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

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

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.

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

International Nuclear Information System (INIS)

Appel, M.J.

2006-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

M. J. Appel

2006-11-02

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

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

Energy Technology Data Exchange (ETDEWEB)

M. J. Appel and J. M. Capron

2007-07-25

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

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.

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)

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.

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

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

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

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

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

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

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

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)

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

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

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.

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)

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.

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

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)

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

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.

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.

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

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

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

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

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

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

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.

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

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

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

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

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

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.

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

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.)

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)

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.

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.

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

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

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.

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

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

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

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

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

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/ .

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

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

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

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

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

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

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

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

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

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.

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.

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.)

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

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

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

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

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.)

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

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

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)

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

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

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

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

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

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 ).

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

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

A history of solid waste packaging at the Hanford Site

International Nuclear Information System (INIS)

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

1995-02-01

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

A 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

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

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

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

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.

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

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.

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

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

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

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

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

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

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...

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

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

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

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

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

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

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

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.

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

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

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.

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

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

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

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.

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…

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

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

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

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.

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

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.

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

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.

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.

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)

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)

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

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

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

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

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

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)

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.

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

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

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

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.

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

Verification of chemistry reference ranges using a simple method in sub-Saharan Africa

Directory of Open Access Journals (Sweden)

Irith De Baetselier

2016-10-01

Full Text Available Background: Chemistry safety assessments are interpreted by using chemistry reference ranges (CRRs. Verification of CRRs is time consuming and often requires a statistical background. Objectives: We report on an easy and cost-saving method to verify CRRs. Methods: Using a former method introduced by Sigma Diagnostics, three study sites in sub- Saharan Africa, Bondo, Kenya, and Pretoria and Bloemfontein, South Africa, verified the CRRs for hepatic and renal biochemistry assays performed during a clinical trial of HIV antiretroviral pre-exposure prophylaxis. The aspartate aminotransferase/alanine aminotransferase, creatinine and phosphorus results from 10 clinically-healthy participants at the screening visit were used. In the event the CRRs did not pass the verification, new CRRs had to be calculated based on 40 clinically-healthy participants. Results: Within a few weeks, the study sites accomplished verification of the CRRs without additional costs. The aspartate aminotransferase reference ranges for the Bondo, Kenya site and the alanine aminotransferase reference ranges for the Pretoria, South Africa site required adjustment. The phosphorus CRR passed verification and the creatinine CRR required adjustment at every site. The newly-established CRR intervals were narrower than the CRRs used previously at these study sites due to decreases in the upper limits of the reference ranges. As a result, more toxicities were detected. Conclusion: To ensure the safety of clinical trial participants, verification of CRRs should be standard practice in clinical trials conducted in settings where the CRR has not been validated for the local population. This verification method is simple, inexpensive, and can be performed by any medical laboratory.

Verification of chemistry reference ranges using a simple method in sub-Saharan Africa.

Science.gov (United States)

De Baetselier, Irith; Taylor, Douglas; Mandala, Justin; Nanda, Kavita; Van Campenhout, Christel; Agingu, Walter; Madurai, Lorna; Barsch, Eva-Maria; Deese, Jennifer; Van Damme, Lut; Crucitti, Tania

2016-01-01

Chemistry safety assessments are interpreted by using chemistry reference ranges (CRRs). Verification of CRRs is time consuming and often requires a statistical background. We report on an easy and cost-saving method to verify CRRs. Using a former method introduced by Sigma Diagnostics, three study sites in sub-Saharan Africa, Bondo, Kenya, and Pretoria and Bloemfontein, South Africa, verified the CRRs for hepatic and renal biochemistry assays performed during a clinical trial of HIV antiretroviral pre-exposure prophylaxis. The aspartate aminotransferase/alanine aminotransferase, creatinine and phosphorus results from 10 clinically-healthy participants at the screening visit were used. In the event the CRRs did not pass the verification, new CRRs had to be calculated based on 40 clinically-healthy participants. Within a few weeks, the study sites accomplished verification of the CRRs without additional costs. The aspartate aminotransferase reference ranges for the Bondo, Kenya site and the alanine aminotransferase reference ranges for the Pretoria, South Africa site required adjustment. The phosphorus CRR passed verification and the creatinine CRR required adjustment at every site. The newly-established CRR intervals were narrower than the CRRs used previously at these study sites due to decreases in the upper limits of the reference ranges. As a result, more toxicities were detected. To ensure the safety of clinical trial participants, verification of CRRs should be standard practice in clinical trials conducted in settings where the CRR has not been validated for the local population. This verification method is simple, inexpensive, and can be performed by any medical laboratory.

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

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

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.

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

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.

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

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

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

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...

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.

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

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

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

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.)

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.

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

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

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

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

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

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

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.

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.

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

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

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

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

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

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.

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

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

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

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

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

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)

A restructuring of the FL package for the MIDAS computer code

International Nuclear Information System (INIS)

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

2005-01-01

The developmental need for a localized severe accident analysis code is on the rise, and KAERI is developing a severe accident code MIDAS, based on MELCOR. The existing data saving method uses pointer variables for a fix-sized storage management, and it deteriorates the readability, maintainability and portability of the code. But new features in FORTRAN90 such as a dynamic allocation have been used for the restructuring. The restructuring of the data saving and transferring method of the existing code makes it easy to understand the code. Before an entire restructuring of the code, a restructuring template for a simple package was developed and tested. The target for the restructuring was the FL package which is responsible for modeling the thermal-hydraulic behavior of a liquid water, water vapor, and gases in MELCOR with the CVH package. The verification was done through comparing the results before and after the restructuring

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)

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)

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.

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...

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

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

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

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

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

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

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

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....

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

Online 3D EPID-based dose verification: Proof of concept

Energy Technology Data Exchange (ETDEWEB)

Spreeuw, Hanno; Rozendaal, Roel, E-mail: r.rozendaal@nki.nl; Olaciregui-Ruiz, Igor; González, Patrick; Mans, Anton; Mijnheer, Ben [Department of Radiation Oncology, The Netherlands Cancer Institute, Amsterdam 1066 CX (Netherlands); Herk, Marcel van [University of Manchester, Manchester Academic Health Science Centre, The Christie NHS Foundation Trust, Manchester M20 4BX (United Kingdom)

2016-07-15

Purpose: Delivery errors during radiotherapy may lead to medical harm and reduced life expectancy for patients. Such serious incidents can be avoided by performing dose verification online, i.e., while the patient is being irradiated, creating the possibility of halting the linac in case of a large overdosage or underdosage. The offline EPID-based 3D in vivo dosimetry system clinically employed at our institute is in principle suited for online treatment verification, provided the system is able to complete 3D dose reconstruction and verification within 420 ms, the present acquisition time of a single EPID frame. It is the aim of this study to show that our EPID-based dosimetry system can be made fast enough to achieve online 3D in vivo dose verification. Methods: The current dose verification system was sped up in two ways. First, a new software package was developed to perform all computations that are not dependent on portal image acquisition separately, thus removing the need for doing these calculations in real time. Second, the 3D dose reconstruction algorithm was sped up via a new, multithreaded implementation. Dose verification was implemented by comparing planned with reconstructed 3D dose distributions delivered to two regions in a patient: the target volume and the nontarget volume receiving at least 10 cGy. In both volumes, the mean dose is compared, while in the nontarget volume, the near-maximum dose (D2) is compared as well. The real-time dosimetry system was tested by irradiating an anthropomorphic phantom with three VMAT plans: a 6 MV head-and-neck treatment plan, a 10 MV rectum treatment plan, and a 10 MV prostate treatment plan. In all plans, two types of serious delivery errors were introduced. The functionality of automatically halting the linac was also implemented and tested. Results: The precomputation time per treatment was ∼180 s/treatment arc, depending on gantry angle resolution. The complete processing of a single portal frame

Online 3D EPID-based dose verification: Proof of concept

International Nuclear Information System (INIS)

Spreeuw, Hanno; Rozendaal, Roel; Olaciregui-Ruiz, Igor; González, Patrick; Mans, Anton; Mijnheer, Ben; Herk, Marcel van

2016-01-01

Purpose: Delivery errors during radiotherapy may lead to medical harm and reduced life expectancy for patients. Such serious incidents can be avoided by performing dose verification online, i.e., while the patient is being irradiated, creating the possibility of halting the linac in case of a large overdosage or underdosage. The offline EPID-based 3D in vivo dosimetry system clinically employed at our institute is in principle suited for online treatment verification, provided the system is able to complete 3D dose reconstruction and verification within 420 ms, the present acquisition time of a single EPID frame. It is the aim of this study to show that our EPID-based dosimetry system can be made fast enough to achieve online 3D in vivo dose verification. Methods: The current dose verification system was sped up in two ways. First, a new software package was developed to perform all computations that are not dependent on portal image acquisition separately, thus removing the need for doing these calculations in real time. Second, the 3D dose reconstruction algorithm was sped up via a new, multithreaded implementation. Dose verification was implemented by comparing planned with reconstructed 3D dose distributions delivered to two regions in a patient: the target volume and the nontarget volume receiving at least 10 cGy. In both volumes, the mean dose is compared, while in the nontarget volume, the near-maximum dose (D2) is compared as well. The real-time dosimetry system was tested by irradiating an anthropomorphic phantom with three VMAT plans: a 6 MV head-and-neck treatment plan, a 10 MV rectum treatment plan, and a 10 MV prostate treatment plan. In all plans, two types of serious delivery errors were introduced. The functionality of automatically halting the linac was also implemented and tested. Results: The precomputation time per treatment was ∼180 s/treatment arc, depending on gantry angle resolution. The complete processing of a single portal frame

Online 3D EPID-based dose verification: Proof of concept.

Science.gov (United States)

Spreeuw, Hanno; Rozendaal, Roel; Olaciregui-Ruiz, Igor; González, Patrick; Mans, Anton; Mijnheer, Ben; van Herk, Marcel

2016-07-01

Delivery errors during radiotherapy may lead to medical harm and reduced life expectancy for patients. Such serious incidents can be avoided by performing dose verification online, i.e., while the patient is being irradiated, creating the possibility of halting the linac in case of a large overdosage or underdosage. The offline EPID-based 3D in vivo dosimetry system clinically employed at our institute is in principle suited for online treatment verification, provided the system is able to complete 3D dose reconstruction and verification within 420 ms, the present acquisition time of a single EPID frame. It is the aim of this study to show that our EPID-based dosimetry system can be made fast enough to achieve online 3D in vivo dose verification. The current dose verification system was sped up in two ways. First, a new software package was developed to perform all computations that are not dependent on portal image acquisition separately, thus removing the need for doing these calculations in real time. Second, the 3D dose reconstruction algorithm was sped up via a new, multithreaded implementation. Dose verification was implemented by comparing planned with reconstructed 3D dose distributions delivered to two regions in a patient: the target volume and the nontarget volume receiving at least 10 cGy. In both volumes, the mean dose is compared, while in the nontarget volume, the near-maximum dose (D2) is compared as well. The real-time dosimetry system was tested by irradiating an anthropomorphic phantom with three VMAT plans: a 6 MV head-and-neck treatment plan, a 10 MV rectum treatment plan, and a 10 MV prostate treatment plan. In all plans, two types of serious delivery errors were introduced. The functionality of automatically halting the linac was also implemented and tested. The precomputation time per treatment was ∼180 s/treatment arc, depending on gantry angle resolution. The complete processing of a single portal frame, including dose verification, took

Reactor physics verification of the MCNP6 unstructured mesh capability

International Nuclear Information System (INIS)

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

2013-01-01

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

Reactor physics verification of the MCNP6 unstructured mesh capability

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

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)

Documentation package for the RFID temperature monitoring system (Model 9977 packages at NTS)

International Nuclear Information System (INIS)

Chen, K.; Tsai, H.

2009-01-01

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

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

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

Complex-wide representation of material packaged in 3013 containers

Energy Technology Data Exchange (ETDEWEB)

Narlesky, Joshua E.; Peppers, Larry G.; Friday, Gary P.

2009-06-01

The DOE sites packaging plutonium oxide materials packaged according to Department of Energy 3013 Standard (DOE-STD-3013) are responsible for ensuring that the materials are represented by one or more samples in the Materials Identification and Surveillance (MIS) program. The sites categorized most of the materials into process groups, and the remaining materials were characterized, based on the prompt gamma analysis results. The sites issued documents to identify the relationships between the materials packaged in 3013 containers and representative materials in the MIS program. These “Represented” documents were then reviewed and concurred with by the MIS Working Group. However, these documents were developed uniquely at each site and were issued before completion of sample characterization, small-scale experiments, and prompt gamma analysis, which provided more detailed information about the chemical impurities and the behavior of the material in storage. Therefore, based on the most recent data, relationships between the materials packaged in 3013 containers and representative materials in the MIS program been revised. With the prompt gamma analysis completed for Hanford, Rocky Flats, and Savannah River Site 3013 containers, MIS items have been assigned to the 3013 containers for which representation is based on the prompt gamma analysis results. With the revised relationships and the prompt gamma analysis results, a Master “Represented” table has been compiled to document the linkages between each 3013 container packaged to date and its representative MIS items. This table provides an important link between the Integrated Surveillance Program database, which contains information about each 3013 container to the MIS items database, which contains the characterization, prompt gamma data, and storage behavior data from shelf-life experiments for the representative MIS items.

Complex-wide representation of material packaged in 3013 containers

International Nuclear Information System (INIS)

Narlesky, Joshua E.; Peppers, Larry G.; Friday, Gary P.

2009-01-01

The DOE sites packaging plutonium oxide materials packaged according to Department of Energy 3013 Standard (DOE-STD-3013) are responsible for ensuring that the materials are represented by one or more samples in the Materials Identification and Surveillance (MIS) program. The sites categorized most of the materials into process groups, and the remaining materials were characterized, based on the prompt gamma analysis results. The sites issued documents to identify the relationships between the materials packaged in 3013 containers and representative materials in the MIS program. These 'Represented' documents were then reviewed and concurred with by the MIS Working Group. However, these documents were developed uniquely at each site and were issued before completion of sample characterization, small-scale experiments, and prompt gamma analysis, which provided more detailed information about the chemical impurities and the behavior of the material in storage. Therefore, based on the most recent data, relationships between the materials packaged in 3013 containers and representative materials in the MIS program have been revised. With the prompt gamma analysis completed for Hanford, Rocky Flats, and Savannah River Site 3013 containers, MIS items have been assigned to the 3013 containers for which representation is based on the prompt gamma analysis results. With the revised relationships and the prompt gamma analysis results, a Master 'Represented' table has been compiled to document the linkages between each 3013 container packaged to date and its representative MIS items. This table provides an important link between the Integrated Surveillance Program database, which contains information about each 3013 container to the MIS items database, which contains the characterization, prompt gamma data, and storage behavior data from shelf-life experiments for the representative MIS items

CTBT on-site inspections

Science.gov (United States)

Zucca, J. J.

2014-05-01

On-site inspection (OSI) is a critical part of the verification regime for the Comprehensive Nuclear-Test-Ban Treaty (CTBT). The OSI verification regime provides for international inspectors to make a suite of measurements and observations on site at the location of an event of interest. The other critical component of the verification regime is the International Monitoring System (IMS), which is a globally distributed network of monitoring stations. The IMS along with technical monitoring data from CTBT member countries, as appropriate, will be used to trigger an OSI. After the decision is made to carry out an OSI, it is important for the inspectors to deploy to the field site rapidly to be able to detect short-lived phenomena such as the aftershocks that may be observable after an underground nuclear explosion. The inspectors will be on site from weeks to months and will be working with many tens of tons of equipment. Parts of the OSI regime will be tested in a field exercise in the country of Jordan late in 2014. The build-up of the OSI regime has been proceeding steadily since the CTBT was signed in 1996 and is on track to becoming a deterrent to someone considering conducting a nuclear explosion in violation of the Treaty.

A restructuring of COR package for MIDAS computer code

International Nuclear Information System (INIS)

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

2004-01-01

The COR package, which calculates the thermal response of the core and the lower plenum internal structures and models the relocation of the core and lower plenum structural materials, 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 a modernized data structure. To do this, the data transferring methods of the current MELCOR code are modified and adopted into the COR package. The data structure of the current MELCOR code using FORTRAN77 has a difficulty in grasping the meaning of the variables as well as a waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which leads to an efficient memory treatment and an easy understanding of the code. Restructuring of the COR package addressed in this paper includes a module development, subroutine modification. The verification has been done by comparing the results of the modified code with those of the existing code. As the trends are similar to each other, it implies that the same approach could be extended to the entire code package. It is expected that the code restructuring will accelerated the code's domestication thanks to a direct understanding of each variable and an easy implementation of the modified or newly developed models. (author)

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

Results of the independent radiological verification survey at 4400 Piehl Road, Ottawa Lake, Michigan (BTO002)

Energy Technology Data Exchange (ETDEWEB)

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

1996-04-01

At the request of the US Department of Energy (DOE), a team from Oak Ridge National Laboratory (ORNL) conducted an independent radiological verification survey at Ottawa Lake, Michigan. The survey was performed in November and December of 1994. The purpose of the survey was to verify that the site was remediated to levels below the DOE guidelines for FUSRAP sites. Results of the independent radiological verification survey at Ottawa Lake, Michigan confirm that the residual uranium contamination at the site is below DOE FUSRAP guidelines for unrestricted use.

System design package for SIMS prototype system 4, solar heating and domestic hot water

Science.gov (United States)

1978-01-01

The system consisted of a modular designed prepackaged solar unit, containing solar collectors, a rock storage container, blowers, dampers, ducting, air-to-water heat exchanger, DHW preheat tank, piping, and system controls. The system was designed to be installed adjacent to a small single family dwelling. The description, performance specification, subsystem drawings, verification plan/procedure, and hazard analysis of the system were packaged for evaluation.

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

International Nuclear Information System (INIS)

Komann, Steffen; Groeke, Carsten; Droste, Bernhard

2013-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

Status of reactor core design code system in COSINE code package

Energy Technology Data Exchange (ETDEWEB)

Chen, Y.; Yu, H.; Liu, Z., E-mail: yuhui@snptc.com.cn [State Nuclear Power Software Development Center, SNPTC, National Energy Key Laboratory of Nuclear Power Software (NEKLS), Beijiing (China)

2014-07-01

For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

Status of reactor core design code system in COSINE code package

International Nuclear Information System (INIS)

Chen, Y.; Yu, H.; Liu, Z.

2014-01-01

For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

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.

DNA Topology and the Initiation of Virus DNA Packaging.

Directory of Open Access Journals (Sweden)

Choon Seok Oh

Full Text Available During progeny assembly, viruses selectively package virion genomes from a nucleic acid pool that includes host nucleic acids. For large dsDNA viruses, including tailed bacteriophages and herpesviruses, immature viral DNA is recognized and translocated into a preformed icosahedral shell, the prohead. Recognition involves specific interactions between the viral packaging enzyme, terminase, and viral DNA recognition sites. Generally, viral DNA is recognized by terminase's small subunit (TerS. The large terminase subunit (TerL contains translocation ATPase and endonuclease domains. In phage lambda, TerS binds a sequence repeated three times in cosB, the recognition site. TerS binding to cosB positions TerL to cut the concatemeric DNA at the adjacent nicking site, cosN. TerL introduces staggered nicks in cosN, generating twelve bp cohesive ends. Terminase separates the cohesive ends and remains bound to the cosB-containing end, in a nucleoprotein structure called Complex I. Complex I docks on the prohead's portal vertex and translocation ensues. DNA topology plays a role in the TerSλ-cosBλ interaction. Here we show that a site, I2, located between cosN and cosB, is critically important for an early DNA packaging step. I2 contains a complex static bend. I2 mutations block DNA packaging. I2 mutant DNA is cut by terminase at cosN in vitro, but in vivo, no cos cleavage is detected, nor is there evidence for Complex I. Models for what packaging step might be blocked by I2 mutations are presented.

The packaging of intermediate and low level radioactive wastes

International Nuclear Information System (INIS)

Flowers, R.H.

1985-01-01

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

Integrated performance assessment model for waste package behavior and radionuclide release

International Nuclear Information System (INIS)

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

1992-01-01

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

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

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

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

Data storage accounting and verification at LHC experiments

Energy Technology Data Exchange (ETDEWEB)

Huang, C. H. [Fermilab; Lanciotti, E. [CERN; Magini, N. [CERN; Ratnikova, N. [Moscow, ITEP; Sanchez-Hernandez, A. [CINVESTAV, IPN; Serfon, C. [Munich U.; Wildish, T. [Princeton U.; Zhang, X. [Beijing, Inst. High Energy Phys.

2012-01-01

All major experiments at the Large Hadron Collider (LHC) need to measure real storage usage at the Grid sites. This information is equally important for resource management, planning, and operations. To verify the consistency of central catalogs, experiments are asking sites to provide a full list of the files they have on storage, including size, checksum, and other file attributes. Such storage dumps, provided at regular intervals, give a realistic view of the storage resource usage by the experiments. Regular monitoring of the space usage and data verification serve as additional internal checks of the system integrity and performance. Both the importance and the complexity of these tasks increase with the constant growth of the total data volumes during the active data taking period at the LHC. The use of common solutions helps to reduce the maintenance costs, both at the large Tier1 facilities supporting multiple virtual organizations and at the small sites that often lack manpower. We discuss requirements and solutions to the common tasks of data storage accounting and verification, and present experiment-specific strategies and implementations used within the LHC experiments according to their computing models.

System design package for IBM system one: solar heating and domestic hot water

Science.gov (United States)

1977-01-01

This report is a collation of documents and drawings that describe a prototype solar heating and hot water system using air as the collector fluid and a pebble bed for heat storage. The system was designed for installation into a single family dwelling. The description, performance specification, subsystem drawings, verification plan/procedure, and hazard analysis of the system was packaged for evaluation of the system with information sufficient to assemble a similar system.

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

VAMOS: The verification and monitoring options study: Current research options for in-situ monitoring and verification of contaminant remediation and containment within the vadose zone

International Nuclear Information System (INIS)

Betsill, J.D.; Gruebel, R.D.

1995-09-01

The Verification and Monitoring Options Study Project (VAMOS) was established to identify high-priority options for future vadose-zone environmental research in the areas of in-situ remediation monitoring, post-closure monitoring, and containment emplacement and verification monitoring. VAMOS examined projected needs not currently being met with applied technology in order to develop viable monitoring and verification research options. The study emphasized a compatible systems approach to reinforce the need for utilizing compatible components to provide user friendly site monitoring systems. To identify the needs and research options related to vadose-zone environmental monitoring and verification, a literature search and expert panel forums were conducted. The search included present drivers for environmental monitoring technology, technology applications, and research efforts. The forums included scientific, academic, industry, and regulatory environmental professionals as well as end users of environmental technology. The experts evaluated current and future monitoring and verification needs, methods for meeting these needs, and viable research options and directions. A variety of high-priority technology development, user facility, and technology guidance research options were developed and presented as an outcome of the literature search and expert panel forums

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, k_s value, and the cementitious leachate impact factor.

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.

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

International Nuclear Information System (INIS)

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

1993-01-01

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

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

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 ...

The role of waste package specifications as a forerunner to ILW repository conditions for acceptance

International Nuclear Information System (INIS)

Barlow, S.V.; Palmer, J.D.

1998-01-01

In the absence of a finalized repository site, design or associated safety case, Nirex is not in a position to issue conditions for acceptance. Nirex has therefore developed a strategy which facilitates packaging of intermediate level waste by providing guidance through waste package specifications, supported by the formal assessment of specific packaging proposals on a case-by-case basis. The waste package specifications are comprehensive and cover all aspects of the waste package including dimensions and other key features, performance standards, wasteform, quality assurance, and data recording requirements. The waste package specifications will be subject to periodic review as repository design and safety cases are finalized and will progressively become site- and design-specific. The waste package specifications will eventually form the basis for conditions for acceptance. The strategy described in this paper has been successfully followed by Nirex and customers for the past ten years and has permitted wastes to be packaged for a deep repository with confidence in the absence of a finalized site and safety cases for the repository. Because the process has its basis in a generic repository concept, it remains robust, despite the increased uncertainty following the March 1997 Secretary of State's decision, as to the siting and time-scale of a deep waste repository, and continues to be an important component of the UK's waste management strategy. (author)

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

International Nuclear Information System (INIS)

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

1992-01-01

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

The packaging and transport of low and intermediate level radioactive wastes

International Nuclear Information System (INIS)

Grover, J.R.; Price, M.S.T.

1985-01-01

Up to the present time, the majority of the radioactive waste which has been transported in the United Kingdom has been low level waste for disposal in the trenches of the shallow burial site operated by British Nuclear Fuels plc at Drigg and also the packaged waste destined for sea disposal in the annual operation. However, the main bulk of the low and intermediate level wastes which have been generated over the last quarter century remain in store at the various nuclear sites where it originated. Before significant packaging and transport of intermediate level wastes takes place it is desirable to examine the sources and types of wastes, the immobilisation and packaging processes and plants, the transport, and the problems of handling of packages at future land repositories. Optimisation of the packaging and transport must take account of both the upstream and downstream con=straints as well as the implications of complying with both the IAEA Transport Regulations and radiological protection guidelines. Packages for sea disposal must in addition comply with the requirements of the London Dumping Convention and the NEA guidelines. (author)

SNF verification requirements imposed through 10 CFR Part 961: Interrelationship with MC ampersand A requirements for the OCRWM safeguards system

International Nuclear Information System (INIS)

Slater, N.; Vance, S.

1994-01-01

Article VI.B.2 of the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (10 CFR Part 961, Standard Contract) provides that spent nuclear fuel shall be subject to verification prior to acceptance by the Department of Energy. As part of the overall process for scheduling deliveries of spent nuclear fuel, the Standard Contract requires contract holders to submit detailed descriptions of the spent nuclear fuel they intend to deliver. Thus, the provision for verification in the Standard contract allows the Department the opportunity to ensure that the spent nuclear fuel intended for deliver is consistent with the description provided, and that the material is properly loaded, packaged, marked, and labeled. The Department is currently evaluating the verification requirements for spent nuclear fuel that it will establish pursuant to Article VI.B.2 of the Standard Contract. The Department recognizes that there may be significant overlap between these verification requirements and the requirements established for material control and accountability (MC ampersand A). Therefore, the Department may recommend modifications to the process established in the Standard Contract to ensure that the verification and MC ampersand A functions and activities are fully integrated

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

Verification of FPGA-Signal using the test board which is applied to Safety-related controller

Energy Technology Data Exchange (ETDEWEB)

Chung, Youn-Hu; Yoo, Kwanwoo; Lee, Myeongkyun; Yun, Donghwa [SOOSAN ENS, Seoul (Korea, Republic of)

2016-10-15

This article aims to provide the verification method for BGA-type FPGA of Programmable Logic Controller (PLC) developed as Safety Class. The logic of FPGA in the control device with Safety Class is the circuit to control overall logic of PLC. Saftety-related PLC must meet the international standard specifications. With this reason, we use V and V according to an international standard in order to secure high reliability and safety. By using this, we are supposed to proceed to a variety of verification courses for extra reliability and safety analysis. In order to have efficient verification of test results, we propose the test using the newly changed BGA socket which can resolve the problems of the conventional socket on this paper. The Verification of processes is divided into verification of Hardware and firmware. That processes are carried out in the unit testing and integration testing. The proposed test method is simple, the effect of cost reductions by batch process. In addition, it is advantageous to measure the signal from the Hi-speed-IC due to its short length of the pins and it was plated with the copper around it. Further, it also to prevent abrasion on the IC ball because it has no direct contact with the PCB. Therefore, it can be actually applied is to the BGA package test and we can easily verify logic as well as easily checking the operation of the designed data.

Type B liquid package technical issues -- Experience with LR-56 safety analysis

International Nuclear Information System (INIS)

Smith, A.C.; Alstine, M.N. van; Gromada, R.J.; Hensel, S.J.; Gupta, N.K.

1997-01-01

In the course of the development of nuclear industry in France, shipment of Type B quantities (i.e., quantities having significant radiological consequences) of radioactive liquids between different, sites became necessary. Based on the experience acquired at the Commissariat a l'Energie Atomique (CEA) nuclear centers, a series of tanker trailers has been developed to meet this need. Similarly, as part of the ongoing program to process wastes to stable end forms, a need exists to move radioactive liquids at several DOE sites. The LR-56, developed by CEA to transport liquids of medium to high activity, was selected for these US applications, based on its design features and successful operating experience in France. No comparable Type B liquid packages are certified in the US Packages employed in transport of Type B quantities of liquids are either only suitable for small volumes, or are used within site boundaries with extensive administrative controls employed to insure that an adequate level of safety is maintained. The requirement is to provide safety equivalent to the level established by federal regulations in 10 CFR 71. Type B radioactive materials packages (RAM packages) are typically simple, rugged containers which are designed and fabricated in accordance with the ASME Boiler and Pressure Vessel Code to provide containment under the normal conditions of transport (NCT) and hypothetical accident conditions (HAC) established by the regulations. Packages designed for liquid contents must address a number of technical issues which are not common to packages for solid contents. This paper reviews the technical issues associated with Type B liquid packages from the perspective of the experience gained from the evaluation of the LR-56 for use at DOE sites

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.

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

System Design Package for SIMS Prototype System 3, Solar Heating and Domestic Hot Water

Science.gov (United States)

1978-01-01

A collation of documents and drawings are presented that describe a prototype solar heating and hot water system using liquid flat plate collectors and a gas or electric furnace energy subsystem. The system was designed for installation into a single-family dwelling. The description, performance specification, subsystem drawings, verification plan/procedure, and hazard analysis of the system are packaged for evaluation of the system with information sufficient to assemble a similar system.

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

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.)

IAEA verification of materials accounting in commercial reprocessing plants

International Nuclear Information System (INIS)

Gutmacher, R.G.; Hakkila, E.A.

1987-01-01

The reprocessing plants currently under International Atomic Energy Agency (IAEA) safeguards have design capacities up to 210 tonnes of heavy metal per year. All of the plants use conventional materials accounting for safeguards. However, several larger commercial reprocessing plants are being designed with capacities of 350 to 1200 tonnes of heavy metal per year. It is likely that many of these plants, as well as some of the existing smaller ones, will adopt near-real-time materials accounting. The major effect of the combination of larger plants and near-real-time accounting on IAEA safeguards will be the demand for greater timeliness of verification. Continuous inspector presence may be required, as well as more on-site measurements by the inspector. In this paper, the authors review what needs to be verified, as well as current inspector activities in the process area. The bulk of the paper describes rapid, easy-to-use measurement techniques and instruments that may be applied to on-site verification measurements

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)

A Restructuring of the HS Package for the MIDAS Computer Code

International Nuclear Information System (INIS)

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

2005-01-01

As one of the processes for a localized severe accident analysis code, KAERI is developing a severe accident code MIDAS, based on MELCOR. The existing data saving method uses pointer variables for a fix-sized storage management, and it deteriorates the readability, maintainability and portability of the code. But new features in FORTRAN90 such as a dynamic allocation have been used for the restructuring. The restructuring of the data saving and transferring method of the existing code makes it easy to understand the code. Before an entire restructuring of the code, a restructuring template for a simple package was developed and tested. The target for the restructuring was the HS package which is responsible for calculation the heat conduction within an intact, solid structure and energy transfer across its boundary surfaces into control volumes. The verification was done through comparing the results before and after the restructuring

The Australian experience following plain packaging: the impact on tobacco branding.

Science.gov (United States)

Greenland, Steven J

2016-12-01

Brands are critical to tobacco marketing. Industry stakeholders predicted that plain packaging, by removing key tangible branding dimensions, would restrict new products and brand differentiation. However, manufacturers respond innovatively to limit regulatory impact. This study investigates brand strategy following plain packaging's introduction to Australia. Brand portfolios were determined using 2006-15 tobacco ingredient reports. These detail the brand and variant names sold and are provided annually as part of a voluntary agreement between the Australian Government and leading manufacturers. Post-plain packaging brand ranges were verified using retail price lists and a supermarket retail audit using a method used previously to verify a period of pre-plain packaging data. The verification process identified some data inaccuracies from one manufacturer which resulted in the issuing of corrected data. After plain packaging the leading manufacturers continued with extensive brand ranges differentiated by price. All launched new products. While total brand numbers fell from 29 to 24, the mean number of variants for the leading 12 brands grew from 8.9 to 9.7. Substantial variant name modifications occurred with 50 new or modified names in 2012-13. Among leading brands, the incidence of variant colour names increased from 49.5 to 79.3%. New brands and variants were not inhibited by the introduction of plain packaging in Australia. After plain packaging, leading brand variant numbers expanded by 9 to 116 and colour variant names increased by 73.6% and became the norm-lighter colours (blue, gold and silver) dominated, perpetuating notions of less harmful cigarettes. [Correction added on 09 September 2016, after first online publication: The figures in the last sentence of the Abstract are now corrected from 'expanded by 116' to 'expanded by 9 to 116'.]. © 2016 Society for the Study of Addiction.

Verification of operating software for cooperative monitoring applications

International Nuclear Information System (INIS)

Tolk, K.M.; Rembold, R.K.

1997-01-01

Monitoring agencies often use computer based equipment to control instruments and to collect data at sites that are being monitored under international safeguards or other cooperative monitoring agreements. In order for this data to be used as an independent verification of data supplied by the host at the facility, the software used must be trusted by the monitoring agency. The monitoring party must be sure that the software has not be altered to give results that could lead to erroneous conclusions about nuclear materials inventories or other operating conditions at the site. The host might also want to verify that the software being used is the software that has been previously inspected in order to be assured that only data that is allowed under the agreement is being collected. A description of a method to provide this verification using keyed has functions and how the proposed method overcomes possible vulnerabilities in methods currently in use such as loading the software from trusted disks is presented. The use of public key data authentication for this purpose is also discussed

Impact of the frequency of online verifications on the patient set-up accuracy and set-up margins

International Nuclear Information System (INIS)

Rudat, Volker; Hammoud, Mohamed; Pillay, Yogin; Alaradi, Abdul Aziz; Mohamed, Adel; Altuwaijri, Saleh

2011-01-01

The purpose of the study was to evaluate the patient set-up error of different anatomical sites, to estimate the effect of different frequencies of online verifications on the patient set-up accuracy, and to calculate margins to accommodate for the patient set-up error (ICRU set-up margin, SM). Alignment data of 148 patients treated with inversed planned intensity modulated radiotherapy (IMRT) or three-dimensional conformal radiotherapy (3D-CRT) of the head and neck (n = 31), chest (n = 72), abdomen (n = 15), and pelvis (n = 30) were evaluated. The patient set-up accuracy was assessed using orthogonal megavoltage electronic portal images of 2328 fractions of 173 planning target volumes (PTV). In 25 patients, two PTVs were analyzed where the PTVs were located in different anatomical sites and treated in two different radiotherapy courses. The patient set-up error and the corresponding SM were retrospectively determined assuming no online verification, online verification once a week and online verification every other day. The SM could be effectively reduced with increasing frequency of online verifications. However, a significant frequency of relevant set-up errors remained even after online verification every other day. For example, residual set-up errors larger than 5 mm were observed on average in 18% to 27% of all fractions of patients treated in the chest, abdomen and pelvis, and in 10% of fractions of patients treated in the head and neck after online verification every other day. In patients where high set-up accuracy is desired, daily online verification is highly recommended

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...

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.

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

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

Novel DNA packaging recognition in the unusual bacteriophage N15

Energy Technology Data Exchange (ETDEWEB)

Feiss, Michael [Department of Microbiology, Roy J. and Lucille A. Carver College of Medicine, University of Iowa, Iowa City, IA 52242 (United States); Geyer, Henriette, E-mail: henriettegeyer@gmail.com [Division of Viral Infections, Robert Koch Institute, Berlin (Germany); Division of Viral Infections, Robert Koch Institute, Berlin (Germany); Klingberg, Franco, E-mail: franco.klingberg@thermofisher.com [Flow Cytometry, Imaging & Microscopy, Thermo Fisher Scientific, Frankfurter Strasse 129B 64293 Darmstadt (Germany); Flow Cytometry, Imaging & Microscopy, Thermo Fisher Scientific, Frankfurter Strasse 129B 64293 Darmstadt (Germany); Moreno, Norma, E-mail: nmoreno@islander.tamucc.edu [Texas A& M University – Corpus Christi, 6300 Ocean Drive, Corpus Christi, TX 78412, United States. (United States); Texas A& M University – Corpus Christi, 6300 Ocean Drive, Corpus Christi, TX 78412, United States. (United States); Forystek, Amanda, E-mail: eamanda-forystek@uiowa.edu [Flow Cytometry, Imaging & Microscopy, Thermo Fisher Scientific, Frankfurter Strasse 129B 64293 Darmstadt (Germany); Room # 2911 JPP, Dept. of Psychiatry, The University of Iowa, 200 Hawkins Drive, Iowa City, Iowa, 52242 (United States); Maluf, Nasib Karl, E-mail: fKarl.Maluf@ap-lab.com [Flow Cytometry, Imaging & Microscopy, Thermo Fisher Scientific, Frankfurter Strasse 129B 64293 Darmstadt (Germany); Alliance Protein Laboratories, Inc. 6042 Cornerstone Court West, Suite ASan Diego, CA 92121, USA. (United States); Sippy, Jean [Department of Microbiology, Roy J. and Lucille A. Carver College of Medicine, University of Iowa, Iowa City, IA 52242 (United States)

2015-08-15

Phage lambda's cosB packaging recognition site is tripartite, consisting of 3 TerS binding sites, called R sequences. TerS binding to the critical R3 site positions the TerL endonuclease for nicking cosN to generate cohesive ends. The N15 cos (cos{sup N15}) is closely related to cos{sup λ}, but whereas the cosB{sup N15} subsite has R3, it lacks the R2 and R1 sites and the IHF binding site of cosB{sup λ}. A bioinformatic study of N15-like phages indicates that cosB{sup N15} also has an accessory, remote rR2 site, which is proposed to increase packaging efficiency, like R2 and R1 of lambda. N15 plus five prophages all have the rR2 sequence, which is located in the TerS-encoding 1 gene, approximately 200 bp distal to R3. An additional set of four highly related prophages, exemplified by Monarch, has R3 sequence, but also has R2 and R1 sequences characteristic of cosB–λ. The DNA binding domain of TerS-N15 is a dimer. - Highlights: • There are two classes of DNA packaging signals in N15-related phages. • Phage N15's TerS binding site: a critical site and a possible remote accessory site. • Viral DNA recognition signals by the λ-like bacteriophages: the odd case of N15.

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

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)

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)

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 ...

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

Cre/loxP-mediated adenovirus type 5 packaging signal excision demonstrates that core element VI is sufficient for virus packaging

International Nuclear Information System (INIS)

Maeda, Yasushi; Kimura, En; Uchida, Yuji; Nishida, Yasuto; Yamashita, Satoshi; Arima, Toshiyuki; Uchino, Makoto

2003-01-01

Previous analyses have demonstrated that packaging of the adenovirus type 5 (Ad5) genome is dependent on at least seven cis-acting elements, called AI to AVII, which are located in the left-end region of the genome. These elements have different packaging efficiencies, and without AI through AV, viral DNA cannot be packaged. Here we report the identification of the cis-acting Ad5 packaging domain in vivo by using the Cre/loxP system. We found that an adenoviral DNA fragment (nt 192 to nt 358), which includes elements AI to AV, is excised by Cre recombinase and packaged into capsids. Furthermore, this mutant adenovirus replicated so efficiently by repetitive propagation that its purification by CsCI equilibrium gradient was possible. This study clarified that the region from nt 358 to nt 454 on the viral genome is sufficient for packaging. Recently, the helper-dependent adenoviral vector (HDAd) construction system has been developed for the purpose of gene therapy. This system uses a helper virus with two parallel loxP sites flanking the packaging signal. This region is eliminated by Cre-mediated excision, which prevents helper virus packaging. Our data provide useful information regarding factors affecting efficient elimination

The US National Resources Defense Council/Soviet Academy of Sciences Nuclear Test Ban Verification Project

International Nuclear Information System (INIS)

Cochran, T.B.

1989-01-01

The first week in September 1987 was an extraordinary one for arms control verification. As part of the co-operative Test Ban Verification Project of the Natural Resources Defense Council (NRDC) and the Soviet Academy of Sciences, fourteen American scientists from the Scripps Institution of Oceanography (at the University of California- San Diego), University of Nevada-Reno and the University of Colorado went to the region of the Soviet's principal nuclear test site near Semipalatinsk. Together with their Soviet counterparts from the Institute of Physics of the Earth (IPE) in Moscow, they fired off three large chemical explosions. The purpose of these explosions was to demonstrate the sensitivity of the three seismic stations surrounding the test site, to study the efficiency with which high-frequency seismic waves propagate in the region, and to study differences between chemical explosions, nuclear explosions and earthquakes in order more firmly to establish procedures for verification of a nuclear test ban. This paper presents a review of the results of these experiments, an update on the status of the joint project, and a review of the significance of high frequency seismic data to test ban verification

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.

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Impact of the frequency of online verifications on the patient set-up accuracy and set-up margins

Directory of Open Access Journals (Sweden)

Mohamed Adel

2011-08-01

Full Text Available Abstract Purpose The purpose of the study was to evaluate the patient set-up error of different anatomical sites, to estimate the effect of different frequencies of online verifications on the patient set-up accuracy, and to calculate margins to accommodate for the patient set-up error (ICRU set-up margin, SM. Methods and materials Alignment data of 148 patients treated with inversed planned intensity modulated radiotherapy (IMRT or three-dimensional conformal radiotherapy (3D-CRT of the head and neck (n = 31, chest (n = 72, abdomen (n = 15, and pelvis (n = 30 were evaluated. The patient set-up accuracy was assessed using orthogonal megavoltage electronic portal images of 2328 fractions of 173 planning target volumes (PTV. In 25 patients, two PTVs were analyzed where the PTVs were located in different anatomical sites and treated in two different radiotherapy courses. The patient set-up error and the corresponding SM were retrospectively determined assuming no online verification, online verification once a week and online verification every other day. Results The SM could be effectively reduced with increasing frequency of online verifications. However, a significant frequency of relevant set-up errors remained even after online verification every other day. For example, residual set-up errors larger than 5 mm were observed on average in 18% to 27% of all fractions of patients treated in the chest, abdomen and pelvis, and in 10% of fractions of patients treated in the head and neck after online verification every other day. Conclusion In patients where high set-up accuracy is desired, daily online verification is highly recommended.

Synergies across verification regimes: Nuclear safeguards and chemical weapons convention compliance

International Nuclear Information System (INIS)

Kadner, Steven P.; Turpen, Elizabeth

2001-01-01

In the implementation of all arms control agreements, accurate verification is essential. In setting a course for verifying compliance with a given treaty - whether the NPT or the CWC, one must make a technical comparison of existing information-gathering capabilities against the constraints in an agreement. Then it must be decided whether this level of verifiability is good enough. Generally, the policy standard of 'effective verification' includes the ability to detect significant violations, with high confidence, in sufficient time to respond effectively with policy adjustments or other responses, as needed. It is at this juncture where verification approaches have traditionally diverged. Nuclear safeguards requirements have taken one path while chemical verification methods have pursued another. However, recent technological advances have brought a number of changes affecting verification, and lately their pace has been accelerating. First, all verification regimes have more and better information as a result of new kinds of sensors, imagery, and other technologies. Second, the verification provisions in agreements have also advanced, to include on-site inspections, portal monitoring, data exchanges, and a variety of transparency, confidence-building, and other cooperative measures, Together these developments translate into a technological overlap of certain institutional verification measures such as the NPT's safeguards requirements and the IAEA and the CWC's verification visions and the OPCW. Hence, a priority of international treaty-implementing organizations is exploring the development of a synergistic and coordinated approach to WMD policy making that takes into account existing inter-linkages between nuclear, chemical, and biological weapons issues. Specific areas of coordination include harmonizing information systems and information exchanges and the shared application of scientific mechanisms, as well as collaboration on technological developments

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.

REVIEW OF AGING DATA ON EPDM O-RINGS IN THE H1616 SHIPPING PACKAGE

Energy Technology Data Exchange (ETDEWEB)

Skidmore, E.

2012-03-27

Currently, all H1616 shipping package containers undergo annual re-verification testing, including containment vessel leak testing to verify leak-tightness (<1 x 10{sup -7} ref cc/sec air) as per ANSI N14.5. The purpose of this literature review is to supplement aging studies currently being performed by SRNL on the EPDM O-rings to provide the technical basis for extending annual re-verification testing for the H1616 shipping package and to predict the life of the seals at bounding service conditions. The available data suggest that the EPDM O-rings can retain significant mechanical properties and sealing force at or below bounding service temperatures (169 F or 76 C) beyond the 1 year maintenance period. Interpretation of available data suggests that a service life of at least 2 years and potentially 4-6 years may be possible at bounding temperatures. Seal lifetimes at lower, more realistic temperatures will likely be longer. Being a hydrocarbon elastomer, EPDM O-rings may exhibit an inhibition period due to the presence of antioxidants. Once antioxidants are consumed, mechanical properties and seal performance could decline at a faster rate. Testing is being performed to validate the assumptions outlined in this report and to assess the long-term performance of O-ring seals under actual service conditions.

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

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

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...

SU-E-T-49: A Multi-Institutional Study of Independent Dose Verification for IMRT

International Nuclear Information System (INIS)

Baba, H; Tachibana, H; Kamima, T; Takahashi, R; Kawai, D; Sugawara, Y; Yamamoto, T; Sato, A; Yamashita, M

2015-01-01

Purpose: AAPM TG114 does not cover the independent verification for IMRT. We conducted a study of independent dose verification for IMRT in seven institutes to show the feasibility. Methods: 384 IMRT plans in the sites of prostate and head and neck (HN) were collected from the institutes, where the planning was performed using Eclipse and Pinnacle3 with the two techniques of step and shoot (S&S) and sliding window (SW). All of the institutes used a same independent dose verification software program (Simple MU Analysis: SMU, Triangle Product, Ishikawa, JP), which is Clarkson-based and CT images were used to compute radiological path length. An ion-chamber measurement in a water-equivalent slab phantom was performed to compare the doses computed using the TPS and an independent dose verification program. Additionally, the agreement in dose computed in patient CT images between using the TPS and using the SMU was assessed. The dose of the composite beams in the plan was evaluated. Results: The agreement between the measurement and the SMU were −2.3±1.9 % and −5.6±3.6 % for prostate and HN sites, respectively. The agreement between the TPSs and the SMU were −2.1±1.9 % and −3.0±3.7 for prostate and HN sites, respectively. There was a negative systematic difference with similar standard deviation and the difference was larger in the HN site. The S&S technique showed a statistically significant difference between the SW. Because the Clarkson-based method in the independent program underestimated (cannot consider) the dose under the MLC. Conclusion: The accuracy would be improved when the Clarkson-based algorithm should be modified for IMRT and the tolerance level would be within 5%

Active Mirror Predictive and Requirements Verification Software (AMP-ReVS)

Science.gov (United States)

Basinger, Scott A.

2012-01-01

This software is designed to predict large active mirror performance at various stages in the fabrication lifecycle of the mirror. It was developed for 1-meter class powered mirrors for astronomical purposes, but is extensible to other geometries. The package accepts finite element model (FEM) inputs and laboratory measured data for large optical-quality mirrors with active figure control. It computes phenomenological contributions to the surface figure error using several built-in optimization techniques. These phenomena include stresses induced in the mirror by the manufacturing process and the support structure, the test procedure, high spatial frequency errors introduced by the polishing process, and other process-dependent deleterious effects due to light-weighting of the mirror. Then, depending on the maturity of the mirror, it either predicts the best surface figure error that the mirror will attain, or it verifies that the requirements for the error sources have been met once the best surface figure error has been measured. The unique feature of this software is that it ties together physical phenomenology with wavefront sensing and control techniques and various optimization methods including convex optimization, Kalman filtering, and quadratic programming to both generate predictive models and to do requirements verification. This software combines three distinct disciplines: wavefront control, predictive models based on FEM, and requirements verification using measured data in a robust, reusable code that is applicable to any large optics for ground and space telescopes. The software also includes state-of-the-art wavefront control algorithms that allow closed-loop performance to be computed. It allows for quantitative trade studies to be performed for optical systems engineering, including computing the best surface figure error under various testing and operating conditions. After the mirror manufacturing process and testing have been completed, the

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

Radiological verification survey results at 14 Peck Ave., Pequannock, New Jersey (PJ001V)

International Nuclear Information System (INIS)

Rodriguez, R.E.; Johnson, C.A.

1995-05-01

The U.S. Department of Energy (DOE) conducted remedial action during 1993 at the Pompton Plains Railroad Spur and eight vicinity properties in the Wayne and Pequannock Townships in New Jersey as part of the Formerly Utilized Sites Remedial Action Program (FUSRAP). These properties are in the vicinity of the DOE-owned Wayne Interim Storage Site (WISS), formerly the W. R. Grace facility. The property at 14 Peck Ave., Pequannock, New Jersey is one of these vicinity properties. At the request of DOE, a team from Oak Ridge National Laboratory conducted an independent radiological verification survey at this property. The purpose of the survey, conducted between September and December 1993, was to confirm the success of the remedial actions performed to remove any radioactive materials in excess of the identified guidelines. The verification survey included surface gamma scans and gamma readings at 1 meter, beta-gamma scans, and the collection of soil and debris samples for radionuclide analysis. Results of the survey demonstrated that all radiological measurements on the property at 14 Peck Ave. were within applicable DOE guidelines. Based on the results of the remedial action data and confirmed by the verification survey data, the portions of the site that had been remediated during this action successfully meet the DOE remedial action objectives

Radiological verification survey results at 3 Peck Ave., Pequannock, New Jersey (PJ002V)

International Nuclear Information System (INIS)

Rodriguez, R.E.; Johnson, C.A.

1995-05-01

The US Department of Energy (DOE) conducted remedial action during 1993 at the Pompton Plains Railroad Spur and eight vicinity properties in the Wayne and Pequannock Townships in New Jersey as part of the Formerly Utilized Sites Remedial Action Program (FUSRAP). These properties are in the vicinity of the DOE-owned Wayne Interim Storage Site (WISS), formerly the W.R. Grace facility. The property at 3 Peck Ave., Pequannock, New Jersey is one of these vicinity properties. At the request of DOE, a team from Oak Ridge National Laboratory conducted an independent radiological verification survey at this property. The purpose of the survey, conducted between September and December 1993, was to confirm the success of the remedial actions performed to remove any radioactive materials in excess of the identified guidelines. The verification survey included surface gamma scans and gamma readings at 1 meter, beta-gamma scans, and the collection of soil and debris samples for radionuclide analysis. Results of the survey demonstrated that all radiological measurements on the property at 3 Peck Ave. were within applicable DOE guidelines. Based on the results of the remedial action data and confirmed by the verification survey data, the portions of the site that had been remediated during this action successfully meet the DOE remedial action objectives

Radiological verification survey results at 15 Peck Ave., Pequannock, New Jersey (PJ005V)

International Nuclear Information System (INIS)

Rodriguez, R.E.; Johnson, C.A.

1995-05-01

The US Department of Energy (DOE) conducted remedial action during 1993 at the Pompton Plains Railroad Spur and eight vicinity properties in the Wayne and Pequannock Townships in New Jersey as part of the Formerly Utilized Sites Remedial Action Program (FUSRAP). These properties are in the vicinity of the DOE-owned Wayne Interim Storage Site (WISS), formerly the W.R. Grace facility. The property at 15 Peck Ave., Pequannock, New Jersey is one of these vicinity properties. At the request of DOE, a team from Oak Ridge National Laboratory conducted an independent radiological verification survey at this property. The purpose of the survey, conducted between September and December 1993, was to confirm the success of the remedial actions performed to remove any radioactive materials in excess of the identified guidelines. The verification survey included surface gamma scans and gamma readings at 1 meter, beta-gamma scans, and the collection of soil and debris samples for radionuclide analysis. Results of the survey demonstrated that all radiological measurements on the property at 15 Peck Ave. were within applicable DOE guidelines. Based on the results of the remedial action data and confirmed by the verification survey data, the portions of the site that had been remediated during this action successfully meet the DOE remedial action objectives

Radiological verification survey results at 17 Peck Ave., Pequannock, New Jersey (PJ006V)

International Nuclear Information System (INIS)

Rodriguez, R.E.; Johnson, C.A.

1995-05-01

The US Department of Energy (DOE) conducted remedial action during 1993 at the Pompton Plains Railroad Spur and eight vicinity properties in the Wayne and Pequannock Townships in New Jersey as part of the Formerly Utilized Sites Remedial Action Program (FUSRAP). These properties are in the vicinity of the DOE-owned Wayne Interim Storage Site (WISS), formerly the W.R. Grace facility. The property at 17 Peck Ave., Pequannock, New Jersey is one of these vicinity properties. At the request of DOE, a team from Oak Ridge National Laboratory conducted an independent radiological verification survey at this property. The purpose of the survey, conducted between September and December 1993, was to confirm the success of the remedial actions performed to remove any radioactive materials in excess of the identified guidelines. The verification survey included surface gamma scans and gamma readings at 1 meter, beta-gamma scans, and the collection of soil and debris samples for radionuclide analysis. Results of the survey demonstrated that all radiological measurements on the property at 17 Peck Ave. were within applicable DOE guidelines. Based on the results of the remedial action data and confirmed by the verification survey data, the portions of the site that had been remediated during this action successfully meet the DOE remedial action objectives

Radiological verification survey results at 7 Peck Ave., Pequannock, New Jersey (PJ003V)

International Nuclear Information System (INIS)

Rodriguez, R.E.; Johnson, C.A.

1995-05-01

The US Department of Energy (DOE) conducted remedial action during 1993 at the Pompton Plains Railroad Spur and eight vicinity properties in the Wayne and Pequannock Townships in New Jersey as part of the Formerly Utilized Sites Remedial Action Program (FUSRAP). These properties are in the vicinity of the DOE-owned Wayne Interim Storage Site (WISS), formerly the W.R. Grace facility. The property at 7 Peck Ave., Pequannock, New Jersey is one of these vicinity properties. At the request of DOE, a team from Oak Ridge National Laboratory conducted an independent radiological verification survey at this property. The purpose of the survey, conducted between September and December 1993, was to confirm the success of the remedial actions performed to remove any radioactive materials in excess of the identified guidelines. The verification survey included surface gamma scans and gamma readings at 1 meter, beta-gamma scans, and the collection of soil samples for radionuclide analysis. Results of the survey demonstrated that all radiological measurements on the property at 7 Peck Ave. were within applicable DOE guidelines. Based on the results of the remedial action data and confirmed by the verification survey data, the portions of the site that had been remediated during this action successfully meet the DOE remedial action objectives

Radiological verification survey results as 13 Peck Ave., Pequannock, New Jersey (PJ004V)

International Nuclear Information System (INIS)

Rodriguez, R.E.; Johnson, C.A.

1995-05-01

The US Department of Energy (DOE) conducted remedial action during 1993 at the Pompton Plains Railroad Spur and eight vicinity properties in the Wayne and Pequannock Townships in New Jersey as part of the Formerly Utilized Sites Remedial Action Program (FUSRAP). These properties are in the vicinity of the DOE-owned Wayne Interim Storage Site (WISS), formerly the W.R. Grace facility. The property at 13 Peck Ave., Pequannock, New Jersey is one of these vicinity properties. At the request of DOE, a team from Oak Ridge National Laboratory conducted an independent radiological verification survey at this property. The purpose of the survey, conducted between September and December 1993, was to confirm the success of the remedial actions performed to remove any radioactive materials in excess of the identified guidelines. The verification survey included surface gamma scans and gamma readings at 1 meter, beta-gamma scans, and the collection of soil and debris samples for radionuclide analysis. Results of the survey demonstrated that all radiological measurements on the property at 13 Peck Ave. were within applicable DOE guidelines. Based on the results of the remedial action data and confirmed by the verification survey data, the portions of the site that had been remediated during this action successfully meet the DOE remedial action objectives

A restructuring of the MELCOR fission product packages for the MIDAS computer code

International Nuclear Information System (INIS)

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

2004-01-01

The RN1/RN2 packages, which are the fission product-related packages in MELCOR, have 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 a modernized data structure. To do this, the data transferring methods of the current MELCOR code are modified and adopted into the RN1/RN2 package. The data structure of the current MELCOR code using FORTRAN77 has a difficulty in grasping the meaning of the variables as well as waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to user-defined data type, which leads to an efficient memory treatment and an easy understanding of the code. Restructuring of the RN1/RN2 package addressed in this paper includes a module development, subroutine modification, and the treatment of MELGEN, which generates the data file, as well as MELCOR, which is processing the calculation. The verification has been done by comparing the results of the modified code with those of the existing code. As the trends are similar to each other, it implies that the same approach could be extended to the entire code package. It is expected that the code restructuring will accelerate the code domestication thanks to a direct understanding of each variable and an easy implementation of the modified or newly developed models. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-11-15

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Conceptual waste packaging options for deep borehole disposal

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

-profile threaded connections at each end. The internal-flush design would be suitable for loading waste that arrives from the originating site in weld-sealed, cylindrical canisters. Internal, tapered plugs with sealing filet welds would seal the tubing at each end. The taper would be precisely machined onto both the tubing and the plug, producing a metal-metal sealing surface that is compressed as the package is subjected to hydrostatic pressure. The lower plug would be welded in place before loading, while the upper plug would be placed and welded after loading. Conceptual Waste Packaging Options for Deep Borehole Disposal July 30, 2015 iv Threaded connections between packages would allow emplacement singly or in strings screwed together at the disposal site. For emplacement on a drill string the drill pipe would be connected directly into the top package of a string (using an adapter sub to mate with premium semi-flush tubing threads). Alternatively, for wireline emplacement the same package designs could be emplaced singly using a sub with wireline latch, on the upper end. Threaded connections on the bottom of the lowermost package would allow attachment of a crush box, instrumentation, etc.

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.)

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.

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

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).

Remote sensing and geoinformation technologies in support of nuclear non-proliferation and arms control verification regimes

Energy Technology Data Exchange (ETDEWEB)

Niemeyer, Irmgard [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung, IEK-6: Nukleare Entsorgung und Reaktorsicherheit (Germany)

2013-07-01

A number of international agreements and export control regimes have been concluded in order to reduce the risk and proliferation of weapons of mass destruction. In order to provide confidence that Member States are complying with the agreed commitments, most of the treaties and agreements include verification provisions. Different types of verification measures exist, e.g. cooperative measures; national technical means; technical monitoring or measurement devices placed at or near sites; on-site inspections; intelligence information; open-source information, such as commercial internet data and satellite imagery. The study reviews the technical progress in the field of satellite imaging sensors and explores the recent advances in satellite imagery processing and geoinformation technologies as to the extraction of significant observables and signatures. Moreover, it discusses how satellite data and geoinformation technologies could be used complementary for confirming information gathered from other systems or sources. The study also aims at presenting the legal and political aspects and the cost benefits of using imagery from both national and commercial satellites in the verification procedure. The study concludes that satellite imagery and geoinformation technologies are expected to enhance the verification efficiency and effectiveness.

Societal Verification: Intellectual Game or International Game-Changer

International Nuclear Information System (INIS)

Hartigan, Kelsey; Hinderstein, Corey

2013-01-01

Within the nuclear nonproliferation and arms control field, there is an increasing appreciation for the potential of open source information technologies to supplement existing verification and compliance regimes. While clearly not a substitute for on-site inspections or national technical means, it may be possible to better leverage information gleaned from commercial satellite imagery, international trade records and the vast amount of data being exchanged online and between publics (including social media) so as to develop a more comprehensive set of tools and practices for monitoring and verifying a state’s nuclear activities and helping judge compliance with international obligations. The next generation “toolkit” for monitoring and verifying items, facility operations and activities will likely include a more diverse set of analytical tools and technologies than are currently used internationally. To explore these and other issues, the Nuclear Threat Initiative has launched an effort that examines, in part, the role that emerging technologies and “citizen scientists” might play in future verification regimes. This paper will include an assessment of past proliferation and security “events” and whether emerging tools and technologies would have provided indicators concurrently or in advance of these actions. Such case studies will be instrumental in understanding the reliability of these technologies and practices and in thinking through the requirements of a 21st century verification regime. Keywords: Verification, social media, open-source information, arms control, disarmament.

Final Hanford Site Transuranic (TRU) Waste Characterization QA Project Plan

International Nuclear Information System (INIS)

GREAGER, T.M.

2000-01-01

The Quality Assurance Project Plan (QAPjP) has been prepared for waste characterization activities to be conducted by the Transuranic (TRU) Project at the Hanford Site to meet requirements set forth in the Waste Isolation Pilot Plan (WIPP) Hazardous Waste Facility Permit, 4890139088-TSDF, Attachment B, including Attachments B1 through B6 (WAP) (DOE, 1999a). The QAPjP describes the waste characterization requirements and includes test methods, details of planned waste sampling and analysis, and a description of the waste characterization and verification process. In addition, the QAPjP includes a description of the quality assurance/quality control (QA/QC) requirements for the waste characterization program. Before TRU waste is shipped to the WIPP site by the TRU Project, all applicable requirements of the QAPjP shall be implemented. Additional requirements necessary for transportation to waste disposal at WIPP can be found in the ''Quality Assurance Program Document'' (DOE 1999b) and HNF-2600, ''Hanford Site Transuranic Waste Certification Plan.'' TRU mixed waste contains both TRU radioactive and hazardous components, as defined in the WLPP-WAP. The waste is designated and separately packaged as either contact-handled (CH) or remote-handled (RH), based on the radiological dose rate at the surface of the waste container. RH TRU wastes are not currently shipped to the WIPP facility

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

International Nuclear Information System (INIS)

Scheglov, A.; Proselkov, V.

2001-01-01

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

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

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.

Site characterization plan:

International Nuclear Information System (INIS)

1988-01-01

The Yucca Mountain site in Nevada is one of three candidate sites for the first geologic repository for radioactive waste. On May 28, 1986, it was recommended for detailed study in a program of site characterization. This site characterization plan (SCP) has been prepared in accordance with the requirements of the Nuclear Waste Policy Act to summarize the information collected to date about the geologic conditions at the site;to describe the conceptual designs for the repository and the waste package;and to present the plans for obtaining the geologic information necessary to demonstrate the suitability of the site for repository, to design the repository and the waste package, to prepare an environmental impact statement, and to obtain from the US Nuclear Regulatory Commission (NRC) an authorization to construct the repository. This introduction begins with a brief section on the process for siting and developing a repository, followed by a discussion of the pertinent legislation and regulations. A description of site characterization is presented next;it describes the facilities to be constructed for the site characterization program and explains the principal activities to be conducted during the program. Finally, the purpose, content, organizing principles, and organization of this site characterization plan are outlined, and compliance with applicable regulations is discussed

NRPB models for calculating the transfer of radionuclides through the environment. Verification and validation

International Nuclear Information System (INIS)

Attwood, C.; Barraclough, I.; Brown, J.

1998-06-01

There is a wide range of models available at NRPB to predict the transfer of radionuclides through the environment. Such models form an essential part of assessments of the radiological impact of releases of radionuclides into the environment. These models cover: the atmosphere; the aquatic environment; the geosphere; the terrestrial environment including foodchains. It is important that the models used for radiological impact assessments are robust, reliable and suitable for the assessment being undertaken. During model development it is, therefore, important that the model is both verified and validated. Verification of a model involves ensuring that it has been implemented correctly, while validation consists of demonstrating that the model is an adequate representation of the real environment. The extent to which a model can be verified depends on its complexity and whether similar models exist. For relatively simple models verification is straightforward, but for more complex models verification has to form part of the development, coding and testing of the model within quality assurance procedures. Validation of models should ideally consist of comparisons between the results of the models and experimental or environmental measurement data that were not used to develop the model. This is more straightforward for some models than for others depending on the quantity and type of data available. Validation becomes increasingly difficult for models which are intended to predict environmental transfer at long times or at great distances. It is, therefore, necessary to adopt qualitative validation techniques to ensure that the model is an adequate representation of the real environment. This report summarises the models used at NRPB to predict the transfer of radionuclides through the environment as part of a radiological impact assessment. It outlines the work carried out to verify and validate the models. The majority of these models are not currently available

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

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

International Nuclear Information System (INIS)

West, M.

2009-01-01

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

Update on Monitoring Technologies for International Safeguards and Fissile Material Verification

International Nuclear Information System (INIS)

Croessmann, C. Dennis; Glidewell, Don D.; Mangan, Dennis L.; Smathers, Douglas C.

1999-01-01

Monitoring technologies are playing an increasingly important part in international safeguards and fissile material verification. The developments reduce the time an inspector must spend at a site while assuring continuity of knowledge. Monitoring technologies' continued development has produced new seal systems and integrated video surveillance advances under consideration for Trilateral Initiative use. This paper will present recent developments for monitoring systems at Embalse, Argentina, VNHEF, Sarov, Russian, and Savannah River Site, Aiken, South Carolina

Safety analysis report for packaging (onsite) transuranic performance demonstration program sample packaging

International Nuclear Information System (INIS)

Mccoy, J.C.

1997-01-01

The Transuranic Performance Demonstration Program (TPDP) sample packaging is used to transport highway route controlled quantities of weapons grade (WG) plutonium samples from the Plutonium Finishing Plant (PFP) to the Waste Receiving and Processing (WRAP) facility and back. The purpose of these shipments is to test the nondestructive assay equipment in the WRAP facility as part of the Nondestructive Waste Assay PDP. The PDP is part of the U. S. Department of Energy (DOE) National TRU Program managed by the U. S. Department of Energy, Carlsbad Area Office, Carlsbad, New Mexico. Details of this program are found in CAO-94-1045, Performance Demonstration Program Plan for Nondestructive Assay for the TRU Waste Characterization Program (CAO 1994); INEL-96/0129, Design of Benign Matrix Drums for the Non-Destructive Assay Performance Demonstration Program for the National TRU Program (INEL 1996a); and INEL-96/0245, Design of Phase 1 Radioactive Working Reference Materials for the Nondestructive Assay Performance Demonstration Program for the National TRU Program (INEL 1996b). Other program documentation is maintained by the national TRU program and each DOE site participating in the program. This safety analysis report for packaging (SARP) provides the analyses and evaluations necessary to demonstrate that the TRU PDP sample packaging meets the onsite transportation safety requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for an onsite Transportation Hazard Indicator (THI) 2 packaging. This SARP, however, does not include evaluation of any operations within the PFP or WRAP facilities, including handling, maintenance, storage, or operating requirements, except as they apply directly to transportation between the gate of PFP and the gate of the WRAP facility. All other activities are subject to the requirements of the facility safety analysis reports (FSAR) of the PFP or WRAP facility and requirements of the PDP

DESIGN ANALYSIS FOR THE NAVAL SNF WASTE PACKAGE

International Nuclear Information System (INIS)

T.L. Mitchell

2000-01-01

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

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

Impact assessment at a hypothetical submergence of a transport package of radioactive materials

International Nuclear Information System (INIS)

Tsumune, Daisuke; Saegusa, Toshiari; Ito, Chihiro

2007-01-01

Under INF code and IAEA standard, radioactive materials are transported safety on the sea. To gain the public acceptance for these transports additionally, impact assessments have been made by assuming that a radioactive material package might be sunk into the sea. A method of the impact assessment consists of the calculation of release rate of radionuclide from a package, calculation of radionuclide concentration in the ocean, and estimation of dose assessment for the public. An ocean general circulation model was used to calculate the radionuclide concentration in the ocean. Background radionuclide concentration by fallout was simulated by the ocean general circulation model in this method for the verification. Agreement between calculation and observation suggests that this method is appropriate for the assessment. In the both cases for a package sunk at the coastal region at the depth of two hundreds meters and for that sunk at the ocean at the depth of several thousands meters, the evaluated result of the dose equivalent by radiation exposure to the public are far below the dose equivalent limit of the ICRP recommendation (1 mSv/year). (author)

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

Verification survey of the 17th Street Drainage Area, Santa Susana Field Laboratory, The Boeing Company, Ventura County, California

International Nuclear Information System (INIS)

John R. Morton

2000-01-01

An independent (third-party) verification of contractor remedial actions of the subject site verifies that remedial actions have been effective in meeting established and site-specific guidelines and that the documentation accurately and adequately describes the radiological conditions at the site

Results of the radiological verification survey of the partial remediation at 90 Avenue C, Lodi, New Jersey (LJ079V)

International Nuclear Information System (INIS)

Foley, R.D.; Johnson, C.A.

1994-02-01

The property at 90 Avenue C, Lodi, New Jersey is one of the vicinity properties of the former Maywood Chemical Works, Maywood, New Jersey designated for remedial action by the US Department of Energy (DOE). In July 1991, Bechtel National, Inc. performed a partial remedial action on this property. At the request of DOE, a team from Oak Ridge National Laboratory conducted an independent radiological verification survey in July, 1991 at this site. The purpose of the verification survey was to ensure the effectiveness of remedial actions performed within FUSRAP and to confirm the site's compliance with DOE guidelines. The radiological survey included surface gamma scans indoors and outdoors, ground-level beta-gamma measurements, and systematic and biased soil and material sampling. Results of the verification survey demonstrated that all radiological measurements on the portions of the property that had been remediated were within DOE guidelines. However, there still remains a portion of the property to be remediated that is not covered by this verification survey

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

A review of technology for verification of waste removal from Hanford Underground Storage Tanks (WHC Issue 30)

International Nuclear Information System (INIS)

Thunborg, S.

1994-09-01

Remediation of waste from Underground Storage Tanks (UST) at the Hanford Waste storage sites will require removal of all waste to a nearly clean condition. Current requirements are 99% clean. In order to meet remediation legal requirements, a means to remotely verify that the waste has been removed to sufficient level is needed. This report discusses the requirements for verification and reviews major technologies available for inclusion in a verification system. The report presents two operational scenarios for verification of residual waste volume. Thickness verification technologies reviewed are Ultrasonic Sensors, Capacitance Type Sensors, Inductive Sensors, Ground Penetrating Radar, and Magnetometers. Of these technologies Inductive (Metal Detectors) and Ground Penetrating Radar appear to be the most suitable for use as waste thickness sensors

A restructuring of the CF/EDF packages for the MIDAS computer code

International Nuclear Information System (INIS)

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

2004-01-01

The CF and EDF packages, which allow the user to define the functions of variables in a database and the usage of an external data file, have 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 a modernized data structure. To restructure the code, the data transferring methods of the current MELCOR code are modified and then partially adopted into the CF/EDF packages. The data structure of the current MELCOR code using FORTRAN77 has a difficulty in grasping the meaning of the variables as pointers are used to define their addresses. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type without pointers leading to an efficient memory treatment and an easy understanding of the code. Restructuring of the CF/EDF packages addressed in this paper includes a module development and subroutine modification. The verification has been done by comparing the results of the modified code with those of the existing code and the trends are almost the same to each other. Therefore the similar approach could be extended to the entire code package for code restructuring. It is expected that the code restructuring will accelerate the code's domestication thanks to a direct understanding of each variable and an easy implementation of the modified or newly developed models. (author)

HyDe: a Python Package for Genome-Scale Hybridization Detection.

Science.gov (United States)

Blischak, Paul D; Chifman, Julia; Wolfe, Andrea D; Kubatko, Laura S

2018-03-19

The analysis of hybridization and gene flow among closely related taxa is a common goal for researchers studying speciation and phylogeography. Many methods for hybridization detection use simple site pattern frequencies from observed genomic data and compare them to null models that predict an absence of gene flow. The theory underlying the detection of hybridization using these site pattern probabilities exploits the relationship between the coalescent process for gene trees within population trees and the process of mutation along the branches of the gene trees. For certain models, site patterns are predicted to occur in equal frequency (i.e., their difference is 0), producing a set of functions called phylogenetic invariants. In this paper we introduce HyDe, a software package for detecting hybridization using phylogenetic invariants arising under the coalescent model with hybridization. HyDe is written in Python, and can be used interactively or through the command line using pre-packaged scripts. We demonstrate the use of HyDe on simulated data, as well as on two empirical data sets from the literature. We focus in particular on identifying individual hybrids within population samples and on distinguishing between hybrid speciation and gene flow. HyDe is freely available as an open source Python package under the GNU GPL v3 on both GitHub (https://github.com/pblischak/HyDe) and the Python Package Index (PyPI: https://pypi.python.org/pypi/phyde).

GUIDEseq: a bioconductor package to analyze GUIDE-Seq datasets for CRISPR-Cas nucleases.

Science.gov (United States)

Zhu, Lihua Julie; Lawrence, Michael; Gupta, Ankit; Pagès, Hervé; Kucukural, Alper; Garber, Manuel; Wolfe, Scot A

2017-05-15

Genome editing technologies developed around the CRISPR-Cas9 nuclease system have facilitated the investigation of a broad range of biological questions. These nucleases also hold tremendous promise for treating a variety of genetic disorders. In the context of their therapeutic application, it is important to identify the spectrum of genomic sequences that are cleaved by a candidate nuclease when programmed with a particular guide RNA, as well as the cleavage efficiency of these sites. Powerful new experimental approaches, such as GUIDE-seq, facilitate the sensitive, unbiased genome-wide detection of nuclease cleavage sites within the genome. Flexible bioinformatics analysis tools for processing GUIDE-seq data are needed. Here, we describe an open source, open development software suite, GUIDEseq, for GUIDE-seq data analysis and annotation as a Bioconductor package in R. The GUIDEseq package provides a flexible platform with more than 60 adjustable parameters for the analysis of datasets associated with custom nuclease applications. These parameters allow data analysis to be tailored to different nuclease platforms with different length and complexity in their guide and PAM recognition sequences or their DNA cleavage position. They also enable users to customize sequence aggregation criteria, and vary peak calling thresholds that can influence the number of potential off-target sites recovered. GUIDEseq also annotates potential off-target sites that overlap with genes based on genome annotation information, as these may be the most important off-target sites for further characterization. In addition, GUIDEseq enables the comparison and visualization of off-target site overlap between different datasets for a rapid comparison of different nuclease configurations or experimental conditions. For each identified off-target, the GUIDEseq package outputs mapped GUIDE-Seq read count as well as cleavage score from a user specified off-target cleavage score prediction

LLNL`s partnership with selected US mines, for CTBT verification: A pictorial and some reflections

Energy Technology Data Exchange (ETDEWEB)

Heuze, F.E.

1996-01-01

The verification of an upcoming Comprehensive Test Ban Treaty (CTBT) will involve seismic monitoring and will provide for on-site inspections which may include drilling. Because of the fact that mining operations can send out strong seismic signals, many mining districts in the US and abroad may come under special scrutiny. The seismic signals can be generated by the use of large quantities of conventional explosives, by the collapse of underground workings, or by sudden energy release in the ground such as in rock bursts and coal bumps. These mining activities may be the cause of false alarms, but may also offer opportunities for evasive nuclear testing. So in preparing for future verification of a CTBT it becomes important to address the mining-related questions. For the United States, these are questions to be answered with respect to foreign mines. But there is a good amount of commonality in mining methods worldwide. Studies conducted at US mine sites can provide good analogs of activities that may be carried out for overseas CTBT verification, save for the expected logistical impediments.

Advances in the Processing of VHR Optical Imagery in Support of Safeguards Verification

International Nuclear Information System (INIS)

Niemeyer, I.; Listner, C.; Canty, M.

2015-01-01

Under the Additional Protocol of the Non-Proliferation Treaty (NPT) complementing the safeguards agreements between States and the International Atomic Energy Agency, commercial satellite imagery, preferably acquired by very high-resolution (VHR) satellite sensors, is an important source of safeguards-relevant information. Satellite imagery can assist in the evaluation of site declarations, design information verification, the detection of undeclared nuclear facilities, and the preparation of inspections or other visits. With the IAEA's Geospatial Exploitation System (GES), satellite imagery and other geospatial information such as site plans of nuclear facilities are available for a broad range of inspectors, analysts and country officers. The demand for spatial information and new tools to analyze this data is growing, together with the rising number of nuclear facilities under safeguards worldwide. Automated computer-driven processing of satellite imagery could therefore add a big value in the safeguards verification process. These could be, for example, satellite imagery pre-processing algorithms specially developed for new sensors, tools for pixel or object-based image analysis, or geoprocessing tools that generate additional safeguards-relevant information. In the last decade procedures for automated (pre-) processing of satellite imagery have considerably evolved. This paper aims at testing some pixel-based and object-based procedures for automated change detection and classification in support of safeguards verification. Taking different nuclear sites as examples, these methods will be evaluated and compared with regard to their suitability to (semi-) automatically extract safeguards-relevant information. (author)

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

DOE handbook: Integrated safety management systems (ISMS) verification. Team leader's handbook

International Nuclear Information System (INIS)

1999-06-01

The primary purpose of this handbook is to provide guidance to the ISMS verification Team Leader and the verification team in conducting ISMS verifications. The handbook describes methods and approaches for the review of the ISMS documentation (Phase I) and ISMS implementation (Phase II) and provides information useful to the Team Leader in preparing the review plan, selecting and training the team, coordinating the conduct of the verification, and documenting the results. The process and techniques described are based on the results of several pilot ISMS verifications that have been conducted across the DOE complex. A secondary purpose of this handbook is to provide information useful in developing DOE personnel to conduct these reviews. Specifically, this handbook describes methods and approaches to: (1) Develop the scope of the Phase 1 and Phase 2 review processes to be consistent with the history, hazards, and complexity of the site, facility, or activity; (2) Develop procedures for the conduct of the Phase 1 review, validating that the ISMS documentation satisfies the DEAR clause as amplified in DOE Policies 450.4, 450.5, 450.6 and associated guidance and that DOE can effectively execute responsibilities as described in the Functions, Responsibilities, and Authorities Manual (FRAM); (3) Develop procedures for the conduct of the Phase 2 review, validating that the description approved by the Approval Authority, following or concurrent with the Phase 1 review, has been implemented; and (4) Describe a methodology by which the DOE ISMS verification teams will be advised, trained, and/or mentored to conduct subsequent ISMS verifications. The handbook provides proven methods and approaches for verifying that commitments related to the DEAR, the FRAM, and associated amplifying guidance are in place and implemented in nuclear and high risk facilities. This handbook also contains useful guidance to line managers when preparing for a review of ISMS for radiological

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

Thermal analysis of Yucca Mountain commercial high-level waste packages

International Nuclear Information System (INIS)

Altenhofen, M.K.; Eslinger, P.W.

1992-10-01

The thermal performance of commercial high-level waste packages was evaluated on a preliminary basis for the candidate Yucca Mountain repository site. The purpose of this study is to provide an estimate for waste package component temperatures as a function of isolation time in tuff. Several recommendations are made concerning the additional information and modeling needed to evaluate the thermal performance of the Yucca Mountain repository system

Field test for treatment verification of an in-situ enhanced bioremediation study

International Nuclear Information System (INIS)

Taur, C.K.; Chang, S.C.

1995-01-01

Due to a leakage from a 12-inch pressurized diesel steel pipe four years ago, an area of approximately 30,000 square meters was contaminated. A pilot study applying the technology of in-situ enhanced bioremediation was conducted. In the study, a field test kit and on-site monitoring equipment were applied for site characterization and treatment verification. Physically, the enhanced bioremediation study consisted of an air extraction and air supply system, and a nutrition supply network. Certain consistent sampling methodology was employed. Progress was verified by daily monitoring and monthly verification. The objective of this study was to evaluate the capabilities of indigenous microorganisms to biodegrade the petroleum hydrocarbons with provision of oxygen and nutrients. Nine extraction wells and eight air sparging wells were installed. The air sparging wells injected the air into geoformation and the extraction wells provided the underground air circulation. The soil samples were obtained monthly for treatment verification by a Minuteman drilling machine with 2.5-foot-long hollow-stem augers. The samples were analyzed on site for TPH-diesel concentration by a field test kit manufactured by HNU-Hanby, Houston, Texas. The analytical results from the field test kit were compared with the results from an environmental laboratory. The TVPH concentrations of the air extracted from the vadose zone by a vacuum blower and the extraction wells were routinely monitored by a Foxboro FID and Cosmos XP-311A combustible air detector. The daily monitoring of TVPH concentrations provided the reliable data for assessing the remedial progress

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.

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.

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

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

Site characterization plan:

International Nuclear Information System (INIS)

1988-01-01

The Yucca Mountain site in Nevada is one of three candidate sites for the first geologic repository for radioactive waste. On May 28, 1986, it was recommended for detailed study in a program of site characterization. This site characterization plan (SCP) has been prepared in acordance with the requirements of the Nuclear Waste Policy Act to summarize the information collected to date about the geologic conditions at the site;to describe the conceptual designs for the repository and the waste package and to present the plans for obtaining the geologic information necessary to demonstrate the suitability of the site for a repository, to design the repository and the waste package, to prepare an environmental impact statement, and to obtain from the US Nuclear Regulatory Commission (NRC) an authorization to construct the repository. This introduction begins with a brief section on the process for siting and eveloping a repository, followed by a discussion of the pertinent legislation and regulations. A description of site characterization is presented next;it describes the facilities to be constructed for the site characterization program and explains the principal activities to be conducted during the program. Finally, the purpose, content, organizing prinicples, and organization of this site characterization plan are outlined, and compliance with applicable regulations is discussed. 880 refs., 130 figs., 25 tabs

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...

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

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...

EQ6 Calculations for Chemical Degradation of Navy Waste Packages

International Nuclear Information System (INIS)

S. LeStrange

1999-01-01

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

EQ6 Calculations for Chemical Degradation of Navy Waste Packages

Energy Technology Data Exchange (ETDEWEB)

S. LeStrange

1999-11-15

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

EURATOM safeguards efforts in the development of spent fuel verification methods by non-destructive assay

Energy Technology Data Exchange (ETDEWEB)

Matloch, L.; Vaccaro, S.; Couland, M.; De Baere, P.; Schwalbach, P. [Euratom, Communaute europeenne de l' energie atomique - CEEA (European Commission (EC))

2015-07-01

The back end of the nuclear fuel cycle continues to develop. The European Commission, particularly the Nuclear Safeguards Directorate of the Directorate General for Energy, implements Euratom safeguards and needs to adapt to this situation. The verification methods for spent nuclear fuel, which EURATOM inspectors can use, require continuous improvement. Whereas the Euratom on-site laboratories provide accurate verification results for fuel undergoing reprocessing, the situation is different for spent fuel which is destined for final storage. In particular, new needs arise from the increasing number of cask loadings for interim dry storage and the advanced plans for the construction of encapsulation plants and geological repositories. Various scenarios present verification challenges. In this context, EURATOM Safeguards, often in cooperation with other stakeholders, is committed to further improvement of NDA methods for spent fuel verification. In this effort EURATOM plays various roles, ranging from definition of inspection needs to direct participation in development of measurement systems, including support of research in the framework of international agreements and via the EC Support Program to the IAEA. This paper presents recent progress in selected NDA methods. These methods have been conceived to satisfy different spent fuel verification needs, ranging from attribute testing to pin-level partial defect verification. (authors)

Hanford site transuranic waste certification plan

International Nuclear Information System (INIS)

GREAGER, T.M.

1999-01-01

As a generator of transuranic (TRU) and TRU mixed waste destined for disposal at the Waste Isolation Pilot Plant (WIPP), the Hanford Site must ensure that its TRU waste meets the requirements of U.S. Department of Energy (DOE) Order 5820.2A, ''Radioactive Waste Management, and the Waste Acceptance Criteria for the Waste Isolation Pilot Plant' (DOE 1996d) (WIPP WAC). The WIPP WAC establishes the specific physical, chemical, radiological, and packaging criteria for acceptance of defense TRU waste shipments at WIPP. The WIPP WAC also requires that participating DOE TRU waste generator/treatment/storage sites produce site-specific documents, including a certification plan, that describe their management of TRU waste and TRU waste shipments before transferring waste to WIPP. The Hanford Site must also ensure that its TRU waste destined for disposal at WIPP meets requirements for transport in the Transuranic Package Transporter41 (TRUPACT-11). The U.S. Nuclear Regulatory Commission (NRC) establishes the TRUPACT-I1 requirements in the ''Safety Analysis Report for the TRUPACT-II Shipping Package'' (NRC 1997) (TRUPACT-I1 SARP)

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

Can self-verification strivings fully transcend the self-other barrier? Seeking verification of ingroup identities.

Science.gov (United States)

Gómez, Angel; Seyle, D Conor; Huici, Carmen; Swann, William B

2009-12-01

Recent research has demonstrated self-verification strivings in groups, such that people strive to verify collective identities, which are personal self-views (e.g., "sensitive") associated with group membership (e.g., "women"). Such demonstrations stop short of showing that the desire for self-verification can fully transcend the self-other barrier, as in people working to verify ingroup identities (e.g., "Americans are loud") even when such identities are not self-descriptive ("I am quiet and unassuming"). Five studies focus on such ingroup verification strivings. Results indicate that people prefer to interact with individuals who verify their ingroup identities over those who enhance these identities (Experiments 1-5). Strivings for ingroup identity verification were independent of the extent to which the identities were self-descriptive but were stronger among participants who were highly invested in their ingroup identities, as reflected in high certainty of these identities (Experiments 1-4) and high identification with the group (Experiments 1-5). In addition, whereas past demonstrations of self-verification strivings have been limited to efforts to verify the content of identities (Experiments 1 to 3), the findings also show that they strive to verify the valence of their identities (i.e., the extent to which the identities are valued; Experiments 4 and 5). Self-verification strivings, rather than self-enhancement strivings, appeared to motivate participants' strivings for ingroup identity verification. Links to collective self-verification strivings and social identity theory are discussed.

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

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

Method to determine the radioactivity of radioactive waste packages. Basic procedure of the method used to determine the radioactivity of low-level radioactive waste packages generated at nuclear power plants: 2007

International Nuclear Information System (INIS)

2008-03-01

This document describes the procedures adopted in order to determine the radioactivity of low-level radioactive waste packages generated at nuclear power plants in Japan. The standards applied have been approved by the Atomic Energy Society of Japan after deliberations by the Subcommittee on the Radioactivity Verification Method for Waste Packages, the Nuclear Cycle Technical Committee, and the Standards Committee. The method for determining the radioactivity of the low-level radioactive waste packages was based on procedures approved by the Nuclear Safety Commission in 1992. The scaling factor method and other methods of determining radioactivity were then developed on the basis of various investigations conducted, drawing on extensive accumulated knowledge. Moreover, the international standards applied as common guidelines for the scaling factor method were developed by Technical Committee ISO/TC 85, Nuclear Energy, Subcommittee SC 5, Nuclear Fuel Technology. Since the application of accumulated knowledge to future radioactive waste disposal is considered to be rational and justified, such body of knowledge has been documented in a standardized form. The background to this standardization effort, the reasoning behind the determination method as applied to the measurement of radioactivity, as well as other related information, are given in the Annexes hereto. This document includes the following Annexes. Annex 1: (reference) Recorded items related to the determination of the scaling factor. Annex 2 (reference): Principles applied to the determining the radioactivity of waste packages. (author)

The Packaging Handbook -- A guide to package design

International Nuclear Information System (INIS)

Shappert, L.B.

1995-01-01

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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...

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

Verification survey report of the south waste tank farm training/test tower and hazardous waste storage lockers at the West Valley demonstration project, West Valley, New York

International Nuclear Information System (INIS)

Weaver, Phyllis C.

2012-01-01

A team from ORAU's Independent Environmental Assessment and Verification Program performed verification survey activities on the South Test Tower and four Hazardous Waste Storage Lockers. Scan data collected by ORAU determined that both the alpha and alpha-plus-beta activity was representative of radiological background conditions. The count rate distribution showed no outliers that would be indicative of alpha or alpha-plus-beta count rates in excess of background. It is the opinion of ORAU that independent verification data collected support the site's conclusions that the South Tower and Lockers sufficiently meet the site criteria for release to recycle and reuse

System design package for SIMS prototype system 3, solar heating and domestic hot water

Energy Technology Data Exchange (ETDEWEB)

1978-11-01

This report is a collation of documents and drawings that describe a prototype solar heating and hot water system using liquid flat plat collectors and a gas or electric furnace energy subsystem. The system was designed for installation into a single-family dwelling. The description, performance specification, subsystem drawings, verification plan/procedure, and hazard analysis of the system are packaged for evaluation of the system with information sufficient to assemble a similar system. The SIMS Prototype Heating and Hot Water System, Model Number 3 has been installed in a residence at Glendo State Park, Glendo, Wyoming.

Toward a regional power plant siting method: AEC-Maryland regional siting factors study, FY 1974 progress report

International Nuclear Information System (INIS)

Yaffee, S.L.; Miller, C.A.

1974-11-01

The ''AEC-Maryland Regional Siting Factors Study'' examines the process of siting in a regional context. It is developing an analysis method to delineate candidate areas for siting of several power plant technology packages, including both fossil-fueled and nuclear options. Tools that are being used include simulation modeling, economic and demographic forecasting, spatial analysis, and computer graphics and numerical manipulation. The approach will describe the trade-offs incurred if a power plant is located in one candidate area rather than in another. In FY 1974, a suitability analysis method was developed which uses engineering and environmental parameters to define a level of environmental cost incurred if a segment of land is used to site a specific technology package. (U.S.)

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.

Estimates of dietary exposure to bisphenol A (BPA) from light metal packaging using food consumption and packaging usage data: a refined deterministic approach and a fully probabilistic (FACET) approach.

Science.gov (United States)

Oldring, P K T; Castle, L; O'Mahony, C; Dixon, J

2014-01-01

The FACET tool is a probabilistic model to estimate exposure to chemicals in foodstuffs, originating from flavours, additives and food contact materials. This paper demonstrates the use of the FACET tool to estimate exposure to BPA (bisphenol A) from light metal packaging. For exposure to migrants from food packaging, FACET uses industry-supplied data on the occurrence of substances in the packaging, their concentrations and construction of the packaging, which were combined with data from a market research organisation and food consumption data supplied by national database managers. To illustrate the principles, UK packaging data were used together with consumption data from the UK National Diet and Nutrition Survey (NDNS) dietary survey for 19-64 year olds for a refined deterministic verification. The UK data were chosen mainly because the consumption surveys are detailed, data for UK packaging at a detailed level were available and, arguably, the UK population is composed of high consumers of packaged foodstuffs. Exposures were run for each food category that could give rise to BPA from light metal packaging. Consumer loyalty to a particular type of packaging, commonly referred to as packaging loyalty, was set. The BPA extraction levels used for the 15 types of coating chemistries that could release BPA were in the range of 0.00005-0.012 mg dm(-2). The estimates of exposure to BPA using FACET for the total diet were 0.0098 (mean) and 0.0466 (97.5th percentile) mg/person/day, corresponding to 0.00013 (mean) and 0.00059 (97.5th percentile) mg kg(-1) body weight day(-1) for consumers of foods packed in light metal packaging. This is well below the current EFSA (and other recognised bodies) TDI of 0.05 mg kg(-1) body weight day(-1). These probabilistic estimates were compared with estimates using a refined deterministic approach drawing on the same input data. The results from FACET for the mean, 95th and 97.5th percentile exposures to BPA lay between the

Design, Implementation, and Verification of the Reliable Multicast Protocol. Thesis

Science.gov (United States)

Montgomery, Todd L.

1995-01-01

This document describes the Reliable Multicast Protocol (RMP) design, first implementation, and formal verification. RMP provides a totally ordered, reliable, atomic multicast service on top of an unreliable multicast datagram service. RMP is fully and symmetrically distributed so that no site bears an undue portion of the communications load. RMP provides a wide range of guarantees, from unreliable delivery to totally ordered delivery, to K-resilient, majority resilient, and totally resilient atomic delivery. These guarantees are selectable on a per message basis. RMP provides many communication options, including virtual synchrony, a publisher/subscriber model of message delivery, a client/server model of delivery, mutually exclusive handlers for messages, and mutually exclusive locks. It has been commonly believed that total ordering of messages can only be achieved at great performance expense. RMP discounts this. The first implementation of RMP has been shown to provide high throughput performance on Local Area Networks (LAN). For two or more destinations a single LAN, RMP provides higher throughput than any other protocol that does not use multicast or broadcast technology. The design, implementation, and verification activities of RMP have occurred concurrently. This has allowed the verification to maintain a high fidelity between design model, implementation model, and the verification model. The restrictions of implementation have influenced the design earlier than in normal sequential approaches. The protocol as a whole has matured smoother by the inclusion of several different perspectives into the product development.

Closure report for housekeeping category, Corrective Action Unit 348, Nevada Test Site

International Nuclear Information System (INIS)

1998-01-01

This Closure Report summarizes the corrective actions which were completed at twelve Corrective Action Sites within Corrective Action Unit 348 at the Nevada Test Site. Current site descriptions, observations and identification of wastes removed are included on FFACO Corrective Action Site housekeeping closure verification forms

Closure report for housekeeping category, Corrective Action Unit 347, Nevada Test Site

International Nuclear Information System (INIS)

1998-01-01

This Closure Report summarizes the corrective actions which were completed at the Corrective Action Sites within Corrective Action Unit 347 at the Nevada Test Site. Current site descriptions, observations and identification of wastes removed are included on FFACO Corrective Action Site housekeeping closure verification forms

Radiological verification survey results at 898 Black Oak Ridge Rd., Wayne, New Jersey (WJ004V)

International Nuclear Information System (INIS)

Rodriguez, R.E.; Johnson, C.A.

1995-05-01

The US Department of Energy (DOE) conducted remedial action during 1993 at the Pompton Plains Railroad Spur and eight vicinity properties in the Wayne and Pequannock Townships in New Jersey as part of the Formerly Utilized Sites Remedial Action Program (FUSRAP). These properties are in the vicinity of the DOE-owned Wayne Interim Storage Site (WISS), formerly the W.R. Grace facility. The property at 898 Black Oak Ridge Road, Wayne, New Jersey is one of these vicinity properties. At the request of DOE, a team from Oak Ridge National Laboratory conducted an independent radiological verification survey at this property. The purpose of the survey, conducted between September and December 1993, was to confirm the success of the remedial actions performed to remove any radioactive materials in excess of the identified guidelines. The verification survey included surface gamma scans and gamma readings at one meter, beta-gamma scans, and the collection of soil and debris samples for radionuclide analysis. Results of the survey demonstrated that all radiological measurements on the property at 898 Black Oak Ridge Road were within applicable DOE guidelines. Based on the results of the remedial action data and confirmed by the verification survey data, the portions of the site that had been remediated during this action successfully meet the DOE remedial action objectives

Radiological verification survey results at the Pompton Plains Railroad Spur, Pequannock, New Jersey (PJ008V)

International Nuclear Information System (INIS)

Rodriguez, R.E.; Johnson, C.A.

1995-05-01

The US Department of Energy (DOE) conducted remedial action during 1993 at the Pompton Plains railroad spur and eight vicinity properties in the Wayne and Pequannock Townships in New Jersey as part of the Formerly Utilized Sites Remedial Action Program (FUSRAP). These properties are in the vicinity of the DOE-owned Wayne Interim Storage Site (WISS), formerly the W.R. Grace facility. The property at the Pompton Plains Railroad Spur, Pequannock, New Jersey is one of these vicinity properties. At the request of DOE, a team from Oak Ridge National Laboratory conducted an independent radiological verification survey at this property. The purpose of the survey, conducted between September and December 1993, was to confirm the success of the remedial actions performed to remove any radioactive materials in excess of the identified guidelines. The verification survey included surface gamma scans and gamma readings at 1 meter, beta-gamma scans, and the collection of soil and debris samples for radionuclide analysis. Results of the survey demonstrated that all radiological measurements on the property at the Pompton Plains railroad spur were within applicable DOE guidelines. Based on the results of the remedial action data and confirmed by the verification survey data, the portions of the site that had been remediated during this action successfully meet the DOE remedial action objectives

37Ar monitoring techniques and on-site inspection system

International Nuclear Information System (INIS)

Duan Rongliang; Chen Yinliang; Li Wei; Wang Hongxia; Hao Fanhua

2001-01-01

37 Ar is separated, purified and extracted from air sample with a low temperature gas-solid chromatographic purifying method, prepared into a radioactive measurement source and its radioactivity is measured with a proportional counter. Based on the monitoring result, a judgement can be made if an nuclear explosion event has happened recently in a spectabilis area. A series of element techniques that are associated the monitoring of the trace element 37 Ar have been investigated and developed. Those techniques include leaked gas sampling, 37 Ar separation and purification, 37 Ar radioactivity measurement and the on-site inspection of 37 Ar. An advanced 37 Ar monitoring method has been developed, with which 200 liters of air can be treated in 2 hours with sensitivity of 0.01 Bq/L for 37 Ar radioactivity measurement. A practical 37 Ar On-site Inspection system has been developed. This research work may provide technical and equipment support for the verification protection, verification supervision and CTBT verification

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

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...

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.

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

Lessons learned in the implementation of Integrated Safety Management at DOE Order Compliance Sites vs Necessary and Sufficient Sites

International Nuclear Information System (INIS)

Hill, R.L.

2000-01-01

This paper summarizes the development and implementation of Integrated Safety Management (ISM) at an Order Compliance Site (Savannah River Site) and a Necessary and Sufficient Site (Nevada Test Site). A discussion of each core safety function of ISM is followed by an example from an Order Compliance Site and a Necessary and Sufficient Site. The Savannah River Site was the first DOE site to have a DOE Headquarters-validated and approved ISM System. The NTS is beginning the process of verification and validation. This paper defines successful strategies for integrating Environment, Safety, and Health management into work under various scenarios

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

International Nuclear Information System (INIS)

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

1995-04-01

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

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

Particularities of Verification Processes for Distributed Informatics Applications

Directory of Open Access Journals (Sweden)

Ion IVAN

2013-01-01

Full Text Available This paper presents distributed informatics applications and characteristics of their development cycle. It defines the concept of verification and there are identified the differences from software testing. Particularities of the software testing and software verification processes are described. The verification steps and necessary conditions are presented and there are established influence factors of quality verification. Software optimality verification is analyzed and some metrics are defined for the verification process.

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.)

Specificity of interactions among the DNA-packaging machine components of T4-related bacteriophages.

Science.gov (United States)

Gao, Song; Rao, Venigalla B

2011-02-04

Tailed bacteriophages use powerful molecular motors to package the viral genome into a preformed capsid. Packaging at a rate of up to ∼2000 bp/s and generating a power density twice that of an automobile engine, the phage T4 motor is the fastest and most powerful reported to date. Central to DNA packaging are dynamic interactions among the packaging components, capsid (gp23), portal (gp20), motor (gp17, large "terminase"), and regulator (gp16, small terminase), leading to precise orchestration of the packaging process, but the mechanisms are poorly understood. Here we analyzed the interactions between small and large terminases of T4-related phages. Our results show that the gp17 packaging ATPase is maximally stimulated by homologous, but not heterologous, gp16. Multiple interaction sites are identified in both gp16 and gp17. The specificity determinants in gp16 are clustered in the diverged N- and C-terminal domains (regions I-III). Swapping of diverged region(s), such as replacing C-terminal RB49 region III with that of T4, switched ATPase stimulation specificity. Two specificity regions, amino acids 37-52 and 290-315, are identified in or near the gp17-ATPase "transmission" subdomain II. gp16 binding at these sites might cause a conformational change positioning the ATPase-coupling residues into the catalytic pocket, triggering ATP hydrolysis. These results lead to a model in which multiple weak interactions between motor and regulator allow dynamic assembly and disassembly of various packaging complexes, depending on the functional state of the packaging machine. This might be a general mechanism for regulation of the phage packaging machine and other complex molecular machines.

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

Verification of statistical method CORN for modeling of microfuel in the case of high grain concentration

Energy Technology Data Exchange (ETDEWEB)

Chukbar, B. K., E-mail: bchukbar@mail.ru [National Research Center Kurchatov Institute (Russian Federation)

2015-12-15

Two methods of modeling a double-heterogeneity fuel are studied: the deterministic positioning and the statistical method CORN of the MCU software package. The effect of distribution of microfuel in a pebble bed on the calculation results is studied. The results of verification of the statistical method CORN for the cases of the microfuel concentration up to 170 cm{sup –3} in a pebble bed are presented. The admissibility of homogenization of the microfuel coating with the graphite matrix is studied. The dependence of the reactivity on the relative location of fuel and graphite spheres in a pebble bed is found.

TH-B-204-03: TG-199: Implanted Markers for Radiation Treatment Verification

International Nuclear Information System (INIS)

Wang, Z.

2016-01-01

Implanted markers as target surrogates have been widely used for treatment verification, as they provide safe and reliable monitoring of the inter- and intra-fractional target motion. The rapid advancement of technology requires a critical review and recommendation for the usage of implanted surrogates in current field. The symposium, also reporting an update of AAPM TG 199 - Implanted Target Surrogates for Radiation Treatment Verification, will be focusing on all clinical aspects of using the implanted target surrogates for treatment verification and related issues. A wide variety of markers available in the market will be first reviewed, including radiopaque markers, MRI compatible makers, non-migrating coils, surgical clips and electromagnetic transponders etc. The pros and cons of each kind will be discussed. The clinical applications of implanted surrogates will be presented based on different anatomical sites. For the lung, we will discuss gated treatments and 2D or 3D real-time fiducial tracking techniques. For the prostate, we will be focusing on 2D-3D, 3D-3D matching and electromagnetic transponder based localization techniques. For the liver, we will review techniques when patients are under gating, shallow or free breathing condition. We will review techniques when treating challenging breast cancer as deformation may occur. Finally, we will summarize potential issues related to the usage of implanted target surrogates with TG 199 recommendations. A review of fiducial migration and fiducial derived target rotation in different disease sites will be provided. The issue of target deformation, especially near the diaphragm, and related suggestions will be also presented and discussed. Learning Objectives: Knowledge of a wide variety of markers Knowledge of their application for different disease sites Understand of issues related to these applications Z. Wang: Research funding support from Brainlab AG Q. Xu: Consultant for Accuray; Q. Xu, I am a consultant

TH-B-204-03: TG-199: Implanted Markers for Radiation Treatment Verification

Energy Technology Data Exchange (ETDEWEB)

Wang, Z. [Duke University Medical Center (United States)

2016-06-15

Implanted markers as target surrogates have been widely used for treatment verification, as they provide safe and reliable monitoring of the inter- and intra-fractional target motion. The rapid advancement of technology requires a critical review and recommendation for the usage of implanted surrogates in current field. The symposium, also reporting an update of AAPM TG 199 - Implanted Target Surrogates for Radiation Treatment Verification, will be focusing on all clinical aspects of using the implanted target surrogates for treatment verification and related issues. A wide variety of markers available in the market will be first reviewed, including radiopaque markers, MRI compatible makers, non-migrating coils, surgical clips and electromagnetic transponders etc. The pros and cons of each kind will be discussed. The clinical applications of implanted surrogates will be presented based on different anatomical sites. For the lung, we will discuss gated treatments and 2D or 3D real-time fiducial tracking techniques. For the prostate, we will be focusing on 2D-3D, 3D-3D matching and electromagnetic transponder based localization techniques. For the liver, we will review techniques when patients are under gating, shallow or free breathing condition. We will review techniques when treating challenging breast cancer as deformation may occur. Finally, we will summarize potential issues related to the usage of implanted target surrogates with TG 199 recommendations. A review of fiducial migration and fiducial derived target rotation in different disease sites will be provided. The issue of target deformation, especially near the diaphragm, and related suggestions will be also presented and discussed. Learning Objectives: Knowledge of a wide variety of markers Knowledge of their application for different disease sites Understand of issues related to these applications Z. Wang: Research funding support from Brainlab AG Q. Xu: Consultant for Accuray; Q. Xu, I am a consultant

Safety analysis report for packaging (onsite) doorstop samplecarrier system

Energy Technology Data Exchange (ETDEWEB)

Obrien, J.H.

1997-02-24

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

Double-Difference Tomography for Sequestration MVA [monitoring, verification, and accounting

Energy Technology Data Exchange (ETDEWEB)

Westman, Erik [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States)

2012-12-31

Analysis of synthetic data was performed to determine the most cost-effective tomographic monitoring system for a geologic carbon sequestration injection site. Double-difference tomographic inversion was performed on 125 synthetic data sets: five stages of CO2 plume growth, five seismic event regions, and five geophone arrays. Each resulting velocity model was compared quantitatively to its respective synthetic velocity model to determine an accuracy value. The results were examined to determine a relationship between cost and accuracy in monitoring, verification, and accounting applications using double-difference tomography. The geophone arrays with widely-varying geophone locations, both laterally and vertically, performed best. Additionally, double difference seismic tomography was performed using travel time data from a carbon sequestration site at the Aneth oil field in southeast Utah as part of a Department of Energy initiative on monitoring, verification, and accounting (MVA) of sequestered CO2. A total of 1,211 seismic events were recorded from a borehole array consisting of 22 geophones. Artificial velocity models were created to determine the ease with which different CO2 plume locations and sizes can be detected. Most likely because of the poor geophone arrangement, a low velocity zone in the Desert Creek reservoir can only be detected when regions of test site containing the highest ray path coverage are considered. MVA accuracy and precision may be improved through the use of a receiver array that provides more comprehensive ray path coverage.

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

Function and horizontal transfer of the small terminase subunit of the tailed bacteriophage Sf6 DNA packaging nanomotor

Science.gov (United States)

Leavitt, Justin C.; Gilcrease, Eddie B.; Wilson, Kassandra; Casjens, Sherwood R.

2013-01-01

Bacteriophage Sf6 DNA packaging series initiate at many locations across a 2 kbp region. Our in vivo studies that show that Sf6 small terminase subunit (TerS) protein recognizes a specific packaging (pac) site near the center of this region, that this site lies within the portion of the Sf6 gene that encodes the DNA-binding domain of TerS protein, that this domain of the TerS protein is responsible for the imprecision in Sf6 packaging initiation, and that the DNA-binding domain of TerS must be covalently attached to the domain that interacts with the rest of the packaging motor. The TerS DNA-binding domain is self-contained in that it apparently does not interact closely with the rest of the motor and it binds to a recognition site that lies within the DNA that encodes the domain. This arrangement has allowed the horizontal exchange of terS genes among phages to be very successful. PMID:23562538

Nuclear waste package design for the Vadose zone in tuff

International Nuclear Information System (INIS)

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

1984-02-01

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

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

Java bytecode verification via static single assignment form

DEFF Research Database (Denmark)

Gal, Andreas; Probst, Christian W.; Franz, Michael

2008-01-01

Java Virtual Machines (JVMs) traditionally perform bytecode verification by way of an iterative data-flow analysis. Bytecode verification is necessary to ensure type safety because temporary variables in the JVM are not statically typed. We present an alternative verification mechanism that trans......Java Virtual Machines (JVMs) traditionally perform bytecode verification by way of an iterative data-flow analysis. Bytecode verification is necessary to ensure type safety because temporary variables in the JVM are not statically typed. We present an alternative verification mechanism...

The Lawrence Livermore National Laboratory DOE-STD-3013 Surveillance Program for the Storage of Plutonium Packages

International Nuclear Information System (INIS)

Riley, D

2005-01-01

This document presents a site-specific DOE-STD-3013 (3013) surveillance program for 3013 material stored at Lawrence Livermore National Laboratory (LLNL) in the B332 Plutonium Facility. The 3013 standard requires the development of a surveillance program to assure the long-term safety of plutonium storage in 3013 compliant containers. A complex-wide Integrated Surveillance Program in Support of Long-Term Storage of Plutonium-Bearing Materials (ISP)(LA-UR-00-3246, Revision 1, March 2001) has been developed to give guidance on an acceptable surveillance approach and to set up a mechanism to integrate surveillance activities and facilitate the sharing of lessons learned. This LLNL 3013 surveillance program has been developed following guidelines established for Storage Sites in the ISP and is sufficient for the storage in the LLNL Plutonium Facility. The LLNL 3013 surveillance program must be coupled with the DOE complex wide Materials Identification and Surveillance (MIS) program and the ISP led by Savannah River Site (SRS). These programs support the technical basis for continuing safe storage of plutonium packages and provide the technical basis for the limited scope of the site-specific LLNL 3013 surveillance program. The LLNL 3013 surveillance program calls for surveillance of 3013 packages to begin approximately three years after packaging of the first oxide. One percent of the stored packages per year will be randomly selected and nondestructively examined (NDE) by LLNL per the guidelines of the ISP. Additional packages may be selected for NDE if recommended by the ISP Steering Committee and agreed upon by the MIS Working Group. One selected package will be shipped to SRS for destructive analysis each year starting when SRS can receive them. This is expected to be in FY2007. We expect to store a maximum of 400 3013 packages. This would result in an expected maximum of 4 surveillances per year. The activities outlined in the program evolved from the current

Progress in forming bottom barriers under waste sites

International Nuclear Information System (INIS)

Carter, E.E.

1997-01-01

The paper describes an new method for the construction, verification, and maintenance of underground vaults to isolate and contain radioactive burial sites without excavation or drilling in contaminated areas. The paper begins with a discussion of previous full-scale field tests of horizontal barrier tools which utilized high pressure jetting technology. This is followed by a discussion of the TECT process, which cuts with an abrasive cable instead of high pressure jets. The new method is potentially applicable to more soil types than previous methods and can form very thick barriers. Both processes are performed from the perimeter of a site and require no penetration or disturbance of the active waste area. The paper also describes long-term verification methods to monitor barrier integrity passively

Formal verification of algorithms for critical systems

Science.gov (United States)

Rushby, John M.; Von Henke, Friedrich

1993-01-01

We describe our experience with formal, machine-checked verification of algorithms for critical applications, concentrating on a Byzantine fault-tolerant algorithm for synchronizing the clocks in the replicated computers of a digital flight control system. First, we explain the problems encountered in unsynchronized systems and the necessity, and criticality, of fault-tolerant synchronization. We give an overview of one such algorithm, and of the arguments for its correctness. Next, we describe a verification of the algorithm that we performed using our EHDM system for formal specification and verification. We indicate the errors we found in the published analysis of the algorithm, and other benefits that we derived from the verification. Based on our experience, we derive some key requirements for a formal specification and verification system adequate to the task of verifying algorithms of the type considered. Finally, we summarize our conclusions regarding the benefits of formal verification in this domain, and the capabilities required of verification systems in order to realize those benefits.

SITE GENERATED RADIOLOGICAL WASTE HANDLING SYSTEM DESCRIPTION DOCUMENT

Energy Technology Data Exchange (ETDEWEB)

S. C. Khamankar

2000-06-20

The Site Generated Radiological Waste Handling System handles radioactive waste products that are generated at the geologic repository operations area. The waste is collected, treated if required, packaged for shipment, and shipped to a disposal site. Waste streams include low-level waste (LLW) in solid and liquid forms, as-well-as mixed waste that contains hazardous and radioactive constituents. Liquid LLW is segregated into two streams, non-recyclable and recyclable. The non-recyclable stream may contain detergents or other non-hazardous cleaning agents and is packaged for shipment. The recyclable stream is treated to recycle a large portion of the water while the remaining concentrated waste is packaged for shipment; this greatly reduces the volume of waste requiring disposal. There will be no liquid LLW discharge. Solid LLW consists of wet solids such as ion exchange resins and filter cartridges, as-well-as dry active waste such as tools, protective clothing, and poly bags. Solids will be sorted, volume reduced, and packaged for shipment. The generation of mixed waste at the Monitored Geologic Repository (MGR) is not planned; however, if it does come into existence, it will be collected and packaged for disposal at its point of occurrence, temporarily staged, then shipped to government-approved off-site facilities for disposal. The Site Generated Radiological Waste Handling System has equipment located in both the Waste Treatment Building (WTB) and in the Waste Handling Building (WHB). All types of liquid and solid LLW are processed in the WTB, while wet solid waste from the Pool Water Treatment and Cooling System is packaged where received in the WHB. There is no installed hardware for mixed waste. The Site Generated Radiological Waste Handling System receives waste from locations where water is used for decontamination functions. In most cases the water is piped back to the WTB for processing. The WTB and WHB provide staging areas for storing and shipping LLW

SITE GENERATED RADIOLOGICAL WASTE HANDLING SYSTEM DESCRIPTION DOCUMENT

International Nuclear Information System (INIS)

S. C. Khamankar

2000-01-01

The Site Generated Radiological Waste Handling System handles radioactive waste products that are generated at the geologic repository operations area. The waste is collected, treated if required, packaged for shipment, and shipped to a disposal site. Waste streams include low-level waste (LLW) in solid and liquid forms, as-well-as mixed waste that contains hazardous and radioactive constituents. Liquid LLW is segregated into two streams, non-recyclable and recyclable. The non-recyclable stream may contain detergents or other non-hazardous cleaning agents and is packaged for shipment. The recyclable stream is treated to recycle a large portion of the water while the remaining concentrated waste is packaged for shipment; this greatly reduces the volume of waste requiring disposal. There will be no liquid LLW discharge. Solid LLW consists of wet solids such as ion exchange resins and filter cartridges, as-well-as dry active waste such as tools, protective clothing, and poly bags. Solids will be sorted, volume reduced, and packaged for shipment. The generation of mixed waste at the Monitored Geologic Repository (MGR) is not planned; however, if it does come into existence, it will be collected and packaged for disposal at its point of occurrence, temporarily staged, then shipped to government-approved off-site facilities for disposal. The Site Generated Radiological Waste Handling System has equipment located in both the Waste Treatment Building (WTB) and in the Waste Handling Building (WHB). All types of liquid and solid LLW are processed in the WTB, while wet solid waste from the Pool Water Treatment and Cooling System is packaged where received in the WHB. There is no installed hardware for mixed waste. The Site Generated Radiological Waste Handling System receives waste from locations where water is used for decontamination functions. In most cases the water is piped back to the WTB for processing. The WTB and WHB provide staging areas for storing and shipping LLW

System specification for the plutonium stabilization and packaging system

International Nuclear Information System (INIS)

1996-01-01

This document describes functional design requirements for the Plutonium Stabilization and Packaging System (Pu SPS), as required by DOE contract DE-AC03-96SF20948 through contract modification 9 for equipment in Building 707 at Rocky Flats Environmental Technology Site (RFETS)

Site characterization plan: Yucca Mountain site, Nevada research and development area, Nevada: Consultation draft, Nuclear Waste Policy Act

Energy Technology Data Exchange (ETDEWEB)

NONE

1988-01-01

Chapter six describes the basis for facility design, the completed facility conceptual design, the completed analytical work relating to the resolution of design issues, and future design-related work. The basis for design and the conceptual design information presented in this chapter meet the requirements of the Nuclear Waste Policy Act of 1982, for a conceptual repository design that takes into account site-specific requirements. This information is presented to permit a critical evaluation of planned site characterization activities. Chapter seven describes waste package components, emplacement environment, design, and status of research and development that support the Nevada Nuclear Waste Storage Investigation (NNWSI) Project. The site characterization plan (SCP) discussion of waste package components is contained entirely within this chapter. The discussion of emplacement environment in this chapter is limited to considerations of the environment that influence, or which may influence, if perturbed, the waste packages and their performance (particularly hydrogeology, geochemistry, and borehole stability). The basis for conceptual waste package design as well as a description of the design is included in this chapter. The complete design will be reported in the advanced conceptual design (ACD) report and is not duplicated in the SCP. 367 refs., 173 figs., 68 tabs.

Site characterization plan: Yucca Mountain site, Nevada research and development area, Nevada: Consultation draft, Nuclear Waste Policy Act

International Nuclear Information System (INIS)

1988-01-01

Chapter six describes the basis for facility design, the completed facility conceptual design, the completed analytical work relating to the resolution of design issues, and future design-related work. The basis for design and the conceptual design information presented in this chapter meet the requiremenrs of the Nuclear Waste Policy Act of 1982, for a conceptual repository design that takes into account site-specific requirements. This information is presented to permit a critical evaluation of planned site characterization activities. Chapter seven describes waste package components, emplacement environment, design, and status of research and development that support the Nevada Nuclear Waste Storage Investigation (NNWSI) Project. The site characterization plan (SCP) discussion of waste package components is contained entirely within this chapter. The discussion of emplacement environment in this chapter is limited to considerations of the environment that influence, or which may influence, if perturbed, the waste packages and their performance (particularly hydrogeology, geochemistry, and borehole stability). The basis for conceptual waste package design as well as a description of the design is included in this chapter. The complete design will be reported in the advanced conceptual design (ACD) report and is not duplicated in the SCP. 367 refs., 173 figs., 68 tabs

MESA: Message-Based System Analysis Using Runtime Verification

Science.gov (United States)

Shafiei, Nastaran; Tkachuk, Oksana; Mehlitz, Peter

2017-01-01

In this paper, we present a novel approach and framework for run-time verication of large, safety critical messaging systems. This work was motivated by verifying the System Wide Information Management (SWIM) project of the Federal Aviation Administration (FAA). SWIM provides live air traffic, site and weather data streams for the whole National Airspace System (NAS), which can easily amount to several hundred messages per second. Such safety critical systems cannot be instrumented, therefore, verification and monitoring has to happen using a nonintrusive approach, by connecting to a variety of network interfaces. Due to a large number of potential properties to check, the verification framework needs to support efficient formulation of properties with a suitable Domain Specific Language (DSL). Our approach is to utilize a distributed system that is geared towards connectivity and scalability and interface it at the message queue level to a powerful verification engine. We implemented our approach in the tool called MESA: Message-Based System Analysis, which leverages the open source projects RACE (Runtime for Airspace Concept Evaluation) and TraceContract. RACE is a platform for instantiating and running highly concurrent and distributed systems and enables connectivity to SWIM and scalability. TraceContract is a runtime verication tool that allows for checking traces against properties specified in a powerful DSL. We applied our approach to verify a SWIM service against several requirements.We found errors such as duplicate and out-of-order messages.

A Syntactic-Semantic Approach to Incremental Verification

OpenAIRE

Bianculli, Domenico; Filieri, Antonio; Ghezzi, Carlo; Mandrioli, Dino

2013-01-01

Software verification of evolving systems is challenging mainstream methodologies and tools. Formal verification techniques often conflict with the time constraints imposed by change management practices for evolving systems. Since changes in these systems are often local to restricted parts, an incremental verification approach could be beneficial. This paper introduces SiDECAR, a general framework for the definition of verification procedures, which are made incremental by the framework...

The relative importance of managerial competencies for predicting the perceived job performance of Broad-Based Black Economic Empowerment verification practitioners

Directory of Open Access Journals (Sweden)

Barbara M. Seate

2016-04-01

Full Text Available Orientation: There is a need for the growing Broad-Based Black Economic Empowerment (B-BBEE verification industry to assess competencies and determine skills gaps for the management of the verification practitioners’ perceived job performance. Knowing which managerial competencies are important for different managerial functions is vital for developing and improving training and development programmes. Research purpose: The purpose of this study was to determine the managerial capabilities that are required of the B-BBEE verification practitioners, in order to improve their perceived job performance. Motivation for the study: The growing number of the B-BBEE verification practitioners calls for more focused training and development. Generating such a training and development programme demands empirical research into the relative importance of managerial competencies. Research approach, design and method: A quantitative design using the survey approach was adopted. A questionnaire was administered to a stratified sample of 87 B-BBEE verification practitioners. Data were analysed using the Statistical Package for Social Sciences (version 22.0 and Smart Partial Least Squares software. Main findings: The results of the correlation analysis revealed that there were strong and positive associations between technical skills, interpersonal skills, compliance to standards and ethics, managerial skills and perceived job performance. Results of the regression analysis showed that managerial skills, compliance to standards and ethics and interpersonal skills were statistically significant in predicting perceived job performance. However, technical skills were insignificant in predicting perceived job performance. Practical/managerial implications: The study has shown that the B-BBEE verification industry, insofar as the technical skills of the practitioners are concerned, does have suitably qualified staff with the requisite educational qualifications. At

Thermal Analysis of the SAFKEG Package for Long Term Storage

International Nuclear Information System (INIS)

NARENDRA, GUPTA

2005-01-01

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

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

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).

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

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.

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

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

Closure report for CAU No. 450: Historical UST release sites, Nevada Test Site. Volume 2

International Nuclear Information System (INIS)

1997-09-01

This report addresses the closure of 11 historical underground storage tank release sites within various areas of the Nevada Test Site. This report contains remedial verification of the soil sample analytical results for the following: Area 11 Tweezer facility; Area 12 boiler house; Area 12 service station; Area 23 bypass yard; Area 23 service station; Area 25 power house; Area 25 tech. services building; Area 25 tech. operations building; Area 26 power house; and Area 27 boiler house

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...

Flammable Gas Refined Safety Analysis Tool Software Verification and Validation Report for Resolve Version 2.5

International Nuclear Information System (INIS)

BRATZEL, D.R.

2000-01-01

The purpose of this report is to document all software verification and validation activities, results, and findings related to the development of Resolve Version 2.5 for the analysis of flammable gas accidents in Hanford Site waste tanks

Results of Remediation and Verification Sampling for the 600-270 Horseshoe Landfill

Energy Technology Data Exchange (ETDEWEB)

W. S. Thompson

2005-12-14

This report presents the results of the 2005 remedial action and verification soil sampling conducted at the 600-270 waste site after removal of soil containing residual concentrations of dichlorodiphenyl trichloroethane and its breakdown products dichlorodiphenyl dichloroethylene and dichlorodiphenyl dichloroethane. The remediation was performed in response to post-closure surface soil sampling performed between 1998 and 2003 that indicated the presence of residual DDT contamination exceeding the Record of Decision for the 1100 Area National Priorities List site cleanup criteria of 1 mg/kg that was established for the original 1994 cleanup activities.

Results of Remediation and Verification Sampling for the 600-270 Horseshoe Landfill

International Nuclear Information System (INIS)

Thompson, W.S.

2005-01-01

This report presents the results of the 2005 remedial action and verification soil sampling conducted at the 600-270 waste site after removal of soil containing residual concentrations of dichlorodiphenyl trichloroethane and its breakdown products dichlorodiphenyl dichloroethylene and dichlorodiphenyl dichloroethane. The remediation was performed in response to post-closure surface soil sampling performed between 1998 and 2003 that indicated the presence of residual DDT contamination exceeding the Record of Decision for the 1100 Area National Priorities List site cleanup criteria of 1 mg/kg that was established for the original 1994 cleanup activities.

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...

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.

Draft site characterization analysis of the site characterization report for the Basalt Waste Isolation Project, Hanford, Washington site. Appendices E through W

International Nuclear Information System (INIS)

1983-03-01

Volume 2 contains Appendices E through W: potential for large-scale pump tests in the Grande Ronde; review of hydrochemical characterization related to flow system interpretation in Hanford basalts; limitations of packer-testing for head evaluation in Hanford basalts; hydrogeologic data integration for conceptual groundwater flow models; drilling mud effects on hydrogeologic testing; site issue analyses related to the nature at the present groundwater system at the Hanford site, Washington; structural and stratigraphic characteristics related to groundwater flow at the Hanford site, Washington; seismic hazard and some examples of hazard studies at Hanford; earthquake swarms in the Columbia Plateau; seismic ground motion at depth; failure modes for the metallic waste package component; degradation mechanisms of borosilicate glass; transport and retardation of radionuclides in the waste package; determination and interpretation of redox conditions and changes in underground high-level repositories; determination and interpretation of sorption data applied to radionuclide migration in underground repositories; solubility of radionuclide compounds presented in the BWIP site characterization report; and release rate from engineered system

Data storage accounting and verification in LHC experiments

CERN Document Server

Ratnikova ,Natalia

2012-01-01

All major experiments at Large Hadron Collider (LHC) need to measure real storage usage at the Grid sites. This information is equally important for the resource management, planning, and operations. To verify consistency of the central catalogs, experiments are asking sites to provide full list of files they have on storage, including size, checksum, and other file attributes. Such storage dumps provided at regular intervals give a realistic view of the storage resource usage by the experiments. Regular monitoring of the space usage and data verification serve as additional internal checks of the system integrity and performance. Both the importance and the complexity of these tasks increase with the constant growth of the total data volumes during the active data taking period at the LHC. Developed common solutions help to reduce the maintenance costs both at the large Tier-1 facilities supporting multiple virtual organizations, and at the small sites that often lack manpower. We discuss requirements...

Packaging design criteria modified fuel spacer burial box. Revision 1

International Nuclear Information System (INIS)

Stevens, P.F.

1994-01-01

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

Methodology and applicability of a safety and demonstration concept for a HAW final repository on clays. Safety concept and verification strategy

International Nuclear Information System (INIS)

Ruebel, Andre; Meleshyn, Artur

2014-08-01

The report describes the site independent frame for a safety concept and verification strategy for a final repository for heat generating wastes in clay rock. In the safety concept planning specifications and technical measures are summarized that are supposed to allow a safe inclusion of radionuclides in the host rock. The verification strategy defines the systematic procedures for the development of fundamentals and scenarios as basis for the demonstration of the safety case and to allow the prognosis of appropriateness. The report includes the boundary conditions, the safety concept for the post-closure phase and the verification strategy for the post-closure phase.

GREEN: A program package for docking studies in rational drug design

Science.gov (United States)

Tomioka, Nobuo; Itai, Akiko

1994-08-01

A program package, GREEN, has been developed that enables docking studies between ligand molecules and a protein molecule. Based on the structure of the protein molecule, the physical and chemical environment of the ligand-binding site is expressed as three-dimensional grid-point data. The grid-point data are used for the real-time evaluation of the protein-ligand interaction energy, as well as for the graphical representation of the binding-site environment. The interactive docking operation is facilitated by various built-in functions, such as energy minimization, energy contribution analysis and logging of the manipulation trajectory. Interactive modeling functions are incorporated for designing new ligand molecules while considering the binding-site environment and the protein-ligand interaction. As an example of the application of GREEN, a docking study is presented on the complex between trypsin and a synthetic trypsin inhibitor. The program package will be useful for rational drug design, based on the 3D structure of the target protein.

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-09-28

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

In-core Instrument Subcritical Verification (INCISV) - Core Design Verification Method - 358

International Nuclear Information System (INIS)

Prible, M.C.; Heibel, M.D.; Conner, S.L.; Sebastiani, P.J.; Kistler, D.P.

2010-01-01

According to the standard on reload startup physics testing, ANSI/ANS 19.6.1, a plant must verify that the constructed core behaves sufficiently close to the designed core to confirm that the various safety analyses bound the actual behavior of the plant. A large portion of this verification must occur before the reactor operates at power. The INCISV Core Design Verification Method uses the unique characteristics of a Westinghouse Electric Company fixed in-core self powered detector design to perform core design verification after a core reload before power operation. A Vanadium self powered detector that spans the length of the active fuel region is capable of confirming the required core characteristics prior to power ascension; reactivity balance, shutdown margin, temperature coefficient and power distribution. Using a detector element that spans the length of the active fuel region inside the core provides a signal of total integrated flux. Measuring the integrated flux distributions and changes at various rodded conditions and plant temperatures, and comparing them to predicted flux levels, validates all core necessary core design characteristics. INCISV eliminates the dependence on various corrections and assumptions between the ex-core detectors and the core for traditional physics testing programs. This program also eliminates the need for special rod maneuvers which are infrequently performed by plant operators during typical core design verification testing and allows for safer startup activities. (authors)

Site Closure Strategy Model for Creosote Site

International Nuclear Information System (INIS)

Coll, F.R.; Gray, D.R.

2009-01-01

In conjunction with RCRA site corrective action at an active wood preserving facility, a risk-based site closure strategy was developed and incorporated the performance of a dense non-aqueous phase liquid (DNAPL) source recovery remedy, a monitored natural attenuation (MNA) remedy for dissolved phase groundwater, and institutional controls. Innovative creosote DNAPL source recovery has been undertaken at the Site since 1998. Pooled creosote DNAPL is present 90 feet below ground within a transmissive sand and gravel aquifer with a saturated thickness of approximately 80 feet. The creosote DNAPL source is situated on the property boundary of the site and has generated a 1/2 mile off-site dissolved phase plume, creating significant NAPL management and remedial technology verification issues. To date, over 120,000 gallons of creosote DNAPL have been recovered from the subsurface utilizing a modified circulation well technology. A mass discharge flux protocol was developed to serve as a major performance metrics for the continuation of source removal efforts and to support the application of monitored natural attenuation as an associated remedial technology for groundwater. The mass removal success has supported the MNA remedy for dissolved phase groundwater and the associated development of institutional controls. The enacted site management strategy outlines the current and future risk management activities for the Site and represents an appropriate site closure strategy for the Site. (authors)

Independent tube verification and dynamic tracking in et inspection of nuclear steam generator

International Nuclear Information System (INIS)

Xiongzi, Li; Zhongxue, Gan; Lance, Fitzgibbons

2001-01-01

The full text follows. In the examination of pressure boundary tubes in steam generators of commercial pressurized water nuclear power plants (PWR's), it is critical to know exactly which particular tube is being accessed. There are no definitive landmarks or markings on the individual tubes. Today this is done manually, it is tedious, and interrupts the normal inspection work, and is difficult due to the presence of water on the tube surface, plug ends instead of tube openings in the field of view, and varying lighting quality. In order to eliminate the human error and increase the efficiency of operation, there is a need to identify tube position during the inspection process, independent of robot encoder position and motion. A process based on a Cognex MVS-8200 system and its application function package has been developed to independently identify tube locations. ABB Combustion Engineering Nuclear Power's Outage Services group, USPPL in collaboration with ABB Power Plant Laboratories' Advanced Computers and Controls department has developed a new vision-based Independent Tube Verification system (GENESIS-ITVS-TM ). The system employ's a model-based tube-shape detection algorithm and dynamic tracking methodology to detect the true tool position and its offsets from identified tube location. GENESIS-ITVS-TM is an automatic Independent Tube Verification System (ITVS). Independent tube verification is a tube validation technique using computer vision, and not using any robot position parameters. This process independently counts the tubes in the horizontal and vertical axes of the plane of the steam generator tube sheet as the work tool is moved. Thus it knows the true position in the steam generator, given a known starting point. This is analogous to the operator's method of counting tubes for verification, but it is automated. GENESIS-ITVS-TM works independent of the robot position, velocity, or acceleration. The tube position information is solely obtained from

Development of a PC code package for the analysis of research and power reactors

International Nuclear Information System (INIS)

Urli, N.

1992-06-01

Computer codes available for performing reactor physics calculations for nuclear research reactors and power reactors are normally suited for running on mainframe computers. With the fast development in speed and memory of the PCs and affordable prices it became feasible to develop PC versions of commonly used codes. The present work performed under an IAEA sponsored research contract has successfully developed a code package for running on a PC. This package includes a cross-section generating code PSU-LEOPARD and 2D and 1D spatial diffusion codes, MCRAC and MCYC 1D. For adapting PSU-LEOPARD for a PC, the binary library has been reorganized to decimal form, upgraded to FORTRAN-77 standard and arrays and subroutines reorganized to conform to PC compiler. Similarly PC version of MCRAC for FORTRAN-77 and 1D code MCYC 1D have been developed. Tests, verification and bench mark results show excellent agreement with the results obtained from mainframe calculations. The execution speeds are also very satisfactory. 12 refs, 4 figs, 3 tabs

In-Package Chemistry Abstraction

Energy Technology Data Exchange (ETDEWEB)

P.S. Domski

2003-07-21

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

In-Package Chemistry Abstraction

International Nuclear Information System (INIS)

P.S. Domski

2003-01-01

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

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.

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

OpenAIRE

Hu, Cuicui

2010-01-01

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

Packaging for Sustainability

CERN Document Server

Lewis, Helen; Fitzpatrick, Leanne

2012-01-01

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

CDIAC catalog of numeric data packages and computer model packages

International Nuclear Information System (INIS)

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

1993-05-01

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

Siting a low-level waste facility

International Nuclear Information System (INIS)

English, M.R.

1988-01-01

In processes to site disposal facilities for low-level radioactive waste, volunteerism and incentives packages hold more promise for attracting host communities than they have for attracting host states. But volunteerism and incentives packages can have disadvantages as well as advantages. This paper discusses their pros and cons and summarizes the different approaches that states are using in their relationships with local governments

Fingerprint verification prediction model in hand dermatitis.

Science.gov (United States)

Lee, Chew K; Chang, Choong C; Johor, Asmah; Othman, Puwira; Baba, Roshidah

2015-07-01

Hand dermatitis associated fingerprint changes is a significant problem and affects fingerprint verification processes. This study was done to develop a clinically useful prediction model for fingerprint verification in patients with hand dermatitis. A case-control study involving 100 patients with hand dermatitis. All patients verified their thumbprints against their identity card. Registered fingerprints were randomized into a model derivation and model validation group. Predictive model was derived using multiple logistic regression. Validation was done using the goodness-of-fit test. The fingerprint verification prediction model consists of a major criterion (fingerprint dystrophy area of ≥ 25%) and two minor criteria (long horizontal lines and long vertical lines). The presence of the major criterion predicts it will almost always fail verification, while presence of both minor criteria and presence of one minor criterion predict high and low risk of fingerprint verification failure, respectively. When none of the criteria are met, the fingerprint almost always passes the verification. The area under the receiver operating characteristic curve was 0.937, and the goodness-of-fit test showed agreement between the observed and expected number (P = 0.26). The derived fingerprint verification failure prediction model is validated and highly discriminatory in predicting risk of fingerprint verification in patients with hand dermatitis. © 2014 The International Society of Dermatology.

Guidelines for the verification and validation of expert system software and conventional software: User's manual. Volume 7

International Nuclear Information System (INIS)

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

1995-03-01

This report provides a step-by-step guide, or user manual, for personnel responsible for the planning and execution of the verification and validation (V ampersand V), and developmental testing, of expert systems, conventional software systems, and various other types of artificial intelligence systems. While the guide was developed primarily for applications in the utility industry, it applies well to all industries. The user manual has three sections. In Section 1 the user assesses the stringency of V ampersand V needed for the system under consideration, identifies the development stage the system is in, and identifies the component(s) of the system to be tested next. These three pieces of information determine which Guideline Package of V ampersand V methods is most appropriate for those conditions. The V ampersand V Guideline Packages are provided in Section 2. Each package consists of an ordered set of V ampersand V techniques to be applied to the system, guides on choosing the review/evaluation team, measurement criteria, and references to a book or report which describes the application of the method. Section 3 presents details of 11 of the most important (or least well-explained in the literature) methods to assist the user in applying these techniques accurately

Closure report for housekeeping category, Corrective Action Unit 349, Area 12, Nevada Test Site

International Nuclear Information System (INIS)

1998-01-01

This Closure Report summarizes the corrective actions which were completed at the Corrective Action Sites within Corrective Action Unit 349 Area 12 at the Nevada Test Site. Current site descriptions, observations and identification of wastes removed are included on FFACO Corrective Action Site housekeeping closure verification forms

Closure report for housekeeping category, Corrective Action Unit 349, Area 12, Nevada Test Site

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-01-01

This Closure Report summarizes the corrective actions which were completed at the Corrective Action Sites within Corrective Action Unit 349 Area 12 at the Nevada Test Site. Current site descriptions, observations and identification of wastes removed are included on FFACO Corrective Action Site housekeeping closure verification forms.

A new verification film system for routine quality control of radiation fields: Kodak EC-L.

Science.gov (United States)

Hermann, A; Bratengeier, K; Priske, A; Flentje, M

2000-06-01

The use of modern irradiation techniques requires better verification films for determining set-up deviations and patient movements during the course of radiation treatment. This is an investigation of the image quality and time requirement of a new verification film system compared to a conventional portal film system. For conventional verifications we used Agfa Curix HT 1000 films which were compared to the new Kodak EC-L film system. 344 Agfa Curix HT 1000 and 381 Kodak EC-L portal films of different tumor sites (prostate, rectum, head and neck) were visually judged on a light box by 2 experienced physicians. Subjective judgement of image quality, masking of films and time requirement were checked. In this investigation 68% of 175 Kodak EC-L ap/pa-films were judged "good", only 18% were classified "moderate" or "poor" 14%, but only 22% of 173 conventional ap/pa verification films (Agfa Curix HT 1000) were judged to be "good". The image quality, detail perception and time required for film inspection of the new Kodak EC-L film system was significantly improved when compared with standard portal films. They could be read more accurately and the detection of set-up deviation was facilitated.

Groundwater flow code verification ''benchmarking'' activity (COVE-2A): Analysis of participants' work

International Nuclear Information System (INIS)

Dykhuizen, R.C.; Barnard, R.W.

1992-02-01

The Nuclear Waste Repository Technology Department at Sandia National Laboratories (SNL) is investigating the suitability of Yucca Mountain as a potential site for underground burial of nuclear wastes. One element of the investigations is to assess the potential long-term effects of groundwater flow on the integrity of a potential repository. A number of computer codes are being used to model groundwater flow through geologic media in which the potential repository would be located. These codes compute numerical solutions for problems that are usually analytically intractable. Consequently, independent confirmation of the correctness of the solution is often not possible. Code verification is a process that permits the determination of the numerical accuracy of codes by comparing the results of several numerical solutions for the same problem. The international nuclear waste research community uses benchmarking for intercomparisons that partially satisfy the Nuclear Regulatory Commission (NRC) definition of code verification. This report presents the results from the COVE-2A (Code Verification) project, which is a subset of the COVE project

Waste package performance assessment

International Nuclear Information System (INIS)

Lester, D.H.

1981-01-01

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

Edible packaging materials.

Science.gov (United States)

Janjarasskul, Theeranun; Krochta, John M

2010-01-01

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

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

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

Environmental restoration risk-based prioritization work package planning and risk ranking methodology. Revision 2

International Nuclear Information System (INIS)

Dail, J.L.; Nanstad, L.D.; White, R.K.

1995-06-01

This document presents the risk-based prioritization methodology developed to evaluate and rank Environmental Restoration (ER) work packages at the five US Department of Energy, Oak Ridge Field Office (DOE-ORO) sites [i.e., Oak Ridge K-25 Site (K-25), Portsmouth Gaseous Diffusion Plant (PORTS), Paducah Gaseous Diffusion Plant (PGDP), Oak Ridge National Laboratory (ORNL), and the Oak Ridge Y-12 Plant (Y-12)], the ER Off-site Program, and Central ER. This prioritization methodology was developed to support the increased rigor and formality of work planning in the overall conduct of operations within the DOE-ORO ER Program. Prioritization is conducted as an integral component of the fiscal ER funding cycle to establish program budget priorities. The purpose of the ER risk-based prioritization methodology is to provide ER management with the tools and processes needed to evaluate, compare, prioritize, and justify fiscal budget decisions for a diverse set of remedial action, decontamination and decommissioning, and waste management activities. The methodology provides the ER Program with a framework for (1) organizing information about identified DOE-ORO environmental problems, (2) generating qualitative assessments of the long- and short-term risks posed by DOE-ORO environmental problems, and (3) evaluating the benefits associated with candidate work packages designed to reduce those risks. Prioritization is conducted to rank ER work packages on the basis of the overall value (e.g., risk reduction, stakeholder confidence) each package provides to the ER Program. Application of the methodology yields individual work package ''scores'' and rankings that are used to develop fiscal budget requests. This document presents the technical basis for the decision support tools and process

Quantitative analysis of patient-specific dosimetric IMRT verification

International Nuclear Information System (INIS)

Budgell, G J; Perrin, B A; Mott, J H L; Fairfoul, J; Mackay, R I

2005-01-01

Patient-specific dosimetric verification methods for IMRT treatments are variable, time-consuming and frequently qualitative, preventing evidence-based reduction in the amount of verification performed. This paper addresses some of these issues by applying a quantitative analysis parameter to the dosimetric verification procedure. Film measurements in different planes were acquired for a series of ten IMRT prostate patients, analysed using the quantitative parameter, and compared to determine the most suitable verification plane. Film and ion chamber verification results for 61 patients were analysed to determine long-term accuracy, reproducibility and stability of the planning and delivery system. The reproducibility of the measurement and analysis system was also studied. The results show that verification results are strongly dependent on the plane chosen, with the coronal plane particularly insensitive to delivery error. Unexpectedly, no correlation could be found between the levels of error in different verification planes. Longer term verification results showed consistent patterns which suggest that the amount of patient-specific verification can be safely reduced, provided proper caution is exercised: an evidence-based model for such reduction is proposed. It is concluded that dose/distance to agreement (e.g., 3%/3 mm) should be used as a criterion of acceptability. Quantitative parameters calculated for a given criterion of acceptability should be adopted in conjunction with displays that show where discrepancies occur. Planning and delivery systems which cannot meet the required standards of accuracy, reproducibility and stability to reduce verification will not be accepted by the radiotherapy community

SU-E-T-48: A Multi-Institutional Study of Independent Dose Verification for Conventional, SRS and SBRT

International Nuclear Information System (INIS)

Takahashi, R; Kamima, T; Tachibana, H; Baba, H; Itano, M; Yamazaki, T; Ishibashi, S; Higuchi, Y; Shimizu, H; Yamamoto, T; Yamashita, M; Sugawara, Y; Sato, A; Nishiyama, S; Kawai, D; Miyaoka, S

2015-01-01

Purpose: To show the results of a multi-institutional study of the independent dose verification for conventional, Stereotactic radiosurgery and body radiotherapy (SRS and SBRT) plans based on the action level of AAPM TG-114. Methods: This study was performed at 12 institutions in Japan. To eliminate the bias of independent dose verification program (Indp), all of the institutions used the same CT-based independent dose verification software (Simple MU Analysis, Triangle Products, JP) with the Clarkson-based algorithm. Eclipse (AAA, PBC), Pinnacle 3 (Adaptive Convolve) and Xio (Superposition) were used as treatment planning system (TPS). The confidence limits (CL, Mean±2SD) for 18 sites (head, breast, lung, pelvis, etc.) were evaluated in comparison in dose between the TPS and the Indp. Results: A retrospective analysis of 6352 treatment fields was conducted. The CLs for conventional, SRS and SBRT were 1.0±3.7 %, 2.0±2.5 % and 6.2±4.4 %, respectively. In conventional plans, most of the sites showed within 5 % of TG-114 action level. However, there were the systematic difference (4.0±4.0 % and 2.5±5.8 % for breast and lung, respectively). In SRS plans, our results showed good agreement compared to the action level. In SBRT plans, the discrepancy between the Indp was variable depending on dose calculation algorithms of TPS. Conclusion: The impact of dose calculation algorithms for the TPS and the Indp affects the action level. It is effective to set the site-specific tolerances, especially for the site where inhomogeneous correction can affect dose distribution strongly