composite variables (structure fields, array elements, the size of the heap areas, the length of lines, the number of initialized array elements in the verification code using static methods. The article pays attention to the identification of dependencies within the framework of the abstract interpretation, as well as gives an overview and analysis of the inference tools.Methods of dynamic analysis such as testing, monitoring and profiling are presented and analyzed. Also some kinds of tools are considered which can be applied to the software when using the methods of dynamic analysis. Based on the work a conclusion is drawn, which describes the most relevant problems of analysis techniques, methods of their solutions and

  19. MARS software package status

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  20. User friendly packaging

    DEFF Research Database (Denmark)

    Geert Jensen, Birgitte

    2010-01-01

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

  1. The ZOOM minimization package

    International Nuclear Information System (INIS)

    Fischler, Mark S.; Sachs, D.

    2004-01-01

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

  2. Plasma physics plotting package

    International Nuclear Information System (INIS)

    Hyman, D.H.

    1981-02-01

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

  3. Active food packaging technologies.

    Science.gov (United States)

    Ozdemir, Murat; Floros, John D

    2004-01-01

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

  4. Quantum money with classical verification

    Energy Technology Data Exchange (ETDEWEB)

    Gavinsky, Dmitry [NEC Laboratories America, Princeton, NJ (United States)

    2014-12-04

    We propose and construct a quantum money scheme that allows verification through classical communication with a bank. This is the first demonstration that a secure quantum money scheme exists that does not require quantum communication for coin verification. Our scheme is secure against adaptive adversaries - this property is not directly related to the possibility of classical verification, nevertheless none of the earlier quantum money constructions is known to possess it.

  5. Quantum money with classical verification

    International Nuclear Information System (INIS)

    Gavinsky, Dmitry

    2014-01-01

    We propose and construct a quantum money scheme that allows verification through classical communication with a bank. This is the first demonstration that a secure quantum money scheme exists that does not require quantum communication for coin verification. Our scheme is secure against adaptive adversaries - this property is not directly related to the possibility of classical verification, nevertheless none of the earlier quantum money constructions is known to possess it

  6. CYPROS - Cybernetic Program Packages

    Directory of Open Access Journals (Sweden)

    Arne Tyssø

    1980-10-01

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

  7. User friendly packaging

    DEFF Research Database (Denmark)

    Geert Jensen, Birgitte

    2010-01-01

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

  8. WASTE PACKAGE TRANSPORTER DESIGN

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  9. WASTE PACKAGE TRANSPORTER DESIGN

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-09-23

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

  10. The BINSYN Program Package

    Directory of Open Access Journals (Sweden)

    Albert P. Linnell

    2012-06-01

    Full Text Available The BINSYN program package, recently expanded to calculate synthetic spectra of cataclysmic variables, is being further extended to include synthetic photometry of ordinary binary stars in addition to binary stars with optically thick accretion disks. The package includes a capability for differentials correction optimization of eclipsing binary systems using synthetic photometry.

  11. The LCDROOT Analysis Package

    International Nuclear Information System (INIS)

    Abe, Toshinori

    2001-01-01

    The North American Linear Collider Detector group has developed simulation and analysis program packages. LCDROOT is one of the packages, and is based on ROOT and the C++ programing language to maximally benefit from object oriented programming techniques. LCDROOT is constantly improved and now has a new topological vertex finder, ZVTOP3. In this proceeding, the features of the LCDROOT simulation are briefly described

  12. Grooming. Learning Activity Package.

    Science.gov (United States)

    Stark, Pamela

    This learning activity package on grooming for health workers is one of a series of 12 titles developed for use in health occupations education programs. Materials in the package include objectives, a list of materials needed, information sheets, reviews (self evaluations) of portions of the content, and answers to reviews. These topics are…

  13. Scalable Techniques for Formal Verification

    CERN Document Server

    Ray, Sandip

    2010-01-01

    This book presents state-of-the-art approaches to formal verification techniques to seamlessly integrate different formal verification methods within a single logical foundation. It should benefit researchers and practitioners looking to get a broad overview of the spectrum of formal verification techniques, as well as approaches to combining such techniques within a single framework. Coverage includes a range of case studies showing how such combination is fruitful in developing a scalable verification methodology for industrial designs. This book outlines both theoretical and practical issue

  14. A Practitioners Perspective on Verification

    Science.gov (United States)

    Steenburgh, R. A.

    2017-12-01

    NOAAs Space Weather Prediction Center offers a wide range of products and services to meet the needs of an equally wide range of customers. A robust verification program is essential to the informed use of model guidance and other tools by both forecasters and end users alike. In this talk, we present current SWPC practices and results, and examine emerging requirements and potential approaches to satisfy them. We explore the varying verification needs of forecasters and end users, as well as the role of subjective and objective verification. Finally, we describe a vehicle used in the meteorological community to unify approaches to model verification and facilitate intercomparison.

  15. Development of waste packages for tuff

    International Nuclear Information System (INIS)

    Rothman, A.J.

    1982-01-01

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

  16. Verification and validation of control system software

    International Nuclear Information System (INIS)

    Munro, J.K. Jr.; Kisner, R.A.; Bhadtt, S.C.

    1991-01-01

    The following guidelines are proposed for verification and validation (V ampersand V) of nuclear power plant control system software: (a) use risk management to decide what and how much V ampersand V is needed; (b) classify each software application using a scheme that reflects what type and how much V ampersand V is needed; (c) maintain a set of reference documents with current information about each application; (d) use Program Inspection as the initial basic verification method; and (e) establish a deficiencies log for each software application. The following additional practices are strongly recommended: (a) use a computer-based configuration management system to track all aspects of development and maintenance; (b) establish reference baselines of the software, associated reference documents, and development tools at regular intervals during development; (c) use object-oriented design and programming to promote greater software reliability and reuse; (d) provide a copy of the software development environment as part of the package of deliverables; and (e) initiate an effort to use formal methods for preparation of Technical Specifications. The paper provides background information and reasons for the guidelines and recommendations. 3 figs., 3 tabs

  17. Waste package performance analysis

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  18. RH Packaging Operations Manual

    International Nuclear Information System (INIS)

    Washington TRU Solutions LLC

    2003-01-01

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

  19. Advanced flip chip packaging

    CERN Document Server

    Lai, Yi-Shao; Wong, CP

    2013-01-01

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

  20. High-level verification

    CERN Document Server

    Lerner, Sorin; Kundu, Sudipta

    2011-01-01

    Given the growing size and heterogeneity of Systems on Chip (SOC), the design process from initial specification to chip fabrication has become increasingly complex. This growing complexity provides incentive for designers to use high-level languages such as C, SystemC, and SystemVerilog for system-level design. While a major goal of these high-level languages is to enable verification at a higher level of abstraction, allowing early exploration of system-level designs, the focus so far for validation purposes has been on traditional testing techniques such as random testing and scenario-based

  1. Survey on Offline Finger Print Verification System

    NARCIS (Netherlands)

    Suman, R.; Kaur, R.

    2012-01-01

    The fingerprint verification, means where "verification" implies a user matching a fingerprint against a single fingerprint associated with the identity that the user claims. Biometrics can be classified into two types Behavioral (signature verification, keystroke dynamics, etc.) And Physiological

  2. Development and verification of a reciprocating test rig designed for investigation of piston ring tribology

    DEFF Research Database (Denmark)

    Pedersen, Michael Torben; Imran, Tajammal; Klit, Peder

    2009-01-01

    This paper describes the development and verification of a reciprocating test rig, which was designed to study the piston ring tribology. A crank mechanism is used to generate a reciprocating motion for a moving plate, which acts as the liner. A stationary block acting as the ring package is loaded......, which is suitable for the study of piston ring tribology....

  3. Packaging Printing Today

    Directory of Open Access Journals (Sweden)

    Stanislav Bolanča

    2015-12-01

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

  4. Consumer response to packaging design

    NARCIS (Netherlands)

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

    2017-01-01

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

  5. Packaging Concerns/Techniques for Large Devices

    Science.gov (United States)

    Sampson, Michael J.

    2009-01-01

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

  6. Hermeticity of electronic packages

    CERN Document Server

    Greenhouse, Hal; Romenesco, Bruce

    2011-01-01

    This is a book about the integrity of sealed packages to resist foreign gases and liquids penetrating the seal or an opening (crack) in the packageùespecially critical to the reliability and longevity of electronics. The author explains how to predict the reliability and the longevity of the packages based on leak rate measurements and the assumptions of impurities. Non-specialists in particular will benefit from the author's long involvement in the technology. Hermeticity is a subject that demands practical experience, and solving one problem does not necessarily give one the background to so

  7. Hermeticity of electronic packages

    CERN Document Server

    Greenhouse, Hal

    2000-01-01

    This is a book about the integrity of sealed packages to resist foreign gases and liquids penetrating the seal or an opening (crack) in the package-especially critical to the reliability and longevity of electronics. The author explains how to predict the reliability and the longevity of the packages based on leak rate measurements and the assumptions of impurities. Non-specialists in particular will benefit from the author's long involvement in the technology. Hermeticity is a subject that demands practical experience, and solving one problem does not necessarily give one the background to so

  8. HDL to verification logic translator

    Science.gov (United States)

    Gambles, J. W.; Windley, P. J.

    1992-01-01

    The increasingly higher number of transistors possible in VLSI circuits compounds the difficulty in insuring correct designs. As the number of possible test cases required to exhaustively simulate a circuit design explodes, a better method is required to confirm the absence of design faults. Formal verification methods provide a way to prove, using logic, that a circuit structure correctly implements its specification. Before verification is accepted by VLSI design engineers, the stand alone verification tools that are in use in the research community must be integrated with the CAD tools used by the designers. One problem facing the acceptance of formal verification into circuit design methodology is that the structural circuit descriptions used by the designers are not appropriate for verification work and those required for verification lack some of the features needed for design. We offer a solution to this dilemma: an automatic translation from the designers' HDL models into definitions for the higher-ordered logic (HOL) verification system. The translated definitions become the low level basis of circuit verification which in turn increases the designer's confidence in the correctness of higher level behavioral models.

  9. Cleanup Verification Package for the 118-H-6:2, 105-H Reactor Ancillary Support Areas, Below-Grade Structures, and Underlying Soils; the 118-H-6:3, 105-H Reactor Fuel Storage Basin and Underlying Soils; the 118-H-6:6 Fuel Storage Basin Deep Zone Side Slope Soils; the 100-H-9, 100-H-10, and 100-H-13 French Drains; the 100-H-11 and 100-H-12 Expansion Box French Drains; and the 100-H-14 and 100-H-31 Surface Contamination Zones

    International Nuclear Information System (INIS)

    Appel, M.J.

    2006-01-01

    This cleanup verification package documents completion of removal actions for the 105-H Reactor Ancillary Support Areas, Below-Grade Structures, and Underlying Soils (subsite 118-H-6:2); 105-H Reactor Fuel Storage Basin and Underlying Soils (118-H-6:3); and Fuel Storage Basin Deep Zone Side Slope Soils. This CVP also documents remedial actions for the following seven additional waste sties: French Drain C (100-H-9), French Drain D (100-H-10), Expansion Box French Drain E (100-H-11), Expansion Box French Drain F (100-H-12), French Drain G (100-H-13), Surface Contamination Zone H (100-H-14), and the Polychlorinated Biphenyl Surface Contamination Zone (100-H-31)

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

    International Nuclear Information System (INIS)

    Blanton, P.

    2000-01-01

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

  11. Radiochemical verification and validation in the environmental data collection process

    International Nuclear Information System (INIS)

    Rosano-Reece, D.; Bottrell, D.; Bath, R.J.

    1994-01-01

    A credible and cost effective environmental data collection process should produce analytical data which meets regulatory and program specific requirements. Analytical data, which support the sampling and analysis activities at hazardous waste sites, undergo verification and independent validation before the data are submitted to regulators. Understanding the difference between verification and validation and their respective roles in the sampling and analysis process is critical to the effectiveness of a program. Verification is deciding whether the measurement data obtained are what was requested. The verification process determines whether all the requirements were met. Validation is more complicated than verification. It attempts to assess the impacts on data use, especially when requirements are not met. Validation becomes part of the decision-making process. Radiochemical data consists of a sample result with an associated error. Therefore, radiochemical validation is different and more quantitative than is currently possible for the validation of hazardous chemical data. Radiochemical data include both results and uncertainty that can be statistically compared to identify significance of differences in a more technically defensible manner. Radiochemical validation makes decisions about analyte identification, detection, and uncertainty for a batch of data. The process focuses on the variability of the data in the context of the decision to be made. The objectives of this paper are to present radiochemical verification and validation for environmental data and to distinguish the differences between the two operations

  12. High-activity liquid packaging design criteria

    International Nuclear Information System (INIS)

    1994-05-01

    In recent studies, it has been acknowledged that there is an emerging need for packaging to transport high-activity liquid off the Hanford Site to support characterization and process development activities of liquid waste stored in underground tanks. These studies have dealt with specimen testing needs primarily at the Hanford Site; however, similar needs appear to be developing at other US Department of Energy (DOE) sites. The need to ship single and multiple specimens to offsite laboratories is anticipated because it is predicted that onsite laboratories will be overwhelmed by an increasing number and size (volume) of samples. Potentially, the specimen size could range from 250 mL to greater than 50 L. Presently, no certified Type-B packagings are available for transport of high-activity liquid radioactive specimens in sizes to support Site missions

  13. Solar Water-Heater Design Package

    Science.gov (United States)

    1982-01-01

    Information on a solar domestic-hot water heater is contained in 146 page design package. System consists of solar collector, storage tanks, automatic control circuitry and auxiliary heater. Data-acquisition equipment at sites monitors day-by-day performance. Includes performance specifications, schematics, solar-collector drawings and drawings of control parts.

  14. Packaging development needs to support environmental restoration

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  15. EQ6 Calculations for Chemical Degradation of Navy Waste Packages

    International Nuclear Information System (INIS)

    S. LeStrange

    1999-01-01

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

  16. EQ6 Calculations for Chemical Degradation of Navy Waste Packages

    Energy Technology Data Exchange (ETDEWEB)

    S. LeStrange

    1999-11-15

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

  17. London 2012 packaging guidelines

    OpenAIRE

    2013-01-01

    These guidelines are intended to provide supplemental advice to suppliers and licensees regarding the provisions of the LOCOG Sustainable Sourcing Code that relate to packaging design and materials selection.

  18. Type B Drum packages

    International Nuclear Information System (INIS)

    Edwards, W.S.

    1995-11-01

    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

  19. Materials for advanced packaging

    CERN Document Server

    Wong, CP

    2017-01-01

    This second edition continues to be the most comprehensive review on the developments in advanced electronic packaging technologies, with a focus on materials and processing. Recognized experts in the field contribute to 22 updated and new chapters that provide comprehensive coverage on various 3D package architectures, novel bonding and joining techniques, wire bonding, wafer thinning techniques, organic substrates, and novel approaches to make electrical interconnects between integrated circuit and substrates. Various chapters also address advances in several key packaging materials, including: Lead-free solders Flip chip underfills Epoxy molding compounds Conductive adhesives Die attach adhesives/films Thermal interface materials (TIMS) Materials for fabricating embedded passives including capacitors, inductors, and resistors Materials and processing aspects on wafer-level chip scale package (CSP) and MicroElectroMechanical system (MEMS) Contributors also review new and emerging technologies such as Light ...

  20. Food irradiation and packaging

    International Nuclear Information System (INIS)

    Kilcast, David

    1988-01-01

    This outline review was written for 'Food Manufacture'. It deals with the known effects of irradiation on current packaging materials (glass, cellulosics, organic polymers and metals), and their implications for the effective application of the process. (U.K.)

  1. FLEXIBLE FOOD PACKAGING LABORATORY

    Data.gov (United States)

    Federal Laboratory Consortium — This laboratory contains equipment to fabricate and test prototype packages of many types and sizes (e.g., bags, pouches, trays, cartons, etc.). This equipment can...

  2. BCRA R Package

    Science.gov (United States)

    BCRA is an R package that projects absolute risk of invasive breast cancer according to NCI’s Breast Cancer Risk Assessment Tool (BCRAT) algorithm for specified race/ethnic groups and age intervals.

  3. The ENSDF Java Package

    International Nuclear Information System (INIS)

    Sonzogni, A.A.

    2005-01-01

    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

  4. Energy and packaging

    Energy Technology Data Exchange (ETDEWEB)

    Boustead, I; Hancock, G F

    1981-01-01

    Information is given on the energy and raw materials required in the production and use of containers used to package beer, cider, and carbonated soft drinks in the United Kingdom. Topics covered include: methodology of energy analysis, primary and secondary fuels, transport, packaging materials, including glass, aluminum, iron, steel, and tinplate, container production, including plastic bottles, distribution of empty containers, filling and packing, distribution, and disposal. (LEW)

  5. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    Washington TRU Solutions LLC

    2003-01-01

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

  6. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    Washington TRU Solutions LLC

    2002-01-01

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

  7. Verification of Simulation Tools

    International Nuclear Information System (INIS)

    Richard, Thierry

    2015-01-01

    Before qualifying a simulation tool, the requirements shall first be clearly identified, i.e.: - What type of study needs to be carried out? - What phenomena need to be modeled? This phase involves writing a precise technical specification. Once the requirements are defined, the most adapted product shall be selected from the various software options available on the market. Before using a particular version of a simulation tool to support the demonstration of nuclear safety studies, the following requirements shall be met. - An auditable quality assurance process complying with development international standards shall be developed and maintained, - A process of verification and validation (V and V) shall be implemented. This approach requires: writing a report and/or executive summary of the V and V activities, defining a validated domain (domain in which the difference between the results of the tools and those of another qualified reference is considered satisfactory for its intended use). - Sufficient documentation shall be available, - A detailed and formal description of the product (software version number, user configuration, other settings and parameters) in the targeted computing environment shall be available. - Source codes corresponding to the software shall be archived appropriately. When these requirements are fulfilled, the version of the simulation tool shall be considered qualified for a defined domain of validity, in a given computing environment. The functional verification shall ensure that: - the computer architecture of the tool does not include errors, - the numerical solver correctly represents the physical mathematical model, - equations are solved correctly. The functional verification can be demonstrated through certification or report of Quality Assurance. The functional validation shall allow the user to ensure that the equations correctly represent the physical phenomena in the perimeter of intended use. The functional validation can

  8. A Roadmap for the Implementation of Continued Process Verification.

    Science.gov (United States)

    Boyer, Marcus; Gampfer, Joerg; Zamamiri, Abdel; Payne, Robin

    2016-01-01

    In 2014, the members of the BioPhorum Operations Group (BPOG) produced a 100-page continued process verification case study, entitled "Continued Process Verification: An Industry Position Paper with Example Protocol". This case study captures the thought processes involved in creating a continued process verification plan for a new product in response to the U.S. Food and Drug Administration's guidance on the subject introduced in 2011. In so doing, it provided the specific example of a plan developed for a new molecular antibody product based on the "A MAb Case Study" that preceded it in 2009.This document provides a roadmap that draws on the content of the continued process verification case study to provide a step-by-step guide in a more accessible form, with reference to a process map of the product life cycle. It could be used as a basis for continued process verification implementation in a number of different scenarios: For a single product and process;For a single site;To assist in the sharing of data monitoring responsibilities among sites;To assist in establishing data monitoring agreements between a customer company and a contract manufacturing organization. The U.S. Food and Drug Administration issued guidance on the management of manufacturing processes designed to improve quality and control of drug products. This involved increased focus on regular monitoring of manufacturing processes, reporting of the results, and the taking of opportunities to improve. The guidance and practice associated with it is known as continued process verification This paper summarizes good practice in responding to continued process verification guidance, gathered from subject matter experts in the biopharmaceutical industry. © PDA, Inc. 2016.

  9. Comparative Packaging Study

    Science.gov (United States)

    Perchonok, Michele H.; Oziomek, Thomas V.

    2009-01-01

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

  10. Burnup verification using the FORK measurement system

    International Nuclear Information System (INIS)

    Ewing, R.I.

    1994-01-01

    Verification measurements may be used to help ensure nuclear criticality safety when burnup credit is applied to spent fuel transport and storage systems. The FORK measurement system, designed at Los Alamos National Laboratory for the International Atomic Energy Agency safeguards program, has been used to verify reactor site records for burnup and cooling time for many years. The FORK system measures the passive neutron and gamma-ray emission from spent fuel assemblies while in the storage pool. This report deals with the application of the FORK system to burnup credit operations based on measurements performed on spent fuel assemblies at the Oconee Nuclear Station of Duke Power Company

  11. Shift Verification and Validation

    Energy Technology Data Exchange (ETDEWEB)

    Pandya, Tara M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Evans, Thomas M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Davidson, Gregory G [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Johnson, Seth R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Godfrey, Andrew T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-07

    This documentation outlines the verification and validation of Shift for the Consortium for Advanced Simulation of Light Water Reactors (CASL). Five main types of problems were used for validation: small criticality benchmark problems; full-core reactor benchmarks for light water reactors; fixed-source coupled neutron-photon dosimetry benchmarks; depletion/burnup benchmarks; and full-core reactor performance benchmarks. We compared Shift results to measured data and other simulated Monte Carlo radiation transport code results, and found very good agreement in a variety of comparison measures. These include prediction of critical eigenvalue, radial and axial pin power distributions, rod worth, leakage spectra, and nuclide inventories over a burn cycle. Based on this validation of Shift, we are confident in Shift to provide reference results for CASL benchmarking.

  12. Online fingerprint verification.

    Science.gov (United States)

    Upendra, K; Singh, S; Kumar, V; Verma, H K

    2007-01-01

    As organizations search for more secure authentication methods for user access, e-commerce, and other security applications, biometrics is gaining increasing attention. With an increasing emphasis on the emerging automatic personal identification applications, fingerprint based identification is becoming more popular. The most widely used fingerprint representation is the minutiae based representation. The main drawback with this representation is that it does not utilize a significant component of the rich discriminatory information available in the fingerprints. Local ridge structures cannot be completely characterized by minutiae. Also, it is difficult quickly to match two fingerprint images containing different number of unregistered minutiae points. In this study filter bank based representation, which eliminates these weakness, is implemented and the overall performance of the developed system is tested. The results have shown that this system can be used effectively for secure online verification applications.

  13. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    2008-01-01

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

  14. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    2009-01-01

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

  15. RH Packaging Program Guidance

    International Nuclear Information System (INIS)

    2006-01-01

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

  16. RH Packaging Program Guidance

    International Nuclear Information System (INIS)

    2008-01-01

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

  17. A restructuring of CF package for MIDAS computer code

    International Nuclear Information System (INIS)

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

    2004-01-01

    CF package, which evaluates user-specified 'control functions' and applies them to define or control various aspects of computation, has been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and modernized data structure. To do this, data transferring methods of current MELCOR code are modified and adopted into the CF package. The data structure of the current MELCOR code using FORTRAN77 causes a difficult grasping of meaning of the variables as well as waste of memory, difficulty is more over because its data is location information of other package's data due to characteristics of CF package. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which lead to an efficient memory treatment and an easy understanding of the code. Restructuring of the CF package addressed in this paper includes module development, subroutine modification, and treats MELGEN, which generates data file, as well as MELCOR, which is processing a calculation. The verification has been done by comparing the results of the modified code with those from the existing code. As the trends are similar to each other, it hints that the same approach could be extended to the entire code package. It is expected that code restructuring will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models

  18. Hypervelocity Wind Tunnel No. 9 Mach 7 Thermal Structural Facility Verification and Calibration

    National Research Council Canada - National Science Library

    Lafferty, John

    1996-01-01

    This report summarizes the verification and calibration of the new Mach 7 Thermal Structural Facility located at the White Oak, Maryland, site of the Dahlgren Division, Naval Surface Warfare Center...

  19. Independent Verification Survey Report for the Operable Unit-1 Landfill Miamisburg Closure Project, Miamisburg, OH

    International Nuclear Information System (INIS)

    Weaver, P.

    2008-01-01

    The objectives of the independent verification survey were to confirm that remedial actions have been effective in meeting established release criteria and that documentation accurately and adequately describes the current radiological and chemical conditions of the MCP site

  20. Application of GA package in functional packaging

    Science.gov (United States)

    Belousova, D. A.; Noskova, E. E.; Kapulin, D. V.

    2018-05-01

    The approach to application program for the task of configuration of the elements of the commutation circuit for design of the radio-electronic equipment on the basis of the genetic algorithm is offered. The efficiency of the used approach for commutation circuits with different characteristics for computer-aided design on radio-electronic manufacturing is shown. The prototype of the computer-aided design subsystem on the basis of a package GA for R with a set of the general functions for optimization of multivariate models is programmed.

  1. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    2005-01-01

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

  2. Food Packaging Materials

    Science.gov (United States)

    1978-01-01

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

  3. Improved verification methods for safeguards verifications at enrichment plants

    International Nuclear Information System (INIS)

    Lebrun, A.; Kane, S. C.; Bourva, L.; Poirier, S.; Loghin, N. E.; Langlands, D.

    2009-01-01

    The International Atomic Energy Agency (IAEA) has initiated a coordinated research and development programme to improve its verification methods and equipment applicable to enrichment plants. The programme entails several individual projects to meet the objectives of the IAEA Safeguards Model Approach for Gas Centrifuge Enrichment Plants updated in 2006. Upgrades of verification methods to confirm the absence of HEU (highly enriched uranium) production have been initiated and, in particular, the Cascade Header Enrichment Monitor (CHEM) has been redesigned to reduce its weight and incorporate an electrically cooled germanium detector. Such detectors are also introduced to improve the attended verification of UF 6 cylinders for the verification of the material balance. Data sharing of authenticated operator weighing systems such as accountancy scales and process load cells is also investigated as a cost efficient and an effective safeguards measure combined with unannounced inspections, surveillance and non-destructive assay (NDA) measurement. (authors)

  4. Improved verification methods for safeguards verifications at enrichment plants

    Energy Technology Data Exchange (ETDEWEB)

    Lebrun, A.; Kane, S. C.; Bourva, L.; Poirier, S.; Loghin, N. E.; Langlands, D. [Department of Safeguards, International Atomic Energy Agency, Wagramer Strasse 5, A1400 Vienna (Austria)

    2009-07-01

    The International Atomic Energy Agency (IAEA) has initiated a coordinated research and development programme to improve its verification methods and equipment applicable to enrichment plants. The programme entails several individual projects to meet the objectives of the IAEA Safeguards Model Approach for Gas Centrifuge Enrichment Plants updated in 2006. Upgrades of verification methods to confirm the absence of HEU (highly enriched uranium) production have been initiated and, in particular, the Cascade Header Enrichment Monitor (CHEM) has been redesigned to reduce its weight and incorporate an electrically cooled germanium detector. Such detectors are also introduced to improve the attended verification of UF{sub 6} cylinders for the verification of the material balance. Data sharing of authenticated operator weighing systems such as accountancy scales and process load cells is also investigated as a cost efficient and an effective safeguards measure combined with unannounced inspections, surveillance and non-destructive assay (NDA) measurement. (authors)

  5. RNA secondary structures of the bacteriophage phi6 packaging regions.

    Science.gov (United States)

    Pirttimaa, M J; Bamford, D H

    2000-06-01

    Bacteriophage phi6 genome consists of three segments of double-stranded RNA. During maturation, single-stranded copies of these segments are packaged into preformed polymerase complex particles. Only phi6 RNA is packaged, and each particle contains only one copy of each segment. An in vitro packaging and replication assay has been developed for phi6, and the packaging signals (pac sites) have been mapped to the 5' ends of the RNA segments. In this study, we propose secondary structure models for the pac sites of phi6 single-stranded RNA segments. Our models accommodate data from structure-specific chemical modifications, free energy minimizations, and phylogenetic comparisons. Previously reported pac site deletion studies are also discussed. Each pac site possesses a unique architecture, that, however, contains common structural elements.

  6. Universal storage/transport/disposal packages

    International Nuclear Information System (INIS)

    Smith, M.L.

    1992-01-01

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

  7. Safety evaluation for packaging (onsite) SERF cask

    International Nuclear Information System (INIS)

    Edwards, W.S.

    1997-01-01

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

  8. Food packages for Space Shuttle

    Science.gov (United States)

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

    1978-01-01

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

  9. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    2006-01-01

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

  10. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    2007-01-01

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

  11. Welding robot package; Arc yosetsu robot package

    Energy Technology Data Exchange (ETDEWEB)

    Nishikawa, S. [Yaskawa Electric Corp., Kitakyushu (Japan)

    1998-09-01

    For the conventional high-speed welding robot, the welding current was controlled mainly for reducing the spatters during short circuits and for stabilizing the beads by the periodic short circuits. However, an increase of deposition amount in response to the speed is required for the high-speed welding. Large-current low-spatter welding current region control was added. Units were integrated into a package by which the arc length is kept in short without dispersion of arc length for welding without defects such as undercut and unequal beads. In automobile industry, use of aluminum parts is extended for the light weight. The welding is very difficult, and automation is not so progressing in spite of the poor environment. Buckling of welding wire is easy to occur, and supply of wire is obstructed by the deposition of chipped powders on the torch cable, which stay within the contact chip resulting in the deposition. Dislocation of locus is easy to occur at the corner of rectangular pipe during the welding. By improving these troubles, an aluminum MIG welding robot package has been developed. 13 figs.

  12. Tomotherapy: IMRT and tomographic verification

    International Nuclear Information System (INIS)

    Mackie, T.R.

    2000-01-01

    include MLC's and many clinics use them to replace 90% or more of the field-shaping requirements of conventional radiotherapy. Now, several academic centers are treating patients with IMRT using conventional MLC's to modulate the field. IMRT using conventional MLC's have the advantage that the patient is stationary during the treatment and the MLC's can be used in conventional practice. Nevertheless, tomotherapy using the Peacock system delivers the most conformal dose distributions of any commercial system to date. The biggest limitation with the both the NOMOS Peacock tomotherapy system and conventional MLC's for IMRT delivery is the lack of treatment verification. In conventional few-field radiotherapy one relied on portal images to determine if the patient was setup correctly and the beams were correctly positioned. With IMRT the image contrast is superimposed on the beam intensity variation. Conventional practice allowed for monitor unit calculation checks and point dosimeters placed on the patient's surface to verify that the treatment was properly delivered. With IMRT it is impossible to perform hand calculations of monitor units and dosimeters placed on the patient's surface are prone to error due to high gradients in the beam intensity. NOMOS has developed a verification phantom that allows multiple sheets of film to be placed in a light-tight box that is irradiated with the same beam pattern that is used to treat the patient. The optical density of the films are adjusted, normalized, and calibrated and then quantitatively compared with the dose calculated for the phantom delivery. However, this process is too laborious to be used for patient-specific QA. If IMRT becomes ubiquitous and it can be shown that IMRT is useful on most treatment sites then there is a need to design treatment units dedicated to IMRT delivery and verification. Helical tomotherapy is such a redesign. Helical tomotherapy is the delivery of a rotational fan beam while the patient is

  13. Reactor physics verification of the MCNP6 unstructured mesh capability

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  14. Reactor physics verification of the MCNP6 unstructured mesh capability

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  15. The joint verification experiments as a global non-proliferation exercise

    International Nuclear Information System (INIS)

    Shaner, J.W.

    1998-01-01

    This conference commemorates the 10th anniversary of the second of two Joint Verification Experiments conducted by the Soviet Union and the US. These two experiments, one at the Nevada test site in the US, and the second here at the Semipalatinsk test site were designed to test the verification of a nuclear testing treaty limiting the size underground explosions to 150 kilotons. By building trust and technical respect between the weapons scientists of the two most powerful adversaries, the Joint Verification Experiment (JVE) had the unanticipated result of initiating a suite of cooperative projects and programs aimed at reducing the Cold War threats and preventing the proliferation of weapons of mass destruction

  16. Numident Online Verification Utility (NOVU)

    Data.gov (United States)

    Social Security Administration — NOVU is a mainframe application that accesses the NUMIDENT to perform real-time SSN verifications. This program is called by other SSA online programs that serve as...

  17. Technical safety requirements control level verification

    International Nuclear Information System (INIS)

    STEWART, J.L.

    1999-01-01

    A Technical Safety Requirement (TSR) control level verification process was developed for the Tank Waste Remediation System (TWRS) TSRs at the Hanford Site in Richland, WA, at the direction of the US. Department of Energy, Richland Operations Office (RL). The objective of the effort was to develop a process to ensure that the TWRS TSR controls are designated and managed at the appropriate levels as Safety Limits (SLs), Limiting Control Settings (LCSs), Limiting Conditions for Operation (LCOs), Administrative Controls (ACs), or Design Features. The TSR control level verification process was developed and implemented by a team of contractor personnel with the participation of Fluor Daniel Hanford, Inc. (FDH), the Project Hanford Management Contract (PHMC) integrating contractor, and RL representatives. The team was composed of individuals with the following experience base: nuclear safety analysis; licensing; nuclear industry and DOE-complex TSR preparation/review experience; tank farm operations; FDH policy and compliance; and RL-TWRS oversight. Each TSR control level designation was completed utilizing TSR control logic diagrams and TSR criteria checklists based on DOE Orders, Standards, Contractor TSR policy, and other guidance. The control logic diagrams and criteria checklists were reviewed and modified by team members during team meetings. The TSR control level verification process was used to systematically evaluate 12 LCOs, 22 AC programs, and approximately 100 program key elements identified in the TWRS TSR document. The verification of each TSR control required a team consensus. Based on the results of the process, refinements were identified and the TWRS TSRs were modified as appropriate. A final report documenting key assumptions and the control level designation for each TSR control was prepared and is maintained on file for future reference. The results of the process were used as a reference in the RL review of the final TWRS TSRs and control suite. RL

  18. Technical safety requirements control level verification; TOPICAL

    International Nuclear Information System (INIS)

    STEWART, J.L.

    1999-01-01

    A Technical Safety Requirement (TSR) control level verification process was developed for the Tank Waste Remediation System (TWRS) TSRs at the Hanford Site in Richland, WA, at the direction of the US. Department of Energy, Richland Operations Office (RL). The objective of the effort was to develop a process to ensure that the TWRS TSR controls are designated and managed at the appropriate levels as Safety Limits (SLs), Limiting Control Settings (LCSs), Limiting Conditions for Operation (LCOs), Administrative Controls (ACs), or Design Features. The TSR control level verification process was developed and implemented by a team of contractor personnel with the participation of Fluor Daniel Hanford, Inc. (FDH), the Project Hanford Management Contract (PHMC) integrating contractor, and RL representatives. The team was composed of individuals with the following experience base: nuclear safety analysis; licensing; nuclear industry and DOE-complex TSR preparation/review experience; tank farm operations; FDH policy and compliance; and RL-TWRS oversight. Each TSR control level designation was completed utilizing TSR control logic diagrams and TSR criteria checklists based on DOE Orders, Standards, Contractor TSR policy, and other guidance. The control logic diagrams and criteria checklists were reviewed and modified by team members during team meetings. The TSR control level verification process was used to systematically evaluate 12 LCOs, 22 AC programs, and approximately 100 program key elements identified in the TWRS TSR document. The verification of each TSR control required a team consensus. Based on the results of the process, refinements were identified and the TWRS TSRs were modified as appropriate. A final report documenting key assumptions and the control level designation for each TSR control was prepared and is maintained on file for future reference. The results of the process were used as a reference in the RL review of the final TWRS TSRs and control suite. RL

  19. Waste package characterisation

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-09-01

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

  20. Waste package characterisation

    International Nuclear Information System (INIS)

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

    1998-09-01

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

  1. Waste package reliability analysis

    International Nuclear Information System (INIS)

    Pescatore, C.; Sastre, C.

    1983-01-01

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

  2. Subsurface barrier verification technologies, informal report

    International Nuclear Information System (INIS)

    Heiser, J.H.

    1994-06-01

    One of the more promising remediation options available to the DOE waste management community is subsurface barriers. Some of the uses of subsurface barriers include surrounding and/or containing buried waste, as secondary confinement of underground storage tanks, to direct or contain subsurface contaminant plumes and to restrict remediation methods, such as vacuum extraction, to a limited area. To be most effective the barriers should be continuous and depending on use, have few or no breaches. A breach may be formed through numerous pathways including: discontinuous grout application, from joints between panels and from cracking due to grout curing or wet-dry cycling. The ability to verify barrier integrity is valuable to the DOE, EPA, and commercial sector and will be required to gain full public acceptance of subsurface barriers as either primary or secondary confinement at waste sites. It is recognized that no suitable method exists for the verification of an emplaced barrier's integrity. The large size and deep placement of subsurface barriers makes detection of leaks challenging. This becomes magnified if the permissible leakage from the site is low. Detection of small cracks (fractions of an inch) at depths of 100 feet or more has not been possible using existing surface geophysical techniques. Compounding the problem of locating flaws in a barrier is the fact that no placement technology can guarantee the completeness or integrity of the emplaced barrier. This report summarizes several commonly used or promising technologies that have been or may be applied to in-situ barrier continuity verification

  3. SPHINX experimenters information package

    International Nuclear Information System (INIS)

    Zarick, T.A.

    1996-08-01

    This information package was prepared for both new and experienced users of the SPHINX (Short Pulse High Intensity Nanosecond X-radiator) flash X-Ray facility. It was compiled to help facilitate experiment design and preparation for both the experimenter(s) and the SPHINX operational staff. The major areas covered include: Recording Systems Capabilities,Recording System Cable Plant, Physical Dimensions of SPHINX and the SPHINX Test cell, SPHINX Operating Parameters and Modes, Dose Rate Map, Experiment Safety Approval Form, and a Feedback Questionnaire. This package will be updated as the SPHINX facilities and capabilities are enhanced

  4. Nirex - The complete package

    International Nuclear Information System (INIS)

    Carr, N.A.; Rarok, M.

    2002-01-01

    Through continuous interaction with its customers, Nirex had identified the need for a comprehensive range of waste containers, reflecting the variety of wastes and operational undertakings. The current range consists of five standard containers. Standardisation is introduced across all waste packages to enable the safe and efficient operation of future waste management facilities. The practical lessons learned during the development of standard containers are in turn reflected in the container design work that Nirex has undertaken. They are also fed into the advice given to customers during evaluation of waste packaging proposals. (author)

  5. Ada Namelist Package

    Science.gov (United States)

    Klumpp, Allan R.

    1991-01-01

    Ada Namelist Package, developed for Ada programming language, enables calling program to read and write FORTRAN-style namelist files. Features are: handling of any combination of types defined by user; ability to read vectors, matrices, and slices of vectors and matrices; handling of mismatches between variables in namelist file and those in programmed list of namelist variables; and ability to avoid searching entire input file for each variable. Principle benefits derived by user: ability to read and write namelist-readable files, ability to detect most file errors in initialization phase, and organization keeping number of instantiated units to few packages rather than to many subprograms.

  6. Organics Verification Study for Sinclair and Dyes Inlets, Washington

    Energy Technology Data Exchange (ETDEWEB)

    Kohn, Nancy P.; Brandenberger, Jill M.; Niewolny, Laurie A.; Johnston, Robert K.

    2006-09-28

    Sinclair and Dyes Inlets near Bremerton, Washington, are on the State of Washington 1998 303(d) list of impaired waters because of fecal coliform contamination in marine water, metals in sediment and fish tissue, and organics in sediment and fish tissue. Because significant cleanup and source control activities have been conducted in the inlets since the data supporting the 1998 303(d) listings were collected, two verification studies were performed to address the 303(d) segments that were listed for metal and organic contaminants in marine sediment. The Metals Verification Study (MVS) was conducted in 2003; the final report, Metals Verification Study for Sinclair and Dyes Inlets, Washington, was published in March 2004 (Kohn et al. 2004). This report describes the Organics Verification Study that was conducted in 2005. The study approach was similar to the MVS in that many surface sediment samples were screened for the major classes of organic contaminants, and then the screening results and other available data were used to select a subset of samples for quantitative chemical analysis. Because the MVS was designed to obtain representative data on concentrations of contaminants in surface sediment throughout Sinclair Inlet, Dyes Inlet, Port Orchard Passage, and Rich Passage, aliquots of the 160 MVS sediment samples were used in the analysis for the Organics Verification Study. However, unlike metals screening methods, organics screening methods are not specific to individual organic compounds, and are not available for some target organics. Therefore, only the quantitative analytical results were used in the organics verification evaluation. The results of the Organics Verification Study showed that sediment quality outside of Sinclair Inlet is unlikely to be impaired because of organic contaminants. Similar to the results for metals, in Sinclair Inlet, the distribution of residual organic contaminants is generally limited to nearshore areas already within the

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

    OpenAIRE

    Shan, Kun; Julius, Joezer

    2015-01-01

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

  8. Experimental inventory verification system

    International Nuclear Information System (INIS)

    Steverson, C.A.; Angerman, M.I.

    1991-01-01

    As Low As Reasonably Achievable (ALARA) goals and Department of Energy (DOE) inventory requirements are frequently in conflict at facilities across the DOE complex. The authors wish, on one hand, to verify the presence of correct amounts of nuclear materials that are in storage or in process; yet on the other hand, we wish to achieve ALARA goals by keeping individual and collective exposures as low as social, technical, economic, practical, and public policy considerations permit. The Experimental Inventory Verification System (EIVSystem) is a computer-based, camera-driven system that utilizes image processing technology to detect change in vault areas. Currently in the test and evaluation phase at Idaho National Engineering Laboratory, this system guards personnel. The EIVSystem continually monitors the vault, providing proof of changed status for objects sorted within the vault. This paper reports that these data could provide the basis for reducing inventory requirements when no change has occurred, thus helping implement ALARA policy; the data will also help describe there target area of an inventory when change has been shown to occur

  9. Woodward Effect Experimental Verifications

    Science.gov (United States)

    March, Paul

    2004-02-01

    The work of J. F. Woodward (1990 1996a; 1996b; 1998; 2002a; 2002b; 2004) on the existence of ``mass fluctuations'' and their use in exotic propulsion schemes was examined for possible application in improving space flight propulsion and power generation. Woodward examined Einstein's General Relativity Theory (GRT) and assumed that if the strong Machian interpretation of GRT as well as gravitational / inertia like Wheeler-Feynman radiation reaction forces hold, then when an elementary particle is accelerated through a potential gradient, its rest mass should fluctuate around its mean value during its acceleration. Woodward also used GRT to clarify the precise experimental conditions necessary for observing and exploiting these mass fluctuations or ``Woodward effect'' (W-E). Later, in collaboration with his ex-graduate student T. Mahood, they also pushed the experimental verification boundaries of these proposals. If these purported mass fluctuations occur as Woodward claims, and his assumption that gravity and inertia are both byproducts of the same GRT based phenomenon per Mach's Principle is correct, then many innovative applications such as propellantless propulsion and gravitational exotic matter generators may be feasible. This paper examines the reality of mass fluctuations and the feasibility of using the W-E to design propellantless propulsion devices in the near to mid-term future. The latest experimental results, utilizing MHD-like force rectification systems, will also be presented.

  10. Verification of hypergraph states

    Science.gov (United States)

    Morimae, Tomoyuki; Takeuchi, Yuki; Hayashi, Masahito

    2017-12-01

    Hypergraph states are generalizations of graph states where controlled-Z gates on edges are replaced with generalized controlled-Z gates on hyperedges. Hypergraph states have several advantages over graph states. For example, certain hypergraph states, such as the Union Jack states, are universal resource states for measurement-based quantum computing with only Pauli measurements, while graph state measurement-based quantum computing needs non-Clifford basis measurements. Furthermore, it is impossible to classically efficiently sample measurement results on hypergraph states unless the polynomial hierarchy collapses to the third level. Although several protocols have been proposed to verify graph states with only sequential single-qubit Pauli measurements, there was no verification method for hypergraph states. In this paper, we propose a method for verifying a certain class of hypergraph states with only sequential single-qubit Pauli measurements. Importantly, no i.i.d. property of samples is assumed in our protocol: any artificial entanglement among samples cannot fool the verifier. As applications of our protocol, we consider verified blind quantum computing with hypergraph states, and quantum computational supremacy demonstrations with hypergraph states.

  11. Packaging and transportation occurrence reporting

    International Nuclear Information System (INIS)

    Needels, T.S.

    1996-01-01

    The US Department of Energy (DOE) Order 231.1 calls for the maintenance of a database for all unclassified occurrence reports (ORs). ORS provide DOE with notice of incidents and accidents that endanger the public, workers, or DOE facility operations. To fulfill this policy, the DOE Occurrence Reporting and Processing System (ORPS) was established to require DOE facilities to report and process information concerning such events. The Oak Ridge National Laboratory (ORNL) provides DOE with data and analysis of occurrence related to packaging and transportation (P and T) safety. This program produces annual reports, lessons learned bulletins, and information for the packaging and transportation home page on the Internet. The analysis and reports provided can be used as a tool for oversight and a means for DOE sites to be proactive and anticipate problems through shared knowledge and lessons learned. To illustrate, some observable trends based on 3 years of the program are given. In summary, this program shows potential problem areas that need correcting, and possible breakdowns of safety

  12. AN ADA NAMELIST PACKAGE

    Science.gov (United States)

    Klumpp, A. R.

    1994-01-01

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

  13. K Basins Spent Nuclear Fuel (SNF) Project Safety Analysis Report for Packaging (SARP) approval plan

    International Nuclear Information System (INIS)

    1995-01-01

    This document delineates the plan for preparation, review, and approval of the K Basins Spent Nuclear Fuel (SNF) Packaging Design Criteria (PDC) document and the on-site Safety Analysis Report for Packaging (SARP). The packaging addressed in these documents is used to transport SNF in a Multi- canister Overpack (MCO) configuration

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

    International Nuclear Information System (INIS)

    Komann, Steffen; Groeke, Carsten; Droste, Bernhard

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  16. Radioactive waste disposal package

    Science.gov (United States)

    Lampe, Robert F.

    1986-11-04

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

  17. ROBATEL RS 24 packaging

    International Nuclear Information System (INIS)

    Robatel, M.; Bochard, C.

    1986-01-01

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

  18. The Swarm Magnetometry Package

    DEFF Research Database (Denmark)

    Merayo, José M.G.; Jørgensen, John Leif; Friis-Christensen, Eigil

    2008-01-01

    The Swarm mission under the ESA's Living Planet Programme is planned for launch in 2010 and consists of a constellation of three satellites at LEO. The prime objective of Swarm is to measure the geomagnetic field with unprecedented accuracy in space and time. The magnetometry package consists...

  19. Learning Activity Package, Algebra.

    Science.gov (United States)

    Evans, Diane

    A set of ten teacher-prepared Learning Activity Packages (LAPs) in beginning algebra and nine in intermediate algebra, these units cover sets, properties of operations, number systems, open expressions, solution sets of equations and inequalities in one and two variables, exponents, factoring and polynomials, relations and functions, radicals,…

  20. Geothermal Greenhouse Information Package

    Energy Technology Data Exchange (ETDEWEB)

    Rafferty, K. [P.E.; Boyd, T. [ed.

    1997-01-01

    This package of information is intended to provide a foundation of background information for developers of geothermal greenhouses. The material is divided into seven sections covering such issues as crop culture and prices, operating costs for greenhouses, heating system design, vendors and a list of other sources of information.

  1. Openability of tamperproof packaging

    NARCIS (Netherlands)

    Del Castillo C., A.; Wever, R.; Buijs, P.J.; Stevels, A.

    2007-01-01

    Communication, product protection and presentation are three key aspects in the world of packaging nowadays. Due to a retail landscape consisting of large stores, displaying packed products on the shelves in self-service environments, these aspects become increasingly important, not only for Fast

  2. Packaging LLW and ILW

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  3. Aquaculture Information Package

    Energy Technology Data Exchange (ETDEWEB)

    Boyd, T.; Rafferty, K. [editors

    1998-01-01

    This package of information is intended to provide background 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.

  4. CH Packaging Maintenance Manual

    International Nuclear Information System (INIS)

    Washington TRU Solutions

    2002-01-01

    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

  5. Printer Graphics Package

    Science.gov (United States)

    Blanchard, D. C.

    1986-01-01

    Printer Graphics Package (PGP) is tool for making two-dimensional symbolic plots on line printer. PGP created to support development of Heads-Up Display (HUD) simulation. Standard symbols defined with HUD in mind. Available symbols include circle, triangle, quadrangle, window, line, numbers, and text. Additional symbols easily added or built up from available symbols.

  6. Type B drum packages

    International Nuclear Information System (INIS)

    McCoy, J.C.

    1994-08-01

    The Type B drum packages (TBD) are conceptualized as a family of containers in which a single 208 L or 114 L (55 gal or 30 gal) drum containing Type B quantities of radioactive material (RAM) can be packaged for shipment. The TBD containers are being developed to fill a void in the packaging and transportation capabilities of the U.S. Department of Energy as no container packaging single drums of Type B RAM exists offering double containment. Several multiple-drum containers currently exist, as well as a number of shielded casks, but the size and weight of these containers present many operational challenges for single-drum shipments. As an alternative, the TBD containers will offer up to three shielded versions (light, medium, and heavy) and one unshielded version, each offering single or optional double containment for a single drum. To reduce operational complexity, all versions will share similar design and operational features where possible. The primary users of the TBD containers are envisioned to be any organization desiring to ship single drums of Type B RAM, such as laboratories, waste retrieval activities, emergency response teams, etc. Currently, the TBD conceptual design is being developed with the final design and analysis to be completed in 1995 to 1996. Testing and certification of the unshielded version are planned to be completed in 1996 to 1997 with production to begin in 1997 to 1998

  7. Waste disposal package

    Science.gov (United States)

    Smith, M.J.

    1985-06-19

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

  8. EQ3/6 software test and verification report 9/94

    International Nuclear Information System (INIS)

    Kishi, T.

    1996-02-01

    This document is the Software Test and Verification Report (STVR) for the EQ3/6 suite of codes as stipulated in the Individual Software Plan for Initial Qualification of EQ3/6 (ISP-NF-07, Revision 1, 11/25/92). The software codes, EQPT, EQ3NR, EQ6, and the software library EQLIB constitute the EQ3/6 software package. This software test and verification project for EQ3/6 was started under the requirements of the LLNL Yucca Mountain Project Software Quality Assurance Plan (SQAP), Revision 0, December 14, 1989, but QP 3.2, Revision 2, June 21, 1994 is now the operative controlling procedure. This is a ''V and V'' report in the language of QP 3.2, Revision 2. Because the author of this report does not have a background in geochemistry, other technical sources were consulted in order to acquire some familiarity with geochemisty, the terminology minology involved, and to review comparable computational methods especially, geochemical aqueous speciation-solubility calculations. The software for the EQ3/6 package consists of approximately 47,000 lines of FORTRAN77 source code and nine on platforms ranging from workstations to supercomputers. The physical control of EQ3/6 software package and documentation is on a SUN SPARC station. Walkthroughs of each principal software packages, EQPT, EQ3NR, and EQ6 were conducted in order to understand the computational procedures involved, to determine any commonality in procedures, and then to establish a plan for the test and verification of EQ3/6. It became evident that all three phases depended upon solving an n x n matrix by the Newton-Raphson Method. Thus, a great deal of emphasis on the test and verification of this procedure was carried out on the first code in the software package EQPT

  9. EQ3/6 software test and verification report 9/94

    Energy Technology Data Exchange (ETDEWEB)

    Kishi, T.

    1996-02-01

    This document is the Software Test and Verification Report (STVR) for the EQ3/6 suite of codes as stipulated in the Individual Software Plan for Initial Qualification of EQ3/6 (ISP-NF-07, Revision 1, 11/25/92). The software codes, EQPT, EQ3NR, EQ6, and the software library EQLIB constitute the EQ3/6 software package. This software test and verification project for EQ3/6 was started under the requirements of the LLNL Yucca Mountain Project Software Quality Assurance Plan (SQAP), Revision 0, December 14, 1989, but QP 3.2, Revision 2, June 21, 1994 is now the operative controlling procedure. This is a ``V and V`` report in the language of QP 3.2, Revision 2. Because the author of this report does not have a background in geochemistry, other technical sources were consulted in order to acquire some familiarity with geochemisty, the terminology minology involved, and to review comparable computational methods especially, geochemical aqueous speciation-solubility calculations. The software for the EQ3/6 package consists of approximately 47,000 lines of FORTRAN77 source code and nine on platforms ranging from workstations to supercomputers. The physical control of EQ3/6 software package and documentation is on a SUN SPARC station. Walkthroughs of each principal software packages, EQPT, EQ3NR, and EQ6 were conducted in order to understand the computational procedures involved, to determine any commonality in procedures, and then to establish a plan for the test and verification of EQ3/6. It became evident that all three phases depended upon solving an n x n matrix by the Newton-Raphson Method. Thus, a great deal of emphasis on the test and verification of this procedure was carried out on the first code in the software package EQPT.

  10. Hazardous materials package performance regulations

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  11. A restructuring of RN1 package for MIDAS computer code

    International Nuclear Information System (INIS)

    Park, S. H.; Kim, D. H.; Kim, K. R.

    2003-01-01

    RN1 package, which is one of two fission product-related packages in MELCOR, has been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and modernized data structure. To do this, data transferring methods of current MELCOR code are modified and adopted into the RN1 package. The data structure of the current MELCOR code using FORTRAN77 causes a difficult grasping of meaning of the variables as well as waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which lead to an efficient memory treatment and an easy understanding of the code. Restructuring of the RN1 package addressed in this paper includes module development, subroutine modification, and treats MELGEN, which generates data file, as well as MELCOR, which is processing a calculation. The verification has been done by comparing the results of the modified code with those from the existing code. As the trends are similar to each other, it hints that the same approach could be extended to the entire code package. It is expected that code restructuring will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models

  12. In-Package Chemistry Abstraction

    Energy Technology Data Exchange (ETDEWEB)

    P.S. Domski

    2003-07-21

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

  13. In-Package Chemistry Abstraction

    International Nuclear Information System (INIS)

    P.S. Domski

    2003-01-01

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

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

    OpenAIRE

    González Pericot, Natalia; Río Merino, Mercedes del

    2011-01-01

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

  15. Packaging design criteria for the Hanford Ecorok Packaging

    International Nuclear Information System (INIS)

    Mercado, M.S.

    1996-01-01

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

  16. Dual Use Packaging, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA seeks down-weighted packaging compatible with microwave preparation and perhaps high hydrostatic pressure processing. New packaging must satisfy NASA's 3-year...

  17. Food packaging history and innovations.

    Science.gov (United States)

    Risch, Sara J

    2009-09-23

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

  18. Fair Package Assignment

    Science.gov (United States)

    Lahaie, Sébastien; Parkes, David C.

    We consider the problem of fair allocation in the package assignment model, where a set of indivisible items, held by single seller, must be efficiently allocated to agents with quasi-linear utilities. A fair assignment is one that is efficient and envy-free. We consider a model where bidders have superadditive valuations, meaning that items are pure complements. Our central result is that core outcomes are fair and even coalition-fair over this domain, while fair distributions may not even exist for general valuations. Of relevance to auction design, we also establish that the core is equivalent to the set of anonymous-price competitive equilibria, and that superadditive valuations are a maximal domain that guarantees the existence of anonymous-price competitive equilibrium. Our results are analogs of core equivalence results for linear prices in the standard assignment model, and for nonlinear, non-anonymous prices in the package assignment model with general valuations.

  19. Anticounterfeit packaging technologies

    Directory of Open Access Journals (Sweden)

    Ruchir Y Shah

    2010-01-01

    Full Text Available Packaging is the coordinated system that encloses and protects the dosage form. Counterfeit drugs are the major cause of morbidity, mortality, and failure of public interest in the healthcare system. High price and well-known brands make the pharma market most vulnerable, which accounts for top priority cardiovascular, obesity, and antihyperlipidemic drugs and drugs like sildenafil. Packaging includes overt and covert technologies like barcodes, holograms, sealing tapes, and radio frequency identification devices to preserve the integrity of the pharmaceutical product. But till date all the available techniques are synthetic and although provide considerable protection against counterfeiting, have certain limitations which can be overcome by the application of natural approaches and utilization of the principles of nanotechnology.

  20. Package materials, waste form

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

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

  1. Irradiation of packaged food

    International Nuclear Information System (INIS)

    Kilcast, D.

    1990-01-01

    Food irradiation is used to improve the safety of food by killing insects and microorganisms, to inhibit sprouting in crops such as onions and potatoes and to control ripening in agricultural produce. In order to prevent re-infestation and re-contamination it is essential that the food is suitably packed. Consequently, the packaging material is irradiated whilst in contact with the food, and it is important that the material is resistant to radiation-induced changes. In this paper the nature of the irradiation process is reviewed briefly, together with the known effects of irradiation on packaging materials and their implications for the effective application of food irradiation. Recent research carried out at the Leatherhead Food RA on the possibility of taint transfer into food is described. (author)

  2. Technology transfer packages

    International Nuclear Information System (INIS)

    Mizon, G.A.; Bleasdale, P.A.

    1994-01-01

    Nuclear power is firmly established in many developed countries'energy policies and is being adopted by emerging nations as an attractive way of gaining energy self sufficiency. The early users of nuclear power had to develop the technology that they needed, which now, through increasing world wide experience, has been rationalised to meet demanding economic and environmental pressures. These justifiable pressures, can lead to existing suppliers of nuclear services to consider changing to more appropriate technologies and for new suppliers to consider licensing proven technology rather then incurring the cost of developing new alternatives. The transfer of technology, under license, is made more straight forward if the owner conveniently groups appropriate technology into packages. This paper gives examples of 'Technology Packages' and suggests criteria for the specification, selection and contractual requirements to ensure successful licensing

  3. CASL Verification and Validation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Mousseau, Vincent Andrew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Dinh, Nam [North Carolina State Univ., Raleigh, NC (United States)

    2016-06-30

    This report documents the Consortium for Advanced Simulation of LWRs (CASL) verification and validation plan. The document builds upon input from CASL subject matter experts, most notably the CASL Challenge Problem Product Integrators, CASL Focus Area leaders, and CASL code development and assessment teams. This document will be a living document that will track progress on CASL to do verification and validation for both the CASL codes (including MPACT, CTF, BISON, MAMBA) and for the CASL challenge problems (CIPS, PCI, DNB). The CASL codes and the CASL challenge problems are at differing levels of maturity with respect to validation and verification. The gap analysis will summarize additional work that needs to be done. Additional VVUQ work will be done as resources permit. This report is prepared for the Department of Energy’s (DOE’s) CASL program in support of milestone CASL.P13.02.

  4. Technical challenges for dismantlement verification

    International Nuclear Information System (INIS)

    Olinger, C.T.; Stanbro, W.D.; Johnston, R.G.; Nakhleh, C.W.; Dreicer, J.S.

    1997-01-01

    In preparation for future nuclear arms reduction treaties, including any potential successor treaties to START I and II, the authors have been examining possible methods for bilateral warhead dismantlement verification. Warhead dismantlement verification raises significant challenges in the political, legal, and technical arenas. This discussion will focus on the technical issues raised by warhead arms controls. Technical complications arise from several sources. These will be discussed under the headings of warhead authentication, chain-of-custody, dismantlement verification, non-nuclear component tracking, component monitoring, and irreversibility. The authors will discuss possible technical options to address these challenges as applied to a generic dismantlement and disposition process, in the process identifying limitations and vulnerabilities. They expect that these considerations will play a large role in any future arms reduction effort and, therefore, should be addressed in a timely fashion

  5. Amdahl 470 Chip Package

    CERN Multimedia

    1975-01-01

    In the late 70s the larger IBM computers were water cooled. Amdahl, an IBM competitor, invented an air cooling technology for it's computers. His company worked hard, developing a computer that was faster and less expensive than the IBM System/360 mainframe computer systems. This object contains an actual Amdahl series 470 computer logic chip with an air cooling device mounted on top. The package leads and cooling tower are gold-plated.

  6. Aquaculture information package

    Energy Technology Data Exchange (ETDEWEB)

    Boyd, T.; Rafferty, K.

    1998-08-01

    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.

  7. The CASA Software Package

    Science.gov (United States)

    Petry, Dirk

    2018-03-01

    CASA is the standard science data analysis package for ALMA and VLA but it can also be used for the analysis of data from other observatories. In this talk, I will give an overview of the structure and features of CASA, who develops it, and the present status and plans, and then show typical analysis workflows for ALMA data with special emphasis on the handling of single dish data and its combination with interferometric data.

  8. Utterance Verification for Text-Dependent Speaker Recognition

    DEFF Research Database (Denmark)

    Kinnunen, Tomi; Sahidullah, Md; Kukanov, Ivan

    2016-01-01

    Text-dependent automatic speaker verification naturally calls for the simultaneous verification of speaker identity and spoken content. These two tasks can be achieved with automatic speaker verification (ASV) and utterance verification (UV) technologies. While both have been addressed previously...

  9. The Ettention software package

    International Nuclear Information System (INIS)

    Dahmen, Tim; Marsalek, Lukas; Marniok, Nico; Turoňová, Beata; Bogachev, Sviatoslav; Trampert, Patrick; Nickels, Stefan; Slusallek, Philipp

    2016-01-01

    We present a novel software package for the problem “reconstruction from projections” in electron microscopy. The Ettention framework consists of a set of modular building-blocks for tomographic reconstruction algorithms. The well-known block iterative reconstruction method based on Kaczmarz algorithm is implemented using these building-blocks, including adaptations specific to electron tomography. Ettention simultaneously features (1) a modular, object-oriented software design, (2) optimized access to high-performance computing (HPC) platforms such as graphic processing units (GPU) or many-core architectures like Xeon Phi, and (3) accessibility to microscopy end-users via integration in the IMOD package and eTomo user interface. We also provide developers with a clean and well-structured application programming interface (API) that allows for extending the software easily and thus makes it an ideal platform for algorithmic research while hiding most of the technical details of high-performance computing. - Highlights: • Novel software package for “reconstruction from projections” in electron microscopy. • Support for high-resolution reconstructions on iterative reconstruction algorithms. • Support for CPU, GPU and Xeon Phi. • Integration in the IMOD software. • Platform for algorithm researchers: object oriented, modular design.

  10. The Ettention software package

    Energy Technology Data Exchange (ETDEWEB)

    Dahmen, Tim, E-mail: Tim.Dahmen@dfki.de [German Research Center for Artificial Intelligence GmbH (DFKI), 66123 Saarbrücken (Germany); Saarland University, 66123 Saarbrücken (Germany); Marsalek, Lukas [Eyen SE, Na Nivách 1043/16, 141 00 Praha 4 (Czech Republic); Saarland University, 66123 Saarbrücken (Germany); Marniok, Nico [Saarland University, 66123 Saarbrücken (Germany); Turoňová, Beata [Saarland University, 66123 Saarbrücken (Germany); IMPRS-CS, Max-Planck Institute for Informatics, Campus E 1.4, 66123 Saarbrücken (Germany); Bogachev, Sviatoslav [Saarland University, 66123 Saarbrücken (Germany); Trampert, Patrick; Nickels, Stefan [German Research Center for Artificial Intelligence GmbH (DFKI), 66123 Saarbrücken (Germany); Slusallek, Philipp [German Research Center for Artificial Intelligence GmbH (DFKI), 66123 Saarbrücken (Germany); Saarland University, 66123 Saarbrücken (Germany)

    2016-02-15

    We present a novel software package for the problem “reconstruction from projections” in electron microscopy. The Ettention framework consists of a set of modular building-blocks for tomographic reconstruction algorithms. The well-known block iterative reconstruction method based on Kaczmarz algorithm is implemented using these building-blocks, including adaptations specific to electron tomography. Ettention simultaneously features (1) a modular, object-oriented software design, (2) optimized access to high-performance computing (HPC) platforms such as graphic processing units (GPU) or many-core architectures like Xeon Phi, and (3) accessibility to microscopy end-users via integration in the IMOD package and eTomo user interface. We also provide developers with a clean and well-structured application programming interface (API) that allows for extending the software easily and thus makes it an ideal platform for algorithmic research while hiding most of the technical details of high-performance computing. - Highlights: • Novel software package for “reconstruction from projections” in electron microscopy. • Support for high-resolution reconstructions on iterative reconstruction algorithms. • Support for CPU, GPU and Xeon Phi. • Integration in the IMOD software. • Platform for algorithm researchers: object oriented, modular design.

  11. MACCS2 development and verification efforts

    International Nuclear Information System (INIS)

    Young, M.; Chanin, D.

    1997-01-01

    MACCS2 represents a major enhancement of the capabilities of its predecessor MACCS, the MELCOR Accident Consequence Code System. MACCS, released in 1987, was developed to estimate the potential impacts to the surrounding public of severe accidents at nuclear power plants. The principal phenomena considered in MACCS/MACCS2 are atmospheric transport and deposition under time-variant meteorology, short-term and long-term mitigative actions and exposure pathways, deterministic and stochastic health effects, and economic costs. MACCS2 was developed as a general-purpose analytical tool applicable to diverse reactor and nonreactor facilities. The MACCS2 package includes three primary enhancements: (1) a more flexible emergency response model, (2) an expanded library of radionuclides, and (3) a semidynamic food-chain model. In addition, errors that had been identified in MACCS version1.5.11.1 were corrected, including an error that prevented the code from providing intermediate-phase results. MACCS2 version 1.10 beta test was released to the beta-test group in May, 1995. In addition, the University of New Mexico (UNM) has completed an independent verification study of the code package. Since the beta-test release of MACCS2 version 1.10, a number of minor errors have been identified and corrected, and a number of enhancements have been added to the code package. The code enhancements added since the beta-test release of version 1.10 include: (1) an option to allow the user to input the σ y and σ z plume expansion parameters in a table-lookup form for incremental downwind distances, (2) an option to define different initial dimensions for up to four segments of a release, (3) an enhancement to the COMIDA2 food-chain model preprocessor to allow the user to supply externally calculated tables of tritium food-chain dose per unit deposition on farmland to support analyses of tritium releases, and (4) the capability to calculate direction-dependent doses

  12. Formal Verification of Continuous Systems

    DEFF Research Database (Denmark)

    Sloth, Christoffer

    2012-01-01

    and the verification procedures should be algorithmically synthesizable. Autonomous control plays an important role in many safety-critical systems. This implies that a malfunction in the control system can have catastrophic consequences, e.g., in space applications where a design flaw can result in large economic...... losses. Furthermore, a malfunction in the control system of a surgical robot may cause death of patients. The previous examples involve complex systems that are required to operate according to complex specifications. The systems cannot be formally verified by modern verification techniques, due...

  13. Biometric Technologies and Verification Systems

    CERN Document Server

    Vacca, John R

    2007-01-01

    Biometric Technologies and Verification Systems is organized into nine parts composed of 30 chapters, including an extensive glossary of biometric terms and acronyms. It discusses the current state-of-the-art in biometric verification/authentication, identification and system design principles. It also provides a step-by-step discussion of how biometrics works; how biometric data in human beings can be collected and analyzed in a number of ways; how biometrics are currently being used as a method of personal identification in which people are recognized by their own unique corporal or behavior

  14. IMRT plan verification in radiotherapy

    International Nuclear Information System (INIS)

    Vlk, P.

    2006-01-01

    This article describes the procedure for verification of IMRT (Intensity modulated radiation therapy) plan, which is used in the Oncological Institute of St. Elisabeth in Bratislava. It contains basic description of IMRT technology and developing a deployment plan for IMRT planning system CORVUS 6.0, the device Mimic (Multilammelar intensity modulated collimator) and the overall process of verifying the schedule created. The aim of verification is particularly good control of the functions of MIMIC and evaluate the overall reliability of IMRT planning. (author)

  15. K Basins Field Verification Program

    International Nuclear Information System (INIS)

    Booth, H.W.

    1994-01-01

    The Field Verification Program establishes a uniform and systematic process to ensure that technical information depicted on selected engineering drawings accurately reflects the actual existing physical configuration. This document defines the Field Verification Program necessary to perform the field walkdown and inspection process that identifies the physical configuration of the systems required to support the mission objectives of K Basins. This program is intended to provide an accurate accounting of the actual field configuration by documenting the as-found information on a controlled drawing

  16. Runtime Verification Through Forward Chaining

    Directory of Open Access Journals (Sweden)

    Alan Perotti

    2014-12-01

    Full Text Available In this paper we present a novel rule-based approach for Runtime Verification of FLTL properties over finite but expanding traces. Our system exploits Horn clauses in implication form and relies on a forward chaining-based monitoring algorithm. This approach avoids the branching structure and exponential complexity typical of tableaux-based formulations, creating monitors with a single state and a fixed number of rules. This allows for a fast and scalable tool for Runtime Verification: we present the technical details together with a working implementation.

  17. Implementation and integration of program packages NAMMU and HYPAC

    International Nuclear Information System (INIS)

    Nedbal, T.

    1986-05-01

    This work is prepared for the Swedish Power Inspectorate (SKI). The SKI has from the Atomic Energy Research Establishment (AERE) at Harwell, U.K., acquired the computer model NAMMU for groundwater hydrology calculations. The code was first implemented on an AMDAHL 470, a IBM compatible computer, and then modified in order to integrate it with HYPAC, which is a program package for pre- and post-processing finite element data, developed by KEMAKTA AB. This report describes the modifications done to both NAMMU and HYPAC, and the verification of the coupled program system NAMMU-HYPAC. (author)

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

    International Nuclear Information System (INIS)

    Smith, R.I.

    1990-06-01

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

  19. Ensuring socially responsible packaging design

    DEFF Research Database (Denmark)

    Geert Jensen, Birgitte

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

  20. Packaging based on polymeric materials

    Directory of Open Access Journals (Sweden)

    Jovanović Slobodan M.

    2005-01-01

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

  1. A Scalable Approach for Hardware Semiformal Verification

    OpenAIRE

    Grimm, Tomas; Lettnin, Djones; Hübner, Michael

    2018-01-01

    The current verification flow of complex systems uses different engines synergistically: virtual prototyping, formal verification, simulation, emulation and FPGA prototyping. However, none is able to verify a complete architecture. Furthermore, hybrid approaches aiming at complete verification use techniques that lower the overall complexity by increasing the abstraction level. This work focuses on the verification of complex systems at the RT level to handle the hardware peculiarities. Our r...

  2. Verification of EPA's " Preliminary remediation goals for radionuclides" (PRG) electronic calculator

    Energy Technology Data Exchange (ETDEWEB)

    Stagich, B. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-03-29

    The U.S. Environmental Protection Agency (EPA) requested an external, independent verification study of their “Preliminary Remediation Goals for Radionuclides” (PRG) electronic calculator. The calculator provides information on establishing PRGs for radionuclides at Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) sites with radioactive contamination (Verification Study Charge, Background). These risk-based PRGs set concentration limits using carcinogenic toxicity values under specific exposure conditions (PRG User’s Guide, Section 1). The purpose of this verification study is to ascertain that the computer codes has no inherit numerical problems with obtaining solutions as well as to ensure that the equations are programmed correctly.

  3. Power Electronics Packaging Reliability | Transportation Research | NREL

    Science.gov (United States)

    Packaging Reliability Power Electronics Packaging Reliability A photo of a piece of power electronics laboratory equipment. NREL power electronics packaging reliability research investigates the electronics packaging around a semiconductor switching device determines the electrical, thermal, and

  4. Development and Deployment of a Compact Eye-Safe Scanning Differential absorption Lidar (DIAL) for Spatial Mapping of Carbon Dioxide for Monitoring/Verification/Accounting at Geologic Sequestration Sites

    Energy Technology Data Exchange (ETDEWEB)

    Repasky, Kevin

    2014-03-31

    A scanning differential absorption lidar (DIAL) instrument for monitoring carbon dioxide has been developed. The laser transmitter uses two tunable discrete mode laser diodes (DMLD) operating in the continuous wave (cw) mode with one locked to the online absorption wavelength and the other operating at the offline wavelength. Two in-line fiber optic switches are used to switch between online and offline operation. After the fiber optic switch, an acousto- optic modulator (AOM) is used to generate a pulse train used to injection seed an erbium doped fiber amplifier (EDFA) to produce eye-safe laser pulses with maximum pulse energies of 66 {micro}J, a pulse repetition frequency of 15 kHz, and an operating wavelength of 1.571 {micro}m. The DIAL receiver uses a 28 cm diameter Schmidt-Cassegrain telescope to collect that backscattered light, which is then monitored using a photo-multiplier tube (PMT) module operating in the photon counting mode. The DIAL instrument has been operated from a laboratory environment on the campus of Montana State University, at the Zero Emission Research Technology (ZERT) field site located in the agricultural research area on the western end of the Montana State University campus, and at the Big Sky Carbon Sequestration Partnership site located in north-central Montana. DIAL data has been collected and profiles have been validated using a co-located Licor LI-820 Gas Analyzer point sensor.

  5. Likelihood-ratio-based biometric verification

    NARCIS (Netherlands)

    Bazen, A.M.; Veldhuis, Raymond N.J.

    2002-01-01

    This paper presents results on optimal similarity measures for biometric verification based on fixed-length feature vectors. First, we show that the verification of a single user is equivalent to the detection problem, which implies that for single-user verification the likelihood ratio is optimal.

  6. Likelihood Ratio-Based Biometric Verification

    NARCIS (Netherlands)

    Bazen, A.M.; Veldhuis, Raymond N.J.

    The paper presents results on optimal similarity measures for biometric verification based on fixed-length feature vectors. First, we show that the verification of a single user is equivalent to the detection problem, which implies that, for single-user verification, the likelihood ratio is optimal.

  7. Economic tour package model using heuristic

    Science.gov (United States)

    Rahman, Syariza Abdul; Benjamin, Aida Mauziah; Bakar, Engku Muhammad Nazri Engku Abu

    2014-07-01

    A tour-package is a prearranged tour that includes products and services such as food, activities, accommodation, and transportation, which are sold at a single price. Since the competitiveness within tourism industry is very high, many of the tour agents try to provide attractive tour-packages in order to meet tourist satisfaction as much as possible. Some of the criteria that are considered by the tourist are the number of places to be visited and the cost of the tour-packages. Previous studies indicate that tourists tend to choose economical tour-packages and aiming to visit as many places as they can cover. Thus, this study proposed tour-package model using heuristic approach. The aim is to find economical tour-packages and at the same time to propose as many places as possible to be visited by tourist in a given geographical area particularly in Langkawi Island. The proposed model considers only one starting point where the tour starts and ends at an identified hotel. This study covers 31 most attractive places in Langkawi Island from various categories of tourist attractions. Besides, the allocation of period for lunch and dinner are included in the proposed itineraries where it covers 11 popular restaurants around Langkawi Island. In developing the itinerary, the proposed heuristic approach considers time window for each site (hotel/restaurant/place) so that it represents real world implementation. We present three itineraries with different time constraints (1-day, 2-day and 3-day tour-package). The aim of economic model is to minimize the tour-package cost as much as possible by considering entrance fee of each visited place. We compare the proposed model with our uneconomic model from our previous study. The uneconomic model has no limitation to the cost with the aim to maximize the number of places to be visited. Comparison between the uneconomic and economic itinerary has shown that the proposed model have successfully achieved the objective that

  8. Nuclear waste package fabricated from concrete

    International Nuclear Information System (INIS)

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

    1987-03-01

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

  9. Module 13: Bulk Packaging Shipments by Highway

    International Nuclear Information System (INIS)

    Przybylski, J.L.

    1994-07-01

    The Hazardous Materials Modular Training Program provides participating United States Department of Energy (DOE) sites with a basic, yet comprehensive, hazardous materials transportation training program for use onsite. This program may be used to assist individual program entities to satisfy the general awareness, safety training, and function specific training requirements addressed in Code of Federal Regulation (CFR), Title 49, Part 172, Subpart H -- ''Training.'' Module 13 -- Bulk Packaging Shipments by Highway is a supplement to the Basic Hazardous Materials Workshop. Module 13 -- Bulk Packaging Shipments by Highway focuses on bulk shipments of hazardous materials by highway mode, which have additional or unique requirements beyond those addressed in the ten module core program. Attendance in this course of instruction should be limited to those individuals with work experience in transporting hazardous materials utilizing bulk packagings and who have completed the Basic Hazardous Materials Workshop or an equivalent. Participants will become familiar with the rules and regulations governing the transportation by highway of hazardous materials in bulk packagings and will demonstrate the application of these requirements through work projects and examination

  10. Structural dynamics of retroviral genome and the packaging.

    Science.gov (United States)

    Miyazaki, Yasuyuki; Miyake, Ariko; Nomaguchi, Masako; Adachi, Akio

    2011-01-01

    Retroviruses can cause diseases such as AIDS, leukemia, and tumors, but are also used as vectors for human gene therapy. All retroviruses, except foamy viruses, package two copies of unspliced genomic RNA into their progeny viruses. Understanding the molecular mechanisms of retroviral genome packaging will aid the design of new anti-retroviral drugs targeting the packaging process and improve the efficacy of retroviral vectors. Retroviral genomes have to be specifically recognized by the cognate nucleocapsid domain of the Gag polyprotein from among an excess of cellular and spliced viral mRNA. Extensive virological and structural studies have revealed how retroviral genomic RNA is selectively packaged into the viral particles. The genomic area responsible for the packaging is generally located in the 5' untranslated region (5' UTR), and contains dimerization site(s). Recent studies have shown that retroviral genome packaging is modulated by structural changes of RNA at the 5' UTR accompanied by the dimerization. In this review, we focus on three representative retroviruses, Moloney murine leukemia virus, human immunodeficiency virus type 1 and 2, and describe the molecular mechanism of retroviral genome packaging.

  11. Structural dynamics of retroviral genome and the packaging

    Directory of Open Access Journals (Sweden)

    Yasuyuki eMiyazaki

    2011-12-01

    Full Text Available Retroviruses can cause diseases such as AIDS, leukemia and tumors, but are also used as vectors for human gene therapy. All retroviruses, except foamy viruses, package two copies of unspliced genomic RNA into their progeny viruses. Understanding the molecular mechanisms of retroviral genome packaging will aid the design of new anti-retroviral drugs targeting the packaging process and improve the efficacy of retroviral vectors. Retroviral genomes have to be specifically recognized by the cognate nucleocapsid (NC domain of the Gag polyprotein from among an excess of cellular and spliced viral mRNA. Extensive virological and structural studies have revealed how retroviral genomic RNA is selectively packaged into the viral particles. The genomic area responsible for the packaging is generally located in the 5’ untranslated region (5’ UTR, and contains dimerization site(s. Recent studies have shown that retroviral genome packaging is modulated by structural changes of RNA at the 5’ UTR accompanied by the dimerization. In this review, we focus on three representative retroviruses, Moloney murine leukemia virus (MoMLV, human immunodeficiency virus type 1 (HIV-1 and 2 (HIV-2, and describe the molecular mechanism of retroviral genome packaging.

  12. Automated Verification of Virtualized Infrastructures

    DEFF Research Database (Denmark)

    Bleikertz, Sören; Gross, Thomas; Mödersheim, Sebastian Alexander

    2011-01-01

    Virtualized infrastructures and clouds present new challenges for security analysis and formal verification: they are complex environments that continuously change their shape, and that give rise to non-trivial security goals such as isolation and failure resilience requirements. We present a pla...

  13. Future of monitoring and verification

    International Nuclear Information System (INIS)

    Wagenmakers, H.

    1991-01-01

    The organized verification entrusted to IAEA for the implementation of the NPT, of the Treaty of Tlatelolco and of the Treaty of Rarotonga, reaches reasonable standards. The current dispute with the Democratic People's Republic of Korea about the conclusion of a safeguards agreement with IAEA, by its exceptional nature, underscores rather than undermines the positive judgement to be passed on IAEA's overall performance. The additional task given to the Director General of IAEA under Security Council resolution 687 (1991) regarding Iraq's nuclear-weapons-usable material is particularly challenging. For the purposes of this paper, verification is defined as the process for establishing whether the States parties are complying with an agreement. In the final stage verification may lead into consideration of how to respond to non-compliance. Monitoring is perceived as the first level in the verification system. It is one generic form of collecting information on objects, activities or events and it involves a variety of instruments ranging from communications satellites to television cameras or human inspectors. Monitoring may also be used as a confidence-building measure

  14. Hot cell verification facility update

    International Nuclear Information System (INIS)

    Titzler, P.A.; Moffett, S.D.; Lerch, R.E.

    1985-01-01

    The Hot Cell Verification Facility (HCVF) provides a prototypic hot cell mockup to check equipment for functional and remote operation, and provides actual hands-on training for operators. The facility arrangement is flexible and assists in solving potential problems in a nonradioactive environment. HCVF has been in operation for six years, and the facility is a part of the Hanford Engineering Development Laboratory

  15. Static Verification for Code Contracts

    Science.gov (United States)

    Fähndrich, Manuel

    The Code Contracts project [3] at Microsoft Research enables programmers on the .NET platform to author specifications in existing languages such as C# and VisualBasic. To take advantage of these specifications, we provide tools for documentation generation, runtime contract checking, and static contract verification.

  16. Verification of safety critical software

    International Nuclear Information System (INIS)

    Son, Ki Chang; Chun, Chong Son; Lee, Byeong Joo; Lee, Soon Sung; Lee, Byung Chai

    1996-01-01

    To assure quality of safety critical software, software should be developed in accordance with software development procedures and rigorous software verification and validation should be performed. Software verification is the formal act of reviewing, testing of checking, and documenting whether software components comply with the specified requirements for a particular stage of the development phase[1]. New software verification methodology was developed and was applied to the Shutdown System No. 1 and 2 (SDS1,2) for Wolsung 2,3 and 4 nuclear power plants by Korea Atomic Energy Research Institute(KAERI) and Atomic Energy of Canada Limited(AECL) in order to satisfy new regulation requirements of Atomic Energy Control Boars(AECB). Software verification methodology applied to SDS1 for Wolsung 2,3 and 4 project will be described in this paper. Some errors were found by this methodology during the software development for SDS1 and were corrected by software designer. Outputs from Wolsung 2,3 and 4 project have demonstrated that the use of this methodology results in a high quality, cost-effective product. 15 refs., 6 figs. (author)

  17. Eggspectation : organic egg verification tool

    NARCIS (Netherlands)

    Ruth, van S.M.; Hoogenboom, L.A.P.

    2011-01-01

    In 2009 RIKILT conducted a study on about 2,000 eggs to evaluate three different analytical verification methods: carotenoid profiling, fatty acid profiling and isotope ratio mass spectrometry. The eggs were collected from about 50 Dutch farms. The selection was based on the farms’ location and

  18. Documentation and analysis for packaging limited quantity ice chests

    International Nuclear Information System (INIS)

    Nguyen, P.M.

    1995-01-01

    The purpose of this Documentation and Analysis for Packaging (DAP) is to document that ice chests meet the intent of the International Air Transport Association (IATA) and the U.S. Department of Transportation (DOT) Code of Federal Regulations as strong, tight containers for the packaging of limited quantities for transport. This DAP also outlines the packaging method used to protect the sample bottles from breakage. Because the ice chests meet the DOT requirements, they can be used to ship LTD QTY on the Hanford Site

  19. Packaging - Materials review

    Energy Technology Data Exchange (ETDEWEB)

    Herrmann, Matthias [Hoppecke Advanced Battery Technology GmbH, 08056 Zwickau (Germany)

    2014-06-16

    Nowadays, a large number of different electrochemical energy storage systems are known. In the last two decades the development was strongly driven by a continuously growing market of portable electronic devices (e.g. cellular phones, lap top computers, camcorders, cameras, tools). Current intensive efforts are under way to develop systems for automotive industry within the framework of electrically propelled mobility (e.g. hybrid electric vehicles, plug-in hybrid electric vehicles, full electric vehicles) and also for the energy storage market (e.g. electrical grid stability, renewable energies). Besides the different systems (cell chemistries), electrochemical cells and batteries were developed and are offered in many shapes, sizes and designs, in order to meet performance and design requirements of the widespread applications. Proper packaging is thereby one important technological step for designing optimum, reliable and safe batteries for operation. In this contribution, current packaging approaches of cells and batteries together with the corresponding materials are discussed. The focus is laid on rechargeable systems for industrial applications (i.e. alkaline systems, lithium-ion, lead-acid). In principle, four different cell types (shapes) can be identified - button, cylindrical, prismatic and pouch. Cell size can be either in accordance with international (e.g. International Electrotechnical Commission, IEC) or other standards or can meet application-specific dimensions. Since cell housing or container, terminals and, if necessary, safety installations as inactive (non-reactive) materials reduce energy density of the battery, the development of low-weight packages is a challenging task. In addition to that, other requirements have to be fulfilled: mechanical stability and durability, sealing (e.g. high permeation barrier against humidity for lithium-ion technology), high packing efficiency, possible installation of safety devices (current interrupt device

  20. Packaging - Materials review

    Science.gov (United States)

    Herrmann, Matthias

    2014-06-01

    Nowadays, a large number of different electrochemical energy storage systems are known. In the last two decades the development was strongly driven by a continuously growing market of portable electronic devices (e.g. cellular phones, lap top computers, camcorders, cameras, tools). Current intensive efforts are under way to develop systems for automotive industry within the framework of electrically propelled mobility (e.g. hybrid electric vehicles, plug-in hybrid electric vehicles, full electric vehicles) and also for the energy storage market (e.g. electrical grid stability, renewable energies). Besides the different systems (cell chemistries), electrochemical cells and batteries were developed and are offered in many shapes, sizes and designs, in order to meet performance and design requirements of the widespread applications. Proper packaging is thereby one important technological step for designing optimum, reliable and safe batteries for operation. In this contribution, current packaging approaches of cells and batteries together with the corresponding materials are discussed. The focus is laid on rechargeable systems for industrial applications (i.e. alkaline systems, lithium-ion, lead-acid). In principle, four different cell types (shapes) can be identified - button, cylindrical, prismatic and pouch. Cell size can be either in accordance with international (e.g. International Electrotechnical Commission, IEC) or other standards or can meet application-specific dimensions. Since cell housing or container, terminals and, if necessary, safety installations as inactive (non-reactive) materials reduce energy density of the battery, the development of low-weight packages is a challenging task. In addition to that, other requirements have to be fulfilled: mechanical stability and durability, sealing (e.g. high permeation barrier against humidity for lithium-ion technology), high packing efficiency, possible installation of safety devices (current interrupt device