The long and short of dry vault storage [for spent fuel

International Nuclear Information System (INIS)

Bradley, N.; O'Tallamhain, C.; Grine, C.J.

1984-01-01

The case has been made for purchasing Magnox reactors outside the UK without any prior commitment to reprocessing, the fuel being stored for an interim period. Two alternative storage concepts using natural draught air-cooled vaults have been presented, one based upon experience of 'buffer' storage at gas-cooled reactors in Britain, and te second based on the long term 'interim' storage developed to a detailed design stage for British AGRs. Although storage scenarios for new Magnox stations are discussed, they may also be of interest in relation to other types of reactor. (U.K.)

Key technical issues relating to safety of spent fuel dry storage in vaults: CASCAD system

Energy Technology Data Exchange (ETDEWEB)

Berge, F [Societe Generale pour les Techniques Nouvelles (SGN), 78 - Saint-Quentin-en-Yvelines (France)

1994-12-31

The operating CASCAD Facility at the Cadarashe site (FR) was commissioned in May 1990. Fuel is received in tight canisters which are transferred to storage pits in the vault and scheduled to be stored for up to 50 years. Canistering operations are performed in a cell of the reactor building.The paper describes the main functions of the facility as: cask receipt and shipping; fuel unloading; fuel conditioning; canisters emplacements in storage location; fuel storage; fuel retrieving and shipping at the end of the storage period; operation system and operation organization. Safety characteristics of the facility discussed are: fuel decay heat removal; subcriticality control and radiological protection. The fuel decay heat removal has two main purposes: (1) maintaining rod cladding temperature below a set limit in order to keep the fuel in its as received condition; (2) maintaining structures and equipment performing a safety function below the design temperature. The features of the sub-criticality control in the storage vault are such that sub-criticality in normal and accidental conditions is provided by the arrangement of pits in the vault. Radiological protection is based on limiting collective and individual annual dose equivalent to ALARA levels ensuring that they remain in any case below the set limits. Radiological protection system described consists in: confinement of radioactive materials for protection against its dissemination; radiation shielding for protection against irradiation. It is pointed out that all technical solutions presented are based on or adapted from proven technologies used in operating facilities in France or in other countries. The solution not only benefits from the experience of SGN in the design, construction and start-up of facilities for fuel or high level waste handling and storage, but also from the experience of the CEA and COGEMA groups in operating such facilities. 2 figs., 1 ref.

Development of evaluation method for heat removal design of dry storage facilities. pt. 1. Heat removal test on vault storage system of cross flow type

International Nuclear Information System (INIS)

Sakamoto, Kazuaki; Koga, Tomonari; Wataru, Masumi; Hattori, Yasuo

1997-01-01

The report describes the result of heat removal test of passive cooling vault storage system of cross flow type using 1/5 scale model. Based on a prospect of steady increase in the amount of spent fuel, it is needed to establish large capacity dry storage technologies for spent fuel. Air flow patterns, distributions of air temperature and velocity were measured, by which heat removal characteristics of the system were made clear. Air flow patterns in the storage module depended on the ratio of the buoyant force to the inertial force; the former generated by the difference of air temperatures and the height of the storage module, the latter by the difference of air densities between the outlet of the storage module and ambience and the height of the chimney of the storage facility. A simple method to estimate air flow patterns in the storage module was suggested, where Ri(Richardson) number was applied to represent the ratio. Moreover, heat transfer coefficient from a model of storage tube to cooling air was evaluated, and it was concluded that the generalized expression of heat transfer coefficient for common heat exchangers could be applied to the vault storage system of cross flow type, in which dozens of storage tubes were placed in a storage module. (author)

Searchable Data Vault: Encrypted Queries in Secure Distributed Cloud Storage

Directory of Open Access Journals (Sweden)

Geong Sen Poh

2017-05-01

Full Text Available Cloud storage services allow users to efficiently outsource their documents anytime and anywhere. Such convenience, however, leads to privacy concerns. While storage providers may not read users’ documents, attackers may possibly gain access by exploiting vulnerabilities in the storage system. Documents may also be leaked by curious administrators. A simple solution is for the user to encrypt all documents before submitting them. This method, however, makes it impossible to efficiently search for documents as they are all encrypted. To resolve this problem, we propose a multi-server searchable symmetric encryption (SSE scheme and construct a system called the searchable data vault (SDV. A unique feature of the scheme is that it allows an encrypted document to be divided into blocks and distributed to different storage servers so that no single storage provider has a complete document. By incorporating the scheme, the SDV protects the privacy of documents while allowing for efficient private queries. It utilizes a web interface and a controller that manages user credentials, query indexes and submission of encrypted documents to cloud storage services. It is also the first system that enables a user to simultaneously outsource and privately query documents from a few cloud storage services. Our preliminary performance evaluation shows that this feature introduces acceptable computation overheads when compared to submitting documents directly to a cloud storage service.

HEU Measurements of Holdup and Recovered Residue in the Deactivation and Decommissioning Activities of the 321-M Reactor Fuel Fabrication Facility at the Savannah River Site

Energy Technology Data Exchange (ETDEWEB)

DEWBERRY, RAYMOND; SALAYMEH, SALEEM R.; CASELLA, VITO R.; MOORE, FRANK S.

2005-03-11

This paper contains a summary of the holdup and material control and accountability (MC&A) assays conducted for the determination of highly enriched uranium (HEU) in the deactivation and decommissioning (D&D) of Building 321-M at the Savannah River Site (SRS). The 321-M facility was the Reactor Fuel Fabrication Facility at SRS and was used to fabricate HEU fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the SRS production reactors. The facility operated for more than 35 years. During this time thousands of uranium-aluminum-alloy (U-Al) production reactor fuel tubes were produced. After the facility ceased operations in 1995, all of the easily accessible U-Al was removed from the building, and only residual amounts remained. The bulk of this residue was located in the equipment that generated and handled small U-Al particles and in the exhaust systems for this equipment (e.g., Chip compactor, casting furnaces, log saw, lathes A & B, cyclone separator, Freon{trademark} cart, riser crusher, ...etc). The D&D project is likely to represent an important example for D&D activities across SRS and across the Department of Energy weapons complex. The Savannah River National Laboratory was tasked to conduct holdup assays to quantify the amount of HEU on all components removed from the facility prior to placing in solid waste containers. The U-235 holdup in any single component of process equipment must not exceed 50 g in order to meet the container limit. This limit was imposed to meet criticality requirements of the low level solid waste storage vaults. Thus the holdup measurements were used as guidance to determine if further decontamination of equipment was needed to ensure that the quantity of U-235 did not exceed the 50 g limit and to ensure that the waste met the Waste Acceptance Criteria (WAC) of the solid waste storage vaults. Since HEU is an accountable nuclear material, the holdup assays and assays of recovered

Radiological performance assessment for the E-Area Vaults Disposal Facility

Energy Technology Data Exchange (ETDEWEB)

Cook, J.R.; Hunt, P.D. [Westinghouse Savannah River Co., Aiken, SC (United States)

1994-04-15

The E-Area Vaults (EAVs) located on a 200 acre site immediately north of the current LLW burial site at Savannah River Site will provide a new disposal and storage site for solid, low-level, non-hazardous radioactive waste. The EAV Disposal Facility will contain several large concrete vaults divided into cells. Three types of structures will house four designated waste types. The Intermediate Level Non-Tritium Vaults will receive waste radiating greater than 200 mR/h at 5 cm from the outer disposal container. The Intermediate Level Tritium Vaults will receive waste with at least 10 Ci of tritium per package. These two vaults share a similar design, are adjacent, share waste handling equipment, and will be closed as one facility. The second type of structure is the Low Activity Waste Vaults which will receive waste radiating less than 200 mR/h at 5 cm from the outer disposal container and containing less than 10 Ci of tritium per package. The third facility, the Long Lived Waste Storage Building, provides covered, long term storage for waste containing long lived isotopes. Two additional types of disposal are proposed: (1) trench disposal of suspect soil, (2) naval reactor component disposal. To evaluate the long-term performance of the EAVs, site-specific conceptual models were developed to consider: (1) exposure pathways and scenarios of potential importance; (2) potential releases from the facility to the environment; (3) effects of degradation of engineered features; (4) transport in the environment; (5) potential doses received from radionuclides of interest in each vault type.

Radiological performance assessment for the E-Area Vaults Disposal Facility

International Nuclear Information System (INIS)

Cook, J.R.; Hunt, P.D.

1994-01-01

The E-Area Vaults (EAVs) located on a 200 acre site immediately north of the current LLW burial site at Savannah River Site will provide a new disposal and storage site for solid, low-level, non-hazardous radioactive waste. The EAV Disposal Facility will contain several large concrete vaults divided into cells. Three types of structures will house four designated waste types. The Intermediate Level Non-Tritium Vaults will receive waste radiating greater than 200 mR/h at 5 cm from the outer disposal container. The Intermediate Level Tritium Vaults will receive waste with at least 10 Ci of tritium per package. These two vaults share a similar design, are adjacent, share waste handling equipment, and will be closed as one facility. The second type of structure is the Low Activity Waste Vaults which will receive waste radiating less than 200 mR/h at 5 cm from the outer disposal container and containing less than 10 Ci of tritium per package. The third facility, the Long Lived Waste Storage Building, provides covered, long term storage for waste containing long lived isotopes. Two additional types of disposal are proposed: (1) trench disposal of suspect soil, (2) naval reactor component disposal. To evaluate the long-term performance of the EAVs, site-specific conceptual models were developed to consider: (1) exposure pathways and scenarios of potential importance; (2) potential releases from the facility to the environment; (3) effects of degradation of engineered features; (4) transport in the environment; (5) potential doses received from radionuclides of interest in each vault type

HEU Transparency Implementation Program and its Radiation Safety Program

International Nuclear Information System (INIS)

Radev, R

2002-01-01

In February 1993, the Governments of the United States (U.S.) and the Russian Federation (R.F.) signed a bilateral Agreement for the U.S. purchase of low enriched uranium (LEU) derived from 500 metric tons (MT) of highly enriched uranium (HEU) resulting from the dismantlement of Russian nuclear weapons. The HEU Purchase Agreement serves important national security and nonproliferation policy imperatives for both countries since its implementation reduces the quantity of surplus Russian HEU that could be stolen and diverted for weapons use. In return, Russia receives much needed U.S. dollars over a 20-year delivery period. In 2001, Russia received over half a billion US dollars from the purchase of the LEU blended from 30 MT HEU. As part of this Agreement, transparency rights were agreed upon that provide confidence to both governments that the nonproliferation objectives of the Agreement are being fulfilled. While the U.S. Department of State, in concert with the U.S. Department of Energy's (DOE) National Nuclear Security Administration (NNSA) is responsible negotiating transparency rights associated with this nuclear material, the NNSA is responsible for implementing those rights. These rights allow U.S. and R.F., personnel (called ''monitors'') to visit the processing facilities and observe the steps for processing the HEU into fuel for nuclear reactors. In this fashion, the processing of HEU to LEU is made ''transparent.'' For DOE, there are three transparency objectives: (1) that the HEU is extracted from nuclear weapons, (2) that this same HEU is oxidized, and (3) that the HEU is blended into LEU. For MINATOM, the transparency objective is: (1) that the LEU is fabricated into fuel for commercial nuclear power reactors: The transparency is based on visits by designated transparency monitors (100 preapproved U.S. and Russian monitors) with specific rights to monitor and to access storage and processing areas to provide confidence that the nonproliferation goals

Dose analysis on high performance vault storage system of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Shibata, Kei-ichiro; Maki, Koichi; Shimizu, Masashi; Oda, Masashi; Kumagai, Naoki [Hitachi, Ltd., Power and Industrial Systems R and D Laboratory, Hitachi, Ibaraki (Japan); Hoshikawa, Tadahiro; Oyama, Kenichi; Kanai, Hidetoshi [Hitachi Ltd., Ibaraki (Japan). Hitachi Works

2000-03-01

The radiation shielding design for the high performance vault storage system is studied in order to satisfy the design targets at the controlled area and the site boundaries. The additional gamma-ray shields in front of the storage tubes and the shielding structures at the entrance of the ducts are installed. The dose rates at the inlet and the outlet of the ducts are estimated by 3D calculation. And the dose rates at the controlled area and the site boundaries are evaluated taking the effect of the direct radiation and the indirect one (skyshine) into consideration. The dose rates at the controlled area and the site boundaries are about 7x10{sup -7} Sv/h and 3x10{sup -10} Sv/h, respectively. Thus, we have the prospect to satisfy the design targets. (author)

HEU to LEU conversion and blending facility: UNH blending alternative to produce LEU oxide for disposal

International Nuclear Information System (INIS)

1995-09-01

The United States Department of Energy (DOE) is examining options for the disposition of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. Disposition is a process of use or disposal of material that results in the material being converted to a form that is substantially and inherently more proliferation-resistant than is the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. This report provides data to be used in the environmental impact analysis for the uranyl nitrate hexahydrate blending option to produce oxide for disposal. This the Conversion and Blending Facility (CBF) alternative will have two missions (1) convert HEU materials into HEU uranyl nitrate (UNH) and (2) blend the HEU uranyl nitrate with depleted and natural assay uranyl nitrate to produce an oxide that can be stored until an acceptable disposal approach is available. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The blended LEU will be produced as a waste suitable for storage or disposal

HEU to LEU conversion and blending facility: UNH blending alternative to produce LEU oxide for disposal

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

The United States Department of Energy (DOE) is examining options for the disposition of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. Disposition is a process of use or disposal of material that results in the material being converted to a form that is substantially and inherently more proliferation-resistant than is the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. This report provides data to be used in the environmental impact analysis for the uranyl nitrate hexahydrate blending option to produce oxide for disposal. This the Conversion and Blending Facility (CBF) alternative will have two missions (1) convert HEU materials into HEU uranyl nitrate (UNH) and (2) blend the HEU uranyl nitrate with depleted and natural assay uranyl nitrate to produce an oxide that can be stored until an acceptable disposal approach is available. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The blended LEU will be produced as a waste suitable for storage or disposal.

Monitored Retrievable Storage conceptual system studies: closed-cycle vault

International Nuclear Information System (INIS)

Washington, J.A.; Ganley, J.T.

1984-02-01

The Nuclear Waste Policy Act of 1982 requires the DOE to submit a proposal to Congress by June 1985 for the construction of one or more Monitored Retrieval Storage (MRS) facilities. In response, the DOE initiated studies to develop system descriptions and cost estimates for preconceptual designs of storage concepts suitable for use at MRS facilities. This report provides a system description and cost estimates for a Closed-Cycle Vault (CCV) MRS facility. The facility description is divided into four parts: (1) the R and H area, (2) the interface facility, (3) the on-site transport system, and (4) the storage system. The MRS facility has been designed to meet handling rates of 1800 and 3000 MTU/yr. The corresponding peak inventories are 15,000 and 72,000 MTU. Three types of cases were considered, based on the material to be stored: (1) Spent fuel only; (2) HLW and TRU waste; and (3) HLW only. For each of these three types, a cost estimate was done for a 15,000 and a 72,000 MTU facility, resulting in six different cost estimates. Section 4 presents the cost analysis of the CCV MRS system. Tables 4-2 through 4-7 give the construction or capital costs for the six cases. Tables 4-8 through 4-13 show the total discounted life-cycle costs for each of the six cases. These life-cycle costs include operating and decommissioning costs. These tables also show the time distribution of the capital costs. Table 2-1 summarizes the capital, operating, and discounted costs for the six cases studied. 2 references, 15 figures, 18 tables

PHASE II VAULT TESTING OF THE ARGONNE RFID SYSTEM

Energy Technology Data Exchange (ETDEWEB)

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

2012-06-25

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

Phase II Vault Testing of the Argonne RFID System

International Nuclear Information System (INIS)

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

2012-01-01

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

HEU to LEU Conversion and Blending Facility: UNH blending alternative to produce LEU UNH for commercial use

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. The nuclear material is converted to a form that is more proliferation-resistant than the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. Five technologies for blending HEU will be assessed. This document provides data to be used in the environmental impact analysis for the UNH blending HEU disposition option. Process requirements, resource needs, employment needs, waste/emissions from plant, hazards, accident scenarios, and intersite transportation are discussed.

Stationary and protable instruments for assay of HEU [highly enriched uranium] solids holdup

International Nuclear Information System (INIS)

Russo, P.A.; Sprinkle, J.K. Jr.; Stephens, M.M.; Brumfield, T.L.; Gunn, C.S.; Watson, D.R.

1987-01-01

Two NaI(Tl)-based instruments, one stationary and one portable, designed for automated assay of highly enriched uranium (HEU) solids holdup, are being evaluated at the scrap recovery facility of the Oak Ridge Y-12 Plant. The stationary instrument, a continuous monitor of HEU within the filters of the chip burner exhaust system, measures the HEU deposits that accumulate erratically and rapidly during chip burner operation. The portable system was built to assay HEU in over 100 m of elevated piping used to transfer UO 3 , UO 2 , and UF 4 powder to, from, and between the fluid bed conversion furnances and the powder storage hoods. Both instruments use two detector heads. Both provide immediate automatic readout of accumulated HEU mass. The 186-keV 235 U gamma ray is the assay signature, and the 60-keV gamma ray from an 241 Am source attached to each detector is used to normalize the 186-keV rate. The measurement geometries were selected for compatibility with simple calibration models. The assay calibrations were calculated from these models and were verified and normalized with measurements of HEU standards built to match geometries of uniform accumulations on the surfaces of the process equipment. This instrumentation effort demonstrates that simple calibration models can often be applied to unique measurement geometries, minimizing the otherwise unreasonable requirements for calibration standards and allowing extension of the measurements to other process locations

SnoVault and encodeD: A novel object-based storage system and applications to ENCODE metadata.

Directory of Open Access Journals (Sweden)

Benjamin C Hitz

Full Text Available The Encyclopedia of DNA elements (ENCODE project is an ongoing collaborative effort to create a comprehensive catalog of functional elements initiated shortly after the completion of the Human Genome Project. The current database exceeds 6500 experiments across more than 450 cell lines and tissues using a wide array of experimental techniques to study the chromatin structure, regulatory and transcriptional landscape of the H. sapiens and M. musculus genomes. All ENCODE experimental data, metadata, and associated computational analyses are submitted to the ENCODE Data Coordination Center (DCC for validation, tracking, storage, unified processing, and distribution to community resources and the scientific community. As the volume of data increases, the identification and organization of experimental details becomes increasingly intricate and demands careful curation. The ENCODE DCC has created a general purpose software system, known as SnoVault, that supports metadata and file submission, a database used for metadata storage, web pages for displaying the metadata and a robust API for querying the metadata. The software is fully open-source, code and installation instructions can be found at: http://github.com/ENCODE-DCC/snovault/ (for the generic database and http://github.com/ENCODE-DCC/encoded/ to store genomic data in the manner of ENCODE. The core database engine, SnoVault (which is completely independent of ENCODE, genomic data, or bioinformatic data has been released as a separate Python package.

HEU to LEU conversion and blending facility: Metal blending alternative to produce LEU oxide for disposal

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. The nuclear material is converted to a form more proliferation- resistant than the original form. Blending HEU (highly enriched uranium) with less-enriched uranium to form LEU has been proposed as a disposition option. Five technologies are being assessed for blending HEU. This document provides data to be used in environmental impact analysis for the HEU-LEU disposition option that uses metal blending with an oxide waste product. It is divided into: mission and assumptions, conversion and blending facility descriptions, process descriptions and requirements, resource needs, employment needs, waste and emissions from plant, hazards discussion, and intersite transportation.

HEU to LEU conversion and blending facility: Metal blending alternative to produce LEU oxide for disposal

International Nuclear Information System (INIS)

1995-09-01

US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. The nuclear material is converted to a form more proliferation- resistant than the original form. Blending HEU (highly enriched uranium) with less-enriched uranium to form LEU has been proposed as a disposition option. Five technologies are being assessed for blending HEU. This document provides data to be used in environmental impact analysis for the HEU-LEU disposition option that uses metal blending with an oxide waste product. It is divided into: mission and assumptions, conversion and blending facility descriptions, process descriptions and requirements, resource needs, employment needs, waste and emissions from plant, hazards discussion, and intersite transportation

An estimation of tritium inventory limits for the E-Area vaults

International Nuclear Information System (INIS)

Yu, A.D.; Cook, J.R.

1991-01-01

At the request of Waste Management, Interim Waste Technology has conducted a modeling study to estimate the tritium inventory limits for the E-Area vaults. These inventory limits are based on the groundwater impact of the planned waste disposal. The tritium inventory limit for an Intermediate Level Tritium Vault (ILTV) is estimated to be 400,000 Curies with a 100 year storage period. During this period, it is assumed that the ILTV will be vented, any leachate will be extracted, and its performance will be carefully monitored. The tritium inventory limits for a Low Activity Waste Vault (LAWV) and an Intermediate Level Non-Tritium Vault (ILNTV) are estimated to be 15,000 and 11,000 Curies, respectively. Venting and leachate extraction were not assumed necessary. These operational alternatives would further enhance the performance of these vaults. These limits are significantly higher than the forecasted maximum tritium inventories for the vaults. Details of the modeling study are described in the attached report

HEU to LEU conversion and blending facility: Oxide blending alternative to produce LEU oxide for commercial use

International Nuclear Information System (INIS)

1995-09-01

The United States Department of Energy (DOE) is examining options for the disposition of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. Disposition is a process of use or disposal of material that results in the material being converted to a form that is substantially and inherently more proliferation-resistant than the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. This document provides data to be used in the environmental impact analysis for the oxide blending HEU disposition option. This option provides for a yearly HEU throughput of 1 0 metric tons (MT) of uranium metal with an average U235 assay of 50% blended with 165 MT of natural assay triuranium octoxide (U 3 O 8 ) per year to produce 177 MT of 4% U235 assay U 3 O 8 , for LWR fuel. Since HEU exists in a variety of forms and not necessarily in the form to be blended, worst case scenarios for preprocessing prior to blending will be assumed for HEU feed streams

Radionuclide limits for vault disposal at the Savannah River Site

International Nuclear Information System (INIS)

Cook, James R.

1992-01-01

The Savannah River Site is developing a facility called the E-Area Vaults which will serve as the new radioactive waste disposal facility beginning early in 1992. The facility will employ engineered below-grade concrete vaults for disposal and above grade storage for certain long-lived mobile radionuclides. This report documents the determination of interim upper limits for radionuclide inventories and concentrations which should be allowed in the disposal structures. The work presented here will aid in the development of both waste acceptance criteria and operating limits for the E-Area Vaults. Disposal limits for forty isotopes which comprise the SRS waste streams were determined. The limits are based on total facility and vault inventories for those radionuclides which impact groundwater) and on waste package concentrations for those radionuclides which could affect intruders. (author)

HEU Holdup Measurements on 321-M A-Lathe

International Nuclear Information System (INIS)

Dewberry, R.A.

2002-01-01

The Analytical Development Section of SRTC was requested by the Facilities Disposition Division (FDD) of the Savannah River Site to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. The 321-M facility was used to fabricate enriched uranium fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the production reactors. The results of the holdup assays are essential for determining compliance with the solid waste Waste Acceptance Criteria, Material Control and Accountability, and to meet criticality safety controls. Three measurement systems were used to determine highly enriched uranium (HEU) holdup. This report covers holdup measurements on the A-Lathe that was used to machine uranium-aluminum-alloy (U-Al). Our results indicated that the lathe contained more than the limits stated in the Waste Acceptance Criteria (WAC) for the solid waste E-Area Vaults. Thus the lathe was decontaminated three times and assayed four times in order to bring the amounts of uranium to an acceptable content. This report will discuss the methodology, Non-Destructive Assay (NDA) measurements, and results of the U-235 holdup on the lathe

HEU to LEU conversion and blending facility: Oxide blending alternative to produce LEU oxide for commercial use

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

The United States Department of Energy (DOE) is examining options for the disposition of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. Disposition is a process of use or disposal of material that results in the material being converted to a form that is substantially and inherently more proliferation-resistant than the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. This document provides data to be used in the environmental impact analysis for the oxide blending HEU disposition option. This option provides for a yearly HEU throughput of 1 0 metric tons (MT) of uranium metal with an average U235 assay of 50% blended with 165 MT of natural assay triuranium octoxide (U{sub 3} O{sub 8}) per year to produce 177 MT of 4% U235 assay U{sub 3} O{sub 8}, for LWR fuel. Since HEU exists in a variety of forms and not necessarily in the form to be blended, worst case scenarios for preprocessing prior to blending will be assumed for HEU feed streams.

31 CFR 540.306 - Highly Enriched Uranium (HEU).

Science.gov (United States)

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Highly Enriched Uranium (HEU). 540...) OFFICE OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.306 Highly Enriched Uranium (HEU). The term highly...

HEU to LEU Conversion and Blending Facility: UF6 blending alternative to produce LEU UF6 for commercial use

International Nuclear Information System (INIS)

1995-09-01

US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials; the nuclear material will be converted to a form more proliferation- resistant than the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. Five technologies for blending HEU will be assessed; blending as UF 6 to produce a UF 6 product for commercial use is one of them. This document provides data to be used in the environmental impact analysis for the UF 6 blending HEU disposition option. Resource needs, employment needs, waste and emissions from plant, hazards, accident scenarios, and intersite transportation are discussed

Development of evaluation method for heat removal design of dry storage facilities. Pt. 4. Numerical analysis on vault storage system of cross flow type

International Nuclear Information System (INIS)

Sakamoto, Kazuaki; Hattori, Yasuo; Koga, Tomonari; Wataru, Masumi

1999-01-01

On the basis of the result of the heat removal test on vault storage system of cross flow type using the 1/5 scale model, an evaluation method for the heat removal design was established. It was composed of the numerical analysis for the convection phenomena of air flow inside the whole facility and that for the natural convection and the detailed turbulent mechanism near the surface of the storage tube. In the former analysis, air temperature distribution in the storage area obtained by the calculation gave good agreement within ±3degC with the test result. And fine turbulence models were introduced in the latter analysis to predict the separation flow in the boundary layer near the surface of the storage tube and the buoyant flow generated by the heat from the storage tube. Furthermore, the properties of removing the heat in a designed full-scale storage facility, such as flow pattern in the storage area, temperature and heat transfer rate of the storage tubes, were evaluated by using each of three methods, which were the established numerical analysis method, the experimental formula demonstrated in the heat removal test and the conventional evaluation method applied to the past heat removal design. As a result, the safety margin and issues included in the methods were grasped, and the measures to make a design more rational were proposed. (author)

Addendum 1 to CSER 96-025: PFP storage of 9.25/9.5 inch tall, 4.4 kg Pu cans on existing Vault 4 pedestals

International Nuclear Information System (INIS)

Hillesland, K.E.

1997-01-01

A nuclear criticality safety analysis has been performed to increase the approved plutonium mass limit for cans stored in Vault number-sign 4 cubicles at PFP. The original CSER 96-025 accommodated the storage of 4.4 kg of plutonium in PuO, (5.0 kg PuO,) in Vault number-sign 4 by requiring that half the cubicles be left vacant. This addendum allows for all the cubicles to be used, but with a fissile plutonium mass limit of 58 kg per cubicle. A mass limit for each cubical allows for storage of a larger number of cans if some have less than the 4.4 kg Pu limit per can. The highest k., calculated is 0.932 + 0.003 when an overbatched can is present in every fourth cubicle. This is below the criticality safety limit of kff 0.935, and consequently, an increase of plutonium mass to 4.4 kg per can is within acceptable safety limits for the given mass limit

Nuclear waste vault sealing

International Nuclear Information System (INIS)

Gyenge, M.

1980-01-01

A nuclear waste vault must be designed and built to ensure adequate isolation of the nuclear wastes from human contact. Consequently, after a vault has been fully loaded, it must be adequately sealed off to prevent radionuclide migration which may be provided by circulating groundwater. Vault sealing entails four major aspects, i.e.: (a) vault grouting; (b) borehole sealing; (c) buffer packing; and (d) backfilling. Of particular concern in vault sealing are the physical and chemical properties of the sealing material, its long-term durability and stability, and the techniques used for its emplacement. Present sealing technology and sealing materials are reviewed in terms of the particular needs of vault sealing. Areas requiring research and development are indicated

252Cf-source-driven noise analysis measurements for characterization of concrete highly enriched uranium (HEU) storage vaults

International Nuclear Information System (INIS)

Valentine, T.E.; Mihalczo, J.T.

1993-01-01

The 252 Cf-source-driven noise analysis method has been used in measurements for subcritical configurations of fissile systems for a variety of applications. Measurements of 25 fissile systems have been performed with a wide variety of materials and configurations. This method has been applied to measurements for (1) initial fuel loading of reactors, (2) quality assurance of reactor fuel elements, (3) fuel preparation facilities, (4) fuel processing facilities, (5) fuel storage facilities, (6) zero-power testing of reactors, and (7) verification of calculational methods for assemblies with the neutron k 252 Cf source and commercially available detectors was feasible and to determine if the measurement could characterize the ability of the concrete to isolate the fissile material

HEU to LEU Conversion and Blending Facility: UF{sub 6} blending alternative to produce LEU UF{sub 6} for commercial use

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials; the nuclear material will be converted to a form more proliferation- resistant than the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. Five technologies for blending HEU will be assessed; blending as UF{sub 6} to produce a UF{sub 6} product for commercial use is one of them. This document provides data to be used in the environmental impact analysis for the UF{sub 6} blending HEU disposition option. Resource needs, employment needs, waste and emissions from plant, hazards, accident scenarios, and intersite transportation are discussed.

HEU/LEU-conversion of BER II successfully finished

International Nuclear Information System (INIS)

Haas, K.; Fischer, C.-O.; Krohn, H.

2000-01-01

The BER II (Berliner Experimental Reactor) research reactor is a swimming pool type reactor located in Berlin, Germany. The reactor operates with a thermal power of 10 MW and is primarily used to produce neutrons for neutron scattering experiments. The conversion from HEU- to LEU-fuel elements began in August, 1997. At the last RERTR Meeting 1999 in Budapest, Hungary, Hahn-Meitner-Institut (HMI) presented a 'Status Report' on the conversion of 10 HEU/LEU mixed cores. In February 2000, HMI finished the HEU/LEU-conversion. Hereby, the first pure LEU-standard-core went into operation. Our second LEU-core just ends its operation at the end of July. The third LEU-core will be built up in the beginning of August. The average burn-up rate was improved from 50 - 55% (HEU) to 60 - 65% (LEU). Therefore, only 14 elements/year are now used instead of 28/year. The following report describes our first steps in building pure LEU-cores from mixed HEU/LEU-cores, as well as our initial experience using the pure LEU-cores. (author)

Analysis of Hydrogen Generation and Accumulation in U-233 Tube Vaults

International Nuclear Information System (INIS)

Ally, M.R.; Willis, K.J.

1999-01-01

The purpose of the 233 U Safe Storage Program is to enhance the safe storage of 233 U-bearing materials. This report describes the work done at the Oak Ridge National Laboratory's Radiochemical Development Facility (RDF) to address questions related to possible hydrogen generation and accumulation in 233 U tube vaults. The objective of this effort was to verify assumptions in the mathematical model used to estimate the hydrogen content of the gaseous atmosphere that possibly could occur inside the tube vaults in Building 3019 and to evaluate proposed measures for mitigating any hydrogen concerns. A mathematical model was developed using conservative assumptions to evaluate possible hydrogen generation and accumulation in the tube vaults. The model concluded that an equilibrium concentration would be established below the lower flammability limit (LFL) of 4.1% hydrogen. The major assumptions used in the model that were validated are as follows: (1) The shield plug does not form a seal with the tube vault wall, thus allowing the hydrogen gas to diffuse past the shield plug to the upper section of the tube vault. (2) The tube vault end-cap leaks sufficiently to allow air to be drawn into the tube vault by the off-gas system, thereby purging hydrogen from the upper section of the tube vault. (3) Any hydrogen gas generated completely mixes with the other gases present in the lower section of the tube vault and does not stratify beneath the shield plug. (4) The diffusion coefficient determined from the literature for constant diffusion of hydrogen in air is valid. The coefficient is corrected for temperatures from 0 to 25 C. Another assumption used in the model, that hydrogen generated by radiolytic decomposition of hydrogen-bearing materials (e.g., moisture and plastic) leaks from the cans under steady-state condition, as opposed to a sudden release resulting from rupture of the can(s), was beyond the scope of this investigation. Several parameters from the original

E-Area Vault Concrete Material Property And Vault Durability/Degradation Projection Recommendations

Energy Technology Data Exchange (ETDEWEB)

Phifer, M. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2014-03-11

Subsequent to the 2008 E-Area Low-Level Waste Facility (ELLWF) Performance Assessment (PA) (WSRC 2008), two additional E-Area vault concrete property testing programs have been conducted (Dixon and Phifer 2010 and SIMCO 2011a) and two additional E-Area vault concrete durability modeling projections have been made (Langton 2009 and SIMCO 2012). All the information/data from these reports has been evaluated and consolidated herein by the Savannah River National Laboratory (SRNL) at the request of Solid Waste Management (SWM) to produce E-Area vault concrete hydraulic and physical property data and vault durability/degradation projection recommendations that are adequately justified for use within associated Special Analyses (SAs) and future PA updates. The Low Activity Waste (LAW) and Intermediate Level (IL) Vaults structural degradation predictions produced by Carey 2006 and Peregoy 2006, respectively, which were used as the basis for the 2008 ELLWF PA, remain valid based upon the results of the E-Area vault concrete durability simulations reported by Langton 2009 and those reported by SIMCO 2012. Therefore revised structural degradation predictions are not required so long as the mean thickness of the closure cap overlying the vaults is no greater than that assumed within Carey 2006 and Peregoy 2006. For the LAW Vault structural degradation prediction (Carey 2006), the mean thickness of the overlying closure cap was taken as nine feet. For the IL Vault structural degradation prediction (Peregoy 2006), the mean thickness of the overlying closure cap was taken as eight feet. The mean closure cap thicknesses as described here for both E-Area Vaults will be included as a key input and assumption (I&A) in the next revision to the closure plan for the ELLWF (Phifer et al. 2009). In addition, it has been identified as new input to the PA model to be assessed in the ongoing update to the new PA Information UDQE (Flach 2013). Once the UDQE is approved, the SWM Key I

E-Area Vault Concrete Material Property And Vault Durability/Degradation Projection Recommendations

International Nuclear Information System (INIS)

Phifer, M. A.

2014-01-01

Subsequent to the 2008 E-Area Low-Level Waste Facility (ELLWF) Performance Assessment (PA) (WSRC 2008), two additional E-Area vault concrete property testing programs have been conducted (Dixon and Phifer 2010 and SIMCO 2011a) and two additional E-Area vault concrete durability modeling projections have been made (Langton 2009 and SIMCO 2012). All the information/data from these reports has been evaluated and consolidated herein by the Savannah River National Laboratory (SRNL) at the request of Solid Waste Management (SWM) to produce E-Area vault concrete hydraulic and physical property data and vault durability/degradation projection recommendations that are adequately justified for use within associated Special Analyses (SAs) and future PA updates. The Low Activity Waste (LAW) and Intermediate Level (IL) Vaults structural degradation predictions produced by Carey 2006 and Peregoy 2006, respectively, which were used as the basis for the 2008 ELLWF PA, remain valid based upon the results of the E-Area vault concrete durability simulations reported by Langton 2009 and those reported by SIMCO 2012. Therefore revised structural degradation predictions are not required so long as the mean thickness of the closure cap overlying the vaults is no greater than that assumed within Carey 2006 and Peregoy 2006. For the LAW Vault structural degradation prediction (Carey 2006), the mean thickness of the overlying closure cap was taken as nine feet. For the IL Vault structural degradation prediction (Peregoy 2006), the mean thickness of the overlying closure cap was taken as eight feet. The mean closure cap thicknesses as described here for both E-Area Vaults will be included as a key input and assumption (I and A) in the next revision to the closure plan for the ELLWF (Phifer et al. 2009). In addition, it has been identified as new input to the PA model to be assessed in the ongoing update to the new PA Information UDQE (Flach 2013). Once the UDQE is approved, the SWM Key I and

TRANSPARENCY: Tracking Uranium under the U.S./Russian HEU Purchase Agreement

International Nuclear Information System (INIS)

Benton, J B; Decman, D J; Leich, D A

2005-01-01

By the end of August, 2005, the Russia Federation delivered to the United States (U.S.) more than 7,000 metric tons (MT) of low enriched uranium (LEU) containing approximately 46 million SWU and 75,000 MT of natural uranium. This uranium was blended down from weapons-grade (nominally enriched to 90% 235 U) highly enriched uranium (HEU) under the 1993 HEU Purchase Agreement that provides for the blend down of 500 MT HEU into LEU for use as fuel in commercial nuclear reactors. The HEU Transparency Program, under the National Nuclear Security Administration (NNSA), monitored the conversion and blending of the more than 250 MT HEU used to produce this LEU. The HEU represents more than half of the 500 MT HEU scheduled to be blended down through the year 2013 and is equivalent to the elimination of more than 10,000 nuclear devices. The HEU Transparency Program has made considerable progress in its mission to develop and implement transparency measures necessary to assure that Russian HEU extracted from dismantled Russian nuclear weapons is blended down into LEU for delivery to the United States. U.S. monitor observations include the inventory of inprocess containers, observation of plant operations, nondestructive assay measurements to determine 235 U enrichment, as well as the examination of Material Control and Accountability (MC and A) documents. During 2005, HEU Transparency Program personnel will conduct 24 Special Monitoring Visits (SMVs) to four Russian uranium processing plants, in addition to staffing a Transparency Monitoring Office (TMO) at one Russian site

Supply of low enriched (LEU) and highly enriched uranium (HEU) for research reactors

International Nuclear Information System (INIS)

Mueller, H.

1997-01-01

Enriched uranium for research reactors in the form of LEU /= low enriched uranium at 19.75% U-235) and HEU (= highly enriched uranium at 90 to 93% U-235) was and is - due to its high U-235 enrichment - a political fuel other than enriched uranium for power reactors. The sufficient availability of LEU and HEU is a vital question for research reactors, especially in Europe, in order to perform their peaceful research reactor programs. In the past the USA were in the Western hemisphere sole supplier of LEU and HEU. Today the USA have de facto stopped the supply of LEU and HEU, for HEU mainly due to political reasons. This paper deals, among others, with the present availability of LEU and HEU for European research reactors and touches the following topics: - historical US supplies, - influence of the RERTR-program, - characteristics of LEU and HEU, - military HEU enters the civil market, -what is the supply situation for LEU and HEU today? - outlook for safe supplies of LEU and HEU. (author)

2009 Annual Health Physics Report for the HEU Transparency Program

International Nuclear Information System (INIS)

Radev, R.

2010-01-01

During the 2009 calendar year, Lawrence Livermore National Laboratory (LLNL) provided health physics support for the Highly Enriched Uranium (HEU) Transparency Program for external and internal radiation protection. LLNL also provided technical expertise related to BDMS radioactive sources and Russian radiation safety regulatory compliance. For the calendar year 2009, there were 159 person-trips that required dose monitoring of the U.S. monitors. Of the 159 person-trips, 149 person-trips were SMVs and 10 person-trips were Transparency Monitoring Office (TMO) trips. There were 4 monitoring visits by TMO monitors to facilities other than UEIE and 10 to UEIE itself. LLNL's Hazard Control Department laboratories provided the dosimetry services for the HEU Transparency monitors. In 2009, the HEU Transparency activities in Russia were conducted in a radiologically safe manner for the HEU Transparency monitors in accordance with the expectations of the HEU Transparency staff, NNSA and DOE. The HEU Transparency Program now has over fifteen years of successful experience in developing and providing health and safety support in meeting its technical objectives.

Vault mobility depends in part on microtubules and vaults can be recruited to the nuclear envelope

International Nuclear Information System (INIS)

Zon, Arend van; Mossink, Marieke H.; Houtsmuller, Adriaan B.; Schoester, Martijn; Scheffer, George L.; Scheper, Rik J.; Sonneveld, Pieter; Wiemer, Erik A.C.

2006-01-01

Vaults are ribonucleoproteins that may function in intracellular transport processes. We investigated the intracellular distribution and dynamics of vaults in non-small cell lung cancer cells in which vaults are labeled with the green fluorescent protein. Immunofluorescence experiments showed that vaults are dispersed throughout the cytoplasm; a small fraction is found in close proximity to microtubules. Immunoprecipitation experiments corroborated these results showing co-precipitation of MVP and β-tubulin. Using quantitative fluorescence-recovery after photobleaching (FRAP), we demonstrated that vault mobility over longer distances in part depends on intact microtubules; vaults moving slower when microtubules are depolymerized by nocodazole. Biochemical fractionation indicated a small fraction of MVP associated with the nucleus, however, no GFP-tagged vaults could be observed inside the nucleus. We observed an accumulation of vaults at the nuclear envelope upon treatment of cells with the protein synthesis inhibitor cycloheximide. Analysis of nucleo-cytoplasmic transport using a fluorescent substrate containing a classical NLS and NES expressed in MVP +/+ and MVP -/- mouse embryonic fibroblasts indicated no differences in nuclear import/export kinetics, suggesting no role for vaults in these processes. We hypothesize that a subset of vaults moves directionally via microtubules, possibly towards the nucleus

PFP vault operations containers for Plutonium Handling and Storage Critical Characteristics

International Nuclear Information System (INIS)

BONADIE, E.P.

2000-01-01

This document specifies the critical characteristics for containers procured for Plutonium Finishing Plant's (PFP's) Vault Operations system as required by HNF-PRO-268 and HNF-PRO-1819. These are the minimum specifications that the equipment must meet in order to perform its safety function

The Experience of Storage and Shipment for Reprocessing of HEU Nuclear Fuel Irradiated in the IRT-M Research Reactor and Pamir-630 Mobile Reactor

Energy Technology Data Exchange (ETDEWEB)

Sikorin, S. N.; Polazau, S. A.; Luneu, A. N.; Hrigarovich, T. K. [Joint Institute for Power and Nuclear Research–Sosny of the National Academy of Sciences of Belarus, Minsk (Belarus)

2014-08-15

At the end of 2010 under the Global Threat Reduction Initiative (GTRI), the Joint Institute for Power and Nuclear Research–“Sosny” (JIPNR–Sosny) of the National Academy of Sciences of the Republic of Belarus repatriated HEU spent nuclear fuel to the Russian Federation. The spent nuclear fuel was from the decommissioned Pamir-630D mobile reactor and IRT-M research reactor. The paper discusses the Pamir-630D spent nuclear fuel; experience and problems of spent nuclear fuel storage; and various aspects of the shipment including legal framework, preparation activities and shipment logistics. The conceptual project of a new research reactor for Belarus is also presented.

An automated magnetic tape vault at CERN computer center

CERN Multimedia

Claudia Marcelloni

2008-01-01

The rapidly changing data processing landscape the underlying long-term storage technology remains the tried and tested magnetic tape. This robust and mature technology is used to store the complete LHC data set, from which a fraction of the data is copied to overlying disk caches for fast and widespread access. The handling of the magnetic tape cartridges is now fully automated, as they are racked in vaults where they are moved between the storage shelves and the tape drives by robotic arms.

2011 Annual Health Physics Report for the HEU transparency Program

International Nuclear Information System (INIS)

Radev, R.

2012-01-01

During the 2008 calendar year, Lawrence Livermore National Laboratory (LLNL) provided health physics support for the Highly Enriched Uranium (HEU) Transparency Program for external and internal radiation protection. They also provided technical expertise related to BDMS radioactive sources and Russian radiation safety regulatory compliance. For the calendar year 2008, there were 158 person-trips that required dose monitoring of the U.S. monitors. Of the 158 person-trips, 148 person-trips were SMVs and 10 person-trips were Transparency Monitoring Office (TMO) trips. There were 6 monitoring visits by TMO monitors to facilities other than UEIE and 8 to UEIE itself. There were three monitoring visits (source changes) that were back-to-back with a total of 24 monitors. LLNL's Hazard Control Department laboratories provided the dosimetry services for the HEU Transparency monitors. In 2008, the HEU Transparency activities in Russia were conducted in a radiologically safe manner for the HEU Transparency monitors in accordance with the expectations of the HEU Transparency staff, NNSA and DOE. The HEU Transparency now has thirteen years of successful experience in developing and providing health and safety support in meeting its technical objectives.

244-AR Vault Interim Stabilization Project Plan

International Nuclear Information System (INIS)

LANEY, T.

2000-01-01

The 244-AR Vault Facility, constructed between 1966 and 1968, was designed to provide lag storage and treatment for the Plutonium-Uranium Extraction Facility (PUREX) tank farm sludges. Tank farm personnel transferred the waste from the 244-AR Vault Facility to B Plant for recovery of cesium and strontium. B Plant personnel then transferred the treatment residuals back to the tank farms for storage of the sludge and liquids. The last process operations, which transferred waste supporting the cesium/strontium recovery mission, occurred in April 1978. After the final transfer in 1978, the 244-AR facility underwent a cleanout. However, 2,271 L (600 gal) of sludge were left in Tank 004AR from an earlier transfer from Tank 241-AX-104. When the cleanout was completed, the facility was placed in a standby status. The sludge had been transferred to Tank 004AR to support Pacific Northwest National Laboratory [PNNL] vitrification work. Documentation of waste transfers suggests that a portion of the sludge may have been moved from Tank 004AR to Tank 002AR in preparation for transfer back to the AX Tank Farm; however, quantities of the sludge that were moved to Tank 002AR from that transfer must be estimated

Seismic scoping evaluation of high level liquid waste tank vaults at the Idaho Chemical Processing Plant

International Nuclear Information System (INIS)

Hashimoto, P.S.; Uldrich, E.D.; McGee, W.D.

1991-01-01

A seismic scoping evaluation of buried vaults enclosing high level liquid waste storage tanks at the Idaho Chemical Processing Plant has been performed. The objective of this evaluation was to scope out which of the vaults could be demonstrated to be seismically adequate against the Safe Shutdown Earthquake (SSE). Using approximate analytical methods, earthquake experience data, and engineering judgement, this study determined that one vault configuration would be expected to meet ICPP seismic design criteria, one would not be considered seismically adequate against the SSE, and one could be shown to be seismically adequate against the SSE using nonlinear analysis

PRTR ion exchange vault column sampling

International Nuclear Information System (INIS)

Cornwell, B.C.

1995-01-01

This report documents ion exchange column sampling and Non Destructive Assay (NDA) results from activities in 1994, for the Plutonium Recycle Test Reactor (PRTR) ion exchange vault. The objective was to obtain sufficient information to prepare disposal documentation for the ion exchange columns found in the PRTR Ion exchange vault. This activity also allowed for the monitoring of the liquid level in the lower vault. The sampling activity contained five separate activities: (1) Sampling an ion exchange column and analyzing the ion exchange media for purpose of waste disposal; (2) Gamma and neutron NDA testing on ion exchange columns located in the upper vault; (3) Lower vault liquid level measurement; (4) Radiological survey of the upper vault; and (5) Secure the vault pending waste disposal

Re-evaluation of monitored retrievable storage concepts

International Nuclear Information System (INIS)

Fletcher, J.F.; Smith, R.I.

1989-04-01

In 1983, as a prelude to the monitored retrievable storage (MRS) facility conceptual design, the Pacific Northwest Laboratory (PNL) conducted an evaluation for the US Department of Energy (DOE) that examined alternative concepts for storing spent LWR fuel and high- level wastes from fuel reprocessing. The evaluation was made considering nine concepts for dry away-from-reactor storage. The nine concepts evaluated were: concrete storage cask, tunnel drywell, concrete cask-in-trench, open-cycle vault, metal casks (transportable and stationary), closed-cycle vault, field drywell, and tunnel-rack vault. The purpose and scope of the re-evaluation did not require a repetition of the expert-based examinations used earlier. Instead, it was based on more detailed technical review by a small group, focusing on changes that had occurred since the initial evaluation was made. Two additional storage concepts--the water pool and the horizontal modular storage vault (NUHOMS system)--were ranked along with the original nine. The original nine concepts and the added two conceptual designs were modified as appropriate for a scenario with storage capacity for 15,000 MTU of spent fuel. Costs, area requirements, and technical and historical data pertaining to MRS storage were updated for each concept

Re-evaluation of monitored retrievable storage concepts

Energy Technology Data Exchange (ETDEWEB)

Fletcher, J.F.; Smith, R.I.

1989-04-01

In 1983, as a prelude to the monitored retrievable storage (MRS) facility conceptual design, the Pacific Northwest Laboratory (PNL) conducted an evaluation for the US Department of Energy (DOE) that examined alternative concepts for storing spent LWR fuel and high- level wastes from fuel reprocessing. The evaluation was made considering nine concepts for dry away-from-reactor storage. The nine concepts evaluated were: concrete storage cask, tunnel drywell, concrete cask-in-trench, open-cycle vault, metal casks (transportable and stationary), closed-cycle vault, field drywell, and tunnel-rack vault. The purpose and scope of the re-evaluation did not require a repetition of the expert-based examinations used earlier. Instead, it was based on more detailed technical review by a small group, focusing on changes that had occurred since the initial evaluation was made. Two additional storage concepts--the water pool and the horizontal modular storage vault (NUHOMS system)--were ranked along with the original nine. The original nine concepts and the added two conceptual designs were modified as appropriate for a scenario with storage capacity for 15,000 MTU of spent fuel. Costs, area requirements, and technical and historical data pertaining to MRS storage were updated for each concept.

Dry storage technologies: keys to choosing among metal casks, concrete shielded steel canister modules and vaults

International Nuclear Information System (INIS)

Roland, V.; Solignac, Y.; Chiguer, M.; Guenon, Y.

2003-01-01

The current international trend towards expanding Spent Fuel Interim Dry Storage capabilities goes with an improvement of the performance of the proposed systems which have to accommodate Spent fuel Assemblies characterized by ever increasing burn-up, fissile isotopes contents, thermal releases, and total inventory. Due to heterogeneous worldwide reactor pools and specific local constraints the proposed solutions have also to cope with a wide fuel design variety. Moreover, the Spent fuel Assemblies stored temporarily for cooling may have to be transported either to reprocessing facilities or to interim storage facilities before direct disposal; it is the reason why the retrievability, including or not transportability of the proposed systems, is often specified by the Utilities for the design of their Storage systems and sometimes by law. This paper shows on examples developed within companies of AREVA Group the key parameters and elements that can direct toward the selection of a technology in a user specific context. Some of the constraints are ability to dry store at once a large number of spent fuel assemblies, readily available, on a given site. No urgent need for further move of the fuel is foreseen. Then clearly a Vault Type Storage system developed and implemented by SGN is an excellent solution: It combines passive safety with immediate large capacity, which allows quick amortization of fuel receiving equipment. In addition the versatile storage position can easily accept in the same facility different fuel types, and also intermediate and High Level Waste. This is the reason why a vault system is often a preferred solution for a long-term dry interim centralized storage, for a multiplicity of spent fuel. It can be also a choice solution when the ISFSI stands on a site that is dedicated permanently to many different nuclear activities.In most cases, the producers of spent fuel require a large capacity that is cumulated over many years, each reload at a

Dry storage of Magnox fuel

International Nuclear Information System (INIS)

1986-09-01

This work, commissioned by the CEGB, studies the feasibility of a combination of short-term pond storage and long-term dry storage of Magnox spent fuel as a cheaper alternative to reprocessing. Storage would be either at the reactor site or a central site. Two designs are considered, based on existing design work done by GEC-ESL and NNC; the capsule design developed by NNC and with storage in passive vaults for up to 100 yrs and the GEC-ESL tube design developed at Wylfa for the interim storage of LWR. For the long-term storage of Magnox spent fuel the GEC-ESL tubed vault all-dry storage method is recommended and specifications for this method are given. (U.K.)

Numerical analysis of a natural convection cooling system for radioactive canisters storage

Energy Technology Data Exchange (ETDEWEB)

Tsal, R.J.; Anwar, S.; Mercada, M.G. [Fluor Daniel Inc., Irvine, CA (United States)

1995-02-01

This paper describes the use of numerical analysis for studying natural convection cooling systems for long term storage of heat producing radioactive materials, including special nuclear materials and nuclear waste. The paper explains the major design philosophy, and shares the experiences of numerical modeling. The strategy of storing radioactive material is to immobilize nuclear high-level waste by a vitrification process, convertion it into borosilicate glass, and cast the glass into stainless steel canisters. These canisters are seal welded, decontaminated, inspected, and temporarily stored in an underground vault until they can be sent to a geologic repository for permanent storage. These canisters generate heat by nuclear decay of radioactive isotopes. The function of the storage facility ventilation system is to ensure that the glass centerline temperature does not exceed the glass transition temperature during storage and the vault concrete temperatures remain within the specified limits. A natural convection cooling system was proposed to meet these functions. The effectiveness of a natural convection cooling system is dependent on two major factors that affect air movement through the vault for cooling the canisters: (1) thermal buoyancy forces inside the vault which create a stack effect, and (2) external wind forces, that may assist or oppose airflow through the vault. Several numerical computer models were developed to analyze the thermal and hydraulic regimes in the storage vault. The Site Model is used to simulate the airflow around the building and to analyze different air inlet/outlet devices. The Airflow Model simulates the natural convection, thermal regime, and hydraulic resistance in the vault. The Vault Model, internal vault temperature stratification; and, finally, the Hot Area Model is used for modeling concrete temperatures within the vault.

CLEARANCE OF ARCHIVE TAPE VAULT (BUILDING 186)

CERN Multimedia

EP Division

1999-01-01

The archive tape vault is being cleared to make way for the LHC experiments silicon facility.Experiments were contacted in February to arrange for the extended storage or destruction of archived volumes. Any who have not yet done so are asked to contactR. Minchin/IT immediately.This is also an opportunity to clear unused tapes or cartridges from around the site and have them destroyed ecologically. Please contact ep.smi.helpdesk@cern.ch to arrange for collection.A.E. Ball/EP-SMI

Behavior of spent nuclear fuel and storage system components in dry interim storage.

Energy Technology Data Exchange (ETDEWEB)

Johnson, A.B. Jr.; Gilbert, E.R.; Guenther, R.J.

1982-08-01

Irradiated nuclear fuel has been handled under dry conditions since the early days of nuclear reactor operation, and use of dry storage facilities for extended management of irradiated fuel began in 1964. Irradiated fuel is currently being stored dry in four types of facilities: dry wells, vaults, silos, and metal casks. Essentially all types of irradiated nuclear fuel are currently stored under dry conditions. Gas-cooled reactor (GCR) and liquid metal fast breeder reactor (LMFBR) fuels are stored in vaults and dry wells. Certain types of fuel are being stored in licensed dry storage facilities: Magnox fuel in vaults in the United Kingdom and organic-cooled reactor (OCR) fuel in silos in Canada. Dry storage demonstrations are under way for Zircaloy-clad fuel from boiling water reactors BWR's, pressurized heavy-water reactors (PHWRs), and pressurized water reactors (PWRs) in all four types of dry storage facilities. The demonstrations and related hot cell and laboratory tests are directed toward expanding the data base and establishing a licensing basis for dry storage of water reactor fuel. This report reviews the scope of dry interim storage technology, the performance of fuel and facility materials, the status of programs in several countries to license dry storage of water reactor fuel, and the characteristics of water reactor fuel that relate to dry storage conditions.

Behavior of spent nuclear fuel and storage-system components in dry interim storage

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Gilbert, E.R.; Guenther, R.J.

1982-08-01

Irradiated nuclear fuel has been handled under dry conditions since the early days of nuclear reactor operation, and use of dry storage facilities for extended management of irradiated fuel began in 1964. Irradiated fuel is currently being stored dry in four types of facilities: dry wells, vaults, silos, and metal casks. Essentially all types of irradiated nuclear fuel are currently stored under dry conditions. Gas-cooled reactor (GCR) and liquid metal fast breeder reactor (LMFBR) fuels are stored in vaults and dry wells. Certain types of fuel are being stored in licensed dry storage facilities: Magnox fuel in vaults in the United Kingdom and organic-cooled reactor (OCR) fuel in silos in Canada. Dry storage demonstrations are under way for Zircaloy-clad fuel from boiling water reactors BWR's, pressurized heavy-water reactors (PHWRs), and pressurized water reactors (PWRs) in all four types of dry storage facilities. The demonstrations and related hot cell and laboratory tests are directed toward expanding the data base and establishing a licensing basis for dry storage of water reactor fuel. This report reviews the scope of dry interim storage technology, the performance of fuel and facility materials, the status of programs in several countries to license dry storage of water reactor fuel, and the characteristics of water reactor fuel that relate to dry storage conditions

Kinematic variables of table vault on artistic gymnastics

Directory of Open Access Journals (Sweden)

Sarah Maria Boldrini FERNANDES

2016-03-01

Full Text Available Abstract The table vault is an event of male and female Artistics Gymnastics. Although it can be performed in a variety of rotations and body positions in different phases, it can be separated in three groups: handspring, Yurchenko and Tsukahara. It is believed that kinematic variables of vault may vary according to group of vault or gymnast body position, but few studies compares the real differences among the three groups of vaults, comparing and describing the variables in different phases. Vault kinematic variables could be diversifying according to the approach or position of the vaulting, but little has been studied about the biomechanical differences, comparing and describing behaviours at different stages. The aim of this study was to organize critical, objective and to systematize the most relevant kinematic variables to performance on vaulting. A Meta analysis over the basis Pubmed, Sport Discus and Web of Science were performed about this issue. From the selected references, we described and analyzed the kinematics of the table vault. Vault can be characterized in seven phases of analysis. Most of the studies are descriptive, and some do not descript all phases. Differences among vault variables according to group vaults, technical level and gender were analysed only in recent studies. There still gaps of knowledge about kinematic variables of table vault, in order to provide comprehensive information about all possibilities of vaults in this gymnastic event. It is concluded that kinematic variables of table vault depends upon vault group and may be considered to the improvement of technical performance. More researches are needed to approach the coaching interface with biomechanics applicable knowledge.

The calculation of the MEU-HEU coupled core in the KUCA

International Nuclear Information System (INIS)

Hayashi, M.; Shiroya, S.; Kanda, K.; Shibata, T.

1984-01-01

The KUCA has a plan for critical experiments of the MEU-HEU coupled core in 1984. The neutronics calculation has been performed for the MEU-HEU coupled core in the KUCA. The GGC-4 and THERMOS were used to generate the four-group constants and the 2D-FEM-KUR, based on the finite-element method, was used for the diffusion calculation. The calculations with four-group constants agreed with experiments within 1.8% for the both single-cores with the MEU and the HEU. (author)

In-situ radiation measurements of the C1 and C2 waste storage tank vault

International Nuclear Information System (INIS)

Yong, L.K.; Womble, P.C.; Weems, L.D.

1996-09-01

In August of 1996, the Applied Radiation Measurements Department (ARMD) of the Waste Management and Remedial Action Division (WMRAD) at Oak Ridge National Laboratory (ORNL) was tasked with characterizing the radiation fields in the C 1 and C 2 Liquid Low Level Waste (LLLW) tank vault located at ORNL. These in-situ measurements were made to provide data for evaluating the potential radiological conditions for personnel working in or around the vault during future planned activities. This report describes the locations where measurements were made, the types of radiation detection instruments used, the methods employed, the problems encountered and resolved, and discusses the results obtained

Fuel cycle cost study with HEU and LEU fuels

International Nuclear Information System (INIS)

Matos, J.E.; Freese, K.E.

1984-01-01

Fuel cycle costs are compared for a range of 235 U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors

RERTR end-game: A win-win framework. Phasing out remaining global HEU commerce by conditionally and temporarily renewing U.S. exports of HEU

International Nuclear Information System (INIS)

Kuperman, Alan J.; Leventhal, Paul L.

1997-01-01

The RERTR program stands on the brink of fulfilling its historic mission. However, a series of missteps and misunderstandings have recently raised the risk that defeat will be snatched from the jaws of victory. Perhaps the most serious threat to the RERTR regime is posed by France's pending import of 625 kilograms of bomb-grade, highly enriched uranium (HEU) from Russia, intended primarily to fuel its high-flux research reactor at the Institute Laue-Langevin in Grenoble, as well as its Orphee research reactor. As the first export of HEU from Russia to a facility outside the former Soviet bloc, this precedential transaction would establish Russia as a new global supplier of bomb-grade uranium, potentially setting the stage for a rise in international HEU commerce, rather than its phase-out as envisioned under the RERTR program. Apparently, France turned to Russia for supply of the fuel because the United States was perceived as unable or unwilling to continue supplying such fuel in the wake of the U.S. Energy Policy Act of 1992, which, pursuant to its so-called Schumer Amendment, places sharp restrictions on HEU exports. Unexplained delays in Russia's shipment of this material to France provide a fortuitous window of opportunity in which efforts can and should be made by France and the United States to resolve present differences in a manner beneficial to each, as well as in the interest of global security. This paper proposes an arrangement under which the United States would renew exports of HEU to France, in exchange for pledges from France enabling the export to comply with the principles and objectives of the RERTR program as embodied in U.S. law. In so doing, the arrangement would obviate the need for Russian HEU export, thereby avoiding its dangerous precedent. By enabling high quality scientific research to continue, while simultaneously helping to fulfill the RERTR program's original goal, such an arrangement would truly be a 'win-win' solution. (author)

Super-light pearl-chain arch vaults

DEFF Research Database (Denmark)

Hertz, Kristian Dahl; Halding, Philip Skov

2014-01-01

Arch vaults are known as optimal and impressive structures, but due to their curved shapes they are extremely costly to produce in countries, where the cost of labor is high. By means of super-light pearl-chain technology arch vaults can be constructed from equal plane prefabricated elements, which...... that it is also applicable as basic elements for super-light pearl-chain vaults. Machines and software have been developed for automatic mass production of the elements, and the first factory has started production in Denmark in 2014 delivering SL-deck elements for a variety of building projects. This means...... is known from curved voluminous elements. The paper describes the first full-scale test of an arch vault made of SL-deck elements demonstrating the principle, documenting the load-bearing capacity, and solving a number of details needed in order to create vaults in practice. Considerations and plans...

Cranial vault thickness in primates: Homo erectus does not have uniquely thick vault bones.

Science.gov (United States)

Copes, Lynn E; Kimbel, William H

2016-01-01

Extremely thick cranial vaults have been noted as a diagnostic characteristic of Homo erectus since the first fossil of the species was identified, but relatively little work has been done on elucidating its etiology or variation across fossils, living humans, or extant non-human primates. Cranial vault thickness (CVT) is not a monolithic trait, and the responsiveness of its layers to environmental stimuli is unknown. We obtained measurements of cranial vault thickness in fossil hominins from the literature and supplemented those data with additional measurements taken on African fossil specimens. Total CVT and the thickness of the cortical and diploë layers individually were compared to measures of CVT in extant species measured from more than 500 CT scans of human and non-human primates. Frontal and parietal CVT in fossil primates was compared to a regression of CVT on cranial capacity calculated for extant species. Even after controlling for cranial capacity, African and Asian H. erectus do not have uniquely high frontal or parietal thickness residuals, either among hominins or extant primates. Extant primates with residual CVT thickness similar to or exceeding H. erectus (depending on the sex and bone analyzed) include Nycticebus coucang, Perodicticus potto, Alouatta caraya, Lophocebus albigena, Galago alleni, Mandrillus sphinx, and Propithecus diadema. However, the especially thick vaults of extant non-human primates that overlap with H. erectus values are composed primarily of cortical bone, while H. erectus and other hominins have diploë-dominated vault bones. Thus, the combination of thick vaults comprised of a thickened diploë layer may be a reliable autapomorphy for members of the genus Homo. Copyright © 2015 Elsevier Ltd. All rights reserved.

A simple method for rapidly processing HEU from weapons returns

Energy Technology Data Exchange (ETDEWEB)

McLean, W. II; Miller, P.E.

1994-01-01

A method based on the use of a high temperature fluidized bed for rapidly oxidizing, homogenizing and down-blending Highly Enriched Uranium (HEU) from dismantled nuclear weapons is presented. This technology directly addresses many of the most important issues that inhibit progress in international commerce in HEU; viz., transaction verification, materials accountability, transportation and environmental safety. The equipment used to carry out the oxidation and blending is simple, inexpensive and highly portable. Mobile facilities to be used for point-of-sale blending and analysis of the product material are presented along with a phased implementation plan that addresses the conversion of HEU derived from domestic weapons and related waste streams as well as material from possible foreign sources such as South Africa or the former Soviet Union.

Rupture loop annex ion exchange RLAIX vault deactivation

Energy Technology Data Exchange (ETDEWEB)

Ham, J.E.; Harris, D.L., Westinghouse Hanford

1996-08-01

This engineering report documents the deactivation, stabilization and final conditions of the Rupture Loop Annex Ion Exchange (RLAIX) Vault located northwest of the 309 Building`s Plutonium Recycle Test Reactor (PRTR). Twelve ion exchange columns, piping debris, and column liquid were removed from the vault, packaged and shipped for disposal. The vault walls and floor were decontaminated, and portions of the vault were painted to fix loose contamination. Process piping and drains were plugged, and the cover blocks and rain cover were installed. Upon closure,the vault was empty, stabilized, isolated.

PRTR ion exchange vault water removal

International Nuclear Information System (INIS)

Ham, J.E.

1995-11-01

This report documents the removal of radiologically contaminated water from the Plutonium Recycle Test Reactor (PRTR) ion exchange vault. Approximately 57,000 liters (15,000 gallons) of water had accumulated in the vault due to the absence of a rain cover. The water was removed and the vault inspected for signs of leakage. No evidence of leakage was found. The removal and disposal of the radiologically contaminated water decreased the risk of environmental contamination

Activity release during the dry storage of fuel assemblies

International Nuclear Information System (INIS)

Valentine, M.K.; Fettel, W.; Gunther, H.

1991-01-01

This paper reports that wet storage is the predominant storage method in the USA for spent fuel assemblies. Nevertheless, most utilities have stretched their storage capacities and several reactors will lose their full-core reserve in the 90's. A great variety of out-of-pool storage methods already exist, including the FUELSTOR vault-type dry storage concept. A FUELSTOR vault relies on double containment of the spent fuel (intact cladding as the primary containment and sealing of assemblies in canisters filled with an inert gas as the secondary containment) to reduce radiation levels at the outside wall of the vault to less than site boundary levels. Investigation of accident scenarios reveals that radiation release limits are only exceeded following complete failure of all canisters and simultaneous cladding breach for more than 40% of the rods (or for more than 1% of failed rods if massive fuel oxidation occurs following cladding failure). Such failures are considered highly improbable. Thus, it can be concluded that this type of dry storage is safe and individual canister monitoring is not required in the facility

Final safety-analysis report for the Fifth Calcined Solids Storage Facility

International Nuclear Information System (INIS)

1982-01-01

Radioactive aqueous wastes generated by the solvent extraction of uranium from expended fuels at ICPP will be calcined in the New Waste Calcining Facility (NWCF). The calcined solids are pneumatically transferred to stainless steel bins enclosed in concrete vaults for interim storage of up to 500 years. The Fifth Calcined Solids Storage Facility (CSSF) provides 1000 m 3 of storage and consists of seven annular stainless steel bins inside a reinforced concrete vault set on bedrock. Storage of calcined solids is essentially a passive operation with very little opportunity for release of radionuclides and with no potential for criticality. There will be no potential for fire or explosion. Shielding has been designed to assure that the radiation levels at the vault exterior surfaces will be limited to less than 0.5 mRem/h. A sump in the vault floor will collect any in-leakage that may occur. Any water that collects in the sump will be sampled then removed with the sump jet. There will be an extremely small chance of release of radioactive particulates into the atmosphere as a result of a bin leak. The Design Basis Accident (DBA) postulates the spill of solids from an eroded fill line into the vault coupled with a failure of the vault cooling air radiation monitor. For the DBA, the maximum calculated radiation dose to an exposed individual near the site boundary is less than 1.2 μRem to the bone and lung

Design of saltstone vaults

International Nuclear Information System (INIS)

Aiyar, G.S.; Hsiu, F.J.

1987-01-01

Radioactive waste from processed spent nuclear fuel at the Savannah River Plant, South Carolina, are stored in underground tanks. The wastes consist of sludge and supernate. Most of the radionuclides and some nonradioactive constituents are removed from the supernate. These along with the sludge are converted into a glass-crystallite matrix and cast into stainless steel cansisters for future disposal in a geological repository. The decontaminated salt solution is mixed with cement, fly ash, and a set-retarding agent, and the resulting grout is transferred to reinforced concrete vaults where it sets into a monolith termed saltstone. The vault is then capped with concrete. A total of 21 vaults measuring 600 x 100 x 27 ft are planned for disposal of the existing supernate plus any additional supernate generated up to the year 2000

A fuel cycle cost study with HEU and LEU fuels

International Nuclear Information System (INIS)

Matos, J.E.; Freese, K.E.

1985-01-01

Fuel cycle costs are compared for a range of 235 U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors. (author)

A fuel cycle cost study with HEU and LEU fuels

Energy Technology Data Exchange (ETDEWEB)

Matos, J E; Freese, K E [Argonne National Laboratory, Argonne, IL (United States)

1985-07-01

Fuel cycle costs are compared for a range of {sup 235}U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors. (author)

Report by the committee assessing fuel storage

International Nuclear Information System (INIS)

Morgan, W.W.

1977-11-01

Various concepts for interim storage of spent nuclear fuel have been considered. Preliminary design studies and cost estimates have been prepared for the following concepts: two with water cooling - prolonged pool storage at a generating station and pool storage at a central site - , three with air cooling at a central site - ''canister'', ''convection vault'', and ''conduction vault'' - and one underground storage scheme in rock salt. Costs (1972 dollars) were estimated including transportation and a perpetual care fund for maintenance and periodical renewal of the storage facility. Part 2 provides details of the concepts and costing methods. All concepts gave moderate costs providing a contribution of about 0.1 m$/kWh to the total unit energy cost. Advantages and disadvantages of the respective schemes are compared. (author)

Implementation of the United States/Russian HEU Agreement: Current Status and Prospects

International Nuclear Information System (INIS)

Rutkowski, E

2003-01-01

During Calendar Year (CY) 2002, the Russian Federation (R.F.) delivered low enriched uranium (LEU) from the conversion and processing of 30 metric tons (MT) of weapons-grade (90% 235 U assay) uranium. Through July 2003, the Highly Enriched Uranium (HEU) Transparency Implementation Program (TIP) will have monitored the conversion of over 190 MT HEU into LEU. This total represents about 38 percent of the projected 500 MT HEU scheduled to be blended down through the year 2013 and is equivalent to the destruction of 7,600 nuclear devices. The National Nuclear Security Administration's (NNSA) HEU-TIP monitors the processing of this HEU at four Russian uranium-processing plants. During CY 2002, United States (U.S.) personnel monitored this process for a total of 194 monitor-weeks by staffing a Transparency Monitoring Office (TMO) located in Novouralsk, and through a series of five-day Special Monitoring Visits (SMV) to the four plants. U.S. monitor observations include the inventory of in-process containers, the observation of operations and non-destructive assay measurements (NDA) to determine 235 U enrichment, as well as the examination and validation of Russian Material Control and Accountability (MC and A) documents. In addition, the U.S. designed Blend Down Monitoring System (BDMS) installed at the Ural Electrochemical Integrated Plant (UEIP) in January 1999 monitored all HEU blended at that facility, which is about 50 percent of the HEU blended into LEU during CY 2002. Recently we installed a BDMS at the Electrochemical Plant (ECP) in Zelenogorsk and plans are underway to install a BDMS at the Siberian Chemical Enterprise (SChE) in Seversk in late 2004. On a very positive note, interpersonal interactions between U.S. and Russian technical experts continues to expand and have proven to be an important element of the transparency regime. On the tenth anniversary of the HEU Purchase Agreement, the Ministry of the R.F. for Atomic Energy (Minatom) also saluted the

A re-evaluation of physical protection standards for irradiated HEU fuel

International Nuclear Information System (INIS)

Lyman, Edwin; Kuperman, Alan

2002-01-01

In the post-September 11 era, it is essential to reconsider all the assumptions upon which the physical protection systems of the past were based and determine whether these assumptions are still appropriate in light of the current terrorist threat. For instance, the U.S. Nuclear Regulatory Commission definition of a 'formula quantity' of special nuclear material is derived from the belief that a terrorist plot to carry out multiple coordinated attacks on different facilities with the goal of acquiring enough SNM for a nuclear weapon is incredible. This assumption has clearly been proven wrong by the September 11 attacks. Another standard that needs to be revisited is the 'self-protection' threshold that determines whether or not an item containing SNM is considered to be 'irradiated' for physical protection purposes. The current value of this threshold, 1 Sv/hr unshielded at 1 meter, is of questionable value as a deterrent to determined terrorists who would be willing to sustain long-term injury as long as they could accomplish their near-term goals. A more credible threshold would be set at a level that would have a high likelihood of disabling the perpetrators before they could complete their mission. Most irradiated nonpower reactor fuels would be unable to meet such a standard. This raises serious questions about the adequacy of the level of physical protection applied today to the large inventories of irradiated HEU fuels now scattered in storage sites around the world. The absence of a coherent global policy for dealing with these materials has created a situation rife with vulnerabilities that terrorists could exploit. The international community, now seized with concern about unused stockpiles of unirradiated HEU fuels around the world, also needs to appreciate the dangers posed by lightly irradiated spent fuels as well. A U.S. proposal to import Russian HEU for supplying U.S. nonpower reactors will only prolong this situation This paper will review policy

Device for providing a leak-tight penetration for electric cables through a reactor vault roof

International Nuclear Information System (INIS)

Eyral, M.; Mahe, A.

1979-01-01

The device for providing a cable penetration through the vault roof of a liquid sodium cooled fast reactor comprises a vertical tube closed at the top end by a flange-plate. Electric cables connected to measuring and detecting instruments are passed through the flange-plate which is joined to the reactor vault roof in leak-tight manner and enclosed within a removable hood. At least one horizontal plate is mounted within the vertical tube and provided with orifices for the leak-tight passage of the cables. Cable storage reels are placed within the tube and can be locked in position or released by controlled mechanical means

Detection of illicit HEU production in gaseous centrifuge enrichment plants using neutron counting techniques on product cylinders

International Nuclear Information System (INIS)

Freeman, Corey R.; Geist, William H.

2010-01-01

Innovative and novel safeguards approaches are needed for nuclear energy to meet global energy needs without the threat of nuclear weapons proliferation. Part of these efforts will include creating verification techniques that can monitor uranium enrichment facilities for illicit production of highly-enriched uranium (HEU). Passive nondestructive assay (NDA) techniques will be critical in preventing illicit HEU production because NDA offers the possibility of continuous and unattended monitoring capabilities with limited impact on facility operations. Gaseous centrifuge enrichment plants (GCEP) are commonly used to produce low-enriched uranium (LEU) for reactor fuel. In a GCEP, gaseous UF 6 spins at high velocities in centrifuges to separate the molecules containing 238 U from those containing the lighter 235 U. Unfortunately, the process for creating LEU is inherently the same as HEU, creating a proliferation concern. Insuring that GCEPs are producing declared enrichments poses many difficult challenges. In a GCEP, large cascade halls operating thousands of centrifuges work together to enrich the uranium which makes effective monitoring of the cascade hall economically prohibitive and invasive to plant operations. However, the enriched uranium exiting the cascade hall fills product cylinders where the UF 6 gas sublimes and condenses for easier storage and transportation. These product cylinders hold large quantities of enriched uranium, offering a strong signal for NDA measurement. Neutrons have a large penetrability through materials making their use advantageous compared to gamma techniques where the signal is easily attenuated. One proposed technique for detecting HEU production in a GCEP is using neutron coincidence counting at the product cylinder take off stations. This paper discusses findings from Monte Carlo N-Particle eXtended (MCNPX) code simulations that examine the feasibility of such a detector.

Characterization of MVP and VPARP assembly into vault ribonucleoprotein complexes.

Science.gov (United States)

Zheng, Chun-Lei; Sumizawa, Tomoyuki; Che, Xiao-Fang; Tsuyama, Shinichiro; Furukawa, Tatsuhiko; Haraguchi, Misako; Gao, Hui; Gotanda, Takenari; Jueng, Hei-Cheul; Murata, Fusayoshi; Akiyama, Shin-Ichi

2005-01-07

Vaults are barrel-shaped cytoplasmic ribonucleoprotein particles composed of three proteins: the major vault protein (MVP), the vault poly(ADP-ribose)polymerase (VPARP), and the telomerase-associated protein 1, together with one or more small untranslated RNAs. To date, little is known about the process of vault assembly or about the stability of vault components. In this study, we analyzed the biosynthesis of MVP and VPARP, and their half-lives within the vault particle in human ACHN renal carcinoma cells. Using an immunoprecipitation assay, we found that it took more than 4h for newly synthesized MVPs to be incorporated into vault particles but that biosynthesized VPARPs were completely incorporated into vaults within 1.5h. Once incorporated into the vault complex, both MVP and VPARP were very stable. Expression of human MVP alone in Escherichia coli resulted in the formation of particles that had a distinct vault morphology. The C-terminal region of VPARP that lacks poly(ADP-ribose)polymerase activity co-sedimented with MVP particles. This suggests that the activity of VPARP is not essential for interaction with MVP-self-assembled vault-like particles. In conclusion, our findings provide an insight into potential mechanisms of physiological vault assembly.

Construction gets underway on Hungary's Modern Vault Dry Store

International Nuclear Information System (INIS)

Anon.

1995-01-01

A construction licence has recently been granted for a Modular Vault Dry Store (MVDS) for spent fuel at the Paks reactor site in Hungary. The store will be used for medium term (50 years) storage of spent fuel from four VVER-440 reactors. It is anticipated that storage capacity for 1350 fuel assemblies will be available by 1996. Two further construction phases will take the capacity to 4950, covering the first ten years of reactor operation. The design provides for further extension to accommodate a total 15000 assemblies, corresponding to 30 years of reactor operation. The MVDS has developed out of the first application of dry store technology to spent Magnox reactor fuel at the Wylfa power station in the United Kingdom 25 years ago. (UK)

A level-playing field for medical isotope production - How to phase-out reliance on HEU

International Nuclear Information System (INIS)

Kuperman, A.J.

1999-01-01

Two decades ago, civilian commerce in highly enriched uranium (HEU) for use as targets in the production of medical isotopes was considered a relatively minor security concern for three reasons. First, the number of producers was small. Second, the amount of HEU involved was small. Third, the amount of HEU was dwarfed by the quantities of HEU in civilian commerce as fuel for nuclear research and test reactors. Now, however, all three variables have changed. First, as the use of medical isotopes has expanded rapidly, production programs are proliferating. Second, as the result of such new producers and the expansion of existing production facilities, the amounts of HEU involved are growing. Third, as the RERTR program has facilitated the phase-out of HEU as fuel in most research and test reactors, the quantities of HEU for isotope production have come to represent a significant percentage of global commerce in this weapons-usable material. Medical isotope producers in several states are cooperating with the RERTR program to convert to low-enriched uranium (LEU) targets within the next few years, and one already relies on LEU for isotope production. However, the three biggest isotope producers - in Canada and the European Union - continue to rely on HEU, creating a double-standard that endangers the goal of the RERTR program. Each of these three producers has expressed economic concerns about being put at a competitive disadvantage if it alone converts. This paper proposes forging a firmer international consensus that all present and future isotope producers should convert to LEU, and calls for codifying such a commitment in a statement of intent to be prepared by producers over the next year. With such a level playing field, no producer would need fear being put at a competitive disadvantage by conversion, or being stigmatized by pressure groups for continued reliance on HEU. The phase-out of all HEU commerce for isotope production could be achieved within about

CSER 94-014: Storage of metal-fuel loaded EBR-II casks in concrete vault on PFP grounds

International Nuclear Information System (INIS)

Hess, A.L.

1994-01-01

A criticality safety evaluation is presented to permit EBR-2 spent fuel casks loaded with metallic fuel rods to be stored in an 8-ft diameter, cylindrical concrete vault inside the PFP security perimeter. The specific transfer of three casks with Pu alloy fuel from the Los Alamos Molten Plutonium Reactor Experiment from the burial grounds to the vault is thus covered. Up to seven casks may be emplaced in the casing with 30 inches center to center spacing. Criticality safety is assured by definitive packaging rules which keep the fissile medium dry and at a low effective volumetric density

Radiological consequence analysis with HEU and LEU fuels

Energy Technology Data Exchange (ETDEWEB)

Woodruff, W.L.; Warinner, D.K.; Matos, J.E.

1984-01-01

A model for estimating the radiological consequences from a hypothetical accident in HEU and LEU fueled research and test reactors is presented. Simple hand calculations based on fission product yield table inventories and non-site specific dispersion data may be adequate in many cases. However, more detailed inventories and site specific data on meteorological conditions and release rates and heights can result in substantial reductions in the dose estimates. LEU fuel gives essentially the same doses as HEU fuel. The plutonium buildup in the LEU fuel does not significantly increase the radiological consequences. The dose to the thyroid is the limiting dose. 10 references, 3 figures, 7 tables.

Radiological consequence analysis with HEU and LEU fuels

International Nuclear Information System (INIS)

Woodruff, W.L.; Warinner, D.K.; Matos, J.E.

1984-01-01

A model for estimating the radiological consequences from a hypothetical accident in HEU and LEU fueled research and test reactors is presented. Simple hand calculations based on fission product yield table inventories and non-site specific dispersion data may be adequate in many cases. However, more detailed inventories and site specific data on meteorological conditions and release rates and heights can result in substantial reductions in the dose estimates. LEU fuel gives essentially the same doses as HEU fuel. The plutonium buildup in the LEU fuel does not significantly increase the radiological consequences. The dose to the thyroid is the limiting dose. 10 references, 3 figures, 7 tables

Closure Report for Corrective Action Unit 135: Areas 25 Underground Storage Tanks, Nevada Test Site, Nevada

Energy Technology Data Exchange (ETDEWEB)

D. H. Cox

2001-06-01

Corrective Action Unit (CAU) 135, Area 25 Underground Storage Tanks, was closed in accordance with the approved Corrective Action Plan (DOE/NV, 2000). CAU 135 consists of three Corrective Action Sites (CAS). Two of these CAS's were identified in the Corrective Action Investigation Data Quality Objective meeting as being improperly identified as underground storage tanks. CAS 25-02-03 identified as the Deluge Valve Pit was actually an underground electrical vault and CAS 25-02-10 identified as an Underground Storage Tank was actually a former above ground storage tank filled with demineralized water. Both of these CAS's are recommended for a no further action closure. CAS 25-02-01 the Underground Storage Tanks commonly referred to as the Engine Maintenance Assembly and Disassembly Waste Holdup Tanks and Vault was closed by decontaminating the vault structure and conducting a radiological verification survey to document compliance with the Nevada Test Site unrestricted use release criteria. The Area 25 Underground Storage Tanks, (CAS 25-02-01), referred to as the Engine Maintenance, Assembly, and Disassembly (E-MAD) Waste Holdup Tanks and Vault, were used to receive liquid waste from all of the radioactive and cell service area drains at the E-MAD Facility. Based on the results of the Corrective Action Investigation conducted in June 1999, discussed in ''The Corrective Action Investigation Plan for Corrective Action Unit 135: Area 25 Underground Storage Tanks, Nevada Test Site, Nevada'' (DOE/NV, 199a), one sample from the radiological survey of the concrete vault interior exceeded radionuclide preliminary action levels. The analytes from the sediment samples exceeded the preliminary action levels for polychlorinated biphenyls, Resource Conservation and Recovery Act metals, total petroleum hydrocarbons as diesel-range organics, and radionuclides. The CAU 135 closure activities consisted of scabbling radiological ''hot spots

Structure of the vault, a ubiquitous celular component.

Science.gov (United States)

Kong, L B; Siva, A C; Rome, L H; Stewart, P L

1999-04-15

The vault is a ubiquitous and highly conserved ribonucleoprotein particle of approximately 13 MDa. This particle has been shown to be upregulated in certain multidrug-resistant cancer cell lines and to share a protein component with the telomerase complex. Determination of the structure of the vault was undertaken to provide a first step towards understanding the role of this cellular component in normal metabolism and perhaps to shed some light on its role in mediating drug resistance. Over 1300 particle images were combined to calculate an approximately 31 A resolution structure of the vault. Rotational power spectra did not yield a clear symmetry peak, either because of the thin, smooth walls or inherent flexibility of the vault. Although cyclic eightfold (C8) symmetry was imposed, the resulting reconstruction may be partially cylindrically averaged about the eightfold axis. Our results reveal the vault to be a hollow, barrel-like structure with two protruding caps and an invaginated waist. Although the normal cellular function of the vault is as yet undetermined, the structure of the vault is consistent with either a role in subcellular transport, as previously suggested, or in sequestering macromolecular assemblies.

Kinematic variables of table vault on artistic gymnastics

OpenAIRE

FERNANDES,Sarah Maria Boldrini; CARRARA,Paulo; SERRÃO,Júlio Cerca; AMADIO,Alberto Carlos; MOCHIZUKI,Luis

2016-01-01

Abstract The table vault is an event of male and female Artistics Gymnastics. Although it can be performed in a variety of rotations and body positions in different phases, it can be separated in three groups: handspring, Yurchenko and Tsukahara. It is believed that kinematic variables of vault may vary according to group of vault or gymnast body position, but few studies compares the real differences among the three groups of vaults, comparing and describing the variables in different phases...

MVP and vaults: a role in the radiation response

International Nuclear Information System (INIS)

Lara, Pedro C; Pruschy, Martin; Zimmermann, Martina; Henríquez-Hernández, Luis Alberto

2011-01-01

Vaults are evolutionary highly conserved ribonucleoproteins particles with a hollow barrel-like structure. The main component of vaults represents the 110 kDa major vault protein (MVP), whereas two minor vaults proteins comprise the 193 kDa vault poly(ADP-ribose) polymerase (vPARP) and the 240 kDa telomerase-associated protein-1 (TEP-1). Additionally, at least one small and untranslated RNA is found as a constitutive component. MVP seems to play an important role in the development of multidrug resistance. This particle has also been implicated in the regulation of several cellular processes including transport mechanisms, signal transmission and immune responses. Vaults are considered a prognostic marker for different cancer types. The level of MVP expression predicts the clinical outcome after chemotherapy in different tumour types. Recently, new roles have been assigned to MVP and vaults including the association with the insulin-like growth factor-1, hypoxia-inducible factor-1alpha, and the two major DNA double-strand break repair machineries: non-homologous endjoining and homologous recombination. Furthermore, MVP has been proposed as a useful prognostic factor associated with radiotherapy resistance. Here, we review these novel actions of vaults and discuss a putative role of MVP and vaults in the response to radiotherapy

MVP and vaults: a role in the radiation response

Directory of Open Access Journals (Sweden)

Zimmermann Martina

2011-10-01

Full Text Available Abstract Vaults are evolutionary highly conserved ribonucleoproteins particles with a hollow barrel-like structure. The main component of vaults represents the 110 kDa major vault protein (MVP, whereas two minor vaults proteins comprise the 193 kDa vault poly(ADP-ribose polymerase (vPARP and the 240 kDa telomerase-associated protein-1 (TEP-1. Additionally, at least one small and untranslated RNA is found as a constitutive component. MVP seems to play an important role in the development of multidrug resistance. This particle has also been implicated in the regulation of several cellular processes including transport mechanisms, signal transmission and immune responses. Vaults are considered a prognostic marker for different cancer types. The level of MVP expression predicts the clinical outcome after chemotherapy in different tumour types. Recently, new roles have been assigned to MVP and vaults including the association with the insulin-like growth factor-1, hypoxia-inducible factor-1alpha, and the two major DNA double-strand break repair machineries: non-homologous endjoining and homologous recombination. Furthermore, MVP has been proposed as a useful prognostic factor associated with radiotherapy resistance. Here, we review these novel actions of vaults and discuss a putative role of MVP and vaults in the response to radiotherapy.

MVP and vaults: a role in the radiation response.

Science.gov (United States)

Lara, Pedro C; Pruschy, Martin; Zimmermann, Martina; Henríquez-Hernández, Luis Alberto

2011-10-31

Vaults are evolutionary highly conserved ribonucleoproteins particles with a hollow barrel-like structure. The main component of vaults represents the 110 kDa major vault protein (MVP), whereas two minor vaults proteins comprise the 193 kDa vault poly(ADP-ribose) polymerase (vPARP) and the 240 kDa telomerase-associated protein-1 (TEP-1). Additionally, at least one small and untranslated RNA is found as a constitutive component. MVP seems to play an important role in the development of multidrug resistance. This particle has also been implicated in the regulation of several cellular processes including transport mechanisms, signal transmission and immune responses. Vaults are considered a prognostic marker for different cancer types. The level of MVP expression predicts the clinical outcome after chemotherapy in different tumour types. Recently, new roles have been assigned to MVP and vaults including the association with the insulin-like growth factor-1, hypoxia-inducible factor-1alpha, and the two major DNA double-strand break repair machineries: non-homologous endjoining and homologous recombination. Furthermore, MVP has been proposed as a useful prognostic factor associated with radiotherapy resistance. Here, we review these novel actions of vaults and discuss a putative role of MVP and vaults in the response to radiotherapy.

A Continuous Automated Vault Inventory System (CAVIS) for accountability monitoring of stored nuclear materials

International Nuclear Information System (INIS)

Pickett, C.A.; Barham, M.A.; Gafford, T.A.; Hutchinson, D.P.; Jordan, J.K.; Maxey, L.C.; Moran, B.W.; Muhs, J.; Nodine, R.; Simpson, M.L.

1994-01-01

Nearly all facilities that store hazardous (radioactive or non-radioactive) materials must comply with prevailing federal, state, and local laws. These laws usually have components that require periodic physical inspections to insure that all materials remain safely and securely stored. The inspections are generally labor intensive, slow, put personnel at risk, and only find anomalies after they have occurred. The system described in this paper was developed for monitoring stored nuclear materials resulting from weapons dismantlement, but its applications extend to any storage facility that meets the above criteria. The traditional special nuclear material (SNM) accountability programs, that are currently used within most of the Department of Energy (DOE) complex, require the physical entry of highly trained personnel into SNM storage vaults. This imposes the need for additional security measures, which typically mandate that extra security personnel be present while SNM inventories are performed. These requirements increase labor costs and put additional personnel at risk to radiation exposure. In some cases, individuals have received radiation exposure equivalent to the annual maximum during just one inventory verification. With increasing overhead costs, the current system is rapidly becoming too expensive to operate, the need for an automated method of inventory verification is evident. The Continuous Automated Vault Inventory System (CAVIS) described in this paper was designed and prototyped as a low cost, highly reliable, and user friendly system that is capable of providing, real-time weight, gamma. and neutron energy confirmation from each item stored in a SNM vault. This paper describes the sensor technologies, the CAVIS prototype system (built at Y- 12 for highly enriched uranium storage), the technical requirements that must be achieved to assure successful implementation, and descriptions of sensor technologies needed for a plutonium facility

A Continuous Automated Vault Inventory System (CAVIS) for accountability monitoring of stored nuclear materials

Energy Technology Data Exchange (ETDEWEB)

Pickett, C.A.; Barham, M.A.; Gafford, T.A.; Hutchinson, D.P.; Jordan, J.K.; Maxey, L.C.; Moran, B.W.; Muhs, J.; Nodine, R.; Simpson, M.L. [and others

1994-12-08

Nearly all facilities that store hazardous (radioactive or non-radioactive) materials must comply with prevailing federal, state, and local laws. These laws usually have components that require periodic physical inspections to insure that all materials remain safely and securely stored. The inspections are generally labor intensive, slow, put personnel at risk, and only find anomalies after they have occurred. The system described in this paper was developed for monitoring stored nuclear materials resulting from weapons dismantlement, but its applications extend to any storage facility that meets the above criteria. The traditional special nuclear material (SNM) accountability programs, that are currently used within most of the Department of Energy (DOE) complex, require the physical entry of highly trained personnel into SNM storage vaults. This imposes the need for additional security measures, which typically mandate that extra security personnel be present while SNM inventories are performed. These requirements increase labor costs and put additional personnel at risk to radiation exposure. In some cases, individuals have received radiation exposure equivalent to the annual maximum during just one inventory verification. With increasing overhead costs, the current system is rapidly becoming too expensive to operate, the need for an automated method of inventory verification is evident. The Continuous Automated Vault Inventory System (CAVIS) described in this paper was designed and prototyped as a low cost, highly reliable, and user friendly system that is capable of providing, real-time weight, gamma. and neutron energy confirmation from each item stored in a SNM vault. This paper describes the sensor technologies, the CAVIS prototype system (built at Y- 12 for highly enriched uranium storage), the technical requirements that must be achieved to assure successful implementation, and descriptions of sensor technologies needed for a plutonium facility.

31 CFR 540.305 - HEU Agreements.

Science.gov (United States)

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false HEU Agreements. 540.305 Section 540.305 Money and Finance: Treasury Regulations Relating to Money and Finance (Continued) OFFICE OF... Federation for Atomic Energy Concerning the Transfer of Source Material to the Russian Federation signed at...

Grouting of nuclear waste vault shafts

International Nuclear Information System (INIS)

Gyenge, M.

1980-01-01

A nuclear waste vault must be designed and built to ensure adequate isolation of the nuclear wastes from human contact. Consequently, after a vault has been fully loaded it must be adequately sealed off to prevent radionuclide migration which may be provided by circulating ground water. Of particular concern in vault sealing are the physical and chemical properties of the sealing materials its long-term durability and stability and the techniques used for its emplacement. Present grouting technology and grout material are reviewed in terms of the particular needs of shaft grouting. Areas requiring research and development are indicated

HEU to LEU conversion experience at the UMass-Lowell research reactor

International Nuclear Information System (INIS)

White, John R.; Bobek, Leo M.

2005-01-01

The UMass-Lowell Research Reactor (UMLRR) operated safely with high-enriched uranium (HEU) fuel for over 25 years. Having reached the end of core lifetime and due to proliferation concerns, the reactor was recently converted to low-enriched uranium silicide (LEU) fuel. The actual process for converting the UMLRR from HEU to LEU fuel covered a period of over 15 years. The conversion effort - from the initial conceptual design studies in the late 1980s to the final offsite shipment of the spent HEU fuel in August 2004 - was a unique experience for the faculty and staff of a small university research reactor. This paper gives a historical view of the process and it highlights several key milestones along the road to successful completion of this project. (author)

A radiological consequence analysis with HEU and LEU fuels

International Nuclear Information System (INIS)

Woodruff, W.L.; Warinner, D.K.; Matos, J.E.

1985-01-01

A model for estimating the radiological consequences from a hypothetical accident in HEU and LEU fueled research and test reactors is presented. Simple hand calculations based on fission product yield table inventories and nonsite specific dispersion data may be adequate in many cases. However, more detailed inventories and site specific data on meteorological conditions and release rates and heights can result in substantial reductions in the dose estimates. LEU fuel gives essentially the same doses as HEU fuel. The plutonium buildup in the LEU fuel does not significantly increase the radiological consequences. The dose to the thyroid is the limiting dose. (author)

Filigree Vaults

DEFF Research Database (Denmark)

2017-01-01

Filigree Vaults takes its point of departure in the use of ceramics in architecture.Today ceramic is in buildings mostly used as uniform modular elements for surface application - bricks and tiles. This project discusses and exemplifies new possibilities for ceramics in architecture, which emerge...

2011 Progress Report on HEU Minimization Activities in Argentina

Energy Technology Data Exchange (ETDEWEB)

Bonini, A.; Cristini, P.; Lio, L. De; Dell' Occhio, L.; Gil, D.; Gonzalez, A.G.; Gonzalez, R.; Varela, C. Komar; Lopez, M.; Novara, O.; Taboada, H. [Comision Nacional de Energia Atomica, Av. Del Libertador 8250 (1429) Buenos Aires (Argentina)

2011-07-01

After the core conversion of the RA-6 reactor finished in March 2008, an extension of the original CNEA-NNSA DoE contract was signed to enhance the final national HEU inventories minimization. Before this process, CNEA reserved a small inventory of HEU for R and D uses in fission chambers, neutronic probes and standards. This minimization comprises that all fresh and irradiated HEU remnant inventories coming from fuels and Mo99 irradiation targets fabrication and irradiated HEU-oxides retained in production filters and solutions will be recovered, down-blended into LEU and purified or dispose as waste whenever its recovery would not be advisable due to cost-benefit consideration. CNEA has a R and D program to develop the fabrication technology of both dispersed U-Mo (Al-Si matrix and Al cladding) and monolithic (Zry-4 cladding) miniplates to support the qualification activities of the RERTR program. Some monolithic 58% enrichment and LEU 8%Mo and U10%Mo miniplates and plates were and are being delivered to INL-DoE to be irradiated in the ATR reactor core. CNEA, a worldwide leader on LEU technology for fission radioisotope production is providing Brazil with 1/3 of the national requirements on Mo99 by weekly deliveries. Australia has started the fission radioisotope production through several batches by week, based on CNEA's LEU technology provided by INVAP SE. CNEA is also committed to improve the diffusion of LEU target and radiochemical technology for radioisotope production and target and process optimization. Future plans include: 1. Plans to recover and purify the LEU based inventories in Mo99 production filters, once the HEU to LEU campaign is over. 2. Fabrication and delivering to INL to be irradiated in the ATR core of U-8%Mo and U-10%Mo monolithic miniplates and development and fabrication of LEU very high density monolithic and dispersed U-Mo fuel plates with Zr cladding for the FUTURE-MONO experiment in the frame of the RERTR program. 3

Overview of Russian HEU transparency issues

International Nuclear Information System (INIS)

Kempf, C.R.; Bieniawski, A.

1993-01-01

The U.S. has signed an agreement with the Russian Federation for the purchase of 500 metric tons of highly-enriched uranium (HEU) taken from dismantled nuclear weapons. The HEU will be blended down to low-enriched uranium and will be transported to the U.S. to be used by fuel fabricators to make fuel for commercial nuclear power plants. Both the U.S. and Russia have been preparing to institute transparency measures to provide assurance that nonproliferation and arms control objectives specified in the agreement are met. This paper provides background information on the original agreement and on subsequent negotiations with the Russians, as well as discussion of technical aspects of developing transparency measures suited to the facilities and processes which are expected to be involved. Transparency has been defined as those agreed-upon measures which build confidence that arms control and non-proliferation objectives shared by the parties are met. Transparency is a departure from exhaustive, detailed arms control verification regimes of past agreements, which were based on a presumption of detecting transgressions as opposed to confirming compliance

Evaluation of concepts for monitored retrievable storage of spent nuclear fuel and high-level radioactive waste

International Nuclear Information System (INIS)

Triplett, M.B.; Smith, R.I.

1984-04-01

The primary mission selected by DOE for the monitored retrieval storage (MRS) system is to provide an alternative means of storage in the event that the repository program is delayed. The MRS concepts considered were the eight concepts included in the MRS Research and Development Report to Congress (DOE 1983). These concepts are: metal cask (stationary and transportable); concrete cask (sealed storage cask); concrete cask-in-trench; field drywell; tunnel drywell; open cycle vault; closed cycle vault; and tunnel rack vault. Conceptual design analyses were performed for the candidate concepts using a common set of design requirements specified in consideration of the MRS mission

Comparison of the FRM-II HEU design with an alternative LEU design

International Nuclear Information System (INIS)

Mo, S.C.; Hanan, N.A.; Matos, J.E.

2004-01-01

The FRM-II reactor design of the Technical University of Munich has a compact core that utilizes fuel plates containing highly-enriched uranium (HEU, 93%). This paper presents an alternative core design utilizing low-enriched uranium (LEU, 3 that provides nearly the same neutron flux for experiments as the HEU design, but has a less favourable fuel cycle economy. If an LEU fuel with a uranium density of 6.0 - 6.5 g/cm 3 . were developed, the alternative design would provide the same neutron flux and use the same number of cores per year as the HEU design. The results of this study show that there are attractive possibilities for using LEU fuel instead of HEU fuel in the FRM-II. Further optimization of the LEU design and near-term availability of LEU fuel with a uranium density greater than 4.8 g/cm 3 would enhance the performance of the LEU core. The REKIR Program is ready to exchange information with the Technical University of Munich to resolve any differences that may exist and to identify design modifications that would optimize reactor performance utilizing LEU fuel. (author)

Sea urchin vault structure, composition, and differential localization during development

Directory of Open Access Journals (Sweden)

Dickey-Sims Carrie

2005-02-01

Full Text Available Abstract Background Vaults are intriguing ribonucleoprotein assemblies with an unknown function that are conserved among higher eukaryotes. The Pacific coast sea urchin, Strongylocentrotus purpuratus, is an invertebrate model organism that is evolutionarily closer to humans than Drosophila and C. elegans, neither of which possesses vaults. Here we compare the structures of sea urchin and mammalian vaults and analyze the subcellular distribution of vaults during sea urchin embryogenesis. Results The sequence of the sea urchin major vault protein (MVP was assembled from expressed sequence tags and genome traces, and the predicted protein was found to have 64% identity and 81% similarity to rat MVP. Sea urchin MVP includes seven ~50 residue repeats in the N-terminal half of the protein and a predicted coiled coil domain in the C-terminus, as does rat MVP. A cryoelectron microscopy (cryoEM reconstruction of isolated sea urchin vaults reveals the assembly to have a barrel-shaped external structure that is nearly identical to the rat vault structure. Analysis of the molecular composition of the sea urchin vault indicates that it contains components that may be homologs of the mammalian vault RNA component (vRNA and protein components (VPARP and TEP1. The sea urchin vault appears to have additional protein components in the molecular weight range of 14–55 kDa that might correspond to molecular contents. Confocal experiments indicate a dramatic relocalization of MVP from the cytoplasm to the nucleus during sea urchin embryogenesis. Conclusions These results are suggestive of a role for the vault in delivering macromolecules to the nucleus during development.

Implementation of safeguards at modular vault dry store at Paks NPP in Hungary

International Nuclear Information System (INIS)

Safar, J.; Czoch, I.; Szoellosi, E.F.; Janov, J.; Sannie, G.; Daniel, G.; Szabo, J.L.

1999-01-01

A safeguards system has been implemented at the GEC-Alsthom designed Modular Vault Dry Store at Paks NPP in Hungary without previous safeguards related experience for this type of spent fuel storage. C/S measures and sealing have primary importance. In addition. spent fuel attribute signatures are detected by a fuel transfer monitor at the cask load/unload port. These are complemented with the corresponding accounting measures. (author)

Dynamic research of masonry vault in a technical scale

Science.gov (United States)

Golebiewski, Michal; Lubowiecka, Izabela; Kujawa, Marcin

2017-03-01

The paper presents preliminary results of dynamic tests of the masonry barrel vault in a technical scale. Experimental studies are intended to identify material properties of homogenized masonry vaults under dynamic loads. The aim of the work is to create numerical models to analyse vault's dynamic response to dynamic loads in a simplest and accurate way. The process of building the vault in a technical scale is presented in the paper. Furthermore a excitation of vibrations with an electrodynamic modal exciter placed on the vault, controlled by an arbitrary waveform function generator, is discussed. Finally paper presents trends in the research for homogenization algorithm enabling dynamic analysis of masonry vaults. Experimental results were compared with outcomes of so-called macromodels (macromodel of a brick masonry is a model in which masonry, i.e. a medium consisting of two different fractions - bricks and mortar, is represented by a homogenized, uniformed, material). Homogenization entail significant simplifications, nevertheless according to the authors, can be a useful approach in a static and dynamic analysis of masonry structures.

Immobilized low-activity waste interim storage facility, Project W-465 conceptual design report

International Nuclear Information System (INIS)

Pickett, W.W.

1997-01-01

This report outlines the design and Total Estimated Cost to modify the four unused grout vaults for the remote handling and interim storage of immobilized low-activity waste (ILAW). The grout vault facilities in the 200 East Area of the Hanford Site were constructed in the 1980s to support Tank Waste disposal activities. The facilities were to serve project B-714 which was intended to store grouted low-activity waste. The existing 4 unused grout vaults, with modifications for remote handling capability, will provide sufficient capacity for approximately three years of immobilized low activity waste (ILAW) production from the Tank Waste Remediation System-Privatization Vendors (TWRS-PV). These retrofit modifications to the grout vaults will result in an ILAW interim storage facility (Project W465) that will comply with applicable DOE directives, and state and federal regulations

Immobilized low-activity waste interim storage facility, Project W-465 conceptual design report

Energy Technology Data Exchange (ETDEWEB)

Pickett, W.W.

1997-12-30

This report outlines the design and Total Estimated Cost to modify the four unused grout vaults for the remote handling and interim storage of immobilized low-activity waste (ILAW). The grout vault facilities in the 200 East Area of the Hanford Site were constructed in the 1980s to support Tank Waste disposal activities. The facilities were to serve project B-714 which was intended to store grouted low-activity waste. The existing 4 unused grout vaults, with modifications for remote handling capability, will provide sufficient capacity for approximately three years of immobilized low activity waste (ILAW) production from the Tank Waste Remediation System-Privatization Vendors (TWRS-PV). These retrofit modifications to the grout vaults will result in an ILAW interim storage facility (Project W465) that will comply with applicable DOE directives, and state and federal regulations.

Expression of the Major Vault Protein (MVP) and Cellular Vault Particles in Fish.

Science.gov (United States)

Margiotta, Alyssa L; Bain, Lisa J; Rice, Charles D

2017-11-01

Cellular vaults are ubiquitous 13 mega Da multi-subunit ribonuceloprotein particles that may have a role in nucleocytoplasmic transport. Seventy percent of the vault's mass consists of a ≈100 kDa protein, the major vault protein (MVP). In humans, a drug resistance-associated protein, originally identified as lung resistance protein in metastatic lung cancer, was ultimately shown to be the previously described MVP. In this study, a partial MVP sequence was cloned from channel catfish. Recombinant MVP (rMVP) was used to generate a monoclonal antibody that recognizes full length protein in distantly related fish species, as well as mice. MVP is expressed in fish spleen, liver, anterior kidney, renal kidney, and gills, with a consistent expression in epithelial cells, macrophages, or endothelium at the interface of the tissue and environment or vasculature. We show that vaults are distributed throughout cells of fish lymphoid cells, with nuclear and plasma membrane aggregations in some cells. Protein expression studies were extended to liver neoplastic lesions in Atlantic killifish collected in situ at the Atlantic Wood USA-EPA superfund site on the southern branch of the Elizabeth River, VA. MVP is highly expressed in these lesions, with intense staining at the nuclear membrane, similar to what is known about MVP expression in human liver neoplasia. Additionally, MVP mRNA expression was quantified in channel catfish ovarian cell line following treatment with different classes of pharmacological agents. Notably, mRNA expression is induced by ethidium bromide, which damages DNA. Anat Rec, 2017. © 2017 Wiley Periodicals, Inc. Anat Rec, 300:1981-1992, 2017. © 2017 Wiley Periodicals, Inc. © 2017 Wiley Periodicals, Inc.

Calculation of mixed HEU-LEU cores for the HOR research reactor with the scale code system

International Nuclear Information System (INIS)

Leege, P.F.A. de; Gibcus, H.P.M.; Hoogenboom, J.E.; Vries, J.W. de

1997-01-01

The HOR reactor of Interfaculty Reactor Institute (IRI), Delft, The Netherlands, will be converted to use low enriched fuel (LEU) assemblies. As there are still many usable high enriched (HEU) fuel assemblies present, there will be a considerable reactor operation time with mixed cores with both HEU and LEU fuel assemblies. At IRI a comprehensive reactor physics code system and evaluated nuclear data is implemented for detailed core calculations. One of the backbones of the IRI code system is the well-known SCALE code system package. Full core calculations are performed with the diffusion theory code BOLD VENTURE, the nodal code SILWER, and the Monte Carlo code KENO Va. Results are displayed of a strategy from a HEU core to a mixed HEU-LEU core and eventually a LEU core. (author)

309 plutonium recycle test reactor ion exchanger vault deactivitation report

International Nuclear Information System (INIS)

Griffin, P.W.

1996-03-01

This report documents the deactivation of the ion exchanger vault at the 309 Plutonium Recycle Test Reactor (PRTR) Facility in the 300 Area. The vault deactivation began in May 1995 and was completed in June 1995. The final site restoration and shipment of the low-level waste for disposal was finished in September 1995. The ion exchanger vault deactivation project involved the removal and disposal of twelve ion exchangers and decontaminating and fixing of residual smearable contamination on the ion exchanger vault concrete surfaces

Computational Fluid Dynamics Model for Saltstone Vault 4 Vapor Space

International Nuclear Information System (INIS)

Lee, Si Young

2005-01-01

Computational fluid dynamics (CFD) methods have been used to estimate the flow patterns for vapor space inside the Saltstone Vault No.4 under different operating scenarios. The purpose of this work is to examine the gas motions inside the vapor space under the current vault configurations. A CFD model took three-dimensional transient momentum-energy coupled approach for the vapor space domain of the vault. The modeling calculations were based on prototypic vault geometry and expected normal operating conditions as defined by Waste Solidification Engineering. The modeling analysis was focused on the air flow patterns near the ventilated corner zones of the vapor space inside the Saltstone vault. The turbulence behavior and natural convection mechanism used in the present model were benchmarked against the literature information and theoretical results. The verified model was applied to the Saltstone vault geometry for the transient assessment of the air flow patterns inside the vapor space of the vault region using the boundary conditions as provided by the customer. The present model considered two cases for the estimations of the flow patterns within the vapor space. One is the reference baseline case. The other is for the negative temperature gradient between the roof inner and top grout surface temperatures intended for the potential bounding condition. The flow patterns of the vapor space calculated by the CFD model demonstrate that the ambient air comes into the vapor space of the vault through the lower-end ventilation hole, and it gets heated up by the Benard-cell type circulation before leaving the vault via the higher-end ventilation hole. The calculated results are consistent with the literature information

Verification experiment on the downblending of high enriched uranium (HEU) at the Portsmouth Gaseous Diffusion Plant. Digital video surveillance of the HEU feed stations

International Nuclear Information System (INIS)

Martinez, R.L.; Tolk, K.; Whiting, N.; Castleberry, K.; Lenarduzzi, R.

1998-01-01

As part of a Safeguards Agreement between the US and the International Atomic Energy Agency (IAEA), the Portsmouth Gaseous Diffusion Plant, Piketon, Ohio, was added to the list of facilities eligible for the application of IAEA safeguards. Currently, the facility is in the process of downblending excess inventory of HEU to low enriched uranium (LEU) from US defense related programs for commercial use. An agreement was reached between the US and the IAEA that would allow the IAEA to conduct an independent verification experiment at the Portsmouth facility, resulting in the confirmation that the HEU was in fact downblended. The experiment provided an opportunity for the DOE laboratories to recommend solutions/measures for new IAEA safeguards applications. One of the measures recommended by Sandia National Laboratories (SNL), and selected by the IAEA, was a digital video surveillance system for monitoring activity at the HEU feed stations. This paper describes the SNL implementation of the digital video system and its integration with the Load Cell Based Weighing System (LCBWS) from Oak Ridge National Laboratory (ORNL). The implementation was based on commercially available technology that also satisfied IAEA criteria for tamper protection and data authentication. The core of the Portsmouth digital video surveillance system was based on two Digital Camera Modules (DMC-14) from Neumann Consultants, Germany

Efflux kinetics and intracellular distribution of daunorubicin are not affected by major vault protein/lung resistance-related protein (vault) expression.

NARCIS (Netherlands)

Zon, van A; Mossink, MH; Schoester, M.; Scheper, R.J.; Sonneveld, P.; Wiemer, EA

2004-01-01

Vaults may contribute to multidrug resistance by transporting drugs away from their subcellular targets. To study the involvement of vaults in the extrusion of anthracyclines from the nucleus, we investigated the handling of daunorubicin by drug-sensitive and drug-resistant non-small lung cancer

Neutronic performance of a 14 MW TRIGA reactor: LEU vs HEU fuel

International Nuclear Information System (INIS)

Bretscher, M.M.; Snelgrove, J.L.; Cornella, R.J.

1983-01-01

A primary objective of the US Reduced Enrichment Research and Test Reactor (RERTR) Program is to develop means for replacing, wherever possible, currently used highly-enriched uranium (HEU) fuel ( 235 U enrichment > 90%) with low-enriched uranium (LEU) fuel ( 235 U enrichment < 20%) without significantly degrading the performance of research and test reactors. The General Atomic Company has developed a low-enriched but high uranium content Er-U-ZrH/sub 1.6/ fuel to enable the conversion of TRIGA reactors (and others) from HEU to LEU. One possible application is to the water-moderated 14 MW TRIGA Steady State Reactor (SSR) at the Romanian Institute for Nuclear Power Reactors. The work reported here was undertaken for the purpose of comparing the neutronic performance of the SSR for HEU fuel with that for LEU fuel. In order to make these relative comparisons as valid as possible, identical methods and models were used for the neutronic calculations

Vault Nanoparticles Packaged with Enzymes as an Efficient Pollutant Biodegradation Technology.

Science.gov (United States)

Wang, Meng; Abad, Danny; Kickhoefer, Valerie A; Rome, Leonard H; Mahendra, Shaily

2015-11-24

Vault nanoparticles packaged with enzymes were synthesized as agents for efficiently degrading environmental contaminants. Enzymatic biodegradation is an attractive technology for in situ cleanup of contaminated environments because enzyme-catalyzed reactions are not constrained by nutrient requirements for microbial growth and often have higher biodegradation rates. However, the limited stability of extracellular enzymes remains a major challenge for practical applications. Encapsulation is a recognized method to enhance enzymatic stability, but it can increase substrate diffusion resistance, lower catalytic rates, and increase the apparent half-saturation constants. Here, we report an effective approach for boosting enzymatic stability by single-step packaging into vault nanoparticles. With hollow core structures, assembled vault nanoparticles can simultaneously contain multiple enzymes. Manganese peroxidase (MnP), which is widely used in biodegradation of organic contaminants, was chosen as a model enzyme in the present study. MnP was incorporated into vaults via fusion to a packaging domain called INT, which strongly interacts with vaults' interior surface. MnP fused to INT and vaults packaged with the MnP-INT fusion protein maintained peroxidase activity. Furthermore, MnP-INT packaged in vaults displayed stability significantly higher than that of free MnP-INT, with slightly increased Km value. Additionally, vault-packaged MnP-INT exhibited 3 times higher phenol biodegradation in 24 h than did unpackaged MnP-INT. These results indicate that the packaging of MnP enzymes in vault nanoparticles extends their stability without compromising catalytic activity. This research will serve as the foundation for the development of efficient and sustainable vault-based bioremediation approaches for removing multiple contaminants from drinking water and groundwater.

SWSA [Solid Waste Storage Area] 6 tumulus disposal demonstration

International Nuclear Information System (INIS)

Van Hoesen, S.D.; Clapp, R.B.

1987-01-01

A facility to demonstrate the above-grade disposal of solid low-level radioactive wastes (LLW) is being constructed in the Solid Waste Storage Area 6 (SWSA 6) at the Oak Ridge National Laboratory (ORNL). The demonstration facility will utilize the ''Tumulus'' technology, which basically involves sealing the waste in concrete vaults, placing the vaults on a grade level concrete pad, and covering the pad with a soil cover after vault placement is complete. Loading of the demonstration unit is scheduled to begin in June, and will continue one to one and a half years until the 28,000 ft 3 capacity is exhausted

Rupture Loop Annex (RLA) ion exchange vault entry and characterization

International Nuclear Information System (INIS)

Ham, J.E.

1996-01-01

This engineering report documents the entry and characterization of the Rupture Loop Annex Ion Exchange (RLAIX) Vault located near the 309 Building's Plutonium Recycle Test Reactor (PRTR). Twelve ion exchange columns were found in the vault. Some of which contained transuranics, Cs 137, and Co 60. The characterization information is necessary for future vault cleanout and column disposal

Ion-induced gammas for photofission interrogation of HEU.

Energy Technology Data Exchange (ETDEWEB)

Doyle, Barney Lee (Sandia National Laboratories, Albuquerque, NM); Antolak, Arlyn J.; Morse, Daniel H.; Provencio, Paula Polyak (Sandia National Laboratories, Albuquerque, NM)

2006-03-01

High-energy photons and neutrons can be used to actively interrogate for heavily shielded special nuclear material (SNM), such as HEU (highly enriched uranium), by detecting prompt and/or delayed induced fission signatures. In this work, we explore the underlying physics for a new type of photon source that generates high fluxes of mono-energetic gamma-rays from low-energy (<500 keV) proton-induced nuclear reactions. The characteristic energies (4- to 18-MeV) of the gamma-rays coincide with the peak of the photonuclear cross section. The source could be designed to produce gamma-rays of certain selected energies, thereby improving the probability of detecting shielded HEU or providing a capability to determine enrichment inside sealed containers. The fundamental physics of such an interrogation source were studied in this LDRD through scaled ion accelerator experiments and radiation transport modeling. The data were used to assess gamma and neutron yields, background, and photofission-induced signal levels from several (p,{gamma}) target materials under consideration.

Department of Energy HEU ES and H vulnerability assessment, Savannah River Site, Site Assessment Team report. Revision 2

International Nuclear Information System (INIS)

Geddes, R.L.; Barone, A.; Shook, H.E. Varner, C.E.; Rollins, R.

1996-01-01

This report fulfills the directive issued by the Secretary of Energy on February 22, 1996 to complete a comprehensive assessment of potential vulnerabilities associated with the management of highly enriched uranium (HEU) throughout the DOE complex. In a subsequent letter instruction, the DOE-SR Field Office formally directed WSRC to conduct an assessment of the HEU materials at SRS. The term ''ES and H vulnerabilities'' is defined for the purpose of this assessment to mean conditions or weaknesses that could lead to unnecessary or increased exposure of workers or the public to radiation or to HEU-associated chemical hazards, or to the release of radioactive materials to the environment. The assessment will identify and prioritize ES and H vulnerabilities, and will serve as an information base for identifying corrective actions for the safe management of HEU. Primary facilities that hold HEU at SRS are H-Canyon, K-Reactor assembly area, K, L, and P-Reactor disassembly basins, and the Receiving Basin for Offsite Fuels (RBOF)

Economics of dry storage systems

International Nuclear Information System (INIS)

Moore, G.R.; Winders, R.C.

1980-01-01

This paper postulates a dry storage application suitable as a regional away-from-reactor storage (AFR), develops an economical system design concept and estimates system costs. The system discussed uses the experience gained in the dry storage research activities and attempts to present a best foot forward system concept. The major element of the system is the Receiving and Packaging Building. In this building fuel assemblies are removed from transportation casks and encapsulated for storage. This facility could be equally applicable to silo, vault, or caisson storage. However the caisson storage concept has been chosen for discussion purposes

Criticality safety of storage barrels for enriched uranium fresh fuel at the RB research reactor

International Nuclear Information System (INIS)

Pesic, M. P.

1997-01-01

Study on criticality safety of fresh low and high enriched uranium (LEU and HEU) fuel elements in the storage/transport barrels at the RB research reactor is carried out by using the well-known MCNP computer code. It is shown that studied arrays of tightly closed fuel barrels, each entirely loaded with 100 fresh (HEU or LEU) fuel slugs, are far away from criticality, even in cases of an unexpected flooding by light water.(author)

Experimental Analysis of Dynamic Effects of FRP Reinforced Masonry Vaults

Science.gov (United States)

Corradi, Marco; Borri, Antonio; Castori, Giulio; Coventry, Kathryn

2015-01-01

An increasing interest in the preservation of historic structures has produced a need for new methods for reinforcing curved masonry structures, such as arches and vaults. These structures are generally very ancient, have geometries and materials which are poorly defined and have been exposed to long-term historical movements and actions. Consequently, they are often in need of repair or reinforcement. This article presents the results of an experimental study carried out in the laboratory and during on-site testing to investigate the behaviour of brick masonry vaults under dynamic loading strengthened with FRPs (Fiber Reinforced Polymers). For the laboratory tests, the brick vaults were built with solid sanded clay bricks and weak mortar and were tested under dynamic loading. The experimental tests were designed to facilitate analysis of the dynamic behaviour of undamaged, damaged and reinforced vaulted structures. On-site tests were carried out on an earthquake-damaged thin brick vault of an 18th century aristocratic residence in the city of L’Aquila, Italy. The provision of FRP reinforcement is shown to re-establish elastic behavior previously compromised by time induced damage in the vaults. PMID:28793697

Experimental Analysis of Dynamic Effects of FRP Reinforced Masonry Vaults.

Science.gov (United States)

Corradi, Marco; Borri, Antonio; Castori, Giulio; Coventry, Kathryn

2015-11-27

An increasing interest in the preservation of historic structures has produced a need for new methods for reinforcing curved masonry structures, such as arches and vaults. These structures are generally very ancient, have geometries and materials which are poorly defined and have been exposed to long-term historical movements and actions. Consequently, they are often in need of repair or reinforcement. This article presents the results of an experimental study carried out in the laboratory and during on-site testing to investigate the behaviour of brick masonry vaults under dynamic loading strengthened with FRPs (Fiber Reinforced Polymers). For the laboratory tests, the brick vaults were built with solid sanded clay bricks and weak mortar and were tested under dynamic loading. The experimental tests were designed to facilitate analysis of the dynamic behaviour of undamaged, damaged and reinforced vaulted structures. On-site tests were carried out on an earthquake-damaged thin brick vault of an 18th century aristocratic residence in the city of L'Aquila, Italy. The provision of FRP reinforcement is shown to re-establish elastic behavior previously compromised by time induced damage in the vaults.

Experimental Analysis of Dynamic Effects of FRP Reinforced Masonry Vaults

Directory of Open Access Journals (Sweden)

Marco Corradi

2015-11-01

Full Text Available An increasing interest in the preservation of historic structures has produced a need for new methods for reinforcing curved masonry structures, such as arches and vaults. These structures are generally very ancient, have geometries and materials which are poorly defined and have been exposed to long-term historical movements and actions. Consequently, they are often in need of repair or reinforcement. This article presents the results of an experimental study carried out in the laboratory and during on-site testing to investigate the behaviour of brick masonry vaults under dynamic loading strengthened with FRPs (Fiber Reinforced Polymers. For the laboratory tests, the brick vaults were built with solid sanded clay bricks and weak mortar and were tested under dynamic loading. The experimental tests were designed to facilitate analysis of the dynamic behaviour of undamaged, damaged and reinforced vaulted structures. On-site tests were carried out on an earthquake-damaged thin brick vault of an 18th century aristocratic residence in the city of L’Aquila, Italy. The provision of FRP reinforcement is shown to re-establish elastic behavior previously compromised by time induced damage in the vaults.

Passive Time Coincidence Measurements with HEU and DU Metal Castings

International Nuclear Information System (INIS)

McConchie, Seth M.; Hausladen, Paul; Mihalczo, John T.; Wright, Michael C.; Archer, Daniel E.

2008-01-01

A Department of Energy sponsored Oak Ridge National Laboratory/Y-12 National Security Complex program of passive time coincidence measurements has been initiated at Y-12 to evaluate the ability to determine the presence of high enriched uranium (HEU) and distinguish it from depleted uranium (DU). This program uses the Nuclear Materials Identification System (NMIS) without an active interrogation source. Previous passive NMIS measurements with Pu metal and Pu oxide have been successful in determining the Pu mass, assuming a known 240Pu content. The spontaneous fission of uranium metal is considerably lower than Pu and measurements of this type have been performed at Lawrence Livermore National Laboratory. This work presents results of measurements of HEU and DU metal castings using moderated 3He detectors.

Material handling for the Los Alamos National Laboratory Nuclear Storage Facility

International Nuclear Information System (INIS)

Pittman, P.; Roybal, J.; Durrer, R.; Gordon, D.

1999-01-01

This paper will present the design and application of material handling and automation systems currently being developed for the Los Alamos National Laboratory (LANL) Nuclear Material Storage Facility (NMSF) renovation project. The NMSF is a long-term storage facility for nuclear material in various forms. The material is stored within tubes in a rack called a basket. The material handling equipment range from simple lift assist devices to more sophisticated fully automated robots, and are split into three basic systems: a Vault Automation System, an NDA automation System, and a Drum handling System. The Vault Automation system provides a mechanism to handle a basket of material cans and to load/unload storage tubes within the material vault. In addition, another robot is provided to load/unload material cans within the baskets. The NDA Automation System provides a mechanism to move material within the small canister NDA laboratory and to load/unload the NDA instruments. The Drum Handling System consists of a series of off the shelf components used to assist in lifting heavy objects such as pallets of material or drums and barrels

Measurements of the HEU and LEU in-core spectra at the Ford Nuclear Reactor

Energy Technology Data Exchange (ETDEWEB)

Wehe, D K [Oak Ridge National Laboratory, Oak Ridge, TN (United States); King, J S; Lee, J C; Martin, W R [Department of Nuclear Engineering, University of Michigan, Ann Arbor, MI (United States)

1985-07-01

The Ford Nuclear Reactor (FNR) at the University of Michigan has been serving as the test site for a low-enriched uranium (LEU) fuel whole-core demonstration. As part of the experimental program, the differential neutron spectrum has been measured in a high-enriched uranium (HEU) core and an LEU core. The HEU and LEU spectra were determined by unfolding the measured activities of foils that were irradiated in the reactor. When the HEU and LEU spectra are compared from meV to 10 MeV, significant differences between the two spectra are apparent below 10 eV. These are probably caused by the additional {sup 238}U resonance absorption in the LEU fuel. No measurable difference occurs in the shape of the spectra above MeV. (author)

Mixed core conversion study with HEU and LEU fuels

International Nuclear Information System (INIS)

Matos, J.E.; Freese, K.E.

1984-01-01

The results of a mixed core study are presented for gradual replacement of HEU fuel with LEU fuel using the IAEA generic 10 MW reactor as an example. The key parameters show that the transition can be accomplished safely and economically

The Effect of the Crystalline Lens on Central Vault After Implantable Collamer Lens Implantation.

Science.gov (United States)

Qi, Meng-Ying; Chen, Qian; Zeng, Qing-Yan

2017-08-01

To identify associations between crystalline lens-related factors and central vault after Implantable Collamer Lens (ICL) (Staar Surgical, Monrovia, CA) implantation. This retrospective clinical study included 320 eyes from 186 patients who underwent ICL implantation surgery. At 1 year after surgery, the central vault was measured using anterior segment optical coherence tomography. Preoperative anterior chamber depth, lens thickness, lens position (lens position = anterior chamber depth + 1/2 lens thickness), and vault were analyzed to investigate the effects of lens-related factors on postoperative vault. The mean vault was 513 ± 215 µm at 1 year after surgery. Vault was positively correlated with preoperative anterior chamber depth (r = 0.495, P lens position (r = 0.371, P lens thickness (r = -0.262, P lens position than eyes in the other two vault groups (which had vaults ≥ 250 µm) (P lens position less than 5.1 mm had greatly reduced vaults (P lens could have an important influence on postoperative vault. Eyes with a shallower anterior chamber and a forward lens position will have lower vaults. [J Refract Surg. 2017;33(8):519-523.]. Copyright 2017, SLACK Incorporated.

New fuel vault criticality analysis at Chinshan nuclear power station with new approaches to improve the storage flexibility

International Nuclear Information System (INIS)

Huang, P. H.

2010-10-01

The Chinshan new fuel vault (NFV) consists of 13 fuel storage racks, each rack may store 10 fuel assemblies. Prior to 2008, the NFV had never been used and the practice by the Taiwan Power Company (TPC) was to temporarily store the fuel assemblies in the shipping containers after received, until the inspection work was performed shortly before the outage, and then assemblies were loaded directly into the spent fuel pool (SFP). Starting from 2009, this practice has been revised since the new fuel contract would only supply a small amount of containers for storage, and the SFP would lose full-core-off load capability soon; therefore, use of NFV to store fuel assemblies following inspection becomes extremely crucial. The original Chinshan NFV criticality analysis was performed for the initial fuel design. Although many new fuel designs had been used (e.g., Atrium-10 reported in PBNC-14), no reanalysis had been performed because it was not anticipated that NFV would be used. Therefore, TPC requested the vendor to perform the analysis for Atrium-10. Originally, the vendor estimated that number of assemblies allowed to be stored would be limited severely to about 60. To enhance storage flexibility, Tpc proposed some new approaches: 1) All assemblies are assumed in vendor's standard method to contain a single limiting lattice for entire fuel length, it is suggested that axially zoned limiting lattices be selected based on characteristics of reloads to be delivered, and this significantly improves flexibility. 2) The maximum k-effective equation used by vendor was corrected (manufacturing tolerances were conservatively mistreated). Also, the vendor typically used 0.95 k-effective as the criterion, it is suggested that NUREG-0800 requirement (≤0.98 for optimum moderation conditions) be applied. After several iterations, all the 130 locations are allowed to store fuel. The analysis report has been approved by the authority in June 2008. (Author)

New fuel vault criticality analysis at Chinshan nuclear power station with new approaches to improve the storage flexibility

Energy Technology Data Exchange (ETDEWEB)

Huang, P. H., E-mail: u808966@taipower.com.t [Taiwan Power Company, Department of Nuclear Generation, 242 Roosevelt Rd., Sec. 3, Taipei, Taiwan (China)

2010-10-15

The Chinshan new fuel vault (NFV) consists of 13 fuel storage racks, each rack may store 10 fuel assemblies. Prior to 2008, the NFV had never been used and the practice by the Taiwan Power Company (TPC) was to temporarily store the fuel assemblies in the shipping containers after received, until the inspection work was performed shortly before the outage, and then assemblies were loaded directly into the spent fuel pool (SFP). Starting from 2009, this practice has been revised since the new fuel contract would only supply a small amount of containers for storage, and the SFP would lose full-core-off load capability soon; therefore, use of NFV to store fuel assemblies following inspection becomes extremely crucial. The original Chinshan NFV criticality analysis was performed for the initial fuel design. Although many new fuel designs had been used (e.g., Atrium-10 reported in PBNC-14), no reanalysis had been performed because it was not anticipated that NFV would be used. Therefore, TPC requested the vendor to perform the analysis for Atrium-10. Originally, the vendor estimated that number of assemblies allowed to be stored would be limited severely to about 60. To enhance storage flexibility, Tpc proposed some new approaches: 1) All assemblies are assumed in vendor's standard method to contain a single limiting lattice for entire fuel length, it is suggested that axially zoned limiting lattices be selected based on characteristics of reloads to be delivered, and this significantly improves flexibility. 2) The maximum k-effective equation used by vendor was corrected (manufacturing tolerances were conservatively mistreated). Also, the vendor typically used 0.95 k-effective as the criterion, it is suggested that NUREG-0800 requirement ({<=}0.98 for optimum moderation conditions) be applied. After several iterations, all the 130 locations are allowed to store fuel. The analysis report has been approved by the authority in June 2008. (Author)

Seismic analysis of the ICPP high level liquid waste tanks and vaults

International Nuclear Information System (INIS)

Uldrich, E.D.; Malik, L.E.

1991-01-01

Two buried, closely spaced, reinforced concrete vaults founded on base rock were evaluated for gravity and safe shutdown earthquake loads. These vaults enclose steel tanks used to store high level radioactive liquid waste. Detailed 3-dimensional finite element models were used for state-of-the-art structure-soil-structure interaction (SSSI) analyses. Three soil dynamic property profiles were used to address soil variability. Vault accelerations are not significantly affected by the variability of soil dynamic properties. Lower bound soil properties yield maximum member forces and moments. Demands on the side closer to the other vault due to horizontal motions are lower due to SSSI effects. Combined gravity and seismic demand on the vault force resisting system was calculated. The vaults were qualified, using member capacities based on current design codes

ADS with HEU in the Vinca Institute

International Nuclear Information System (INIS)

Pesic, M.; Sobolevsky, N.

2000-01-01

The 'Conceptual design of ADS' is a new project proposed in the Vin.a Institute for the next three years. In this paper, an option in the project - an idea of high-enriched uranium (HEU) - H 2 O low-flux ADS is shown. Preliminary results of design study and calculations of the beam-target interaction and neutronics of proposed sub-critical system are given. (author)

324 Radiochemical engineering cells and high level vault tanks mixed waste compliance status

International Nuclear Information System (INIS)

1994-01-01

The 324 Building in the Hanford 300 Area contains Radiochemical Engineering Cells and High Level Vault tanks (the open-quotes REC/HLVclose quotes) for research and development activities involving radioactive materials. Radioactive mixed waste within this research installation, found primarily in B-Cell and three of the high level vault tanks, is subject to RCRA/DWR (open-quotes RCRAclose quotes) regulations for storage. This white paper provides a baseline RCRA compliance summary of MW management in the REC/HLV, based on best available knowledge. The REC/HLV compliance project, of which this paper is a part, is intended to achieve the highest degree of compliance practicable given the special technical difficulties of managing high activity radioactive materials, and to assure protection of human health and safety and the environment. The REC/HLV was constructed in 1965 to strict standards for the safe management of highly radioactive materials. Mixed waste in the REC/HLV consists of discarded tools and equipment, dried feed stock from nuclear waste melting experiments, contaminated particulate matter, and liquid feed stock from various experimental programs in the vault tanks. B-Cell contains most of these materials. Total radiological inventory in B-Cell is estimated at 3 MCi, about half of which is potentially open-quotes dispersibleclose quotes, that is, it is in small pieces or mobile particles. Most of the mixed waste currently in the REC/HLV was generated or introduced before mixed wastes were subjected to RCRA in 1987

Exploratory Investigation of Impact Loads During the Forward Handspring Vault

Directory of Open Access Journals (Sweden)

Penitente Gabriella

2015-06-01

Full Text Available The purpose of this study was to examine kinematic and kinetic differences in low and high intensity hand support impact loads during a forward handspring vault. A high-speed video camera (500 Hz and two portable force platforms (500 Hz were installed on the surface of the vault table. Two-dimensional analyses were conducted on 24 forward handspring vaults performed by 12 senior level, junior Olympic program female gymnasts (16.9 ±1.4 yr; body height 1.60 ±0.1 m; body mass 56.7 ±7.8 kg. Load intensities at impact with the vault table were classified as low (peak force 0.8 × body weight. These vaults were compared via crucial kinetic and kinematic variables using independent t-tests and Pearson correlations. Statistically significant (p < 0.001 differences were observed in peak force (t(24 = 4.75, ES = 3.37 and time to peak force (t(24 = 2.07, ES = 1.56. Statistically significant relationships between the loading rate and time to peak force were observed for high intensity loads. Peak force, time to peak force, and a shoulder angle at impact were identified as primary variables potentially involved in the determination of large repetitive loading rates on the forward handspring vault.

Major vault protein in cardiac and smooth muscle.

Science.gov (United States)

Shults, Nataliia V; Das, Dividutta; Suzuki, Yuichiro J

Major vault protein (MVP) is the major component of the vault particle whose functions are not well understood. One proposed function of the vault is to serve as a mechanism of drug transport, which confers drug resistance in cancer cells. We show that MVP can be found in cardiac and smooth muscle. In human airway smooth muscle cells, knocking down MVP was found to cause cell death, suggesting that MVP serves as a cell survival factor. Further, our laboratory found that MVP is S-glutathionylated in response to ligand/receptor-mediated cell signaling. The S-glutathionylation of MVP appears to regulate protein-protein interactions between MVP and a protein called myosin heavy chain 9 (MYH9). Through MYH9 and Vsp34, MVP may form a complex with Beclin-1 that regulates autophagic cell death. In pulmonary vascular smooth muscle, proteasome inhibition promotes the ubiquitination of MVP, which may function as a mechanism of proteasome inhibition-mediated cell death. Investigating the functions and the regulatory mechanisms of MVP and vault particles is an exciting new area of research in cardiovascular/pulmonary pathophysiology.

Mapping genetic variants for cranial vault shape in humans

DEFF Research Database (Denmark)

Roosenboom, Jasmien; Lee, Myoung Keun; Hecht, Jacqueline T

2018-01-01

The shape of the cranial vault, a region comprising interlocking flat bones surrounding the cerebral cortex, varies considerably in humans. Strongly influenced by brain size and shape, cranial vault morphology has both clinical and evolutionary relevance. However, little is known about the geneti...

Experiments with HEU (93.14 wt.%) metal annuli with internal graphite cylinder

Energy Technology Data Exchange (ETDEWEB)

Liu, Xiaobo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wehmann, Udo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mihalczo, John T. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-04-01

A variety of critical experiments were constructed of enriched uranium metal (oralloy ) during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, only three experimental configurations are described here. They are internal graphite reflected metal uranium assemblies with three different diameter HEU annuli (15-9 inches, 15-7 inches and 13-7 inches). These experiments can be found in Reference 1 and in their associated logbook

Corrective Action Plan for Corrective Action Unit 135: Area 25 Underground Storage Tanks, Nevada Test Site, Nevada

International Nuclear Information System (INIS)

Cox, D. H.

2000-01-01

The Area 25 Underground Storage Tanks site Corrective Action Unit (CAU) 135 will be closed by unrestricted release decontamination and verification survey, in accordance with the Federal Facility Agreement and Consert Order (FFACO, 1996). The CAU includes one Corrective Action Site (CAS). The Area 25 Underground Storage Tanks, (CAS 25-02-01), referred to as the Engine-Maintenance Assembly and Disassembly (E-MAD) Waste Holdup Tanks and Vault, were used to receive liquid waste from all of the radioactive drains at the E-MAD Facility. Based on the results of the Corrective Action Investigation conducted in June 1999 discussed in the Corrective Action Investigation Plan for Corrective Action Unit 135: Area 25 Underground Storage Tanks, Nevada Test Site, Nevada (DOE/NV,1999a), one sample from the radiological survey of the concrete vault interior exceeded radionuclide preliminary action levels. The analytes from the sediment samples that exceeded the preliminary action levels are polychlorinated biphenyls, Resource Conservation and Recovery Act metals, total petroleum hydrocarbons as diesel-range organics, and radionuclides. Unrestricted release decontamination and verification involves removal of concrete and the cement-lined pump sump from the vault. After verification that the contamination has been removed, the vault will be repaired with concrete, as necessary. The radiological- and chemical-contaminated pump sump and concrete removed from the vault would be disposed of at the Area 5 Radioactive Waste Management Site. The vault interior will be field surveyed following removal of contaminated material to verify that unrestricted release criteria have been achieved

Corrective Action Plan for Corrective Action Unit 135: Area 25 Underground Storage Tanks, Nevada Test Site, Nevada

Energy Technology Data Exchange (ETDEWEB)

D. H. Cox

2000-07-01

The Area 25 Underground Storage Tanks site Corrective Action Unit (CAU) 135 will be closed by unrestricted release decontamination and verification survey, in accordance with the Federal Facility Agreement and Consert Order (FFACO, 1996). The CAU includes one Corrective Action Site (CAS). The Area 25 Underground Storage Tanks, (CAS 25-02-01), referred to as the Engine-Maintenance Assembly and Disassembly (E-MAD) Waste Holdup Tanks and Vault, were used to receive liquid waste from all of the radioactive drains at the E-MAD Facility. Based on the results of the Corrective Action Investigation conducted in June 1999 discussed in the Corrective Action Investigation Plan for Corrective Action Unit 135: Area 25 Underground Storage Tanks, Nevada Test Site, Nevada (DOE/NV,1999a), one sample from the radiological survey of the concrete vault interior exceeded radionuclide preliminary action levels. The analytes from the sediment samples that exceeded the preliminary action levels are polychlorinated biphenyls, Resource Conservation and Recovery Act metals, total petroleum hydrocarbons as diesel-range organics, and radionuclides. Unrestricted release decontamination and verification involves removal of concrete and the cement-lined pump sump from the vault. After verification that the contamination has been removed, the vault will be repaired with concrete, as necessary. The radiological- and chemical-contaminated pump sump and concrete removed from the vault would be disposed of at the Area 5 Radioactive Waste Management Site. The vault interior will be field surveyed following removal of contaminated material to verify that unrestricted release criteria have been achieved.

Evaluation of economics of spent fuel storage techniques

International Nuclear Information System (INIS)

Yamaji, Kenji; Nagano, Koji

1988-01-01

Various spent fuel storage techniques are evaluated in terms of required costs. The unit storage cost for each spent fuel storage scenario is calculated based on the total cost required for the scenario including capital expenditure, operation cost, maintenance cost and transport cost. Intermediate storage may be performed in relatively small facilities in the plant or in independent large-scale facilities installed away from the plant. Dry casks or water pools are assumed to be used in in-plant storage facilities while vaults may also be employed in independent facilities. Evaluation is made for these different cases. In in-plant facilities, dry cask storage is found to be more economical in all cases than water pool storage, especially when large-sized casks are employed. In independent facilities, on the other hand, the use of vaults is the most desirable because the required capital expenditure is the lowest due to the effect of scale economics. Dry cask storage is less expensive than water pool storage also in independent facilities. The annual discount rate has relatively small influence on the unit cost for storage. An estimated unit cost for storage in independent storage facilities is shown separately for facilities with a capacity of 1,000 tons, 3,000 tons or 5,000 tons. The report also outlines the economics of spent fuel storage in overseas facilities (Finland, Sweden and U.S.A.). (Nogami, K.)

HEU and Leu FueL Shielding Comparative Study Applied for Spent Fuel Transport

International Nuclear Information System (INIS)

Margeanu, C.A.; Margeanu, S.; Barbos, D.

2009-01-01

INR Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies. The dual core concept involves the operation of a 14 MW TRIGA steady-state, high flux research and material testing reactor at one end of a large pool, and the independent operation of an annular-core pulsing reactor (TRIGA-ACPR) at the other end of the pool. The steady-state reactor is mostly used for long term testing of power reactor fuel components (pellets, pins, subassemblies and fuel assemblies) followed by post-irradiation examination. Following the general trend to replace the He fuel type (High Enriched Uranium) by Leu fuel type (Low Enriched Uranium), in the light of international agreements between IAEA and the states using He fuel in their nuclear reactors, Inr Past's have been accomplished the TRIGA research reactor core full conversion on May 2006. The He fuel repatriation in US in the frame of Foreign Research Reactor Spent Nuclear Fuel Return Programme effectively started in 1999, the final stage being achieved in summer of 2008. Taking into account for the possible impact on the human and environment, in all activities associated to nuclear fuel cycle, the spent fuel or radioactive waste characteristics must be well known. Shielding calculations basic tasks consist in radiation doses calculation, in order to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper is a comparative study of Leu and He fuel utilization effects for the shielding analysis during spent fuel transport. A comparison against the measured data for He spent fuel, available from the last stage of the spent fuel repatriation, is presented. All the geometrical and material data related on the spent fuel shipping cask were considered according to the Nac-Lt Cask approved model. The shielding analysis estimates radiation doses to shipping cask wall surface

Concept study: Use of grout vaults for disposal of long-length contaminated equipment

International Nuclear Information System (INIS)

Clem, D.K.

1994-01-01

Study considers the potential for use of grout vaults for disposal of untreated long length equipment removed from waste tanks. Looks at ways to access vaults, material handling, regulatory aspects, and advantages and disadvantages of vault disposal

Heat-transfer analysis of the existing HEU and proposed LEU cores of Pakistan research reactor

International Nuclear Information System (INIS)

Khan, L.A.; Nabbi, R.

1987-02-01

In connection with conversion of Pakistan Research Reactor (PARR) from the use of Highly Enriched Uranium (HEU) fuel to the use of Low Enriched Uranium (LEU) fuel, steady-state thermal hydraulic analysis of both existing HEU and proposed LEU cores has been carried out. Keeping in mind the possibility of power upgrading, the performance of proposed LEU core, under 10 MW operating conditions, has also been evaluated. Computer code HEATHYD has been used for this purpose. In order to verify the reliability of the code, IAEA benchmark 2 MW reactor was analyzed. The cooling parameters evaluated include: coolant velocity, critical velocity, pressure drop, temperature distribution in the core, heat fluxes at onset of nucleate boiling, flow instability and burnout and corresponding safety margins. From the results of the study it can be concluded that the conversion of the core to LEU fuel will result in higher safety margins, as compared to existing HEU core, mainly because the increased number of fuel plates in the proposed design will reduce the average heat flux significantly. Anyhow upgrading of the reactor power to 10 MW will need the flow rate to be adjusted between 850 to 900 m 3 /hr, to achieve reasonable safety margins, at least, comparable with the existing HEU core. (orig.)

Geometric principles in additive systems for construction of vaults

DEFF Research Database (Denmark)

Jaeger, Thomas Arvid

2015-01-01

della Fiore. The Barouque style followed with the domes of Borromini; Bernini and Guarini. The Period of Enlightenment had iconic tunnel- and dome vault projects by Etienne l. Boulée. All these stone vaults were made by additive systems in bricks and mortar. In the 20.th century concrete was introduced...

Vaginal vault recurrences of endometrial cancer in non-irradiated patients

DEFF Research Database (Denmark)

Hardarson, Hordur Alexander; Nyhøj Heidemann, Lene; Christensen, René dePont

2015-01-01

are few and limited to previously irradiated patients or patients with advanced disease. Investigation of surgical treatment for isolated vaginal vault recurrence is practically nonexistent. The aim of this study is to evaluate the efficacy of RT and ST in a non-irradiated group with recurrent endometrial...... cancer limited to the vaginal vault. METHODS: Patients treated for recurrent endometrial cancer at Odense University Hospital, Denmark between 2003 and 2012 were identified, n = 118. Thirty-three patients had an isolated vaginal vault recurrence and were treated with either RT, ST or both. Re...

A mixed core conversion study with HEU and LEU fuels

International Nuclear Information System (INIS)

Matos, J.E.; Freese, K.E.

1985-01-01

The results of a mixed core study are presented for gradual replacement of HEU fuel with LEU fuel using the IAEA generic 10 MW reactor as an example. The key parameters show that the transition can be accomplished safely and economically. (author)

Performance of an active well coincidence counter for HEU samples

International Nuclear Information System (INIS)

Ferrari, Francesca; Peerani, Paolo

2010-01-01

Neutron coincidence counting is the reference NDA technique used in nuclear safeguards to measure the mass of nuclear material in samples. For high-enriched uranium (HEU) samples active neutron interrogation is generally performed and the most common device used by nuclear inspectors is the Active Well Coincidence Counter (AWCC). Within her master thesis at the Polytechnic of Milan, the first author performed an intensive study on the characteristics and performances of the AWCC in order to assess the 235 U mass in HEU oxide samples at the PERLA laboratory of JRC. The work has been summarised in this paper that starts with the optimisation of the use of AWCC for nuclear safeguards, describing the calibration procedure, reporting results of a series of verification measurements, summarising the performances that can be obtained with this instruments during inspections at fuel production plants and concluding with the discussion of uncertainties related to these measurements.

Selection of concepts for monitored retrievable storage of spent nuclear fuel and high-level radioactive wastes

International Nuclear Information System (INIS)

1984-04-01

The monitored retrievable storage (MRS) concepts considered are: metal cask (stationary and transportable); concrete cask (sealed storage cask); concrete cask-in-trench; field drywell; tunnel drywell; open cycle vault; closed cycle vault; and tunnel rack vault. These concepts were compared primarily upon the relative performance of the storage units on seven criteria which together encompass the key considerations for selecting an MRS concept, namely their ability to satisfy the MRS mission requirements. These criteria were: safety and licensing; environmental impacts; socioeconomic impacts; siting requirements; cost; concept maturity; and flexibility. Evaluations of the candidate concepts indicate that all of the concepts could satisfactorily serve in an MRS facility. However, using the above criteria, the two concepts selected for further design studies are the concrete cask (primary concept), better entitled the sealed storage cask, and the field drywell (alternate concept). It was recognized that the transportable metal storage cask may be used to supplement at-reactor storage until such time as the repository or MRS becomes available. Consequently, a hybrid storage facility may be required (e.g., one using concrete casks or field drywells, with the capability of receiving and storing the transportable cask). Both the concrete cask and the field drywell concepts can easily accommodate the transportable cask. Further design efforts will ensure the compatibility of the MRS designs with the transportable cask

What the difference to use LEU and HEU fuel elements separately or together in a research reactor

International Nuclear Information System (INIS)

Kaya, S.; Uestuen, G.

2005-01-01

Concerning of nuclear material safety, most of the research reactors are advised to shift from HEU (high enriched-%93 U-235) to LEU (low enriched-%20 U-235) fuel elements. When LEU and HEU fuel elements are to be used together in a research reactor, some design and safety problems are encountered. According to use of the reactor, some research reactors such as MTR type may not show any considerable difference for HEU or LEU fuel elements, but the efficiency of radioisotope production generated by thermal neutron interaction may decrease about twenty-thirty percent when LEU fuel elements are used. Here, fine mesh-sized 3D neutronic analysis of TR-2 research reactor is presented to indicate the arising problem when LEU end HEU fuel elements are used together in a research reactor. Partial thermohydraulic analysis of the reactor is also given to show the betterness of the LEU fuel element design. However, there might be some points that should be noticed for safer operation of plate type fuelled research reactors. (author)

Status of HEU-LEU conversion of FRJ-2

International Nuclear Information System (INIS)

Damm, G.; Nabbi, R.

2002-01-01

The operator of the German FRJ-2 research reactor, 'Research Center Juelich', has participated from the beginning in the RERTR programme and made comprehensive contributions to the test and use of LEU fuel for HEU-LEU-conversion measures. The originally planned time scale for the conversion of FRJ-2 was significantly delayed because of a change of the manufacturer of the LEU fuel elements and a 4 years shutdown of the reactor for refurbishment purposes. In the meantime the new LEU fuel elements are qualified and tested in the reactor. In the moment calculations for the safety report are made and it is planned to apply for the license of FRJ-2 operation with LEU fuel at the beginning of 2003. In order to get most reliable results a sophisticated computational method based on a MCNP model coupled with the depletion code BURN was developed for reactor physical calculations, core conversion studies and fuel element performance analysis and applied to the mixed and LEU core. The licensing schedule and results of latest calculations for the conversion study will be presented. The simulations shows that the thermal flux in the LEU core is about 19% resulting in a lower burnup rate. But in the reflector area around the core and in the center of the cold n source the neutron flux reduction remains limited to 6%. Due to a harder neutron spectrum in the LEU core the kinetic and safety related parameters are slightly reduced. Using the ORIGEN code it could be shown that the increase of the total fission products inventory amounts to about 6% compared to a HEU core. As a consequence of the high amount of U-238, the amount of U-235 in the LEU core has to be about 27% higher than in the HEU core but the U-235 burnup is approx. 5% lower due to the contribution of fissile plutonium. (author)

Comparison of HEU and LEU neutron spectra in irradiation facilities at the Oregon State TRIGA® Reactor

International Nuclear Information System (INIS)

Schickler, R.A.; Marcum, W.R.; Reese, S.R.

2013-01-01

Highlights: • The Oregon State TRIGA ® Reactor neutron spectra is characterized herein. • Neutron spectra between highly enriched uranium and low enriched uranium cores are compared. • Discussion is given as to differences between HEU and LEU core spectra results and impact on experiments. -- Abstract: In 2008, the Oregon State TRIGA ® Reactor (OSTR) was converted from highly enriched uranium (HEU) fuel lifetime improvement plan (FLIP) fuel to low-enriched uranium (LEU) fuel. This effort was driven and supported by the Department of Energy's (DoE's) Reduced Enrichment for Research and Test Reactors (RERTR) program. The basis behind the RERTR program's ongoing conversion effort is to reduce the nuclear proliferation risk of civilian research and test reactors. The original intent of the HEU FLIP fuel was to provide fuel to research reactors that could be utilized for many years before a necessary refueling cycle. As a research reactor, the OSTR provides irradiation facilities for a variety of applications, such as activation analysis, fission-track dating, commercial isotope production, neutron radiography, prompt gamma characterization, and many others. In order to accurately perform these research functions, several studies had been conducted on the HEU FLIP fuel core to characterize the neutron spectra in various experimental facilities of the OSTR (Tiyapun, 1997; Ashbaker, 2005). As useful as these analyses were, they are no longer valid due to the change in fuel composition and the resulting alteration of core performance characteristics. Additionally, the core configuration (fuel reconfiguration) was altered between the HEU and LEU cores. This study characterizes the neutron spectra in various experimental facilities within and around the current LEU core. It also compares the spectra to that which was yielded in the HEU core through use of Monte Carlo n-Particle 5 (MCNP5) and experimental adjustment via a least-squares technique. The quantification of

Comparison of thermohydraulic and nuclear aspects in a standard HEU core and a typical LEU core for the HFR Petten. A case study

International Nuclear Information System (INIS)

Pruimboom, H.; Tas, A.

1985-01-01

Within the framework of the RERTR program various HEU-LEU core calculations have been performed by ANL in a cooperative effort with ECN and JRC Petten. The main purpose of this work has been to gain competence in analysing HEU-LEU core conversion for high power Materials Testing Reactors and to assist in a possible HEU-LEU conversion of the HFR Petten. For reference purposes the present HFR standard core (HEU) in the 'old' vessel geometry was calculated at first. As a next step the new vessel geometry and the increased fuel weights were taken into account. Subsequently various LEU HFR core options have been analysed. Main parameters in the LEU study were the uranium loading in the meat, the fuel type, the thickness of the meat, the number of fuel plates per element and the type of burnable poison applied. Though the study has not yet been completed, one of its striking preliminary results concerns the increased power peaking in the LEU fuel elements as compared with the HEU situation. A preliminary analysis of the thermal characteristics of a typical LEU core as compared with a standard HEU core has been made and is presented in the paper. A short survey of the various HEU and LEU calculations is given. The thermal safety analysis procedure for the HFR, as based on the flow instability criterion, is clarified. Finally, the thermal comparison HEU versus LEU and the resulting conclusions are presented. (author)

Seismic review of vault for MFTF upgrade project

International Nuclear Information System (INIS)

Franklin, H.A.

1983-01-01

This letter report was prepared in accordance with the scope of work for the preconceptual seismic evaluation of the (α + T) Tandem Mirror Fusion Machine concrete vault. The scope of the work was developed with the assistance of the Bechtel site representative Dr. Sunil Ghose. The report contains comments and preconceptual recommendations on wall upgrading for an 150-ton crane installation, concrete vault seismic capability for (α + T) conditions, and recommendations for future work

Mechanical design of the storage tubes in the HWVP canister storage building

International Nuclear Information System (INIS)

Divona, C.J.; Fages, R.; Janicek, G.P.; Mullally, J.A.

1993-01-01

Canisters of high-level waste from the Hanford Waste Vitrification Plant (HWVP) will be stored in an adjacent facility, the Canister Storage Building (CSB). The canisters are stored vertically in an array of tubes within the shielded vault area of the CSB. This paper describes the mechanical design of the storage tubes, the shield floor plugs that confine the waste within the tubes and the impact absorber system used to assure that the canisters are not breached in the event of an accidental drop. Installation and testing of the components is also discussed

Effect of materials and manufacturing on the bending stiffness of vaulting poles

Science.gov (United States)

Davis, C. L.; Kukureka, S. N.

2012-09-01

The increase in the world record height achieved in pole vaulting can be related to the improved ability of the athletes, in terms of their fitness and technique, and to the change in materials used to construct the pole. For example in 1960 there was a change in vaulting pole construction from bamboo to glass fibre reinforced polymer (GFRP) composites. The lighter GFRP pole enabled the athletes to have a faster run-up, resulting in a greater take-off speed, giving them more kinetic energy to convert into potential energy and hence height. GFRP poles also have a much higher failure stress than bamboo, so the poles were engineered to bend under the load of the athlete, thereby storing elastic strain energy that can be released as the pole straightens, resulting in greater energy efficiency. The bending also allowed athletes to change their vaulting technique from a style that involved the body remaining almost upright during the vault to one where the athlete goes over the bar with their feet upwards. Modern vaulting poles can be made from GFRP and/or carbon fibre reinforced polymer (CFRP) composites. The addition of carbon fibres maintains the mechanical properties of the pole, but allows a reduction in the weight. The number and arrangement of the fibres determines the mechanical properties, in particular the bending stiffness. Vaulting poles are also designed for an individual athlete to take into account each athlete’s ability and physical characteristics. The poles are rated by ‘weight’ to allow athletes to select an appropriate pole for their ability. This paper will review the development of vaulting poles and the requirements to maximize performance. The properties (bending stiffness and pre-bend) and microstructure (fibre volume fraction and lay-up) of typical vaulting poles will be discussed. Originally published as Davis C L and Kukureka S N (2004) Effect of materials and manufacturing on the bending stiffness of vaulting poles The Engineering of

Meeting the regulatory challenges of mixed waste storage and monitoring: A novel approach

International Nuclear Information System (INIS)

Wilkinson, Dennis; Shaw, Mark

1992-01-01

This paper describes an original approach to providing safe storage of Remote Handled TRU Mixed Waste that is required to meet the EPA double liner and leachate collection system standards. This system, known as the 'Environmental Vault Liner', also allows a cost effective means of complying with the EPA's inspection requirements per 40 CFR 265.170, Use and Management of Containers. This approach is modular in nature, allowing additional storage capacity to be added on a demand basis, thereby eliminating significant upfront costs associated with large storage facilities built on estimated needs over many years. It reduces the financial and technical risks associated with large storage construction projects, allows modifications to new Liners put into service based on changing regulations and technologies. The Environmental Vault Liner offers additional benefits including easy waste retrieval, a 300 year design life, continuous below ground liquid detection and monitoring, replaceable instrumentation, inert (Nitrogen) atmosphere for container storage, continuous air monitoring, and remote visual container inspections. (author)

On the importance of ending the use of HEU in the nuclear fuel cycle: An updated assessment

International Nuclear Information System (INIS)

Glaser, Alexander; Hippel, Frank von

2002-01-01

The events of September 2001 have created a renewed urgency with regard to the disposition and future use and management of nuclear-weapons-usable materials. Highly enriched uranium (HEU) has received particular attention because it is relatively easy to use in a nuclear weapon and therefore an obvious candidate for diversion or theft by state or nonstate actors. The role of the RERTR program in this context and its contribution to global security can hardly be overemphasized. This article reviews existing or proposed activities to reduce the threat posed by HEU, how these activities are linked to the RERTR program, and outlines the most urgent steps to be taken to approach the ultimate objective of eliminating non-weapons HEU inventories in the world. (author)

Borated concrete for ZPPR fuel storage

International Nuclear Information System (INIS)

Gasidlo, J.M.

1985-01-01

Fuel handling at the Zero Power Plutonium Reactor (ZPPR) led to two requirements for storage of ZPPR fuel: a low neutron multiplication and shielded storage to minimize personnel doses. Boron-poisoned concrete was chosen as the storge medium with boron frit as the poisoning agent. The calculated effects of water content and boron concentration led to specifying a concrete with a water content that was higher than ordinary concrete. The finite size of the boron frit particles caused concern about reduced effectiveness due to self-shielding. The self-shielding was evaluated using optical path lengths for spheres and tabulated self-shielding for slabs. The results showed that the finite-sized particles were at least 80% as effective as infinitely dilute absorption. Neutron and gamma dose rates measured in the vault verified that personnel could work in the vault on a regular basis without exceeding personnel dose limits. 4 refs., 3 figs., 7 tabs

The Acoustics of the Double Elliptical Vault of the Royal Palace of Caserta (Italy

Directory of Open Access Journals (Sweden)

Umberto Berardi

2017-03-01

Full Text Available This work investigates the acoustic characteristics of the double elliptical vault, which overlooks the Grand Staircase of the Royal Palace of Caserta (Italy. The Royal Palace was built by the architect Luigi Vanvitelli in the Seventeenth Century and it is the largest royal building in Italy. The double elliptical vault presents a great scenography effect. Inside the vault, on the planking level, musicians used to play for the king and his guests when the royal procession, going up the grand staircase, entered the royal apartments, creating astonishment among the guests who heard the music without understanding from where it was coming. Since the musicians were inside the vault, the long reverberation made the listeners perceive the vault to be enveloped by the music. To investigate this effect, the acoustic characteristics of the double vault were measured, putting the sound source on the planking level of the vault, while the microphones were put along the staircase and in the vestibule towards the royal apartments. Finally, the spatial distribution of several acoustic parameters is evaluated also using architectural acoustic simulations.

NGLW RCRA Storage Study

International Nuclear Information System (INIS)

Waters, R.J.; Ochoa, R.; Fritz, K.D.; Craig, D.W.

2000-01-01

The Idaho Nuclear Technology and Engineering Center (INTEC) at the Idaho National Engineering and Environmental Laboratory contains radioactive liquid waste in underground storage tanks at the INTEC Tank Farm Facility (TFF). INTEC is currently treating the waste by evaporation to reduce the liquid volume for continued storage, and by calcination to reduce and convert the liquid to a dry waste form for long-term storage in calcine bins. Both treatment methods and activities in support of those treatment operations result in Newly Generated Liquid Waste (NGLW) being sent to TFF. The storage tanks in the TFF are underground, contained in concrete vaults with instrumentation, piping, transfer jets, and managed sumps in case of any liquid accumulation in the vault. The configuration of these tanks is such that Resource Conservation and Recovery Act (RCRA) regulations apply. The TFF tanks were assessed several years ago with respect to the RCRA regulations and they were found to be deficient. This study considers the configuration of the current tanks and the RCRA deficiencies identified for each. The study identifies four potential methods and proposes a means of correcting the deficiencies. The cost estimates included in the study account for construction cost; construction methods to minimize work exposure to chemical hazards, radioactive contamination, and ionizing radiation hazards; project logistics; and project schedule. The study also estimates the tank volumes benefit associated with each corrective action to support TFF liquid waste management planning

NGLW RCRA Storage Study

Energy Technology Data Exchange (ETDEWEB)

R. J. Waters; R. Ochoa; K. D. Fritz; D. W. Craig

2000-06-01

The Idaho Nuclear Technology and Engineering Center (INTEC) at the Idaho National Engineering and Environmental Laboratory contains radioactive liquid waste in underground storage tanks at the INTEC Tank Farm Facility (TFF). INTEC is currently treating the waste by evaporation to reduce the liquid volume for continued storage, and by calcination to reduce and convert the liquid to a dry waste form for long-term storage in calcine bins. Both treatment methods and activities in support of those treatment operations result in Newly Generated Liquid Waste (NGLW) being sent to TFF. The storage tanks in the TFF are underground, contained in concrete vaults with instrumentation, piping, transfer jets, and managed sumps in case of any liquid accumulation in the vault. The configuration of these tanks is such that Resource Conservation and Recovery Act (RCRA) regulations apply. The TFF tanks were assessed several years ago with respect to the RCRA regulations and they were found to be deficient. This study considers the configuration of the current tanks and the RCRA deficiencies identified for each. The study identifies four potential methods and proposes a means of correcting the deficiencies. The cost estimates included in the study account for construction cost; construction methods to minimize work exposure to chemical hazards, radioactive contamination, and ionizing radiation hazards; project logistics; and project schedule. The study also estimates the tank volumes benefit associated with each corrective action to support TFF liquid waste management planning.

Cryoelectron microscopy imaging of recombinant and tissue derived vaults: localization of the MVP N termini and VPARP.

Science.gov (United States)

Mikyas, Yeshi; Makabi, Miriam; Raval-Fernandes, Sujna; Harrington, Lea; Kickhoefer, Valerie A; Rome, Leonard H; Stewart, Phoebe L

2004-11-12

The vault is a highly conserved ribonucleoprotein particle found in all higher eukaryotes. It has a barrel-shaped structure and is composed of the major vault protein (MVP); vault poly(ADP-ribose) polymerase (VPARP); telomerase-associated protein 1 (TEP1); and small untranslated RNA (vRNA). Although its strong conservation and high abundance indicate an important cellular role, the function of the vault is unknown. In humans, vaults have been implicated in multidrug resistance during chemotherapy. Recently, assembly of recombinant vaults has been established in insect cells expressing only MVP. Here, we demonstrate that co-expression of MVP with one or both of the other two vault proteins results in their co-assembly into regularly shaped vaults. Particles assembled from MVP with N-terminal peptide tags of various length are compared. Cryoelectron microscopy (cryoEM) and single-particle image reconstruction methods were used to determine the structure of nine recombinant vaults of various composition, as well as wild-type and TEP1-deficient mouse vaults. Recombinant vaults with MVP N-terminal peptide tags showed internal density that varied in size with the length of the tag. Reconstruction of a recombinant vault with a cysteine-rich tag revealed 48-fold rotational symmetry for the vault. A model is proposed for the organization of MVP within the vault with all of the MVP N termini interacting non-covalently at the vault midsection and 48 copies of MVP forming each half vault. CryoEM difference mapping localized VPARP to three density bands lining the inner surface of the vault. Difference maps designed to localize TEP1 showed only weak density inside of the caps, suggesting that TEP1 may interact with MVP via a small interaction region. In the absence of atomic-resolution structures for either VPARP or TEP1, fold recognition methods were applied. A total of 21 repeats were predicted for the TEP1 WD-repeat domain, suggesting an unusually large beta-propeller fold.

HYDRAULIC AND PHYSICAL PROPERTIES OF SALTSTONE GROUTS AND VAULT CONCRETES

International Nuclear Information System (INIS)

Dixon, K.; Harbour, J.; Phifer, M.

2008-01-01

The Saltstone Disposal Facility (SDF), located in the Z-Area of the Savannah River Site (SRS), is used for the disposal of low-level radioactive salt solution. The SDF currently contains two vaults: Vault 1 (6 cells) and Vault 4 (12 cells). Additional disposal cells are currently in the design phase. The individual cells of the saltstone facility are filled with saltstone. Saltstone is produced by mixing the low-level radioactive salt solution, with blast furnace slag, fly ash, and cement (dry premix) to form a dense, micro-porous, monolithic, low-level radioactive waste form. The saltstone is pumped into the disposal cells where it subsequently solidifies. Significant effort has been undertaken to accurately model the movement of water and contaminants through the facility. Key to this effort is an accurate understanding of the hydraulic and physical properties of the solidified saltstone. To date, limited testing has been conducted to characterize the saltstone. The primary focus of this task was to estimate the hydraulic and physical properties of three types of saltstone and two vault concretes. The saltstone formulations included saltstone premix batched with (1) Deliquification, Dissolution, and Adjustment (DDA) salt simulant (w/pm 0.60), (2) Actinide Removal Process (ARP)/Modular Caustic Side Solvent Extraction Unit (MCU) salt simulant (w/pm 0.60), and (3) Salt Waste Processing Facility (SWPF) salt simulant (w/pm 0.60). The vault concrete formulations tested included the Vault 1/4 concrete and two variations of the Vault 2 concrete (Mix 1 and Mix 2). Wet properties measured for the saltstone formulations included yield stress, plastic viscosity, wet unit weight, bleed water volume, gel time, set time, and heat of hydration. Hydraulic and physical properties measured on the cured saltstone and concrete samples included saturated hydraulic conductivity, moisture retention, compressive strength, porosity, particle density, and dry bulk density. These properties

Fusée ceramic vaults and domes in the Netherlands

NARCIS (Netherlands)

Kamerling, M.W.

2011-01-01

This paper gives an overview of the fusée Ceramique vaults and domes constructed in The Netherlands just after World War II and describes the technique of construction as well as the advantages and disadvantages. Using only a minimum of concrete and steel, the environmental load of the vaults is

Shipment of VINCA Institute's HEU fresh fuel to Russia

International Nuclear Information System (INIS)

Pesic, Milan; Sotic, Obrad

2002-01-01

This paper shows, for the first time, the basic data related to the recent shipment of the fresh HEU fuel elements from Yugoslavia back to Russia for uranium down blending. In this way, Yugoslavia gives its contribution to the RERTR program and to the world's joint efforts to prevent possible terrorist action against nuclear material potentially usable for production of nuclear weapons. (author)

The prospects for dry fuel storage

International Nuclear Information System (INIS)

Harris, G.G.; Elliott, D.

1994-01-01

Dry storage of spent nuclear fuels is one method of dealing with radioactive waste. This article reports from a one day seminar on future prospects for dry fuel storage held in November 1993. Dry storage in an inert gas or air environment in vaults or casks, is an alternative to wet storage in water-filled ponds. Both wet and dry storage form part of the Interim Storage option for radioactive waste materials, and form alternatives to reprocessing or direct disposal in a deep repository. It has become clear that a large market for dry fuel storage will exist in the future. It will therefore be necessary to ensure that the various technical, safety, commercial, legislative and political constraints associated with it can be met effectively. (UK)

Comparison of the FRM-II HEU design with an alternative LEU design. Attachment

International Nuclear Information System (INIS)

Hanan, N.A.; Mo, S.C.; Smith, R.S.; Matos, J.E.

2004-01-01

After presentation of the foregoing paper by Dr. Nelson Hanan of Argonne National Laboratory (ANL) proposing an alternative LEU core with one fuel ring and a power level of 33 MW, a presentation was made by Dr. Klaus Boning of the Technical University of Munich comparing the FRM-II HEU design with an LEU design by Tlm that had two fuel rings and a power level of 40 MW. Dr. Boning raised the following issues concerning the use of LEU fuel in FRM-H reactor designs: (1) qualification of HEU and LEU silicide fuels, (2) gamma heating in the heavy water reflector, (3) the radiological consequences of hypothetical accidents, and (4) cost and schedule. These issues are addressed in this Attachment. In his presentation, Dr. Hanan mentioned that ANL was also investigating other LEU designs. This work led to a second alternative LEU design that has the same neutron flux performance (8 x 10 14 n/cm 2 /s peak neutron flux in the reflector) and the same fuel lifetime (50 full power days) as the HEU design, but uses LEU silicide fuel with a uranium density of only 4.5 g/cm 3 . This design was achieved by using a fuel plate that has a fuel meat thickness of 0.76 mm, a cladding thickness of 0.38 mm, and a water channel gap of 2.2 mm. A comparison is shown of the main characteristics of this second alternative LEU design with those of the FRM-II HEU design. The ANL core again has one fuel ring with the same dimensions. With this LEU design, a two stage process is no longer necessary because LEU silicide fuel with a uranium density of 4.5 g/cm 3 is fully qualified, licensable, and available now for use in a high flux reactor such as the FRM-II

HEU age determination by the activity ratio {sup 227}Th/{sup 235}U

Energy Technology Data Exchange (ETDEWEB)

Li, Junjie; Zeng, Lina; Wu, Jian; Zheng, Chun; Li, Jiansheng, E-mail: lastljj@hotmail.com

2014-02-15

It is important to measure the age of a highly enriched uranium (HEU) assembly for authentication of the material in the frame of arms control inspections. A new non-destructive gamma spectrometric method for HEU age-dating is reported. This method relies on measuring the daughter/parent activity ratio {sup 227}Th/{sup 235}U by high-resolution gamma spectrometry. Only a narrow gamma range of energy of uranium from 230 keV to 242 keV will be used for analysis. The relative efficiency of every characteristic gamma ray changes in a small range because it has a near energy, which makes the results more accurate in theory. It provides a quick and reliable method for HEU age determination. Several gamma spectra of the same HEU assembly have been measured with different conditions (gain settings, distance and measurement time). When a branching ratio of 12.6% was chosen for the 235.96 keV line of {sup 227}Th, we obtained the activity ratios of (5.61 ± 0.40) × 10{sup −4}, (5.17 ± 0.39) × 10{sup −4}, (5.26 ± 0.39) × 10{sup −4}, (5.10 ± 0.35) × 10{sup −4}, (5.50 ± 0.44) × 10{sup −4} and (5.47 ± 0.42) × 10{sup −4}, respectively. These ratios correspond to ages of 52.2 ± 2.4 years, 49.7 ± 2.3 years, 50.1 ± 2.3 years, 49.3 ± 2.2 years, 51.6 ± 2.5 years and 51.5 ± 2.4 years, respectively, which are consistent with the known age of this material and the results of the U–Bi method.

The performance assessment and the design of an intermediate level tritium disposal vault

International Nuclear Information System (INIS)

Yu, A.D.

1991-01-01

The topic of this report is the assessment of the performance and design of the tritium disposal vault for the Westinghouse River Company at the Savannah River Laboratory. This paper describes how the groundwater modeling has affected the design of a tritium disposal vault at the Savannah River Site and this new vault will meet the regulatory performance requirements. (MB)

Automation in a material processing/storage facility

International Nuclear Information System (INIS)

Peterson, K.; Gordon, J.

1997-01-01

The Savannah River Site (SRS) is currently developing a new facility, the Actinide Packaging and Storage Facility (APSF), to process and store legacy materials from the United States nuclear stockpile. A variety of materials, with a variety of properties, packaging and handling/storage requirements, will be processed and stored at the facility. Since these materials are hazardous and radioactive, automation will be used to minimize worker exposure. Other benefits derived from automation of the facility include increased throughput capacity and enhanced security. The diversity of materials and packaging geometries to be handled poses challenges to the automation of facility processes. In addition, the nature of the materials to be processed underscores the need for safety, reliability and serviceability. The application of automation in this facility must, therefore, be accomplished in a rational and disciplined manner to satisfy the strict operational requirements of the facility. Among the functions to be automated are the transport of containers between process and storage areas via an Automatic Guided Vehicle (AGV), and various processes in the Shipping Package Unpackaging (SPU) area, the Accountability Measurements (AM) area, the Special Isotope Storage (SIS) vault and the Special Nuclear Materials (SNM) vault. Other areas of the facility are also being automated, but are outside the scope of this paper

Ohio State University Nuclear Reactor Laboratory HEU fuel shipment summary. Final

International Nuclear Information System (INIS)

1997-01-01

In November 1988, OSURR converted from HEU fuel to LEU fuel. As a result they needed to get rid of their HEU fuel by shipping it to Savannah River. The players in the fuel shipping game are: OSURR as the keeper of the fuel; DOE as the owner of fuel and shipper of record; Tri-State Motor Transit Co. for transporting the cask; Muth Brothers as the rigger responsible for getting the cask on and off the truck and in and out of the building; Hoffman LaRoche/Cintichem as the owner of the cask; Savannah River as the receiver of the fuel; and the NRC for approval of the Security Plan, QA Plan, etc. This report gives a chronological history of the events from February 1989 to June 1, 1995, the actual day of shipment. The cask was received at Savannah River on June 2, 1995

Fuel element burnup determination in HEU-LEU mixed TRIGA research reactor core

International Nuclear Information System (INIS)

Zagar, Tomaz; Ravnik, Matjaz

2000-01-01

This paper presents the results of a burnup calculations and burnup measurements for TRIGA FLIP HEU fuel elements and standard TRIGA LEU fuel elements used simultaneously in small TRIGA Mark II research reactor in Ljubljana, Slovenija. The fuel element burnup for approximately 15 years of operation was calculated with two different in house computer codes TRIGAP and TRIGLAV (both codes are available at OECD NEA Data Bank). The calculation is performed in one-dimensional radial geometry in TRIGAP and in two-dimensional (r,φ) geometry in TRIGLAV. Inter-comparison of results shows important influence of in-core water gaps, irradiation channels and mixed rings on burnup calculation accuracy. Burnup of 5 HEU and 27 LEU fuel elements was also measured with reactivity method. Measured and calculated burnup values are inter-compared for these elements (author)

Desalination of painted brick vaults

DEFF Research Database (Denmark)

Larsen, Poul Klenz

The subject of the thesis is salt and moisture movement that causes damage to wall paintings on church vaults. The deterioration was studied in the churches of Fanefjord, Kirkerup and Brarup. A desalination method was tested om location. The salt and moisture transfer was examined in detail...

Technical investigation of a pyrophoric event involving corrosion products from HEU ZPPR fuel plates

International Nuclear Information System (INIS)

Totemeier, T. C.

2000-01-01

A pyrophoric event recently occurred which involved corrosion products collected from highly-enriched uranium (HEU) fuel plates used in the Zero Power Physics Reactor (ZPPR). This paper summarizes the event and its background, and presents the results of an investigation into its source and mechanism. The investigation focused on characterization of corrosion product samples similar to those involved in the event using thermo-gravimetric analysis (TGA). Burning curve TGA tests were performed to measure the ignition temperature and hydride fractions of corrosion products in several different conditions to assess the effects of passivation treatment and long-term storage on chemical reactivity. The hydride fraction and ignition temperature of the corrosion products were found to be strongly dependent on the corrosion extent of the source metal. The results indicate that the energy source for the event was a considerable quantity of uranium hydride present in the corrosion products, but the specific ignition mechanism could not be identified

Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

Energy Technology Data Exchange (ETDEWEB)

Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

2006-01-01

The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

Orsphere: Physics Measurments For Bare, HEU(93.2)-Metal Sphere

Energy Technology Data Exchange (ETDEWEB)

Marshall, Margaret A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Briggs, J. Blair [Idaho National Lab. (INL), Idaho Falls, ID (United States); White, Christine E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dyrda, James P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tancock, Nigel P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mihalczo, John [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-03-01

In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an attempt to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s (HEU-MET-FAST-001). The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. “The very accurate description of this sphere, as assembled, establishes it as an ideal benchmark for calculational methods and cross-section data files” (Reference 1). While performing the ORSphere experiments care was taken to accurately document component dimensions (±0.0001 inches), masses (±0.01 g), and material data. The experiment was also set up to minimize the amount of structural material in the sphere proximity. Two, correlated spheres were evaluated and judged to be acceptable as criticality benchmark experiments. This evaluation is given in HEU-MET-FAST-100. The second, smaller sphere was used for additional reactor physics measurements. Worth measurements (Reference 1, 2, 3 and 4), the delayed neutron fraction (Reference 3, 4 and 5) and surface material worth coefficient (Reference 1 and 2) are all measured and judged to be acceptable as benchmark data. The prompt neutron decay (Reference 6), relative fission density (Reference 7) and relative neutron importance (Reference 7) were measured, but are not evaluated. Information for the evaluation was compiled from References 1 through 7, the experimental logbooks 8 and 9 ; additional drawings and notes provided by the experimenter; and communication with the lead experimenter, John T. Mihalczo.

Neutronic analysis of the conversion of HEU to LEU fuel for a 5-MW MTR core

International Nuclear Information System (INIS)

Pazirandeh, A.; Bartsch, G.

1987-01-01

In recent years, due to cessation of highly enriched uranium (HEU) fuel supply, practical steps have been taken to substitute HEU fuel in almost all research reactors by medium-enriched uranium or low-enriched uranium (LEU) fuels. In this study, a neutronic calculation of a 5-MW research reactor core fueled with HEU (93% 235 U) is presented. In order to assess the performance of the core with the LEU ( 235 U loadings were examined. The core consists of 22 standard fuel elements (SFEs) and 6 control fuel elements (CFEs). Each fuel elements has 18 curved plates of which two end plates are dummies. Initial 235 U content is 195 g 235 U/SFE and 9.7 g 235 U/CFE or /PFE. In all calculations the permitted changes to the fuel elements are (a) 18 active plates per SFE, (b) fuel plates assumed to be flat, and (c) 8 or 9 active plates per CFE

Composite analysis E-area vaults and saltstone disposal facilities

Energy Technology Data Exchange (ETDEWEB)

Cook, J.R.

1997-09-01

This report documents the Composite Analysis (CA) performed on the two active Savannah River Site (SRS) low-level radioactive waste (LLW) disposal facilities. The facilities are the Z-Area Saltstone Disposal Facility and the E-Area Vaults (EAV) Disposal Facility. The analysis calculated potential releases to the environment from all sources of residual radioactive material expected to remain in the General Separations Area (GSA). The GSA is the central part of SRS and contains all of the waste disposal facilities, chemical separations facilities and associated high-level waste storage facilities as well as numerous other sources of radioactive material. The analysis considered 114 potential sources of radioactive material containing 115 radionuclides. The results of the CA clearly indicate that continued disposal of low-level waste in the saltstone and EAV facilities, consistent with their respective radiological performance assessments, will have no adverse impact on future members of the public.

Composite analysis E-area vaults and saltstone disposal facilities

International Nuclear Information System (INIS)

Cook, J.R.

1997-09-01

This report documents the Composite Analysis (CA) performed on the two active Savannah River Site (SRS) low-level radioactive waste (LLW) disposal facilities. The facilities are the Z-Area Saltstone Disposal Facility and the E-Area Vaults (EAV) Disposal Facility. The analysis calculated potential releases to the environment from all sources of residual radioactive material expected to remain in the General Separations Area (GSA). The GSA is the central part of SRS and contains all of the waste disposal facilities, chemical separations facilities and associated high-level waste storage facilities as well as numerous other sources of radioactive material. The analysis considered 114 potential sources of radioactive material containing 115 radionuclides. The results of the CA clearly indicate that continued disposal of low-level waste in the saltstone and EAV facilities, consistent with their respective radiological performance assessments, will have no adverse impact on future members of the public

Transition from HEU to LEU fuel in Romania's 14-MW TRIGA reactor

International Nuclear Information System (INIS)

Bretscher, M.M.; Snelgrove, J.L.

1995-01-01

The 14-MW TRIGA steady state reactor (SSR) located in Pitesti, Romania, first went critical in the fall of 1979. Initially, the core configuration for full power operation used 29 fuel clusters each containing a 5 x 5 square array of HEU U (10 wt% - ZrH - Er 2.8 wt%) fuel-moderator rods (1.295 cm o.d.) clad in Incoloy. With a total inventory of 35 HEU fuel clusters, burnup, considerations required a gradual expansion of the core from 29 to 32 and finally to 35 clusters before the reactor was shut down because of insufficient excess reactivity. At this time each of the original 29 fuel clusters had an average 235 U burnup in the range from 50 to 62%. Because of the U.S. policy regarding the export of highly enriched uranium, fresh HEU TRIGA replacement fuel is not available. After a number of safety-related measurements, the SSR is expected to resume full power operation in the near future using a mixed core containing five LEU TRIGA clusters of the same geometry as the original fuel but with fuel-moderator rods containing 45 wt% U (19.7% 235 U enrichment) and 1.1 wt% Er. Rods for 14 additional LEU fuel clusters will be fabricated by General Atomics. In support of the SSR mixed core operation numerous neutronic calculations have been performed. This paper presents some of the results of those calculations. (author)

Comparison of HEU and LEU neutron spectra in irradiation facilities at the Oregon State TRIGA{sup ®} Reactor

Energy Technology Data Exchange (ETDEWEB)

Schickler, R.A., E-mail: robert.schickler@oregonstate.edu; Marcum, W.R., E-mail: wade.marcum@oregonstate.edu; Reese, S.R.

2013-09-15

Highlights: • The Oregon State TRIGA{sup ®} Reactor neutron spectra is characterized herein. • Neutron spectra between highly enriched uranium and low enriched uranium cores are compared. • Discussion is given as to differences between HEU and LEU core spectra results and impact on experiments. -- Abstract: In 2008, the Oregon State TRIGA{sup ®} Reactor (OSTR) was converted from highly enriched uranium (HEU) fuel lifetime improvement plan (FLIP) fuel to low-enriched uranium (LEU) fuel. This effort was driven and supported by the Department of Energy's (DoE's) Reduced Enrichment for Research and Test Reactors (RERTR) program. The basis behind the RERTR program's ongoing conversion effort is to reduce the nuclear proliferation risk of civilian research and test reactors. The original intent of the HEU FLIP fuel was to provide fuel to research reactors that could be utilized for many years before a necessary refueling cycle. As a research reactor, the OSTR provides irradiation facilities for a variety of applications, such as activation analysis, fission-track dating, commercial isotope production, neutron radiography, prompt gamma characterization, and many others. In order to accurately perform these research functions, several studies had been conducted on the HEU FLIP fuel core to characterize the neutron spectra in various experimental facilities of the OSTR (Tiyapun, 1997; Ashbaker, 2005). As useful as these analyses were, they are no longer valid due to the change in fuel composition and the resulting alteration of core performance characteristics. Additionally, the core configuration (fuel reconfiguration) was altered between the HEU and LEU cores. This study characterizes the neutron spectra in various experimental facilities within and around the current LEU core. It also compares the spectra to that which was yielded in the HEU core through use of Monte Carlo n-Particle 5 (MCNP5) and experimental adjustment via a least

A programme for Euratom safeguards inspectors, used in the assay of high enriched (H.E.U.) and low enriched (L.E.U.) uranium fuel materials by active neutron interrogation

International Nuclear Information System (INIS)

Vocino, V.; Farese, N.; Maucq, T.; Nebuloni, M.

1991-01-01

The programme AECC (Active Euratom Coincidence Counters) has been developed at the Joint Research Center, Ispra by the Euratom Safeguards Directorate, Luxembourg and the Safety Technology Institute, Ispra for the acquisition, evaluation, management and storage of measurement data originating from active neutron interrogation of HEU and LEU fuel materials. The software accommodates the implementation of the NDA (Non Destructive Assay) procedures for the Active Well Coincidence Counters and Active Neutron Coincidence Counters deployed by the Euratom Safeguards Directorate, Luxembourg

Seismic Behaviour of Masonry Vault-Slab Structures

International Nuclear Information System (INIS)

Chesi, Claudio; Butti, Ferdinando; Ferrari, Marco

2008-01-01

Spandrel walls typically play a structural role in masonry buildings, transferring load from a slab to the supporting vault. Some indications are given in the literature on the behaviour of spandrels under the effect of vertical loads, but little attention is given to the effect coming from lateral forces acting on the building. An opportunity to investigate this problem has come from the need of analyzing a monumental building which was damaged by the Nov. 24, 2004 Val Sabbia earthquake in the north of Italy. The finite element model set up for the analysis of the vault-spandrel-slab system is presented and the structural role resulting for the spandrels is discussed

Dry storage of irradiated nuclear fuels and vitrified wastes

International Nuclear Information System (INIS)

Deacon, D.

1982-01-01

A review is given of the work of GEC Energy Systems Ltd. over the years in the dry storage of irradiated fuel. The dry-storage module (designated as Cell 4) for irradiated magnox fuel recently constructed at Wylfa nuclear power station is described. Development work on the long-term dry storage of irradiated oxide fuels is reported. Four different methods of storage are compared. These are the pond, vault, cask and caisson stores. It is concluded that there are important advantages with the passive air-cooled ESL dry stove. (U.K.)

Experiments for post accident hydrogen dispersion in F.M. vault using helium

International Nuclear Information System (INIS)

Bajaj, S.S.; Bhattacharyya, D.; Mishra, S.

1994-01-01

Under certain postulated accident scenarios involving a Loss of Coolant Accident (LOCA) simultaneous with impairment of Emergency Core Cooling (ECC), generation of hydrogen due to reaction between the zirconium clad and coolant is predicted in the coolant channel. The hydrogen generated in the coolant channels would eventually get released either in Fuelling Machine (FM) vault or in the pump room atmosphere depending on the location of the break. Analytical studies carried out so far to estimate the time dependent hydrogen concentration in the accident FM Vault consider the entire vault as a single volume. Tests were, therefore, planned to assess the mixing within the FM vault atmosphere with and without the availability of cooling fan units by releasing a known quantity of helium (instead of hydrogen) at selected locations and monitoring the relative concentration of helium in air at various locations. Test was conducted by releasing about 360 1 helium over a period of to 4 minutes at preselected locations and by measuring the relative concentration (leak rates indicated by helium leak detectors) at various locations in the FM vault. The results of cases with fans operating indicate repeatable and consistent trends of good mixing in the vault. For other cases (non turbulent, still condition) the results are sensitive to various factors including orientation of release. The former set of cases (turbulent. fans operating) are more relevant for postulated accident conditions. (author). 1 tab., 18 figs

Health physics challenges involved with opening a "seventeen-inch" concrete waste vault.

Science.gov (United States)

Sullivan, Patrick T; Pizzulli, Michelle

2005-05-01

This paper describes the various activities involved with opening a sealed legacy "Seventeen-inch" concrete vault and the health physics challenges and solutions employed. As part of a legacy waste stream that was removed from the former Hazardous Waste Management Facility at Brookhaven National Laboratory, the "Seventeen-inch" concrete vault labeled 1-95 was moved to the new Waste Management Facility for ultimate disposal. Because the vault contained 239Pu foils with a total activity in excess of the transuranic waste limits, the foils needed to be removed and repackaged for disposal. Conventional diamond wire saws could not be used because of facility constraints, so this project relied mainly on manual techniques. The planning and engineering controls put in place enabled personnel to open the vault and remove the waste while keeping dose as low as reasonably achievable.

Experience with the transport and storage casks CASTOR (registered) MTR 2 for spent nuclear fuel assemblies from research reactors

Energy Technology Data Exchange (ETDEWEB)

Jack, Allen; Rettenbacher, Katharina; Skrzyppek, Juergen [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

2011-07-01

The CASTOR (registered) MTR 2 cask was designed and manufactured by the company GNS during the 1990's for the transport and interim storage of spent nuclear fuel assemblies from various types of research reactors. Casks of this type have been used at the VKTA Research Centre in Rossendorf near Dresden, Germany as well as at the European Commission's Joint Research Centre at Petten and at the HOR reactor at Delft in the Netherlands. A total of 24 units have been used for the functions of transport and storage with various spent fuel types (VVER, HFR-HEU, and HOR-HEU) for more than ten years now. This type of packaging for radioactive material is a member of the CASTOR (registered) family of spent nuclear fuel casks used worldwide. Over 1000 units are loaded and in storage in Europe, Asia, Africa and North America. This paper presents the experience from the use of the casks for transport and storage in the past, as well as the prospects for the future. (author)

Analyses for inserting fresh LEU fuel assemblies instead of fresh HEU fuel assemblies in the Dalat Nuclear Research Reactor in Vietnam

International Nuclear Information System (INIS)

Hanan, N. A.; Deen, J.R.; Matos, J.E.

2005-01-01

Analyses were performed by the RERTR Program to replace 36 burned HEU (36%) fuel assemblies in the Dalat Nuclear Research Reactor in Vietnam with either 36 fresh fuel assemblies currently on-hand at the reactor or with LEU fuel assemblies to be procured. The study concludes that the current HEU (36%) WWR-M2 fuel assemblies can be replaced with LEU WWR-M2 fuel assemblies that are fully-qualified and have been commercially available since 2001 from the Novosibirsk Chemical Concentrates Plant in Russia. The current reactor configuration using re-shuffled HEU fuel began in June 2004 and is expected to allow normal operation until around August 2006. If 36 HEU assemblies each with 40.2 g 235 U are inserted without fuel shuffling over the next five operating cycles, the core could operate for an additional 10 years until June 2016. Alternatively, inserting 36 LEU fuel assemblies each containing 49.7 g 235 U without fuel shuffling over five operating cycles would allow normal operation for about 14 years from August 2006 until October 2020. The main reason for the longer service life of the LEU fuel is that its 235 U content is higher than the 235 U content needed simply to match the service life of the HEU fuel. Fast neutron fluxes in the experiment regions would be very nearly the same in both the HEU and LEU cores. Thermal neutron fluxes in the experiment regions would be lower by 1-5%, depending on the experiment type and location. (author)

Highly enriched uranium (HEU) politics: An enigma wrapped up in a warhead and boxed in political chaos

International Nuclear Information System (INIS)

Anon.

1993-01-01

It could be fairly said that while the Cold War arose in an atmosphere of mutual mistrust and hostility, it is ending with an equal amount of confusion and uncertainty. More than a year has passed since the US and Russia signed a tentative HEU agreement in August 1992. Many of the details have been worked out, but major questions remain. And they're not just on the Russian side. The fine points of President Clinton's overall nuclear policy are only now beginning to emerge. In his first major foreign policy address, before the United Nations in late September, Clinton called for a worldwide ban on the production of plutonium and HEU for nuclear weapons. open-quotes Growing global stockpiles of plutonium and highly enriched uranium are raising the danger of nuclear terrorism for all nations,close quotes said Clinton before the UN. open-quotes We will press for an international agreement that would ban production of these materials for weapons forever.close quotes As the veil lifts from Clinton's nuclear policy, it appears the Administration realizes that Russia may have more HEU than originally thought. That possibility has been confirmed by Minatom Minister Mikhailov's disclosures to the NUKEM Market Report, which brought a greater degree of certainty to estimates that had been floating around for some time. When the Bush Administration signed the HEU pact, it apparently thought the 500 metric tons comprised most of the former Soviet Union's nuclear arsenal. Now that the number appears higher, Clinton may propose to accelerate and enlarge the HEU deal. He is due to summit with Yeltsin, if Yeltsin survives, next spring. The 500-metric-ton deal may only be the first step

Special Analysis: Revision of Saltstone Vault 4 Disposal Limits (U)

Energy Technology Data Exchange (ETDEWEB)

Cook, J

2005-05-26

New disposal limits have been computed for Vault 4 of the Saltstone Disposal Facility based on several revisions to the models in the existing Performance Assessment and the Special Analysis issued in 2002. The most important changes are the use of a more rigorous groundwater flow and transport model, and consideration of radon emanation. Other revisions include refinement of the aquifer mesh to more accurately model the footprint of the vault, a new plutonium chemistry model accounting for the different transport properties of oxidation states III/IV and V/VI, use of variable infiltration rates to simulate degradation of the closure system, explicit calculation of gaseous releases and consideration of the effects of settlement and seismic activity on the vault structure. The disposal limits have been compared with the projected total inventory expected to be disposed in Vault 4. The resulting sum-of-fractions of the 1000-year disposal limits is 0.2, which indicates that the performance objectives and requirements of DOE 435.1 will not be exceeded. This SA has not altered the conceptual model (i.e., migration of radionuclides from the Saltstone waste form and Vault 4 to the environment via the processes of diffusion and advection) of the Saltstone PA (MMES 1992) nor has it altered the conclusions of the PA (i.e., disposal of the proposed waste in the SDF will meet DOE performance measures). Thus a PA revision is not required and this SA serves to update the disposal limits for Vault 4. In addition, projected doses have been calculated for comparison with the performance objectives laid out in 10 CFR 61. These doses are 0.05 mrem/year to a member of the public and 21.5 mrem/year to an inadvertent intruder in the resident scenario over a 10,000-year time-frame, which demonstrates that the 10 CFR 61 performance objectives will not be exceeded. This SA supplements the Saltstone PA and supersedes the two previous SAs (Cook et al. 2002; Cook and Kaplan 2003).

Development of spent fuel dry storage technology

International Nuclear Information System (INIS)

Maruoka, Kunio; Matsunaga, Kenichi; Kunishima, Shigeru

2000-01-01

The spent fuels are the recycle fuel resources, and it is very important to store the spent fuels in safety. There are two types of the spent fuel interim storage system. One is wet storage system and another is dry storage system. In this study, the dry storage technology, dual purpose metal cask storage and canister storage, has been developed. For the dual purpose metal cask storage, boronated aluminum basket cell, rational cask body shape and shaping process have been developed, and new type dual purpose metal cask has been designed. For the canister storage, new type concrete cask and high density vault storage technology have been developed. The results of this study will be useful for the spent fuel interim storage. Safety and economical spent fuel interim storage will be realized in the near future. (author)

Risk Management in the Original Extreme Sporting Event: The Pole Vault

Science.gov (United States)

Bemiller, Jim; Hardin, Robin

2010-01-01

The pole vault was considered the ultimate test of physical ability and daring before the advent of modern extreme sports such as skateboarding, snowboarding, and mountain biking. The inherent risks of the pole vault have been well documented. The National Center for Catastrophic Sport Injury Research reported in 2007 that the catastrophic injury…

A comparison of the radiological consequences of a HEU and LEU fueled research reactor

International Nuclear Information System (INIS)

Kollas, J.G.

1985-01-01

An analysis of the design basis accident radiological consequences of the HEU and LEU fueled Greek Research Reactor is presented. Doses and individual cancer risk from exposure to the passing radioactive cloud are estimated up to a distance of 20 km from the reactor site. Collective exposure and latent health effects are estimated for the total Athens area of 3081000 inhabitants. The results indicate that the plutonium isotopes buildup in the LEU fuel does not increase appreciably the consequences in respect to the HEU fueled reactor. The plutonium impact concerns mainly bone effects and secondly lung and whole body effects. The contribution to the limiting thyroid dose and the corresponding thyroid effects is insignificant. (author)

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

International Nuclear Information System (INIS)

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

1977-09-01

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

3D SURVEYING AND GEOMETRIC ASSESSMENT OF A GOTHIC NAVE VAULTING FROM POINT CLOUDS

Directory of Open Access Journals (Sweden)

A. Costa-Jover

2017-02-01

Full Text Available The development of massive data captures techniques (MDC in recent years, such as the Terrestrial laser Scanner (TLS, raises the possibility of developing new assessment procedures for architectural heritage. The 3D models that it is able to obtain is a great potential tool, both for conservation purposes and for historical and architectural studies. The paper proposes a simple, non-invasive methodology for the assessment of masonry vaults from point clouds which makes it possible to obtain relevant data about the formal anomalies. The methodology is tested in Tortosa’s Gothic Cathedral’s vaults, where the geometrical differences between vaults, a priori equal, are identified and related with the partially known construction phases. The procedure can be easily used on any other vaulted construction of any kind, but is especially useful to deal with the complex geometry of Gothic masonry vaults.

Nitrate Diffusional Releases from the Saltstone Facility, Vault 2, with Respect to Different Concrete Wall Thicknesses

International Nuclear Information System (INIS)

ROBERT, HIERGESELL

2005-01-01

To assist the Saltstone Vault 2 Design Team, an investigation was conducted to evaluate the effectiveness of alternative concrete wall thicknesses in limiting nitrate diffusion away from the planned facility. While the current design calls for 18-inch concrete walls, alternative thicknesses of 12-in, 8-in, and 6-in were evaluated using a simplified 1-D numerical model. To serve as a guide for Saltstone Vault 2 conceptual design, the results of this investigation were applied to Saltstone Vault 4 to determine what the hypothetical limits would be for concrete wall thicknesses thinner than the planned 18-inches. This was accomplished by adjusting the Vault 4 Limits, based on the increased nitrate diffusion rates through the thinner concrete walls, such that the 100-m well limit of 44 mg/L of nitrate as nitrate was not exceeded. The implication of these preliminary results is that as thinner vault walls are implemented there is a larger release of nitrate, thus necessitating optimal vault placement to minimize the number of vaults placed along a single groundwater flow path leading to the discharge zone

Operation of automated NDA instruments for in-line HEU accounting at Y-12

International Nuclear Information System (INIS)

Russo, P.A.; Strittmatter, R.B.; Sandford, E.L.; Jeter, I.W.; McCullough, E.; Bowers, G.L.

1983-01-01

Two automated nondestructive assay instruments developed at Los Alamos in support of nuclear materials accounting needs are currently operating in-line at the Y-12 Plant for recovery of highly enriched uranium. One instrument provides the HEU inventory in the secondary solvent extraction system, and the other monitors HEU concentration in the secondary intermediate evaporator. Both instruments were installed in December 1982. Operational evaluation of these instruments has been a joint effort of Y-12 and Los Alamos. This has included comparison of the solvent extraction system inventories with direct measurement performed on the dumped solution components of the solvent extraction system, as well as comparisons of concentration assay results with the external assays of samples withdrawn from the process. The function, design, and preliminary results of the operational evaluation are reported

A preliminary assessment of temperature distributions associated with a radioactive waste vault

International Nuclear Information System (INIS)

Tammemagi, H.Y.

1978-09-01

The temperature distributions of models which simulated radioactive waste vaults were determined, using a finite difference computer code to solve the transient heat conduction equation. Input parameters to the code included thermal properties for granitic rock and heat generation decay data for wastes that would be separated from CANDU fuel if it were recycled. Due to the preliminary nature of the study, only simple models were analysed. A disc source was utilized to approximate a one-level repository. Various parameters were investigated such as depth of disc, thermal properties of rock, and long-term effects. It was shown that, for a vault at 500 m depth with an initial areal heat flux of 31 W/m 2 , a maximum temperature increase of about 80 deg C occurs at the vault level about 30 years after waste emplacement; maximum increases near the earth's surface occur after about 1000 years and are less than 1 deg C. Modelling the vault by a number of vertical waste boreholes on one horizontal level, instead of by a disc, with the gross areal heat flux again 31 W/m 2 , did not cause serious local temperature increases as long as the initial heat generation rate of each container was less than about 750 W. It was also shown that, by using the vertical dimension available in granitic plutons and constructing either multiple-level vaults or very deep boreholes, initial areal heat fluxes greater than 31 W/m 2 can be utilized without exceeding the 80 deg C maximum temperature increase anywhere in the vault. (author)

Electrochemical desalination of the ionic mixture measured in the vault of Rørby Church - Laboratory scale

DEFF Research Database (Denmark)

Rörig-Dalgaard, Inge; Bøllingtoft, P.

2011-01-01

Salt induced deterioration of the vault in Rørby Church has resulted in extensive deterioration of the vault itself and of the murals on the lower surface of the vault. The upper surface of the vault has previously been covered with cement mortar. After removing this cement mortar some...... of the bricks erode and form brick powder. The brick powder was collected in 2002, 2005 and 2011 and measured for the most common salts in Danish vault structures. On the basis of the actual measured ions two different salt mixtures were made and used for contamination of brick specimens. By use of an applied...

Preoperative Biometric Parameters Predict the Vault after ICL Implantation: A Retrospective Clinical Study.

Science.gov (United States)

Zheng, Qian-Yin; Xu, Wen; Liang, Guan-Lu; Wu, Jing; Shi, Jun-Ting

2016-01-01

To investigate the correlation between the preoperative biometric parameters of the anterior segment and the vault after implantable Collamer lens (ICL) implantation via this retrospective study. Retrospective clinical study. A total of 78 eyes from 41 patients who underwent ICL implantation surgery were included in this study. Preoperative biometric parameters, including white-to-white (WTW) diameter, central corneal thickness, keratometer, pupil diameter, anterior chamber depth, sulcus-to-sulcus diameter, anterior chamber area (ACA) and central curvature radius of the anterior surface of the lens (Lenscur), were measured. Lenscur and ACA were measured with Rhinoceros 5.0 software on the image scanned with ultrasound biomicroscopy (UBM). The vault was assessed by UBM 3 months after surgery. Multiple stepwise regression analysis was employed to identify the variables that were correlated with the vault. The results showed that the vault was correlated with 3 variables: ACA (22.4 ± 4.25 mm2), WTW (11.36 ± 0.29 mm) and Lenscur (9.15 ± 1.21 mm). The regressive equation was: vault (mm) = 1.785 + 0.017 × ACA + 0.051 × Lenscur - 0.203 × WTW. Biometric parameters of the anterior segment (ACA, WTW and Lenscur) can predict the vault after ICL implantation using a new regression equation. © 2016 The Author(s) Published by S. Karger AG, Basel.

Spent LWR fuel-storage costs

International Nuclear Information System (INIS)

Clark, H.J.

1981-01-01

Expanded use of existing storage basins is clearly the most economic solution to the spent fuel storage problem. The use of high-density racks followed by fuel disassembly and rod storage is an order of magnitude cheaper than building new facilities adjacent to the reactor. The choice of a new storage facility is not as obvious; however, if the timing of expenditures and risk allowance are to be considered, then modular concepts such as silos, drywells, and storage casks may cost less than water basins and air-cooled vaults. A comparison of the costs of the various storage techniques without allowances for timing or risk is shown. The impact of allowances for discounting and early resumption of reprocessing is also shown. Economics is not the only issue to be considered in selecting a storage facility. The licensing, environmental impact, timing, and social responses must also be considered. Each utility must assess all of these issues for their particular reactors before the best storage solution can be selected

ORSPHERE: CRITICAL, BARE, HEU(93.2)-METAL SPHERE

Energy Technology Data Exchange (ETDEWEB)

Margaret A. Marshall

2013-09-01

In the early 1970’s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an attempt to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950’s (HEU-MET-FAST-001). The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. “The very accurate description of this sphere, as assembled, establishes it as an ideal benchmark for calculational methods and cross-section data files.” (Reference 1) While performing the ORSphere experiments care was taken to accurately document component dimensions (±0. 0001 in. for non-spherical parts), masses (±0.01 g), and material data The experiment was also set up to minimize the amount of structural material in the sphere proximity. A three part sphere was initially assembled with an average radius of 3.4665 in. and was then machined down to an average radius of 3.4420 in. (3.4425 in. nominal). These two spherical configurations were evaluated and judged to be acceptable benchmark experiments; however, the two experiments are highly correlated.

Multilateral nonproliferation cooperation: US - Led effort to remove HEU/LEU fresh and spent fuels from the Republic of Georgia to Dounreay, Scotland

International Nuclear Information System (INIS)

Shelton, Thomas A.; Viebrock, James M.; Riedy, Alexander W.; Moses, Stanley D.; Bird, Helen M.

1998-01-01

This paper presents the efforts led by United States for removing HEU/LEU fresh and spent fuel from dhe Republic of Georgia to Dounreay, Scotland. These efforts are resulted from a plan approved by the United States Government, in cooperation with the United Kingdom and Georgia Governments to rapidly retrieve and transport circa 4.3 kilograms of enriched uranium. This material consisted largely of highly enriched uranium (HEU) and a small amount of low enriched uranium (LEU) fresh fuel, as well as about 800 grams of HEU/LEU-based spent fuel from a shutdown IR T-M research reactor on the outskirts of Table's, Georgia. The technical team lead by DOE consisted of HEU handling, packaging and transportation experts from the Oak Ridge Y-12 plant, managed and operated by Lockheed Martin Energy Systems, and fuel handling and transportation experts from Nac International in Norcross, Georgia, United States

Design analysis report for the 244-AR vault Interim Stabilization interior transfer system

International Nuclear Information System (INIS)

CARLSON, A.B.

2002-01-01

The purpose of this calculation note is to verify that the 244-AR Vault Interior Transfer System piping installed in the vault meets ASME B31.3 code requirements. This calculation also evaluates the pipe support loads

Coordination Between the HEU Transparency Program and the Material Protection, Control and Accountability Program

International Nuclear Information System (INIS)

Glaser, J.; Hernandez, J.; Dougherty, D.; Bieniawski, A.; Cahalane, P.; Mastal, E.

2000-01-01

DOE sponsored programs such as Material Protection Control and Accountability (MPC and A) and implementation of the Highly-Enriched Uranium (HEU) Transparency Program send US personnel into Russian nuclear facilities and receive Russian representatives from these programs. While there is overlap in the Russian nuclear facilities visited by these two programs, there had not been any formal mechanism to share information between them. Recently, an MPC and A/HEU Working Group was developed to facilitate the sharing of appropriate information and to address concerns expressed by Minatom and Russian facility personnel such as US visit scheduling conflicts. This paper discusses the goals of the Working Group and ways it has helped to allow the programs to work more efficiently with the Russian facilities

A COMPARISON OF UPPER-EXTREMITY REACTION FORCES BETWEEN THE YURCHENKO VAULT AND FLOOR EXERCISE

Directory of Open Access Journals (Sweden)

Matthew Kirk Seeley

2005-06-01

Full Text Available The purpose of this study was to examine reaction forces transmitted to the upper extremities of high-level gymnasts during the round-off phase of the Yurchenko vault. A secondary purpose of this study was to compare reaction forces during the Yurchenko vault to reaction forces observed in a tumbling pass during the floor exercise. Ten high-level, female gymnasts volunteered to participate. Conditions of the independent variable were the Yurchenko vault and floor exercise; dependent variables were peak vertical and peak anterior-posterior reaction forces. Each participant performed three trials of both conditions with the trail hand contacting a force platform. Vertical and anterior-posterior reaction forces, normalized to body weight, were greater (p < 0.05 during the round-off phase of the Yurchenko vault (2.38 than during the floor exercise round-off (2.15. Vertical reaction forces during the round-off phase of the Yurchenko vault and floor exercise round-off are similar to reaction forces transmitted to upper extremities during other gymnastic skills and ground reaction forces transmitted to lower extremities while running and walking at various speeds. Results of this study reveal a need for further research considering methods aimed at reducing reaction forces transmitted to the upper extremities during the Yurchenko vault and floor exercise.

Geochemical factors influencing vault design and layout

International Nuclear Information System (INIS)

Gascoyne, M.; Stroes-Gascoyne, S.; Sargent, F.P.

1995-01-01

The design and construction of a vault for used nuclear fuel in crystalline rock may be influenced by a number of geochemical factors. During the siting stage, information is needed regarding the rock type, heterogeneities in its composition and the mineralogy of permeable zones because these will cause variations in thermal conductivity, strength and radionuclide sorptive properties of the rock. These factors may affect decisions regarding depth of vault construction, tunnel dimensions and spacing of panels and waste containers. The decision on whether groundwaters are allowed to flow freely into a planned excavation may depend on measurements of their chemical compositions, microbiological contents and presence of hazardous or corrosive constituents. During site characterization, borehole drilling from the surface and subsequent hydraulic testing will introduce both chemical and microbiological contaminants that may further influence this decision. During vault construction, the geochemistry of the rock may cause changes to the characterization, design and construction of the vault. For example, high salinity fluids in micropores in the rock could prevent the use of radar surveys to detect fractures in the surrounding rock. High rock salinity may also cause unacceptably high total dissolved solids loadings in water discharged from the facility. Again, the presence of toxic, corrosive or radioactive constituents in inflowing groundwater may require grouting or, if inflow is needed for service operations, development of treatment facilities both above and below ground. In addition, the use of explosives will cause high organic and nitrate loadings in service water as well as the possible impregnation of these chemicals in the damaged wall-rock surrounding an excavation. These chemicals may remain despite cleaning efforts and act as nutrients to promote microbial activity in the post-closure phase. In the operational phase, further design and construction, changes

Sensor concentrator unit for the Continuous Automated Vault Inventory System

Energy Technology Data Exchange (ETDEWEB)

Nodine, R.N.; Lenarduzzi, R.

1997-06-01

The purpose of this document is to describe the use and operation of the sensor concentrator in the Continuous Automated Vault Inventory System (CAVIS). The CAVIS electronically verifies the presence of items of stored special nuclear material (SNM). US Department of Energy orders require that stored SNM be inventoried periodically to provide assurance that the material is secure. Currently this inventory is a highly manual activity, requiring personnel to enter the storage vaults. Using a CAVIS allows the frequency of physical inventories to be significantly reduced, resulting in substantial cost savings, increased security, and improved safety. The electronic inventory of stored SNM requires two different types of sensors for each item. The two sensors measure different parameters of the item, usually weight and gamma rays. A CAVIS is constructed using four basic system components: sensors, sensor concentrators, a data collection unit, and a database/user interface unit. One sensor concentrator supports the inventory of up to 20 items (40 sensors) and continuously takes readings from the item sensors. On request the sensor concentrator outputs the most recent sensor readings to the data collection unit. The information transfer takes place over a RS485 communications link. The data collection unit supports from 1 to 120 sensor concentrators (1 to 2,400 items) and is referred to as the Sensor Polling and Configuration System (SPCS). The SPCS is connected by a secure Transmission Control Protocol/Internet Protocol (TCP/IP) network to the database/user interface unit, which is referred to as the Graphical Facility Information Center (GraFIC). A CAVIS containing more than 2,400 items is supported by connecting additional SPCS units to the GraFIC.

Sensor concentrator unit for the Continuous Automated Vault Inventory System

International Nuclear Information System (INIS)

Nodine, R.N.; Lenarduzzi, R.

1997-06-01

The purpose of this document is to describe the use and operation of the sensor concentrator in the Continuous Automated Vault Inventory System (CAVIS). The CAVIS electronically verifies the presence of items of stored special nuclear material (SNM). US Department of Energy orders require that stored SNM be inventoried periodically to provide assurance that the material is secure. Currently this inventory is a highly manual activity, requiring personnel to enter the storage vaults. Using a CAVIS allows the frequency of physical inventories to be significantly reduced, resulting in substantial cost savings, increased security, and improved safety. The electronic inventory of stored SNM requires two different types of sensors for each item. The two sensors measure different parameters of the item, usually weight and gamma rays. A CAVIS is constructed using four basic system components: sensors, sensor concentrators, a data collection unit, and a database/user interface unit. One sensor concentrator supports the inventory of up to 20 items (40 sensors) and continuously takes readings from the item sensors. On request the sensor concentrator outputs the most recent sensor readings to the data collection unit. The information transfer takes place over a RS485 communications link. The data collection unit supports from 1 to 120 sensor concentrators (1 to 2,400 items) and is referred to as the Sensor Polling and Configuration System (SPCS). The SPCS is connected by a secure Transmission Control Protocol/Internet Protocol (TCP/IP) network to the database/user interface unit, which is referred to as the Graphical Facility Information Center (GraFIC). A CAVIS containing more than 2,400 items is supported by connecting additional SPCS units to the GraFIC

ROLE OF MESH REPAIR IN PELVIC ORGAN PROLAPSE, WITH PARTICULAR REFERENCE TO VAULT PROLAPSE & NULLIPAROUS PROLAPSE

Directory of Open Access Journals (Sweden)

Geeta Vandana Reddi

2017-12-01

Full Text Available BACKGROUND Pelvic organ prolapse is a common complaint in gynaecological practice. Nulliparous prolapse is seen in 2% of female population and vault prolapse in 0.5% following hysterectomy. Various surgical procedures have been described for the repair of vault prolapse e.g. transvaginal sacrospinous colpopexy, transabdominal sacral colpopexy, Le Forte's operation, colpoclesis, posterior intravaginal sling plasty etc. The introduction of synthetic mesh like Prolene, Mersilene for the repair of vault prolapse have the advantage of tensile strength. In nulliparous prolapse and uterovaginal prolapse also, the supports of uterus are weakened, so there seems to be a definite advantage of mesh repair over sling surgeries. MATERIALS AND METHODS A prospective study done in 50 patients with vault prolapse, Nulliparous prolapse and patients with UV prolapse who want to retain their menstrual and reproductive function were identified and repair of defects were done. 2 cases who underwent both hysterectomy and sacral colpopexy in two sittings were included in the study. They were followed up for a period of 6 weeks - 29 months. RESULTS In the present study, the mean age for vault prolapse was 52.14 years, mean parity was 3.36. Vault prolapses were found to be more common after abdominal hysterectomy 60.97% compared to vaginal 39.02%. Common indication for hysterectomy is pelvic organ prolapse followed by DUB in this study. The mean time between the hysterectomy and vault prolapse in this study was 5.46 years. Importance should be given to proper enterocoele repair and vault suspense time of primary surgery. In this study incidence of smoking in vault prolapse cases was 34.14% significant. In cases of sacral colpopexy for vault prolapse in this study there were no major mesh related complications and the cure rate was 97.5%. So, sacral colpopexy with prolene mesh i.e. abdominal mesh repair can be employed as the primary surgery for vault prolapse. The safety

Vault Safety and Inventory System users manual, PRIME 2350. Revision 1

International Nuclear Information System (INIS)

Downey, N.J.

1994-01-01

This revision is issued to request review of the attached document: VSIS User Manual, PRIME 2350, which provides user information for the operation of the VSIS (Vault Safety and Inventory System). It describes operational aspects of Prime 2350 minicomputer and vault data acquisition equipment. It also describes the User's Main Menu and menu functions, including REPORTS. Also, system procedures for the Prime 2350 minicomputer are covered

Vault Safety and Inventory System users manual, PRIME 2350. Revision 1

Energy Technology Data Exchange (ETDEWEB)

Downey, N.J.

1994-12-14

This revision is issued to request review of the attached document: VSIS User Manual, PRIME 2350, which provides user information for the operation of the VSIS (Vault Safety and Inventory System). It describes operational aspects of Prime 2350 minicomputer and vault data acquisition equipment. It also describes the User`s Main Menu and menu functions, including REPORTS. Also, system procedures for the Prime 2350 minicomputer are covered.

WWER spent fuel storage

Energy Technology Data Exchange (ETDEWEB)

Bower, C C; Lettington, C [GEC Alsthom Engineering Systems Ltd., Whetstone (United Kingdom)

1994-12-31

Selection criteria for PAKS NPP dry storage system are outlined. They include the following: fuel temperature in storage; sub-criticality assurance (avoidance of criticality for fuel in the unirradiated condition without having to take credit for burn-up); assurance of decay heat removal; dose uptake to the operators and public; protection of environment; volume of waste produced during operation and decommissioning; physical protection of stored irradiated fuel assemblies; IAEA safeguards assurance; storage system versus final disposal route; cost of construction and extent of technology transfer to Hungarian industry. Several available systems are evaluated against these criteria, and as a result the GEC ALSTHOM Modular Vault Dry Store (MVDS) system has been selected. The MVDS is a passively cooled dry storage facility. Its most important technical, safety, licensing and technology transfer characteristics are outlined. On the basis of the experience gained some key questions and considerations related to the East European perspective in the field of spent fuel storage are discussed. 8 figs.

WWER spent fuel storage

International Nuclear Information System (INIS)

Bower, C.C.; Lettington, C.

1994-01-01

Selection criteria for PAKS NPP dry storage system are outlined. They include the following: fuel temperature in storage; sub-criticality assurance (avoidance of criticality for fuel in the unirradiated condition without having to take credit for burn-up); assurance of decay heat removal; dose uptake to the operators and public; protection of environment; volume of waste produced during operation and decommissioning; physical protection of stored irradiated fuel assemblies; IAEA safeguards assurance; storage system versus final disposal route; cost of construction and extent of technology transfer to Hungarian industry. Several available systems are evaluated against these criteria, and as a result the GEC ALSTHOM Modular Vault Dry Store (MVDS) system has been selected. The MVDS is a passively cooled dry storage facility. Its most important technical, safety, licensing and technology transfer characteristics are outlined. On the basis of the experience gained some key questions and considerations related to the East European perspective in the field of spent fuel storage are discussed. 8 figs

The mechanism of vault opening from the high resolution structure of the N-terminal repeats of MVP.

Science.gov (United States)

Querol-Audí, Jordi; Casañas, Arnau; Usón, Isabel; Luque, Daniel; Castón, José R; Fita, Ignasi; Verdaguer, Nuria

2009-11-04

Vaults are ubiquitous ribonucleoprotein complexes involved in a diversity of cellular processes, including multidrug resistance, transport mechanisms and signal transmission. The vault particle shows a barrel-shaped structure organized in two identical moieties, each consisting of 39 copies of the major vault protein MVP. Earlier data indicated that vault halves can dissociate at acidic pH. The crystal structure of the vault particle solved at 8 A resolution, together with the 2.1-A structure of the seven N-terminal domains (R1-R7) of MVP, reveal the interactions governing vault association and provide an explanation for a reversible dissociation induced by low pH. The structural comparison with the recently published 3.5 A model shows major discrepancies, both in the main chain tracing and in the side chain assignment of the two terminal domains R1 and R2.

GOTHIC analysis of post-accident hydrogen mixing behaviour in CANDU fuelling machine vault

International Nuclear Information System (INIS)

Yim, K.; Wong, R.C.; Fluke, R.J.

1997-01-01

The GOTHIC code was used to assess the post-accident hydrogen gas mixing patterns in a CANDU reactor containment and demonstrate the acceptability of Ontario Hydro Nuclear's hydrogen mitigation methods. The fuelling machine vault, being a small volume room containing major reactor piping, is the room of most concern with respect to hydrogen concentrations. Detailed three dimensional modelling of the gas mixing patterns in the fuelling machine vault was completed. Results showed that, even without forced air circulation, there is enough dispersion of hydrogen to other parts of containment to preclude the build-up of sensitive mixtures in the vault. For a brief time during the peak period of hydrogen release, hydrogen concentrations rise to close to the lower ignition limit in a small portion of the vault, but these hydrogen-steam-air mixtures are considered acceptable. Natural mixing alone is sufficient to preclude damaging hydrogen burns. (author)

Beyond BLASTing: Tertiary and Quaternary Structure Analysis Helps Identify Major Vault Proteins

Science.gov (United States)

Daly, Toni K.; Sutherland-Smith, Andrew J.; Penny, David

2013-01-01

We examine the advantages of going beyond sequence similarity and use both protein three-dimensional (3D) structure prediction and then quaternary structure (docking) of inferred 3D structures to help evaluate whether comparable sequences can fold into homologous structures with sufficient lateral associations for quaternary structure formation. Our test case is the major vault protein (MVP) that oligomerizes in multiple copies to form barrel-like vault particles and is relatively widespread among eukaryotes. We used the iterative threading assembly refinement server (I-TASSER) to predict whether putative MVP sequences identified by BLASTp and PSI Basic Local Alignment Search Tool are structurally similar to the experimentally determined rodent MVP tertiary structures. Then two identical predicted quaternary structures from I-TASSER are analyzed by RosettaDock to test whether a pair-wise association occurs, and hence whether the oligomeric vault complex is likely to form for a given MVP sequence. Positive controls for the method are the experimentally determined rat (Rattus norvegicus) vault X-ray crystal structure and the purple sea urchin (Strongylocentrotus purpuratus) MVP sequence that forms experimentally observed vaults. These and two kinetoplast MVP structural homologs were predicted with high confidence value, and RosettaDock predicted that these MVP sequences would dock laterally and therefore could form oligomeric vaults. As the negative control, I-TASSER did not predict an MVP-like structure from a randomized rat MVP sequence, even when constrained to the rat MVP crystal structure (PDB:2ZUO), thus further validating the method. The protocol identified six putative homologous MVP sequences in the heterobolosean Naegleria gruberi within the excavate kingdom. Two of these sequences are predicted to be structurally similar to rat MVP, despite being in excess of 300 residues shorter. The method can be used generally to help test predictions of homology via

The TROVE module: a common element in Telomerase, Ro and Vault ribonucleoproteins.

Science.gov (United States)

Bateman, Alex; Kickhoefer, Valerie

2003-10-16

Ribonucleoproteins carry out a variety of important tasks in the cell. In this study we show that a number of these contain a novel module, that we speculate mediates RNA-binding. The TROVE module--Telomerase, Ro and Vault module--is found in TEP1 and Ro60 the protein components of three ribonucleoprotein particles. This novel module, consisting of one or more domains, may be involved in binding the RNA components of the three RNPs, which are telomerase RNA, Y RNA and vault RNA. A second conserved region in these proteins is shown to be a member of the vWA domain family. The vWA domain in TEP1 is closely related to the previously recognised vWA domain in VPARP a second component of the vault particle. This vWA domain may mediate interactions between these vault components or bind as yet unidentified components of the RNPs. This work suggests that a number of ribonucleoprotein components use a common RNA-binding module. The TROVE module is also found in bacterial ribonucleoproteins suggesting an ancient origin for these ribonucleoproteins.

Finding of no significant impact: Interim storage of enriched uranium above the maximum historical level at the Y-12 Plant Oak Ridge, Tennessee

International Nuclear Information System (INIS)

1995-01-01

The US Department of Energy (DOE) has prepared an Environmental Assessment (EA) for the Proposed Interim Storage of Enriched Uranium Above the Maximum Historical Storage Level at the Y-12 Plant, Oak Ridge, Tennessee (DOE/EA-0929, September, 1994). The EA evaluates the environmental effects of transportation, prestorage processing, and interim storage of bounding quantities of enriched uranium at the Y-12 Plant over a ten-year period. The State of Tennessee and the public participated in public meetings and workshops which were held after a predecisional draft EA was released in February 1994, and after the revised pre-approval EA was issued in September 1994. Comments provided by the State and public have been carefully considered by the Department. As a result of this public process, the Department has determined that the Y-12 Plant-would store no more than 500 metric tons of highly enriched uranium (HEU) and no more than 6 metric tons of low enriched uranium (LEU). The bounding storage quantities analyzed in the pre-approval EA are 500 metric tons of HEU and 7,105.9 metric tons of LEU. Based on-the analyses in the EA, as revised by the attachment to the Finding of No Significant Impact (FONSI), DOE has determined that interim storage of 500 metric tons of HEU and 6 metric tons of LEU at the Y-12 Plant does not constitute a major Federal action significantly affecting the quality of the human environment, within the meaning of the National Environmental Policy Act (NEPA) of 1969. Therefore, an Environmental Impact Statement (EIS) is not required and the Department is issuing this FONSI

Data Vaults: Database Technology for Scientific File Repositories

NARCIS (Netherlands)

M.G. Ivanova (Milena); M.L. Kersten (Martin); S. Manegold (Stefan); Y. Kargin (Yagiz)

2013-01-01

htmlabstractCurrent data-management systems and analysis tools fail to meet scientists’ data-intensive needs. A "data vault" approach lets researchers effectively and efficiently explore and analyze information.

Inventory extension considerations for long-term storage at the nuclear materials storage facility

International Nuclear Information System (INIS)

Olinger, C.T.; Stanbro, W.D.; Longmire, V.; Argo, P.E.; Nielson, S.M.

1996-01-01

Los Alamos National Laboratory is in the process of modifying its nuclear materials storage facility to a long-term storage configuration. In support of this effort, we examined technical and administrative means to extend periods between physical inventories. Both the frequency and sample size during a physical inventory could significantly impact required sizing of the non-destructive assay (NDA) laboratory as well as material handling capabilities. Several options are being considered, including (1) treating each storage location as a separate vault, (2) minimizing the number of items returned for quantitative analysis by optimizing the use of in situ confirmatory measurements, and (3) utilizing advanced monitoring technologies. Careful consideration of these parameters should allow us to achieve and demonstrate safe and secure storage while minimizing the impact on facility operations and without having to increase the size of the NDA laboratory beyond that required for anticipated shipping and receiving activities

Preliminary modelling study of geochemical interactions between a used-fuel disposal vault and the surrounding geosphere

International Nuclear Information System (INIS)

McMurry, J.

1995-10-01

In the Environmental Impact Statement (EIS) and the related documents that describe the Canadian nuclear fuel waste disposal concept (AECL 1994), it has been assumed that a disposal vault would have no significant geochemical impact on the geosphere, and so no such effects were included explicitly in the postclosure assessment model. The purpose of this study was to estimate the general magnitude and significance of vault-induced geochemical changes over an expected range of temperatures. The results of the preliminary modelling are used broadly to evaluate the implications of these changes for the migration of radionuclides through the geosphere. The geochemical modelling program PHREEQE was used to calculate the changes in mineral solubilities that would result from the transfer of aqueous species from the vault to the geosphere or that would result from groundwater-granite interactions enhanced by vault-derived elevated temperatures. Twelve representative vault water compositions, derived from predicted interactions with buffer material and backfill over a range of temperatures up to 95 deg C, were used in the modelling. For the conditions modelled it was determined that the interactions of the geosphere with dissolved vault constituents, and the relatively modest maximum increase in groundwater temperature produced by a vault, would have a limited impact on the geosphere. The conclusions of this preliminary study are qualified by some of the simplifying assumptions used in the modelling. More realistic modelling of natural systems requires a more detailed representation of water-solid interactions with a variety of vault materials at elevated temperatures. (author) 48 refs., 13 tabs, 4 figs

Energy–angle correlation of neutrons and gamma-rays emitted from an HEU source

Energy Technology Data Exchange (ETDEWEB)

Miloshevsky, G., E-mail: gennady@purdue.edu; Hassanein, A.

2014-06-01

Special Nuclear Materials (SNM) yield very unique fission signatures, namely correlated neutrons and gamma-rays. A major challenge is not only to detect, but also to rapidly identify and recognize SNM with certainty. Accounting for particle multiplicity and correlations is one of standard ways to detect SNM. However, many parameter data such as joint distributions of energy, angle, lifetime, and multiplicity of neutrons and gamma-rays can lead to better recognition of SNM signatures in the background radiation noise. These joint distributions are not well understood. The Monte Carlo simulations of the transport of neutrons and gamma-rays produced from spontaneous and interrogation-induced fission of SNM are carried out using the developed MONSOL computer code. The energy spectra of neutrons and gamma-rays from a bare Highly Enriched Uranium (HEU) source are investigated. The energy spectrum of gamma-rays shows spectral lines by which HEU isotopes can be identified, while those of neutrons do not show any characteristic lines. The joint probability density function (JPDF) of the energy–angle association of neutrons and gamma-rays is constructed. Marginal probability density functions (MPDFs) of energy and angle are derived from JPDF. A probabilistic model is developed for the analysis of JPDF and MPDFs. This probabilistic model is used to evaluate mean values, standard deviations, covariance and correlation between the energy and angle of neutrons and gamma-rays emitted from the HEU source. For both neutrons and gamma-rays, it is found that the energy–angle variables are only weakly correlated.

Thermal calculations for the study of the heat evacuation in the vaults building of the centralised temporary storage (ATC); Calculos termicos para el estudio de la evacuacion de calor en el Edificio de Bovedas del Almacen Temporal Centralizado (ATC)

Energy Technology Data Exchange (ETDEWEB)

Perez Dominguez, J. R.; Garcia Gonzalez, M.; Huelamo, E.

2015-07-01

This article presents the thermal analyses of the vaults building at the future Spanish Nuclear Waste Storage facility (ATC) in which spent nuclear fuel and high activity nuclear wastes are to be stored efficiency, safety and securely. the analyses have been carried out by means of computational fluid dynamics (CFD) simulation codes, for the purpose of confirming the adequate design of the storage buildings and in order to obtain the air flow rate required to guarantee that the different thermal criteria are met. The design relies on natural convection in order to remove residual heat from the nuclear waste. The simulation model allows the designer to perform sensitivity analyses to evaluate the impact of different design parameters, to optimize the heat load per fuel canister and to provide an optimal loading plan for the facility. (Author)

Forensic analysis of a smuggled HEU sample interdicted in Bulgaria

International Nuclear Information System (INIS)

Niemeyer, S.; Hutcheon, I.

2002-01-01

Full text: A sample of HEU was seized in Rousse, Bulgaria on May 29, 1999, at a border crossing between Bulgaria and Romania. A search of the suspect's vehicle uncovered a lead canister hidden in the car trunk. The initial examination of the contents by Bulgarian scientists indicated that the sample was indeed HEU, and subsequently arrangements were made for a U.S. team of nuclear forensics scientists from several national laboratories to conduct a thorough examination. This report gives a summary of the results. The HEU sample was contained in a glass ampoule that was embedded in a yellow wax, and in turn the wax filled the inside of the cylindrical lead container. A broad set of techniques was used to examine both nuclear and non-nuclear materials. Our general experimental approach has been previously described at meetings of the Nuclear Smuggling International Technical Working Group (ITWG), but this case represents the application of the most diverse set of nuclear forensic measurements for an actual seized sample. Analysis of the HEU itself included particle characterization, stoichiometry, impurity elements, residual nuclides, age-dating, and U and Pu isotopics. Measurements by XRD, SEM, and TEM show that sample is mostly U3O8, with minor amounts of two other phases. The powder is extremely fine-grained (160 nm mean) and quite uniform in size. Most grains (95%) are equidimensional, with the remainder rod-or plate-shaped. The U is 72.7% U-235 with a high U-236 abundance of 12.1%. The sample is reprocessed, reactor-irradiated material. The original U enrichment was 90% and the irradiation burned up about 50% of the initial U-235. Pu is present at a very low-level (3 ppb); the Pu-239 abundance is 82% with 240/239=0.12. Three fission products were detected at low levels, giving unambiguous evidence of fuel recycling. The total impurity content is about 600 ppmw (mostly S, Cl, Fe, and Br), which we interpret as indicating a batch processing operation because the

The disposal of Canada's nuclear fuel waste: the vault model for postclosure assessment

International Nuclear Information System (INIS)

Johnson, L.H.; LeNeveu, D.M.; Shoesmith, D.W.; Oscarson, D.W.; Gray, M.N.; Lemire, R.J.; Garisto, N.C.

1994-01-01

The concept for disposal of Canada's nuclear fuel waste involves emplacing the waste in a vault excavated at a depth of 500 to 1000 m in plutonic rock of thc Canadian Shield. The solid waste would be isolated from the biosphere by a multibarrier system consisting of engineered barriers, including corrosion-resistant containers and clay- and cement-based sealing materials, and the natural barrier provided by the massive geological formation. The technical feasibility of this concept, and its impact on the environment and human health, is being documented in an Environmental Impact Statement (EIS) that will be submitted for review under the federal Environmental Assessment and Review Process. The present report is one of nine EIS primary references. The report describes the vault model, which is used to calculate the time-dependent release of radioactive and non-radioactive contaminants from the engineered barrier system (vault) into the surrounding rock (geosphere). The model calculations presented are for a specific reference vault design that comprises used CANDU (CANada Deuterium Uranium) fuel bundles in Grade-2 titanium containers, which would be emplaced in boreholes in the floor of a mined excavation located at a depth of 500 m in plutonic rock. The containers would be surrounded by a compacted buffer material that is a mixture of 50 wt. % sand and 50 wt. % bentonite. Disposal rooms and tunnels would be sealed with a layer of backfill mixture composed of 25 % glacial lake clay and 75% crushed granite and an overlying layer of buffer material. The vault model is a computer code that calculates the failure times of titanium containers, the rate of release of radionuclides from used-fuel bundles into the groundwater that would flow into the failed containers, and the rate of transport of radionuclides from the fuel through the groundwater-saturated buffer and backfill materials and into the surrounding rock. The vault model uses distributed or probabilistic

Neutron field inside a PET Cyclotron vault room

International Nuclear Information System (INIS)

Vega C, H.R.; Mendez, R.; Iniguez, M.P.; Climent, J.M.; Penuelas, I.; Barquero, R.

2006-01-01

The neutron field around a Positron Emission Tomography cyclotron was investigated during 18 F radioisotope production with an 18 MeV proton beam. In this study the Ion Beam Application cyclotron, model Cyclone 18/9, was utilized. Measurements were carried out with a Bonner sphere neutron spectrometer with pairs of thermoluminescent dosemeters (TLD600 and TLD700) as thermal neutron detector. The TLDs readouts were utilized to unfold the neutron spectra at three different positions inside the cyclotron's vault room. With the spectra the Ambient dose equivalent was calculated. Neutron spectra unfolding were performed with the BUNKIUT code and the UTA4 response matrix. Neutron spectra were also determined by Monte Carlo calculations using a detailed model of cyclotron and vault room. (Author)

First experiences with electrochemical in-situ desalination of bricks in a church vault construction

DEFF Research Database (Denmark)

Rörig-Dalgaard, Inge

2012-01-01

was chosen. Salt profiles from three different bricks within this area clarified two bricks with low ion contents and one with a high ion content which is representative for church vault constructions. The idea of using a climate chamber for dissolution of present salts for minimizing additional water supply...... problematic in relation to church vault constructions with murals as the surface deterioration can result in loss of valuable cultural heritage. An electrochemical method has been investigated with focus on possible use for desalination of salt loaded vault constructions with murals in laboratory scale...

Vault lining for 340 waste handling facility, 300 area

International Nuclear Information System (INIS)

Hollenbeck, R.G.

1997-01-01

Coating systems by Protection Enterprises, Ameron, Carboline, and Steelcote were evaluated. Each manufacturer has a coating system that is acceptable for use in the 340 Vault (see Appendix A). The choice of which system to use will be made after in-place adhesion tests are complete. The Protection Enterprises coating has the greatest potential for acceptable adhesion with minimal surface preparation. Total project cost for engineering and construction is $1,220,000 including 50% for contingency (see Appendix B). If the existing vault coverblock access hatch can satisfy entry requirements, $95,000 can be saved from the removal of coverblocks and the erection and disassembly of the greenhouse

The status of HEU and LEU core conversion activities at the Jamaica SLOWPOKE

Energy Technology Data Exchange (ETDEWEB)

Preston, J.; Grant, C., E-mail: john.preston@uwimona.edu.jm [Univ. of the West Indies, Mona Campus, International Centre for Environmental and Nuclear Sciences, Kingston (Jamaica)

2013-07-01

The SLOWPOKE reactor in Jamaica has been operated by the International Centre for Environmental and Nuclear Sciences, University of the West Indies since 1984, mainly for the purpose of Neutron Activation Analysis. The HEU core with current utilization has another 14 years of operation, before the addition of a large beryllium annulus would be required to further extend the life-time by 15 years. However, in keeping with the spirit of the Reduced Enrichment for Research and Test Reactors (RERTR) program, the decision was taken in 2003 to convert the core from HEU to LEU, inline with those at the Ecole Polytechnic and RMC SLOWPOKE facilities. This paper reports on the current status of the conversion activities, including key fuel manufacture and regulatory issues, which have seen substantial progress during the last year. A timetable for the complete process is given, and provided that the fuel fabrication can be completed in the estimated 18 months, the core conversion should be accomplished by the end of 2014. (author)

Vault submodel for the second interim assessment of the Canadian concept for nuclear fuel waste disposal

International Nuclear Information System (INIS)

LeNeveu, D.M.

1986-02-01

The consequences to man and the environment of the disposal of nuclear fuel waste are being studied within the Canadian Nuclear Fuel Waste Management Program. The concept being assessed is that of a sealed disposal vault at a depth of 1000 m in plutonic rock in the Canadian Shield. To determine the consequences, the vault and its environment are simulated using a SYstem Variability Analysis Code (SYVAC), a stochastic model of the disposal system. SYVAC contains three submodels that represent the three major parts of the disposal system: the vault, the geosphere and the biosphere. This report documents the conceptual and mathematical framework of the vault submodel

Architecture - re-Construction - Geometry: a knowledge path applied to the study of Plantery's vaults in Turin

Directory of Open Access Journals (Sweden)

Roberta Spallone

2012-06-01

Full Text Available Plantery's vaults, by the name of the architect working in Turin since first years of '700, are composed by a main vault intersected by some axial and angular secondary vaults. Plantery's vaults allow to roof over unitary spaces obtaining effects of lightness also by small rises. Their diffusion, up to the '80s of the century, in boardrooms as palaces halls, and the richness of variations on the theme, justifies a research, that makes use of survey and digital modeling. This research is focused on the relationships between geometric models and built architecture, connected by use of materials, construction techniques and decorative styles.

Preliminary assessment of alternative dry storage methods for the storage of commercial spent nuclear fuel

International Nuclear Information System (INIS)

1981-11-01

This report presents the results of an assessment of the (1) state of technology, (2) licensability, (3) implementation schedule, and (4) costs of alternative dry methods for storage of spent fuel at a reactor location when used to supplement reactor pool storage facilities. The methods of storage that were considered included storage in casks, drywells, concrete silos and air-cooled vaults. The impact of disassembly of spent fuel and storage of consolidated fuel rods was also determined. The economic assessments were made based on the current projected storage requirements of Virginia Electric and Power Company's Surry Station for the period 1985 to 2009, which has two operating pressurized water reactors (824 MWe each). It was estimated that the unit cost for storage of spent fuel in casks would amount to $117/kgU and that such costs for storage in drywells would amount to $137/kgU. However, based on the overall assessment it was concluded both storage methods were equal in merit. Modular methods of storage were generally found to be more economic than those requiring all or most of the facilities to be constructed prior to commencement of storage operations

Association of lens vault with narrow angles among different ethnic groups.

Science.gov (United States)

Lee, Roland Y; Huang, Guofu; Cui, Qi N; He, Mingguang; Porco, Travis C; Lin, Shan C

2012-06-01

To compare lens vault between open-angle and narrow-angle eyes in African-, Caucasian-, Hispanic-, Chinese- and Filipino-Americans. In this prospective study, 436 patients with open angle and narrow angle based on the Shaffer gonioscopic grading classification underwent anterior-segment optical coherence tomography. The Zhongshan Angle Assessment Program was used to calculate lens vault. The narrow-angle group included 32 Chinese-Americans, 22 Filipino-Americans, 26 African-Americans, 24 Hispanic-Americans and 73 Caucasian-Americans. The open-angle group included 56 Chinese-Americans, 29 Filipino-Americans, 45 African-Americans, 27 Hispanic-Americans and 102 Caucasian-Americans. Linear mixed effect regression models, accounting for the use of both eyes and adjusting for age, sex, pupil diameter and spherical equivalent, were used to test for the ethnicity and angle coefficients. Tukey's multiple comparison test was used for pairwise comparisons among the open-angle racial groups. Significant difference in lens vault was found among the open-angle racial groups (P = 0.022). For the open-angle patients, mean values for the lens vault measurements were 265 ± 288 µm for Chinese-Americans, 431 ± 248 µm for Caucasian-Americans, 302 ± 213 µm for Filipino-Americans, 304 ± 263 µm for Hispanic-Americans and 200 ± 237 µm for African-Americans. Using Tukey's multiple comparison for pairwise comparisons among the open-angle racial groups, a significant difference was found between African-American and Caucasian-Americans groups (P values for the rest of the pairwise comparisons were not statistically significant. No significant difference was found among the narrow-angle racial groups (P = 0.14). Comparison between the open angle and narrow angle within each racial group revealed significant difference for all racial groups (P < 0.05). Among all the ethnicities included in this study, narrow-angle eyes have greater lens vault compared to open

The TROVE module: A common element in Telomerase, Ro and Vault ribonucleoproteins

Directory of Open Access Journals (Sweden)

Bateman Alex

2003-10-01

Full Text Available Abstract Background Ribonucleoproteins carry out a variety of important tasks in the cell. In this study we show that a number of these contain a novel module, that we speculate mediates RNA-binding. Results The TROVE module – Telomerase, Ro and Vault module – is found in TEP1 and Ro60 the protein components of three ribonucleoprotein particles. This novel module, consisting of one or more domains, may be involved in binding the RNA components of the three RNPs, which are telomerase RNA, Y RNA and vault RNA. A second conserved region in these proteins is shown to be a member of the vWA domain family. The vWA domain in TEP1 is closely related to the previously recognised vWA domain in VPARP a second component of the vault particle. This vWA domain may mediate interactions between these vault components or bind as yet unidentified components of the RNPs. Conclusions This work suggests that a number of ribonucleoprotein components use a common RNA-binding module. The TROVE module is also found in bacterial ribonucleoproteins suggesting an ancient origin for these ribonucleoproteins.

Foreign research reactor spent nuclear fuel inventories containing HEU and LEU of US-origin

International Nuclear Information System (INIS)

Matos, J.E.

1995-01-01

This paper provides estimates of the quantities and types of foreign research reactor spent nuclear fuel containing HEU and LEU of US-origin that are anticipated during the period beginning in January 1996 and extending for 10-15 years

Fuel conversion of JRR-4 from HEU to LEU

International Nuclear Information System (INIS)

Ichikawa, Hiroki; Nakajima, Teruo

1997-01-01

Japanese JRR-4 (Japan Research Reactor No.4) is a pool type, light water moderated and cooled, ETR type fuel reactor used for Shielding experiments, isotope production, neutron activation analyses, Si doping, reactor students training. It acieved first criticality on January 28, 1965 with maximum thermal power 3.5MW. The standard core consistes of 20 Fuel elements, 7 control rods 5 Irradiation holes, neutron source, graphite reflectors. Available thermal flux is 7x1013 n/cm2/s. Within the RERTR program plans are made for core conversion from HEU to LEU

Radiation survey of first Hi-Art II Tomotherapy vault design in India

International Nuclear Information System (INIS)

Kinhikar, Rajesh A.; Jamema, S.V.; Pai, Rajeshree; Sharma, P.K. Dash; Deshpande, Deepak D.

2009-01-01

A vault as per government-regulation compliance with adequate shielding needs was designed and constructed for Hi-Art II Tomotherapy machine being the first in India. Radiation measurements around this Tomotherapy treatment vault were carried out to check the shielding adequacy of the source housing and the vault. It was mandatory to get this un-conventional machine 'Type Approved' by Atomic Energy Regulatory Board (AERB) in India. The aim of this paper was to report on the radiation levels measured during the radiation survey carried out for this machine. The radiation levels in and around the vault were measured for stationary as well as rotational treatment procedures with the largest open field size (5 cm x 40 cm) at the isocenter with and without scattering medium. The survey was performed at three locations near each wall surrounding the vault as well. The leakage radiation from the source housing was measured both in the patient plane outside the treatment field and one meter distance from the source outside the patient plane. The radiation levels both for stationary as well as rotational procedures were within 1 mR/h. No significance difference was observed in the radiation levels measured for rotational procedures with and without scattering medium. The leakage radiation in the patient plane was found to be 0.04% (Tolerance 0.2%), while the head leakage was 0.007% (Tolerance 0.5%) of the dose rate at the isocenter. The treatment delivery with Tomotherapy does play safe radiation levels around the installation layout and also passes the leakage criteria as well.

Expert's statement on the research reactor Munich II (FRM-II); Gutachterliche Stellungnahme zum Forschungsreaktor Muenchen II (FRM-II)

Energy Technology Data Exchange (ETDEWEB)

Liebert, Wolfgang; Friess, Friederike; Gufler, Klaus; Arnold, Nikolaus [Univ. fuer Bodenkultur (BOKU), Wien (Austria). Inst. fuer Sicherheits- und Risikowissenschaften (ISR)

2017-12-15

The Expert's statement on the research reactor FRM-II covers the following issues: The situation in Germany with respect to HEU (highly enriched uranium) fuel elements, the proliferation problems related to HEU fuel and the generated high-level radioactive wastes, possible safety hazards of an interim storage of HEU containing wastes, for instance in the interim storage facility Ahaus, possible safety hazards of final disposal of HEU containing radioactive wastes, possibilities to avoid the use of HEU fuel in order to prevent further production of these wastes, requirement of processing spent HEU containing fuel elements for final disposal.

Managing Spent Nuclear Fuel at the Idaho National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Thomas Hill; Denzel L. Fillmore

2005-10-01

The Idaho National Laboratory (INL) has a large inventory of diverse types of spent nuclear fuel (SNF). This legacy derives from the history of the INL as the National Reactor Testing Station, and from its mission to recover HEU from SNF and to test and examine SNF after irradiation. The INL also has a large diversity of SNF storage facilities, some 50 years old. SNF at INL has many forms—from intact assemblies down to metallurgical mounts, and some fuel has been wet stored for over 40 years. SNF is stored bare or in metal cans under water, or dry in vaults, caissons or casks. Inspection shows varying corrosion and degradation of the SNF and its storage cans. SNF has been stored in 10 different facilities: 5 pools, one cask storage pad, one vault, two generations of caisson facilities, and one modular Independent Spent Fuel Storage Installation (ISFSI). The pools range in age from 40 years old to the most modern in the US Department of Energy (DOE) complex. The near-term objective is to move SNF from older pools to interim dry storage, allowing shutdown and decommissioning of the older facilities. This move involves drying methods that are dependent on fuel type. The long-term objective is to have INL SNF in safe dry storage and ready to be shipped to the National Repository. The unique features of the INL SNF requires special treatments and packaging to meet the proposed repository acceptance criteria and SNF will be repackaged in standardized canisters for shipment and disposal in the National Repository. Disposal will use the standardized canisters that can be co-disposed with High Level Waste glass logs to limit the total fissile material in a repository waste package. The DOE standardized canister also simplifies the repository handling of the multitude of DOE SNF sizes and shapes.

Skull morphometry and vault sutures of Myrmecophaga tridactyla and Tamandua tetradactyla

Directory of Open Access Journals (Sweden)

Camila M. de S. Hossotani

2017-10-01

Full Text Available ABSTRACT This study aimed to examine the relationship between skull size and the level of cranial vault suture closure. A total of 50 Myrmecophaga tridactyla Linnaeus, 1758 and 178 Tamandua tetradactyla Linnaeus, 1758 skulls were analyzed in relation to 18 skull dimensions. The skulls were grouped into three levels of suture closure: no sutures closed (level 0, one or all the fallowing sutures closed: interfrontalis, sagitalis and coronalis (level 1 and all sutures closed (level 2. The results indicated that among the 18 variables measured, 17 showed significant differences (p ≤ 0.01 between level 0 and level 1 skulls of T. tetradactyla; as well as between level 0 and level 1, and level 0 and level 2 skulls of M. tridactyla. M. tridactyla level 1 and level 2 had no significant difference among any of the 18 dimensions. The foramen magnum height in both species showed no significant difference (p > 0.05 among any suture categories. In principle, suture closure level and cranial dimensions are related. The specimens with larger cranial dimensions showed greater number of cranial vault sutures closed for both species of anteaters. Tamandua tetradactyla and M. tridactyla specimens with none of the cranial vault suture closed have a foramen magnum height similar to those with cranial vault suture closed.

Knowledge Transfer and Culture Exchange between HEU and TAMU through a Summer School on Nuclear Power Engineering

International Nuclear Information System (INIS)

Gao, P.; Zhang, Z.; Kurwitz, R. C.; Shao, L.

2016-01-01

Full text: Since 2012, Harbin Engineering University (HEU) and Texas A&M University (TAMU) hold an annual Summer School on Nuclear Engineering. By now, the activity has been held four times. Each year, 15–20 students are selected from their respective institutions and paired with a counterpart to form partners. They study lectures in the first week at HEU and tour three nuclear power plants (NPP) in the second week, visit the NPP simulators, and learn the nuclear safety culture. This activity expands the scale of international nuclear educational exchange, provide a platform for the students from different countries to communicate and exchange insights into their respective culture. (author

Interim spent-fuel storage options at commercial nuclear power plants

International Nuclear Information System (INIS)

Thakkar, A.R.; Hylko, J.M.

1991-01-01

Although spent fuel can be stored safely in waterfilled pools at reactor sites, some utilities may not possess sufficient space for life-of-plant storage capability. In-pool storage capability may be increased by reracking assemblies, rod consolidation, double tiering spent-fuel racks, and by shipping spent fuel to other utility-owned facilities. Long-term on-site storage capability for spent fuel may be provided by installing (dry-type) metal casks, storage and transportation casks, concrete casks, horizontal concrete modules, modular concrete vaults, or by constructing additional (pool-type) storage installations. Experience to date has provided valuable information regarding dry-type or pool-type installations, cask handling and staffing requirements, security features, decommissioning activities, and radiological issues

Experimental modelling of core debris dispersion from the vault under a PWR pressure vessel. Pt. 2

International Nuclear Information System (INIS)

Rose, P.W.

1987-12-01

In previous experiments, done on a 1/25 scale model in Perspex of the vault under a PWR pressure vessel, the instrument tubes support structure built into the vault was not included. It consists of a number of grids made up of fairly massive steel girders. These have now been added to the model and experiments performed using water to simulate molten core debris assumed to have fallen on to the vault floor and high-pressure air to simulate the discharge of steam or gas from the assumed breach at the bottom of the pressure vessel. The results show that the tubes support structure considerably reduces the carry-over of liquid via the vault access shafts. (author)

Multi-Purpose Storage Complex description

International Nuclear Information System (INIS)

Nyman, D.H.

1993-01-01

The Multi-Purpose Storage Complex will provide interim storage of radioactive material (irradiated fuel, cesium/strontium capsules, plutonium residuals, canisters of vitrified high-level waste glass, and other radioactive material) at the Hanford Site near Richland, Washington. A Storage Preparation and Shipping Facility is included that will have the capability to stabilize failed metal fuel, segregate high-level solid waste, and package/repackage any of the materials for interim storage/final disposal or subsequent processing. Current technology, both domestic and foreign, will be adapted with the expectation that no new technology will be required. This cost-effective approach will use fuel casks, transport systems, and/or modular vaults that have been licensed in the United States. The complex will have a central control room, and appropriate safeguards and security measures will be incorporated. A specific design objective will be to minimize the amount of secondary waste

Fluxes at experiment facilities in HEU and LEU designs for the FRM-II

International Nuclear Information System (INIS)

Hanan, N. A.

1998-01-01

An Alternative LEU Design for the FRM-II proposed by the RERTR Program at Argonne National Laboratory (ANL) has a compact core consisting of a single fuel element that uses LEU silicide fuel with a uranium density of 4.5 g/cm 3 and has a power level of 32 MW. Both the HEU design by the Technical University of Munich (TUM) and the alternative LEU design by ANL have the same fuel lifetime(50 days) and the same neutron flux performance (8 x 10 14 n/cm 2 -s in the reflector). LEU silicide fuel with 4.5 g/cm 3 has been thoroughly tested and is fully-qualified, licensable, and available now for use in a high flux reactor such as the FRM-II. Several issues that were raised by TUM have been addressed in Refs. 1-3. The conclusions of these analyses are summarized below. This paper addresses four additional issues that have been raised in several forums, including Ref 4: heat generation in the cold neutron source (CNS), the gamma and fast neutron fluxes which are components of the reactor noise in neutron scattering experiments in the experiment hall of the reactor, a fuel cycle length difference, and the reactivity worth of the beam tubes and other experiment facilities. The results show that: (a) for the same thermal neutron flux, the neutron and gamma heating in the CNS is smaller in the LEU design than in the HEU design, and cold neutron fluxes as good or better than those of the HEU design can be obtained with the LEU design; (b) the gamma and fast neutron components of the reactor noise in the experiment hall are about the same in both designs; (c) the fuel cycle length is 50 days for both designs; and (d) the absolute value of the reactivity worth of the beam tubes and other experiment facilities is smaller in the LEU design, allowing its fuel cycle length to be increased to 53 or 54 days. Based on the excellent results for the Alternative LEU Design that were obtained in all analyses, the RERTR Program reiterates its conclusion that there are no major technical

Disruption of the murine major vault protein (MVP/LRP) gene does not induce hypersensitivity to cytostatics.

Science.gov (United States)

Mossink, Marieke H; van Zon, Arend; Fränzel-Luiten, Erna; Schoester, Martijn; Kickhoefer, Valerie A; Scheffer, George L; Scheper, Rik J; Sonneveld, Pieter; Wiemer, Erik A C

2002-12-15

Vaults are ribonucleoprotein particles with a distinct structure and a high degree of conservation between species. Although no function has been assigned to the complex yet, there is some evidence for a role of vaults in multidrug resistance. To confirm a direct relation between vaults and multidrug resistance, and to investigate other possible functions of vaults, we have generated a major vault protein (MVP/lung resistance-related protein) knockout mouse model. The MVP(-/-) mice are viable, healthy, and show no obvious abnormalities. We investigated the sensitivity of MVP(-/-) embryonic stem cells and bone marrow cells derived from the MVP-deficient mice to various cytostatic agents with different mechanisms of action. Neither the MVP(-/-) embryonic stem cells nor the MVP(-/-) bone marrow cells showed an increased sensitivity to any of the drugs examined, as compared with wild-type cells. Furthermore, the activities of the ABC-transporters P-glycoprotein, multidrug resistance-associated protein and breast cancer resistance protein were unaltered on MVP deletion in these cells. In addition, MVP wild-type and deficient mice were treated with the anthracycline doxorubicin. Both groups of mice responded similarly to the doxorubicin treatment. Our results suggest that MVP/vaults are not directly involved in the resistance to cytostatic agents.

The Application of Carbon Composites in the Rehabilitation of Historic Baroque Vaults

OpenAIRE

Jiří Witzany; Jiří Brožovský; Tomáš Čejka; Klára Kroftová; Jan Kubát; Daniel Makovička; Radek Zigler

2015-01-01

The stabilization and reinforcement of damaged barrel vaults with lunettes over an arcaded walk, applying composite strips based on high-strength carbon fibers and epoxy resin, was performed during the restoration of a historic monastery. The application of reinforcing composite strips in the soffit of damaged barrel vaults was preceded by relatively extensive experimental research and theoretical analyses. This method significantly reduces the interventions into and the degradation of the or...

Seismic analysis of the mirror fusion test facility shielding vault

International Nuclear Information System (INIS)

Gabrielsen, B.L.; Tsai, K.

1981-04-01

This report presents a seismic analysis of the vault in Building 431 at Lawrence Livermore National Laboratory which houses the mirror Fusion Test Facility. The shielding vault structure is approximately 120 ft long by 80 ft wide and is constructed of concrete blocks approximately 7 x 7 x 7 ft. The north and south walls are approximately 53 ft high and the east wall is approximately 29 ft high. These walls are supported on a monolithic concrete foundation that surrounds a 21-ft deep open pit. Since the 53-ft walls appeared to present the greatest seismic problem they were the first investigated

Hazard Evaluation for 244-CR Vault

Energy Technology Data Exchange (ETDEWEB)

GRAMS, W.H.

1999-08-19

This document presents the results of a hazards identification and evaluation performed on the 244-CR Vault to close a USQ (USQ No.TF-98-0785, Potential Inadequacy in Authorization Basis (PIAB): To Evaluate Miscellaneous Facilities Listed In HNF-2503 And Not Addressed In The TWRS Authorization Basis) that was generated as part of an evaluation of inactive TWRS facilities.

Hazard Evaluation for 244-CR Vault

International Nuclear Information System (INIS)

GRAMS, W.H.

1999-01-01

This document presents the results of a hazards identification and evaluation performed on the 244-CR Vault to close a USQ (USQ No.TF-98-0785, Potential Inadequacy in Authorization Basis (PIAB): To Evaluate Miscellaneous Facilities Listed In HNF-2503 And Not Addressed In The TWRS Authorization Basis) that was generated as part of an evaluation of inactive TWRS facilities

A comparative and predictive study of the annual fuel cycle costs for HEU and LEU fuels in the High Flux Reactor, Petten, 1985-1993

Energy Technology Data Exchange (ETDEWEB)

Moss, R L; May, P

1985-07-01

The internationally agreed constraint on availability of supply of HEU fuels to Research and Test Reactors has necessitated that a cost analysis be carried out to determine the financial effect of converting the core of the HFR from HEU to LEU fuels. A computer program, written at Petten and based on information extracted from studies in Europe and the USA, identifies the major cost variables to be manufacturing, uranium, reprocessing and transport costs. Comparison between HEU and LEU cores have been carried out and includes the effects of inflation and exchange rate fluctuations. Conversion of the HFR core to LEU fuels is shown to be financially disadvantageous. (author)

A comparative and predictive study of the annual fuel cycle costs for HEU and LEU fuels in the High Flux Reactor, Petten, 1985-1993

International Nuclear Information System (INIS)

Moss, R.L.; May, P.

1985-01-01

The internationally agreed constraint on availability of supply of HEU fuels to Research and Test Reactors has necessitated that a cost analysis be carried out to determine the financial effect of converting the core of the HFR from HEU to LEU fuels. A computer program, written at Petten and based on information extracted from studies in Europe and the USA, identifies the major cost variables to be manufacturing, uranium, reprocessing and transport costs. Comparison between HEU and LEU cores have been carried out and includes the effects of inflation and exchange rate fluctuations. Conversion of the HFR core to LEU fuels is shown to be financially disadvantageous. (author)

The potential for microbial life in a Canadian high-level nuclear fuel waste disposal vault

International Nuclear Information System (INIS)

Stroes-Gascoyne, S.

1989-12-01

Recent studies have concluded that microbial contamination of a nuclear fuel waste disposal vault is inevitable. Factors that will affect the development of substantial population of micro-organisms include: physiological tolerance of microbes; fluid movement in a vault; availability of nutrients; and availability of energy sources. It is difficult to resolve whether microbial growth will either positively or negatively affect the performance of a vault. One of the necessary steps towards ultimately answering this question is to assess the potential for microbial growth in a disposal vault, based on a nutrient and energy budget. This report gives a quantitative (but conservative) inventory of nutrients and potential energy sources present in a Canadian nuclear fuel waste vault, which hypothetically could support the growth of micro-organisms. Maximum population densities are calculated based on these inventories and assuming that all conditions for microbial growth are optimal, although this will certainly not be the case. Laboratory studies under the vault-relevant conditions are being performed to put realistic boundaries on the calculated numbers. Initial results from these studies, combined with data from a natural analogue site indicate that the calculated population densities could be overestimated by four to five orders of magnitude. Limited data show no effect of the presence of microbes on the transport of Tc, I, and Sr in backfill sand columns. Additional work is needed to address transport effects on buffer and backfill clay columns

Disposition of Highly Enriched Uranium (HEU) and Pu from nuclear weapons

International Nuclear Information System (INIS)

Neff, T.L.

1992-01-01

Last year, as the Soviet Union began to crumble and the Bush-Gorbachev talks had advanced to consider the dismantling of actual warheads instead of mere delivery systems, Dr. Thomas L. Neff, a senior member of MIT's Center for International Studies, tinkered with the notion that one day soon the US could be buying Russian uranium from scrapped nuclear arms. He also considered the costly business of dismantlement and set to work on an ingenious proposal. The basic idea was simply to purchase the HEU from Russia using funds derived from savings in the US Department of Energy's enrichment enterprise. The proposal, now part of an umbrella agreement between the US and Russia announced in early September, promises large strategic benefits for the US in terms of both financing dismantlement (to the tune of $100 million annually), as well as political oversight for the operation itself. In the words of Dr. Neff, who made briefings to both governments on the proposal, open-quotes It's a budget-neutral, win-win solution.close quotes What follows is an illustrated, step-by-step analysis of the proposal, as well as a reprint of Dr. Neff's paper, Disposition of HEU and Pu from Nuclear Weapons, as presented to the Uranium Institute's annual symposium last month

Sealing of process valves for the HEU downblending verification experiment at Portsmouth

International Nuclear Information System (INIS)

Baldwin, G.T.; Bartberger, J.C.; Jenkins, C.D.; Perlinski, A.W.; Schoeneman, J.L.; Gordon, D.M.; Whiting, N.E.; Bonner, T.N.; Castle, J.M.

1998-01-01

At the Portsmouth Gaseous Diffusion Plant in Piketon, Ohio, USA, excess inventory of highly-enriched uranium (HEU) from US defense programs is being diluted to low-enriched uranium (LEU) for commercial use. The conversion is subject to a Verification Experiment overseen by the International Atomic Energy Agency (IAEA). The Verification Experiment is making use of monitoring technologies developed and installed by several DOE laboratories. One of the measures is a system for sealing valves in the process piping, which secures the path followed by uranium hexafluoride gas (UF 6 ) from cylinders at the feed stations to the blend point, where the HEU is diluted with LEU. The Authenticated Item Monitoring System (AIMS) was the alternative proposed by Sandia National Laboratories that was selected by the IAEA. Approximately 30 valves were sealed by the IAEA using AIMS fiber-optic seals (AFOS). The seals employ single-core plastic fiber rated to 125 C to withstand the high-temperature conditions of the heated piping enclosures at Portsmouth. Each AFOS broadcasts authenticated seal status and state-of-health messages via a tamper-protected radio-frequency transmitter mounted outside of the heated enclosure. The messages are received by two collection stations, operated redundantly

Radioactive air emissions notice of construction use of a portable exhauster at 244-AR vault. Revision 2

International Nuclear Information System (INIS)

Carrell, D.J.

1997-01-01

This document serves as a notice of construction (NOC), pursuant to the requirements of Washington Administrative Code (WAC) 246-247-060, and as a request for approval to construct, pursuant to 40 Code of Federal Regulations (CFR) 61.96, a portable exhauster at the 244-AR Vault. The exhauster would be used during air jetting of accumulated liquids from the cell sumps into the tanks and to make transfers among the tanks within the vault when needed. The 244-AR Vault is considered to be a double-contained receiver tank (OCRT) based on its functional characteristics, although it is not listed as one of the five designated DCRTs in the 200 Area Tank Farm systems. Process operations at the vault have been inactive since 1978 and the vault's two stacks have not operated since 1993. Since cessation of vault operations an extremely large amount of rain water and snow melt have accumulated in the cell sumps. The water level in the sumps is substantially above their respective operating levels and there is concern for leakage to the environment through containment failure due to corrosion from backed-up sump liquid. Active ventilation is required to provide contamination control during air jetting operations within the vault. It has been determined that it would not be cost effective to repair the existing exhaust systems to an operational condition; thus, a portable exhauster will be used to support the intermittent operations

The status of HEU to LEU core conversion activities at the Jamaica SLOWPOKE

Energy Technology Data Exchange (ETDEWEB)

Preston, J.; Grant, C., E-mail: john.preston@uwimona.edu.jm [Univ. of the West Indies, Mona Campus, International Centre for Environmental and Nuclear Sciences, Mona (Jamaica)

2012-12-15

The SLOWPOKE reactor in Jamaica has been operated by the International Centre for Environmental and Nuclear Sciences, University of the West Indies since 1984, mainly for the purpose of Neutron Activation Analysis. The HEU core with current utilization has another 14 years of operation, before the addition of a large beryllium annulus would be required to further extend the life-time by 15 years. However, in keeping with the spirit of the Reduced Enrichment for Research and Test Reactors (RERTR) program, the decision was taken in 2003 to convert the core from HEU to LEU, in line with those at the Ecole Polytechnic and RMC SLOWPOKE facilities. This paper reports on the current status of the conversion activities, including key fuel manufacture and regulatory issues, which have seen substantial progress during the last year. A timetable for the complete process is given, and provided that the fuel fabrication can be completed in the estimated 18 months, the core conversion should be accomplished by the end of 2014. (author)

Accounting for cranial vault growth in experimental design.

Science.gov (United States)

Power, Stephanie M; Matic, Damir B; Holdsworth, David W

2014-05-01

Earlier studies have not accounted for continued growth when using the rat calvarial defect model to evaluate bone healing in vivo. The purpose of this study was: 1) to calculate rat cranial vault growth over time; and 2) to determine the effects of accounting for growth on defect healing. Bilateral parietal defects were created in 10 adult Wistar rats. Serial microscopic computerized tomography scans were performed. Bone mineral content (BMC) measured according to standard technique and repeated accounting for cranial growth over time was compared with the use of parametric and nonparametric tests. Cranial vault growth continued through 22 weeks of age, increasing 7.5% in width and 9.1% in length, and calvarial defects expanded proportionately. BMC was greater within defects accounting for growth 2-12 weeks postoperatively (P accounting for cranial growth given advances in serial imaging techniques. Crown Copyright © 2014. Published by Mosby, Inc. All rights reserved.

The evolution of cerebellar tonsillar herniation after cranial vault remodeling surgery.

Science.gov (United States)

Leikola, J; Hukki, A; Karppinen, A; Valanne, L; Koljonen, V

2012-10-01

We sought to examine the pre- and postoperative changes of cerebellar tonsillar herniation by MR imaging in asymptomatic pediatric patients with nonsyndromic, single-suture craniosynostosis (N-SSSC), who underwent cranial vault remodeling surgery without suboccipital decompression. We required cerebellar tonsillar herniation through foramen magnum ≥3 mm for Chiari type I malformation (CMI). We hypothesized that the increase of intracranial volume by cranial vault remodeling would correct the asymptomatic CMI. We identified 9 patients among 121 N-SSSC children undergoing craniofacial surgery from January 2004 to October 2010 with CMI. However, two of them were excluded from the study due to missing postoperative MR images. In the final study population, six were males, five were scaphocephalic, while two were diagnosed with coronal synostosis. In four of the cases, the CMI was decreased in postoperative MR imaging varying from 6 to 12 mm. In three cases, the herniation remained stable. The median change of cerebellar tonsillar herniation was -6.5 mm. We conclude that asymptomatic patients with existing CMI may benefit from cranial vault remodeling surgery alone increasing the intracranial volume.

Backfill formulations for a nuclear fuel waste disposal vault

International Nuclear Information System (INIS)

Yong, R.N.; Boonsinsuk, P.; Wong, G.; Ming, X.D.; Caporuscio, F.; Lytle, P.

1987-01-01

Atomic Energy of Canada Limited and Ontario Hydro are studying the concept of disposing of nuclear fuel wastes in a vault within the Canadian Shield. After nuclear waste containers have been emplaced in a vault, the vault will have to be backfilled permanently. A suitable backfill material should have low hydraulic conductivity and high radionuclide sorption capacity. The research was done with a goal of recommending a specification for formulating this backfill material. This report suggests that such a backfill material should be a mixture of coarse aggregates and swelling clay. Actual trial mixtures were prepared using crushed granite and natural Lake Agassiz clay. Various trial mixtures were subjected to constant-head permeability tests. The results indicate that the hydraulic conductivity of the aggregate-clay mixtures could be close to those of the clay (by itself) when the clay content was in the range of 25% or more. The resulting hydraulic conductivity of about 10 -10 m/s is considered to be low, especially since the maximum grain size is 19.1 mm. Selected mixtures were evaluated for free swell and swelling pressure, both of which increased with increasing clay content. When the clay content was 25%, the free swell was about 4%, compared with 6% for the 100% clay. The corresponding swelling pressure was about 16 kPa - in comparison to 48 kPa for the 100% clay. These results indicate that the proposed backfill material should contain about 25% clay, with a maximum grain size of 19.1 mm. The selected mixture was also tested to evaluate the effects of mixing methods, load-carrying capacity and compaction techniques suitable for the underground vault conditions. The proposed backfill material appeared to perform satisfactorily according to the criteria demanded. The backfill material proposed was further tested for its behaviour during water intake. The unsaturated hydraulic conductivity was found to be approximately 10 -10 m/s and the swelling pressure was

The Application of Carbon Composites in the Rehabilitation of Historic Baroque Vaults

Directory of Open Access Journals (Sweden)

Jiří Witzany

2015-12-01

Full Text Available The stabilization and reinforcement of damaged barrel vaults with lunettes over an arcaded walk, applying composite strips based on high-strength carbon fibers and epoxy resin, was performed during the restoration of a historic monastery. The application of reinforcing composite strips in the soffit of damaged barrel vaults was preceded by relatively extensive experimental research and theoretical analyses. This method significantly reduces the interventions into and the degradation of the original historic structure (surface application and is reversible.

A Protective Vaccine against Chlamydia Genital Infection Using Vault Nanoparticles without an Added Adjuvant.

Science.gov (United States)

Jiang, Janina; Liu, Guangchao; Kickhoefer, Valerie A; Rome, Leonard H; Li, Lin-Xi; McSorley, Stephen J; Kelly, Kathleen A

2017-01-19

Chlamydia trachomatis genital infection is the most common sexually transmitted bacterial disease, causing a significant burden to females due to reproductive dysfunction. Intensive screening and antibiotic treatment are unable to completely prevent female reproductive dysfunction, thus, efforts have become focused on developing a vaccine. A major impediment is identifying a safe and effective adjuvant which induces cluster of differentiation 4 (CD4) cells with attributes capable of halting genital infection and inflammation. Previously, we described a natural nanocapsule called the vault which was engineered to contain major outer membrane protein (MOMP) and was an effective vaccine which significantly reduced early infection and favored development of a cellular immune response in a mouse model. In the current study, we used another chlamydial antigen, a polymorphic membrane protein G-1 (PmpG) peptide, to track antigen-specific cells and evaluate, in depth, the vault vaccine for its protective capacity in the absence of an added adjuvant. We found PmpG-vault immunized mice significantly reduced the genital bacterial burden and histopathologic parameters of inflammation following a C. muridarum challenge. Immunization boosted antigen-specific CD4 cells with a multiple cytokine secretion pattern and reduced the number of inflammatory cells in the genital tract making the vault vaccine platform safe and effective for chlamydial genital infection. We conclude that vaccination with a Chlamydia -vault vaccine boosts antigen-specific immunities that are effective at eradicating infection and preventing reproductive tract inflammation.

A Protective Vaccine against Chlamydia Genital Infection Using Vault Nanoparticles without an Added Adjuvant

Directory of Open Access Journals (Sweden)

Janina Jiang

2017-01-01

Full Text Available Chlamydia trachomatis genital infection is the most common sexually transmitted bacterial disease, causing a significant burden to females due to reproductive dysfunction. Intensive screening and antibiotic treatment are unable to completely prevent female reproductive dysfunction, thus, efforts have become focused on developing a vaccine. A major impediment is identifying a safe and effective adjuvant which induces cluster of differentiation 4 (CD4 cells with attributes capable of halting genital infection and inflammation. Previously, we described a natural nanocapsule called the vault which was engineered to contain major outer membrane protein (MOMP and was an effective vaccine which significantly reduced early infection and favored development of a cellular immune response in a mouse model. In the current study, we used another chlamydial antigen, a polymorphic membrane protein G-1 (PmpG peptide, to track antigen-specific cells and evaluate, in depth, the vault vaccine for its protective capacity in the absence of an added adjuvant. We found PmpG-vault immunized mice significantly reduced the genital bacterial burden and histopathologic parameters of inflammation following a C. muridarum challenge. Immunization boosted antigen-specific CD4 cells with a multiple cytokine secretion pattern and reduced the number of inflammatory cells in the genital tract making the vault vaccine platform safe and effective for chlamydial genital infection. We conclude that vaccination with a Chlamydia-vault vaccine boosts antigen-specific immunities that are effective at eradicating infection and preventing reproductive tract inflammation.

Side loading vault system and method for the disposal of radioactive waste

International Nuclear Information System (INIS)

Meess, D.C.; Jones, B.J.; Mello, R.M.; Weiss, T.G. Jr.; Wright, J.B.

1990-01-01

This patent describes a method for the disposal of hazardous radioactive waste. It comprises: constructing a floor slab in the earth; constructing an elongated wall assembly over the floor slab having sidewalls and a front wall and a back wall at either end the side walls being longer than the front and back walls; providing an accessway in the front wall; constructing a ceiling slab over the wall assembly that is supported at least in part by the wall assembly to form a vault cell; inspecting the vault cell for structural defects, introducing hazardous radioactive waste through the accessway in the front wall and loading the cell with the waste from the back wall to the front wall in rows, each of which is substantially parallel to the back wall to minimize radiation exposure to workers loading the cell, and closing the accessway of the vault cell by constructing a removable wall structure within the accessway

The "Akopian" vault performed by elite male gymnasts: Which biomechanical variables are related to a judge's score?

Directory of Open Access Journals (Sweden)

Roman Farana

2015-03-01

Full Text Available Background: A vaulting performance takes a short time and it is influenced by and affects the quantity of mechanical variables. The significant relationships between the vaulting score and specific aspects of the gymnast's vault should conduct coaches to monitor these variables as a part of training or routine testing. Objective: The aim of the current study was to determine the biomechanical variables that are related to a successful performance of the Akopian vaults performed by top level male gymnasts during the World Cup competition. Methods: Fifteen top-level male gymnasts participated in this study. For the 3D analysis, two digital camcorders with a frame rate of 50 Hz were used. The data were digitized by the Simi motion software. The Hay and Reid method was used to identify the biomechanical variables that determine the linear and angular motions of the handspring and front somersault vaults. A correlation analysis was used to establish the relationship between the biomechanical variables and the judges' scores. The level of statistical significance was determined at the value of p < .05. Results: In the Akopian vaults, in five out of 24 variables arising from the deterministic model showed a significant relationship to the score. A significant correlation was found in the maximum height of the body center of mass in the second flight phase, in the height of the body center of mass at the mat touchdown, in the change of the vertical velocity during the take-off from the vaulting table, and in the duration of the second flight phase. Conclusions: The results of the study suggest that a successful execution of Akopian vaults and the achievement of a higher score required: to maximize the change in vertical velocity in the table contact phase and maximize vertical velocity in the table take-off phase; to maximize the amplitude of the second flight phase, which is determined by the duration of the second flight phase, by the maximum

Introduction of microbial nutrients in a nuclear fuel waste disposal vault as a result of excavation and operation activities

Energy Technology Data Exchange (ETDEWEB)

Stroes-Gascoyne, S; Gascoyne, M; Onagi, D; Thomas, D A; Hamon, C J; Watson, R; Porth, R J

1996-08-01

A nuclear fuel waste disposal vault would not likely be a sterile environment. Bacterial activity would be expected in those areas of the vault conducive to bacterial life, i.e., where effects of heat, moisture content, radiation and compaction would not prevent or severely restrict bacterial life and where suitable and sufficient nutrients would be present. An inventory of bacterial nutrients that would be emplaced `intentionally` with vault materials (fuel waste, waste containers, buffer and backfill materials) has been made previously. This report assesses bacterial nutrients that would be added `inadvertently` to a vault in the form of residues of materials used to excavate and operate a vault. Measurements of blasting material residues in the various water supplies, excavated broken rock (muck) and in cores drilled in old and new tunnel walls were made at AECL`s Underground Research Laboratory. Results show that the largest potential nutrient addition (both carbon and nitrogen) to a vault would result from using untreated excavated broken rock as part of the backfill. (author). 16 refs., 4 tabs., 10 figs.

Introduction of microbial nutrients in a nuclear fuel waste disposal vault as a result of excavation and operation activities

International Nuclear Information System (INIS)

Stroes-Gascoyne, S.; Gascoyne, M.; Onagi, D.; Thomas, D.A.; Hamon, C.J.; Watson, R.; Porth, R.J.

1996-08-01

A nuclear fuel waste disposal vault would not likely be a sterile environment. Bacterial activity would be expected in those areas of the vault conducive to bacterial life, i.e., where effects of heat, moisture content, radiation and compaction would not prevent or severely restrict bacterial life and where suitable and sufficient nutrients would be present. An inventory of bacterial nutrients that would be emplaced 'intentionally' with vault materials (fuel waste, waste containers, buffer and backfill materials) has been made previously. This report assesses bacterial nutrients that would be added 'inadvertently' to a vault in the form of residues of materials used to excavate and operate a vault. Measurements of blasting material residues in the various water supplies, excavated broken rock (muck) and in cores drilled in old and new tunnel walls were made at AECL's Underground Research Laboratory. Results show that the largest potential nutrient addition (both carbon and nitrogen) to a vault would result from using untreated excavated broken rock as part of the backfill. (author). 16 refs., 4 tabs., 10 figs

Disruption of the murine major vault protein (MVP/LRP) gene does not induce hypersensitivity to cytostatics

NARCIS (Netherlands)

M.H. Mossink (Marieke); A. van Zon (Arend); A.A. Fränzel-Luiten (Erna); M. Schoester (Martijn); V.A. Kickhoefer; G.L. Scheffer (George); R.J. Scheper (Rik); P. Sonneveld (Pieter); E.A.C. Wiemer (Erik)

2002-01-01

textabstractVaults are ribonucleoprotein particles with a distinct structure and a high degree of conservation between species. Although no function has been assigned to the complex yet, there is some evidence for a role of vaults in multidrug resistance. To confirm a direct

Hazard evaluation for 244-AR vault facility

International Nuclear Information System (INIS)

BRAUN, D.J.

1999-01-01

This document presents the results of a hazard identification and evaluation performed on the 244-AR Vault Facility to close a USQ (USQ No. TF-98-0785, Potential Inadequacy in Authorization Basis (PIAB): To Evaluate Miscellaneous Facilities Listed in HNF-2503 And Not Addressed In The TWRS Authorization Basis) that was generated as part of an evaluation of inactive TWRS facilities

Addendum to the Composite Analysis for the E-Area Vaults and Saltstone Disposal Facilities

International Nuclear Information System (INIS)

Cook, J.R.

2002-01-01

Revision 1 of the Composite Analysis (CA) Addendum has been prepared to respond to the U.S. Department of Energy (DOE) Low-Level Waste Disposal Facilities Federal Review Group review of the CA. This addendum to the composite analysis responds to the conditions of approval. The composite analysis was performed on the two active SRS low-level radioactive waste disposal facilities. The facilities are the Z-Area Saltstone Disposal Facility and the E-Area Vaults Disposal Facility. The analysis calculated potential releases to the environment from all sources of residual radioactive material expected to remain in the General Separations Area (GSA). The GSA is the central part of the Savannah River Site and contains all of the waste disposal facilities, the chemical separation facilities and associated high-level waste storage facilities, as well as numerous other sources of radioactive material

SIIOS in Alaska - Testing an `In-Vault' Option for a Europa Lander Seismometer.

Science.gov (United States)

Bray, V. J.; Weber, R. C.; DellaGiustina, D. N.; Bailey, H.; Schmerr, N. C.; Pettit, E. C.; Dahl, P.; Albert, D.; Avenson, B.; Byrne, S.; Siegler, M.; Bland, M. T.; Patterson, G. W.; Selznick, S.

2017-12-01

The surface environment of Europa within the radiation-heavy jovian system, poses extreme technical challenges for potential landed missions. The need for radiation shielding and protection from the cold requires instruments to be housed within a thermally insulated and radiation protected `vault'. Unfortunately, this is non-ideal for seismometers as instrument-to-surface coupling is an important factor in the quality of returned data. Delivering a seismic package to an icy world would therefore benefit from the development of a cold-tolerant, radiation-hardened sensor that can survive outside of a protective vault. If such an instrument package were not technologically mature enough, or if lander safety considerations prevent deployment on lander legs, an in-vault location is still a viable option. For such a case, a better understanding of the transmission of seismic signals received through the lander legs is necessary for interpretation of the received signals. The performance, mass, and volume of the `Seismometer to investigate ice and ocean structure' (SIIOS) already meet or exceed flight requirements identified in lander studies for the icy moon Europa. We are testing this flight-candidate in several configurations around and within a lander mock-up, assuming a 1x1 meter vault with extended legs. We compare the received signals from a SIIOS device on the ice with those received by an identical sensor directly above it in the `vault'. We also compare the data from these single-point receivers to that received by two short base-line arrays - A 4-point "in-vault" array and another 4-point array arranged at the ice surface at the base of the lander legs. Our field-testing is performed at Gulkana Glacier, Alaska. The summer melt season provides kilometer-scale regions of coexisting ice, water, and silicate material, thereby providing seismic contrasts analogous to the ice-water layers and possible sub-surface lakes expected at Europa. We demonstrate the

Dispersion, mixing and intentional ignition of hydrogen in the Darlington reactor vault

International Nuclear Information System (INIS)

Lee, J.H.S.; Knystautas, R.

1989-03-01

The present report reviews the Darlington Safety Report (DSR) which has been used as basis for decisions regarding intentional ignition in the Darlington reactor vault. The validity of the assumptions in the DSR regarding mixing of contents is assessed and possible hydrogen release scenarios, specific to the Darlington reactor vault, are examined. The combustion analysis in the DSR vent code calculations are reviewed in the light of existing state of the art information on high speed turbulent flames and transition to detonation. Limitations of the vent code, in this context, are identified and improvements recommended

Experimental modelling of core debris dispersion from the vault under a PWR pressure vessel: Part 1

International Nuclear Information System (INIS)

Macbeth, R.V.; Trenberth, R.

1987-12-01

Modelling experiments have been done on a 1/25 scale model in Perspex of the vault under a PWR pressure vessel. Various liquids have been used to simulate molten core debris assumed to have fallen on to the vault floor from a breach at the bottom of the pressure vessel. High pressure air and helium have been used to simulate the discharge of steam and gas from the breach. The dispersion of liquid via the vault access shafts has been measured. Photographs have been taken of fluid flow patterns and velocity profiles have been obtained. The requirements for further experiments are indicated. (author)

Thickened cranial vault and parasagittal keeling: correlated traits and autapomorphies of Homo erectus?

Science.gov (United States)

Balzeau, Antoine

2013-06-01

Homo erectus sensu lato (s.l.) is a key species in the hominin fossil record for the study of human evolution, being one of the first species discovered and perhaps the most documented, but also because of its long temporal range and having dispersed out of Africa earlier than any other human species. Here I test two proposed autapomorphic traits of H. erectus, namely the increased thickness of the upper cranial vault and parasagittal keeling. The definition of these two anatomical features and their expression and variation among hominids are discussed. The results of this study indicate that the upper vault in Asian H. erectus is not absolutely thicker compared with fossil anatomically modern Homo sapiens, whereas Broken Hill and Petralona have values above the range of variation of H. erectus. Moreover, this anatomical region in Asian H. erectus is not significantly thicker compared with Pan paniscus. In addition, these results demonstrate that cranial vault thickness should not be used to make hypotheses regarding sexual attribution of fossil hominin specimens. I also show that the relation between relief on the external surface of the upper vault, parasagittal keeling and bregmatic eminence, and bone thickness is complex. In this context, the autapomorphic status of the two analysed traits in H. erectus may be rejected. Nevertheless, different patterns in the distribution of bone thickness on the upper vault were identified. Some individual variations are visible, but specificities are observable in samples of different species. The pattern of bone thickness distribution observed in Asian H. erectus, P. paniscus, possibly australopiths, and early Homo or Homo ergaster/erectus appears to be shared by these different species and would be a plesiomorphic trait among hominids. In contrast, two apomorphic states for this feature were identified for Neandertals and H. sapiens. Copyright © 2013 Elsevier Ltd. All rights reserved.

Choosing the right materials for a dry vault store

International Nuclear Information System (INIS)

Walters, J.

1985-01-01

Britain's National Nuclear Corporation has been treating various materials to see if they would be suitable for the construction of a dry vault store for spent fuel and/or vitrified waste. The factors influencing the choice of materials are considered. (UK)

Overexpression of the human major vault protein in gangliogliomas.

NARCIS (Netherlands)

Aronica, E; Gorter, J.A.; Vliet, van EA; Spliet, WG; Veelen, van CW; Rijen, van PC; Leenstra, S.; Ramkema, MD; Scheffer, G.L.; Scheper, R.J.; Sisodiya, SM; Troost, D.

2003-01-01

PURPOSE: Recent evidence has been obtained that the major vault protein (MVP) may play a role in multidrug resistance (MDR). We investigated the expression and cellular localization of MVP in gangliogliomas (GGs), which are increasingly recognized causes of chronic pharmacoresistant epilepsy.

Overexpression of the human major vault protein in gangliogliomas

NARCIS (Netherlands)

Aronica, Eleonora; Gorter, Jan A.; van Vliet, Erwin A.; Spliet, Wim G. M.; van Veelen, Cees W. M.; van Rijen, Peter C.; Leenstra, Sieger; Ramkema, Marja D.; Scheffer, George L.; Scheper, Rik J.; Sisodiya, Sanjay M.; Troost, Dirk

2003-01-01

Purpose: Recent evidence has been obtained that the major vault protein (MVP) may play a role in multidrug resistance (MDR). We investigated the expression and cellular localization of MVP in gangliogliomas (GGs), which are increasingly recognized causes of chronic pharmacoresistant epilepsy.

Data vaults: a database welcome to scientific file repositories

NARCIS (Netherlands)

Ivanova, M.; Kargın, Y.; Kersten, M.; Manegold, S.; Zhang, Y.; Datcu, M.; Espinoza Molina, D.

2013-01-01

Efficient management and exploration of high-volume scientific file repositories have become pivotal for advancement in science. We propose to demonstrate the Data Vault, an extension of the database system architecture that transparently opens scientific file repositories for efficient in-database

Correlation of Palatal Rugoscopy with Gender, Palatal Vault Height ...

African Journals Online (AJOL)

Allen in 1889[3] discussed their role as an identification method. ... a typical orientation pattern, once formed; only changed in its length due to ... along with palatal vault forms as well as ABO blood groups for racial and individualistic soft.

Time-correlated neutron analysis of a multiplying HEU source

Energy Technology Data Exchange (ETDEWEB)

Miller, E.C., E-mail: Eric.Miller@jhuapl.edu [Johns Hopkins University Applied Physics Laboratory, Laurel, MD (United States); Kalter, J.M.; Lavelle, C.M. [Johns Hopkins University Applied Physics Laboratory, Laurel, MD (United States); Watson, S.M.; Kinlaw, M.T.; Chichester, D.L. [Idaho National Laboratory, Idaho Falls, ID (United States); Noonan, W.A. [Johns Hopkins University Applied Physics Laboratory, Laurel, MD (United States)

2015-06-01

The ability to quickly identify and characterize special nuclear material remains a national security challenge. In counter-proliferation applications, identifying the neutron multiplication of a sample can be a good indication of the level of threat. Currently neutron multiplicity measurements are performed with moderated {sup 3}He proportional counters. These systems rely on the detection of thermalized neutrons, a process which obscures both energy and time information from the source. Fast neutron detectors, such as liquid scintillators, have the ability to detect events on nanosecond time scales, providing more information on the temporal structure of the arriving signal, and provide an alternative method for extracting information from the source. To explore this possibility, a series of measurements were performed on the Idaho National Laboratory's MARVEL assembly, a configurable HEU source. The source assembly was measured in a variety of different HEU configurations and with different reflectors, covering a range of neutron multiplications from 2 to 8. The data was collected with liquid scintillator detectors and digitized for offline analysis. A gap based approach for identifying the bursts of detected neutrons associated with the same fission chain was used. Using this approach, we are able to study various statistical properties of individual fission chains. One of these properties is the distribution of neutron arrival times within a given burst. We have observed two interesting empirical trends. First, this distribution exhibits a weak, but definite, dependence on source multiplication. Second, there are distinctive differences in the distribution depending on the presence and type of reflector. Both of these phenomena might prove to be useful when assessing an unknown source. The physical origins of these phenomena can be illuminated with help of MCNPX-PoliMi simulations.

Time-correlated neutron analysis of a multiplying HEU source

Science.gov (United States)

Miller, E. C.; Kalter, J. M.; Lavelle, C. M.; Watson, S. M.; Kinlaw, M. T.; Chichester, D. L.; Noonan, W. A.

2015-06-01

The ability to quickly identify and characterize special nuclear material remains a national security challenge. In counter-proliferation applications, identifying the neutron multiplication of a sample can be a good indication of the level of threat. Currently neutron multiplicity measurements are performed with moderated 3He proportional counters. These systems rely on the detection of thermalized neutrons, a process which obscures both energy and time information from the source. Fast neutron detectors, such as liquid scintillators, have the ability to detect events on nanosecond time scales, providing more information on the temporal structure of the arriving signal, and provide an alternative method for extracting information from the source. To explore this possibility, a series of measurements were performed on the Idaho National Laboratory's MARVEL assembly, a configurable HEU source. The source assembly was measured in a variety of different HEU configurations and with different reflectors, covering a range of neutron multiplications from 2 to 8. The data was collected with liquid scintillator detectors and digitized for offline analysis. A gap based approach for identifying the bursts of detected neutrons associated with the same fission chain was used. Using this approach, we are able to study various statistical properties of individual fission chains. One of these properties is the distribution of neutron arrival times within a given burst. We have observed two interesting empirical trends. First, this distribution exhibits a weak, but definite, dependence on source multiplication. Second, there are distinctive differences in the distribution depending on the presence and type of reflector. Both of these phenomena might prove to be useful when assessing an unknown source. The physical origins of these phenomena can be illuminated with help of MCNPX-PoliMi simulations.

Time-correlated neutron analysis of a multiplying HEU source

International Nuclear Information System (INIS)

Miller, E.C.; Kalter, J.M.; Lavelle, C.M.; Watson, S.M.; Kinlaw, M.T.; Chichester, D.L.; Noonan, W.A.

2015-01-01

The ability to quickly identify and characterize special nuclear material remains a national security challenge. In counter-proliferation applications, identifying the neutron multiplication of a sample can be a good indication of the level of threat. Currently neutron multiplicity measurements are performed with moderated 3 He proportional counters. These systems rely on the detection of thermalized neutrons, a process which obscures both energy and time information from the source. Fast neutron detectors, such as liquid scintillators, have the ability to detect events on nanosecond time scales, providing more information on the temporal structure of the arriving signal, and provide an alternative method for extracting information from the source. To explore this possibility, a series of measurements were performed on the Idaho National Laboratory's MARVEL assembly, a configurable HEU source. The source assembly was measured in a variety of different HEU configurations and with different reflectors, covering a range of neutron multiplications from 2 to 8. The data was collected with liquid scintillator detectors and digitized for offline analysis. A gap based approach for identifying the bursts of detected neutrons associated with the same fission chain was used. Using this approach, we are able to study various statistical properties of individual fission chains. One of these properties is the distribution of neutron arrival times within a given burst. We have observed two interesting empirical trends. First, this distribution exhibits a weak, but definite, dependence on source multiplication. Second, there are distinctive differences in the distribution depending on the presence and type of reflector. Both of these phenomena might prove to be useful when assessing an unknown source. The physical origins of these phenomena can be illuminated with help of MCNPX-PoliMi simulations

WVNS Tank Farm Process Support: Experimental evaluation of an inert gas (nitrogen) to mitigate external corrosion of high-level waste storage tanks

International Nuclear Information System (INIS)

Elmore, M.R.

1996-02-01

Corrosion of the carbon steel waste storage tanks at West Valley Nuclear Services continues to be of concern, especially as the planned duration of waste storage time increases and sludge washing operations are conducted. The external surfaces of Tanks 8D-1 and 8D-2 have been exposed for more than 10 years to water that has intruded into the tank vaults. Visual inspection of the external tank surfaces using a remote video camera has shown indications of heavy corrosion in localized areas on the tank walls. Tests on mild steel specimens under simulated tank vault conditions showed that corrosion is related to the availability of oxygen for the corrosion reactions; consequently, removing oxygen as one of the reactants should effectively eliminate corrosion. In terms of the waste tanks, excluding oxygen from the annular vault space, such as by continuous flushing with an inert gas, should substantially decrease corrosion of the external surfaces of the mild steel tanks (100% exclusion of oxygen is probably not practicable). Laboratory corrosion testing was conducted at Pacific Northwest National Laboratory to give a preliminary assessment of the ability of nitrogen-inerting to reduce steel corrosion. This report summarizes test results obtained after 18-month corrosion tests comparing open-quotes nitrogen-inertedclose quotes corrosion with open-quotes air-equilibratedclose quotes corrosion under simulated tank vault conditions

Heat removal tests on dry storage facilities for nuclear spent fuels

International Nuclear Information System (INIS)

Wataru, M.; Saegusa, T.; Koga, T.; Sakamoto, K.; Hattori, Y.

1999-01-01

In Japan, spent fuel generated in NPP is controlled and stored in dry storage facility away-from reactor. Natural convection cooling system of the storage facility is considered advantageous from both safety and economic point of view. In order to realize this type of facility it is necessary to develop an evaluation method for natural convection characteristics and to make a rational design taking account safety and economic factors. Heat removal tests with the reduces scale models of storage facilities (cask, vault and silo) identified the the flow pattern in the test modules. The temperature and velocity distributions were obtained and the heat transfer characteristics were evaluated

ICPP calcined solids storage facility closure study. Volume III: Engineering design files

International Nuclear Information System (INIS)

1998-02-01

The following information was calculated to support cost estimates and radiation exposure calculations for closure activities at the Calcined Solids Storage Facility (CSSF). Within the estimate, volumes were calculated to determine the required amount of grout to be used during closure activities. The remaining calcine on the bin walls, supports, piping, and floor was also calculated to approximate the remaining residual calcine volumes at different stages of the removal process. The estimates for remaining calcine and vault void volume are higher than what would actually be experienced in the field, but are necessary for bounding purposes. The residual calcine in the bins may be higher than was is experienced in the field as it was assumed that the entire bin volume is full of calcine before removal activities commence. The vault void volumes are higher as the vault roof beam volumes were neglected. The estimations that follow should be considered rough order of magnitude, due to the time constraints as dictated by the project's scope of work. Should more accurate numbers be required, a new analysis would be necessary

ICPP calcined solids storage facility closure study. Volume III: Engineering design files

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-02-01

The following information was calculated to support cost estimates and radiation exposure calculations for closure activities at the Calcined Solids Storage Facility (CSSF). Within the estimate, volumes were calculated to determine the required amount of grout to be used during closure activities. The remaining calcine on the bin walls, supports, piping, and floor was also calculated to approximate the remaining residual calcine volumes at different stages of the removal process. The estimates for remaining calcine and vault void volume are higher than what would actually be experienced in the field, but are necessary for bounding purposes. The residual calcine in the bins may be higher than was is experienced in the field as it was assumed that the entire bin volume is full of calcine before removal activities commence. The vault void volumes are higher as the vault roof beam volumes were neglected. The estimations that follow should be considered rough order of magnitude, due to the time constraints as dictated by the project`s scope of work. Should more accurate numbers be required, a new analysis would be necessary.

Expression profiles of vault components MVP, TEP1 and vPARP and their correlation to other multidrug resistance proteins in ovarian cancer.

Science.gov (United States)

Szaflarski, Witold; Sujka-Kordowska, Patrycja; Pula, Bartosz; Jaszczyńska-Nowinka, Karolina; Andrzejewska, Małgorzata; Zawierucha, Piotr; Dziegiel, Piotr; Nowicki, Michał; Ivanov, Pavel; Zabel, Maciej

2013-08-01

Vaults are cytoplasmic ribonucleoprotein particles composed of three proteins (MVP, TEP1, vPARP) and vault‑associated RNAs (vRNAs). Although the cellular functions of vaults remain unclear, vaults are strongly linked to the development of multidrug resistance (MDR), the major obstacle to the efficient treatment of cancers. Available published data suggest that vaults and their components are frequently upregulated in broad variety of multidrug-resistant cancer cell lines and tumors of different histological origin. Here, we provide detailed analysis of vault protein expression in post-surgery ovarian cancer samples from patients that were not exposed to chemotherapy. Our analysis suggests that vault proteins are expressed in the ovaries of healthy individuals but their expression in cancer patients is changed. Specifically, MVP, TEP1 and vPARP mRNA levels are significantly decreased in cancer samples with tendency of lower expression in higher-grade tumors. The pattern of vault protein mRNA expression is strongly correlated with the expression of other MDR-associated proteins such as MDR1, MRP1 and BCRP. Surprisingly, the protein levels of MVP, TEP1 and vPARP are actually increased in the higher‑grade tumors suggesting existence of post-transcriptional regulation of vault component production.

A comparison of Coulomb and pseudo-Coulomb friction implementations: Application to the table contact phase of gymnastics vaulting.

Science.gov (United States)

Jackson, M I; Hiley, M J; Yeadon, M R

2011-10-13

In the table contact phase of gymnastics vaulting both dynamic and static friction act. The purpose of this study was to develop a method of simulating Coulomb friction that incorporated both dynamic and static phases and to compare the results with those obtained using a pseudo-Coulomb implementation of friction when applied to the table contact phase of gymnastics vaulting. Kinematic data were obtained from an elite level gymnast performing handspring straight somersault vaults using a Vicon optoelectronic motion capture system. An angle-driven computer model of vaulting that simulated the interaction between a seven segment gymnast and a single segment vaulting table during the table contact phase of the vault was developed. Both dynamic and static friction were incorporated within the model by switching between two implementations of the tangential frictional force. Two vaulting trials were used to determine the model parameters using a genetic algorithm to match simulations to recorded performances. A third independent trial was used to evaluate the model and close agreement was found between the simulation and the recorded performance with an overall difference of 13.5%. The two-state simulation model was found to be capable of replicating performance at take-off and also of replicating key contact phase features such as the normal and tangential motion of the hands. The results of the two-state model were compared to those using a pseudo-Coulomb friction implementation within the simulation model. The two-state model achieved similar overall results to those of the pseudo-Coulomb model but obtained solutions more rapidly. Copyright © 2011 Elsevier Ltd. All rights reserved.

Evaluating the potential for large-scale fracturing at a disposal vault: an example using the underground research laboratory

Energy Technology Data Exchange (ETDEWEB)

Martin, C D; Chandler, N A; Brown, Anton

1994-09-01

The potential for large-scale fracturing (> 10 m{sup 2}) around a nuclear fuel waste disposal vault is investigated in this report. The disposal vault is assumed to be located at a depth of 500 m in the plutonic rocks of the Canadian Shield. The rock mass surrounding the disposal vault is considered to have similar mechanical properties and in situ stress conditions to that found at a depth of 420 m at the Underground Research Laboratory. Theoretical, experimental and field evidence shows that Mode I fractures propagate in a plane perpendicular to {sigma}{sub 3} and only if the tensile stress at the tip of the advancing crack is sufficient to overcome the tensile strength of the rock. Because the stress state at a depth of 500 m or more is compressive, and will very probably stay so during the 10,000 year life of the disposal vault, there does not appear to be any mechanism which could propagate large-scale Mode I fracturing in the rock mass surrounding the vault. In addition because {sigma}{sub 3} is near vertical any Mode I fracture propagation that might occur would be in a horizontal plane. The development of either Mode I or large-scale shear fractures would require a drastic change in the compressive in situ stress state at the depth of the disposal vault. The stresses developed as a result of both thermal and glacial loading do not appear sufficient to cause new fracturing. Glacial loading would reduce the shear stresses in the rock mass and hence improve the stability of the rock mass surrounding the vault. Thus, it is not feasible that large-scale fracturing would occur over the 10,000 year life of a disposal vault in the Canadian Shield, at depths of 500 m or greater, where the compressive stress state is similar to that found at the Underground Research Laboratory. 107 refs., 44 figs.

Evaluating the potential for large-scale fracturing at a disposal vault: an example using the underground research laboratory

International Nuclear Information System (INIS)

Martin, C.D.; Chandler, N.A.; Brown, Anton.

1994-09-01

The potential for large-scale fracturing (> 10 m 2 ) around a nuclear fuel waste disposal vault is investigated in this report. The disposal vault is assumed to be located at a depth of 500 m in the plutonic rocks of the Canadian Shield. The rock mass surrounding the disposal vault is considered to have similar mechanical properties and in situ stress conditions to that found at a depth of 420 m at the Underground Research Laboratory. Theoretical, experimental and field evidence shows that Mode I fractures propagate in a plane perpendicular to σ 3 and only if the tensile stress at the tip of the advancing crack is sufficient to overcome the tensile strength of the rock. Because the stress state at a depth of 500 m or more is compressive, and will very probably stay so during the 10,000 year life of the disposal vault, there does not appear to be any mechanism which could propagate large-scale Mode I fracturing in the rock mass surrounding the vault. In addition because σ 3 is near vertical any Mode I fracture propagation that might occur would be in a horizontal plane. The development of either Mode I or large-scale shear fractures would require a drastic change in the compressive in situ stress state at the depth of the disposal vault. The stresses developed as a result of both thermal and glacial loading do not appear sufficient to cause new fracturing. Glacial loading would reduce the shear stresses in the rock mass and hence improve the stability of the rock mass surrounding the vault. Thus, it is not feasible that large-scale fracturing would occur over the 10,000 year life of a disposal vault in the Canadian Shield, at depths of 500 m or greater, where the compressive stress state is similar to that found at the Underground Research Laboratory. 107 refs., 44 figs

Treatment vault shielding for a flattening filter-free medical linear accelerator

Science.gov (United States)

Kry, Stephen F.; Howell, Rebecca M.; Polf, Jerimy; Mohan, Radhe; Vassiliev, Oleg N.

2009-03-01

The requirements for shielding a treatment vault with a Varian Clinac 2100 medical linear accelerator operated both with and without the flattening filter were assessed. Basic shielding parameters, such as primary beam tenth-value layers (TVLs), patient scatter fractions, and wall scatter fractions, were calculated using Monte Carlo simulations of 6, 10 and 18 MV beams. Relative integral target current requirements were determined from treatment planning studies of several disease sites with, and without, the flattening filter. The flattened beam shielding data were compared to data published in NCRP Report No. 151, and the unflattened beam shielding data were presented relative to the NCRP data. Finally, the shielding requirements for a typical treatment vault were determined for a single-energy (6 MV) linac and a dual-energy (6 MV/18 MV) linac. With the exception of large-angle patient scatter fractions and wall scatter fractions, the vault shielding parameters were reduced when the flattening filter was removed. Much of this reduction was consistent with the reduced average energy of the FFF beams. Primary beam TVLs were reduced by 12%, on average, and small-angle scatter fractions were reduced by up to 30%. Head leakage was markedly reduced because less integral target current was required to deliver the target dose. For the treatment vault examined in the current study, removal of the flattening filter reduced the required thickness of the primary and secondary barriers by 10-20%, corresponding to 18 m3 less concrete to shield the single-energy linac and 36 m3 less concrete to shield the dual-energy linac. Thus, a shielding advantage was found when the linac was operated without the flattening filter. This translates into a reduction in occupational exposure and/or the cost and space of shielding.

Treatment vault shielding for a flattening filter-free medical linear accelerator

Energy Technology Data Exchange (ETDEWEB)

Kry, Stephen F; Howell, Rebecca M; Polf, Jerimy; Mohan, Radhe; Vassiliev, Oleg N [Department of Radiation Physics, University of Texas M. D. Anderson Cancer Center, Houston, TX (United States)], E-mail: sfkry@mdanderson.org

2009-03-07

The requirements for shielding a treatment vault with a Varian Clinac 2100 medical linear accelerator operated both with and without the flattening filter were assessed. Basic shielding parameters, such as primary beam tenth-value layers (TVLs), patient scatter fractions, and wall scatter fractions, were calculated using Monte Carlo simulations of 6, 10 and 18 MV beams. Relative integral target current requirements were determined from treatment planning studies of several disease sites with, and without, the flattening filter. The flattened beam shielding data were compared to data published in NCRP Report No. 151, and the unflattened beam shielding data were presented relative to the NCRP data. Finally, the shielding requirements for a typical treatment vault were determined for a single-energy (6 MV) linac and a dual-energy (6 MV/18 MV) linac. With the exception of large-angle patient scatter fractions and wall scatter fractions, the vault shielding parameters were reduced when the flattening filter was removed. Much of this reduction was consistent with the reduced average energy of the FFF beams. Primary beam TVLs were reduced by 12%, on average, and small-angle scatter fractions were reduced by up to 30%. Head leakage was markedly reduced because less integral target current was required to deliver the target dose. For the treatment vault examined in the current study, removal of the flattening filter reduced the required thickness of the primary and secondary barriers by 10-20%, corresponding to 18 m{sup 3} less concrete to shield the single-energy linac and 36 m{sup 3} less concrete to shield the dual-energy linac. Thus, a shielding advantage was found when the linac was operated without the flattening filter. This translates into a reduction in occupational exposure and/or the cost and space of shielding.

Treatment vault shielding for a flattening filter-free medical linear accelerator

International Nuclear Information System (INIS)

Kry, Stephen F; Howell, Rebecca M; Polf, Jerimy; Mohan, Radhe; Vassiliev, Oleg N

2009-01-01

The requirements for shielding a treatment vault with a Varian Clinac 2100 medical linear accelerator operated both with and without the flattening filter were assessed. Basic shielding parameters, such as primary beam tenth-value layers (TVLs), patient scatter fractions, and wall scatter fractions, were calculated using Monte Carlo simulations of 6, 10 and 18 MV beams. Relative integral target current requirements were determined from treatment planning studies of several disease sites with, and without, the flattening filter. The flattened beam shielding data were compared to data published in NCRP Report No. 151, and the unflattened beam shielding data were presented relative to the NCRP data. Finally, the shielding requirements for a typical treatment vault were determined for a single-energy (6 MV) linac and a dual-energy (6 MV/18 MV) linac. With the exception of large-angle patient scatter fractions and wall scatter fractions, the vault shielding parameters were reduced when the flattening filter was removed. Much of this reduction was consistent with the reduced average energy of the FFF beams. Primary beam TVLs were reduced by 12%, on average, and small-angle scatter fractions were reduced by up to 30%. Head leakage was markedly reduced because less integral target current was required to deliver the target dose. For the treatment vault examined in the current study, removal of the flattening filter reduced the required thickness of the primary and secondary barriers by 10-20%, corresponding to 18 m 3 less concrete to shield the single-energy linac and 36 m 3 less concrete to shield the dual-energy linac. Thus, a shielding advantage was found when the linac was operated without the flattening filter. This translates into a reduction in occupational exposure and/or the cost and space of shielding.

Remote tooling for inspection and repair in Pickering NGS-A calandria vault

International Nuclear Information System (INIS)

Hadji-Mirzai, M.; Tokarz, A.; Vandenberg, J.P.

1993-01-01

In recent years it has been necessary to develop capabilities for the inspection and repair of carbon steel components located within calandria vaults at Ontario Hydro's Pickering Nuclear Generating Station 'A'. Concerns about corrosion of piping and some of the structural components have made necessary the development of remote manipulators to inspect and repair carbon steel components within the vaults to ensure continued reliable operation of the units. Remote manipulators for this program have been designed to perform a number of inspection and repair tasks, and several versions have been developed to specialise in detailed inspection techniques and precision tooling module manipulation. (author)

244-AR vault cooling water stream-specific report

International Nuclear Information System (INIS)

1990-08-01

The proposed wastestream designation for the 244-AR Vault cooling water wastestream is that this stream is not a dangerous waste, pursuant to the Washington (State) Administration Code (WAC) 173-303, Dangerous Waste Regulations. A combination of process knowledge and sampling data was used to make this determination. 21 refs., 6 figs., 7 tabs

Radiation protection of cyclotron vault with maze in PET Cyclotron Center

International Nuclear Information System (INIS)

Fueloep, Marko

2003-01-01

The PET Cyclotron center (PCC) is a complex for production, research and application of positron radiopharmaceuticals for PET (Positron Emission Tomography), which was commissioned this year (2004) in Bratislava, Slovak Republic. Positron radionuclides are produced by 18/9 MeV proton/deuteron cyclotron CYCLONE 18/9. Radiation protection of personnel and inhabitants against ionizing radiation in the PCC is solved with regard to the ICRP recommendations and Slovak regulatory system, protection rules and criteria and optimization of radiation protection. In the article comparisons of calculated and measured neutron and gamma dose equivalent rates around the CYCLONE 18/9 and at various points behind the shielding of cyclotron vault with maze are presented. Description of the CYCLONE 18/9 as a source of angular distribution of neutron energy spectra (production of 18 F was considered) was simulated by Monte Carlo code MCNPX. Code MCNP4B was used for shielding calculation of cyclotron vault with maze. Neutron energy spectra behind the shielding were measured by Bonner spectrometer. The values of neutron dose equivalent, which were calculated and measured around the CYCLONE 18/9 and at various points behind the shielding of cyclotron vault with maze, are within the range of factor 2. (authors)

Development of vault model 'VERMIN' for post closure behaviour of repositories for non-heat generating radioactive wastes

International Nuclear Information System (INIS)

1983-10-01

The computer model VERMIN has been developed to simulate the post closure time dependent behaviour of the vault section of a Land 3 and a Land 2 type repository. Development was carried out within the constraints of the computer code SYVAC. Output from the new model, in terms of radionuclide fluxes versus time, provides the source term for that code. A number of conceptual designs for different geological conditions were produced and used to develop the model. Unlike SYVAC, the boundary of the vault was considered to be at the interface between the damaged rock zone and the undamaged host rock. VERMIN treats the vault as a series of engineered barriers namely: waste matrix, waste package, backfill material liner and the damaged rock zone. For the Land 3 repository, the vault was considered to be fully saturated and consequently corrosion, leading to eventual package failure, will occur. VERMIN allows for package failure and subsequent leaching and then calculates the migration of nuclides from within the vault out to its boundary. One dimensional advection and two dimensional diffusion/dispersion are modelled allowing for retardation due to sorption radionuclide saturation and radionuclide decay. Radioactive decay chains up to eight members can be modelled by VERMIN. (author)

SU-E-T-267: Construction and Evaluation of a Neutron Wall to Shield a 15 MV Linac in a Low-Energy Vault.

Science.gov (United States)

Speiser, M; Hager, F; Foster, R; Solberg, T

2012-06-01

To design and quantify the shielding efficacy of an inner Borated Polyethylene (BPE)wall for a 15 MV linac in a low energy vault. A Varian TrueBeam linac with a maximum photon energy of 15 MV was installed in asmaller, preexisting vault. This vault originally housed a low-energy machine and did not havesufficient maze length recommended for neutron attenuation. Effective dose rate calculationswere performed using the Modified Kersey's Method as detailed in NCRP Report No. 151 andfound to be unacceptably high. An initial survey following the machine installation confirmedthese calculations. Rather than restrict the linac beam energy to 10 MV, BPE was investigatedas a neutron moderating addition. An inner wall and door were planned and constructed using4'×8'×1″ thick 5% BPE sheets. The resulting door and wall had 2″ of BPE; conduits and ductwork were also redesigned and shielded. A survey was conducted following construction of thewall. The vault modification reduced the expected effective dose at the vault door from 36.23to 0.010 mSv/week. As specific guidelines for vault modification are lacking, this project quantitativelydemonstrates the potential use of BPE for vault modification. Such modifications may provide alow-cost shielding solution to allow for the use of high energy modes in smaller treatment vaults. © 2012 American Association of Physicists in Medicine.

Correlation of Palatal Rugoscopy with Gender, Palatal Vault Height ...

African Journals Online (AJOL)

Background: Palatal rugae (PR) are asymmetrical irregular elevations, recorded during maxillary cast fabrication , that can be used for identification purpose if previous comparative sources are available. Aim: This study investigated uniqueness of PR patterns in relation to gender, palatal vault forms, and ABO blood groups ...

Melton Valley Storage Tanks Capacity Increase Project, Oak Ridge National Laboratory, Oak Ridge, Tennessee

International Nuclear Information System (INIS)

1995-04-01

The US Department of Energy (DOE) proposes to construct and maintain additional storage capacity at Oak Ridge National Laboratory (ORNL), Oak Ridge, Tennessee, for liquid low-level radioactive waste (LLLW). New capacity would be provided by a facility partitioned into six individual tank vaults containing one 100,000 gallon LLLW storage tank each. The storage tanks would be located within the existing Melton Valley Storage Tank (MVST) facility. This action would require the extension of a potable water line approximately one mile from the High Flux Isotope Reactor (HFIR) area to the proposed site to provide the necessary potable water for the facility including fire protection. Alternatives considered include no-action, cease generation, storage at other ORR storage facilities, source treatment, pretreatment, and storage at other DOE facilities

A new method for generating an invariant iris private key based on the fuzzy vault system.

Science.gov (United States)

Lee, Youn Joo; Park, Kang Ryoung; Lee, Sung Joo; Bae, Kwanghyuk; Kim, Jaihie

2008-10-01

Cryptographic systems have been widely used in many information security applications. One main challenge that these systems have faced has been how to protect private keys from attackers. Recently, biometric cryptosystems have been introduced as a reliable way of concealing private keys by using biometric data. A fuzzy vault refers to a biometric cryptosystem that can be used to effectively protect private keys and to release them only when legitimate users enter their biometric data. In biometric systems, a critical problem is storing biometric templates in a database. However, fuzzy vault systems do not need to directly store these templates since they are combined with private keys by using cryptography. Previous fuzzy vault systems were designed by using fingerprint, face, and so on. However, there has been no attempt to implement a fuzzy vault system that used an iris. In biometric applications, it is widely known that an iris can discriminate between persons better than other biometric modalities. In this paper, we propose a reliable fuzzy vault system based on local iris features. We extracted multiple iris features from multiple local regions in a given iris image, and the exact values of the unordered set were then produced using the clustering method. To align the iris templates with the new input iris data, a shift-matching technique was applied. Experimental results showed that 128-bit private keys were securely and robustly generated by using any given iris data without requiring prealignment.

Homology of the cranial vault in birds: new insights based on embryonic fate-mapping and character analysis

Science.gov (United States)

Maddin, Hillary C.; Piekarski, Nadine; Sefton, Elizabeth M.; Hanken, James

2016-08-01

Bones of the cranial vault appear to be highly conserved among tetrapod vertebrates. Moreover, bones identified with the same name are assumed to be evolutionarily homologous. However, recent developmental studies reveal a key difference in the embryonic origin of cranial vault bones between representatives of two amniote lineages, mammals and birds, thereby challenging this view. In the mouse, the frontal is derived from cranial neural crest (CNC) but the parietal is derived from mesoderm, placing the CNC-mesoderm boundary at the suture between these bones. In the chicken, this boundary is located within the frontal. This difference and related data have led several recent authors to suggest that bones of the avian cranial vault are misidentified and should be renamed. To elucidate this apparent conflict, we fate-mapped CNC and mesoderm in axolotl to reveal the contributions of these two embryonic cell populations to the cranial vault in a urodele amphibian. The CNC-mesoderm boundary in axolotl is located between the frontal and parietal bones, as in the mouse but unlike the chicken. If, however, the avian frontal is regarded instead as a fused frontal and parietal (i.e. frontoparietal) and the parietal as a postparietal, then the cranial vault of birds becomes developmentally and topologically congruent with those of urodeles and mammals. This alternative hypothesis of cranial vault homology is also phylogenetically consistent with data from the tetrapod fossil record, where frontal, parietal and postparietal bones are present in stem lineages of all extant taxa, including birds. It further implies that a postparietal may be present in most non-avian archosaurs, but fused to the parietal or supraoccipital as in many extant mammals.

The cascad spent fuel dry storage facility

International Nuclear Information System (INIS)

Guay, P.; Bonnet, C.

1991-01-01

France has a wide variety of experimental spent fuels different from LWR spent fuel discharged from commercial reactors. Reprocessing such fuels would thus require the development and construction of special facilities. The French Atomic Energy Commission (CEA) has consequently opted for long-term interim storage of these spent fuels over a period of 50 years. Comparative studies of different storage concepts have been conducted on the basis of safety (mainly containment barriers and cooling), economic, modular design and operating flexibility criteria. These studies have shown that dry storage in a concrete vault cooled by natural convection is the best solution. A research and development program including theoretical investigations and mock-up tests confirmed the feasibility of cooling by natural convection and the validity of design rules applied for fuel storage. A facility called CASCAD was built at the CEA's Cadarache Nuclear Research Center, where it has been operational since mid-1990. This paper describes the CASCAD facility and indicates how its concept can be applied to storage of LWR fuel assemblies

Data Vaults: a Database Welcome to Scientific File Repositories

NARCIS (Netherlands)

M.G. Ivanova (Milena); Y. Kargin (Yagiz); M.L. Kersten (Martin); S. Manegold (Stefan); Y. Zhang (Ying); M. Datcu (Mihai); D. Espinoza Molina

2013-01-01

textabstractEfficient management and exploration of high-volume scientific file repositories have become pivotal for advancement in science. We propose to demonstrate the Data Vault, an extension of the database system architecture that transparently opens scientific file repositories for efficient

Numerical and experimental analysis of an in-scale masonry cross-vault prototype up to failure

Energy Technology Data Exchange (ETDEWEB)

Rossi, Michela; Calderini, Chiara; Lagomarsino, Sergio [Department of Civil, Chemical and Environmental Engineering, University of Genoa, Via Montallegro 1, Genoa (Italy); Milani, Gabriele [Department of Architecture, Built Environment and Construction Engineering, Milan Polytechnic University, Piazza Leonardo da Vinci 32, Milan (Italy)

2015-12-31

A heterogeneous full 3D non-linear FE approach is validated against experimental results obtained on an in-scale masonry cross vault assembled with dry joints, and subjected to various loading conditions consisting on imposed displacement combinations to the abutments. The FE model relies into a discretization of the blocks by means of few rigid-infinitely resistant parallelepiped elements interacting by means of planar four-noded interfaces, where all the deformation (elastic and inelastic) occurs. The investigated response mechanisms of vault are the shear in-plane distortion and the longitudinal opening and closing mechanism at the abutments. After the validation of the approach on the experimentally tested cross-vault, a sensitivity analysis is conducted on the same geometry, but in real scale, varying mortar joints mechanical properties, in order to furnish useful hints for safety assessment, especially in presence of seismic action.

Safety aspects of dry spent fuel storage and spent fuel management

International Nuclear Information System (INIS)

Botsch, W.; Smalian, S.; Hinterding, P.; Voelzke, H.; Wolff, D.; Kasparek, E.

2014-01-01

The storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Safety aspects like safe enclosure of radioactive materials, safe removal of decay heat, nuclear criticality safety and avoidance of unnecessary radiation exposure must be achieved throughout the storage period. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. In Germany dual purpose casks for SF or HLW are used for safe transportation and interim storage. TUV and BAM, who work as independent experts for the competent authorities, present the storage licensing process including sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel, based on their long experience in these fields (authors)

Sex determination by three-dimensional geometric morphometrics of the vault and midsagittal curve of the neurocranium in a modern Greek population sample.

Science.gov (United States)

Chovalopoulou, Maria-Eleni; Valakos, Efstratios D; Manolis, Sotiris K

2016-06-01

The aim of this study is to assess sexual dimorphism of adult crania in the vault and midsagittal curve of the vault using three-dimensional geometric morphometric methods. The study sample consisted of 176 crania of known sex (94 males, 82 females) belonging to individuals who lived during the 20th century in Greece. The three-dimensional co-ordinates of 31 ecto-cranial landmarks and 30 semi-landmarks were digitized using a MicroScribe 3DX contact digitizer. Generalized Procrustes analysis (GPA) was used to obtain size and shape variables for statistical analysis. Shape, size and form analyses were carried out by logistic regression and three discriminant function analyses. Results indicate that there are shape differences between sexes. Females in the region of the parietal bones are narrower and the axis forming the frontal and occipital bones is more elongated; the frontal bone is more vertical. Sex-specific shape differences give better classification results in the vault (79%) compared with the midsagittal curve of the neurocranium (68.8%). Size alone yielded better results for cranial vault (82%), while for the midsagittal curve of the vault the result is poorer (68.1%). As anticipated, the classification accuracy improves when both size and shape are combined (89.2% for vault, and 79.4% for midsagittal curve of the vault). These latter findings imply that, in contrast to the midsagittal curve of the neurocranium, the shape of the cranial vault can be used as an indicator of sex in the modern Greek population. Copyright © 2016. Published by Elsevier GmbH.

Fire Hazards Analysis for the 200 Area Interim Storage Area

International Nuclear Information System (INIS)

JOHNSON, D.M.

2000-01-01

This documents the Fire Hazards Analysis (FHA) for the 200 Area Interim Storage Area. The Interim Storage Cask, Rad-Vault, and NAC-1 Cask are analyzed for fire hazards and the 200 Area Interim Storage Area is assessed according to HNF-PRO-350 and the objectives of DOE Order 5480 7A. This FHA addresses the potential fire hazards associated with the Interim Storage Area (ISA) facility in accordance with the requirements of DOE Order 5480 7A. It is intended to assess the risk from fire to ensure there are no undue fire hazards to site personnel and the public and to ensure property damage potential from fire is within acceptable limits. This FHA will be in the form of a graded approach commensurate with the complexity of the structure or area and the associated fire hazards

Vault-Scale Modelling of pH Buffering Capacity in Crushed Granite Backfills

International Nuclear Information System (INIS)

Benbow, Steven; Savage, David; Robinson, Peter; Watson, Sarah

2004-04-01

consist of grains of quartz of uniform size of either 4 or 32 mm diameter. The cement pore fluid diffusing from the waste package was assumed to be pure water saturated with portlandite [Ca(OH) 2 ] at 25 deg C. The pore fluid saturating the backfill was assumed to be pure water equilibrated with quartz at 25 deg C. A number of different simulations were carried out for both 4 and 32 mm diameter gravel backfill grains: 1. A 2D horizontal slice through the vault length with a transmissive feature in the host rock parallel to the vault length, but positioned roughly two vault widths from the vault. 2. As (1) above, but considering a 2D vertical slice through the vault. 3. A 2D horizontal slice through the vault width with a transmissive feature in the host rock normal to the vault length. 4. A 2D horizontal slice through the vault length with a transmissive feature in the host rock intersecting the vault at 45 deg to the vault. An additional simulation was also carried out for the geometry considered in model variant (4) above where reaction of quartz was excluded so that only the reaction of Ca(OH) 2 -saturated fluid with ambient groundwater was simulated. The results of the model variants incorporating reaction of quartz were broadly similar, with few differences apparent for the different orientations of the transmissive feature in the host rock. Most simulations showed that the gravel backfill was capable of maintaining pH 4 moles after 10,000 years in a representative 1 m thick slice through the backfill) were dissolved in the gravel adjacent to the waste package to achieve pH buffering. All of the dissolved quartz was converted to CSH solids (tobermorite and gyrolite). Reactions in simulations with the 4 mm diameter backfill grains were slightly faster, but the results were otherwise identical to those for the 32 mm diameter grains. The model variant with no chemical reaction of the backfill particles was considerably different from the other simulations with pH > 11

Neutronic calculations in core conversion of the IAN-R1 research reactor from MTR HEU to TRIGA LEU fuel

International Nuclear Information System (INIS)

Sarta Fuentes, Jose A.; Castiblanco, L.A.

2003-01-01

With cooperation of the International Atomic Energy Agency (IAEA), neutronic calculations were carried out for conversion of the Ian-R1 Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to establish a staff for neutronic calculation at the Instituto de Cancan's Nucleares y Energia s Alternatives (INEA) a program was established. This program included training, acquisition of hardware, software and calculation for the core with MTR-HEU fuel , enriched nominally to 93% and calculation for several arrangements with the TRIGA-LEU fuel, enriched to 19.7%. The results were verified and compared with several groups of calculation at the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico, and General Atomics (GA) in United States. As a result of this program, several technical reports have been wrote. (author)

Creep in the sparsely fractured rock between a disposal vault and a zone of highly fractured rock

International Nuclear Information System (INIS)

Wilkins, B.J.S.; Rigby, G.L.

1993-08-01

AECL Research is responsible for investigating the feasibility and safety of the disposal of Canada's nuclear fuel waste deep in the plutonic rock of the Canadian Shield. The excavation of the disposal vault, the installation of sealing systems and the heat generated by the fuel waste will all perturb the in situ stress state of the rock mass. This computer codes HOTROK, MCROC and MCDIRC are used to analyze the influence of these stress perturbations on the mechanical behaviour of the rock mass. Time-dependent microcracking of the rock mass will lead to creep around openings in the vault. The analysis specifically estimates the resulting creep strain in the sparsely fractured rock between the edge of the disposal vault and a postulated zone of highly fractured rock. The estimates are extremely conservative. The conclusion reached is that the rock mass more than 3 m beyond the edge of the vault will experience < 0.001 creep strain 100 000 years after the fuel waste is emplaced. (author). 10 refs., 4 tabs., 4 figs

The evaluation, design and implementation of an automated storage and retrieval system for uranium trioxide powder (UO3) at Sellafield

International Nuclear Information System (INIS)

Fitt, C.R.; Mather, K.

1993-01-01

The paper initially sets out the methods used to evaluate the requirements for an automated system to store and retrieve drums of radioactive Uranium Trioxide (UO3) power arising from the Thermal Oxide Reprocessing Plant (THORP) at Sellafield Cumbria. This is followed by a description of the configuration of storage vaults used and of the development of a Self Guided Vehicle (SGV) to operate remotely within these vaults. The system evolved is based on a combination of well proven mechanical equipment and control techniques and the implementation of the design together with testing and control procedures are described. (author)

Safety assessment of a vault-based disposal facility using the ISAM methodology

International Nuclear Information System (INIS)

Kelly, E.; Kim, C.-L.; Lietava, P.; Little, R.; Simon, I.

2002-01-01

As part of the IAEA's Co-ordinated Research Project (CRP) on Improving Long-term of Safety Assessment Methodologies for Near Surface Waste Disposal Facilities (ISAM), three example cases were developed. The aim was to testing the ISAM safety assessment methodology using as realistic as possible data. One of the Test Cases, the Vault Test Case (VTC), related to the disposal of low level radioactive waste (LLW) to a hypothetical facility comprising a set of above surface vaults. This paper uses the various steps of the ISAM safety assessment methodology to describe the work undertaken by ISAM participants in developing the VTC and provides some general conclusions that can be drawn from the findings of their work. (author)

Development of a New Cryptographic Construct Using Palmprint-Based Fuzzy Vault

Directory of Open Access Journals (Sweden)

Amioy Kumar

2009-01-01

Full Text Available The combination of cryptology and biometrics has emerged as promising component of information security. Despite the current popularity of palmprint biometric, there has not been any attempt to investigate its usage for the fuzzy vault. This paper therefore investigates the possible usage of palmprint in fuzzy vault to develop a user friendly and reliable crypto system. We suggest the use of both symmetric and asymmetric approach for the encryption. The ciphertext of any document is generated by symmetric cryptosystem; the symmetric key is then encrypted by asymmetric approach. Further, Reed and Solomon codes are used on the generated asymmetric key to provide some error tolerance while decryption. The experimental results from the proposed approach on the palmprint images suggest its possible usage in an automated palmprint-based key generation system.

SHAPE AND CONSTRUCTION OF BRICK VAULTS. CRITERIA, METHODS AND TOOLS FOR A POSSIBLE CATALOGUE

Directory of Open Access Journals (Sweden)

R. Brumana

2017-05-01

Full Text Available The use of cloister vaults in the construction of noble buildings, as covering elements for square or rectangular rooms, is widespread and well-known. The geometric continuity at the intrados makes generally possible the execution all over the span of frescoes, stucco and decorations. The construction of brick vaults, from the late Middle Age, was sped up by limiting the centering to the wooden planks arches that were instrumental in the profile determination. Nowadays, the availability of several procedures, phases and tools for carrying out a survey allows to draw reliable assumptions about the construction methods and the execution time. It is mandatory to determine the properties of the binders, the shape and dimensions of the bricks, and to carry out a comparison between the geometry of the intrados surface and the evidences emerging at the extrados. The support of the laser scanner technique allows to accurately identify the surface profile and thickness. All these indications, in turn, are useful, in view of an interpretation of the structural behavior, to identify weaknesses, and to highlight contributing factors of instability (if any. The paper focuses on a well-documented case, the Magio Grasselli palace in Cremona in which the cloister vaults of two main rooms show different construction systems, although they were built almost at the same time. The thermographic recordings and laser-scanner surveys highlight the various arrangements used for the cloister vaults.

From the semicircular vault to the flattened vault in masonry bridges. The influence of rise/span ratio and the resistant backfill in the breaking load

Directory of Open Access Journals (Sweden)

J. R. Urruchi-Rojo

2017-04-01

Full Text Available The shape of the vaults in masonry bridges has evolved during the history, so that we can find bibliography where, besides recommending one way of build instead of others depending on the period, is also prescribed the way to carry out the backfill of the extrados. All of this emerged initially as a result of the experiences acquired during the construction and the observation of bridges. Hence, during the 18th century new formulations started to arise, but they led to one result in some cases and to the opposite in others. This article tries to make a comparison between some of these formulations, as well as carrying out an analysis of the influence of both, the lowering rise/span ratio of the vaults and the presence of a resistant backfill in its extrados, on the variation of the breaking load.

Fibrous dysplasia of the cranial vault: quantitative analysis based on neural networks

International Nuclear Information System (INIS)

Arana, E.; Marti-Bonmati, L.; Paredes, R.; Molla, E.

1998-01-01

To assess the utility of statistical analysis and neural networks in the quantitative analysis of fibrous dysplasia of the cranial vault. Ten patients with fibrous dysplasia (six women and four men with a mean age of 23.60±17.85 years) were selected from a series of 167 patients with lesions of the cranial vault evaluated by plain radiography and computed tomography (CT). Nineteen variables were taken from their medical records and radiological study. Their characterization was based on statistical analysis and neural network, and was validated by means of the leave-one-out method. The performance of the neural network was estimated by means of receiver operating characteristics (ROC) curves, using as a parameter the area under the curve A z . Bivariate analysis identified age, duration of symptoms, lytic and sclerotic patterns, sclerotic margin, ovoid shape, soft-tissue mas and periosteal reaction as significant variables. The area under the neural network curve was 0.9601±0.0435. The network selected the matrix and soft-tissue mass a variables that were indispensable for diagnosis. The neural network presents a high performance in the characterization of fibrous dysplasia of the cranial vault, disclosing occult interactions among the variables. (Author) 24 refs

49 CFR 192.187 - Vaults: Sealing, venting, and ventilation.

Science.gov (United States)

2010-10-01

... explosive mixture might be ignited, and there must be a means for testing the internal atmosphere before removing the cover; (2) If the vault or pit is vented, there must be a means of preventing external sources... ventilating effect of a pipe 4 inches (102 millimeters) in diameter; (2) The ventilation must be enough to...

Documentation Experiences for Jamaican SLOWPOKE-2 Conversion from HEU to LEU

International Nuclear Information System (INIS)

Warner, T.-A.; Dennis, H.; Antoine, J.

2015-01-01

The Jamaican SLOWPOKE–2 (JM–1) is a 20 kW research reactor manufactured by Atomic Energy of Canada Limited and has been operating since March 1984, in the department of the International Centre for Environmental and Nuclear Sciences (ICENS), at the University of the West Indies, Mona Campus in Kingston, Jamaica. The pool type reactor has been primarily used for Neutron Activation Analysis in environmental, agricultural, geochemical, health-related studies and mineral exploration. The University, assisted by the IAEA under the GTRI/RERTR program, is currently in the process of converting from HEU to LEU. Extensive documentation on policies, general requirements, elements of the conversion quality assurance (QA) system and conversion QA administrative procedures is required for the conversion. The core conversion activities are being carried out in accordance with current international standards and regulatory guidelines of the newly established Jamaican Radiation Safety Authority (RSA) with agreement between the RSA and IAEA or DOE related to Nuclear Safety and Control. The documentation structure has taken into consideration nuclear safety and licensing, LEU fuel design and conversion analysis, LEU fuel procurement and fabrication, removal of HEU fuel and reactor maintenance and conversion and commissioning, with the conversion QA manual at the apex of the structure. To a large extent, the documentation format will adhere to that of the IAEA applicable regulatory standards and guidance documents. The major challenge of the conversion activities, it is envisioned, will come from the absence of any previous regulatory framework in Jamaica; however, a timeline for the process, which includes training and equipping of regulators, will guide operation. (author)

Expression and proteasomal degradation of the major vault protein (MVP) in mammalian oocytes and zygotes.

Science.gov (United States)

Sutovsky, Peter; Manandhar, Gaurishankar; Laurincik, Jozef; Letko, Juraj; Caamaño, Jose Nestor; Day, Billy N; Lai, Liangxue; Prather, Randall S; Sharpe-Timms, Kathy L; Zimmer, Randall; Sutovsky, Miriam

2005-03-01

Major vault protein (MVP), also called lung resistance-related protein is a ribonucleoprotein comprising a major part (>70%) of the vault particle. The function of vault particle is not known, although it appears to be involved in multi-drug resistance and cellular signaling. Here we show that MVP is expressed in mammalian, porcine, and human ova and in the porcine preimplantation embryo. MVP was identified by matrix-assisted laser-desorption ionization-time-of-flight (MALDI-TOF) peptide sequencing and Western blotting as a protein accumulating in porcine zygotes cultured in the presence of specific proteasomal inhibitor MG132. MVP also accumulated in poor-quality human oocytes donated by infertile couples and porcine embryos that failed to develop normally after in vitro fertilization or somatic cell nuclear transfer. Normal porcine oocytes and embryos at various stages of preimplantation development showed mostly cytoplasmic labeling, with increased accumulation of vault particles around large cytoplasmic lipid inclusions and membrane vesicles. Occasionally, MVP was associated with the nuclear envelope and nucleolus precursor bodies. Nucleotide sequences with a high degree of homology to human MVP gene sequence were identified in porcine oocyte and endometrial cell cDNA libraries. We interpret these data as the evidence for the expression and ubiquitin-proteasome-dependent turnover of MVP in the mammalian ovum. Similar to carcinoma cells, MVP could fulfill a cell-protecting function during early embryonic development.

Grout to meet physical and chemical requirements for closure at Hanford grout vaults. Final report

International Nuclear Information System (INIS)

1994-01-01

The US Army Engineer Waterways Experiment Station (WES) developed a grout based on portland cement, Class F fly ash, and bentonite clay, for the Hanford Grout Vault Program. The purpose of this grout was to fill the void between a wasteform containing 106-AN waste and the vault cover blocks. Following a successful grout development program, heat output, volume change, and compressive strength were monitored with time in simulated repository conditions and in full-depth physical models. This research indicated that the cold-cap grout could achieve and maintain adequate volume stability and other required physical properties in the internal environment of a sealed vault. To determine if contact with 106-AN liquid waste would cause chemical deterioration of the cold-cap grout, cured specimens were immersed in simulated waste. Over a period of 21 days at 150 F, specimens increased in mass without significant changes in volume. X-ray diffraction of reacted specimens revealed crystallization of sodium aluminum silicate hydrate. Scanning electron microscopy used with X-ray fluorescence showed that clusters if this phase had formed in grout pores, increasing grout density and decreasing its effective porosity. Physical and chemical tests collectively indicate a sealing component. However, the Hanford Grout Vault Program was cancelled before completion of this research. This report summarizes close-out Waterways Experiment Station when the Program was cancelled

Grout for closure of the demonstration vault at the US DOE Hanford Facility. Final report

International Nuclear Information System (INIS)

Wakeley, L.D.; Ernzen, J.J.

1992-08-01

The Waterways Experiment Station (WES) developed a grout to be used as a cold- (nonradioactive) cap or void-fill grout between the solidified low-level waste and the cover blocks of a demonstration vault for disposal of phosphate-sulfate waste (PSW) at the US Department of Energy (DOE) Hanford Facility. The project consisted of formulation and evaluation of candidate grouts and selection of the best candidate grout, followed by a physical scale-model test to verify grout performance under project-specific conditions. Further, the project provided data to verify numerical models (accomplished elsewhere) of stresses and isotherms inside the Hanford demonstration vault. Evaluation of unhardened grout included obtaining data on segregation, bleeding, flow, and working time. For hardened grout, strength, volume stability, temperature rise, and chemical compatibility with surrogate wasteform grout were examined. The grout was formulated to accommodate unique environmental boundary conditions (vault temperature = 45 C) and exacting regulatory requirements (mandating less than 0.1% shrinkage with no expansion and no bleeding); and to remain pumpable for a minimum of 2 hr. A grout consisting of API Class H oil-well cement, an ASTM C 618 Class F fly ash, sodium bentonite clay, and a natural sand from the Hanford area met performance requirements in laboratory studies. It is recommended for use in the DOE Hanford demonstration PSW vault

Vaginal vault drainage after complicated single-port access laparoscopic-assisted vaginal hysterectomy

Directory of Open Access Journals (Sweden)

Soo-Jeong Lee

2017-05-01

Conclusion: Vaginal vault drainage could be a safe alternative that allows for the management of postoperative morbidity and retains the advantages of minimally invasive surgery after complicated SPA-LAVH.

At-reactor storage of spent fuel for life-of-plant

International Nuclear Information System (INIS)

Fuierer, A.A.

1990-01-01

The management of commercial spent fuel is a fairly broad topic beginning with the discharge from a reactor, its storage on-site, its transport from the reactor site to a U.S. Department of Energy (DOE) facility, and its ultimate disposal in a geologic repository. This paper discusses spent-fuel management in the at-reactor phase. There are two basic methods for at-reactor storage of spent fuel. The first is wet storage in a pool, and the second is dry storage external to the plant in some form of cask or vault. Spent-fuel consolidation will impact the utility and the DOE waste system. Some of these impacts have a positive effect and some have a negative effect, and each will vary somewhat for each utility. Spent-fuel consolidation and life-of-plant storage will be an increased burden to utilities but will likely result in significant cost savings to the overall waste management system and by proper integration can result in significant institutional benefits

Novel Technique for the UX15 Cavern Vault Support System

CERN Document Server

Rammer, H

2000-01-01

The overall LHC project schedule requires the civil engineering work to begin before the final LEP shutdown. The new caverns for the ATLAS experiment will be built in and around the existing underground structures at point 1. In order to make the best possible use of the time available for the LHC civil engineering before the shutdown of LEP, a particular arrangement for the construction of the UX15 cavern vault has been developed. The basic concept of this arrangement consists of the excavation of the cavern top heading and the installation of the concrete vault immediately afterwards, prior to the subsequent bench excavation after LEP shutdown. A temporary support of the dead weight of the concrete roof will be achieved by the suspension of the roof by 38 no. pre-stressed ground anchors of 225 tons capacity each. This support system will work up to the construction of the cavern base slab and walls and the completion of the permanent concrete lining.

Conceptual design of shaft seals for a nuclear waste disposal vault

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-04-01

The concept of a disposal vault in the Canadian Shield for the effective isolation of nuclear fuel wastes is being assessed as part of the Canadian Nuclear Fuel Waste Management Program. The vault would be accessed from the surface by a number of shafts, which would likely penetrate the vault environment and intersect significant rock fractures and thereby form preferential pathways for the migration of radionuclides from the disposal area to the biosphere. Golder Associates were retained to conduct a conceptual design study of sealing and backfilling the shafts. The first volume of this report reviews current shaft sinking and lining technologies, and recommends the preferred construction methods for the shafts. Factors that could affect the design of a shaft seal system are reviewed, and a conceptual shaft seal is proposed. The second volume addresses the performance assessment of a shaft seal system. While there are no specific performance criteria against which to compare the anticipated containment characteristics of the shaft seal system proposed, the methodology developed for the performance assessment of the reference design should enable the design to be modified to meet performance criteria as they are developed. The report estimates that it will cost $133.7 million in 1986 Canadian dollars to seal three reference shafts, including $18 million for labour and equipment, $103.4 million for backfill and sealing materials, $9.5 million for project indirect costs, and $2.8 million project management. (author). 53 refs., 36 tabs., 43 figs.

Conceptual design of shaft seals for a nuclear waste disposal vault

International Nuclear Information System (INIS)

1993-04-01

The concept of a disposal vault in the Canadian Shield for the effective isolation of nuclear fuel wastes is being assessed as part of the Canadian Nuclear Fuel Waste Management Program. The vault would be accessed from the surface by a number of shafts, which would likely penetrate the vault environment and intersect significant rock fractures and thereby form preferential pathways for the migration of radionuclides from the disposal area to the biosphere. Golder Associates were retained to conduct a conceptual design study of sealing and backfilling the shafts. The first volume of this report reviews current shaft sinking and lining technologies, and recommends the preferred construction methods for the shafts. Factors that could affect the design of a shaft seal system are reviewed, and a conceptual shaft seal is proposed. The second volume addresses the performance assessment of a shaft seal system. While there are no specific performance criteria against which to compare the anticipated containment characteristics of the shaft seal system proposed, the methodology developed for the performance assessment of the reference design should enable the design to be modified to meet performance criteria as they are developed. The report estimates that it will cost $133.7 million in 1986 Canadian dollars to seal three reference shafts, including $18 million for labour and equipment, $103.4 million for backfill and sealing materials, $9.5 million for project indirect costs, and $2.8 million project management. (author). 53 refs., 36 tabs., 43 figs

Operational readiness review of the Low Level Waste vaults at Savannah River Site: A case study

International Nuclear Information System (INIS)

Ahmad, M.; McVay, C.; Venkatesh, S.

1994-01-01

Low Level radioactive Waste (LLW) at the Savannah River Site at Aiken, South Carolina, has traditionally been disposed of using engineered trenches in accordance with the guidelines and technology existing at the time. Recently, subgrade concrete vaults known as E-Area Vaults (EAV) have been constructed at SRS. The EAV project is a comprehensive effort for upgrading LLW disposal at SRS based on meeting the requirements of current Department of Energy (DOE) Orders, and addressing more stringent federal and state regulations. The EAV is a first of its kind state-of-the-art facility designed and built in the United States to receive LLW. Westinghouse Savannah River Company (WSRC) conducted an Operational Readiness Review (ORR) of the vaults prior to startup. The objective of the EAV ORR was to perform a comprehensive review of the operational readiness of the facilities per DOE guidelines including Defense Nuclear Facilities Safety Board (DNFSB) recommendations. This review included assessing construction of the vaults as per design, adequate approved procedures, and training of all the personnel associated with the facility operations. EAV ORR incorporated the lessons learned from other DOE ORRs, included DNFSB recommendations, used a graded approach, and utilized subject matter experts for each functional area of assessment

The effect of organic matter in clay sealing materials on the performance of a nuclear fuel waste disposal vault

International Nuclear Information System (INIS)

Oscarson, D.W.; Stroes-Gascoyne, S.; Cheung, S.C.H.

1986-12-01

The potential effect of organic matter in clay sealing materials on the performance of a nuclear fuel waste disposal vault was examined. The available data indicate that the engineering properties of clays are not significantly affected by the relatively low levels of organic matter (< 1.2 wt.%) present in the clay sealing materials. Complexing of radionuclides by organic substances that are released from the clay sealing materials or produced by microorganisms will likely inhibit rather than promote radionuclide mobility in the compacted sealing materials because of the relatively large size of organic complexing species. Decreasing the level of organic matter in the clay sealing materials will not eliminate microorganisms, and perhaps not decrease their numbers significantly, because chemolithotrophic microorganisms (microorganisms that utilize inorganic forms of C) will be present in a disposal vault. Furthermore, an examination of the nutrient budget in a disposal vault indicates that N, rather than C, will likely be the limiting nutrient for microbial growth. Finally, there is not suitable, proven method for decreasing the level of organic matter in the large amounts of clay needed to seal a vault. It is concluded that the organic matter present in the clay sealing material will not adversely affect the performance of a disposal vault

The potential for vault-induced seismicity in nuclear fuel waste disposal: experience from Canadian mines

International Nuclear Information System (INIS)

Martin, C.D.; Chandler, N.A.

1996-12-01

A seismic event which causes damage to an underground opening is called a rockburst. Practical experience indicates that these damaging seismic events are associated with deep mines where extraction ratios are greater than 0.6. For the arrangement being considered by AECL for nuclear fuel waste disposal vaults, extraction ratios, for the room and pillar design, will be less than 0.3. At this extraction ratio the stress magnitudes will not be sufficient to induce seismic events that can damage the underground openings. Documented world-wide experience shows that unless the underground opening is very close to the source of a naturally occurring seismic event, such as an earthquake, the opening will also not experience any significant damage. Backfilling a disposal vault will improve its resistance to earthquake damage. Backfilling a disposal vault will also reduce the total convergence of the openings caused by thermal loads and hence minimize the potential for thermally-induced seismic events. (author)

TARGET-DIRECTED RUNNING IN GYMNASTICS: THE ROLE OF THE SPRINGBOARD POSITION AS AN INFORMATIONAL SOURCE TO REGULATE HANDSPRINGS ON VAULT

Directory of Open Access Journals (Sweden)

T. Heinen

2011-11-01

Full Text Available Empirical evidence highlights the role of visual information to control gymnastics vaulting and thus neglects a stereotyped approach run. However, there is no evidence on which informational source this regulation is based on. The aim of this study was to examine the position of the springboard as an informational source in the regulation of the handspring on vault. The hypothesis tested was that the action of running towards the springboard brings about changes in the approach run kinematics and handspring kinematics that relate directly to the position of the springboard. Therefore, kinematics of N = 14 female expert gymnasts’ handsprings on vault and their approach runs were examined while manipulating the position of the springboard. The results revealed that expert gymnasts placed their feet on average in the same position on the springboard and adapted to the springboard position during the last three steps of the approach run. A smaller springboard distance to the front edge of the vaulting table resulted in a different hand placement on the vaulting table, a shorter first flight phase, a take-off angle closer to 90° and a longer second flight phase. Findings suggest that the position of the springboard is a relevant informational source in gymnastics vaulting. We state that knowledge about relationships between informational sources in the environment and the resulting regulatory processes in athletes may help coaches to develop specific training programmes in order to optimize performance in complex skills.

The Casa Pizano and the Construction of Light Vaults in Colombia. The Origin of a Modern Tradition

Directory of Open Access Journals (Sweden)

Julián García

2018-05-01

Full Text Available At the beginning of the 20th Century several buildings were covered with tile vaults all through Latin America. However, two houses designed and built by Le Corbusier (the Maisons Sarabhai (Ahmadabad 1955 and Jaoul (Paris 1955 have traditionally been considered the main influence for later vaulted buildings covered with this technique in the area in the 60s and 70s. Since there were built examples contemporary or previous to the buildings by the Swiss architect, it seems logical to think that the influence of the Jaoul or Sarabhai houses in Latin American vault construction might have been somehow less relevant. There are several examples that support that idea: the most relevant is the Casa Pizano, a missing Bogota building designed by the local architect Francisco Pizano de Brigard, which was Le Corbusier's main inspiration for the aforementioned houses. The hypothesis of this paper is that local networks were also essential in the task of connecting similar technical initiatives regarding vault construction. To verify this hypothesis, several examples of buildings from different Latin American countries, starting from the Casa Pizano, will be studied, and some possible contacts between architects will be proposed.

A THERMAL MODEL OF THE IMMOBILIZATION OF LOW-LEVEL RADIOACTIVE WASTE AS GROUT IN CONCRETE VAULTS

Energy Technology Data Exchange (ETDEWEB)

Shadday, M

2008-10-27

Salt solution will be mixed with cement and flyash/slag to form a grout which will be immobilized in above ground concrete vaults. The curing process is exothermic, and a transient thermal model of the pouring and curing process is herein described. A peak temperature limit of 85 C for the curing grout restricts the rate at which it can be poured into a vault. The model is used to optimize the pouring.

International symposium on storage of spent fuel from power reactors. Book of extended synopses

International Nuclear Information System (INIS)

1998-11-01

This book of extended synopses includes papers presented at the International Symposium on Storage of Spent Fuel from Power Reactors organized by IAEA and held in Vienna from 9 to 13 November 1998. It deals with the problems of spent fuel management being an outstanding stage in the nuclear fuel cycle, strategy of interim spent fuel storage, transportation and encapsulation of spent fuel elements from power reactors. Spent fuel storage facilities at reactor sites are always wet while spent fuel storage facilities away from reactor are either wet or dry including casks and vaults. Different design solutions and constructions of storage or transportation casks as well as storing facilities are presented, as well as status of spent fuel storage together with experiences achieved in a number of member states, in the frame of safety, licensing and regulating procedures

International symposium on storage of spent fuel from power reactors. Book of extended synopses

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-11-01

This book of extended synopses includes papers presented at the International Symposium on Storage of Spent Fuel from Power Reactors organized by IAEA and held in Vienna from 9 to 13 November 1998. It deals with the problems of spent fuel management being an outstanding stage in the nuclear fuel cycle, strategy of interim spent fuel storage, transportation and encapsulation of spent fuel elements from power reactors. Spent fuel storage facilities at reactor sites are always wet while spent fuel storage facilities away from reactor are either wet or dry including casks and vaults. Different design solutions and constructions of storage or transportation casks as well as storing facilities are presented, as well as status of spent fuel storage together with experiences achieved in a number of member states, in the frame of safety, licensing and regulating procedures Refs, figs, tabs

Spent fuel storage for ISER plant

International Nuclear Information System (INIS)

Nakajima, Takasuke; Kimura, Yuzi

1987-01-01

ISER is an intrinsically safe reactor basing its safety only on physical laws, and uses a steel reactor vessel in order to be economical. For such a new type reactor, it is essentially important to be accepted by the society by showing that the reactor is more profitable than conventional reactors to the public in both technical and economic viewpoint. It is also important that the reactor raises no serious problem in the total fuel cycle. Reprocessing seems one of the major worldwide fuel cycle issues. Spent fuel storage is also one of the key technologies for fuel cycle back end. Various systems for ISER spent fuel storages are examined in the present report. Spent fuel specifications of ISER are similar to those of LWR and therefore, most of LWR spent fuel technologies are basically applicable to ISER spent fuel. Design requirements and examples of storage facilities are also discussed. Dry storage seems to be preferable for the relatively long cooling time spent fuel like ISER's one from economical viewpoint. Vault storage will possibly be the most advantageous for large storage capacity. Another point for discussion is the location and international collaboration for spent fuel storages: ISER expected to be a worldwide energy source and therefore, international spent fuel management seems to be fairly attractive way for an energy recipient country. (Nogami, K.)

Studies and research concerning BNFP: spent fuel dry storage studies at the Barnwell Nuclear Fuel Plant

International Nuclear Information System (INIS)

Anderson, K.J.

1980-09-01

Conceptual designs are presented utilizing the Barnwell Nuclear Fuel Plant for the dry interim storage of spent light water reactor fuel. Studies were conducted to determine feasible approaches to storing spent fuel by methods other than wet pool storage. Fuel that has had an opportunity to cool for several years, or more, after discharge from a reactor is especially adaptable to dry storage since its thermal load is greatly reduced compared to the thermal load immediately following discharge. A thermal analysis was performed to help in determining the feasibility of various spent fuel dry storage concepts. Methods to reject the heat from dry storage are briefly discussed, which include both active and passive cooling systems. The storage modes reviewed include above and below ground caisson-type storage facilities and numerous variations of vault, or hot cell-type, storage facilities

Selected concrete spent fuel storage cask concepts and the DOE/PSN Cooperative Cask Testing Program

International Nuclear Information System (INIS)

Creer, J.M.; McKinnon, M.A.; Collantes, C.E.

1990-01-01

To date, water pools, metal casks, horizontal concrete modules, and modular vaults have been used to store the major quantity of commercial light water reactor spent nuclear fuel. Recently, vertical concrete dry storage casks have received consideration for storage of spent nuclear fuel. This paper reviews the evolution of the development of selected vertical concrete dry storage casks and outlines a cooperative cask testing (heat transfer and shielding) program involving the US Department of Energy and Pacific Sierra Nuclear Associates. Others participating in the cooperative program are Pacific Northwest Laboratory; EG ampersand G Idaho, Inc.; Wisconsin Electric Power Company; and the Electric Power Research Institute. 28 refs., 14 figs

Alternative disposal technologies for new low-level radioactive waste disposal/storage facilities at the Savannah River Plant

International Nuclear Information System (INIS)

Cook, J.R.

1987-01-01

A Draft Environmental Impact Statement for Waste Management Activities for groundwater protection has been prepared for the Savannah River Plant. Support documentation for the DEIS included an Environmental Information Document on new radioactive waste disposal and storage facilities in which possible alternative disposal technologies were examined in depth. Six technologies that would meet the needs of the Savannah River Plant that selected for description and analysis include near surface disposal, near surface disposal with exceptions, engineered storage, engineered disposal, vault disposal of untreated waste, and a combination of near surface disposal, engineered disposal, and engineered storage. 2 refs

Mobile Pit verification system design based on passive special nuclear material verification in weapons storage facilities

Energy Technology Data Exchange (ETDEWEB)

Paul, J. N.; Chin, M. R.; Sjoden, G. E. [Nuclear and Radiological Engineering Program, George W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, 770 State St, Atlanta, GA 30332-0745 (United States)

2013-07-01

A mobile 'drive by' passive radiation detection system to be applied in special nuclear materials (SNM) storage facilities for validation and compliance purposes has been designed through the use of computational modeling and new radiation detection methods. This project was the result of work over a 1 year period to create optimal design specifications to include creation of 3D models using both Monte Carlo and deterministic codes to characterize the gamma and neutron leakage out each surface of SNM-bearing canisters. Results were compared and agreement was demonstrated between both models. Container leakages were then used to determine the expected reaction rates using transport theory in the detectors when placed at varying distances from the can. A 'typical' background signature was incorporated to determine the minimum signatures versus the probability of detection to evaluate moving source protocols with collimation. This established the criteria for verification of source presence and time gating at a given vehicle speed. New methods for the passive detection of SNM were employed and shown to give reliable identification of age and material for highly enriched uranium (HEU) and weapons grade plutonium (WGPu). The finalized 'Mobile Pit Verification System' (MPVS) design demonstrated that a 'drive-by' detection system, collimated and operating at nominally 2 mph, is capable of rapidly verifying each and every weapon pit stored in regularly spaced, shelved storage containers, using completely passive gamma and neutron signatures for HEU and WGPu. This system is ready for real evaluation to demonstrate passive total material accountability in storage facilities. (authors)

Decontamination of the activation product based on a legal revision of the cyclotron vault room on the non-self-shield compact medical cyclotron

International Nuclear Information System (INIS)

Komiya, Isao; Umezu, Yoshiyuki; Fujibuchi, Toshiou; Nakamura, Kazumasa; Baba, Shingo; Honda, Hiroshi

2016-01-01

The non-self-shield compact medical cyclotron and the cyclotron vault room were in operation for 27 years. They have now been decommissioned. We efficiently implemented a technique to identify an activation product in the cyclotron vault room. Firstly, the distribution of radioactive concentrations in the concrete of the cyclotron vault room was estimated by calculation from the record of the cyclotron operation. Secondly, the comparison of calculated results with an actual measurement was performed using a NaI scintillation survey meter and a high-purity germanium detector. The calculated values were overestimated as compared to the values measured using the Nal scintillation survey meter and the high-purity germanium detector. However, it could limit the decontamination area. By simulating the activation range, we were able to minimize the concrete core sampling. Finally, the appropriate range of radioactivated area in the cyclotron vault room was decontaminated based on the results of the calculation. After decontamination, the radioactive concentration was below the detection limit value in all areas inside the cyclotron vault room. By these procedures, the decommissioning process of the cyclotron vault room was more efficiently performed. (author)

Fuel depletion analyses for the HEU core of GHARR-1: Part II: Fission product inventory

International Nuclear Information System (INIS)

Anim-Sampong, S.; Akaho, E.H.K.; Boadu, H.O.; Intsiful, J.D.K.; Osae, S.

1999-01-01

The fission product isotopic inventories have been estimated for a 90.2% highly enriched uranium (HEU) fuel lattice cell of the Ghana Research Reactor-1 (GHARR-1) using the WIMSD/4 transport lattice code. The results indicate a gradual decrease in the Xe 135 inventory, and saturation trend for Sm 149 , Cs 134 and Cs 135 inventories as the fuel is depleted to 10,000 MWd/tU. (author)

HEU core conversion of Russian production reactors: a major threat to the international RERTR regime

International Nuclear Information System (INIS)

Kuperman, Alan J.; Leventhal, Paul L.

1998-01-01

This paper calls the attention for the major threat to the International Reduced Enrichment for Research and Test Reactors (RERTR) program, represented by the HEU core conversion of russian production reactors. This program aims to reduce and eventually eliminate international civilian commerce in nuclear weapons-usable, highly enriched uranium , and thereby significantly lower risks of the material being stolen or diverted by terrorist or states for producing nuclear weapons

Global ex-situ crop diversity conservation and the Svalbard Global Seed Vault: assessing the current status.

Directory of Open Access Journals (Sweden)

Ola T Westengen

Full Text Available Ex-situ conservation of crop diversity is a global concern, and the development of an efficient and sustainable conservation system is a historic priority recognized in international law and policy. We assess the completeness of the safety duplication collection in the Svalbard Global Seed Vault with respect to data on the world's ex-situ collections as reported by the Food and Agriculture Organization of the United Nations. Currently, 774,601 samples are deposited at Svalbard by 53 genebanks. We estimate that more than one third of the globally distinct accessions of 156 crop genera stored in genebanks as orthodox seeds are conserved in the Seed Vault. The numbers of safety duplicates of Triticum (wheat, Sorghum (sorghum, Pennisetum (pearl millet, Eleusine (finger millet, Cicer (chickpea and Lens (lentil exceed 50% of the estimated numbers of distinct accessions in global ex-situ collections. The number of accessions conserved globally generally reflects importance for food production, but there are significant gaps in the safety collection at Svalbard in some genera of high importance for food security in tropical countries, such as Amaranthus (amaranth, Chenopodium (quinoa, Eragrostis (teff and Abelmoschus (okra. In the 29 food-crop genera with the largest number of accessions stored globally, an average of 5.5 out of the ten largest collections is already represented in the Seed Vault collection or is covered by existing deposit agreements. The high coverage of ITPGRFA Annex 1 crops and of those crops for which there is a CGIAR mandate in the current Seed Vault collection indicates that existence of international policies and institutions are important determinants for accessions to be safety duplicated at Svalbard. As a back-up site for the global conservation system, the Seed Vault plays not only a practical but also a symbolic role for enhanced integration and cooperation for conservation of crop diversity.

Global ex-situ crop diversity conservation and the Svalbard Global Seed Vault: assessing the current status.

Science.gov (United States)

Westengen, Ola T; Jeppson, Simon; Guarino, Luigi

2013-01-01

Ex-situ conservation of crop diversity is a global concern, and the development of an efficient and sustainable conservation system is a historic priority recognized in international law and policy. We assess the completeness of the safety duplication collection in the Svalbard Global Seed Vault with respect to data on the world's ex-situ collections as reported by the Food and Agriculture Organization of the United Nations. Currently, 774,601 samples are deposited at Svalbard by 53 genebanks. We estimate that more than one third of the globally distinct accessions of 156 crop genera stored in genebanks as orthodox seeds are conserved in the Seed Vault. The numbers of safety duplicates of Triticum (wheat), Sorghum (sorghum), Pennisetum (pearl millet), Eleusine (finger millet), Cicer (chickpea) and Lens (lentil) exceed 50% of the estimated numbers of distinct accessions in global ex-situ collections. The number of accessions conserved globally generally reflects importance for food production, but there are significant gaps in the safety collection at Svalbard in some genera of high importance for food security in tropical countries, such as Amaranthus (amaranth), Chenopodium (quinoa), Eragrostis (teff) and Abelmoschus (okra). In the 29 food-crop genera with the largest number of accessions stored globally, an average of 5.5 out of the ten largest collections is already represented in the Seed Vault collection or is covered by existing deposit agreements. The high coverage of ITPGRFA Annex 1 crops and of those crops for which there is a CGIAR mandate in the current Seed Vault collection indicates that existence of international policies and institutions are important determinants for accessions to be safety duplicated at Svalbard. As a back-up site for the global conservation system, the Seed Vault plays not only a practical but also a symbolic role for enhanced integration and cooperation for conservation of crop diversity.

A lime based mortar for thermal insulation of medieval church vaults

DEFF Research Database (Denmark)

Larsen, P.K.; Hansen, Tessa Kvist

A new mortar for thermal insulation of medieval church vaults was tested in a full scale experiment in Annisse Church, DK. The mortar consists of perlite, a highly porous aggregate, mixed with slaked lime. These materials are compatible with the fired clay bricks and the lime mortar joints....... The lambda-value of the insulation mortar is 0.08 W/m K or twice the lambda-value for mineral wool. The water vapour permeability is equal to a medieval clay brick, and it has three times higher capacity for liquid water absorption. The mortar was applied to the top side of the vaults in a thickness of 10 cm......, despite a water vapour pressure gradient up to 500 Pa between the nave and attic. There was no reduction in energy consumption the first winter, possibly due to the increased heat loss related to the drying of the mortar....

Quality assurance plan for placement of cold-cap grout, demonstration vault, Hanford Grout Vault Program. Final report

Energy Technology Data Exchange (ETDEWEB)

Harrington, P.T.; Wakeley, L.D.; Ernzen, J.J.; Walley, D.M.

1992-08-01

During FY 91, the U.S. Army Engineer Waterways Experiment Station (WES) developed a grout to be used as a cold cap, a nonradioactive layer, between the solidified waste and the cover blocks of a demonstration waste disposal vault at the U.S. Department of Energy Hanford Facility. This document recommends requirements for a quality assurance (QA) plan for field mixing and placing of the cold-cap grout during final closure of the demonstration vault. Preplacement activities emphasize selection and testing of materials that will match the performance of materials used in the WES grout. Materials sources and applicable American Society of Testing and Materials, American Concrete Institute, and American Petroleum Institute specifications and requirements are provided. Archiving of physical samples of materials is essential, in addition to careful maintenance of test reports and laboratory data. Full-scale field trial mixing and a detailed preconstruction conference are recommended. Placement activities focus on production and placement of a grout that remains sufficiently constant throughout all batches and meets performance requirements. QA activities must be coordinated between the batch plant and delivery site. Recommended sampling during placement includes cylinders cast for subsequent tests of compressive strength and for nondestructive evaluation and prisms cast for monitoring volume stability. A minimum of two lifts is recommended. Postplacement activities include long-term monitoring of the properties of grout specimens cast during placement. Minimum testing of cylinders includes pulse velocity, fundamental frequency, and unconfined compressive strength. Monitoring characteristics of the microstructure also are recommended. The QA plan should designate an organization to have responsibility for maintaining complete records, reports, and archived samples, including details of deviations from plans written before field placement.

Giant osteoma of the skull vault: A rare case of mixed variety

Directory of Open Access Journals (Sweden)

Harisha P.N.

2016-12-01

Full Text Available Osteoma is the most common primary bone tumor in the craniofacial skeleton. However, most of these are small, asymptomatic and arise from the facial bones or in relation to the paranasal sinuses. Cranial vault osteomas, that too giant and symptomatic are much rarer. We report a case of sixty year-old gentleman presented with a very slowly increasing, painless, hard swelling on the left side of his head. Computerized tomography scan showed the left parietal calvarial tumor to be having large exostotic and enostotic components. He underwent an en-bloc excision of the tumor and cranioplasty. Giant, symptomatic cranial vault osteoma with concurrent exostotic and enostotic components is extremely rare. These lesions can be safely and completely excised with careful planning and attention to detail.

Containers for short-term storage of nuclear materials at the Los Alamos plutonium facility

International Nuclear Information System (INIS)

Hagan, R.; Gladson, J.

1997-01-01

The Los Alamos Plutonium Facility for the past 18 yr has stored nuclear samples for archiving and in support of nuclear materials research and processing programs. In the past several years, a small number of storage containers have been found in a deteriorated condition. A failed plutonium container can cause personnel contamination exposure and expensive physical area decontamination. Containers are stored in a physically secure radiation area vault, making close inspection costly in the form of personnel radiation exposure and work time. A moderate number of these containers are used in support of plutonium processing and must withstand daily handling abuse. A 2-yr evaluation of failed containers and those that have shown no deterioration has been conducted. Based on that study, a program was established to formalize our packing methods and materials and standardize the size and shape of containers that are used for short-term use. A standardized set of containers was designed, evaluated, tested, and procured for use in the facility. This paper reviews our vault storage problems, shows some failed containers, and presents our planned solutions to provide safe and secure containment of nuclear materials

Determination of Dancoff correction thermal utilization and thermal disadvantage factors of HEU and LEU cores of an MNSR

International Nuclear Information System (INIS)

Ofori, Y. T.

2013-07-01

Ghana Research Reactor-1 (GHARR-1), an MNSR (Miniature Neutron Source Reactor) is to be converted from HEU (Highly Enriched Uranium) to LEU (Low Enriched Uranium) fuel, along with all current MNSRs in various other countries. The purpose of the conversion is to minimize the use of HEU for non-proliferation of high-grade nuclear fuel. In this research work, a comparative study has been performed for the determination of the Dancoff, thermal utilization and thermal disadvantage factors of highly enriched uranium (HEU) and potential low enriched uranium (LEU) cores of GHARR-1. A one group transport theory and collision probability based methodologies was used to develop mathematical formulations for thermal utilization factor and thermal disadvantage factor assuming isotropic scattering. This methodology was implemented in a FORTRAN 95 based computer program THERMCALC, which uses Bessell and BesselK as subroutines developed to calculate the modified Bessel functions I n and K n respectively using the polynomial approximation method. Furthermore, a Dancoff correction factor of 0.1519 thermal utilization factor of 0.9767 and a thermal disadvantage factor of 1.894 were obtained for the 90.2% highly enriched Uranium core of GHARR-1. The results compare favorably with literature. Thus THERMCALC can be used as a reliable tool for the calculation of Dancoff, thermal utilization and disadvantage factors of MNSR cores. Other potential LEU cores; UO 2 (with different fuel meat densities and enrichments) and U 3 Si 2 have also been analysed. UO 2 with 12.6% of Uranium-235 was chosen as the most potential LEU core for the GHARR-1. (au)

Dry spent fuel storage licensing

International Nuclear Information System (INIS)

Sturz, F.C.

1995-01-01

In the US, at-reactor-site dry spent fuel storage in independent spent fuel storage installations (ISFSI) has become the principal option for utilities needing storage capacity outside of the reactor spent fuel pools. Delays in the geologic repository operational date at or beyond 2010, and the increasing uncertainty of the US Department of Energy's (DOE) being able to site and license a Monitored Retrievable Storage (MRS) facility by 1998 make at-reactor-site dry storage of spent nuclear fuel increasingly desirable to utilities and DOE to meet the need for additional spent fuel storage capacity until disposal, in a repository, is available. The past year has been another busy year for dry spent fuel storage licensing. The licensing staff has been reviewing 7 applications and 12 amendment requests, as well as participating in inspection-related activities. The authors have licensed, on a site-specific basis, a variety of dry technologies (cask, module, and vault). By using certified designs, site-specific licensing is no longer required. Another new cask has been certified. They have received one new application for cask certification and two amendments to a certified cask design. As they stand on the brink of receiving multiple applications from DOE for the MPC, they are preparing to meet the needs of this national program. With the range of technical and licensing options available to utilities, the authors believe that utilities can meet their need for additional spent fuel storage capacity for essentially all reactor sites through the next decade

Recruitment of the major vault protein by InlK: a Listeria monocytogenes strategy to avoid autophagy.

Directory of Open Access Journals (Sweden)

Laurent Dortet

2011-08-01

Full Text Available L. monocytogenes is a facultative intracellular bacterium responsible for listeriosis. It is able to invade, survive and replicate in phagocytic and non-phagocytic cells. The infectious process at the cellular level has been extensively studied and many virulence factors have been identified. Yet, the role of InlK, a member of the internalin family specific to L. monocytogenes, remains unknown. Here, we first show using deletion analysis and in vivo infection, that InlK is a bona fide virulence factor, poorly expressed in vitro and well expressed in vivo, and that it is anchored to the bacterial surface by sortase A. We then demonstrate by a yeast two hybrid screen using InlK as a bait, validated by pulldown experiments and immunofluorescence analysis that intracytosolic bacteria via an interaction with the protein InlK interact with the Major Vault Protein (MVP, the main component of cytoplasmic ribonucleoproteic particules named vaults. Although vaults have been implicated in several cellular processes, their role has remained elusive. Our analysis demonstrates that MVP recruitment disguises intracytosolic bacteria from autophagic recognition, leading to an increased survival rate of InlK over-expressing bacteria compared to InlK(- bacteria. Together these results reveal that MVP is hijacked by L. monocytogenes in order to counteract the autophagy process, a finding that could have major implications in deciphering the cellular role of vault particles.

EFFECTS OF RUN-UP VELOCITY ON PERFORMANCE, KINEMATICS, AND ENERGY EXCHANGES IN THE POLE VAULT

Directory of Open Access Journals (Sweden)

Nicholas P. Linthorne

2012-06-01

Full Text Available This study examined the effect of run-up velocity on the peak height achieved by the athlete in the pole vault and on the corresponding changes in the athlete's kinematics and energy exchanges. Seventeen jumps by an experienced male pole vaulter were video recorded in the sagittal plane and a wide range of run-up velocities (4.5-8.5 m/s was obtained by setting the length of the athlete's run-up (2-16 steps. A selection of performance variables, kinematic variables, energy variables, and pole variables were calculated from the digitized video data. We found that the athlete's peak height increased linearly at a rate of 0.54 m per 1 m/s increase in run-up velocity and this increase was achieved through a combination of a greater grip height and a greater push height. At the athlete's competition run-up velocity (8.4 m/s about one third of the rate of increase in peak height arose from an increase in grip height and about two thirds arose from an increase in push height. Across the range of run-up velocities examined here the athlete always performed the basic actions of running, planting, jumping, and inverting on the pole. However, he made minor systematic changes to his jumping kinematics, vaulting kinematics, and selection of pole characteristics as the run-up velocity increased. The increase in run-up velocity and changes in the athlete's vaulting kinematics resulted in substantial changes to the magnitudes of the energy exchanges during the vault. A faster run-up produced a greater loss of energy during the take-off, but this loss was not sufficient to negate the increase in run-up velocity and the increase in work done by the athlete during the pole support phase. The athlete therefore always had a net energy gain during the vault. However, the magnitude of this gain decreased slightly as run-up velocity increased

Follow-up after repair of vaginal vault prolapse with abdominal colposacropexy

DEFF Research Database (Denmark)

Pilsgaard, K; Mouritsen, L

1999-01-01

BACKGROUND: Vaginal vault prolapse after hysterectomy is a rare complication, with a prevalence of less than 1/2%. The vaginal depth an axis can be restored by colposacropexy with preservation of coital function. The aim of this follow-up study was to assess the results of colposacropexy with spe...

Is Postoperative Intensive Care Unit Care Necessary following Cranial Vault Remodeling for Sagittal Synostosis?

Science.gov (United States)

Wolfswinkel, Erik M; Howell, Lori K; Fahradyan, Artur; Azadgoli, Beina; McComb, J Gordon; Urata, Mark M

2017-12-01

Of U.S. craniofacial and neurosurgeons, 94 percent routinely admit patients to the intensive care unit following cranial vault remodeling for correction of sagittal synostosis. This study aims to examine the outcomes and cost of direct ward admission following primary cranial vault remodeling for sagittal synostosis. An institutional review board-approved retrospective review was undertaken of the records of all patients who underwent primary cranial vault remodeling for isolated sagittal craniosynostosis from 2009 to 2015 at a single pediatric hospital. Patient demographics, perioperative course, and outcomes were recorded. One hundred ten patients met inclusion criteria with absence of other major medical problems. Average age at operation was 6.7 months, with a mean follow-up of 19.8 months. Ninety-eight patients (89 percent) were admitted to a general ward for postoperative care, whereas the remaining 12 (11 percent) were admitted to the intensive care unit for preoperative or perioperative concerns. Among ward-admitted patients, there were four (3.6 percent) minor complications; however, there were no major adverse events, with none necessitating intensive care unit transfers from the ward and no mortalities. Average hospital stay was 3.7 days. The institution's financial difference in cost of intensive care unit stay versus ward bed was $5520 on average per bed per day. Omitting just one intensive care unit postoperative day stay for this patient cohort would reduce projected health care costs by a total of $540,960 for the study period. Despite the common practice of postoperative admission to the intensive care unit following cranial vault remodeling for sagittal craniosynostosis, the authors suggest that postoperative care be considered on an individual basis, with only a small percentage requiring a higher level of care. Therapeutic, III.

Neutronic calculations for the conversion of the University of Florida Training Reactor from HEU to LEU fuel

International Nuclear Information System (INIS)

Dugan, E.T.; Diaz, N.J.; Kniedler, G.S.

1983-01-01

The University of Florida Training Reactor (UFTR) is located on the University of Florida campus in Gainesville, Florida. The reactor is the Argonaut type, heterogeneous in design and currently fueled with 93% enriched, uranium-aluminum alloy MTR plate-type fuel. Investigations are being performed to examine te feasibility of replacing the highly-enriched fuel of the current UFTR with 4.8% enriched, cylindrical pin SPERT fuel. The SPERT fuel is stainless steel clad and contains uranium dioxide (UO 2 ) pellets. On a broad spectrum, training reactor conversion from high enrichment uranium (HEU) to low enrichment uranium (LEU) fueled facilities has been a continuing concern in the International Atomic Energy Agency (IAEA) and significant work has been done in this area by the Argonne RERTR Program. The International Atomic Energy Agency cites three reasons for reactor conversion to low-enriched uranium. The main reason is the desire to reduce the proliferation potential of research reactor fuels. The second is to increase the assurance of continued fuel availability in the face of probable restrictions on the supply of highly-enriched uranium. The third reason is the possible reduction in requirements for physical security measures during fabrication, transportation, storage and use. This same IAEA report points out that the three reasons stated for the conversion of the fuel of research reactors are interrelated and cannot be considered individually. The concerns of the Nuclear Engineering Sciences Department at the University of Florida relating to the HEU fuel of the UFTR coincide with those of the International Atomic Energy Agency. The primary reason for going to low-enriched pin-type fuel is the concern with proliferation provoked by the highly-enriched plate fuel which has led to tighter security of nuclear facilities such as the UFTR. A second reason for changing to the pin-type fuel is because of difficulties that are being encountered in the supply of the

Neutronic calculations for the conversion of the University of Florida Training Reactor from HEU to LEU fuel

Energy Technology Data Exchange (ETDEWEB)

Dugan, E T; Diaz, N J [Department of Nuclear Engineering Sciences, University of Florida, Gainesville, FL (United States); Kniedler, G S [Reactor Analysis Group, TVA, Chattanooga, TN (United States)

1983-09-01

The University of Florida Training Reactor (UFTR) is located on the University of Florida campus in Gainesville, Florida. The reactor is the Argonaut type, heterogeneous in design and currently fueled with 93% enriched, uranium-aluminum alloy MTR plate-type fuel. Investigations are being performed to examine te feasibility of replacing the highly-enriched fuel of the current UFTR with 4.8% enriched, cylindrical pin SPERT fuel. The SPERT fuel is stainless steel clad and contains uranium dioxide (UO{sub 2}) pellets. On a broad spectrum, training reactor conversion from high enrichment uranium (HEU) to low enrichment uranium (LEU) fueled facilities has been a continuing concern in the International Atomic Energy Agency (IAEA) and significant work has been done in this area by the Argonne RERTR Program. The International Atomic Energy Agency cites three reasons for reactor conversion to low-enriched uranium. The main reason is the desire to reduce the proliferation potential of research reactor fuels. The second is to increase the assurance of continued fuel availability in the face of probable restrictions on the supply of highly-enriched uranium. The third reason is the possible reduction in requirements for physical security measures during fabrication, transportation, storage and use. This same IAEA report points out that the three reasons stated for the conversion of the fuel of research reactors are interrelated and cannot be considered individually. The concerns of the Nuclear Engineering Sciences Department at the University of Florida relating to the HEU fuel of the UFTR coincide with those of the International Atomic Energy Agency. The primary reason for going to low-enriched pin-type fuel is the concern with proliferation provoked by the highly-enriched plate fuel which has led to tighter security of nuclear facilities such as the UFTR. A second reason for changing to the pin-type fuel is because of difficulties that are being encountered in the supply of

A thermal model of the immobilization of low-level radioactive waste as grout in concrete vaults

International Nuclear Information System (INIS)

Shadday, Martin A.

2009-01-01

Salt solution, from radioactive waste generated by the production of plutonium and tritium in nuclear reactors at the Savannah River Site, will be mixed with cement and flyash/slag to form a grout which will be poured into above ground concrete vaults. The curing process is exothermic, and a transient thermal model of the pouring and curing process is herein described. A peak temperature limit of 85 o C for the curing grout restricts the rate at which it can be poured into a vault. The model is used to optimize the pouring.

Spent fuel storage options: a critical appraisal

International Nuclear Information System (INIS)

Singh, K.P.; Bale, M.G.

1990-01-01

The delayed decisions on nuclear fuel reprocessing strategies in the USA and other countries have forced the development of new long-term irradiated fuel storage techniques, to allow a larger volume of fuel to be held on the nuclear station site after removal from the reactor. The nuclear power industry has responded to the challenge by developing several viable options for long-term onsite storage, which can be employed individually or in tandem. They are: densification of storage in the existing spent fuel pool; building another fuel pool facility at the plant site; onsite cask park, and on site vault clusters. Desirable attributes of a storage option are: Safety: minimise the number of fuel handling steps; Economy: minimise total installed, and O and M cost; Security: protection from anti-nuclear protesters; Site adaptability: available site space, earthquake characteristics of the region and so on; Non-intrusiveness: minimise required modifications to existing plant systems; Modularisation: afford the option to adapt a modular approach for staged capital outlays; and Maturity: extent of industry experience with the technology. A critical appraisal is made of each of the four aforementioned storage options in the light of these criteria. (2 figures, 1 table, 4 references) (Author)

Safety assessment for deep underground disposal vault-pathways analysis

International Nuclear Information System (INIS)

Lyon, R.B.; Rosinger, E.L.J.

1980-01-01

The concept verification phase of the Canadian programme for the disposal of nuclear fuel waste encompasses a period of about three years before the start of site selection. During this time, the methodology for Environmental and Safety Assessment studies is being developed by focusing on a model site. Pathways analysis is an important component of these studies. It involves the prediction of the rate at which radionuclides might be released from a disposal vault and travel through the geosphere and biosphere to reach man. The pathways analysis studies cover three major topics: geosphere pathways analysis, biosphere pathways analysis and potentially-disruptive-phenomena analysis. Geosphere pathways analysis includes a total systems analysis, using the computer program GARD2, vault analysis, which considers container failure and waste leaching, hydrogeological modelling and geochemical modelling. Biosphere pathways analysis incorporates a compartmental modelling approach using the computer program RAMM, and a food chain analysis using the computer program FOOD II. Potentially-disruptive-phenomena analysis involves the estimation of the probability and consequences of events such as earthquakes which might reduce the effectiveness of the barriers preventing the release of radionuclides. The current stage of development of the required methodology and data is discussed in each of the three areas and preliminary results are presented. (author)

Experimental investigation of the long term dissolution properties of a cement-based vault backfill

International Nuclear Information System (INIS)

Butcher, E.J.; Borwick, J.; Thorburn, A.A.; Williams, S.J.

2012-01-01

One concept for the long-term management of packages of intermediate-level radioactive waste (ILW) is to place them in underground vaults in a Geological Disposal Facility (GDF). After the packaged waste is placed in the vault it is planned to fill the space around the waste packages with a cement-based backfill prior to closure of the facility. The currently specified backfill is the NRVB (Nirex Reference Vault Backfill), composed of a blend of Portland cement, limestone flour and hydrated lime. Leaching trials are ongoing at the UK National Nuclear Laboratory (NNL) to investigate the dissolution of NRVB until pH values typical of calcium carbonate are achieved. The objective of this work is to determine the buffering capacity of samples of the NRVB, and to allow the sequential release of the alkalinity from the intact samples to be monitored. Leaching is being performed utilising three leachants: deionised water and two saline solutions. Trials have been performed in duplicate for each leachant to allow an initial assessment to be made of the reproducibility of the data produced. In order to simulate the conditions expected in the GDF the approach being used in the trials is a flow through experiment utilising a flexible wall permeameter, within a 35 deg temperature-controlled cell

Assessment of spent WWER-440 fuel performance under long-term storage conditions

Energy Technology Data Exchange (ETDEWEB)

Takats, F [TS Enercon Kft. (Hungary)

2012-07-01

Paks Nuclear Power Plant is the only NPP in Hungary. It has four WWER-440 type reactor units. The fresh fuel is imported from Russia so far. The spent fuel assemblies were shipped back to Russia until 1997 after about 6 years cooling at the plant. A dry storage facility (MVDS type) has been constructed and is operational since then. By 1 January 2008, there were 5107 assemblies in dry storage. The objectives are: 1) Wet AR storage of spent fuel from the NPP Paks: Measurements of conditions for spent fuel storage in the at-reactor (AR) storage pools of Paks NPP (physical and chemical characteristics of pool water, corrosion product data); Measurements and visual control of storage pool component characteristics; Evaluation of storage characteristics and conditions with respect to long-term stability (corrosion of fuel cladding, construction materials); 2) Dry AFR storage at Paks NPP: Calculation and measurement of spent fuel conditions during the transfer from the storage pool to the modular vault dry storage (MVDS) on the site; Calculation and measurement of spent fuel conditions during the preparation of fuel for dry storage (drying process), such as crud release, activity build-up; Measurement of spent fuel conditions during the long-term dry storage, activity data in the storage tubes and amount of crud.

Effect of Plyometric Training on Handspring Vault Performance and Functional Power in Youth Female Gymnasts.

Directory of Open Access Journals (Sweden)

Emma Hall

Full Text Available This study aimed to determine the effect of plyometric training (PT when added to habitual gymnastic training (HT on handspring vault (HV performance variables. Twenty youth female competitive gymnasts (Age: 12.5 ± 1.67 y volunteered to participate and were randomly assigned to two independent groups. The experimental plyometric training group (PTG undertook a six-week plyometric program, involving two additional 45 min PT sessions a week, alongside their HT, while the control group (CG performed regular HT only. Videography was used (120 Hz in the sagittal plane to record both groups performing three HVs for both the baseline and post-intervention trials. Furthermore, participants completed a countermovement jump test (CMJ to assess the effect of PT on functional power. Through the use of Quintic biomechanics software, significant improvements (P < 0.05 were found for the PTG for run-up velocity, take-off velocity, hurdle to board distance, board contact time, table contact time and post-flight time and CMJ height. However, there were no significant improvements on pre-flight time, shoulder angle or hip angle on the vault for the PTG. The CG demonstrated no improvement for all HV measures. A sport-specific PT intervention improved handspring vault performance measures and functional power when added to the habitual training of youth female gymnasts. The additional two hours plyometric training seemingly improved the power generating capacity of movement-specific musculature, which consequently improved aspects of vaulting performance. Future research is required to examine the whether the improvements are as a consequence of the additional volume of sprinting and jumping activities, as a result of the specific PT method or a combination of these factors.

Near-field thermal transient and thermomechanical stress analysis of a disposal vault in crystalline hard rock

International Nuclear Information System (INIS)

Tsui, K.K.; Tsai, A.; Lee, C.F.

1981-01-01

The Canadian Nuclear Fuel Waste Management Program currently focuses on the development of a disposal vault in crystalline hard rock at a reference depth of 1 km below the surface in a suitable pluton in the Canadian Shield. As part of Ontario Hydro's technical assistance to the Atomic Energy of Canada Limited in this program, studies are being carried out to determine the effects of radiogenic heat on the near-field behaviour of a disposal vault. This paper presents the study results obtained to date. Temperature and stress fields were computed and cross-checked by several finite element codes. A comparison between vertical and horizontal borehole emplacement concepts is made. The effects of material non-linearity (temperature dependence) and three-dimensionality on the thermomechanical response are evaluated. Case histories of thermal spalling or fracturing in rock were summarized and discussed to illustrate the possible mechanisms and processes involved in thermal fracturing. An assessment of the thermomechanical stability of the rock mass around a disposal vault under a state of high horizontal in-situ stress is also presented

The solubility of U, Np, Pu, Th and Tc in a geological disposal vault for used nuclear fuel

International Nuclear Information System (INIS)

Lemire, R.J.; Garisto, F.

1989-12-01

This document describes the solubility model used to calculate the concentrations of uranium, thorium, technetium, neptunium and plutonium in a geological disposal vault for used nuclear fuel. This model is incorporated in the vault model of SYVAC3-CC3 - the third generation of the Systems Variability Analysis Code used to assess the long-term safety of the disposal of Canada's nuclear fuel waste. The data for the solubility model and the sources for these data are also reported

Distribution of residual long-lived radioactivity in the inner concrete walls of a compact medical cyclotron vault room.

Science.gov (United States)

Fujibuchi, Toshioh; Nohtomi, Akihiro; Baba, Shingo; Sasaki, Masayuki; Komiya, Isao; Umedzu, Yoshiyuki; Honda, Hiroshi

2015-01-01

Compact medical cyclotrons have been set up to generate the nuclides necessary for positron emission tomography. In accelerator facilities, neutrons activate the concrete used to construct the vault room; this activation increases with the use of an accelerator. The activation causes a substantial radioactive waste management problem when facilities are decommissioned. In the present study, several concrete cores from the walls, ceiling and floor of a compact medical cyclotron vault room were samples 2 years after the termination of operations, and the radioactivity concentrations of radionuclides were estimated. Cylindrical concrete cores 5 cm in diameter and 10 cm in length were bored from the concrete wall, ceiling and floor. Core boring was performed at 18 points. The gamma-ray spectrum of each sample was measured using a high-purity germanium detector. The degree of activation of the concrete in the cyclotron vault room was analyzed, and the range and tendency toward activation in the vault room were examined. (60)Co and (152)Eu were identified by gamma-ray spectrometry of the concrete samples. (152)Eu and (60)Co are produced principally from the stable isotopes of europium and cobalt by neutron capture reactions. The radioactivity concentration did not vary much between the surface of the concrete and at a depth of 10 cm. Although the radioactivity concentration near the target was higher than the clearance level for radioactive waste indicated in IAEA RS-G-1.7, the mean radioactivity concentration in the walls and floor was lower than the clearance level. The radioactivity concentration of the inner concrete wall of the medical cyclotron vault room was not uniform. The areas exceeding the clearance level were in the vicinity of the target, but most of the building did not exceed the clearance levels.

Studies concerning the degradation of concrete vaults for intermediate-level radioactive waste disposal

International Nuclear Information System (INIS)

Duffo, Gustavo S.; Arva, Esteban A; Giordano, Celia M.; Lafont, Claudio J.

2007-01-01

The National Atomic Energy Commission (CNEA) is the responsible for developing a management nuclear waste disposal programme. This programme contemplates the design and construction of a facility for the final disposal of intermediate-level radioactive wastes. The proposed model is the near-surface monolithic repository similar to those in operation in El Cabril, Spain. The design of this type of repository is based on the use of multiple, independent and redundant barriers. Since the vault and cover are major components of the engineered barriers, the durability of this concrete structures is an important aspect for the facilities integrity. This work presents a laboratory and field investigation performed for the last 6 years on reinforced concrete specimens, in order to predict the service life of the intermediate level radioactive waste disposal vaults from data obtained from electrochemical techniques. On the other hand, the development of sensors that allow on-line measurements of rebar corrosion potential and corrosion current density; incoming oxygen flow that reaches the metal surface; concrete electrical resistivity and chloride concentration is shown. Those sensors, properly embedded in a new full scale vault (nowadays in construction), will allow the monitoring of the corrosion process of the steel rebars embedded in the structure. All the information obtained from the sensors will be used for the final design of the container in order to achieve a service life more or equal than the foreseen durability for this type of facilities. (author) [es

GOTHIC CHURCHES IN PARIS ST GERVAIS ET ST PROTAIS IMAGE MATCHING 3D RECONSTRUCTION TO UNDERSTAND THE VAULTS SYSTEM GEOMETRY

Directory of Open Access Journals (Sweden)

M. Capone

2015-02-01

benefits and the troubles. From a methodological point of view this is our workflow: - theoretical study about geometrical configuration of rib vault systems; - 3D model based on theoretical hypothesis about geometric definition of the vaults' form; - 3D model based on image matching 3D reconstruction methods; - comparison between 3D theoretical model and 3D model based on image matching;

IRT‑Sofia, HEU to LEU conversion: regulatory approval tasks solution overview

International Nuclear Information System (INIS)

Mitev, Mladen; Belousov, Sergey; Dimitrov, Dobromir

2014-01-01

The HEU to LEU conversion of the IRT–Sofia research reactor of the Institute for Nuclear Research and Nuclear Energy of the Bulgarian Academy of Sciences was jointly studied with the Argonne National Laboratory as a part of the RERTR Programme. The main purpose of the collaboration consisted in accomplishment of safety analyses and preparation of documents used for regulatory approval tasks solution. The main steps and results which are fundamental for the preparation of IRT–Sofia Safety Analyses Report including Operating Limits and Conditions are presented in this paper. The documents prepared by INRNE in accordance with the European nuclear safety requirements and IAEA recommendations were submitted for approval to the Bulgarian Nuclear Regulatory Agency at the end of 2010. Key words: research reactor, safety analyses report, Nuclear Regulatory Agency

Lessons learned from 50 years period the storage of the spent fuel from nuclear research reactor VVR-S

International Nuclear Information System (INIS)

Dragusin, M.

2010-01-01

The nuclear research reactor VVR-S was commissioned in July 1957. This reactor is in permanent shutdown since December 1997 and will be decommissioned. The duration of the decommissioning project is 11 years. The first year of decommissioning project is 2010. The spent nuclear fuels resulting from the 40 years of operating the nuclear research reactor are stored under wet conditions. The chemical and physical water parameters monitored are: transparency, conductibility, pH, chloride content, oxygen content, temperature, dry residual content, Al, Mn, Mg, Fe, Vn, Cr. Residual dry content must be maintained in requested range in order to prevent degradation and corrosion both of the clads, assemblies and linen material of the ponds. Two types of the nuclear fuel assemblies were used: LEU type -EK-10 and HEU type S-36 Russian origin. All spent nuclear fuel assemblies HEU-S-36 type were repatriated in Russian Federation in June 2009 in safety and security conditions without any problems due of the wet storage, after 25 years storage in wet conditions. The spent nuclear fuel assemblies types LEU EK-10 were stored in wet conditions more than 50 years. This paper describes the lessons learned during the 50 years management of the spent nuclear fuel resulted from the operation the research reactor VVR-S. The management was based on the maintenance of water parameters by water filtration, using at all times air HEPA filter incorporated in technological ventilation system and by monitoring the level, temperature, physical and chemical parameters of the water storage from ponds and by controlling ponds linen physical integrity. Also we have used the discs having the same compositions with materials from assemblies stored in the same ponds, in order to verify degradation and corrosion phenomena induced due to the quality of storage water. The paper will described these results obtained by metallographic, visual, XRF analysis onto discs and dry residual samples from storage

From laboratory experiments to a geological disposal vault: calculation of used nuclear fuel dissolution rates

International Nuclear Information System (INIS)

Sunder, S.; Shoesmith, D.W.; Kolar, M.; Leneveu, D.M.

1998-01-01

Calculation of used nuclear fuel dissolution rates in a geological disposal vault requires a knowledge of the redox conditions in the vault. For redox conditions less oxidizing than those causing UO 2 oxidation to the U 3 O 7 , stage, a thermodynamically-based model is appropriate. For more oxidizing redox conditions a kinetic or an electrochemical model is needed to calculate these rates. The redox conditions in a disposal vault will be affected by the radiolysis of groundwater by the ionizing radiation associated with the fuel. Therefore, we have calculated the alpha-, beta- and gamma-dose rates in water in contact with the reference used fuel in the Canadian Nuclear Fuel Waste Management Program (CNFWMP) as a function of cooling time. Also, we have determined dissolution rates of UO 2 fuel as a function of alpha and gamma dose rates from our electrochemical measurements. These room-temperature rates are used to calculate the dissolution rates of used fuel at 100 o C, the highest temperature expected in a container in the CNFWMP, as a function of time since emplacement. It is shown that beta radiolysis of water will be the main cause of oxidation of used CANDU fuel in a failed container. The use of a kinetic or an electrochemical corrosion model, to calculate fuel dissolution rates, is required for a period of ∼1000 a following emplacement of copper containers in the geologic disposal vault envisaged in the CNFWMP. Beyond this time period a thermodynamically-based model adequately predicts the fuel dissolution rates. The results presented in this paper can be adopted to calculate used fuel dissolution rates for other used UO 2 fuels in other waste management programs. (author)

ITER pressure and thermal loads to containment HTS vault LOCA analysis. Draft final report EC Task SEA 3, Subtask 3-4

Energy Technology Data Exchange (ETDEWEB)

Blomquist, R; Shen, K; Sjoeberg, A

1995-03-01

This study has been performed within the framework of the EC Task SEA 3 and its objective is to provide necessary data in supporting the design solution of the ITER secondary confinement around the primary heat transfer system equipment. These data relate to the required dimensions for the blow-out panels, the vent lines and the suppression tank following a LOCA in one of the HTS vaults, namely the first wall/shielding blanket(FW/SB) vault, divertor vault and vacuum vessel (VV) vault. In this report, we present the design and operational input and describe the identified accident sequences. The input data are in correspondence with ITER design data of November 1994. The computer codes used are RELAP5 (LOCA flows) and CONTAIN (secondary confinement thermal-hydraulics) and models of calculations are given. The results in the form of diagrams demonstrating transients of various variables after a LOCA, are presented. After some discussions of the results, we indicate some topics for the continuing study with the emphasis on optimization of the containment system. 10 refs, 29 figs.

ITER pressure and thermal loads to containment HTS vault LOCA analysis. Draft final report EC Task SEA 3, Subtask 3-4

International Nuclear Information System (INIS)

Blomquist, R.; Shen, K.; Sjoeberg, A.

1995-03-01

This study has been performed within the framework of the EC Task SEA 3 and its objective is to provide necessary data in supporting the design solution of the ITER secondary confinement around the primary heat transfer system equipment. These data relate to the required dimensions for the blow-out panels, the vent lines and the suppression tank following a LOCA in one of the HTS vaults, namely the first wall/shielding blanket(FW/SB) vault, divertor vault and vacuum vessel (VV) vault. In this report, we present the design and operational input and describe the identified accident sequences. The input data are in correspondence with ITER design data of November 1994. The computer codes used are RELAP5 (LOCA flows) and CONTAIN (secondary confinement thermal-hydraulics) and models of calculations are given. The results in the form of diagrams demonstrating transients of various variables after a LOCA, are presented. After some discussions of the results, we indicate some topics for the continuing study with the emphasis on optimization of the containment system. 10 refs, 29 figs

The effects of vault drainage on postoperative morbidity after vaginal hysterectomy for benign gynaecological disease: a randomised controlled trial.

Science.gov (United States)

Dua, A; Galimberti, A; Subramaniam, M; Popli, G; Radley, S

2012-02-01

To evaluate the efficacy of vault drainage in reducing the immediate postoperative morbidity associated with vaginal hysterectomy carried out for benign gynaecological conditions. Randomised controlled trial. A tertiary referral gynaecology centre in UK. A total of 272 women who underwent vaginal hysterectomy for benign conditions between March 2005 and June 2010. The 272 women were randomised to have a drain inserted or not inserted, 'drain' or 'no drain', respectively, before vault closure during vaginal hysterectomy, using a sealed envelope technique. The surgical procedures were performed using the surgeons' standard technique and postoperative care was delivered according to the unit's protocol. The primary outcome measure was reduction in postoperative febrile morbidity. Secondary outcome measures were hospital readmission rate, blood transfusion, change in postoperative haemoglobin and length of stay. In all, 135 women were randomised to have a drain and 137 to 'no drain'. There were no differences in the incidence of febrile morbidity, length of stay, change in haemoglobin or need for postoperative blood transfusion between the two groups. The routine use of vault drain at vaginal hysterectomy for benign disorders has no significant effect on postoperative morbidity. The use of vault drain in this context is not recommended. © 2011 The Authors BJOG An International Journal of Obstetrics and Gynaecology © 2011 RCOG.

A design study for a fission product fixation plant and storage at Windscale

International Nuclear Information System (INIS)

Hill, K.M.; Ridley, G.; Adler, D.G.

1961-01-01

A flowsheet for a production plant has been worked out and the problems of designing a reliable production plant utilising it have been assessed. The properties of current glass compositions have been considered in relation to problems of permanent storage for periods of 800 to 1000 years and various storage proposals assessed. The performance of the preferred type of storage, i.e. dry storage in natural draught air-cooled vaults, has been estimated and consideration given to a number of variables affecting the optimum design. The design problems of a storage of the required life have been investigated in relation to suitable types of site and proposals made as to the types of construction to be used. Design work on tunnel loading facilities has been carried to the state of demonstrating feasibility. The capital costs and the economics of this type of waste storage have been investigated for unalloyed Magnox fuel and brief consideration given to alloyed and cermet fuels. Attention has been given to development problems of the whole system where they arise. (author)

Nuclear and radiological safety in the substitution process of the fuel HEU to LEU 30/20 in the Reactor TRIGA Mark III of the ININ

International Nuclear Information System (INIS)

Hernandez G, J.

2012-10-01

Inside the safety initiative in the international ambit, with the purpose of reducing the risks associated with the use of high enrichment nuclear fuels (HEU) for different proposes to the peaceful uses of the nuclear energy, Mexico contributes by means of the substitution of the high enrichment fuel HEU for low enrichment fuel LEU 30/20 in the TRIGA Mark III Reactor, belonging to Instituto Nacional de Investigaciones Nucleares (ININ). The conversion process was carried out by means of the following activities: analysis of the proposed core, reception and inspection of the fuel LEU 30/20, the discharge of the fuels of the mixed reactor core, shipment of the fuels HEU fresh and irradiated to the origin country, reload activities with the fuels LEU 30/20 and parameters measurement of the core operation. In order to maintaining the personnel's integrity and infrastructure associated to the Reactor, during the whole process the measurements of nuclear and radiological safety were controlled to detail, in execution with the license requirements of the installation. This work describes the covering activities and radiological inspections more relevant, as well as the measurements of radiological control implemented with base in the estimate of the equivalent dose of the substitution process. (Author)

Radiological Assessment for the Vance Road Facility Source Vault, Oak Ridge Institute for Science and Education, Oak Ridge, Tennessee

International Nuclear Information System (INIS)

Morton, J. R.

2000-01-01

From the 1950s, the Vance Road laboratories had been used for a broad range of nuclear medicine research involving numerous radionuclides. These radionuclides were stored in the a source vault located on the first floor of the facility. The Environmental Survey and Site Assessment Program (ESSAP) of ORISE performed a radiological assessment survey of the source vault after it had been remediated and in preparation for converting the area to office space

Radiological Assessment Survey of the Vance road Facility Source Vault Building Materials, Oak Ridge Institute for Science and Education, Oak Ridge, Tennessee

International Nuclear Information System (INIS)

Morton, J. R.

2000-01-01

From the 1950s, the Vance Road laboratory was the site of extensive nuclear medical research and involved the used of numerous radionuclides. These nuclides were stored in a source vault stored on the first floor of the facility. Nuclear medical research is no longer conducted in this facility, and the source vault was remediated in preparation for converting the area to office space and general use. The Environmental Survey and Site Assessment Program (ESSAP) of ORISE performed a radiological assessment survey of the source vault and its associated miscellaneous building materials and laboratory equipment in preparation for the conversion to general use space

Microbially influenced corrosion of copper nuclear fuel waste containers in a Canadian disposal vault

Energy Technology Data Exchange (ETDEWEB)

King, F

1996-11-01

An assessment of the potential for microbially influenced corrosion (MIC) of copper nuclear fuel waste containers in a Canadian disposal vault is presented. The assessment is based on a consideration of the microbial activity within a disposal vault, the reported cases of MIC of Cu alloys in the literature and the known corrosion behaviour of Cu. Because of the critical role of biofilms in the reported cases of MIC, their formation and properties are discussed in detail. Next, the literature on the MIC of Cu alloys is briefly reviewed. The various MIC mechanisms proposed are critically discussed and the implications for the corrosion of Cu containers considered. In the majority of literature cases, MIC depends on alternating aerated and deaerated environments, with accelerated corrosion being observed when fresh aerated water replaces stagnant water, e.g., the MIC of Cu-Ni heat exchangers in polluted seawater and the microbially influenced pitting of Cu water pipes. Finally, because of the predominance of corrosion by sulphate-reducing bacteria (SRB) in the MIC literature, the abiotic behaviour of Cu alloys in sulphide solutions is also reviewed. The effect of the evolving environment in a disposal vault on the extent and location of microbial activity is discussed. Biofilm formation on the container surface is considered unlikely throughout the container lifetime, but especially initially when the environmental conditions will be particularly aggressive. Microbial activity in areas of the vault away from the container is possible, however. Corrosion of the container could then occur if microbial metabolic by-products diffuse to the container surface. Sulphide, produced by the action of SRB are considered to be the most likely cause of container corrosion. It is concluded that the only likely form of MIC of Cu containers will result from sulphide produced by SRB diffusing to the container surface. A modelling procedure for predicting the extent of corrosion is

Microbially influenced corrosion of copper nuclear fuel waste containers in a Canadian disposal vault

International Nuclear Information System (INIS)

King, F.

1996-11-01

An assessment of the potential for microbially influenced corrosion (MIC) of copper nuclear fuel waste containers in a Canadian disposal vault is presented. The assessment is based on a consideration of the microbial activity within a disposal vault, the reported cases of MIC of Cu alloys in the literature and the known corrosion behaviour of Cu. Because of the critical role of biofilms in the reported cases of MIC, their formation and properties are discussed in detail. Next, the literature on the MIC of Cu alloys is briefly reviewed. The various MIC mechanisms proposed are critically discussed and the implications for the corrosion of Cu containers considered. In the majority of literature cases, MIC depends on alternating aerated and deaerated environments, with accelerated corrosion being observed when fresh aerated water replaces stagnant water, e.g., the MIC of Cu-Ni heat exchangers in polluted seawater and the microbially influenced pitting of Cu water pipes. Finally, because of the predominance of corrosion by sulphate-reducing bacteria (SRB) in the MIC literature, the abiotic behaviour of Cu alloys in sulphide solutions is also reviewed. The effect of the evolving environment in a disposal vault on the extent and location of microbial activity is discussed. Biofilm formation on the container surface is considered unlikely throughout the container lifetime, but especially initially when the environmental conditions will be particularly aggressive. Microbial activity in areas of the vault away from the container is possible, however. Corrosion of the container could then occur if microbial metabolic by-products diffuse to the container surface. Sulphide, produced by the action of SRB are considered to be the most likely cause of container corrosion. It is concluded that the only likely form of MIC of Cu containers will result from sulphide produced by SRB diffusing to the container surface. A modelling procedure for predicting the extent of corrosion is

Dry storage facility for spent fuel or high-level wastes

International Nuclear Information System (INIS)

Geoffroy, J.; Dobremelle, M.; Fabre, J.C.; Bonnet, C.

1989-01-01

The French Atomic Energy Commission (CEA) has specific irradiated fuels which, due to their properties, cannot be reprocessed directly in existing industrial facilities. Accordingly, for the spent fuels from the EL4 and OSIRIS power plants, the CEA has been faced with the problem of selecting a process that will allow the storage of these materials under satisfactory technical and economic conditions. The authors discuss how three conditions must be satisfied to store irradiated fuels releasing heat: containment of radioactive materials, biological shielding, and thermal cooling to guarantee an acceptable temperature- level throughout. In view of the need for an interim storage facility using a simple cooling process requiring only minimal maintenance and monitoring, dry storage in a concrete vault cooled by natural convection was selected. This choice was made within the framework of a research and development program in which theoretical heat transfer investigations and mock-up tests confirmed the feasibility of cooling by natural convection

EVALUATION OF SULFATE ATTACK ON SALTSTONE VAULT CONCRETE AND SALTSTONESIMCO TECHNOLOGIES, INC. PART1 FINAL REPORT

International Nuclear Information System (INIS)

Langton, C.

2008-01-01

This report summarizes the preliminary results of a durability analysis performed by SIMCO Technologies Inc. to assess the effects of contacting saltstone Vaults 1/4 and Disposal Unit 2 concretes with highly alkaline solutions containing high concentrations of dissolved sulfate. The STADIUM(reg s ign) code and data from two surrogate concretes which are similar to the Vaults 1/4 and Disposal Unit 2 concretes were used in the preliminary durability analysis. Simulation results for these surrogate concrete mixes are provided in this report. The STADIUM(reg s ign) code will be re-run using transport properties measured for the SRS Vaults 1/4 and Disposal Unit 2 concrete samples after SIMCO personnel complete characterization testing on samples of these materials. Simulation results which utilize properties measured for samples of Vaults 1/4 and Disposal Unit 2 concretes will be provided in Revision 1 of this report after property data become available. The modeling performed to date provided the following information on two concrete mixes that will be used to support the Saltstone PA: (1) Relationship between the rate of advancement of the sulfate front (depth of sulfate ion penetration into the concrete) and the rate of change of the concrete permeability and diffusivity. (2) Relationship between the sulfate ion concentration in the corrosive leachate and the rate of the sulfate front progression. (3) Equation describing the change in hydraulic properties (hydraulic conductivity and diffusivity) as a function of sulfate ion concentration in the corrosive leachate. These results have been incorporated into the current Saltstone PA analysis by G. Flach (Flach, 2008). In addition, samples of the Saltstone Vaults 1/4 and Disposal Unit 2 concretes have been prepared by SIMCO Technologies, Inc. Transport and physical properties for these materials are currently being measured and sulfate exposure testing to three high alkaline, high sulfate leachates provided by SRNL is

Addendum to the composite analysis for the E-Area Vaults and Saltstone Disposal Facilities

Energy Technology Data Exchange (ETDEWEB)

Cook, J.R.

2000-03-13

This report documents the composite analysis performed on the two active SRS low-level radioactive waste disposal facilities. The facilities are the Z-Area Saltstone Disposal Facility and the E-Area Vaults Disposal Facility.

Addendum to the composite analysis for the E-Area Vaults and Saltstone Disposal Facilities

International Nuclear Information System (INIS)

Cook, J.R.

2000-01-01

This report documents the composite analysis performed on the two active SRS low-level radioactive waste disposal facilities. The facilities are the Z-Area Saltstone Disposal Facility and the E-Area Vaults Disposal Facility

Cranial vault trauma and selective mortality in medieval to early modern Denmark

DEFF Research Database (Denmark)

Boldsen, Jesper L; Milner, George R; Weise, Svenja

2015-01-01

to interpersonal violence in past populations. Three medieval to early modern Danish skeletal samples are used to estimate the effect of selective mortality on males with cranial vault injuries who survived long enough for bones to heal. The risk of dying for these men was 6.2 times higher than...

Why does the need of HEU for high flux research reactors remain?

International Nuclear Information System (INIS)

Glaeser, W.

1991-01-01

It has shown that high performance high flux reactors need an ongoing supply of highly enriched uranium. The new fuel materials in their highly enriched version offer prospective for advanced and better neutron sources vital for the future of neutron research. This is another very attractive result of the RERTR programme. One-sided restriction would only provide marginal or no values for research. If we adopt the sometimes expressed views that high enriched RERTR developed fuel should only be made available when unique benefits to mankind could be obtained, then certainly basic research at the forefront belongs to this category. HEU would only pose theoretical difficulties, if it would remain under proper safeguards and obviously this is the way to be pursued. (orig.)

Improving the Aesthetic Outcome in Scaphocephaly Correction: Hairline Lowering During Vault Remodeling Procedures.

Science.gov (United States)

Wittig, Joern; Duncan, Christian

2017-06-01

The bossed forehead in patients with scaphocephaly often leads to a high hairline. A new technique to improve the aesthetic outcome of patients undergoing scaphocephaly correction is described. Sixteen patients with scaphocephaly and having a high hairline due to frontal bossing who underwent scaphocephaly correction by subtotal or total vault remodeling were analyzed. The median age at surgery was 18 months. The mean distance between the nasofrontal suture and the hairline was preoperatively 70 mm (range 58-91). An obvious lowering of the hairline could be achieved in all 16 patients. The mean postoperative distance nasofrontal suture to hairline was 59 mm (range 50-73). There were no complications associated with the technique. The hairline lowering technique is a useful addition to vault remodeling techniques and can improve the postoperative aesthetic appearance considerably. The authors recommend this technique in scaphocephaly patients, who present with a high hairline.

Adding EUNIS and VAULT rocket data to the VSO with Modern Perl frameworks

Science.gov (United States)

Mansky, Edmund

2017-08-01

A new Perl code is described, that uses the modern Object-oriented Moose framework, to add EUNIS and VAULT rocket data to the Virtual Solar Observatory website. The code permits the easy fixing of FITS header fields in the case where some FITS fields that are required are missing from the original data files. The code makes novel use of the Moose extensions “before” and “after” to build in dependencies so that database creation of tables occurs before the loading of data, and that the validation of file-dependent tables occurs after the loading is completed. Also described is the computation and loading of the deferred FITS field CHECKSUM into the database following the loading and validation of the file-dependent tables. The loading of the EUNIS 2006 and 2007 flight data, and the VAULT 2.0 flight data is described in detail as illustrative examples.

Influence of the program of prophylaxis of violations of vaults of foot on physical preparedness of children of midchildhood

Directory of Open Access Journals (Sweden)

Valkevich A.V.

2012-06-01

Full Text Available Purpose of work - to probe influence of the program of prophylaxis of violations of vaults of foot on physical preparedness of children of midchildhood. 40 children which studied in a tourist class took part in experiment. All of children on the state a health were attributed to the basic group. In the process of testing determined the indexes of speed (run 30 meters, flexibility (forerake from position «sitting», adroitness (shuttle run 4x9 meters, speed-power qualities (broad jump from a place, force (undercutting on a low cross-beam and endurance (run, meters. The substantive provisions of the program of prophylaxis of violations of vaults of foot are reflected for children. Certain and analysed dynamics of indexes of physical preparedness of schoolboys on completion of the program of prophylaxis of violations of vaults of foot in educational process.

Radiological performance assessment for the E-Area Vaults Disposal Facility

International Nuclear Information System (INIS)

Cook, J.R.

1994-01-01

These document contains appendices A-M for the performance assessment. They are A: details of models and assumptions, B: computer codes, C: data tabulation, D: geochemical interactions, E: hydrogeology of the Savannah River Site, F: software QA plans, G: completeness review guide, H: performance assessment peer review panel recommendations, I: suspect soil performance analysis, J: sensitivity/uncertainty analysis, K: vault degradation study, L: description of naval reactor waste disposal, M: porflow input file

324 Building radiochemical engineering cells, high-level vault, low-level vault, and associated areas closure plan

International Nuclear Information System (INIS)

Barnett, J.M.

1998-01-01

The Hanford Site, located adjacent to and north of Richland, Washington, is operated by the US Department of Energy, Richland Operations Office (RL). The 324 Building is located in the 300 Area of the Hanford Site. The 324 Building was constructed in the 1960s to support materials and chemical process research and development activities ranging from laboratory/bench-scale studies to full engineering-scale pilot plant demonstrations. In the mid-1990s, it was determined that dangerous waste and waste residues were being stored for greater than 90 days in the 324 Building Radiochemical Engineering Cells (REC) and in the High-Level Vault/Low-Level Vault (HLV/LLV) tanks. [These areas are not Resource Conservation and Recovery Act of 1976 (RCRA) permitted portions of the 324 Building.] Through the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Milestone M-89, agreement was reached to close the nonpermitted RCRA unit in the 324 Building. This closure plan, managed under TPA Milestone M-20-55, addresses the identified building areas targeted by the Tri-Party Agreement and provides commitments to achieve the highest degree of compliance practicable, given the special technical difficulties of managing mixed waste that contains high-activity radioactive materials, and the physical limitations of working remotely in the areas within the subject closure unit. This closure plan is divided into nine chapters. Chapter 1.0 provides the introduction, historical perspective, 324 Building history and current mission, and the regulatory basis and strategy for managing the closure unit. Chapters 2.0, 3.0, 4.0, and 5.0 discuss the detailed facility description, process information, waste characteristics, and groundwater monitoring respectively. Chapter 6.0 deals with the closure strategy and performance standard, including the closure activities for the B-Cell, D-Cell, HLV, LLV; piping and miscellaneous associated building areas. Chapter 7.0 addresses the

Isotope Enrichment Detection by Laser Ablation - Laser Absorption Spectrometry: Automated Environmental Sampling and Laser-Based Analysis for HEU Detection

International Nuclear Information System (INIS)

Anheier, Norman C.; Bushaw, Bruce A.

2010-01-01

The global expansion of nuclear power, and consequently the uranium enrichment industry, requires the development of new safeguards technology to mitigate proliferation risks. Current enrichment monitoring instruments exist that provide only yes/no detection of highly enriched uranium (HEU) production. More accurate accountancy measurements are typically restricted to gamma-ray and weight measurements taken in cylinder storage yards. Analysis of environmental and cylinder content samples have much higher effectiveness, but this approach requires onsite sampling, shipping, and time-consuming laboratory analysis and reporting. Given that large modern gaseous centrifuge enrichment plants (GCEPs) can quickly produce a significant quantity (SQ ) of HEU, these limitations in verification suggest the need for more timely detection of potential facility misuse. The Pacific Northwest National Laboratory (PNNL) is developing an unattended safeguards instrument concept, combining continuous aerosol particulate collection with uranium isotope assay, to provide timely analysis of enrichment levels within low enriched uranium facilities. This approach is based on laser vaporization of aerosol particulate samples, followed by wavelength tuned laser diode spectroscopy to characterize the uranium isotopic ratio through subtle differences in atomic absorption wavelengths. Environmental sampling (ES) media from an integrated aerosol collector is introduced into a small, reduced pressure chamber, where a focused pulsed laser vaporizes material from a 10 to 20-(micro)m diameter spot of the surface of the sampling media. The plume of ejected material begins as high-temperature plasma that yields ions and atoms, as well as molecules and molecular ions. We concentrate on the plume of atomic vapor that remains after the plasma has expanded and then cooled by the surrounding cover gas. Tunable diode lasers are directed through this plume and each isotope is detected by monitoring absorbance

On the tenth value distance of the photon field along the maze of high-energy linear accelerator vaults.

Science.gov (United States)

Han, Zhaohui; Chin, Lee M

2018-03-01

There is a wide range in the reported photon tenth value distance (TVD) in the maze of high-energy linear accelerator vaults. In order to gain insight into the appropriate use of the TVD value during door design, we performed measurements of the photon dose in the maze of four vaults. In addition, our study represents the first to describe a scenario where an inner borated polyethylene (BPE) door for neutron shielding is installed in the maze downstream to Point A, the point on the maze centerline that is just visible from the isocenter. The measurements were made along the maze centerline at 1 m above the floor. In all cases, the accelerator operated at a nominal energy of 15 MV. Of the four vaults, three were equipped with an inner BPE door at a distance of 1.0-2.1 m downstream to Point A. The door was made of 10.16 cm (4″) BPE sandwiched between two 0.635 cm (1/4″) steel face plates. The photon dose in the maze without a BPE door decreases exponentially with a characteristic TVD of 6 m beyond a distance of 2.5 m from Point A. The presence of a BPE door in an identical vault not only reduces the photon intensity in the maze by about an order of magnitude, but also softens the energy spectrum with a shortened TVD of 4.7 m, significantly lessening the shielding burden at the outer maze entrance. In contrast to the common use of Point A as the reference point to specify distance, the photon dose in the maze with a BPE door located downstream to Point A can be satisfactorily described as exponential functions of the distance measured from the door, which shows good consistency among the three vaults of different room parameters. © 2018 The Authors. Journal of Applied Clinical Medical Physics published by Wiley Periodicals, Inc. on behalf of American Association of Physicists in Medicine.

Continental glaciation and its potential impact on a used-fuel disposal vault in the Canadian Shield

Energy Technology Data Exchange (ETDEWEB)

Ates, Y.; Bruneau, D.; Ridgway, W.R

1997-09-01

AECL has been assessing the concept of nuclear fuel waste disposal in a vault excavated at a depth ranging between 500 m and 1000 m in a plutonic rock mass of the Canadian Shield. Glaciation is a natural process that has occurred in the past, and is likely to occur in the future, thus causing changes in the loading conditions on the rock mass hosting the disposal vault. Because the rock mass is a natural barrier to the migration of radionuclides, it is important to evaluate its integrity under load changes caused by the glaciation process. Assuming that the magnitude and extent of the future glaciation will be similar to those of the past, we have reviewed published data pertaining to the last continental ice sheet that covered a large area of North America. Estimates have been madefor the magnitude of stresses due to ice sheet loading for a vault located at depths of 500 to 1000 m. These analyses have shown that the uniform loading of a continental ice sheet would reduce the deviatoric stresses in the Canadian Shield, creating more favourable conditions than those existing at the present time, namely, high horizontal stresses. The effects of surface erosion and increase in the in-situ shear stresses have also been examined. Based on the existing data and structural modelling studies, there would be no significant structural effect on a disposal vault located at 1000-m depth in a plutonic rock. At its maximum size, an ice sheet comparable to the Laurentide ice sheet could reactivate the faults and fracture zones along the perimeter areas. Our analyses have been based on fully drained conditions only. At a potential disposal site, it would be important also to consider the potential for excess pore pressure in the analyses. (author)

Continental glaciation and its potential impact on a used-fuel disposal vault in the Canadian Shield

International Nuclear Information System (INIS)

Ates, Y.; Bruneau, D.; Ridgway, W.R.

1997-09-01

AECL has been assessing the concept of nuclear fuel waste disposal in a vault excavated at a depth ranging between 500 m and 1000 m in a plutonic rock mass of the Canadian Shield. Glaciation is a natural process that has occurred in the past, and is likely to occur in the future, thus causing changes in the loading conditions on the rock mass hosting the disposal vault. Because the rock mass is a natural barrier to the migration of radionuclides, it is important to evaluate its integrity under load changes caused by the glaciation process. Assuming that the magnitude and extent of the future glaciation will be similar to those of the past, we have reviewed published data pertaining to the last continental ice sheet that covered a large area of North America. Estimates have been made for the magnitude of stresses due to ice sheet loading for a vault located at depths of 500 to 1000 m. These analyses have shown that the uniform loading of a continental ice sheet would reduce the deviatoric stresses in the Canadian Shield, creating more favourable conditions than those existing at the present time, namely, high horizontal stresses. The effects of surface erosion and increase in the in-situ shear stresses have also been examined. Based on the existing data and structural modelling studies, there would be no significant structural effect on a disposal vault located at 1000-m depth in a plutonic rock. At its maximum size, an ice sheet comparable to the Laurentide ice sheet could reactivate the faults and fracture zones along the perimeter areas. Our analyses have been based on fully drained conditions only. At a potential disposal site, it would be important also to consider the potential for excess pore pressure in the analyses. (author)

Grout for closure of waste-disposal vaults at the US DOE Hanford Site

International Nuclear Information System (INIS)

Wakeley, L.D.; Ernzen, J.J.; McDaniel, E.W.; Voogd, J.

1991-01-01

For permanent disposal of radioactive wastes from reprocessing, the US Department of Energy (DOE) has chosen to grout wastes in concrete vaults within a subsurface multiple-barrier system. The subject of this research is the non-radioactive, or ''cold cap'' grout, which fills the upper 120 cm of these vaults, and provides support for overlying barriers. Because of the heat evolved by the wasteform, this void-filling grout must perform at temperatures higher than those of usual large-volume grouting operations. It must have: low potential for thermal expansion and heat retention; a low modulus to withstand thermal and mechanical stresses without cracking; strength adequate to support overlying barrier-system components; and minimal potential for shrinkage. In addition, it must be pumpable, self-leveling, and non-segregating. Materials for formulation included a large percentage of Class F fly ash, and coarsely ground oil-well cement. Grout development included chemical and physical characterization, and physical and thermal modeling

Anterior vaulting of single piece in-the-bag intraocular lens after blunt trauma

Directory of Open Access Journals (Sweden)

Seung Hoon Oh

2012-01-01

Full Text Available A 64-year-old man complained of difficulty with far vision in the right eye after blunt trauma by his hand during sleep. He had undergone uneventful phacoemulsification and intraocular lens (IOL implantation of both eyes one week before the event. On slit lamp examination, the anterior chamber was shallow compared to postoperative one day or the left eye. After mydriasis, IOL position was perfect, within the capsulorrhexis margin, and no haptics were protruded. SE of his right eye was −4.00 diopter after trauma, although spherical equivalent (SE of both eyes was near emmetropia one day after cataract surgery. We decided on IOL reposition or exchange. Intraoperatively, we pushed the IOL posteriorly; the IOL was vaulted posteriorly, as expected. After surgery, his SE was +0.25 with the same diopter IOL. We report the case on the possibility of 4-point haptic IOL vaulting anteriorly by blunt trauma within the capsular bag.

Final characterization report for the 104-B-1 Tritium Vault and 104-B-2 Tritium Laboratory

International Nuclear Information System (INIS)

Encke, D.B.; Harris, R.A.

1996-11-01

This report is a compilation of the characterization data collected from the 104-B-1 Tritium Vault and the 104-B-2 Trillium Laboratory. The characterization activities were organized and implemented to evaluate the radiological status and identify any hazardous materials. The data contained in this report reflects the current conditions and status of the 104-B-1 Tritium Vault and 104-B-2 Tritium Laboratory. This information is intended to be utilized in support of future building decontamination and demolition, to allow for proper disposal of the demolition debris as required by the Washington Administrative Code, WAC 173-303, the Hanford Site Solid Waste Acceptance Criteria, WHC-EP-0063, and the Environmental Restoration Disposal Facility Waste Acceptance Criteria, BHI-00139. Based on the historical information and facility inspections, the only hazardous materials sampling and analysis activities necessary were to identify lead paint and asbestos containing materials (ACM) in the 104-B-1 Tritium Vault and the 104-B-2 Tritium Laboratory. Asbestos samples were obtained from the outer boundary of the roof areas to confirm the presence and type of asbestos containing fibers. Lead paint samples were obtained to confirm the presence and quantity of lead paint on the roof trim, doors and vents

NSun2-Mediated Cytosine-5 Methylation of Vault Noncoding RNA Determines Its Processing into Regulatory Small RNAs

Directory of Open Access Journals (Sweden)

Shobbir Hussain

2013-07-01

Full Text Available Autosomal-recessive loss of the NSUN2 gene has been identified as a causative link to intellectual disability disorders in humans. NSun2 is an RNA methyltransferase modifying cytosine-5 in transfer RNAs (tRNAs, yet the identification of cytosine methylation in other RNA species has been hampered by the lack of sensitive and reliable molecular techniques. Here, we describe miCLIP as an additional approach for identifying RNA methylation sites in transcriptomes. miCLIP is a customized version of the individual-nucleotide-resolution crosslinking and immunoprecipitation (iCLIP method. We confirm site-specific methylation in tRNAs and additional messenger and noncoding RNAs (ncRNAs. Among these, vault ncRNAs contained six NSun2-methylated cytosines, three of which were confirmed by RNA bisulfite sequencing. Using patient cells lacking the NSun2 protein, we further show that loss of cytosine-5 methylation in vault RNAs causes aberrant processing into Argonaute-associated small RNA fragments that can function as microRNAs. Thus, impaired processing of vault ncRNA may contribute to the etiology of NSun2-deficiency human disorders.

Safety Analysis Report for Packaging, Y-12 National Security Complex, Model ES-3100 Package with Bulk HEU Contents

Energy Technology Data Exchange (ETDEWEB)

Anderson, James [Y-12 National Security Complex, Oak Ridge, TN (United States); Goins, Monty [Y-12 National Security Complex, Oak Ridge, TN (United States); Paul, Pran [Y-12 National Security Complex, Oak Ridge, TN (United States); Wilkinson, Alan [Y-12 National Security Complex, Oak Ridge, TN (United States); Wilson, David [Y-12 National Security Complex, Oak Ridge, TN (United States)

2015-09-03

This safety analysis report for packaging (SARP) presents the results of the safety analysis prepared in support of the Consolidated Nuclear Security, LLC (CNS) request for licensing of the Model ES-3100 package with bulk highly enriched uranium (HEU) contents and issuance of a Type B(U) Fissile Material Certificate of Compliance. This SARP, published in the format specified in the Nuclear Regulatory Commission (NRC) Regulatory Guide 7.9 and using information provided in UCID-21218 and NRC Regulatory Guide 7.10, demonstrates that the Y-12 National Security Complex (Y-12) ES-3100 package with bulk HEU contents meets the established NRC regulations for packaging, preparation for shipment, and transportation of radioactive materials given in Title 10, Part 71, of the Code of Federal Regulations (CFR) [10 CFR 71] as well as U.S. Department of Transportation (DOT) regulations for packaging and shipment of hazardous materials given in Title 49 CFR. To protect the health and safety of the public, shipments of adioactive materials are made in packaging that is designed, fabricated, assembled, tested, procured, used, maintained, and repaired in accordance with the provisions cited above. Safety requirements addressed by the regulations that must be met when transporting radioactive materials are containment of radioactive materials, radiation shielding, and assurance of nuclear subcriticality.

Analysis of ICPP tank farm infiltration

International Nuclear Information System (INIS)

Richards, B.T.

1993-10-01

This report addresses water seeping into underground vaults which contain high-level liquid waste (HLLW) storage tanks at the Idaho Chemical Processing Plant (ICPP). Each of the vaults contains from one to three sumps. The original purpose of the sumps was to serve as a backup leak detection system for release of HLLW from the storage tanks. However, water seeps into most of the vaults, filling the sumps, and defeating their purpose as a leak detection system. Leak detection for the HLLW storage tanks is based on measuring the level of liquid inside the tank. The source of water leaking into the vaults was raised as a concern by the State of Idaho INEL Oversight Group because this source could also be leaching contaminants released to soil in the vicinity of the tank farm and transporting contaminants to the aquifer. This report evaluates information concerning patterns of seepage into vault sumps, the chemistry of water in sumps, and water balances for the tank farm to determine the sources of water seeping into the vaults

Safety of interim storage solutions of used nuclear fuel during extended term

Energy Technology Data Exchange (ETDEWEB)

Shelton, C.; Bader, S.; Issard, H.; Arslan, M. [AREVA, 7135 Minstrel Way, Suite 300 Columbia, MD 21045 (United States)

2013-07-01

In 2013, the total amount of stored used nuclear fuel (UNF) in the world will reach 225,000 T HM. The UNF inventory in wet storage will take up over 80% of the available total spent fuel pool (SFP) capacity. Interim storage solutions are needed. They give flexibility to the nuclear operators and ensure that nuclear reactors continue to operate. However, we need to keep in mind that they are also an easy way to differ final decision and implementation of a UNF management approach (recycling or final disposal). In term of public perception, they can have a negative impact overtime as it may appear that nuclear industry may have significant issues to resolve. In countries lacking an integrated UNF management approach, the UNF are being discharged from the SFPs to interim storage (mostly to dry storage) at the same rate as UNF is being discharged from reactors, as the SFPs at the reactor sites are becoming full. This is now the case in USA, Taiwan, Switzerland, Spain, South Africa and Germany. For interim storage, AREVA has developed different solutions in order to allow the continued operation of reactors while meeting the current requirements of Safety Authorities: -) Dry storage canisters on pads, -) Dual-purpose casks (dry storage and transportation), -) Vault dry storage, and -) Centralized pool storage.

Archaeological finds from the vaulted building at Krševica

Directory of Open Access Journals (Sweden)

Popović Petar

2008-01-01

Full Text Available The completely preserved building with barrel vault was discovered in 2008 in the course of investigation of the remains of an urban settlement dating from the 4th /3rd centuries BC at the site Kale in the village Krševica (southeast Serbia. We are presenting in this work the archaeological finds discovered in this structure. They included pottery, worked stone and many skeletons of horses and dogs that are assumed to be the ritual sacrifice.

Alternative dispositioning methods for HEU spent nuclear fuel at the Savannah River Site

International Nuclear Information System (INIS)

Krupa, J.F.; McKibben, J.M.; Parks, P.B.; DuPont, M.E.

1995-01-01

The United States has a strong policy on prevention of the international spread of nuclear weapons. This policy was announced in Presidential Directive PDD-13 and summarized in a White House press release September 27, 1993. Two cornerstones of this policy are: seek to eliminate where possible the accumulation of stockpiles of highly- enriched uranium or plutonium; propose hor-ellipsis prohibiting the production of highly-enriched uranium (HEU) or plutonium for nuclear explosives purposes or outside international safeguards. The Department of Energy is currently struggling to devise techniques that safely and efficiently dispose of spent nuclear fuel (SNF) while satisfying national non-proliferation policies. SRS plans and proposals for disposing of their SNF are safe and cost effective, and fully satisfy non-proliferation objectives

Long-term quality-of-life outcome after mesh sacrocolpopexy for vaginal vault prolapse.

LENUS (Irish Health Repository)

Thomas, Arun Z

2009-12-01

To evaluate the long-term outcome of mesh sacrocolpopexy (MSC, which aims to restore normal pelvic floor anatomy to alleviate prolapse related symptoms) and its effect on patient\\'s quality of life, as women with vaginal vault prolapse commonly have various pelvic floor symptoms that can affect urinary, rectal and sexual function.

Remote inspection of underground storage tanks

International Nuclear Information System (INIS)

Griebenow, B.L.; Martinson, L.M.

1992-01-01

Westinghouse Idaho Nuclear Company, Inc. (WINCO) operates the Idaho Chemical Processing Plant (ICPP) for the US Department of Energy. The ICPP's mission is to process government-owned spent nuclear fuel. The process involves dissolving the fuel, extracting off uranium, and calcining the waste to a solid form for storage, Prior to calcining, WINCO temporarily stores the liquid waste from this process in eleven 1,135,600-l(300,000-gal), 15,2-m (50-ft)-diam, high-level liquid waste tanks. Each of these stainless steel tanks is contained within an underground concrete vault. The only access to the interior of the tanks is through risers that extend from ground level to the dome of the tanks. WINCO is replacing these tanks because of their age and the fact that they do not meet all of the current design requirements. The tanks will be replaced in two phases. WINCO is now in the Title I design stage for four new tank and vault systems to replace five of the existing systems. The integrity of the six remaining tanks must be verified to continue their use until they can be replaced in the second phase. To perform any integrity analysis, the inner surface of the tanks must be inspected. The remote tank inspection (RTI) robotic system, designed by RedZone Robotics of Pittsburgh, Pennsylvania, was developed to access the interior of the tanks and position various end effectors required to perform tank wall inspections

The military health system's personal health record pilot with Microsoft HealthVault and Google Health.

Science.gov (United States)

Do, Nhan V; Barnhill, Rick; Heermann-Do, Kimberly A; Salzman, Keith L; Gimbel, Ronald W

2011-01-01

To design, build, implement, and evaluate a personal health record (PHR), tethered to the Military Health System, that leverages Microsoft® HealthVault and Google® Health infrastructure based on user preference. A pilot project was conducted in 2008-2009 at Madigan Army Medical Center in Tacoma, Washington. Our PHR was architected to a flexible platform that incorporated standards-based models of Continuity of Document and Continuity of Care Record to map Department of Defense-sourced health data, via a secure Veterans Administration data broker, to Microsoft® HealthVault and Google® Health based on user preference. The project design and implementation were guided by provider and patient advisory panels with formal user evaluation. The pilot project included 250 beneficiary users. Approximately 73.2% of users were Microsoft® HealthVault, and 81 (32.4%) selected Google® Health as their PHR of preference. Sample evaluation of users reflected 100% (n = 60) satisfied with convenience of record access and 91.7% (n = 55) satisfied with overall functionality of PHR. Key lessons learned related to data-transfer decisions (push vs pull), purposeful delays in reporting sensitive information, understanding and mapping PHR use and clinical workflow, and decisions on information patients may choose to share with their provider. Currently PHRs are being viewed as empowering tools for patient activation. Design and implementation issues (eg, technical, organizational, information security) are substantial and must be thoughtfully approached. Adopting standards into design can enhance the national goal of portability and interoperability.

Spent fuel and materials performance in wet and dry storage

Energy Technology Data Exchange (ETDEWEB)

Zuloaga, P [ENRESA (Spain)

2012-07-01

According to the 6th General Radioactive Waste Plan, spent fuel in Spain shall have to be gathered in a Centralised Temporary Storage (CTS) during some decades in order to have time for a decision concerning its final fate: direct disposal at a geological repository or partitioning and transmutation if technology opens this possibility when the decision will be taken, expected in 2050. The CTS technology has already been chosen as a vault type building based in spent fuel dry storage. To support the use of this technology, a number of programmes have been completed or are still in progress, mostly concerned about high burnup fuel issues and new cladding materials. These programmes are directly managed by ENRESA alone or in joint venture with other parties, at a national and international level. Apart from that, there are contacts with other countries organisms who share similar interests with Spanish ones. The objectives are: Review of spent fuel data relevant for future storage in Spain; Perform destructive and non-destructive examinations on irradiated and non-irradiated fuel rods relevant to Spanish spent fuel management.

Benchmarking of HEU mental annuli critical assemblies with internally reflected graphite cylinder

Directory of Open Access Journals (Sweden)

Xiaobo Liu

2017-01-01

Full Text Available Three experimental configurations of critical assemblies, performed in 1963 at the Oak Ridge Critical Experiment Facility, which are assembled using three different diameter HEU annuli (15-9 inches, 15-7 inches and 13-7 inches metal annuli with internally reflected graphite cylinder are evaluated and benchmarked. The experimental uncertainties which are 0.00057, 0.00058 and 0.00057 respectively, and biases to the benchmark models which are − 0.00286, − 0.00242 and − 0.00168 respectively, were determined, and the experimental benchmark keff results were obtained for both detailed and simplified models. The calculation results for both detailed and simplified models using MCNP6-1.0 and ENDF/B-VII.1 agree well to the benchmark experimental results within difference less than 0.2%. The benchmarking results were accepted for the inclusion of ICSBEP Handbook.

The Centralized Temporary Storage (ATC) Spanish. General and operational readiness in the management of high level radioactive waste

International Nuclear Information System (INIS)

Reyes, S.; Diaz, A.; Dilla, A.; Gonzalez, J. R.; Haro, R.

2014-01-01

The high-level waste will go to Centralized Temporary Storage (ATC) in Villar de Canas (Cuenca), which is the solution to your management until finished designing its final disposal in the Deep Geological Storage (AGP). Transport containers of high activity radioactive waste will be stored on arrival at the ATC. In time of management, waste will be taken remotely, will be encapsulated in secure conditions and stored in vaults designed to be cooled by natural convection of air confinement. The buildings in the facility will have a general provision optimize the movement of packages and people for the development of all these activities. (Author)

Radiological performance assessment for the E-Area Vaults Disposal Facility

International Nuclear Information System (INIS)

Cook, J.R.

2000-01-01

This report is the first revision to ''Radiological Performance Assessment for the E-Area Vaults Disposal Facility, Revision 0'', which was issued in April 1994 and received conditional DOE approval in September 1994. The title of this report has been changed to conform to the current name of the facility. The revision incorporates improved groundwater modeling methodology, which includes a large data base of site specific geotechnical data, and special Analyses on disposal of cement-based wasteforms and naval wastes, issued after publication of Revision 0

Shielding evaluation of a medical linear accelerator vault in preparation for installing a high-dose rate 252Cf remote after-loader

International Nuclear Information System (INIS)

Melhus, C. S.; Rivard, M. J.; KurKomelis, J.; Liddle, C. B.; Masse, F. X.

2005-01-01

In support of the effort to begin high-dose rate 252 Cf brachytherapy treatments at Tufts-New England Medical Center, the shielding capabilities of a clinical accelerator vault against the neutron and photon emissions from a 1.124 mg 252 Cf source were examined. Outside the clinical accelerator vault, the fast neutron dose equivalent rate was below the lower limit of detection of a CR-39 etched track detector and below 0.14 ± 0.02 μSv h -1 with a proportional counter, which is consistent, within the uncertainties, with natural background. The photon dose equivalent rate was also measured to be below background levels (0.1 μSv h -1 ) using an ionisation chamber and an optically stimulated luminescence dosemeter. A Monte Carlo simulation of neutron transport through the accelerator vault was performed to validate measured values and determine the thermal-energy to low-energy neutron component. Monte Carlo results showed that the dose equivalent rate from fast neutrons was reduced by a factor of 100,000 after attenuation through the vault wall, and the thermal-energy neutron dose equivalent rate would be an additional factor of 1000 below that of the fast neutrons. Based on these findings, the shielding installed in this facility is sufficient for the use of at least 5.0 mg of 252 Cf. (authors)

Bilateral sacrospinous vault fixation with tailored synthetic mesh arms: clinical outcomes at one year.

Science.gov (United States)

Geoffrion, Roxana; Hyakutake, Momoe T; Koenig, Nicole A; Lee, Terry; Cundiff, Geoffrey W

2015-02-01

Bilateral sacrospinous fixation with tailored mesh arms (bSSVF) uses polypropylene mesh to suspend the vault to the sacrospinous ligaments bilaterally with minimal tension, recreating nulliparous midline anatomy. It can be used with uterine conservation. Our primary objective was to determine objective cure rate at one year following bSSVF compared with a control group undergoing abdominal sacrocolpopexy (ASC). Secondary objectives were to compare symptoms, quality of life, sexual function, pain, and global satisfaction before and after surgery and between bSSVF and ASC groups at one year. This prospective cohort study enrolled patients with symptomatic prolapse who chose to undergo bSSVF or ASC. Baseline demographics were obtained. Prolapse quantification, validated symptom questionnaire scores, and McGill pain scores were obtained at baseline, six weeks, and one-year postoperatively. Global satisfaction was recorded. The primary outcome measure was the difference in cure rate (vault stage ≤ 1) between groups. Fifty patients were recruited: 30 underwent bSSVF and 17 ASC. Forty-three patients were available for one-year follow-up. Baseline data were similar. There was no difference in vault stage between bSSVF and ASC groups at one year. Five women who underwent bSSVF had cervical elongation, and four of these were classified as POP recurrence. Women who underwent bSSVF had more anterior recurrences but fewer postoperative complications, shorter hospital stay, and less use of narcotics than controls. Questionnaire scores were similar at one year. All respondents felt subjective improvement after either surgical procedure. Objective and subjective cure rates are comparable after bSSVF and ASC. Hysteropexy may cause cervical elongation that merits further research.

Sustainable Solutions for Nuclear used Fuels Interim Storage

International Nuclear Information System (INIS)

Arslan, Marc; Favet, Dominique; Issard, Herve; Le Jemtel, Amaury; Drevon, Caroline

2014-01-01

AREVA has a unique experience in providing sustainable solutions for used fuel management, fitted with the needs of different customers in the world and with regulation in different countries. These solutions entail both recycling and interim storage technologies. In a first part, we will describe the various types of solutions for Interim Storage of UNF that have been implemented around the world for interim storage at reactor or centralized Pad solution in canisters dry storage, vault type storages for dry storage, dry storage of transportation casks (dual purpose) pools for wet storage, The experience for all these different families of interim storages in which AREVA is involved is extensive and will be discussed with respect to the new challenges: increase of the duration of the interim storage (long term interim storage) increase of burn up of the fuels In a second part of the presentation, special recycling features will be presented. In that case, interim storage of the used fuels is ensured in pools. This provides in the long term good conditions for the behaviour of the fuel and its retrievability. With recycling, the final waste (Universal Canister of vitrified fission products and compacted hulls and end pieces): is stable and licensed in many countries for the final disposal (France, UK, Belgium, NL, Switzerland, Germany, Japan, upcoming: Spain, Australia, Italy). Presents neither safety criticality risks nor proliferation risks (AREVA conditioned HLW and LL-ILW are free of IAEA safeguard constraints thanks to AREVA process high recovery and purification yields). It can therefore be safely stored in interim storage for more than 100 years before final disposal. Some economic considerations will also be discussed. In particular, in the case of long term interim storage of used fuels, there are growing uncertainties regarding the future needs of repackaging and transportation, which can result in future cost overruns. Meanwhile, in the recycling policy

Bridging the Gap: Capturing the Lyα Counterpart of a Type-II Spicule and its Heating Evolution with VAULT2.0 and IRIS Campaign Observations

Science.gov (United States)

Chintzoglou, G.; De Pontieu, B.; Martinez-Sykora, J.; Mendes Domingos Pereira, T.; Vourlidas, A.; Tun Beltran, S.

2017-12-01

We present the analysis of data from the observing campaign in support to the VAULT2.0 sounding rocket launch on September 30, 2014. VAULT2.0 is a Lyα (1216 Å) spectroheliograph capable of providing fast cadence spectroheliograms of high-spectral purity. High resolution Lyα observations are highly complementary with the IRIS observations of the upper chromosphere and the low transition region but have previously been unavailable. The VAULT2.0 data provide critical, new upper-chromospheric constraints for numerical models. The observing campaign was closely coordinated with the IRIS mission. Taking advantage of this simultaneous multi-wavelength coverage of target AR 12172 and by using state-of-the-art radiative-MHD simulations of spicules, we are able to perform a detailed investigation of a type-II spicule associated with a fast apparent network jet recorded in the campaign observations during the VAULT2.0 flight. Our unique analysis suggests that spicular material exists suspended in lower temperatures until it rapidly gets heated and becomes visible in transition-region temperatures as an apparent network jet.

Nonproliferation and arms control assessment of weapons-usable fissile material storage and excess plutonium disposition alternatives

International Nuclear Information System (INIS)

1997-01-01

This report has been prepared by the Department of Energy's Office of Arms Control and Nonproliferation (DOE-NN) with support from the Office of Fissile Materials Disposition (DOE-MD). Its purpose is to analyze the nonproliferation and arms reduction implications of the alternatives for storage of plutonium and HEU, and disposition of excess plutonium, to aid policymakers and the public in making final decisions. While this assessment describes the benefits and risks associated with each option, it does not attempt to rank order the options or choose which ones are best. It does, however, identify steps which could maximize the benefits and mitigate any vulnerabilities of the various alternatives under consideration

Conversion and Blending Facility highly enriched uranium to low enriched uranium as uranyl nitrate hexahydrate. Revision 1

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-07-05

This Conversion and Blending Facility (CBF) will have two missions: (1) convert HEU materials to pure HEU uranyl nitrate (UNH) and (2) blend pure HEU UNH with depleted and natural UNH to produce HEU UNH crystals. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. To the extent practical, the chemical and isotopic concentrations of blended LEU product will be held within the specifications required for LWR fuel. Such blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry. Otherwise, blended LEU Will be produced as a waste suitable for storage or disposal.

Russian RERTR program as a part of Joint US DOE-RF MINATOM collaboration on elimination of the threat connected to the use of HEU in research reactors

International Nuclear Information System (INIS)

Arkhangelsky, N.

2002-01-01

achieved when the fresh fuel from research reactor in Yugoslavia was shipped to Russia for the safe storage and subsequent blending. The future plans include: continuation of the development of pin type fuel on the base of U-Mo fuel; conversion of reactor in Uzbekistan and shipment of SNF from the site of this reactor to Russia; conversion of other Russian design reactors to LEU; Continuation of investigations of the possibility of conversion of Russian domestic reactors to LEU; consecutive implementation of the take-back program of Russian origin fuel; study of other aspects of elimination of the use of HEU in research reactors. Russian RERTR program has reached the definite progress but its main goal is not yet achieved. Recent successes in the development and qualification of WWR-M2 fuel elements allows to begin delivery of fuel elements with LEU

YB-1 facilitates basal and 5-fluorouracil-inducible expression of the human major vault protein (MVP) gene.

Science.gov (United States)

Stein, Ulrike; Bergmann, Stephan; Scheffer, George L; Scheper, Rik J; Royer, Hans-Dieter; Schlag, Peter M; Walther, Wolfgang

2005-05-19

Vaults have been suggested to play a direct role in multidrug resistance (MDR) to anticancer drugs. The human major vault protein (MVP) also known as lung resistance-related protein (LRP) represents the predominant component of vaults that may be involved in the defense against xenobiotics. Here, we demonstrate that besides MDR-related cytostatics, also the non-MDR-related drug 5-fluorouracil (5-FU) was able to induce MVP mRNA and protein expression. Treatment with 5-FU amplified the binding activity and interaction of the transcription factor Y-box binding protein-1 (YB-1) with the Y-box of the human MVP gene promoter in a time-dependent manner. 5-FU also induced reporter expressions driven by a panel of newly generated MVP promoter deletion mutants. Interestingly, stably YB-1 overexpressing cell clones showed enhanced binding of YB-1 to the Y-box motif, associated with enhanced basal as well as 5-FU-inducible MVP promoter-driven reporter expressions. Moreover, transduction of YB-1 cDNA led to increased expression of endogenous MVP protein. Under physiological conditions, we observed a strong coexpression of MVP and YB-1 in human colon carcinoma specimen. In summary, our data demonstrate a direct involvement of YB-1 in controlling basal and 5-FU-induced MVP promoter activity. Therefore, YB-1 is directly linked to MVP-mediated drug resistance.

Intensive care unit course of infants and children after cranial vault reconstruction for craniosynostosis

Directory of Open Access Journals (Sweden)

Jansen David A

2011-09-01

Full Text Available Abstract Background Craniosynostosis (CSS results from the premature closure of one or more cranial sutures, leading to deformed calvaria at birth. It is a common finding in children with an incidence of one in 2000 births. Surgery is required in order to release the synostotic constraint and promote normal calvaria growth. Cranial vault remodeling is the surgical approach to CSS repair at our institution and it involves excision of the frontal, parietal, and occipital bones. The purpose of this article is to describe the post-operative course of infants and children admitted to our PICU after undergoing cranial vault remodeling for primary CSS. Findings Complete data was available for analyses in only 82 patients, 44 males (M and 38 females (F; M: F ratio was 1:1.2. Patients (pts age in months (mo ranged from 2 mo to 132 mo, mean 18.2 ±-24.9 mo and weights (wt ranged from 4.7 kg to 31.4 kg, mean 10.24 ± 5.5 Kg.. Duration of surgery (DOS ranged from 70 minutes to 573 minutes mean 331.6 ± 89.0 minutes. No significant correlation exist between duration of surgery, suture category, patient's age or use of blood products (P > 0.05. IOP blood loss was higher in older pts (P 3 days in 32%. Pts with fever had prolonged LOS (P Conclusions Post-op morbidities from increased use of blood products can be minimized if cranial vault remodeling is done at a younger age in patients with primary CSS. PICU length of stay is determined in part by post-op pyrexia and it can be reduced if extensive evaluations of post-op fever are avoided.

Preliminary safety evaluation for the plutonium stabilization and packaging system

International Nuclear Information System (INIS)

Shapley, J.E.

1997-01-01

This Preliminary Safety Evaluation (PSE) describes and analyzes the installation and operation of the Plutonium Stabilization and Packaging System (SPS) at the Plutonium Finishing Plant (PFP). The SPS is a combination of components required to expedite the safe and timely storage of Plutonium (Pu) oxide. The SPS program will receive site Pu packages, process the Pu for storage, package the Pu into metallic containers, and safely store the containers in a specially modified storage vault. The location of the SPS will be in the 2736- ZB building and the storage vaults will be in the 2736-Z building of the PFP, as shown in Figure 1-1. The SPS will produce storage canisters that are larger than those currently used for Pu storage at the PFP. Therefore, the existing storage areas within the PFP secure vaults will require modification. Other modifications will be performed on the 2736-ZB building complex to facilitate the installation and operation of the SPS

Structural Integrity Program for the 300,000-Gallon Radioactive Liquid Waste Storage Tanks at the Idaho Nuclear Technology and Engineering Center

International Nuclear Information System (INIS)

Bryant, J.W.; Nenni, J.A.; Yoder, T.S.

2003-01-01

This report provides a record of the Structural Integrity Program for the 300,000-gal liquid waste storage tanks and associated equipment at the Idaho Nuclear Technology and Engineering Center, as required by U.S. Department of Energy M 435.1-1, ''Radioactive Waste Management Manual.'' This equipment is known collectively as the Tank Farm Facility. The conclusion of this report is that the Tank Farm Facility tanks, vaults, and transfer systems that remain in service for storage are structurally adequate, and are expected to remain structurally adequate over the remainder of their planned service life through 2012. Recommendations are provided for continued monitoring of the Tank Farm Facility

Safety Problems of Disposal of Disused Sealed Sources in the Baldone Near Surface Repository

International Nuclear Information System (INIS)

Dreimanis, A.

2003-01-01

Current Latvian regulations encourage re-export of DSS, however, up to now in the repository has been disposed a lot of DSS. Baldone repository has received RW (mainly DSS) from ∼ 300 Latvia objects and from Kaliningrad region (from 1964 to 1973). DSS cause the major part of activity of the repository - from the total activity in facility ∼ 400 TBq more than 300 TBq is due to DSS. A long term Safety Assessment (SA) of the Baldone Radon-type near surface disposal facility has been performed by the consortium CASSIOPEE and recommendations are given. The waste disposal system consists of: vaults 1-6 (closed, as permanent disposal site), vault 7 (as a long-term retrievable storage site); 2 main components of the engineered barriers: the capping system and the vaults. Vaults 1 and 3-6 (of re-fabricated concrete elements, closed with reinforced concrete slabs, covered with hydro isolating layer completed by a sand/ soil layer). The vault No.7 (ten 130 m 3 underground concrete walled storage cells, protected from the weather by the building. These 10 underground storage cells are adjacent (2x5) with additional concrete walls in some of them. The cells are covered by 40 cm concrete slabs placed side to side. The assessment context and development of scenarios are presented in the paper. The results from the SA are presented. For the waste pathway scenarios - the resulting doses - ∼ 4 mSv/y at 30 y in the current status without cover, justify the implantation of a cover for the closure period of the repository. For air pathway scenarios - the basic dose target 1 mSv/y (for public) is not satisfied for all scenarios; the resulting effective dose for the on-site 'Western' residential scenario for the vaults with highest content of DSS: vault 3 - 303 mSv/y; vault 7 - 204 mSv/y. The main recommendations from the SA are: General Advice - to dispose sources with half-lives < 5,3 y; for the spent sources - to build a new long-term storage; 2. To move DSS from Vault 7 to the

What sampling device is the most appropriate for vaginal vault cytology in gynaecological cancer follow up?

International Nuclear Information System (INIS)

Lino, Del Pup; Vincenzo, Canzonieri; Diego, Serraino; Elio, Campagnutta

2012-01-01

In women with cancer-related hysterectomy, the vaginal vault cytology has a low efficacy - when performed by conventional methods – for the early detection of vaginal recurrence. The amount of exfoliated cells collected is generally low because of atrophy, and the vaginal vault corners can be so narrow that the commonly used Ayres spatula cannot often penetrate deeply into them. This prospective study aimed at identifying the advantages obtained in specimens collection using the cytobrush, as compared to the Ayres’s spatula. 141 gynaecologic cancer patients were studied to compare samplings collected with Ayre’s spatula or with cytobrush. In a pilot setting of 15 patients, vaginal cytology samples obtained by both Ayre’s spatula and cytobrush were placed at the opposite sites of a single slide for quali-quantitative evaluation. Thereafter, the remaining 126 consecutive women were assigned to either group A (spatula) or B (cytobrush) according to the order of entry. The same gynaecologist performed all the procedures. In all 15 pilot cases, the cytobrush seemed to collect a higher quantity of material. The comparative analysis of the two complete groups indicated that the cytobrush technique was more effective than the spatula one. The odds ratio (OR) for an optimal cytology using the cytobrush was 2.8 (95% confidence interval -C.I. 1.3–6.2; chi-square test, p=0.008). Vaginal vault cytology with cytobrush turned out to better perform than the traditional Ayre’s spatula to obtain an adequate sampling in gynecological cancer patients

Comment on the contribution of S.C. Mo, N.A. Hanan and J.E. Matos: 'Comparison of the FRM-II HEU design with an alternative LEU design'

International Nuclear Information System (INIS)

Boening, K.

2004-01-01

The results of the reference paper, which came to our attention for the first time during this RERTR Meeting, are more or less consistent with neutronic data we have obtained earlier within the FRM-II project (i.e. with own calculations and extrapolations). However, a realistic comparison of the HEU design of the FR.M-II (HEU = highly enriched uranium, 93 % U-235) with an alternative LEU design (LEU = low enriched uranium, 20 % U-235) is only possible on the basis of identical assumptions on the input parameters and has to consider more than neutronic data only. Serious scientists and experts should not confuse the politicians with academic studies touching some aspects of the full story only. The comparison has shown that the performance and reliability of the FRM-II design, which uses HEU fuel, is so advantageous that it can not - not even approximately - be met by an alternative design using LEU fuel. A change of the FRM-II design from HEU to LEU fuel with the results as shown above - i.e. less performance, higher costs, more nuclear waste and higher risk potential, and all of this with a delay of at least 5 years this could never be justified. If a future development of more advanced fuels should allow us to achieve our scientific goals at the conditions as identified above also with uranium of reduced enrichment - there would be no objection to a corresponding later conversion. Activities to realize a new neutron source in Germany go back to the late 70's with the project of a new middle flux beam reactor (MSR), which was abandoned shortly later in favour of an ambitious new spallation neutron source (SNQ). After this project also having been terminated around 1985 because of too high costs and technological risks, the hopes of the German community of neutron scientists focussed on the FRM-II. If non-technical pressure would damage this project this would equally provide irreversible damage to the large and still prospering field of neutron research in Germany

Mattia Preti’s vault in St John’s, Valletta: New light on a complex iconography

DEFF Research Database (Denmark)

Svenningsen, Jesper

2015-01-01

The article aims at casting new light on Mattia Preti’s well-known decoration of the vault of the conventual church of the Knight Hospitallers, now the Co-cathedral of Valletta. To a modern viewer the iconography of the decoration is all but simple. The iconographic scheme, I argue, seeks to clai...

Nonproliferation and arms control assessment of weapons-usable fissile material storage and excess plutonium disposition alternatives

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-01-01

This report has been prepared by the Department of Energy`s Office of Arms Control and Nonproliferation (DOE-NN) with support from the Office of Fissile Materials Disposition (DOE-MD). Its purpose is to analyze the nonproliferation and arms reduction implications of the alternatives for storage of plutonium and HEU, and disposition of excess plutonium, to aid policymakers and the public in making final decisions. While this assessment describes the benefits and risks associated with each option, it does not attempt to rank order the options or choose which ones are best. It does, however, identify steps which could maximize the benefits and mitigate any vulnerabilities of the various alternatives under consideration.

Pilonidal cyst on the vault: case report

Directory of Open Access Journals (Sweden)

BORGES GUILHERME

1999-01-01

Full Text Available Pilonidal cysts and sinuses are described as dermoid cysts which contain follicles of hairs and sebaceous glands. They clinically present as a classic case of inflammation which comes with pain, local infection and redness. The origin of pilonidal disease remains controverse. There are many hypothesis as lack of hygiene on the affected area and a penetration and growth of a hair in the subcutaneus tissue caused by constant friction or direct trauma on the damaged area. The option for clinical treatment is very frequent. However, taking into consideration the incidence and the possibility of recidive, surgical treatment is presently recommended. Complications include cellulitis and abscess formation. Pilonidal cysts are mostly found on the sacral region. In the literature is found description of pilonidal cysts on the penis, interdigital region on the hands as well as on the cervical region. We present a case of pilonidal cyst located on the vault biparietal region, without malignant degeneration.

Current state of WWER SNF storage in Russia and the perspectives

International Nuclear Information System (INIS)

Anisimov, O.; Kozlov, Y.; Razmashkin, N.; Safutin, V.; Tikhonov, N.

2006-01-01

In the Russian Federation WWER-440 Spent Nuclear Fuel (SNF) is reprocessed at RT-1 plant near Cheliabinsk. WWER-1000 SNF is supposed to be reprocessed at RT-2 plant, which will be built about 2020. The information on the capacity and fill up level of the at-reactor pools at NPP with WWER reactors considering its modification up to May 2005 is given. The regulatory requirements to all SNF 'wet' storage facilities; the principle design and engineering solutions as well as the complex of measures for radiation safety and the environmental protection of spent fuel storage are presented. WWER-440 SNF management, WWER-1000 SNF management and dry storage of WWER-1000 SNF are discussed. In the conclusion it is noted than neither Russia, nor any other country have the experience of construction of vault-type 'dry' storage facilities of such a capacity to store WWER-1000 SNF (9000 tU). The experience and design solutions approved earlier in creation of other dangerous facilities were used. The calculations were based on conservative assumptions allowing with a large assurance to guarantee the nuclear and radiation safety and the environmental protection. At present, a program is developed for scientific-technical support of the dry storage facility design and operation, aimed at the studies whose results will allow to optimize the taken technical decisions, simplify SNF management technology and, possibly, to reduce the cost of the storage facility itself

Two parallel, pragmatic, UK multicentre, randomised controlled trials comparing surgical options for upper compartment (vault or uterine) pelvic organ prolapse (the VUE Study): study protocol for a randomised controlled trial.

Science.gov (United States)

Glazener, Cathryn; Constable, Lynda; Hemming, Christine; Breeman, Suzanne; Elders, Andrew; Cooper, Kevin; Freeman, Robert; Smith, Anthony R B; Hagen, Suzanne; McDonald, Alison; McPherson, Gladys; Montgomery, Isobel; Kilonzo, Mary; Boyers, Dwayne; Goulao, Beatriz; Norrie, John

2016-09-08

One in three women who have a prolapse operation will go on to have another operation, though not necessarily in the same compartment. Surgery can result in greater impairment of quality of life than the original prolapse itself (such as the development of new-onset urinary incontinence, or prolapse at a different site). Anterior and posterior prolapse surgery is most common (90 % of operations), but around 43 % of women also have a uterine (34 %) or vault (9 %) procedure at the same time. There is not enough evidence from randomised controlled trials (RCTs) to guide management of vault or uterine prolapse. The Vault or Uterine prolapse surgery Evaluation (VUE) study aims to assess the surgical management of upper compartment pelvic organ prolapse (POP) in terms of clinical effectiveness, cost-effectiveness and adverse events. VUE is two parallel, pragmatic, UK multicentre, RCTs (Uterine Trial and Vault Trial). Eligible for inclusion are women with vault or uterine prolapse: requiring a surgical procedure, suitable for randomisation and willing to be randomised. Randomisation will be computer-allocated separately for each trial, minimised on: requiring concomitant anterior and/or posterior POP surgery or not, concomitant incontinence surgery or not, age (under 60 years or 60 years and older) and surgeon. Participants will be randomly assigned, with equal probability to intervention or control arms in either the Uterine Trial or the Vault Trial. Uterine Trial participants will receive either a vaginal hysterectomy or a uterine preservation procedure. Vault Trial participants will receive either a vaginal sacrospinous fixation or an abdominal sacrocolpopexy. Participants will be followed up by postal questionnaires (6 months post surgery and 12 months post randomisation) and also reviewed in clinic 12 months post surgery. The primary outcome is the participant-reported Pelvic Organ Prolapse Symptom Score (POP-SS) at 12 months post randomisation

Safety Aspects of Long Term Spent Fuel Dry Storage

International Nuclear Information System (INIS)

Botsch, Wolfgang; Smalian, S.; Hinterding, P.; Drotleff, H.; Voelzke, H.; Wolff, D.; Kasparek, E.

2014-01-01

As a consequence of the lack of a final repository for spent nuclear fuel (SF) and high level waste (HLW), long term interim storage of SF and HLW will be necessary. As with the storage of all radioactive materials, the long term storage of SF and HLW must conform to safety requirements. Safety aspects such as safe enclosure of radioactive materials, safe removal of decay heat, sub-criticality and avoidance of unnecessary radiation exposure must be achieved throughout the complete storage period. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. After the events of Fukushima, the advantages of passively and inherently safe dry storage systems have become more obvious. In Germany, dry storage of SF in casks fulfils both transport and storage requirements. Mostly, storage facilities are designed as concrete buildings above the ground; one storage facility has also been built as a rock tunnel. In all these facilities the safe enclosure of radioactive materials in dry storage casks is achieved by a double-lid sealing system with surveillance of the sealing system. The safe removal of decay heat is ensured by the design of the storage containers and the storage facility, which also secures to reduce the radiation exposure to acceptable levels. TUV and BAM, who work as independent experts for the competent authorities, inform about spent fuel management and issues concerning dry storage of spent nuclear fuel, based on their long experience in these fields. All relevant safety issues such as safe enclosure, shielding, removal of decay heat and sub-criticality are checked and validated with state-of-the-art methods and computer codes before the license approval. In our presentation we discuss which of these aspects need to be examined closer for a long term interim storage. It is shown

Slight conductive hearing loss in children with narrowed maxilla and deep palatal vault.

Science.gov (United States)

Kiki, A; Kiliç, N; Oktay, H

2015-01-01

PROBLEM/OBJECTIVES: Maxillary constriction and high palatal arch are associated with increased risk of chronic eustachian tube dysfunction and conductive hearing loss (CHL) due to chronic effusion. However, this relationship has not been clearly demonstrated. This study assessed CHL in school children with a narrowed maxilla and deep palatal vault. Thirty-two children with maxillary constriction were randomly selected for the study group and 28 children with normal transverse maxillary development were selected for the control group. Pure-tone audiograms were obtained for all children, and hearing levels and air-bone gaps were measured. Air-bone gap measurements in the control group ranged from 5.50 to 14.50 decibels (dB), and in the study group they were between 5.00 and 24.00 dB. In the study group, 14 (43.8%) children had slight CHL, and the remaining 18 (56.2%) children had normal hearing levels. In the control group, all of the children had normal hearing levels. Hearing levels and air-bone gaps were greater in the study group than the control group. This study showed that children with a narrowed maxilla and deep palatal vault may have slight CHL. Therefore, the onset of CHL should be followed with hearing screening programs.

Siim Nestor soovitab : Stereo ÖÖ. Mutant Disco. Beats from the Vault. Turbodisko / Siim Nestor

Index Scriptorium Estoniae

Nestor, Siim, 1974-

2006-01-01

Oma debüütalbumit "Migration" tutvustav ameerika diskor Alex Attias üritustel "Stereo ÖÖ" 19. mail Tallinnas Von Krahlis. Üritustest "Mutant Disco" 19. mail Tallinnas klubis Privé ja Tartus 20. mail klubis Illusion, "Beats from the Vault" 19. mail Tallinnas klubis Võit ja "Turbodisko" 20. mail Tallinnas klubis KuKu

Department of Energy, highly enriched uranium ES ampersand H vulnerability assessment, Idaho National Engineering Laboratory site assessment team report

International Nuclear Information System (INIS)

1996-01-01

In accordance with the February 22, 1996 directive issued by Secretary of Energy O'Leary on the Vulnerability Assessment of Highly Enriched Uranium (HEU) Storage, the Idaho National Engineering Laboratory conducted an assessment of the site's HEU holdings and any associated vulnerabilities. The assessment was conducted between April 25 and May 24, 1996. The scope of this assessment, as defined in the Assessment Plan, included all HEU, and any spent fuel not evaluated in the Spent Fuel Vulnerability Assessment. Addressed in this assessment were all of the holdings at the Idaho National Engineering Laboratory (INEL) except any located at Argonne National Laboratory-West (ANL-W) and the Naval Reactors Facility. Excluded from the assessment were those HEU holdings previously assessed in the Idaho National Engineering Laboratory Spent Nuclear Fuel Inventory and Vulnerability Site Assessment Report and any HEU holdings evaluated in the Plutonium Vulnerability Assessment Report

IAEA Mission Sees High Commitment to Safety at Ghana's Research Reactor After HEU to LEU Fuel Conversion

International Nuclear Information System (INIS)

2018-01-01

An International Atomic Energy Agency (IAEA) team of experts said the operator of Ghana’s research reactor has demonstrated a high commitment to safety following the conversion of the reactor core to use low enriched uranium (LEU) as fuel instead of high enriched uranium (HEU). The team also made recommendations for further safety enhancements. The Integrated Safety Assessment for Research Reactors (INSARR) team concluded a five-day mission today to assess the safety of the GHARR-1 research reactor, originally commissioned in 1994. The 30 kW reactor, operated by the Ghana Atomic Energy Commission (GAEC) at the National Nuclear Research Institute in the capital Accra, is used primarily for trace element analysis for industrial or agricultural purposes, research, education and training. In 2017, the reactor core was converted in a joint effort by Ghana, the United States and China, with assistance from the IAEA. The IAEA supported the operation to eliminate proliferation risks associated with HEU, while maintaining important scientific research. The team made recommendations for improvements to the GAEC, including: • Completing the revision of reactor safety and operating documents to reflect the results of the commissioning of the reactor after the core fuel conversion. • Enhancing the training and qualification programme for operating personnel. • Improving the capability for monitoring operational safety parameters under all conditions. • Strengthening radiation protection by establishing an effective radiation monitoring of workplace. The GAEC said it will request a follow-up INSARR mission by 2020.

Major vault protein/lung resistance-related protein (MVP/LRP) expression in nervous system tumors.

Science.gov (United States)

Sasaki, Tsutomu; Hankins, Gerald R; Helm, Gregory A

2002-01-01

Lung resistance-related protein (LRP) was identified as the major vault protein (MVP), the main component of multimeric vault particles. It functions as a transport-associated protein that can be associated with multidrug resistance. In previous studies, expression of MVP/LRP has been documented in tumors of various types. In general, good correlations have been reported for expression of MVP/LRP and decreased sensitivity to chemotherapy and poor prognosis. MVP/LRP expression has been documented in glioblastomas, but its expression in nervous system tumors in general has not been well characterized. Immunohistochemistry using anti-human MVP/LRP antibody (LRP-56) was performed on formalin-fixed, paraffin-embedded archival tissue from 69 primary central nervous system tumors. Expression of MVP/LRP was observed in 81.2% (56/69) of primary nervous system tumors, including astrocytomas (11/13), oligodendrogliomas (1/2), oligoastrocytomas (5/5), ependymoma (1/1), meningiomas (35/45), schwannomas (2/2), and neurofibroma (1/1). Various degrees and distributions of immunoreactivity to MVP/ LRP were observed. Neither the presence nor the degree of immunoreactivity to MVP/LRP showed any correlation with either tumor grade or the presence of brain invasion.

Frescoed Vaults: Accuracy Controlled Simplified Methodology for Planar Development of Three-Dimensional Textured Models

Directory of Open Access Journals (Sweden)

Marco Giorgio Bevilacqua

2016-03-01

Full Text Available In the field of documentation and preservation of cultural heritage, there is keen interest in 3D metric viewing and rendering of architecture for both formal appearance and color. On the other hand, operative steps of restoration interventions still require full-scale, 2D metric surface representations. The transition from 3D to 2D representation, with the related geometric transformations, has not yet been fully formalized for planar development of frescoed vaults. Methodologies proposed so far on this subject provide transitioning from point cloud models to ideal mathematical surfaces and projecting textures using software tools. The methodology used for geometry and texture development in the present work does not require any dedicated software. The different processing steps can be individually checked for any error introduced, which can be then quantified. A direct accuracy check of the planar development of the frescoed surface has been carried out by qualified restorers, yielding a result of 3 mm. The proposed methodology, although requiring further studies to improve automation of the different processing steps, allowed extracting 2D drafts fully usable by operators restoring the vault frescoes.

Influence of the xanthan gum addition on the technological and sensory quality of baking products during the freezing storage

Directory of Open Access Journals (Sweden)

Tatiana Bojňanská

2016-07-01

Full Text Available The influence of the 0.16% xanthan gum addition in the recipe of the bread production and its influence on the baking and sensory quality of products was monitored during the process of our research. Prepared dough was inserted in the freezing box directly (-18°C and it was stored for one, two, three, four, five and six months. When the baking process was finished, the products with xanthan gum and the products without it were compared and evaluated by both objective and subjective methods. It was found that freezing, cooling and storage of the products without xanthan gum addition influenced the volume, vaulting and general appearance of the products in a negative way and loaves of bread were evaluated as unacceptable after four months of freezing. The quality of experimental loaves of bread with xanthan gum was, even after six months of freezing storage, comparable with freshly baked products. Despite the freezing, the volume of the products had an increased value. After first month of freezing the volume increased by 28.6% and after two months of freezing it increased by 23.8% both compared to the control. The vaulting in products processed by freezing was in the required optimal level during the whole period of freezing. Sensory evaluation results of loaves of bread with xanthan gum were the best after three, four and five months of storage in a freezer, when 98 points were achieved. During the monitored period of freezing, the addition of 0.16% of xanthan gum markedly contributed to the preservation of sensory and baking quality of the frozen wheat dough.

Benchmark calculations on nuclear characteristics of JRR-4 HEU core by SRAC code system

International Nuclear Information System (INIS)

Arigane, Kenji

1987-04-01

The reduced enrichment program for the JRR-4 has been progressing based on JAERI's RERTR (Reduced Enrichment Research and Test Reactor) program. The SRAC (JAERI Thermal Reactor Standard Code System for Reactor Design and Analysis) is used for the neutronic design of the JRR-4 LEU Core. This report describes the benchmark calculations on the neutronic characteristics of the JRR-4 HEU Core in order to validate the calculation method. The benchmark calculations were performed on the various kind of neutronic characteristics such as excess reactivity, criticality, control rod worth, thermal neutron flux distribution, void coefficient, temperature coefficient, mass coefficient, kinetic parameters and poisoning effect by Xe-135 build up. As the result, it was confirmed that these calculated values are in satisfactory agreement with the measured values. Therefore, the calculational method by the SRAC was validated. (author)

Special Analysis for Disposal of High-Concentration I-129 Waste in the Intermediate-Level Vaults at the E-Area Low-Level Waste Facility

Energy Technology Data Exchange (ETDEWEB)

Collard, L.B.

2000-09-26

This revision was prepared to address comments from DOE-SR that arose following publication of revision 0. This Special Analysis (SA) addresses disposal of wastes with high concentrations of I-129 in the Intermediate-Level (IL) Vaults at the operating, low-level radioactive waste disposal facility (the E-Area Low-Level Waste Facility or LLWF) on the Savannah River Site (SRS). This SA provides limits for disposal in the IL Vaults of high-concentration I-129 wastes, including activated carbon beds from the Effluent Treatment Facility (ETF), based on their measured, waste-specific Kds.

Special Analysis for Disposal of High-Concentration I-129 Waste in the Intermediate-Level Vaults at the E-Area Low-Level Waste Facility

International Nuclear Information System (INIS)

Collard, L.B.

2000-01-01

This revision was prepared to address comments from DOE-SR that arose following publication of revision 0. This Special Analysis (SA) addresses disposal of wastes with high concentrations of I-129 in the Intermediate-Level (IL) Vaults at the operating, low-level radioactive waste disposal facility (the E-Area Low-Level Waste Facility or LLWF) on the Savannah River Site (SRS). This SA provides limits for disposal in the IL Vaults of high-concentration I-129 wastes, including activated carbon beds from the Effluent Treatment Facility (ETF), based on their measured, waste-specific Kds

Vault-poly-ADP-ribose polymerase in the Octopus vulgaris brain: a regulatory factor of actin polymerization dynamic.

Science.gov (United States)

De Maio, Anna; Natale, Emiliana; Rotondo, Sergio; Di Cosmo, Anna; Faraone-Mennella, Maria Rosaria

2013-09-01

Our previous behavioural, biochemical and immunohistochemical analyses conducted in selected regions (supra/sub oesophageal masses) of the Octopus vulgaris brain detected a cytoplasmic poly-ADP-ribose polymerase (more than 90% of total enzyme activity). The protein was identified as the vault-free form of vault-poly-ADP-ribose polymerase. The present research extends and integrates the biochemical characterization of poly-ADP-ribosylation system, namely, reaction product, i.e., poly-ADP-ribose, and acceptor proteins, in the O. vulgaris brain. Immunochemical analyses evidenced that the sole poly-ADP-ribose acceptor was the octopus cytoskeleton 50-kDa actin. It was present in both free, endogenously poly-ADP-ribosylated form (70kDa) and in complex with V-poly-ADP-ribose polymerase and poly-ADP-ribose (260kDa). The components of this complex, alkali and high salt sensitive, were purified and characterized. The kind and the length of poly-ADP-ribose corresponded to linear chains of 30-35 ADP-ribose units, in accordance with the features of the polymer synthesized by the known vault-poly-ADP-ribose polymerase. In vitro experiments showed that V-poly-ADP-ribose polymerase activity of brain cytoplasmic fraction containing endogenous actin increased upon the addition of commercial actin and was highly reduced by ATP. Anti-actin immunoblot of the mixture in the presence and absence of ATP showed that the poly-ADP-ribosylation of octopus actin is a dynamic process balanced by the ATP-dependent polymerization of the cytoskeleton protein, a fundamental mechanism for synaptic plasticity. © 2013 Elsevier Inc. All rights reserved.

Active Sites Environmental Monitoring Program: Mid-FY 1991 report

Energy Technology Data Exchange (ETDEWEB)

Ashwood, T.L.; Wickliff, D.S.; Morrissey, C.M.

1991-10-01

This report summarizes the activities of the Active Sites Environmental Monitoring Program (ASEMP) from October 1990 through March 1991. The ASEMP was established in 1989 by Solid Waste Operations and the Environmental Sciences Division to provide early detection and performance monitoring at active low-level radioactive waste (LLW) disposal sites in Solid Waste Storage Area (SWSA) 6 and transuranic (TRU) waste storage sites in SWSA 5 as required by chapters II and III of US Department of Energy Order 5820.2A. Monitoring results continue to demonstrate the no LLW is being leached from the storage vaults on the tumulus pads. Loading of vaults on Tumulus II began during this reporting period and 115 vaults had been loaded by the end of March 1991.

Subcritical Neutron Multiplication Measurements of HEU Using Delayed Neutrons as the Driving Source

International Nuclear Information System (INIS)

Hollas, C.L.; Goulding, C.A.; Myers, W.L.

1999-01-01

A new method for the determination of the multiplication of highly enriched uranium systems is presented. The method uses delayed neutrons to drive the HEU system. These delayed neutrons are from fission events induced by a pulsed 14-MeV neutron source. Between pulses, neutrons are detected within a medium efficiency neutron detector using 3 He ionization tubes within polyethylene enclosures. The neutron detection times are recorded relative to the initiation of the 14-MeV neutron pulse, and subsequently analyzed with the Feynman reduced variance method to extract singles, doubles and triples neutron counting rates. Measurements have been made on a set of nested hollow spheres of 93% enriched uranium, with mass values from 3.86 kg to 21.48 kg. The singles, doubles and triples counting rates for each uranium system are compared to calculations from point kinetics models of neutron multiplicity to assign multiplication values. These multiplication values are compared to those from MC NP K-Code calculations

Quantitative estimation of the state of vault feet gymnasts on the different stages of the long-term training

Directory of Open Access Journals (Sweden)

O.V. Makarova

2013-10-01

Full Text Available The analysis of the vaults given about the state is conducted feet gymnasts on the different stages of the long-term training. 93 sportswomen of different qualification took part in research. The system of Big foot was used. It is set that on the early stages of the long-term training for gymnasts observed flattening heights of unevenness of navicular bone above the floor. With growth of qualification of sportswomen to avoid development of pathological changes of vaults feet actually not possibly. It is conditioned the rules of competitions to complication of competition compositions and technique of execution of elements of calisthenics. It is marked that appearance of flattening feet requires: corrections in the system of training; introduction of the specially developed methods on the removal of existent deformations; prophylaxis of flat-footedness; strengthening of musculoskeletal system feet.

Range and Image Based Modelling: a way for Frescoed Vault Texturing Optimization

Science.gov (United States)

Caroti, G.; Martínez-Espejo Zaragoza, I.; Piemonte, A.

2015-02-01

In the restoration of the frescoed vaults it is not only important to know the geometric shape of the painted surface, but it is essential to document its chromatic characterization and conservation status. The new techniques of range-based and image-based modelling, each with its limitations and advantages, offer a wide range of methods to obtain the geometric shape. In fact, several studies widely document that laser scanning enable obtaining three-dimensional models with high morphological precision. However, the quality level of the colour obtained with built-in laser scanner cameras is not comparable to that obtained for the shape. It is possible to improve the texture quality by means of a dedicated photographic campaign. This procedure, however, requires to calculate the external orientation of each image identifying the control points on it and on the model through a costly step of post processing. With image-based modelling techniques it is possible to obtain models that maintain the colour quality of the original images, but with variable geometric precision, locally lower than the laser scanning model. This paper presents a methodology that uses the camera external orientation parameters calculated by image based modelling techniques to project the same image on the model obtained from the laser scan. This methodology is tested on an Italian mirror (a schifo) frescoed vault. In the paper the different models, the analysis of precision and the efficiency evaluation of proposed methodology are presented.

Technical basis for storage of Zircaloy-clad spent fuel in inert gases

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Gilbert, E.R.

1983-09-01

The technical bases to establish safe conditions for dry storage of Zircaloy-clad fuel are summarized. Dry storage of fuel with zirconium alloy cladding has been licensed in Canada, the Federal Republic of Germany, and Switzerland. Dry storage demonstrations, hot cell tests, and modeling have been conducted using Zircaloy-clad fuel. The demonstrations have included irradiated boiling water reactor, pressurized heavy-water reactor, and pressurized water reactor fuel assemblies. Irradiated fuel has been emplaced in and retrieved from metal casks, dry wells, silos, and a vault. Dry storage tests and demonstrations have involved about 15,000 fuel rods, and about 5600 rods have been monitored during dry storage in inert gases with maximum cladding temperatures ranging from 50 to 570 0 C. Although some tests and demonstrations are still in progress, there is currently no evidence that any rods exposed to inert gases have failed (one PWR rod exposed to an air cover gas failed at about 270 0 C). Based on this favorable experience, it is concluded that there is sufficient information on fuel rod behavior, storage conditions, and potential cladding failure mechanisms to support licensing of dry storage in the US. This licensing position includes a requirement for inert cover gases and a maximum cladding temperature guideline of 380 0 C for Zircaloy-clad fuel. Using an inert cover gas assures that even if fuel with cladding defects were placed in dry storage, or if defects develop during storage, the defects would not propagate. Tests and demonstrations involving Zircaloy-clad rods and assemblies with maximum cladding temperatures above 400 0 C are in progress. When the results from these tests have been evaluated, the viability of higher temperature limits should be examined. Acceptable conditions for storage in air and dry storage of consolidated fuel are issues yet to be resolved

Experimental research into dynamic properties of masonry barrel vaults non-reinforced and reinforced with carbon composite strips

Czech Academy of Sciences Publication Activity Database

Zigler, R.; Witzany, J.; Makovička, D.; Urushadze, Shota; Pospíšil, Stanislav; Kubát, J.; Kroftová, K.

2016-01-01

Roč. 2016, č. 3 (2016) ISSN 1805-2576 Institutional support: RVO:68378297 Keywords : rehabilitation * strengthening * masonry vaults * FRP fabrics * dynamic load * static load Subject RIV: JM - Building Engineering https:// doaj .org/article/a4197d0d1b974dbeadf9188398d45fc8

Techno-economical Analysis of High Level Waste Storage and Disposal Options

International Nuclear Information System (INIS)

Bace, M.; Trontl, K.; Vrankic, K.

2002-01-01

Global warming and instability of gas and oil prices are redefining the role of nuclear energy in electrical energy production. A production of high-level radioactive waste (HLW), during the nuclear power plant operation and a danger of high level waste mitigation to the environment are considered by the public as a main obstacle of accepting the nuclear option. As economical and technical aspects of the back end of fuel cycle will affect the nuclear energy acceptance the techno-economical analysis of different methods for high level waste storage and disposal has to be performed. The aim of this paper is to present technical and economical characteristics of different HLW storage and disposal technologies. The final choice of a particular HLW management method is closely connected to the selection of a fuel cycle type: open or closed. Wet and dry temporary storage has been analyzed including different types of spent fuel pool capacity increase methods, different pool location (at reactor site and away from reactor site) as well as casks and vault system of dry storage. Since deep geological deposition is the only disposal method with a realistic potential, we focused our attention on that disposal technology. Special attention has been given to the new idea of international and regional disposal location. The analysis showed that a coexistence of different storage methods and deep geological deposition is expected in the future, regardless of the fuel cycle type. (author)

POLE.VAULT: A Semantic Framework for Health Policy Evaluation and Logical Testing.

Science.gov (United States)

Shaban-Nejad, Arash; Okhmatovskaia, Anya; Shin, Eun Kyong; Davis, Robert L; Buckeridge, David L

2017-01-01

The major goal of our study is to provide an automatic evaluation framework that aligns the results generated through semantic reasoning with the best available evidence regarding effective interventions to support the logical evaluation of public health policies. To this end, we have designed the POLicy EVAlUation & Logical Testing (POLE.VAULT) Framework to assist different stakeholders and decision-makers in making informed decisions about different health-related interventions, programs and ultimately policies, based on the contextual knowledge and the best available evidence at both individual and aggregate levels.

Design of double containment canister cask storage system

International Nuclear Information System (INIS)

Asami, M.; Matsumoto, T.; Oohama, T.; Kuriyama, K.; Kawakami, K.

2004-01-01

interim storage facility to continue the storage. Furthermore, the primary canister will be able to restore in an advanced system such as high capacity vault system adequate for fuels with decreased decay heat and dose rate

Gamma irradiation tests of concrete material recommended for storage casks of spent nuclear fuel arising from Cernavoda NPP

International Nuclear Information System (INIS)

Dulama, M.; Deneanu, N.; Dulama, C.; Baboescu, E.

2001-01-01

Considerable effort is being devoted to the Romania's Nuclear Spent Fuel and Waste Management R and D Program to develop engineered barriers for the containment of nuclear fuel waste under conditions of deep geological disposal. Engineering practice suggests that the concrete should fulfil the requirements of long term physical stability and resistance to radiation. With an appropriate system of metal reinforcement, it should be possible to obtain the tensile and impact strength required, avoiding the risk of mechanical damage during handling and emplacement. In accordance with the concept developed by CITON-Bucharest, presently, the dry storage of spent nuclear fuel is thought by two choices: - The alternative of dry storage type MMB3; - The alternative of dry storage type TRANSTOR. By using ORIGEN and PELSHIE computer codes, we evaluated the gamma radiation dose absorbed by the concrete walls of the storage vault both in MMB3 and in TRANSTOR designing variants. The irradiation tests were performed at the Gamma Irradiation Facility of the Institute for Nuclear Research. (authors)

Effect of Plyometric Training on Handspring Vault Performance and Functional Power in Youth Female Gymnasts.

Science.gov (United States)

Hall, Emma; Bishop, Daniel C; Gee, Thomas I

2016-01-01

This study aimed to determine the effect of plyometric training (PT) when added to habitual gymnastic training (HT) on handspring vault (HV) performance variables. Twenty youth female competitive gymnasts (Age: 12.5 ± 1.67 y) volunteered to participate and were randomly assigned to two independent groups. The experimental plyometric training group (PTG) undertook a six-week plyometric program, involving two additional 45 min PT sessions a week, alongside their HT, while the control group (CG) performed regular HT only. Videography was used (120 Hz) in the sagittal plane to record both groups performing three HVs for both the baseline and post-intervention trials. Furthermore, participants completed a countermovement jump test (CMJ) to assess the effect of PT on functional power. Through the use of Quintic biomechanics software, significant improvements (P plyometric training seemingly improved the power generating capacity of movement-specific musculature, which consequently improved aspects of vaulting performance. Future research is required to examine the whether the improvements are as a consequence of the additional volume of sprinting and jumping activities, as a result of the specific PT method or a combination of these factors.

How the radiologic and nuclear medical communities can improve nuclear security.

Science.gov (United States)

Kahn, Laura H; von Hippel, Frank

2007-04-01

Highly enriched uranium (HEU) is used to manufacture technetium-99m, the most widely used medical radioisotope in the world. Highly enriched uranium is also used to make nuclear bombs; 50 kg of HEU is enough to make a Hiroshima-type bomb. It is generally agreed that this technology is within the reach of a terrorist group; the main obstacle is acquiring HEU. Currently, as a legacy of the US and Soviet Atoms for Peace Program, there are civilian users of HEU in 40 countries, and about 1,000 kg are still being shipped each year. Unfortunately, the major international manufacturers of technetium-99m have been refusing to convert their production facilities to use low-enriched uranium (LEU), which cannot be used to make a nuclear bomb. Only 1% to 2% of the HEU is consumed in the process of producing technetium-99m. The remainder is accumulating in radioactive waste storage facilities. The radiologic and nuclear medical communities could make a tremendous contribution to a safer world by supporting the replacement of HEU with LEU in the production of technetium-99m. Low-enriched uranium is just as cost effective as HEU for the manufacture of technetium-99m and does not contribute to the risk for nuclear terrorism.

Radiological performance assessment for the E-Area Vaults Disposal Facility. Appendices A through M

Energy Technology Data Exchange (ETDEWEB)

Cook, J.R.

1994-04-15

These document contains appendices A-M for the performance assessment. They are A: details of models and assumptions, B: computer codes, C: data tabulation, D: geochemical interactions, E: hydrogeology of the Savannah River Site, F: software QA plans, G: completeness review guide, H: performance assessment peer review panel recommendations, I: suspect soil performance analysis, J: sensitivity/uncertainty analysis, K: vault degradation study, L: description of naval reactor waste disposal, M: porflow input file. (GHH)

HEU to LEU fuel conversion. Final report

International Nuclear Information System (INIS)

Mulder, R.U.

1994-10-01

The Nuclear Regulatory Commission issued a ruling, effective March 27, 1986, that all U.S. non-power reactors convert from HEU fuel to LEU fuel. A Reduced Enrichment for Research and Test Reactors Program was conducted by the Department of Energy at Argonne National Laboratory to coordinate the development of the high density LEU fuel and assist in the development of Safety Analysis Reports for the smaller non-power reactors. Several meetings were held at Argonne in 1987 with the non-power reactor community to discuss the conversion and to set up a conversion schedule for university reactors. EG ampersand G at Idaho was assigned the coordination of the fuel element redesigns. The fuel elements were manufactured by the Babcock ampersand Wilcox Company in Lynchburg, Virginia. The University of Virginia was awarded a grant by the DOE Idaho Operations Office in 1988 to perform safety analysis studies for the LEU conversion for its 2 MW UVAR and 100 Watt CAVALIER reactors. The University subsequently decided to shut down the CAVALIER reactor. A preliminary SAR on the UVAR, along with Technical Specification changes, was submitted to the NRC in November, 1990. An updated SAR was approved by the NRC in January, 1991. In September, 1992, representatives from the fuel manufacturer (B ampersand W) and the fuel designer (EG ampersand G, Idaho) came to the UVAR facility to observe trial fittings of new 22 plate LEU mock fuel elements. B ampersand W fabricated two non-fuel bearing elements, a regular 22 plate element and a control rod element. The elements were checked against the drawings and test fitted in the UVAR grid plate. The dimensions were acceptable and the elements fit in the grid plate with no problems. The staff made several suggestions for minor construction changes to the end pieces on the elements, which were incorporated into the final design of the actual fuel elements. Selected papers are indexed separately for inclusion in the Energy Science and Technology

HEU to LEU fuel conversion. Final report

Energy Technology Data Exchange (ETDEWEB)

Mulder, R.U.

1994-10-01

The Nuclear Regulatory Commission issued a ruling, effective March 27, 1986, that all U.S. non-power reactors convert from HEU fuel to LEU fuel. A Reduced Enrichment for Research and Test Reactors Program was conducted by the Department of Energy at Argonne National Laboratory to coordinate the development of the high density LEU fuel and assist in the development of Safety Analysis Reports for the smaller non-power reactors. Several meetings were held at Argonne in 1987 with the non-power reactor community to discuss the conversion and to set up a conversion schedule for university reactors. EG&G at Idaho was assigned the coordination of the fuel element redesigns. The fuel elements were manufactured by the Babcock & Wilcox Company in Lynchburg, Virginia. The University of Virginia was awarded a grant by the DOE Idaho Operations Office in 1988 to perform safety analysis studies for the LEU conversion for its 2 MW UVAR and 100 Watt CAVALIER reactors. The University subsequently decided to shut down the CAVALIER reactor. A preliminary SAR on the UVAR, along with Technical Specification changes, was submitted to the NRC in November, 1990. An updated SAR was approved by the NRC in January, 1991. In September, 1992, representatives from the fuel manufacturer (B&W) and the fuel designer (EG&G, Idaho) came to the UVAR facility to observe trial fittings of new 22 plate LEU mock fuel elements. B&W fabricated two non-fuel bearing elements, a regular 22 plate element and a control rod element. The elements were checked against the drawings and test fitted in the UVAR grid plate. The dimensions were acceptable and the elements fit in the grid plate with no problems. The staff made several suggestions for minor construction changes to the end pieces on the elements, which were incorporated into the final design of the actual fuel elements. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

Neutron spectra in two beam ports of a TRIGA Mark III reactor with HEU fuel

International Nuclear Information System (INIS)

Vega C, H. R.; Hernandez D, V. M.; Paredes G, L.; Aguilar, F.

2012-10-01

Before to change the HEU for Leu fuel of the ININ's TRIGA Mark III nuclear reactor the neutron spectra were measured in two beam ports using 5 and 10 W. Measurements were carried out in a tangential and a radial beam port using a Bonner sphere spectrometer. It was found that neutron spectra are different in the beam ports, in radial beam port the amplitude of thermal and fast neutrons are approximately the same while, in the tangential beam port thermal neutron peak is dominant. In the radial beam port the fluence-to-ambient dose equivalent factors are 131±11 and 124±10 p Sv-cm 2 for 5 and 10 W respectively while in the tangential beam port the fluence-to-ambient dose equivalent factor is 55±4 p Sv-cm 2 for 10 W. (Author)

Conceptual design report, plutonium stabilization and handling,project W-460

Energy Technology Data Exchange (ETDEWEB)

Weiss, E.V.

1997-03-06

Project W-460, Plutonium Stabilization and Handling, encompasses procurement and installation of a Stabilization and Packaging System (SPS) to oxidize and package for long term storage remaining plutonium-bearing special nuclear materials currently in inventory at the Plutonium Finishing Plant (PFP), and modification of vault equipment to allow storage of resulting packages of stabilized SNM for up to fifty years. This Conceptual Design Report (CDR) provides conceptual design details for the vault modification, site preparation and site interface with the purchased SPS. Two concepts are described for vault configuration; acceleration of this phase of the project did not allow completion of analysis which would clearly identify a preferred approach.

SU-E-T-670: Radiotherapy Vault Shielding Evaluation Method for a Flattening Filter-Free (FFF) Linac-Practical Considerations and Recommendations

Energy Technology Data Exchange (ETDEWEB)

Mihailidis, D [CAMC Cancer Center-Alliance Oncology and West Virginia University, Charleston, WV (United States)

2015-06-15

Purpose: To date, there isn’t formal approach for flattening filter-free (FFF) linac vault shielding evaluation, thus, we propose an extension to NCRP#151 to accommodate the recent large number of FFF linac installations.Methods and Materials: We extended the approach in NCRP#151 to design two Truebeam vaults in our new cancer center for hypofractionated treatments. Monte Carlo calculations have characterized primary, scattered, leakage and neutron radiations from FFF-modes. These calculations have shown that: a) FFF primary beam is softer on the central-axis compared to flattening filtered (FF), b) the lateral dose profile is peaked on the central axis and less integral target current is required to generate the same tumor dose with the FF beam. Thus, the TVLs for FFF mode are smaller than those of the FF mode and the scatter functions of the FF mode (NCRP#151) may not be appropriate for FFF-mode, c) the neutron source strength and fluence for 18X-FFF is smaller than 18X-FF, but it is not of a concern here, no 18X-FFF-mode is available on the linac under investigation. Results: These barrier thickness are smaller (12% reduction on the average) than those computed for conventional FF mode with same realistic primary workload since, the primary TVLs used here are smaller and the WL is smaller than the conventional (almost half reduced), keeping the TADR in tolerance. Conclusions: A comprehensive method for shielding barrier calculations based on dedicated data for FFF-mode linacs is highly desired. Meanwhile, we provide an extension to NCRP#151 to accommodate the shielding design of such installations. It is also shown that if a vault is already designed for IMRT/VMAT and SABR hypofractionated treatments with FFF-mode linac, the vault can also be used for a FFF mode linac replacement, leaving some leeway for slightly higher workload on the FFF linac.

SU-E-T-670: Radiotherapy Vault Shielding Evaluation Method for a Flattening Filter-Free (FFF) Linac-Practical Considerations and Recommendations

International Nuclear Information System (INIS)

Mihailidis, D

2015-01-01

Purpose: To date, there isn’t formal approach for flattening filter-free (FFF) linac vault shielding evaluation, thus, we propose an extension to NCRP#151 to accommodate the recent large number of FFF linac installations.Methods and Materials: We extended the approach in NCRP#151 to design two Truebeam vaults in our new cancer center for hypofractionated treatments. Monte Carlo calculations have characterized primary, scattered, leakage and neutron radiations from FFF-modes. These calculations have shown that: a) FFF primary beam is softer on the central-axis compared to flattening filtered (FF), b) the lateral dose profile is peaked on the central axis and less integral target current is required to generate the same tumor dose with the FF beam. Thus, the TVLs for FFF mode are smaller than those of the FF mode and the scatter functions of the FF mode (NCRP#151) may not be appropriate for FFF-mode, c) the neutron source strength and fluence for 18X-FFF is smaller than 18X-FF, but it is not of a concern here, no 18X-FFF-mode is available on the linac under investigation. Results: These barrier thickness are smaller (12% reduction on the average) than those computed for conventional FF mode with same realistic primary workload since, the primary TVLs used here are smaller and the WL is smaller than the conventional (almost half reduced), keeping the TADR in tolerance. Conclusions: A comprehensive method for shielding barrier calculations based on dedicated data for FFF-mode linacs is highly desired. Meanwhile, we provide an extension to NCRP#151 to accommodate the shielding design of such installations. It is also shown that if a vault is already designed for IMRT/VMAT and SABR hypofractionated treatments with FFF-mode linac, the vault can also be used for a FFF mode linac replacement, leaving some leeway for slightly higher workload on the FFF linac

Experience on wet storage spent fuel sipping at IEA-R1 Brazilian research reactor

International Nuclear Information System (INIS)

Perrotta, J.A.; Terremoto, L.A.A.; Zeituni, C.A.

1998-01-01

The IEA-R1 research reactor of the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) is a pool type reactor of B and W design, that has been operating since 1957 at a power of 2 MW. Irradiated (spent) fuels have been stored at the facility during the various years of operation. At present there are 40 spent fuel assemblies at dry storage, 79 spent fuel assemblies at wet storage and 30 fuel assemblies in the core. The oldest fuels are of United States origin, made with U-Al alloy, both of LEU and HEU MTR fuel type. Many of these fuel assemblies have corrosion pits along their lateral fuel plates. These pits originate by galvanic corrosion between the fuel plate and the stainless steel storage racks. As a consequence of the possibility of sending the irradiated old fuels back the U.S.A., sipping tests were performed with the spent fuel assemblies. The reason for this was to evaluate their 137 Cs leaking rate, if any. This work describes the procedure and methodology used to perform the sipping tests with the fuel assemblies at the storage pool, and presents the results obtained for the 137 Cs sipping water activity for each fuel assembly. A correlation is made between the corrosion pits and the activity values measured. A 137 Cs leaking rate is determined and compared to the criteria established for canning spent fuel assemblies before shipment

The conversion of NRU from HEU to LEU fuel

International Nuclear Information System (INIS)

Sears, D.F.; Atfield, M.D.; Kennedy, I.C.

1990-01-01

The program at Chalk River Nuclear Laboratories (CRNL) to develop and test low-enriched uranium fuel (LEU, 3 Si, USiAl, USi Al and U 3 Si 2 (U-3.96 wt% Si; U-3.5 wt% Si-1.5 wt% AL; U-3.2 wt%; Si-3 wt% Al; U-7.3 wt% Si, respectively). Fuel elements were fabricated with uranium loadings suitable for NRU, 3.15 gU/cm 3 , and for NRX, 4.5 gU/cm 3 , and were irradiated under normal fuel-operating conditions. Eight experimental irradiations involving 100 mini-elements and 84 full-length elements (7X12-element rods) were completed to qualify the LEU fuel and the fabrication technology. Post irradiation examinations confirmed that the performance of the LEU fuel, and that of a medium enrichment uranium (MEU, 45% U-235) alloy fuel tested as a back-up, was comparable to the HEU fuel. The uranium silicide dispersion fuel swelling was approximately linear up to burnups exceeding NRU's design terminal burnup (80 at%). NRU was partially converted to LEU fuel when the first 31 prototype fuel rods manufactured with industrial scale production equipment were installed in the reactor. The rods were loaded in NRU at a fuelling rate of about two rods per week over the period 1988 September to December. This partial LEU core (one third of a full NRU core) has allowed the reactor engineers and physicists to evaluate the bulk effects of the LEU conversion on NRU operations. As expected, the irradiation is proceeding without incident